ML20154M332

Summary of ACRS 304th Meeting on 850808-10 in Washington,Dc Re Maint & Surveillance Programs,Nrc long-range Plan,Mgt & Disposal of Radwastes,Class 9 Accidents & Scram Circuit Breaker Reliability.Apps Encl

ML20154M332
Person / Time
Issue date:	08/08/1985
From:	Advisory Committee on Reactor Safeguards
To:	Advisory Committee on Reactor Safeguards
References
ACRS-2347, NUDOCS 8603140070
Download: ML20154M332 (362)

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1 TABLE OF CONTENTS T MINUTES OF THE 304th ACRS MEETING h) I

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I. Chairman's Report ................................... I II. Alvin W. Vogtle Nuclear Plant, Units 1 and 2 OL Review (0 pen) ....................................... 2 III. San Onofre Unit 1 SEP Review (0 pen) ................. 10 IV. General Electric Standard Safety Analysis Report (GESSAR-II)(0 pen) ................................... 16 V. Seismic Qualification of Equipment in Operating Plants (0 pen) ....................................... 21 VI. NRC Maintenance and Surveillance Program Plan (0 pen) 27 VII. NRC Long Range Plan (0 pen) .......................... 30, VIII. Management and Disposal of Radioactive Wastes (0 pen) . 31 IX. Class Nine Accidents Subcomittee Meeting (0 pen) .... 33

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X. INPO Radiation Protection Program (0 pen) ............ 34 XI. Scram Circuit Breaker Reliability (0 pen) ............ 35 XII. Meeting of the ECCS Subcomi ttee (0 pen) . . . . . . . . . . . . . 36 XIII. Executive Sessions (0 pen) ........................... 37 A. Subcomittee Assignments

1. Federal Training Academy for Nuclear Power Plant Operators ........................... 37

2. FTOL Conversion for Millstone Nuclear Station Unit 1 .................................... 3I

3. Enforcement Policy on Vendors - SECY-85-256 . 37

4. Dissolution of the SEP Subcomittee ........ 37 .

5. Scram Circuit Breaker Reliability . . . . . . .. . . . 38 B. Reports, Letters, and Memoranda ................ 38 3 ^. ^ ' ' ' " '

8603140070 050000 Cortinca ry--[ - - - . - - - .

PDR 2347 ACRB PDR

TABLE OF CONTENTS (C nt.)

MINUTES OF THE 304th ACRS MEETING

1. ACRS Report on the Vogtle Electric Generating Plant, Units 1 and 2 ....................... 38

2. ACRS Report on the Systematic Evaluation Program Review of the SAN Onofre Nuclear Generating Station Unit 1 ................. 38

3. ACRS Comments on the Status of USI A-46 (Seismic Qualification of Equipment in Ope ra ti ng Pl a nts ) . . . . . . . . . . . . . . . . . . . . . . . . . . 38

4. ACRS Role on the NRC High-Level Radioactive Waste Program .............................. 39

5. ACRS Comments on the NRC Maintenance and Surveillance Program Plan .................. 39

6. Status Report on Long-Range Planning ....... 39

7. INP0 Program on Radiation Protection ....... 39

8. Systematic Analysis of Operating Plants .... 39

9. Reply to G. Petrangeli's, ENEA, Letter dated June 9, 1985 ........................ 39' C. Generic Issues ................................. 40

1. Proposed Amendments on Physical Protection for Category II Quantities of High Enrichment Uranium ................................ 40 D. Future Schedule ................................ 40

1. Future Agenda ............................. 40

2. Future Subcommittee Activities ............ 40 E. Administrative and Procedural Items ............ 40 F. Ccmanche Peak .................................. 40 G. Activities of ACRS Members ..................... 40 Proprietary Supplement - GESSAR-II Sabotage Considerations .... 40

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TABLE OF CONTENTS APPENDIXES TO MINUTES OF THE 304th ACRS MEETING August 8-10, 1985 Appendix I - List of Attendees ........................ A-1 Appendix II - Future Agenda .............................. A-2 Appendix III - Schedule of ACRS Subcommittee Meetings ..... A-4 Appendix IV - Vogtl e S tatus Report . . . . . . . . . . . . . . . . . . . . . . . A-36 Appendix V _

Principle & Unique Design Features of Plant Vogtle ............................... A-45 Appendix VI - NRC Regional Evaluation of Construction .... A-67 Appendix VII - NRC Licensing Review of Plant Vogtle ....... A-75 Appendix VIII - Vogtle Project Organization ................ A-90 Appendix IX - Management Philosophy at Georgia Power Company .................................... A-92 Aopendix X - Georgia Power Corporate Organization ....... A-97 Appendix XI - Vogtle Plant Operation Organization and Training ................................... A-103 Appendix XII - Plant Vogtle Quality Assurance . . . . . . . . . . . . . A-112 Appendix XIII - Vogtle Quality Concern Program . . . . . . . . . . . . . A-121 Appendix XIV - Vogtle Readiness Review Program ............ A-126 Appendix XV - NRC Participation in Vogtle Readiness Review A-145 Appendix XVI - UT Examination of Cast Stainless Steel Piping ..................................... A-146 Appendix XVII - San Onofre Plant Overview .................. A-148 Appendix XVIII - NRC Presentation on SEP Integrated Assessment l San Crofre ................................. A-154 l Appendix XVIX - GESSAR II Severe Accident Issues Presentation to the ACRS ................................ A-218 1

j Appendix XX - GESSAR II PRA Review Detailed Discussion of

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Hydrogen by T. Pratt ....................... A-220 L

TABLE OF CONTENTS (Cont.)

APPENDIXES TO MINUTES OF THE 304TH ACRS MEETING Appendix XXI -

GESSAR II PRA Review Effect of a Core Melt on Vessel Support Integrity by T. Pratt .... A-237 Appendix XXII -

Staff Presentation on USI A-46 Resolution .. A-248 Appendix XXIII - SQUG Presentation .......................... A-261 Appendix XXIV - Summary of the Effects of the Great Chile Earthquake of 1985 ......................... A-268 Appendix XXV - Nuclear Regulatory Maintenance & Surveillance Program .................................... A-279 Appendix XXVI - GE Proprietary Information GESSAR II Sabotage Cons i dera ti ons . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A-301 Appendix XXVII - Additional Documents Provided fcr ACRS' Use .A-308 l

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c 36@30780 Federal " dater / VJ1. 50. No.145 / Mond:y Jul3 1985 / Notices

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d Th3 amendm:nts maka technical and 1015 e m -Coffee Bre:k 9 other changes in the eligibility th0 AM-200 P.M;NRClong Ronge j cnnditions and other terms for the to45 a m.-Biotechnology, ferome S Schults 11.30 a m -Envimnmental Enginuring Plan (Open}-The Commitlee wdl hear

! Edward H. Bryan the report ofits subcommittee on administration of the Ceneral Operating i Support and Museum Assesstnent

$8 Pm " proposed ACRS comments regarding a programs for museums and remove p, sh oe Tnhnology. T K. long range plan for NRC regulatory unneeded provisions. As revised. the Custafon activities. Members of the NRC Staff regulations published on [une 17,1963 2.15 p rn.-a.oecgineenna and research to and and invited experts will participate as the Handicapped. Wilham Freedman oppropnate.

will apply to the sword of grants for 3 oo p m. Coffee Break Fiscal Year 19es. 2sp AM-J 00 PN. Seismic 3 30 p m.-Sy stems Engineenna for tarse q,iolsficolion ofEquipmentin Orseroting Further Infortsation 4 15 INa Ien ## # ### @E'"}~ '

a de ifatard For further information contact Kristine K. Ramaekers. Museum Program Officer. Institute of Museum 3 g, saj'n chael P. Cous Wednesday. Ausust tv. Rooms 1241-A and

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regarding proposed methodology for Sersices.1100 Pennsylvania Avenue, 17'J-8 se smic quahfication of equipment in NW Room 609. Washmgton. D.C. 2050& 9 oo a m.-12 Noon-Emergma Erytineerms nuclear power plants. Representatives i Telephone. (202) 786-0533. Sptems Committee Room :242-A of the NRC Sta!I and the nuclear (Catalog of Federal Domestic Assistance No.

8 industry will participate as appropriate.

45 301 Instatute of M.oseurn Seruceal terns Co m t R m 24 -8 #00EM-#00EM#NOC Dated. luly 25,1945. 12 Noon-Lunch Maintenonew andSurveil/once Pmgram 130 p m -Continuation of Momina Plan (Open)--11ie miembers wdl hear Senna E. Phallips. Discussions Precte. !sst: cre:fMacum Seri,cea. 5 00 p o - AJ:eum the report of its subcommittee regarding the propoecd NRC program plan for IFR Doc. 85-1?925- Filed 7-26-85. e 45 mm) M Rebeua Winkler,

,w,,, coo, ,,y,,, maintenance end surveillance of nuclear committeeMonosement ofFcer, power plants. Members of the NRC Staff

--.-- .= lFR Doc. 85-17s35 Feeld 7-26-83. e45 eml wdl perticipate as appropriate.

m w uscoon m ms-a j NATIONAL SCIENCE FOUNDATION _ _ _ _ . . _ __. ,_

a 00 PM-e 30 PM: nature ACRS Activities (Open)-The members of the s

Joint Meeting of the Advloory Committee w1!! discuss anticipated NUCLEAR REGULATORY ACRS Subcommittee actisity and items Committees for Civil and CO SSION proposed for consideration by the full

)t Environmental Engineer *ng andEarthquaka dvisory Committee Nazard on ReactorMitJgat2on; Committee.

Open \

Meeting Safeguards; Meeting Agenda N

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In accordance with the Federal Advisory Committee Act. Pub. L 92-463.

in accordance with the purposes of sections 29 and 182b. of the Atomic g[ ~[ j # '

the National Science Found. tion Enertry Act (42 U.S C. 2039. 2232b). the members wdl hear and discuss the 8

announces the followmg meetmg. Adusory Committee on Reactor te art ofits subcommittee on the SEP Name Adutory Comaruffees for Civil nd ' I" * ** *** " resiew of this nuclear plant.

Fs al F sineertns and Far*hqi.nhe A a O m Representatives of the NRCStaff and g the I censee wi I m ke presentations and D c of Place. Rooms saa 1242-A.1:42 0. National this meeting was pubbshed in the Sciem.e Foundation.1a00 C Street. NW., Federol Regfsier on July 23.1985. 1J.JO AR-42Jo PMa Indian Point

'. Washington. D C. 2 ossa l D.ie August 1119a5-900 e m to s ao The agenda for the subject meetmg Nac/cor /h er Station (Open)--The p m; August 14.196%90n a m to 12 00 will be as fo!!aws: members of the Committee wdldiscuss g

l Noon. Thursday. August 8,1965 proposed ACRS comments regardma Trve of Mutses Open implementation of the results of the PRA z

,, Contact Person Dr Arthur A- 830 A M-8 45 AM: Report of A CRS of the laduan Point Nuclear Station.

Erre Director. Division of Fundamental Chairman (Open)-The ACRS Chairman 130 P M-2J0 PN; Management will report briefly regarding items of andD'8Posolo/Radioactne Wastes nee te T. .. na Soente n rrent inte+ cst to the Committee. (Oper)-The rnembers of the Committec

{ Foundalson. Room 1132. tsin G S#reet. NW.

Washms'on. D C 20m Telephone. 202/137.

1. 43 A M-12(UNoon; Cenem/ will hear the report of 4ts subcommittee g E!cctricStordordSofety Analysis regardmg proposed ACRS actigities in 9545 I Report /CESSAR /// (Open/ Closed)- support of the NRC regulatory program Puryone of Meetins To proud edv6ce and Members of the Committee wdl hear for handLng and disposal of radioacttve recommendanons concernire fund mental end discuss the report of the copelzant g wastes. Representat2ves of the NRC eneerth in emerairs and crit 4 cal ensineenns ACRS Subcommittee regardir.3 -

• . 'Mm'- Staff and the Department of Energy wdi review of this prolect for a PDA. participate as appropriate.

the con ac p oa ea se ,'j '

Members of the NRC Staff and 220PM-&JaPM: Alvin W. Vogtle

? addens representatives of the applicant m!! Nuclear Plant. Units ! andJ(Open/

I A sead* make presentations and participate in C/osed/-The members will hear and the discussion. discuss the report of its subcomrnittee Tuesday. A.quer n Awom se Portions of this session wdl tse closed

- i regardmit the request for an operating e00 a m -welcome and introduction Nam P us necessary to discuss Proprietary license for this nuclear plant. Members -

i, Suh Information applicable to this project

% of the NRC Staff and repreuntatnes of 9 IS a m --Oserview. Arthur A bte and detailed irformation regardmg the applicant will make presentahons

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910 a nt--Earthquase Fepineerinaw Mechsel P Caus accurity provisions for this type of ruclear steam supply system.

and perficipale in the discuseson as a ppropriata.

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30770 Fed:ral R:gister / Vol. 50. No.145 / hiond.iy luly 29,19ti5 / Notices Portions of this session will be closed ACRS Esecutise Director if such portions of the meeting may be closed as necessary to discuss Proprietary rescheduhng wo;.!d result in major for the discussion ofindividuals as information and detailed security incons enience. potential consultants to the Waste arrangements for this plant. I have determined in accordance with htanagement Subcommittee. All other Saturday. August 10,1985 subsection 10(d) Pub. L 92-463 that it is items remain the same as previously ,

necessary to close portions of this announced.

&Jo A Af.-1200 Noon: ACRS Reports meeting as noted above ta discuss Further information regarding topics I to NRC(Open/ Closed)-This portion nf Proprietary Information (5 U.S C. to be discussed, a hether the meeting the meeting will be to discuss proposed 55:bic)(4ll detailed security information I ACRS reports to the NRC regarding has been cancelled or rescheduled the t

[5 U.5 C. 55:b(c)(3)l. to d,iscuss Chairman's ruling on requests for the items considered during this meeting information that will be involved in an Portions of this session will be closed adjudicatory proceeding [5 U.S C. pportunity to present oraj statements to discuss Preprietary Information and the time allotted therefor can be 55:ble)(10)]. and to discuss infermation ,

applicable to the matters being the release of which would represent an obtained by a prepaid telephone call to ,

considered and detailed secunty unwarranted invasion of personal the cognizant ACRS staff member, htr.

arrangements for the projects being privacy [5 U S.C. 552b(c)(6)l Owen S. kierrill (telephone 202/634- I resiew ed. Further information regarding topics 1414) or Atr. R F. Fraley 1:02/634-3:65)

J 00PN-J 00 PA1.. ACRS to be discussed, whether the meeting between 815 a.m. an 15.00 p m. Persons Subcommittee Activities (Open) ~lhe has been cance!!ed or rescheduled, the planning to attend this meeting are members wdl hear and discuss the Chairman's ruling on requests for the urged to contact the above named reports of designated whcommittees opportunity to present oral statements individu.at one or Iwo days before the regarding ongoing actisities including and the time allotted can be obtained by scheduled meeting to be advised of any ECC systems esaluation. ACRS a prepaid telephone call to the ACRS changes in schedule, etc., which may procedures and practices, scram s) stent Esecutive Director, htr. Raymond F. have occurred.

reliability, the radiation protection Fraley (telephone 202/634-3265).

program of 1.NPO. and the source term between 815 A ht and 5 00 P ht. EDT.

used in accident evaluation. Deted lot3 31953.

DatcJ ivty 24 194 lohn C. Mdinley.

300 P Al-3 30 P A1: Actinties of ACRS Als:nbers (Open/Clased)-The Samu*l1 C 'lk- Chief 1%er t Penew Bwh Na r ,

Committee will discuss proposed $',yi d' i'#) C"*'" f*'T ~'""' p % mn p.x m W owe caos rsews-as

. o et m nt e r o ees." " lFR Doc &5-P9:3 Filed 7-: rwa 5 8 45 am)

• oCo** "***

Portions of this sessmn will be closed as necess.iry to discuss information the - - - - -

release of which would represent .in ' """I u arranted invasion of personal Advisory Committee on Reactor Safeguarda; Suticommittees on Waste g tea $.

1 ty 3, j Procedures for the conduct of a M8"89'm*"I 8"d Pf0C'du8 8"d Before Administrative ludges lohn H Fryc. I participation in ACRS mcetings were Administration;Revloed Agende lit. Chairmen. Dr lamn H Carpenter. Dr

• published in the Federal Reginter on 'Ihc Federal Register pubbshed on Peter A. M:rris October 3,1984 (43 FR 193) In htonday. July 22.1985 (50 FR 29 73) in e4 wner of kerr M ca Chemu el accordance w 'th these procedures. or41 contained notice of a joint meeting of Cerpove' ion (West Chicago Rere Earths or wntten statements may be presented the ARCS Subcommittees on Waste Idc'Id)l Dxhet No ao-2oet ML ASIEP %

by members nf the public.recordmgs hlanagement, and Procedures and amSan-ML and Kerr McCu Chemical will be permitted only during those Administration to be held on Tuesday. C"'Po"*" IK'ese Crnk Deconteminetmal portions of the meeting when a luly 30.1965. 8 30 a.m . Room 1046.1717 D*het Nn 404oet-5C. A$tJP No 44-Sn2-tr4nscript is being b ept, and question 9 it Street. NW., Washington, DC. in m-SC may be asked only by members of the addition to the ACRS Role in the Committee. its consultants. and St4ff. Civdain llish.Les el Radosactive Waste Please take notice that preheanna Persons desinng to snake or.al klenagement Program, the followmg conferences in these proceedings will be statements should notify the ACRS items base been added to the agend4; held on September 11 and 12.19% at Esecutive Director as far in adsance as (1) ACRS Annual Report to Congress the NRC heanna room, fifth floor. 4350 practicable so that appropriate on the NRC Safety Research Prostam Ea st. West liighway. Bethesda, arrangements can be made to allnw the and Dudget-discuss scope and detail of Ntaryland The conference In the West necessary time during the meeting for this report.

such statements Use of still, motion Chicago plu<. reding will begin at 9 30 (2) Appomtment of ACRS A Af. on September 11 and will be picture and television cameras durtng Sutcommittee to consider risk this meeting may be limited to selected followed by the conference in the Kress perspecthe m regulatory requirements Creek proceeding.

portions of the meeting as determined (3) ACRS consideration of proposed by the Chairman Information regardms changes m NRC Standard Revtew The purpose of the conferences is to the time to be set aside for this purpose Plan-discuss apphenble procedures hear argument of the parties regarding may be obtained by a prepaid telephone (4) Testing of NRC Operator discovery disputes.

call to the ACRS Executive Director. Candidates-procedures for ACRS lohn 11 Frie lif.

R F. Fraley. prior to the meeting in stew consideration of natural abihty testing of the possibihty that the schedule for (5) ACRS activities-discuss ACRS Ch*#**"' d d*'"'8 "' /"#8'-

ACRS meetings may be adjusted by the action regarding items carned oser from Idly 31981 Chairman as necessary to facilitate the eather ACRS assignmen's corutuct of the meeting. persons The meeting will, for the most part, be IHI D* SS tM21 Filed 74W. e e amj planning to attend shoubl r. heck with the open to pubhc attendance. lloweser. **u=o cooe ne*4 .es

/ a anco,*g UNITE 3 STATES

!" c. NUCLEAR RECULATORY COMMISSION y . ,I ADVISORY COMMITTEE ON REACTOR SAFEGUARDS o WASHINGTON, D. C. 20585 August 1, 1985 SCHEDULE AND OUTLINE FOR DISCUSSION 304TH ACRS MEETING AUGUST 8-10, 1985 WASHINGTON, D. C.

Thursday, August 8, 1985, Room 1046, 1717 H Street, NW, Washington, D.C.

1) 8:30 A.M. - 8:45 A.M. Report of ACRS Chairman (0 pen) 1.1) Opening Statement (DAW) 1.2) Items of current interest (DAW /RFF)

2) 8:45 A.M. - 12:00 Noon General Electric Standard Safety Analysis Report (GE55AR II) (0 pen) 7.1) Report of ACR5 Subcomittee regarding the FDA requested for this project (D0/RKM) 2.2) Meeting with representatives of the NRC Staff and the Applicant (Note: Portions of this session may be closed as necessary to discuss Proprietary Information appIfcable to this proposal and detailed security provisions for this type of nuclear steam supply system.)

12:00 Noon - 1:00 P.M. LUNCH

3) 1:00 P.M. - 4:00 P.M. Seismic Qualification of Equipment in Operating Plants (open) 3.1) Report of ACRS Subcomittee on proposed procedures for seismic qualification of equipment in operating plants (CJW/AJC) 3.2) Meeting with representatives of the NRC Staff and the nuclear industry as appropriate

4) 4:00 P.M. - 4:20 P.M. NRCLongRangePlan(0 pen) 4.1) Report of ACRS Subcomittee regard-ing coments on the proposed NRC Long Range Plan (MWC/RKM)

5) 4:20 P.M. - 5:20 P.M. NRC Maintenance and Surveillance Program Plan -

W (0 pen) Report of ACRS Subcomittee ~

regarding the proposed NRC program plan for maintenance and surveil-lance requirements in nuclear power plants (GAR /HA) 5.2) Meeting with representatives of the NRC Staff and the nuclear industry

304th ACRS Meeting Agenda .

6) 5:20 P.M. - 6:00 P.M. Future Activities (0 pen) 6.1) Discuss anticipated Subcomunittee activities (MWL) 6.2) Discuss proposed ACRS activities (RFF)

O ae

, l 304th ACRS Meeting Agenda Friday, August 9, 1985, Room 1046, 1717 H Street, NW, Washington, D.C.

7) 8:30 A.M. - 11:30 A.M. San Onofre, Unit 1 (0 pen) 7.1) Report of ACRS Subcommittee regarding the SEP review for this project (CPS /HA) 7.2) Meeting with representatives of the NRC Staff and the Licensee

8) 11:30 A.M. - 12:30 P.M. Ap)lication of PRA to Indian Point and Otler Nuclear Power stations (open) 8.1) Discuss proposed ACR5 connents regarding application of the PRA done for the Indian Point Nuclear Power Station and other nuclear plants (D0/RPS) 12:30 P.M. - 1:30 P.M. LUNCH

9) 1:30 P.M. - 2:30 P.M. Management and Disposal of Radioactive Wastes (0 pen) .

9.1) Report of ACRS Subconnittees on Waste Management and on Procedures and Administration regarding pro-posed ACRS participation in NRC regulation of the DOE program for disposal of high-level civilian radioactivewaste(DWM/ DAW /OSM/RFF)

10) 2:30 P.M. - 6:30 P.M. Alvin W. Vogtle Nuclear Plant, Units 1 and 2 (0 pen)

TO.1) Paport of ACRS Subconnittee the OL request for this regarding(JCE/J05) facility 10.2) Meeting with representatives of the NRC Staff and the Applicant (Note: Portions of this session will be closed as required to discuss Proprietary Information applicable to this project and detailed security arrangements for this facility.)

304th ACR5 Meeting Agenda Saturday, August 10, 1985, Room 1046, 1717 H Street, NW, Washington, D.C.

11) 8:30 A.M. - 12:00 Noon Preparation of ACRS Reports (0 pen / Closed) 11.1) Discuss proposed ACRS reports on:

11.1-1) Seismic Qualification of Equipment in Opera ting)

Plants (CJW/AJC)(Open 11.1-2) San Onofre Nuclear Plant.

Unit 1 - SEP (CPS /HA)

(Closed) 11.1-3) Maintenance and Surveil-lance Program Plan (GAR /HA) 11.1-4) Vogtle Nuclear Power Plant (JCE/JOS)(Closed) 11.1-5) ACRS participation in NRC regulation of DOE progrom for disposal of high-level radioactive waste (DWM/OSM)

(0 pen)

(Note: Portions of this session will be closed as required to discuss information involved in an adjudicatory proceeding.)

12:00 Noon - 1:00 P.M. LUNCH

12) 1:00 P.M. - 3:00 P.M. ACRS Subcomittee Activities (0 pen) 12.1) Report of ACR5 ECC5 Subcommittee re-garding proposed amendment of 10 CFR 50, Appendix X, related ECCS matters, and the Davis-Besse loss of feedwater incident (DAW /PAB) 12.2) Report of ACRS Subcommittee on ACRS Procedures and Practices (DAW /RFF) 12.3) Report of ACRS Subcomittee regarding scram circuit breaker reliability (WK/PAB) 12.4) Report of ACRS Subcomittee regard-ing the INPO Radiation Protection Program (DWM/OSM) 12.5) Report of ACRS Subcomittee on the source term used in nuclear accident evaluation (WK/RPS/0H)

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MINUTES OF THE 304th ACRS MEETING l[ $'

AUGUST 8-10, 1985 WASHINGTON, D.C.

The 304th meeting of the Advisory Committee on Reactor Safeguards, held at 1717 H Street, N.W., Washington, D.C. was convened by Chairman D. A.

Ward at 8:30 a.m., Thursday, August 8, 1985.

[ Note: For a list of attendees, see Appendix I. D. Okrent and C.

Michelson did not attend the meeting. H. W. Lewis was not present on Saturday, August 10.]

Chairman D. A. Ward noted the existence of the published agenda for the meeting, and identified the items to be discussed. He noted that the meeting was being held in conformance with the Federal Advisory Comittee Act and the Government in the Sunshine Act, Public Laws92-463 and 94-409, respectively. He also noted that a transcript of some of the public porticns of the meeting was being taken, and would be available in the NRC's Public Document Room at 1717 H Street, N.W.

Washington, D.C.

[ Note: Copies of the transcript taken at this meeting are also available for purchase from Ann Riley & Associates, Ltd., 1615 I Street, N.W., Suite 921, Washington, D.C. 20006.]

I. Chairman's Report

[ Note: R. F. Fraley was the Designated Federal Official for this portionofthemeeting.]

Chairman D. A. Ward indicated that the Comission authorized the issuance of a full power operating license for Diablo Canyon Unit 2 on August 1, 1985. The Comissioners voted 3 to 2 in favor of the issuance of the Backfit Rule after incorporation of certain modifications, one of which involved the broadening of the definition of backfit. Comissioners Asselstine and Bernthal voted against issuance of the Backfit Rule suggesting several major modifications. Chairman Palladino subsequently reconsidered his vote and asked that the Backfit Rule not be issued until at least August 15 when the recomendations of Comissioners Asselstine and Bernthal could be considered.

D. A. Ward reported that on August 5, the Comission withdrew an -

advance notice of proposed rulemaking entitled " Severe Accident Design Criteria" and issued as of August 5,1985 the final policy statement " Policy Statement on Severe Reactor Accidents Regarding Future Designs and Existing Plants."

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MINUTES OF THE 304th ACRS MEETING AUGUST 8-10, 1985 II. AlvinW.VogtleNuclearPlant, Units 1and2OLReview(0 pen)

[ Note: J. O. Schiffgens was the Designated Federal Official for this portion of the meeting.]

J. C. Ebersole described the Vogtle Electric Generating Plant as consisting of two 1157 MWe Westinghouse pressurized water reactors similar to those at Comanche Peak and the SNUPPS units typified by the Wolf Creek Plant. He presented a brief discussion of the site, noted that the plant utilizes natural draf t cooling towers for normal operation, but additionally has two 100 percent capacity Class I mechanical draft cooling towers that serve as the ultimate heat sink. He indicated that the Subcommittee did not find any technical issues that would inhibit the issuance of an operating license. The plant appears to be a conservative design and the proposed staffing appears adequate (see Appendix IV).

J. C. Ebersole indicated that there will be a contested hearing for these operating licenses. At its July 18-19 meeting the Subcommittee heard two oral statements from members of the public.

Mr. T. Jones, challenged the utility's Readiness Review Program and pointed out that there is a growing list of allegations. A second speaker, Mr. Lawless, expressed concerns regarding radioactive contamination from the DOE Savannah River Plant and also concerns regarding the assurance of the integrity of the marl layer under the Class I Vogtle structures as a barrier between plant releases and the Tuscaloosa Aquifer for the modeling of groundwater flow rate and direction. The applicant indicated he has a defense against the presumed second problem.

O. Batum, Georgia Power, discussed the principle and unique design features of Plant Vcgtle (see Appendix V).

Ruble Thomas, Southern Services, explained that . natural draft towers provide cooling for the main turbine condensers and nonsafety related services. All safety related cooling is provided by the redundant mechanical draft nuclear service cooling water towers. H. Etherington asked if the forced draft tower has any capacity as a natural draft tower. Ruble Thomas indicated thr.t the plant does not count on that. J. C. Ebersole asked if there is redundancy since he noted that there are three 50 percent pumps per tower. Ruble Thomas indicated that there are redundant towers and O. Batum noted that there is a transfer pump between towers.

O. Batum mentioned the Aerojet Energy Conversion System supplied the Radwaste Volume Reduction System which uses fluidized bed combustion and a calcination process. The solidification system is made by Stock Equipment Company and uses a polymer solidification for dry solids and cement solidification for liquids. In answer to -

questions by D. W. Moeller, O. Batum indicated that the system is primarily for resins and dry wastes. The resins are processed by fluidized bed combustion. A reduction factor of about 10 is achievable. D. W. Moeller asked if the State of Georgia is part of a low level waste compact. Ruble Thomas indicated that there is a 2-l

" ' MINUTES OF THE 304th ACRS MEETING AUGUST 8-10, 1985 Southeastern Regional Compact which includes states from Virginia to Louisiana and Tennessee. These states have made a compact among themselves under the law but it is yet to be approved by Congress.

O. Batum indicated that, as of July 21,1985, Unit 1 construction is approximately 82 percent complete and Unit II construction is about 48 percent complete. Fuel loading for Unit 1 is currently scheduled for December 1986 and commercial operation for June 1987.

M. V. Sinkule, NRC Region II Section Chief of the Project Section, discussed the results of inspections done at Plant Vogtle. He explained that the inspection program is primarily aimed at a review of safety-related activities but has been expanded into areas that are important to safety or relate to safety. The review primarily consists of verifying that the coninitments made in the FSAR and the requirements of the NRC regulations are being implemented at the site (see Appendix VI). He described the NRC resident and specialist inspection efforts and noted that the resident inspectors currently spend 40-50 percent of their time observing activities in the plant. The resident inspection program began at Plant Vogtle in November 1982. Resident inspectors observe work on back shifts and week-ends as well as regular day shifts. He indicated that the NRC has not issued any stop work orders, and the Licensee has been very responsive to NRC concerns.

Vogtle has actually had a lower number of non-compliance items than other plants in the same phase of construction as Vogtle. ' Concerns found were classified as Severity Level 4 or Level 5. Severity Level 5 implied minor or no safety significance while severity Level 4 implied the existence of safety potential. The Staff is tracking by computer about 140 open items, construction deficiencies and other noncompliance items. Region II does not anticipate a problem in resolving these items.

M. V. Sinkule described the process of NRC handling of allegations.

The process involves recording an item, evaluating it, and tracking it to conclusion. J. C. Ebersole asked if the allegations are hampering the inspection effort. M. V. Sinkule indicated that the inspector usually tries to factor the evaluation of an allegation into his normal inspection process. W. Kerr asked if the Staff's followup of allegations has indicated that any could have had a significant effect on plant safety. M. V. Sinkule thought it had not. He characterized the majority of the 20 percent of the allegations that could be substantiated as having minor safety significance.

M. V. Sinkule indicated that four Systematic Analysis of Licensee Performance (SALP) reviews have been conducted at Plant Vogtle and when compared against four other plants in Region II, Vogtle was found to be above average. F. J. Remick asked if there had been a INPO construction performance evaluation of either Vogtle Unit. D.

l 0. Foster, Georgia Power, indicated that the Vogtle Plant was the "

i initial project for an INP0 pilot evaluation and a self-initiated l evaluation was performed as well as an additional INP0 assessment l in 1984. Another INPO review is due on Septerrber 9,1985 which l will be the fourth review. F. J. Remick asked if there were any t

L.

MINUTES OF THE 304th ACRS MEETING AUGUST 8-10, 1985 significant findings. D. O. Foster indicated that there were a few findings which involved a limited number of hardware changes or corrections, particularly in the piping area. F. J. Remick asked if the INP0 results were consistent with the NRC's inspection results. D. O. Foster indicated that he thought so. He mentioned a problem with shop welds in the piping area identified by INP0 which was subsequently corrected. F. J. Remick asked if any suggestions were made on quality assurance management or the plant organization. D. O. Foster indicated that some constructive suggestions were made by INP0 which involve a limited number of hardware changes or corrections. V. Sinkule stated Region 11 conclusions that the Vogtle inspection and enforcement programs are on schedule and that there will be sufficient NRC resources to complete the remaining inspection work.

M. Miller, NRC Licensing Project Manager for Vogtle Units 1 and 2, presented an overview of the Vogtle licensing review noting that all open items for the most part have been resolved (see Appendix VII). She indicated that the SER issued this past June had 14 open items, 50 confirmatory issues and 11 license conditions. W. Kerr asked about the issue of training of emergency diesel generator personnel. M. Miller indicated that the issue primarily concerned replacement personnel and their hands-on experience. The Staff was concerned that hands-on experience would not be readily available 5 to 7 years down the road when new maintenance personnel are hired.

The licensee promised that new personnel will be under the direction of a qualified individual for the applicable maintenance task. J. C. Ebersole raised some question about the Applicant's reanalysis of sump bicckage and NPSH margin. Ruble Thomas assured the Comittee and the NRC that the issue has been adequately addressed. D. W. Moeller inquired about the issue of toxic gas evaluation of chemicals and the procedures for teaching control room personnel how to recognize chlorine gas emissions. P. G.

Shewmon asked regarding the issue of diesel fuel oil storage tank cathodic protection. Ruble Thomas indicated that Georgia Power is trying to justify that it does not need the zinc or other anode coating over the tank. M. Miller indicated that Georgia Power is using a coating other than zinc and a tar epoxy oler that. W. Kerr asked for the soil resistivity. R. Thomas indicated that it varies from 281 to 4000 ohm meters. He noted that 200 to 500 ohm meters is mildly corrosive and higher than 500 ohm meters is not corrosive. The resistivity for most of the soil at the site is 400 to 500 ohm meters but there are a few points in the mildly corrosive range. He pointed out that in addition all pipes are coated and wrapped with tape. W. Kerr expressed interest in why there was an open item regarding the emergency response capability since Georgia Power is also responsible for the Hatch plant for which a plan was drawn up in the 1980 to 1981 time frame. D. W.

Moeller thought it would be a good idea if Georgia Power had a cooperative arrangement with the Savannah River plant which is ~

nearby. It would be to the advantage of Savannah River to know the

! Vogtle River Emergency Team in case there were a release from the Savannah River site. S. Ewald, Georgia Power, indicated that Vogtie does have a formal agreement with DOE's Savannah River 1

. MINUTES OF THE 304th ACRS MEETING AUGUST 8-10 0 1985 Plant. It is one of the Appendixes to the Vogtle Emergency Plan.

R. Axtmann asked about Human Factors Engineering items which were open. M. Miller indicated that these referred to the detailed control room design review and the safety parameter display system (SPDS). The Applicant has agreed to incorporate the Staff's feedback and is preparing a summary report. J. C. Ebersole pointed to the Davis-Besse incident where there were two buttons which if pushed could instantly starve the plant of feedwater. He asked if a review of the control boards took account of such vulnerabilities as these which could kill critical functions. D. Becker, NRC, indicated that human engineering deficiencies associated with the steam and feedwater rupture control system at Davis-Besse were previously identified deficiencies in the Davis-Besse control room design review. This type of problem was looked at in the Vogtle control room design review.

F. J. Remick asked if the human engineering deficiencies identified for the Vogtle control room suggested any major modifications. M.

Miller indicated that four identified human engineering deficiencies were considered safety significant and one did involve a number of wiring changes. D. W. Moeller noted the use of sodium hypochlorite at some nuclear plants in part because of the hazard of chlorine intrusion into the control room. He asked the Applicant if the Staff had encouraged the use of sodium hypochlorite over chlorine for algae or slime control. R. Thomas indicated that 2e Staff had never discouraged the use of chlorine and chlorine is used at the Farley and Hatch plants as well as Vogtle. He noted that the sodium hypochlorite is quite a bit more expensive than chlorine. D. W. Moeller comented that if the costs of installation of chlorine monitors or the cost of an incident involving chlorine monitors inadvertently actuating and putting a control room on emergency recycle were factored into the calculations, sodium hypochlorite might appear more cost effective.

M. Miller indicated that as a result of the Subcommittee meeting a question remains regarding Staff criteria on battery duration in the event of a station blackout. J. Knight, NRC, indicated that there are no specific criteria to dictate capacity or length. of time a battery ought to last. The length of time a battery ought to last is dictated by plant design requirements. About the only criterion the Staff has is the single failure criterion which requires that if there is a loss of offsite power the charger will come bark on and DC will be supplied by the chargers. J. C.

Ebersole recalled a ncminal number of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for battery life which could be extended by dropping a number of nonessential loads from the DC bus. He thought it would be well for the Staff to develop sharply defined criteria for battery duration irrespective of whether the battery is unloaded. R. Thomas indicated that the expected life of the battery at Vogtle was 2 3/4 hours. -

D. O. Foster gave a brief overview of the Vogtle project organization and its responsibilities to the Vogtle project (see Appendix VIII). D. Beckham, Vice President and General Manager of Nuclear Operations for Georgia Power Company, dicussed Georgia

, MINUTES OF THE 304th ACRS MEETING AUGUST 8-10, 1985 Power's management philosophy. The key point is that safety is first and it takes priority in the construction and operation of Georgia Power nuclear facilities (see Appendix IX). In answer to a question by D. W. Moeller, J. T. Beckham indicated that, by definition, Georgia Power strives for excellence. J. C. Ebersole asked how Georgia Power controls the eager-beaver type individual who does things which turn out like the Davis-Besse incident. J.

T. Beckham indicated that there would be disciplinary acticr. taken against an individual who does not follow procedures or recognize that the planning of an action is just as important as actually carrying out that operation. There is an organized program for all operators to inculcate the Georgia Power philosophies of excellence and attention. He indicated Georgia Power is very proud of its record in health physics, as is exemplified by the low exposure experience at some of their large projects during normal operation.

He indicated that Georgia Power intends to keep exposures low. D.

W. Moeller praised Georgia Power's management comitment to excellence and to good radiation protection since both are very important to success.

J. T. Beckham discussed the corporate organization (see Appendix X). G. Bockhold, General Manager of Nuclear Operations for Georgia Power, discussed the plant operation organization, staffing and training (see Appendix XI). He discussed the Yogtle-specific dedicated training center which contains an office, a simulator, a classroom wing, a laboratory wing, and in the basement of the classroom wing, an emergency operations facility. He listed the various training programs available, programs in the area of engineering, quality control, health physics and chemistry, instrumentation and control, mechanic, electrician, nonlicensed reactor operator, senior reactor operator, shift technical advisor, fire brigade and general employee training. F. J. Remick asked if these training programs are performance-based or, based on job and task analysis. G. Bockhold indicated that Georgia Power has done some initial job and task analysis and is continuing the process to achieve accreditation from INPO with a complete set of requirements. However, much of the initial training is based upon the older experienced-based training. F. J. Remick asked if all of these training programs have retraining program components. G.

Bockhold indicated that, where appropriate, and in the case of the training programs for which accreditation from INP0 is sought, there will be retraining programs. In some cases, a retraining component would not be appropriate. F. J. Remick asked how many shifts are planned for the Vogtle plant. G. Bockhold indicated that there are five shifts but the plant can accomodate a considerably higher number because of the number of shift supervisors and reactor operators. The plan is to utilize the people most effectively and minimize their stress. Therefore, there will be more shifts as more people become available although F. J. Remick asked if

~

the plant is comitted to five shift .

Georgia Power has considered what it would do if the Comission provides the option of using an STA as a second SRO. He asked if Georgia Power has decided which way it will go or whether it will stick with a separate STA. G. Bockhold indicated that on a limited 1

MINUTES OF THE 304th ACRS MEETING AUGUST 8-10, 198S basis Plant Vogtle intends to use the SR0s with degrees as separate STAS. He indicated that the STA path for plant Vogtle is the path for the engineer to get on shift as a shift supervisor and then to progress upward toward eventually becoming a plant manager. F. J.

Remick asked if there is encouragement or a requirement that STAS be licensed SR0s. G. Buckhold indicated that Georgia Power strongly encourages most of the STAS to be licensed as SR0s. F. J.

Remick asked if there are selection criteria for hiring especially in the area of maintenance or reactor operations. G. Buckhold indicated that Georgia Power uses the POSS test, the Plant Operator Selection System from the Edison Electric Institute, for operators.

F. J. Remick asked if any kind of psychological screening is used for employees. G. Buckhold indicated that Georgia Power uses the MMPI for aberrant behavior psychological screening as part of its security program. F. J. Remick asked if supervisors are trained in observation of aberrant behavior. G. Buckhold indicated affirmatively. P. G. Shewman asked it their are any mechanical aptitude tests for the maintenance people. G. Buckhold explained that Georgia Power does not yet use the MASS test.

P. D. Rice, Vice President and General Manager of Quality Assurance for Georgia Po.ver, discussed the quality assurance organization (see Appendix XII). He indicated that all quality assurance activities on the Vogtle project are under the umbrella of the Georgia Power Company Vogtle Quality Assurance Manager. He.briefly discussed the experience of the Vogtle quality assurance (QA) organization regarding the formal five phase training program for the quality assurance organization which includes written as well as oral examination boards. F. J. Remick asked if the training program has a continuing training aspect. P. D. Rice indicated that there is no formal requalification but there is a continuing training program that parallels ANSI requirements with regard to minimum qualification requirements. He indicated that over 900 audits have been conducted associated with Vogtle under the control of the Vogtle Quality Assurance Manager. The organization has tried to be innovative with a prevention-oriented philosophy by bringing in individuals who have expertise from outside when areas are being examined for which current expertise within the organization does not exist. J. C. Ebersole mentioned an event which occurred at the Hatch plant on August 25, 1982 where a dump valve failed and heated up the plant environment producing an uncomfortable situation for a long period of time. He asked if this incident had been analyzed in the context of operational quality assurance. P. D. Rice indicated that the incident was investigated and as a result more emphasis has been placed on the surveillance program and in particular performance-type surveillance as opposed to area-type surveillances. J. C. Ebersole asked if the Vogtle plant belongs to the NPRDS system. J. T.

Beckham indicated that Georgia Power is comitted to NPRDS both at Hatch and Vogtle. J. C. Ebersole asked about followu) toward -

individual equipment failure in a QA context. P. D. Rice indicated that the responsibility for input of data to NPRDS is part of the line organization of the company. The quality assurance organization is involved with NPRDS with both surveillance and 7

MINUTES OF THE 304th ACRS MEETING AUGUST 8-10, 1985 audit programs for continuing followup. G. A. Reed asked how quality control is handled et Plant Vogtle. D. Rice indicated that

t. the quality assurance organization' for Vogtle and within Georgia Power Company is a line organization along with the craftsman and skilled individuals.

P. D. Rice defined the Vogtle Quality Concern Program as one that identifies concerns by employees, that addresses those problems with resolution by corrective actions, and that provides feedback to the individual or individuals who identified the concern. He noted that Georgia Power has an employee concern program to solicit concerns (allegations) and get them resolved. This program evolved to the point where in 1983 it was formalized as the Quality Concern Program. Employee concerns are routinely published in the plant newspaper "the Blazer." D. W. Moeller asked what percentage of coments received are unfounded. P. D. Rice indicated that Georgia Power takes a very conservative approach which has led to a rejection rate of about 20 percent as being unsubstantiated. F. J.

Remick thought that the Quality Concern Program was commendable.

P. D. Rice indicated that the program is very beneficial and worthwhile and has a very clear payoff.

P. D. Rice explained that the Vogtle Readiness Review Program is to provide added assurance that Georgia Power will be ready to satisfactorily operate Plant Vogtle. It provides a systematic ar.d interactive mechanism between Georgia Power Company and the NRC for their independent reviews, verification of Georgia Power work processes, and the phased approval of work activities. It is a program intended to assess and measure compliance with comitments in the FSAR (see Appendix XIV). He indicated that the Readiness Review Program Manager, in addition to discipline managers with a number of years of design experience who carryout the day-to-day readiness review activities, had a small quality assurance staff to oversee the activities within the program. The program manager also has an independent design review group made up of experienced Stone and Webster Engineers to carry out the design verification process. He indicated that Georgia Power has made a significant commitment of resources to this program and expects to spend 150 man-years to complete the Readiness Review Program activities. He spoke in detail of the preparacion of readiness review modules, a self-assessment process of problem identification and correction, which is reviewed and eventually submitted to the NRC. He noted that the NRC review was described and defined in a policy letter from the E00 to the NRC Comissioners dated April 8,1985. Once the NRC review has been completed and Georgia Power Company has responded satisfactorily to any questions or concerns raised by NRC, NRC would accept the scope of the work covered by that module.

D. A. Ward questioned whether this program which is costing Georgia Power $20,000,000 is a cost-effective effort for both Georgia Power and the NRC. He noted that the NRC Staff is spending 10 man-years -

participating in the Vogtle readiness review, and typically spends 20 man-years on an operating license review including support for the hearings. He questioned whether this is an effective use of 10 man-years of NRC effort. P. D. Rice suggested that many of the 8-

MINUTES OF THE 304th ACRS MEETING AUGUST 8-10, 1985 actions that are being taken in this program will be done anyway.

This program will in effect substitute for the IDVP or IDI evaluations done on other plants but will be done in a timely manner instead of as a crash program. M. V. Sinkule briefly explained NRC's participation in the Vogtle readiness review (see Appendix XV). R. Thomas indicated that it is Georgia Power's intention to do a careful assessment early in the project to be sure that ccmitments are incorporated into the design documents and the specifications, and procurement is done correctly the first time. W. Kerr thought that this was an inherent part of the process of building nuclear power plants and was normally handled by a quality assurance organization. He hinted that this was an additional layer of effort. M. V. Sinkule defended the NRC participation as representing a prudent comitment of resources early in the process when compared to resources expended at other nuclear plants later in the process. He suggested that the overall cost of this program to the NRC may be less than a pure add-on program at another plant and that the readiness review is a review usually done for the NRC by I&E. J. C. Ebersole thought that he would rather see the $20,000,000 spent in search of an improved residual heat removal system for the plant.

F. J. Remick asked where besides the Vogtle control room there will be readouts for the SPOS. G. Buckhold indicated that there were readouts in the technical support center and the efnergency operations facilities in addition to a simulated SPDS as part of the plant-specific simulator. D. A. Ward asked about the l reliability of the SPOS and whether there was redundancy in the computers. G. Buckhold indicated 99 percent plus reliability even without redundancy in the computers. He stated that there are good spare parts as well as a spare computer on site. D. W. Moeller asked if Georgia Power is familiar with studies being conducted by the NRC Staff in the area of control room habitability. R. Thomas indicated that a NUREG document on control room habitability was considered in the Vogtle assessment. D. W. Moeller noted that Georgia Power did not appear to be familiar with the Working Group Report on control room habitability and he suggested that a copy of the report be acquired as soon as possible.

J. C. Ebersole expressed interest in the probable challenge frequency for the auxiliary feedwater system. R. Thomas indicated that Georgia Power expects about 10 to 15 challenges per year when the Vogtle Plant first starts up and to have that reduced to between 5 and 10 shortly af terward, J. C. Ebersole asked if the Staff had any comment on this matter. W. Lefave, Auxiliary Systems Branch, indicated that the Staff's consultants had done a reliability study on the auxiliary fydwater system and came up with a very conservative number of 10 challenges per year. This con asts with failure rates determined by the Applicant of about 10 -

i J. C. Ebersole noted that the Vogtle Plant is somewhat unique in having safety-grade power operated relief valves (PORV). He asked

[ how Georgia Power intends to use these in the safety context. R.

l  !

t

MINUTES OF THE 304th ACRS HEETING AUGUST 8-10, 1985 Thomas indicated that their principal function is to provide the capability to go to a cold shutdown using only safety grade equipment. Such qualified valves provide more confidence regarding their operability. J. C. Ebersole noted that there are safety-grade PORVs on the secondary side as well. He suggested that this qualification makes the Vogtle Plant considerably more conservative regarding rapid depressurization than the Palo Verde Nuclear Plant. G. A. Reed pointed out that these safety-grade PORVs are internal pilot-operated relief valves whose performance in borated hydrogenated environments is suspect. R. Thomas explained that the pilot-operated relief valves are water-sealed.

There is a seal between the steam space in the pressurizer and the PORV. This separates them from the hydrogen.

T. Epps, Manager of Inspection Testing and Engineering with Southern Company Services, explained Georgia Power's own method of UT examination of cast stainless steel piping (see Appendix XVI).

He indicated that Southern Company Services recognized early en that some concern existed in the nuclear industry regarding the effective examination of cast stainless steel material. Therefore, an internal program was launched to find the best technique possible to demonstrate effective examination of that material at the Vogtle site. P. G. Shewmon asked if this technique has been tried on weld repaired pipes, valves, or elbows. He noted that all statically cast stainless steel pipe is usually weld repaired. T.

N. Epps noted that Georgia Power's technique is able to see the counter bors on the statically cast elbows on that side of the weld. He indicated that the signal came back consistently all around the elbow.

W. Kerr asked if Georgia Power has done a PRA on Plant Vogtle. R.

Thomas indicated that a PRA has not been done. A detailed tault tree type review of the Bechtel and Westinghouse design has been done for about 28 systems and their support systems.

III. San Onofre Unit I SEP Review (0 pen)

[ Note: H. Alderman was the Designated Federal Official for this portionofthemeeting.]

C. P. Siess reviewed the Systematic Evaluation Program (SEP) noting that for the 11 plants in the program SEP was intended to determine the extent of compliance with current criteria. One plant, Humboldt Bay, dropped out of the SEP and the remaining 10 plants were divided into two groups. Group 2, which was composed of the five oldest plants still in operation, contained Yankee Rowe, Big Rock Point, LaCross, Haddom Neck and San Onofre Unit 1. The five plants in group 1 were of somewhat more recent vintage. Five of the SEP plants have not been issued an FTOL but have been operating under provisional operating licenses. San Onofre Unit 1, the last of the SEP plants to come before the ACRS is one of those plants with a provisional operating license. The continuation of the SEP, the Integrated Safety Assessment Program (ISAP), is now underway on MINUTES OF THE 304th ACRS MEETING AUGUST 8-10, 1985 a trial basis for the Millstone 1 and Haddam Neck Plants. ISAP is expected to continue subject to budget limitations.

• C. P. Siess explained that the SEP was carried out in two phases.

The first phase derived a set of 137 review items. In phase two each plant was assessed against current criteria for those 137 items which were applicable (some were BWR items and others PWR items). It was decided that TMI Action Plan items would be han.iled generically. Generic items and USI's are being looked at separately by the Staff and would be excluded from the SEP review.

Where a plant-specific PRA was available, it was used. Where a PRA was not available, the SEP used insights from a PRA for the nearest-plant type with due regard to uncertainties. The review currently before the ACRS is the next step in the program, an integrated plant safety assessment which is documented in an integrated plant safetyassessmentreport(IPSAR),

C. P. Siess indicated that the Subcommittee on San Onofre met in November,1984 to review the Unit I seismic reevaluation program.

Seismic reevaluation has 'been an issue for every SEP plant as has flooding and tornadoes. This is because these plants were not specifically designed for extreme external phenomena. Other topics were covered at a Subcommittec meeting held in June,1985. J. C.

Mark noted that San Onofre does not have a full term license and he asked if that license conversion can be discussed at this session.

C. P. Siess explained that the license conversion is a separate operation which will come later as it did for Ginna. At that time TMI items, USIs, and other generic issues will be handled. J. C.

Mark asked regarding the severe accident policy statement whether the SEP review will be concerned with determining a core melt frequency for the San Onofre plant. C. P. Siess pointed out that there are no current criteria that speak to core melt probabilities. J. C. Mark asked if San Onofre 1 is currently a subject of contention with the populace. M. Medford, Southern California Edison (SCE), indicated that San Onofre 1 is currently the subject of litigation on the part of the Sierra Club and others in the Ninth Circuit Federal Court regarding restart of the plant in late 1984.

M. Medford reviewed the history of the San Onofre plant (see AppendixXVII). J. C. Mark expressed interest in whether the beach front adjacent to the plant site has controlled access. M. Medford indicated that the beach is controlled (an issue in the San Onofre Units 2 and 3 proceedings). He explained that the beach is just on the other side a seawall. The current configuration has a beach walkway imediately seaward of that seawall. There is fencing which extends out into the ocean at either end of the site and signs posteo all along the walkway. Fencing defines the walk- way to discourage public use of that section of the beach. Site .

~ ~ security forces monitor that section of the beach and discourage use of it.

M. Medford indicated that the San Onofre 1 continues to use fuel that is stainless steel clad. J. C. Ebersole asked why the plant MINUTES OF THE 304th ACRS MEETING AUGUST 8-10, 1985 still uses stainless steel instead of going to zircoloy cladding

which he thought was less expensive. The response was that the plant stays with stainless steel because it has worked well. The change would necessitate modification of the licensing bases of the plant and its operation. M. Medford thought that the changes that would be required to go to zircoloy cladding would not be worth the economic benefits. G. A. Reed noted that there is often a greater tritium release with stainless steel cladding. He asked if San Onofre Unit I has had such difficulties. M. Medford indicated that zircoloy cladding is used at San Onofre Units 2 and 3 and there has been more difficulty with San Onofre Unit 3 which has some fuel leaks than at the other two units. The answer is no. J. C.

Ebersole asked if the auxiliary feedwater system was a safety-grade system when first installed. It was not. J. Rainsbury, SCE, discussed how the auxiliary feedwater system was upgraded to safety-grade by replacement and seismic upgrading of the discharge and suction piping. The operation of pumps has been revised to provide automatic start both for the turbine-driven pump and the electric-driven pump. A new seismically qualified auxiliary feedwater tank was also installed. J . ~ C. Ebersole expressed concern that San Onofre 1 has only one turbine-driven and one electric-driven pump on the auxiliary feedwater system. J.

Rainsbury explained that a third auxiliary feedwater pump will be installed as part of the fire protection modifications at the next outage. In answer to an additional question by J. C. Ebersole, J.

Rainsbury indicated that the main feedwater pumps as well as booster pumps in front of them, are electrically-driven. J. C.

l Ebersole asked if San Onofre Unit I has had trouble with scram breakers as has occurred at Salem. M. Medford indicated that they have not had such a problem and attribute that to good maintenance.

l C. p. Siess briefly described the seismic upgrade program which took Unit 1 from 0.59 to 0.679 and a modified Housner spectrum.

This was not quite the level of the regulatory guide spectra required for San Onofre Units 2 and 3. The Ccnnission allowed Unit 1 to start up in 1984. All of the modifications in the seismic upgrade to provide the capability to achieve a hot standby condition from full power operation using only seismically-qualified equipment had been completed. At that time not all of the equipment needed for accident mitigation had been qualified. The Licensee was to complete that work by the the end of the next refueling outage. C. Grimes, NRC, indicated that a long term service plan, proposed by the Licensee, is currently under Staff review. J. C. Ebersole asked if the issue of relay chatter has been laid to rest at this point. C. P. Siess explained that that will be part of the resolution of USI A-46. The Licensee has been looking at equipment anchorage but the relay chatter question has not been settled. In answer to F. J. Remick, M.

Medford indicated that the Staff's authorization in late 1984 for

• the plant to return to service without completing the seismic upgrades was under litigation. The Sierra Club contended that the public should have had a right to an ASLB hearing before restart was authorized.

i  !

MINUTES OF THE 304th ACRS MEETING AUGUST 8-10, 1985 E. McKenna, NRC SEP Integrated Assessment Project Manager, San Onofre Unit 1, presented a sumary of the 137 topics in the Phase 2 SEPReview(seeAppendixXVIII). She indicated that 86 issues were

• identified from the 36 topics which were reviewed for the Plant.

Forty one of the issues required no additional action. There were hardware modifications identified for five issues. Procedural or technical specification changes were identified for 12 issues and 26 issues were identified for further evaluation for potential backfits. Two issues were identified as unresolved. C. P. Siess noted that the plant-specific topic of technical specifications was deleted from the SEP program and would be addressed after completion of the integrated assessment. C. Grimes indicated that the reason for handling the technical specifications separately was more procedural than technical. In addition, there was a policy decision made that the standard technical specifications would not be forcibly backfit to operating plants.

J. C. Mark noted that a modification made during the SEP review was the removal of the Tsunami gates. E. McKenna explained that the issue arose because if the Tsunami gates were inadvertently closed one would isolate the Plant from its ultimate heat sink, the Pacific Ocean. The Licensee decided that they did not need the Tsunami gates in the first place. So they removed them.

D. W. Moeller brought up the issue of radiation monitoring of the component cooling water system. J. Rainsbury explained that the component cooling water system is the intermediate closed loop i between the reactor coolant system and salt water cooling (ocean water). Radiation monitoring is used to detect leakage from the reactor coolant to the component cooling system. There are three pumps and two heat exchangers in the safety grade component cooling water system. It is a treated water system to prevent corrosion.

J. C. Ebersole asked if there is an emergency means to fill the component cooling water system on an open cycle basis from the service water system. J. Rainsbury indicated that there was not.

J. Rainsbury indicated that SCE does not plan on losing all the water from the component cooling water system. This system is included in the seismic upgrade program and is provided with a surge tank for making up leakage. J. C. Ebersole pointed out that there is no cross tie to a infinite water supply for emergency cooling. J. Rainsbury agreed. C. Grimes pointed out that it is rot a question of core cooling but maintaining pumps and valves.

Core cooling heat is rejected through the steam generators or throuch the auxiliary feedwater system. J. Rainsbury noted that this particular system was not one of the systems required to maintain the reactor in hot standby. In the event the system is lost, there are procedures available to handle the situation. The critical item is protecting the reactor coolant pump seals. The

,, auxiliary feedwater pumps do not depend on component cooling water. .

The Staff ranked a number of topics based upon insights from its limited probabilistic risk assessment. Those that ranked high (having an effect on core melt frequency) were ventilation systems for "4160 volt switch gear / cable spreading room" and the " battery MINUTES OF THE 304th ACRS MEETING AUGUST 8-10, 1985 and inverter room." D. W. Moeller asked why the control room emergency ventilation system was not included in a PRA. C. Grimes indicated that control room habitability was a multi-plant action and a TMI Action Plan requirement and would be handled outside of the SEP. C. Grimes recalled that there is an issue of " control room ventilation system" as a single train with redundant active components which will be addressed under TMI Action Plan 111.D.3.4.

J. Rainsbury noted that the control room ventilation system does not have redundant active components. J. C. Ebersole expressed

.oncern regarding salt intrusion and corrosion in plant electrical systems since the San Onofre plant is on the Pacific Coast. C.

Grimes indicated that, during the Staff's review of San Onofre Unit 1 plant electrical systems including the new diesel generator systems, he could not recall either operating experience or design differences that would lead the Staff to believe that there was a potential problem with salt water intrusion in electrical components. It was pointed out that the Pacific Coast does not experience the hurricane type storms that the Gulf and Atlantic Coasts experience. ,

J. C. Mark noted that an issue of wind and tornado loadings (load combinations) was in a category of issues requiring further evaluation. M. Medford noted that tornadoes originate over flat terrain. The San Onofre site is pinned between the Pacific Ocean on one side and a range of fairly substantial hills within a few miles on the other side. A site specific tornado study had been done and is currently a matter of discussion between SCE and the NRC Staff. The Comittee discussed the severity of wind conditions the plant is designed to withstand. C. P. Siess noted that the issue here as with other SEP plants is tornado missiles. According to C. Grimes, tht: tornado studies are based upon Mcdonald's report which derives statistical measures and offers expert opinions on how the wind speeds at various probability 1evels are distributed.

H. W. Lewis thought that the misuse of statistics and the confidence limits from Mcdonald's report were not convincing. J.

C. Mark asked the historical cause of the change in the safe shutdown earthquake from 0.59 to 0.679 . C. P. Siess inoicated that ,

it was the result of the discovery of an offshore fault (not the Hosgri fault). M. Medford indicated that the controlling feature  ;

was the offshore zone of deformation and a change in the assrissment of the event that could be caused by that feature. There were no new discoveries about the feature itself. The change came about as a result of the review on Units 2 and 3 in 1972. ,

D. W. Moeller asked regarding the issue of toxic gas monitors. E.

McKenna indicated that the control room does not have toxic gas monitors or automatic isolation of the ventilation systen. During the review of potential industrial hazards from shipments on the highway and the railroad, it was determined that the probability of a toxic gas cloud reaching the control room intake was higher than 4 current criteria. C. P. Siess noted that J. Hendrie, NRC Staff consultant, was not very enthusiastic about toxic gas monitors because of the potential for false alarms. E, McKenna, in answer to a question by D. W. Moeller, noted that there are toxic gas l

14

MINUTES OF THE 304th ACRS MEETING AUGUST 8-10, 1085 monitors at Units 2 and 3 and quick comunication among the three control rooms such that if toxic gas affects Units 2 and/or 3 they would call this to the attention of Unit 1. D. W. Moeller asked if the Staff would accept such a solution. C. Grimes noted that this is a matter being discussed at the present time.

E. McKenna briefly discussed two areas where agreement has not been reached between the NRC Staff and the Licensee. The first is the area of tornado missile protection regarding the hazard functions to be used in an analysis. She pointed out that the Staff is in the process of doing an evaluation of the effects of both the tornado wind and missiles on plant equipment. C. Grimes indicated that the number and types of missiles to be evaluated is dependent upon which shutdown system or which shutdown approach the Licensee plans to use, given a loss of specific equipment. One possible alternative to resolve this issue would be to take the EPRI methodology, which considers both the missile sources and strike probabilities in order to define the critical accident sequence, and put that into context. One then gets a probability of multiple nissile strikes. C. P. Siess suggested that one has a precedent for a probabilistic approach. C. Grimes indicated that the precedent was actually established via the use of the EPRI methodology on NTOL applications. It is considered current criteria. J. Rainsbury indicated that SCE is using the EPRI method.

The other issue that has not been resolved involves the ESF switchover from injection mode to recirculation mode. The issue involves the dual function of the main feedwater ] umps. In the case of a large break LOCA, the feedwater pumps are lined up to the primary icop in the safety injection mode. M. Medford indicated that San Onofre Unit I has comitted to providing two sets of pumps so that the feedwater pumps will no longer serve a dual function.

The issue came up because it requires a manual switching procedure and the refueling water storage tank is rapidly deplete:1. There is a question whether there is sufficient time for the operator to complete all of the valve realignments to establish the alternate flow paths. The Staff is seeking automatic tripping of the pumps to provide more time, to realign the flow paths, and to have double alarms since it is necessary to reestablish suction af ter the refueling water storage tank is drained before there is damage to the charging pumps.

C. P. Siess noted the incompleteness of a list of issues which required hardware modifications. He suggested that there must have been some seismic modifications made to the plant at substantial cost. M. Medford indicated that it had cost SCE $125,000,000 to date. C. P. Siess asked how much of that cost was analysis. M.

Medford indicated that about 10 percent or a little over

$10,000,000 was for analytical work. .

J. C. Ebersole asked if San Onofre Unit 1 had degradable fibrous thermal insulation on the high temperature piping inside containment and if the insulation could beccme a slurry in the

~. - - - _ _ _

MINUTES OF THE 304th ACRS MEETING AUGUST 8-10, 1985 event of a LOCA. C. P. Siess noted that this was a containment sump issue and C. Grimes indicated that it was included in generic issue USI A-43. Reflective insulation is used inside containment

• and plaster insulation is used only outside containment. M. Short, NRC, indicated that nearly all the insulation is the reflective metal type. There are areas where coated cloth covers fiberglass type insulation. it is typically used on the bell portions of pumps or the hemispherical portions of the vessel and steam generators which are difficult to cover with reflective insulation.

K. Baskin, SCE, had mixed opinions with regard to the SEP program.

From a positive standpoint it forced Soutnern California Edison as well as the Staff to look at certain aspects of the plant on an integrated basis and identified some concerns that would not have been looked at otherwise. A good example of this was the need for some improvement in ventilation systems. He also thought that the basic philosophi of comparing against current criteria and then allowing the St 'f f to use judgment as far as the degree of compliance is a very positive aspect. On the negative side the overall program was really overwhelmed by the seismic reevaluation effort. The seismic reevaluation was begun in 1972 with analyses of the reactor coolant system, the containment, and the reactor building. In the 1977 to 1978 time frame San Onofre Unit I was included in the SEP. As a result SCE lost two years in its seismic reevaluation effort with no compensating benefit. Overall, he thought that the SEP program should be viewed as a slight negative from SCE's perspective. C. P. Siess asked if SCE intends to continue in the ISAP program. K. Baskin indicated that the decision has not been made yet. C. Grimes pointed out that SCE has proposed an integrated schedule that combines their current requirements and plant improvements with the SEP items. They are therefore half way in between an ISAP and an integrated schedule.

C. Baskin commented that SCE has invested in excess of $500,000,000 in backfitting San Onofre Unit 1. Since about 1977. The original cost of the plant was $87,000,000. The result has been some question on the part of the California Public Utilities Comission regarding the viability of backfits. There is a question as to whether the California Public Utilities Comission will allow an additional $50,000,000 to $100,000,000 budgeted next year to be passed through into the rate base. Because of the significant expenditure associated with seismic backfit, SCE is under an obligation to obtain the Public Utility Commission's approval for additional capital outlays prior to their expenditure.

The Comittee completed its review and prepared a report to the Comission (see ACRS letter dated August 13,1985).

IV. General Electric Standard Safety Analysis Report (GESSAR !!) (0 pen)

~~[Fote: R. K. hjor was the Designated Federal Official for this portionofthemeeting.]

J. C. Ebersole Indicated that the GESSAR 11 Subcommittee met on August 7, 1985 to review portions of Supplement 4 of the SER.

MINUTES OF THE 304th ACRS MEETING AUGUST 8-10, 1985 Topics discussed by the Staff included core melt frequency and the containment performance criteria for the review of the new standard plant. The Subcomittee discussed design modifications including

• the Staff's proposal to GE of numerous potential design improvements of which a few have been adopted and are now in the design. Also discussed were the seismic capability, the Ultimate Plant Protection System (UPPS), and hydrogen control. He noted that very few potential design improvements have been adopted because they were subjected to a cost-benefit analysis which considered only the reduction in the dose guideline of $1,000 per person-rem and did not include consideration of onsite averted costs which might have made some improvements more attractive.

R. Bernero, Director, Division of Systems Integration, NRR, mentioned the release by the Comission of the severe accident policy statement and it's emphasis on future plants such as GESSAR II. He mentioned that the focus of the evaluation of severe accident risk would be the discovery and suppression of outliers (severe accident vulnerabilities that stand out from an acceptable background level). He explained that the General Electric (GE)

GESSAR !! review has proceeded in parallel with the develo went and promulgation of the policy statement. He noted the diffsculty in addressing severe accident issues as part of an overall safety judgment that must be made about the acceptability of GESSAR !! for future licensing or referenceability in the FDA. He mentioned the controversy and difficulty in a future plant review of the use of a PRA to display what is known and unknown about plant vulnerabilities.

R. Bernero explained that if one refers to SER Supplement IV it becomes apparent that GE's portrayal of risk and the Staff's portrayal of risk are not the same. GE chooses core melt frequency values and accident release values that are different from those chosen by the Staff. The Staff is roughly one order of magnitude higher in the core melt frequency with regard to source terms.

With regard to the severe accident releases, the Staff is about two orders of magnitude higher than the GE portrayal. The Staff in the source term area is characterizing a higher range of outcomes and chooses to err on the side of caution. Since the Staff chooses an order of magnitude higher core melt frequency, when a cost benefit analysis is donc for a modification, the Staff is more likely to justify the alteration. He noted that the Staff's estimate of uncertainty is somewhat skewed and later information may show that ine Staff s conservatism has been superfluous. The Staff has chosen and GE has acknowledged further additions to the design philosophy and design approach for GESSAR. The UPPS and certain features associated with hydrogen control and de power which the Staff believes significantly enhance the safety of the plant and,

~,are therefore justified, have become part of the FDA. With regard .

to criteria for acceptability, there was a close call in the area of cost-benefit analysis. The Staff chose not to include averted economic loss which would have made it easier to justify a design improvement. The signals from the Commission are that they are not E.

I MINUTES OF THE 304th ACRS MEETING AUGUST 8-10, 1985 pursuing the idea of averted economic loss, although the Staff has recomended it.

W. Kerr noted that the statement was made regarding erring on the conservative side in using PRAs. He stated that he thought PRAs were supposed to be best estimates. He asked if the Staff does an additional conservative PRA once it gets to the cost-benefit stage in the analysis. R. Bernero explained that one tries to make a realistic estimate but one alwsys approaches realism from a pessimistic side. As the uncertainty gets very large, one must become conservative. H. W. Lewis suggested that the approach to realism from the conservative direction is logically flawed and easily can lead to nonconservative results. He suggested that it would be better that if a large uncertainty exists to aim squarely at the middle (rather than approach from the conservative direction) an;t to acknowledge the uncertainty. J. C. Mark pointed out the arbitrariness of the $1,000 per person-rem and its inapplicability to the relay chatter contribution to the seismic risk as well as sabotage risk. R. Bernero suggested that cost-benefit analysis can only display in a consistent fashion what is known in order that judgment can be focused and exercised. W.

Kerr agreed with J. C. Mark that one must be careful when dealing with severe accident issues for new plants since the analysis basis depends rather heavily on PRA and on cost-benefit analysis. F. J.

Remick asked if GESSAR 11 is a modification of an existing FDA or is it a review of a new FDA. R. Bernero indicated that existing FDAs are backward referenced. This GESSAR !! design is substantially modified from the original GESSAR design. It is a substantially different FDA because one has to consider the severe accident policy statement which was for a forward referenced FDA and the CP/ML rule. G. Sherwood,'GE, indicated that GE engineers avoid cost-benefit analysis. They take great pride in their work and essentially do their engineering by relying on good engineering judgment. Cost-benefit analyses are undertaken because GE is forced to do them.

D. Hankins, GE, discussed hydrogen and related matters such as scrubbing efficiency of the suppression pool. She indicated that the hydrogen generation rates that GE sees for the full core melt scenarios (as opposed to degraded core scenarios) var 0.4 to 1.6 pounds mass per second (see Appendix XIX)y .

from The about total in-vessel hydrogen produced varies by sequence from about 1300 to 2300 pounds mass. There is only enough oxygen in the Mark III containment to support the combustion of about 2480 pounds mass of hydrogen equivalent, to about 67 percent of the active clad as opposed to the hydrogen control rule which states one must analyze 100 percent metal-water reaction in the active clad.

D. Hankins indicated that GE assumed that there was about a 26 ,

• percent chance of a global hydrogen detonation which would cause simultaneous failure of the drywell and the wetwell. GE says that the only way to fail the drywell through hydrogen phenomena is by a global hydrogen detonation. The Staff assumed a certain fraction of the time one could have a local detonation that was capable of UTES OF THE 304th ACRS MEETING AUGUST 8-10, 1985 failing the drywell. GE thinks there is insufficient energy to fail the drywell during a local detonation. As a result, the Staff is showing a high seismic risk which can be reduced by a factor of two with the installation of ignitors which would burn the hydrogen in low concentrations and preclude the possibility of detonations.

GE disagrees with that analysis and does not show the same type of risk reduction for ignitors. D. Hankins pointed out that since GE has comitted to implementation of the UPPS design, addition of ignitors with a backup power supply as mandated by the NRC Staff will result in essentially zero risk reduction. The large risk reduction for seismic events that the Staff calculates is related to the assumption of local detonations failing the drywell.

Despite disagreements with the NRC Staff, GE has committed to provide a hydrogen control system that is consistent with the outcome of the Hydrogen Control Owners Group Program and the NRC review of that experimental and analytical program. The NRC is requiring a diverse power supply for the ignitors for the case of station blackout (the dominant core rrelt event). GE finds no technical justification for that backup power supply but will comply.

D. Hankins indicated one of the questions involved with hydrogen phenomena in a SWR concerns the bubbling of steam and hydrogen delivered to the pool up through the pool with the stripping out of steam and the release at the surface of the pool of essentially pure hydrogen. In the presence of an ignition scurce such as an ignitor, the hydrogen can burn in what would be called diffusion flames. The impact of those flames on the probability of pool bypass by either failing a seal or a penetration in the drywell is in question. She indicated that the drywell equipment hatch in the GESSAR design has a five-foot concrete shield plug between the wetwell air space and the seals. GE does not believe there is any potential for the flames impacting those seals. The personnel airlock also has a cement shield plug and electrical penetrations which are five feet long and are potted with a Portland cement mixture. This Portland cement material can tolerate very high temperatures without losing its integrity.

D. Hankins discussed the question whether temperature effects could have any impact on the drywell structure and the molten core in the pedestal region. Could the thermal gradient cause stresses in the drywell structure and cause loss of integrity? She indicated that GE does not have any concern regarding thermal stress but decided to discuss the issue of ablation of the pedestal region which might result in loss of drywell structure and, potentially, containment integrity. She explained that the GESSAR pedestal is a steel concrete composite consisting of two concentric steel rings each one and one-half inches thick. They are connected with steel sheer ties which are concrete-filled. The primary structural support comes from the steel, not from the concrete. GE assumed that the ..

first 1.4 meters of concrete had been ablated with the assumed loss of the inner steel ring. The only support would be provided by the outer steel shell which would then be at a radial distance of 1.8 meters and that steel shell would be at a high temperature. GE

MINUTES OF THE 304th ACRS MEETING AUGUST 8-10, 1985 looked at the loads, the weight of the reactor pressure vessel, the shield wall and other equipment, and the weight of the pedestal

~

itself. It was found that the yield strength of steel at 1100

• degrees F. is about 21 ksi. It was found from tte analysis that the load per unit area of the outer steel shell was about 3.4 ksi.

Since there is such a substantial margin between 3 and 21, it can be assumed that the pedestal can, in fact, carry the loads. GE believes there will be no loss of the pedestal, the drywell, or the containment as a result of this ablation challenge. J. C. Ebersole asked if GE had considered another structural material between the two shells besides concrete, such as a core ladle. D. Hankins explained that GE found that placing a material there that would decrease the penetration rate of the corium would drive heat in the upward direction raising concern about failure of the drywell due to the heat. GE used an absorptive concept but also considered mitigating damage to the basemat. D. W. Moeller asked the NRC Staff how they could justify requiring ignitors if GE contends there is no benefit to tt.em. C. Thomas, NRC, indicated that the difference is the numbers that are assumed or the numbers that result from the calculations. The Staff's numbers de show an incremental benefit for ignitors.

C. Thomas pointed out that the GESSAR design includes containment sprays but not a dedicated spray separate from the residual heat removal system for the purposes of cooling for severe accidents and hydrogen burns. D. Hankins indicated that the string of the

containment sprays is based on an assumed bypass from the drywell

! to the wetwell air space and assumes steam bypass under LOCA conditions. GE does not realistically believe that bypass exists.

But, for licensing and design basis purposes, it is included. The sprays would be used in cases where one needed additional containment heat removal.

T. Pratt, Brookhaven National Laboratory, indicated that tt.e Hydrogen Control Owners Group /NRC interactions related to hydrogen control deal with degraded core events and maintaining containment integrity. Very crucial is the amount of hydrogen produced and the rates of hydrogen production. Issues related to deliberate ignition of hydrogen involve optimum ignition sources, type of power source, limitations on ignition sources, and the effect of standing flames (see Appendix XX). He indicated that when BNL looked at hydrogen generation, there were no large differences between the BNL assessment of f jarogen generation and that by GE.

Both were derived using the MARCH code and gave the usual generation rates. Both analyses looked at the potential for local distributions to occur in the containment building, and discussed l

the characteristics of hydrogen detonation which result in a shock l wave.

j T. pratt indicated that BNL looked at the potential for detonations to occur. There was a potential when the hydrogen concentration exceeded 18 to 20 percent. BNL also looked at the magnitude of the shock loading. The equivalent dynamic shock loading was converted

MINUTES OF THE 304th ACRS MEETING AUGUST 8-10, 1985 into an equivalent static load and that equivalent static load compared to the estimates of the capability of the structures that

the shock wave would see. He mentioned the NRC Staff's )osition on acceptable hydrogen release histories as defined in a ' etter from R. Bernero to Mr. Hobbs, dated June 24, 1985. He indicated that the Hydrogen Control Owners Group test program is to confirm the adequacy of deliberate ignition but will not test for optimum ignition sources.

T. Pratt indicated that BNL is in agreement with GE regarding the effect of standing wetwell hydrogen flames. Calculations reported in Appendix A to NUREG-1037. The Containment Performance Working Group Report, indicate that the seal temperatures will increase but were still below the estimated failure point. He did note that later in a degraded core event, high drywell temperatures during core-concrete interactions may cause the seals to exceed their failure limit and introduce pool bypass at that time.

T. Pratt discussed BNL's review of the effect of a core melt on the reactor vessel's support integrity. He referred to the previous GE presentation on the ablation of the support structure noting that J. Rosenthal of the Staff had his calculations based on a corcrete support structure. The GE analysis indicates that support would be transmitted to the steel, and if the outer steel supports the vessel, there may not be a problem (see Appendix XXI). He stated that if one lost containment integrity for all of the accident sequences that result from this phenomenon early on, shortly after vessel failure the increase in risk (without UPPS installed) would be very modest. If one loses containment integrity, plus the drywell wall, the person-rem estimates would go up by less than a factor of two. This result assumes the very conservative dssumptions that one would lose drywell and containment integrity as a result of slippage of the vessel early on. The calculation as presented by GE indicated this event would not occur early, but would occur much later in the accident sequence. The BNL scenario gives an upward bound to the consequences of the accident.

Venting of the containment building would be designed to address the Class II accident sequences, those accidents that refer to loss of containment heat removal. By attempting to vent, one prevents core damage and the ATWS sequences for the same reason. He indicated a preference for trying to manage an ATWS sequence rather than going to a venting procedure because of the uncertainty associated with the ability to mitigate an ATWS by venting. J. C.

Ebersole expressed concern that the venting concest of the UPPS is just a concept since the system is but an assembly of loose parts that have not yet been integrated and organized into a module.

The meeting moved into closed session in order to discuss

.. proprietary information by GE regarding GESSAR !! Sabotage -

Considerations (see Proprietary Supplement to the Minutes).

V. Seismic Oualification of Equipment in Operating Plants (0 pen)

MINUTES OF THE 304th ACRS MEETING AUGUST 8-10, ICDS

[ Note: A. J. Cappucci was the Designated Federal Official for this portionofthemeeting.]

C. J. Wylie explained that a meeting of the Qualification Program for Safety Related Equipment Subcommittee was held on August 6, 1985 to discuss the status of the NRC proposed resolution of US!

A-46, Seismic Qualification of Equipment in Operating Plants, the Seismic Qualification Utilities Group (SQUG) program for seismic qualification, and the result of a SQUG evaluation of the March 3, 1985 Chilean earthquake. He briefly defined the scope of the task action plan developed for the resolution of US! A-46 noting that it is of limited scope and requires the verification of the seismic adequacy of equipment required to bring the plant to a safe shutdown condition and to maintain the plant in a safe shutdown condition following an SSE. It is limited to active mechanical and electrical components and is further limited by excluding those issues and items which are addressed elsewhere by other US!s, generic letters, and IE bulletins. He indicated that the ACRS was briefed by SQUG in July 1983 regarding their pilot program and approved a letter to the EDO which encouraged continued work along the lines presented by SQUG. During the 289th meeting in May 1984 the ACRS was briefed on the status of US! A-46 and wrote a letter to the Comission endorsing the approach being taken by the Staff based on the SQUG efforts. The NRC Staff has presented a revised resolution package which consists of NUREG-1030, a draft report for comment entitled " Seismic Qualification of Equipment in Operating Nuclear Plants. This document is the supporting technical basis for the resolution of USI A-46. The Staff has prepared the regulatory analysis for A-46 which covers an analysis, a plan, and an implementation procedure. A draft generic letter has been prepared to send to licensees following resolution of public comments. The CRGR has reviewed the package and approved sending NUREG-1030 and the regulatory analysis out for public comment. He indicated that no action is required on the part of the ACRS although it may wish to consider writing a letter at the conclusion of this presentation. C. P. Siess pointed out that A-46 deals only with the qualification of equipnent that is already in operating plants. J. C. Ebersole pointed out that you have to replace this equipment as it wears out. C. P. Siess suggested that the replacement would presumably come under the same criteria as new plants. N. Anderson, NRC, indicated that he thought that the Staff's proposed position with regard to replacement parts is that the utility would have the option of either using the SQUG data base developed for A-46 or qualifying the equipment to current criteria. C. P. Siess indicated that the SQUG approach to the resolution of A-46 makes sense for equipment that is in operating plants for which qualification under existing criteria would be very difficult or almost impossible. He did not see a similar justification for replacement of equipment which could be tested at laboratories renote to the facility. He thought this particularly ,

applied to the case of relays which will undergo a rigorous research program and may be qualified on some other basis. J.

Thomas, SQUG, indicated that the SQUG program is developing specific criteria to verify the ruggedness of replacement MINUTES OF THE 304th ACRS MEETING AUGUST 8-10, 1::S equipment. He suggested that engineering judgment from plant walkdowns can give a fairly good estimate of ruggedness at least as well as that developed from shake table tests. It was noted by all

• that new equipment is out of the scope of A-46.

T. Y. Chang, NRC Task Manager for US! A-46, indicated that US! A-46 was set up to answer the question whether equipment in existing plants can perform its safety function during an earthquake. He pointed out that it is probably not practical to impose the current licensing criteria on all the equipreent in existing plants.

Because the current requirement calls for testing that equipment on a laboratory shake table one may not be able to obtain a piece of equipment similar to the ones in the plant and there is consideration of the downtime once the equipment is taken out of the plant (see Appendix XXII). The objective of US! A-46 is to look for alternative methods in lieu of using the current criteria to assess the seismic adequacy of equipment. D. W. Moeller asked what acceleration forces are used during shake table tests. T. Y.

Chang indicated that this has to be site specific. Normally a number of OBE tests are done on top of the SSE. The amplification of the seismic input through the building has to be considered. H.

Etherington asked if an actual earthquake spectrum is put into the test procedure. T. Y. Chang indicated that for each site those doing the test have to generate the required response spectra.

T. Y. Chang explained that after looking at several potential alternatives the Staff concluded that the only viable and practical way other than using the current criteria and method is to use the seismic experience data approach. The feasibility of this approach was established by two independent studies, one done by Lawrence SQUG. There was Livermore Laboratories some question regardingandthis the other SQUGinitiated pro,1cct by(which looked at earthquakes in California) as to whether enough information could be gathered from California earthquakes and whether there was any similarity between equipment in the non-nuclear plants examined and equipment in nuclear plants.

T. Y. Chang explained that in the SQUG pilot program eight classes of mechanical and electrical equipment were examined. This equipment was chosen in be representative of equipment in the examined plants and also to constitute a large percentage of all the equipment in the plant. The eight types of equipment chosen were motor control centers, low voltage switchgear, metal clad switchgear, unit substation transformers, motor operated and air-operated valves, horizontal pumps and motors, and vertical pumps and motors. J. Thomas indicated that the SQUG project picked the equipment that was most utilized in the plant systems and might be most susceptible to seismic type forces. T. Y. Chang indicated

,,that most of the experience data were taken during the San Fernando .

and Imperial Valley earthquakes ard in the 1983 Coalinga earthquake. Data were also taken in 1984 in the Morgan Hill and recent Chilean earthquakes in order to expand the data base.

MINUTES OF THE 304th ACRS MEETING AUGUST 8-10, 1985 T. Y. Chang indicated that a Senior Seismic Review Advisory Panel (SSRAP) was formed in June 1983, jointly selected by SQUG and NRC, to provide expert opinion and advice on how to use the experience data especially in the Staff's review. $5 RAP published in its January 1985 report bounding spectra, a conservative estimation from those earthquakes akin to test spectra for use in seismic qualification. 55 RAP concluded from the SQUG pilot program that equipment installed in nuclear power plants is generally similar to ,

and at least as rugged as that installed at conventional power plants. With some reservations this equipment when properly anchored has an inherent seismic rugiledness and has a demonstrated capability to withstand substant<al seismic motion without structural damage. W. Kerr asked for an interpretation of these conclusions. T. Y. Chang indicated that the anchorage has been identified as an area that can cause some problems. If the anchorage is not adequate the equipment is likely to shif t during i earthquakes or even overturn.

T. Y. Chang indicated that $$ RAP concluded that for the eight equipment classes studied in the $QUG pilot program, functional' ty af ter the strong shaking by an earthquake has ended has been demonstrated but the absence of relay chatter during such strong shaking has not been demonstrated. This means that there is evidence to show that devices like relays always function after an earthquake when they are called upon to function. But information is not available to demonstrate that the relays would function during an earthquake. D. W. Moeller asked whether relays have been shake table tested and whether they chatter during the shaking. T.

Y. Chang indicated that during tests of relays for new p' ants, chattering has been observed. D. W. Moeller suggested that one could expect that there would be chattering in the relays in i operating plants. T. Y. Chang indicated that it depended upon the function of the relay in the system.

T. Y. Chang concluded that there are at least two areas of concern regarding the seismic adequacy of these eight classes of equipments anchorage and the functionality of the equipment and the devices such as relays. lie explained that equipment other than the eight  ;

classes singled out in the SQUG program is being examined in more detail and the data presented in a systematic way. Nevertheless.

- the Staff is confident of the ruggedness of such equipment and has  ;

decided not to require any additional collection of seismic  !

eAperience data for the other Classes of equipment. The Staff will l require the utility to document their basis for seismic adequacy l for the remaining classes of equipment either through verification that the equipment exists in the data base plant or prove that a i similar piece of equipment is in entstence in those plants surveyed in the California earthquakes. The alternative to this method is comparing with the test data currently being collected by EPRI and attempting to produce a conservative envelope for the test spectra ,

that they call GEPS, Generic Equipment Ruggedness Spectra. These spectra would be similar to the bounding spectra that $5 RAP is proposing. T. Y. Chang indicated that a third cuncern involves 24 -

o .

MINUTES OF THE 304th ACRS MEETING AUGUST 8-10, 1985 identifying outliers, equipment in the eight classes identified by SQUG which are outside the data base.

T. Y. Chang classified the proposed Staff resolution of A-46 by the two groups of plants involved. He indicated that there is no requirement from an A-46 point of view for new plant licenses.

Operating plants will have to develop equipment lists, perfonn walk through inspections, verify anchorages and the functionality of equipment (relays), and identify and address deficiencies for outliers. W. Kerr noted that the anchorage analysis is generic and is not an analysis for each piece of equipment separately. N.

Anderson agreed. T. Y. Chang noted the difference between deficiencies which are obvious inadequacies from a walk through inspection and outliers which are equipment that is not included in the dats base. He indicated that the Staff is relying on additional studies, such as testing, in the case of outliers.

T. Y. Chang discussed the scope of the A 46 seismic adequacy review. He explained that the review does not consider an earthquake beyond the S!! and assumes that an SSE earthquake does not cause a LCCA. J. C. Ebersole asked if LOCA also pertains to the secondary systems of feedwater and main steam. T. Y. Chang indicated that the LOCA mentioned here is only with regard to the primary coolant. J. C. Ebersole asked if this assumes that an SSE can bring about a failure on the secondary side. T. Y. Chang indicated that the Staff does not believe piping will break regardless of what kind because all of it was reevaluated during a 1979 IE Bulletin review. J. '. Ebersole suggested that secondary and lower grades of piping have not been analyzed as intensively as those in a primary icop. T. Y. Chang indicated that piping up to the main steam and feedwater pipes was considered during the 1979 review. N. Anderson indicated that the Staff's assumption with regard to pipe break also applies to secondary and feedwater systems. T. Y. Chang indicated that the Staff postulates that a LOCA does not cccur simultaneously with or during an SSE. The Staff believes that the possibility is too low. The Staff also assumes that offsite power will be lost during or following an earthquake. J. C. Cbersole suggested that the assumption should be that offsite power will be lost and normal heat sinks will also be lost. N. Anderson agreed that, to be rigorous, such an approach would be no:essary.

T. Y. Chang indicated that equipment included under A 46, active electrical and rechanical components including instrumentation and controls, is needed to achieve and maintain hot shutdown for a minimum of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. J. C. Ebersole suggested that the Staff ought to look at system interfaces to see whether nonqualified equipment could fall and damage qualified equipment. This could be done during the walk through process. T. Y. Chang indicated that that falls under the scope of US! A 17, Systems Interaction. J. C. ,

Ebersole suggested that there may be control systems which interact '

under special conditions of earthquakes and these would be the ones that ralfunction. N. Anderson indicated that this is the subject of USl A 47, tafety implications and Control Systems and will not 25 . .

MINUTES OF THE 304th ACRS MEETING AUGUST 8-10, 1985 be examined under this generic issue. He did note that the seismic concerns in A-47 should be covered under the A-46 data base because the equipment and instruments concerned with A-47 are represented

• at least by type in A-46.

The Committee briefly discussed the review of relays. T. Y. Chang indicated that evidence from a review of the Chilean earthquake pointed to the fact that general purpose relays functioned without problems. Some problems were seen, however, with protection relays, and big relays for power systems.

J. Thomas, Duke Power Company, indicated that the conclusions of the SQUG experience data is that seismic resistance of standard power plant equipment, when properly anchored, has been verified and explicit seismic qualification of the equipment in the operating plants to current standards is not justified. Based upon the depth of the SQUG survey, the seismic qualification issue does not appear to be a significant concern. He pointed out that one of the purposes of the SQUG survey was to totally deft the equipment that would be required to perform in a seismic event. In pursuing this issue, SQUG began work on the definition of generic equipment required to achieve safety shutdown to envelope all plants. SQUG has enlisted help from EPRI to gather additional test data to develop a rationale to assure the seismic ruggedness of equipment beyond the eight classes defined in the pilot program. He explained that the surveys really were to look for structural type damage or uamage to the equipment to address operability during strongmotion(seeAppendixXXI!!).

J. Thomas described the SQUG follow up of the Chilean earthquake of 7.8 magnitude in 1985. He explained that the SQUG group considered that one of the main areas of interest was the operability issue during strong motion with an investigation of relay chatter in particular (see Appendix XXIV). He discussed earthquake damage to large oil storage tanks at a Chilean oil refinery. Tanks that sat on the ground ruptured. One of the failures developed is called an elephants foot where the tank at its base just moves over and buckles at the base. There is separation of the wall and the bottom plate and the contents are lost very quickly. In a number >

of cases, implosion occurred at the top of the tank. It was noted i that about 10 percent of the tanks failed, rustured and lost their contents but a much larger percentage had distress bucklino, J.

Thomas mentioned a diesel control system at a hospital in San Antonio, Chile that added substantially to the data base regarding the functionality of various types of general purpose relays. Also mentioned were data gathered at a copper refinery regarding motor-operated valves at high levels. He noted that no problems were found as far as the functionality of the controls and equipment at

,,the copper refinery. .

Another example of the durability of general purpose relays was noted in a burner control system manufactured by Forney Engineering in Dallas, Texas which was installed in a fossil generating plant.

The only damage to relays was found at a hydro substation where l

MINUTES OF THE 304th ACRS MEETING AUGUST 8-10, 1985 i

line protective relays for transmission lines were broken. He

, attributed the breakage to inertial forces rather than interaction

. and also noted that they were an antique type relay. J. C. -

Ebersole asked if this was the only kind of relay failure seen. J.

Thomas indicated that at the refinery one of the protective relays caused an emergency power generator to trip during a start-up operation. The operator innediately reset that particular relay and the turbine generator came on-line. That was a protective .

relay and amounted to the only malfunction of such a relay. In answer to an inquiry by C. J. Wylie, he indicated that these were older type relays, not modern solid state ones. <

J. Thomas mentioned EPRI programs that have been initiated to support SQUG work to screen an essential list of equipment needed to assure a hot shutdown during a seismic event. He indicated that he thought that total classes of generic equipment required in this survey would be limited to about 21 classes. Battery charges and inverters were lumped into one of those classes. J. C. Ebersole asked regarding station type batteries what would be the most fragile installation one could find in any nuclear plant. The response was that these would be inert stationary static devices.

J. Thomas indicated that 500G experience has been that no case has been found where failures of such batteries has occurred, even though anchorage problems were found. He noted that one of the final objectives of the future SQUG work is to attempt to limit the scope of the relay chatter issue, and to focus on what systems could actually be affected by relay chatter. He indicated the SQUG's generic implementation plan would involve the development of plant welkdown procedures that could be used generically as well as  !

the development of plans on how to audit the results of such L walkdowns.

VI. NRC Maintenance and Survet11ance Program Plan (0 pen)

(Note: H. Alderman was the Designated Federal Official for this l portionofthemeeting.]

G. A. Reed stated that there appears to have been a major refocusing of the Staff's maintenance program over the past 18 months. They have selected six technica' issues or projects which are to be the major regulatory activities in Phase 1 of the program. He noted that, < n addition to the six technical issues, there are two issues that the Committee might want to discuss.

These are the issue of a U.S./ Japanese comparison of maintenance experience and whether the Japanese experience is closely relatable 4

! to the United States nuclear power plant sitt.ation. The other  ;

issue, which is also connected to the Japanese experience, is the i natural ability selection issue which has previous been taken up in

..a combined meeting of the ACRS Human Factors and the Maintenance - i Subcommittees. He noted that the last meeting on the Maintenance Program Plan was held on June 18, 1985.  ;

H. R. Bocher, NRR, discussed development of the maintenance program which began on November 19,1983(seeAppendixXXV). He mentioned  !

c

MINUTES OF THE 304th ACRS MEETING AUGUST 8-10, 1985 l

a pilot study during the time frame, November 1983 to December 1984, which developed a list of maintenance indicators which was complemented by an industry list of some 10 maintenance indicators.

He noted that the NRC staff was unable to share the data used by the industry-formed NUMARC organization to make an independent evaluation of the industry indicators.

H. R. Booher identified the NRC's specific objectives in the Maintenance and Surveillance Program Plan (MSPP). He noted that the NUMARC organization is in accord with these objectives. The primary focus would be on maintenance effectiveness. H. R. Bocher indicated that the NRC Staff intends to carry out a systematic survey of surveillance and test activities. He explained that in the U.S. the use of the surveillance and test programs is to assure availability, whereas the Japanese use surveillances and tests for preventive maintenance. The NRC wants to go beyond just safety-related equipment in this program and to include balance of plant equipment. He noted that the Japanese have considerable coverage with their preventive maintenance programs not only of safety-related equipment but of balance of plant equipment.

H. R. Bocher indicated that the Davis-Besse incident suggested potential problems with maintenance performance. He noted that the number of reports of abnormal maintenance events has been increasing cver the past 10 years and that trend has persisted. One of the main issues from these abnormal events reports is the challenge to safety systems in terms of the trip rates. U.S. trip rates compare very unfavorably with those in Japan. There are cases where the trip rate is 10 to 20 times greater in some U.S.

plants than in comparable plants in Japan. Surprisingly, some plants in the U.S. are performing just as well as the Japanese plants. The problem may be outlier U.S. plants with difficulties related to maintenance practices. Another problem for examination in the maintenance program is occupational exposures. Maintenance personnel account for 75 percent of such exposures and if there is a way to reduce this exposure systematically, it sbculd be ccnsidered in this program.

H. Bocher indicated that the imediate focus during Phase I of the program plan would be on six technical issues:

Human Error in Maintenance Indicators of Maintenance Performance Usefulness of Counteracting Aging and Service Wear Effects Management Organizational Impacts

• Lack of Adequate Criteria and Standards in the Area of .

Maintenance

• Maintenance / Operations Interface MINUTES OF THE 304th ACRS MEETING AUGUST 8-10, 1985 He indicated that the Staff is taking a phased approach. Phase I is a survey of current maintenance practices and the evaluation of their effectiveness (report due April 1986). Phase II will be an identification of distinct maintenance problems and the determination of their impact on pisnt safety and worker doses. H.

W. Lewis noted there have been a number of fairly egregious maintenance-induced accidents at nuclear power plants in the last year or so. He asked if the Staff has made any effort to look at case histories regarding the attitude of management toward the findings of human error involved in those incidents. H. R. Booher indicated that the Staff has been in a reactive mode in looking at such case histories in the past and has found it very difficult to measure the quality of management's attitude.

H. R. Booher indicated that, of several Phase I projects, the survey of current practices and maintenance performance indicators are perhaps the most important. A site survey of current maintenance practices being conducted by Pacific Northwest Laboratory involves the Kewaunee, Millstone, Brunswick, Turkey Point, Arkansas 1 and Rancho Seco plants. G. A. Reed noticed that the survey was heavily skewed toward PWRs rather than BWRs (4 to 2). G. C. Cwalina, NRC, ex)lained that the Staff wanted to get a wide range of plants and t1erefore selected two General Electric plants, two Westinghouse plants,1 B&W plant and 1 CE plant in order to include all of the major vendors. G. A. Reed thought that was a good idea because it would show the differences in radiation exposure to maintenance workers which vary considerably from BWRs to PWRs. J. C. Ebersole asked if the Staff was looking at maintenance training programs. He noted that there is an extensive maintenance training program at Georgia Power and he wondered if this is a comon practice. H. R. Booher indicated that maintenance training programs are becoming more important because of INPO accreditation. J. C. Ebersole asked if the Staff thought there was adequate incentive in terms of compensation and career advancement for maintenance workers at a nuclear plant. H. R. Bocher indicted that the lack of incentive has been one of the largest complaints he has heard when attending maintenance superintendent ucrkshops at INPO. W. Kerr asked if the Staff noted anything unusual during its trip to Japan to visit the sister plant to the U.S. Kewaunee Plant.

D. Brinkman, I&E, indicated that he was surprised at the level of interest and motivation of maintenance workers and also quite s'urprised at the small staff at the plant. G. A. Reed pointed out that the Kewaunee Plant is a plant that has used natural ability s' election testing for many years.

H. R. Bocher indicated that the objective of the project on maintenance performance indicators was to derive quantitative nfeasurements to complement the Staff's subjective expert opinions.

The Comittee briefly discussed the NRC's lack of accessibility to the data on the 10 NUMARC indicators. It was collected by INPO and ..

is considered proprietary. H. R. Bocher indicated that the Staff should have an independent look at the industry data to avoid duplication in the collection of such data.

. 29 -

MINUTES:0F THE.304th ACRS MEETING AUGUST 8-10. 1985 H. R. -Booher mentioned additional projects in Phase 1 involving participation in standards groups, ~ program integration of the NRC industry work, and the analysis of U.S./ Japanese maintenance programs. He indicated that the Staff has published NUREG/CR-3883 and 3883P, which contain results of the comparison the Staff made

. of ' Japanese and. U.S. maintenance practice. Work has stopped in this area, .however, since the U.S. nuclear industry has sent its own team over to look at Japanese practices. D. A. Ward ask if the Japanese are adopting any of the U.S. attitudes toward maintenance.

H. R. Booher indicated they are considering doing less preventive maintenance because of the resultant long outages.

I H. R. Booher indicated .that the Staff has . undertaken additional

. Phase 1 projects in maintenance personnel qualifications pairing comparable plants in the U.S. and Japan and a feasibility study of

! the human factors' associated with the INPO round-robins on

ultrasonic testing for pipe cracking. An investigation of human f

errors (Genericis underway Issue 102)on events

. He involving explained wrong that the unit goalorinwrong this train last i project is to look at short term solutions with low costs. This is 4

being done _ by starting with LERs and trying to analyze the root

causes sometimes by interview with the person who committed the error. Preliminary findings on this last project are not surprising. Problem areas identified were the area of -labeling, procedures, and fatigue by involved personnel. ,,

H. R. Bocher recomended that the ACRS invite the nuclear industry or INPO to brief the Comittee on their efforts on maintenance practices. G. A. Reed indicated that he was pleased with the Staff's progress in the last 18 months and did not feel it 4 necessary for the' ACRS to request an INP0 briefing. H. R. Booher

. thought that the INPO briefing would be a way in which the ACRS could supply additional pressure on INPO to release supporting data on its maintenance indicators. G. A. Reed indicated that the Subcomittee on Maintenance Practices would like to be kept i informed before any final Staff conclusions on the entire i maintenance issue are released for public coment. He requested that the Staff provide draft proposed releases and documents as they become available.

VII. NRC Long Range Plan (0 pen) 4

[ Note: R. K. Major was the Designated Federal Official for this portion of the meeting.]

M. W. Carbon indicated that the Office of Policy Evaluation has developed a five year plan outline based upon guidance from the Comissioners and the NRC Staff. It is considered proprietary and predecisional and OPE does not wish to have the outline discussed by the ACRS in open session. There was a brief discussion of ..

Subcommittee coments which M. W. Carbon intended to pass on to OPE. The Comittee endorsed the comments and he indicated his intent to pass the coments along orally.

MINUTES OF THE 304th ACRS MEETING AUGUST 8-10, 1985 M. W. Carbon discussed his plans to turn off the ACRS initiative on the long range plan and henceforth confine the Subcomittee's business to review of OPE written plans in open session. Input to the long range plan would be by comenting on OPE's outline / plans.

D. A. Ward indicated that he would like to continue to develop ACRS positions on various important subjects and pass them on to OPE.

He thought that OPE needed a distillation of the data gathering effort conducted by the ACRS Subcomittee. H. W. Lewis indicated that he thought the Comittee was told to limit ACRS effort to technical input, not organizational issues. Other members including D. A. Ward thought that the Comittee did not need to limit itself to technical issues per se. M. W. Carbon noted that the Subcomittee members unanimously Wreed that the ACRS was not competent to write an entire long range plan.

H. W. Lewis thought that compartmentalism in the NRC was a real problem, but that the Comittee could not address this issue because it was of an organizational nature. D. A. Ward and J. C.

Mark both thought that the Committee could say anything it wanted to about deficiencies. D. A. Ward expressed his belief that the ACRS ought to help to drive the process (which included the OPE outline) in the right direction. C. J. Wylie as well as other members pointed to the adversarial situation between NRC and the nuclear industry as another big problem that needed to be addressed. The ACRS Executive Director suggested that the Comittee could work up white papers on selected issues and feed them into the prccess. W. Kerr suggested that the Subcomittee members pick out a few (four) important issues for full Comittee consideration. D. A. Ward thought that the full Comittee ought to look at a list of issues and select a manageable number on which the Subcomittee could develop position papers.

C. P. Siess requested that he be replaced as member of the Subcoc=ittee on the Long Range Plan for the NRC. M. W. Carbon indicated that he would be unable to work on issue papers until the October or November meeting. D. A. Ward thought it would be all right to defer action on this matter since the issue will take a long time to develop anyway.

VIII. Management and Disposal of Radioactive Wastes (0 pen)

~

[ Note: 0. S. Merrill was the Designated Federal Official for this portion of the meeting.]

D. W. Moeller indicated that a combined meeting of the ACRS Subccmittees on Waste Management and Procedures and Administration was held on July 30, 1985 to review, define, and make reccmendations to the full Comittee regarding the ACRS role in the civilian high level radioactive waste management program. He indicated that the NRC Staff made a presentation based on their May 31, 1985 paper on this subject, SECY-85-197. This presentation dealt with the necessary technical expertise for overseeing the high level waste repository program, the present ACRS structure for waste management oversight, and a two-level review process as the

,' MINUTES OF THE 304th'ACRS MEETING AUGUST 8-10, 1985' proposed solution to the oversight problem. He mentioned that an outstanding presentation had been made to the Subcomittees by J.

Kotra, ACRS Fellow, which dealt with the goals of the Nuclear Waste Policy Act of 1982 and issues- relevant to the implementation of- .

this Act that are potentially subject to ACRS review. Additional comments on the subject were provided in a letter to D. W. Moeller from Dr. Martin Steindler, ACRS consultant, and in memoranda to D.

W. Moeller_ from J. S. Parry, ACRS Fellow, and J. C. Mark, ACRS Member.

D. W. Moeller indicated that M. Bell and H. Miller from the NRC Division of Waste Management made an NRC Staff presentation and were present during the ensuing discussion. The Staff believes it would be desirable for ACRS members to be knowledgeable on high

' level waste, particularly in the earth sciences area. The Staff is concerned about the direction things are going; they see the ACRS moving in and out of the process and indicated that continuing, uninterrupted oversight by ACRS is necessary for the ACRS to be effective in this role. D. A. Ward . suggested that the ACRS restrict itself to providing oversight and not peer review. One NRC Staff comment was that. they recognized that the ACRS Waste Management Subcomittee had added consultants to enhance expertise in the technical disciplines needed for licensing a high level waste repository. It was noted that the NRC Staff has 60 full time individuals in the waste management area plus consultants. There are two to ~ three professionals per technical discipline." C. P.

Siess requested that the Staff be requested to provide the ACRS-with a roster of their . staff and consultants, listing their disciplines and credentials.

D. W. Moeller discussed potential sources of technical information for the licensing effort. He noted that during the Subcomittee meeting J. C. Mark suggested that data on the migration of fission products might be available from underground nuclear test data as well as the Rulison Gas Stimulation Project conducted in the State of Mississippi. D. W. Moeller thought -that the National Research Council might be a source of information.

D. W. Moeller indicated that it was brought out during the Subcommittee meeting that the high level waste task is not appreciably different from other tasks undertaken by the ACRS. The major handicap the ACRS has in functioning effectively in an oversight role in the high-level waste management program is lack of resources -- both manpower and funding. The feasibility of a new Comittee was briefly discussed. It was suggested that, in addition to high-level waste matters, it could also handle low level waste, mill tailings, transportation of radioactive materials and related NRC research. D. W. Moeller suggested the need for a meeting with the Commissioners and the Staff te define specific ,

tasks the ACRS should undertake. He pointed out that there are many questions to be addressed regarding a Multiple Retrieveable Storage facility and the retrievability of wastes from a geologic MINUTES OF THE 304th ACRS MEETING AUGUST 8-10, 1985 repository. There are also many other issues relevant to the DOE / EPA high-level waste activities that need to be addressed.

IX. Class Nine Accidents Subcommittee Meeting (0 pen)

[ Note: D. Houston was the Designated Federal Official for this portion of the meeting.]

W. Kerr explained that the meeting of the ACRS Subcomittee on Class 9 Accidents on August 1, 1985 continued the discussion of the NRC Report NUREG-0956 " Reassessment of the Technical Bases for Estimating Source Terms" which is scheduled to be issued for coment on August 8, 1985. The NUREG discusses the use of a Battelle suite of codes described in BMI-2104 to calculate the source term associated with selected accident sequences in five referenced plants. Results of a more complete calculation for the Surry plant are included in the report as is a comparison of the results that one would get using the WASH 1400 approach coupled with that for the Battelle suite of codes. Accident analysis results using these codes along with new containment performance concepts indicated a reduction in risk of about an order of magnitude when compared to previous WASH 1400 results. The containment was assumed to be stronger and the breach of containment to occur at a higher pressure. He explained that the Subccmittee asked that RES members emphasize how and for what purpose such calculational techniques will be used and how long they will be appropriate.

W. Kerr indicated that the cornerstone of the BMI-210A approach is the MARCH code and the MARCH code is only a simplified view of a badly damaged core. He pointed out that the Battelle suite of codes has several shortcomings which include its inability to model containment performance mechanistically during a severe accident.

The MARCH code is unable to model natural circulation. Another major deficiency in the BMI-2104 approach is the omission of the risk produced by external initiators. The Staff claims that they do not have the time to include external initiators because of scheduling constraints.

W. Kerr indicated that the Subcomittee heard presentations on August 2, 1985 from two contractors being employed by IDCOR on methodology development (one concentrating on PWRs and one on BWRs) regarding the IDCOR method for the systematic evaluation of operating nuclear plants. IDCOR does not make use of the BMI-2104 suite of codes or NUREG-0956. 10COR contracted with Westinghouse regarding methodology development for PWRs and with Delian for BWRs. He indicated that IDCOR/NRC interactions have led to disagreements on several issues, disagreements which should be resolved within several months. NRR discussed its plans and schedule for an analysis of four reference plants. -

W. Kerr indicated that there is need for another Subcomittee meeting to decide on the adequacy of the new approach. It is difficult because the interconnections with the suite of codes. He MINUTES OF THE 304th ACRS MEETING AUGUST 8-10, 1@85 noted that the Staff plans to continue with the risk re-baselining as outlined in NUREG-0956 regarding the recalculation of risk for the five referenced plants. The Staff plans to extrapolate insights to other operating plants. He cautioned that the study would be deficient unless external initiators were considered since external initiators have proved to be important contributors to risk from existing PRAs. He stated that the ACRS needs to state early on whether it believes external initiators ought to be included at the risk of re-baselining effort.

X. INP0 Radiation Protection Program (0 pen)

[ Note: 0. S. Merrill was the Designated Federal Official for this portion of the meeting.]

D. W. Moeller indicated that, under a memorandum of understanding with INPO, NRC has been working with INP0 in developing and implementing a self-regulation program. INP0 was to develop a program to assist commercial nuclear power plant licensees in the implementati,on of improved radiation protection programs, and the NRC was to evaluate the effectiveness of the INP0 program. The NRC evaluation will be completed in September 1985.

The Reactor Radiological Effect Subconsnittee during its meeting on July 31, 1985 reviewed both the INPO Radiation Program and the NRC evaluation of it. The Subcommittee was very impressed sith the depth and thoroughness of the INPO program ar.d the professionalism of the INPO approach. D. W. Moeller explained that INP0 had developed the following indicators of performance.

Collective dose Average dose per worker Square feet of contaminated area Volume of low level waste generated

• Number of workers receiving inhalation (internal) doses Number of workers receiving greater than 5 rem per year

• Number of workers with a lifetime dose greater than 100 rem Quality of personnel monitoring Man-rems per hour worked

• Involvement of management in the plant radiation protection program ..

D. W. Moeller noted that the third indicator, " square feet of contaminated area," was somewhat in conflict with the following indicator, " volume of low level waste generated" because INPO in 1

.7.: .

MINUTES OF THE 304th ACRS MEETING AUGUST 8-10,~ 1985 l l

1 effect was encouraging plants to clean up and decrease the square l footage of contaminated area. This approach - would' necessarily involve an increase in the.. volumes of low level waste generated, over the . short . term, but - will result in less radioactive waste production, once' an area has been cleaned .up. He ' explained that INP0 has done a good job tracking radiological incidents. Their review has identified six radiological protection errors that were  ;

precursors comon to most exposure incidents. '

' Inaccurate or incomplete surveys Inadequate or nonexistant radiological work permits Insufficiently trained radiological protection technicians Workers not complying with rules Supervisors not feeling accountable Station management not involved in radiation protection  :

He' pointed out that increased station management involvement in radiation protection invariably reduces the occurrence of ^ the preceding five errors and, therefore, of exposure incidents as' l well. He also stated that INPO has used extreme care to avoid bias in their data. He indicated that INP0's goal for next year is a 10 percent reduction in plant collective doses with fewer workers involved.

XI. Scram Circuit Breaker Reliability (0 pen)

[ Note: P. A. Boehnert was the Designated Federal Official for this portion of the meeting.] q W. Kerr explained that the Subcommittee on Scram System Reliability met on August 10, 1985 to discuss the results of its July 17, 1985 meeting. At the July meeting representatives of the NRC/NRR and Westinghcuse discussed PWR scram breaker reliability. He indicated i that Westinghouse asserted that there has never been a failure of their DB or DS breakers to function on demand. The failures have been confined to the undervoltage trip attachment on the breakers. <

Westinghouse concluded that breaker reliability rates are higher than initially assumed and failure rates are consistent with values used in PRAs. No failures have been reported in shunt trip devices or reactor trip breakers excluding the undervoltage trip attachments. Westinghouse indicated that they had performed life  ;

cycle confimation tests on the undervoltage trip attachments for the 08-50 breakers' and arbitrarily chosen the cycle life as half the test numbers (16 years). This number was deemed a conservative value to take account of potential maintenance problems.

i W. Kerr explained that the staff agreed that the trip breakers

! performed as expected, especially with the shunt attachment. AE00, in particular, hdd no quarrel with the Westinghouse co 'clusions.

_. _ _. _ . _ _ _ , . _ _ _ . _ _ . _ _ _ _ _ _ _ _ _ ,s

MINUTES OF THE 304th ACRS MEETING AUGUST 8-10, 1985 He asked for Comittee guidance regarding future activity on this topic. He suggested looking at BWR scram breakers and then doing an extensive study of scram systems contribution to risk for BWRs.

W. Kerr mentioned IDCOR claims about the low contribution from breaker failure rates to the core melt probability for PWRs and BWRs. Regarding a recent reactor protection system incident at the Sequoyah nuclear plant, Westinghouse is proposing to install a fusible link to protect the transistor which was rendered inoperable during a maintenance operation at that plant.

W. Kerr and C. J. Wylie discussed changes in the manufacturing and certification process which should increase the reliability of Westinghouse scram breakers. It was noted that the Subcomittee would continue to examine the reliabi'ity of scram systems for both BWR and PWR systems.

XII. Meeting of the ECCS Subcommittee (0 pen)

[ Note: P. A. Boehnert was the Designated Federal Official for this portionofthemeeting.]

D. A. Ward explained that the ECCS Subcommittee met on July 31, 1985 to review the proposed revision to 10 CFR 50.46 and Appendix K, to review the report of the NRC investigation team regarding the June 9, 1985 loss of all feedwater event at the Davis-Besse plant, and to discuss resolution of the issue of reactor coolant pump trip given a small break LOCA. He indicated that representatives of B&W, CE and Westinghouse made presentations regarding approaches on whether the reactor coolant pumps should be tripped imediately after the onset of a small break LOCA incident. Westinghouse has performed analyses to justify manual reactor coolant pump trip.

The principal requirements for the trip criteria are 1) trip reactor coolants pumps when needed for small break LOCAs and 2) maintain reactor coolant pumps operational when beneficial to plant recovery. He indicated that the Staff has approved the Westinghouse approach noting that the operator's discretion will be preserved because of the uncertainty regarding the indication of the onset of a small break LOCA. The B&W and CE approaches were also discussed. He thought that the NRC Staff's review of this issue has been satisfactory, that it was not necessary for additional full Comittee action, and that the Staff ought to be allowed to proceed to complete their review of this issue.

D. A. Ward indicated that members of the Subcomittee were briefed by the NRC investigative team on their examination of the June 9, 1985 loss of all feedwater event at the Davis-Besse plant. P. G.

Shewmon asked what administrative actions may result from this event. D. A. Ward stated that the ED0 plans to review the investigative team report, NUREG-1154, sumarize actions fo'r ..

follow-on of the Davis-Besse event, and assign these actions to given NRC offices. He noted that the main cause of the event appears to be the Licensee's lack of attention to detail in the care of plant equipment. He noted the suggestion on the part of MINUTES OF THE 304th ACRS MEETING AUGUST 8-10, 1985 )

some to fault the plant itself because of certain characteristics of B&W plants and certain B&W plant equipment (such as the once-through stearn generators). The NRC Staff contended that the plant will operate well if equipment has been properly maintained.

C. P. Siess pointed out the vulnerability of the plant to the failure of its turbine-driven feedwater pumps. C. J. Wylie noted that Davis-Besse was about to install an electrically-driven pump just before the incident occurred. This might have mitigated or forestalled the problem. It was noted that the safety parameter display system (SPDS) was down for repairs but would have been of only marginal benefit during the incident. He noted the lack of confidence in the SPDS because of the considerable downtime it has experienced. D. A. Ward noted that the SPDS was added to improve safety but was not required to be a safety grade piece of equipment. He indicated that the investigative team uncovered human factors oroblems which probably contributed to the incident, such as the poor arrangement of buttons in the control rocm and the philosophical problem with manual initiation of automatic functions.

XIII. Executive Sessions (0 pen)

[ Note: R. F. Fraley was the Designated Federal Official for this portionofthemeeting.]

A. Subcomittee Assignments -

1. Federal Training Academy for Nuclear Power Plant Operators The Comittee assigned the Subcomittee on Human Factors to review Senator Moynihan's proposal to establ!sh a federally (NRC) sponsored academy for training nuclear power plant operators and report back to the full Comittee regarding a proposed Comittee position regarding this matter.

2. FTOL Conversion for Millstone Nuclear Station, Unit 1 R. Hernon, NRR, indicated that the NRC Staff will be ready to discuss the FTOL conversion for Millstone Unit 1 at the 306th ACRS meeting in October 1985. It was noted that a four volume plant-specific PRA is available for this plant and a combined meeting of the Millstone and Reliability and Probabilistic Assessment Subcommittees is indicated.

3. Enforcement Policy on Vendors - SECY-85-256 G. A. Reed has proposed that the ACRS review and coment on SECY-85-256, particularly the position that nuclear power plant owners rather than the vendors should be held responsible for equipment quality. Review of this issue was _,

assigned to the Subcomitee on Regulatory Policies and Practices.

4 Dissolution of the SEP Subcomittee MINUTES OF THE 304th ACRS MEETING AUGUST 8-10, 1985 Now that the Comittee's review of Phase II of the SEP as it has been applied to the San Onofre Nuclear Generating Station, Unit 1 is complete, all of the 11 SEP plants have been before the ACRS. The FTOL reviews for some of the SEP plants which currently have provisional operating licenses is to be turned back to the project subcomittees. The project subcomittees will complete resolution of outstanding ISAP issues and determine whether any action by the full Comittee is necessary.

5. Scram Circuit Breaker Reliability The ACRS Subcomittee on Scram System Reliability reported on their meeting on July 17, 1985 to review the reliability of Westinghouse scram breakers. W. Kerr and C. J. Wylie discussed changes in the manufacturing and certification process which will increase the reliability of these breakers.

It was noted that the Subcomittee would continue to examine the reliability of the scram systems for both BWR and PWR plants.

B. Reports, Letters, and Memoranda ACRS Report on the Vogtle Electric Generating Plant, Units 1 and 2

1. The Comittee prepared a report to the Comissioners 'of its review of the application of the Georgia Power Company (the Applicant), acting on behalf of itself and as agent for the Municipal Electric Authority of Georgia, the Oglethorpe Power Corporation, and the City of Dalton, Georgia, for licenses to operate the Vogtle Electric Generating Plant, Units 1 and 2.

ACRS Report on the Systematic Evaluation Program Review of the San Onofre Nuclear Generating Station, Unit 1

2. The Comittee prepared a report to the Comissioners of its review of the results of Phase II of the Systematic Evaluation Program as it has been applied to the San Onofre Nuclear Generating Station, Unit 1. The ACRS will review the Full-Tenn Operating License for the San Onofre Nuclear Generating Station, Unit I when the NRC Staff has completed its actions on the remaining SEP topics and the USI and TMI Action Plan items.

ACRS Coments on the Status of USI A-46 (Seismic Qualification of Equipment in Operating Plants)

3. The Comittee prepared a report to the Comissioners of its review of the status of the proposed resolution of USI A-46 ..

(Seismic Qualification of Equipment in Operating Plants). The ACRS plans to follow the status of this issue and intends to review and comment on the final proposed resolution after the public comment phase.

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MINUTES OF THE 304th ACRS MEETlNG AUGUST 8-10, 1985 ACRS Advisory Role on the NRC High-Level Radioactive Waste Program

4. The Comittee prepared a report to the Comissioners regarding the proposed ACRS role in advising the NRC on matters pertaining to the civilian high-level radioactive waste (HLW) management program.

ACRS Comments on the NRC Maintenance and Surveillance Program Plan

5. The Comittee prepared a letter to the EDO regarding its review of the April 12, 1985 draft version of the NRC Maintenance and Surveillance Program Plan (MSPP). The ACRS has a continuing interest in lessons that may be learned from Japar,ese maintenance evaluations and wishes to continue interfacing with the Staff and to have the opportunity for discussion of final positions before they are published.

Status Report on Long-Range Planning

6. The Comittee prepared a report to the Comissioners of the deliberations of the Subcomittee on Long-Range Planning which has been working for the last several months to assist the Comission with its long-range planning efforts. M. W. Carbon was asked to select a few specific issues for the Comittee to discuss at its October meeting. The ACRS plans to prepare position papers for consideration by the Comissioners ~in its discussions on long-rance planning.

INP0 Program on Radiation Protection

7. The Comittee prepared a report to the Comissioners of its review of the adequacy and success of the INPO program to assist comercial nuclear power plant licensees in the implementation of improved radiation protection programs.

Systematic Analysis of Operating Plants

8. The Committee prepared a memorandum to the EDO of its review of a method being developed by IDCOR for systematic analysis of operating plants.

Reply to G. Petrangeli's, ENEA, Letter dated June 9, 1985

9. The Comittee approved a letter to G. Petrangeli, ENEA, regarding his proposed " Core Rescue System" for a simplied means for decay heat removal directly from the primary system of pressurized water reactors. The ACRS suggested that G. A.

Reed, who was in favor of a more substantive reply to G.

Petrangeli, could forward additional coments as an individual member of the public, should be desire to do so. ..

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MINUTES OF THE 304th ACRS MEETING AUGUST 8-10, 1985 C. Generic Issues

1. Proposed Amendments on Physical Protection for Category II Quantities of High Enrichment Uranium J. C. Mark was assigned to review the proposed final rule regarding protection of highly enriched uranium nuclear fuel at non-power reactors and report back to the full Comittee at the September meeting his recommendation as to whether the proposed final rule merits full Comittee review.

1 D. Future Schedule

1. Future Agenda The Comittee agreed on tentative agenda items for the 305th ACRS meeting, September 12-14, 1985 (see Appendix II).

2. Future Subcomittee Activities A schedule of future Subcomittee activities was distributed to Members (see Appendix III).

E. Administrative and Procedural Items Chairman D. A. Ward advised those remaining Members present on Saturday, August 10 that the issues discussed during the joint meeting of the ACRS Subcomittees on Waste Management and Procedures and Administration on July 30, 1985 were of sufficient import that it would be better to defer discussion of them until the September meeting when the maximum number of Members would be present. He requested that a copy of the sumary of the administrative and procedural items discussed be sent to each Member so that they may review the agreements and assignments made during the July 30 Subcomittee meeting, prior to the September full Comittee meeting.

F. Comanche Peak R. F. Fraley, the ACRS Executive Director, noted a request by AS&LB Judge Peter B. Bloch that NRR forward to the ACRS for its consideration certain SER supplements which refer to 144 Quality Assurance / Quality Control deficiencies and 389 allegations associated with the Operating License hearing on Comanche Peak. F. J. Remick proposed that the relevant SERs on Comanche Peak be supplied to all Comittee members for their consideration. The Comittee concurred.

G. Activities of ACRS Members D. W. Moeller indicated that he has been invited to accompany an INPO team when it conducts an operational evaluation of the Pilgrim Nuclear Plant similar to a previous visit he made to the J. M. Farley Nuclear Plant last October. He recomended that other Members consider accompanying INP0 teams as an

, MINUTES OF THE 304th ACRS MEETING AUGUST 8-10, 1985~

observer. R. F. Fraley, the ACRS Executive Director suggested that D. W. Moeller ask INPO to inform the ACRS office of planned operational evaluations so that ACRS Members with an interest could be included in future plant visits / evaluations.

The Comittee endorsed D. W. Moeller's participation in the INP0 operational evaluation.

G. A. Reed indicated that he has been invited to attend the February ANS/ ENS meeting in San Diego and present his paper on a feed and bleed decay heat removal system. This paper was presented recently at a meeting of the ACRS with the RSK and GPR. The Comittee agreed to support his attendance at the ANS/ ENS meeting.

D. A. Ward has been asked to participate as a consultant in the NAS study of research needs on Human Factors. He will perform this work as an employee.of E. I. du Pont de Nemours and Company. Some Members worried about conflict of interest considerations should the ACRS be asked to coment on the report that results from the NAS study but offered no objection to D. A. Ward's participation.

The 304th ACRS Meeting was adjourned at 2:15 p.m., Saturday, August 10, 1985.

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APPENDIXES TO MINUTES OF THE 304TH ACRS MEETING AUGUST 8-10, 1985 l-O

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(7. TABLE OF CONTENTS

\ '# l APPENDIXES TO MINUTES OF THE 304th ACRS MEETING August 8-10, 1985

' Appendix ! -

List of Attendees ........................ A-1 Appendix II -

Future Agenda .............................. A-2 Appendix III -

Schedule of ACRS Subcomittee Meetings ..... A-4 Appendix IV -

Vogtle Status Report ....................... A-36 Appendix V _ Principle & Unique Design Features of

'P l a n t Vog tl e . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A- 4 5 Appendix VI -

NRC Regional Evaluation of Construction .... A-67 Appendix VII -

NRC Licensing Review of Plant Vogtle ....... A-75 Appendix VIII -

Vogtle Project Organization . . . . . . . . . . . . . . . . A-90 Appendix IX -

Management Philosophy at Georgia Power Company .................................... A-92

[\ Appendix X -

Georgia Power Corporate Organization ....... A-97 Appendix XI -

Vogtle Plant Operation Organization and Training ................................... A-103 Appendix XII -

Plant Vogtle Quali ty Assurance . . . . . . . . . . . . . A-112 Appendix XIII -

Vogtl e Quality Concern Program . . . . . . . . . . . . . A-121 Appendix XIV -

Vogtle Readiness Review Program ............ A-126 Appendix XV -

NRC Participation in Vogtle Readiness Review A-145 Appendix XVI -

UT Examination of Cast Stainless Steel Piping ..................................... A-146 Appendix XVII -

San Onofre Plant Overview .................. A-148 Appendix XVIII - .'C Presentation on SEP Integrated Assessment San Onofre ................................. A-154 Appendix XVIX -

GESSAR II Severe Accident Issues Presentation to the ACRS ................................ A-218 Appendix XX -

GESSAR II PRA Review Detailed Discussion of Hydrogen by T. Pratt ....................... A-220

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TABLE OF CONTENTS (Cont.)

APPENDIXES TO MINUTES OF THE 304TH ACRS MEETING

/'3 Appendix XXI -

GESSAR II FRA Review Effect of a Core Melt

( on Vessel Support Integrity by T. Pratt .... A-237 Appendix XXII -

Staff Presentation on USI A-46 Resolution .. A-248 Appendix XXIII - SQUG Presentation .......................... A-261 Appendix XXIV -

Sunnary of the Effects of the Great Chile Earthquake of 1985 ......................... A-268 Appendix XXV -

Nuclear Regulatory Maintenance & Surveillance Program .................................... A-279 Appendix XXVI -

GE Proprietary Information GESSAR II Sabotage Considerations ............................. A-301 Appendix XXVII - Additional Documents Provided for ACRS' Use .A-308 O,.y o

V

APPENDIX 1 ATTENDEES lj s,

s ATTENDEES 304th ACRS MEETING August 8-10, 1985 ADVISORY COP 941TTEE ON REACTOR SAFEGUARDS David A. Ward, Chairman Harold W. Lewis, Vice-Chairman Robert C. Axtmann Max W. Carbon Jesse C. Ebersole William Kerr Carson Mark Dade W. Moeller Glenn A. Reed Forrest J. Remick Paul G. Shewmon Chester P. Siess Charles J. Wylie ACRS Staff Raymond F. Fraley, Executive Dire.ctor M. Norman Schwartz, Technical Secretary Herman Aldennan Paul A. Boehnert Anthony J. Cappucci Robert Cushman Monideep De Sam Duraiswamy Medhat M. El-Zeftawy John Flack John T. Gilbert James A. Jeffries Janet Kotra Morton W. Libarkin Richard K. Major John A. MacEvoy Thomas G. McCreless John C. McKinley Owen S. Merrill .

Austin Newsom Sidney J.S. Parry Gary R. Quittschreiber Richard Savio Stanley Schofer O

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4 NRC ATTENDEES 304TH ACRS MEETING Thursday, August 8,1985 0FFICE OF NUCLEAR REACTOR REGULATION R. M. Bernero C. O. Thomas , DL D. C. Sca11ett, DL R. W. Hernan, PPAS W. B. Hardin, DSI N. Anderson, DST T. Y. Chang, DST D. Persinko, DHFS G. Cwalina,DHFS H. Booher, DHFS OFFICE OF NUCLEAR MATERIAL SAFETY

& SAFEGUARDS C. Gaskin BROOKHAVEN NATIONAL LABORATORY K. Shiu R. Jauney T. Pratt i

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1 INYITED ATTENDEES' 304TH ACRS MEETING ,

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1. Thursday, August 8,1985 1

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i. GENERAL ELECTRIC .!

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f. G. G. Sherwood ,

j' .G. F. Yeazell l

D. A. Hawkins  ;

D. C. Foreman I l D. J. Robare i 1  ;

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PUBLIC ATTENDEES 304TH ACRS MEETING Thursday, Aug. 8, 1985 l

E. Linderian. McGraw Hill E. Thibau, D. I. T. T. , Inc.

P. Birnie B..Schroeder D. Ferris-C. Baty . . -

R. E. Schaffstall, KMC J. E. Thortas, Duke Power i t

D. Ferris,.LILCO M. Beaumont, Westinghouse i

D. Varner, Boston Edison Company A. J. Prestesky, ANS

'L. Connor, DSA J. Trotter, NUS Corp. '

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NRC ATTENDEES

, 304TH ACRS MEETING Friday, August 9,1985 OFFICE OF NUCLEAR REACTOR REGULATION '

G. Lear, DE . V. Benaroya, DE '

R. Hernan, PPAS T. Novak, DE C. Grimes, DL F. Burrown, DSI T. Cheng, DL J. Knight, DSI D. Muller DSI W. LeFave, DSI M. Miller, DL P. Sobel, DE '

E. Adensam, DL i

R. Housh, DSI S. Spikler, DSI L. Heller, DE S. N. Saba. HFEB J. Lazevnick, DSI M. R. Hum, DE D. H. Beckham, DHFS B. Youngblood, DL D. L. Chery, DE OFFICE OF INSPECTION & ENFORCEMENT E. F. Williams REGION II e M. V. Sinkule e V. L. Brownlee J. F. Rogge

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, APPLICANT ATTENDEES

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304TH ACRS MEETING '

Friday, August 9, 1985 BECHTEL POWER CORPORATION WESTINGHOUSE 1

R. C. Snow R. Faas J. T. Luke J. McInerney D. S. Jagannathan R. J. Morrison D. L. Houghton D. W. Call G. J. Baenteli J. Larry Vota M. L. Larson C. Benton L. S. Watson W. Tauche C. McClure G. f.ang T. W. Crosby ser f GEORGIA POWER COMPANY R. A. Schilling D. O. Foster P. Rushton H. Silberman R. J. Kelly S. J. Cereghino M. Sheibani W. R. Ferris W. Burns C. Lesser s D. H. Evans SOUTHERN COMPANY SERVICES D. Townley R. L. George P. D. Rice

0. Batum R. A. Thomas K. Kopecky J. T. Beckham, Jr.

J. A. Bailey G. Bockhold T. N. Epps S. C. Ewald OGLETHORPE POWER CORP. LOWENBERG ASSOCIATES, INC.

C. Bubba McCall H. Gamky J. Jcer SOUTHERN CALIFORNIA EDISON G. E. Hammond J. Rainsberry M. O. Medeord M. P. Short K. P. Baskin O

l PUBLIC ATTENDEES 304TH ACRS MEETING i

Friday, August 10, 1985 j

C M. Larson E. Kurtz, Duquesne J. Sorrells, Morris Communications Corp.

D. Wisenberg, McGraw-Hill E. F. Kurtz, Duquesne Light H. Dezfuli, NUS Corp.

D. Runkle, Morgan Associates S. Sholly, Union of Concerned Scientists L. Connor, PSA E. Putman, Associated Press A. J. Presseky, ANS

APPENDIX 11

- Future Agenda

(~N APPENDIX A FUTURE AGENDA SEPTEMBER ACRS MEETING Report of Panel on ACRS Effectiveness -- 2 hrs Summary / discussion of report Briefing on approach to proposed revisions li hr to Appendix K and related ECCS issues Application of PRAs to Nuclear Power Stations -- li hrs Proposed ACRS coments on application of PRAs to Indian Point and other nuclear power plants 10 CFR 50. Appendix A - General Design Criteria 4 I hr Interim Schedular Exemptions -- ACRS coments regarding proposed changes in criterion for primary systems failure General Electric Standard Safety Analysis Report 4 hrs (GESSAR II) -- Continue ACRS review of the FDA for this standardized NSSS Use of Aptitude Testing of Nuclear Power Plant 2 hrs Operators -- Briefing / discussion regarding ACRS endorsement of the use of aptitude testing for O- nuclear power plant operators Regulatory Guide 1.99. Rev. 2. Effects of Residual i hr Elements on Predicted Radiation Damage to Reactor Vessel Materials -- ACRS subcommittee report ,

State of the Nuclear Power Industry -- Report of 2 hrs Ad Hoc subcomittee regarding the most important safety-related issues in nuclear power plant safety River Bend Nuclear Plant -- OL open issues regarding 2 hrs H2 generation and control Briefing by NRC Staff of Recent Events at Operating 2 hrs Nuclear Power Plants Briefing by H. Denton regarding the recent I hr reorganization of NRR Meeting with NRC Comissioners to discuss the ACRS 2 hrs report on ACRS participation in NRC regulation of radweste, the Severe Accident Policy Implementation Plan, and the Human Factor research program Proposed Amendments on Physical Protection for i hr Category II Quantities of Highly Enriched Uranium -

O Proposed final rule regarding protection of highly enriched fuel at non-power reactors e

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• Accident Source Term -- ACRS coments on NUREG-0956 defer to h source term reassessment for water-cooled nuclear power plants October Davis-Besse Nuclear Plant -- Briefing regarding defer to causes for loss of feedwater and proposed October /

corrective action November

• IE Inspection Program -- Briefing by representatives defer of IE Proposed Regulatory Policy for Advanced Reactors -- defer to ACRS comments regarding SECY-84-453A dated October February 26, 1955 Containment Sump Performance - (USI A-43) -- defer to Discur.sien of containment emergency sump October performance implementation plan Extreme Environmental Phenomena -- ACRS defer to comments regarding consideration of extreme October environmental events in emergency planning Quantitative Safety Goals -- ACRS coments on defer EDO recommendations regarding evaluation of quantitative safety goals for nuclear power (n

V) plants per ACRS report dateo July 17, 1985 Regulatory Guides -- R. G. 1.23, Rev. 2, defer

" Meteorological Measurement Programs for Nuclear Power Plants" and Proposed R. G. (Task No. IC 609-5),

Criteria for Power, Instrumentation, and Cgntrol Portions of Safety Systems"

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APPENDIX III

.. .;a [ /DIf5 SCHEDULE OF ACRS SUBCOMMITTEE MTGS SCHEDULE OF ACRS SUBCO M ITTEE MEETINGS AUGUST 27 JOINT ECCS AND FLUID DYNAMICS (BOEHNERT) - Ward, Ebersole.

Etherington.

Purpose:

(1) To review the status of the hydro-dynamic loads issue for plants with Mark I-!!! containments; (2) To review the AE00 report on Interfacing LOCAs; and (3) To review the US! A-43, "Cc ntainment Emergency Sump Perfonnance,"

implernentation proposal.

29 STATEOFNUCLEARPOWERSAFETY(LOSANGELES,CA)(CAPPUCCI)-

Kerr, Lewis, Okrent, Shewmon.

Purpose:

To discuss draft Subconnittee report.

SEPTEMBER 4&5 METAL COMPONENTS (IGNE) - Shewman, Ward.

Purpose:

To review Reg. Guide 1.99, Rev. 2 and other related concerns, and to discuss the status of the NDT of piping program and the HSST program. The Subconnittee will also review steam generators girth weld cracks.

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\'# l' 9 REACTOR OPERATIGNS (MAJOR) - Ebersole, Kerr, Michelson, (1:00P.M.) Okrent, Reed, Remick, Ward, Wylie.

Purpose:

To discuss recent operating experiences.

10 WESTINGHOUSE WATER REACTORS (CAPPUCCI) - Ebersole, (Closed) Etherington, Michelson, Shewmon, Ward (part-time).

Purpose:

To begin the PDA review of the Westinghouse Advanced

, PWR (RESAR SP/90).

LONG RANGE PLAN FOR NRC (MAJOR) - Carbon, Lewis, Moeller, Remic , Si , 1 o  : The ttee will con-tinu d sc i de op gc long range plan for he 3 l o di 15?d rily technical issue elated to the regulation of nuclear power plant safety and safety regulation over the next 5 to 10 years.

11 RIVER BEND (SAVIO) - Okrent, Ebersole, Shewmon.

Purpose:

To review Gulf States Utilities Company's application for an OL.

12 - 14 30STH ACRS MEETING l 17 & 18 HUMAN FACTORS TOUR (RUSSELVILLE, AR) (SCHIFFGENS) - Ward, l (Closed) Lewis, Michelson, Moeller, Reed, Remick, Wylie.

Purpose:

l g\ This wil1 be a tour and examination of ANO-l's emergency procedures (symptom based) and facilities.

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[V ') SCHEDULE OF ACRS SUBCOMMITTEE MEETINGS SEPTEMBER (CONT'D) 23 & 24 JOINT STRUCTURAL ENGINEERING AND SEISMIC DESIGN OF PIPING (IGNE) - Siess, Ebersole, Mark, Shewmon.

Purpose:

To review the status of research programs on containment integrity, seismic margins, piping reliability, and other related matters.

25 ADVANCED REACTORS (EL-ZEFTAWY) - Park, Carbon (tent.), Siess.

Purpose:

To discuss the proposed policy for regulation of advanced nuclear power plants.

26 WASTE MANAGEMENT (MERRILL) - Moeller, Axtmann, Carbon, Kerr, Mark, Remick, Shewnon.

Purpose:

(1) To review SECY-85-147A, Regulatory-(2) 25, 1985, To review the proposed DOE position paper (andE NRC's coments on it), Retrievability and Retrieval for Geologic Repository, and (3) (if available by meetinp date) --

To review the Comission's request or directive to the ACRS regarding the ir role in the HLW Management Geologic Repository Program.

O b OCTOBER 3&4 MILLSTONE POINT 1-3 (WATERFORD, CN) (SCHIFFGENS) - Shewmon, Kerr, Ward.

Purpose:

To review the Northeast Nuclear Energy kt Company's application for conversion of the Provisional Operating License (POL) for Millstone Unit I to a Full Tem Operating License (FTOL).

8 RELIABILITY ASSURANCE (VALVES) (MAJOR) - Michelson, Kerr, Reed, Ward.

Purpose:

To continue discussions on valve reliability. A risk perspective on valve perfomance will be sought. Also to be studied is the importance of valves from a safety standpoint.

8 SITE EVALUATION (MERRILL) - Moeller, EihmmmmMguemuunt.

Purpose:

(1) To evaluate, from a probabilistic approach, the relative importance of the full range of natural phenomena in terms of their potential impacts on Emergency Planning, (2) to provide coments on the numerical values proposed in a July 5, 1985 memo from the EDO to the Comissioners on this subject, and (3) to develop a strong ACh5 consensus view on this rulemakin Planning)g The (Consideration of Seismic preceding were requestedEvents in Emergency by Comissioners Bernthal, Asselstine, and Zech, respectively, at the ACRS meeting with the Comissioners, July 11, 1935.

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SCHEDULE OF ACRS SUBCOMMITTEE MEETING l

DATE SUBCOMMITTEE MEETING STAFF ENGR. & MFMBERS OCTOBER 3 & 4, 1985 MILLSTONE UNITS 1-3 (SCHIFFGENS) Shewmon, Kerr, Ward PURPOSE: To review the Northeast Nuclear Energy Company's application for conversion of the Provisional Operating Licelse (POL) for Millstone Unit I to a Full "

Term Operating License (FTOL).

LOCATION: WATERFORD, CN BACKGROUND:

What a'ction is requested; by what date is it needed?

Subcommittee conversion review in time for Comittee consideration.at the 307th, Nov. 7-9, 1985 ACRS meeting.

What will be done at this meeting?

Review the conversion application and draft a letter.

hat would be the consequence of postponing this meeting?

The NRC Staff schedule currently has license issuance on Dec. 11, 1985. I don't think this schedule can be met if the meeting is postponed.

PERTINENT PUBLICATIONS AND THEIR AVAILABILITY:

1. The Draft SER has been distributed.

2. A PRA Study has been issued and will be out this month (August).

3. A Draft Suppl. to IPSAR (NUREG-0824) will be made available before the end of August.

4 A Draft ISAP report will be made available by mid-September.

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) SCHEDULE OF ACRS SU3 COMMITTEE MEETINGS OCTOBER (CONT'D) 9 (tentative) COMANCHE PEAK (location to be determined) (MERRILL) - Okrent, Michelson, plus others (to be determined).

Purpose:

To review SER Supplements 7 through 11, and provide coments thereon as requested by Peter B. Bloch, Chaiman and Administrative Judge Atomic Safety and Licensing Boa.rd (ASLB).

9 REGULATORY ACTIVITIES (DURAISWAMY) - Siess,Kerr(part-time),

Moeller (part-time), Ward, Wylie (part-time).

Purpose:

To review: (1) Reg. Guide 1.23, Rev. 1, " Meteorological Measurement Programs for Nuclear Power Plants," (2) proposed.

Reg. Guide (Task No. IC 609-5), " Criteria for Power, Instru-mentation, and Control Portions of Safety Systems," and (3)

Reg. Guide 1.105, Rev. 2, " Instrument Setpoints for Safety-Related Systetrs" (tent.).

9 LONG RANGE PLAN FOR NRC (MAJOR) Carbon, Lewis, Moeller (part-time), Remick, Wylie (part-time).

Purpose:

The Subcomitter will continue discu:sions on developing coments

(, , ' , on a long range plan for the NRC. Topics to be discussed are

(" / primarily technical issues related to the regulation of nuclear power plant safety and safety regulation over the next 5 to 10 years.

10 - 12 306TH ACRS MEETING 15 NUCLEAR PLANT CHEMISTRY (ALDERMAN) - Axtmann, Etherington, Mark, Shewmon, Reed.

Purpose:

To discuss radiation chemistry of aerosol behavior.

18 JOINT REACTOR RADIOLOGICAL EFFECTS AND FIRE PROTECTION (MERRILL/ ALDERMAN) - Moeller, Axtmann, Carbon, Ebersole, Michelson, Okrent, Reed, Siess, Wylie.

Purpose:

To review the increased N-16 radioactivity and fire protection problems in using H 2addition to BWRs to reduce IGSCC.

31 (site visit) BEAVER VALLEY 2 (PITTSBURGH, PA) (ALDERMAN) - Wylie, Axtmann, (1:00P.M.) Ebersole, Kerr, Shewmon.

Purpose:

Site visit and to review Nov.1 (meeting) for an operating license.

NOVEMBER 6 (tentative) QUALIFICATION PROGRAM FOR SAFETY-RELATED EQUIPMENT

,m (CAPPUCCI) - Wylie, Ebersole, Michelson, Reed, Siess, Ward.

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Purpose:

To discuss resolution and implementation of USI V A-46.

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(w/ ) SCHEDULE OF ACRS SUBCOMMITTEE MEETINGS DATES TO BE DETERMINED (September / October) HUMAN FACTORS (SCHIFFGENS) - Ward, Reed, Remick, Wylie.

Purpose:

To explore methods for deciding what actions should be automated in nuclear power plant operation.

(October) SCRAM SYSTEMS RELIABILITY (BOEHNERT) - Kerr, Ebersole, Michelson, Okr'ent, Reed, Ward, Wylle.

Purpose:

To continue the review of the status of ATWS Rule implementation effort and related issues.

(October)(tent.) QUALITY AND QUALITY ASSURANCE IN DESIGN AND CONSTRUCTION (MAJOR) - Remick, Michelson, Okrent, Reed, Siess, Ward, Wylie.

Purpose:

(1) To review the final Rule on the "Important To Safety Issue," and (2) To be briefed on the "NRC Quality

, Assurance Program Implementation Plan."

(October) STANDARD PLANT DESIGN (ALDERMAN) - Wylie, Kerr, Michelson, Reed, Siess.

Purpose:

To be briefed by the NRC Staff on the status of standard plant design.

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(early November) DECAY HEAT REMOVAL SYSTEMS (BOEHNERT) - Ward, Ebersole,

(~' (tentative) Etherington, Reed.

Purpose:

To continue the review of NRR resolution position for USI A-45.

(November) ECCS (PALO ALTO, CA) (B0EHNERT) - Ward, Ebersole, Etherington, Reed.

Purpose:

To continue the review of the joint NRC/B&WOG/EPRI/B&W joint IST Program. A visit is planned to the EPRI Stanford Research Institute facilities supporting this Program.

(Fall) RELIABILITY & PROBABILISTIC ASSESSMENT (location to be (tent.) determined) (SAVIO) - Okrent, Kerr, Ebersole, Lewis, Mark, Michelson, Siess, Ward, Wylie.

Purpose:

To review the probabilistic risk assessment for Millstone 3.

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p._--w s _, SCHEDULE OF ACRS SUBCOMMITTEE MEETING DATE SUBCOMMITTEE MEETING STAFF ENGR. & MEMBERS AUGUST 27, 1985 JOINT ECCS AND (B0EHNERT) Ward, FLUID DYNAMICS Ebersole, Etherington e Cons.: Catton, Schrock, Sullivan, Theofanous Tien PURPOSE: (1) To' review the status of the hydrodynamic loads issue for plants with Mark I-III containments.

(2) To review the AEOD report on Interfacing LOCAs.

> (3) To review the USI A-43, " Containment Emergency Sump Performance,"

implementation proposal.

LOCATION: WASHINGTON, DC BACKGROUND:

- What action is requested; by what date is it needed?

- do specific action date needed.

What will be done at this meeting?

(1) Review the status of hydrodynamic loads issue for Mark I-III containment plants.

(2) Review AE0D report on Interfacing LOCAs.

(3) Review implementation plan for USI A-43.

What would be the consequence of postponing this meeting?

Loss of timely review of Item 3 per NRR's schedule.

PERTINENT PUBLICATIONS AND THEIR AVAILABILITY:

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f i ) SCHEDULE OF ACRS SUBCOMMITTEE MEETING DATE SUBCOP91ITTEE MEETING STAFF ENGR. & MEMBERS AUGUST 29, 1985 STATE OF NUCLEAR POWER SAFETY (CAPPUCCI) Kerr, Lewis, Okrent, Shewmon PURPOSE: To discuss the draft report on the " State of Nuclear Power Safety" in preparation by the Subcommittee.

LOCATION: LOS ANGELES, CA BACKGROUND:

What action is requested; by what date is it needed?

Connittee initiated action. Report to Chaiman Palladino to be written at the 305th (September 1985) ACRS meeting on the State of Nuclear Power Safety.

What will be done at this meeting?

CTo finalize the draf t report for full Comittee consideration.

U What would be the consequence of postponing this meeting?

Would not meet schedules set by Committee.

PERTINENT PUBLICATIONS AND THEIR AVAILABILITY:

Memo from W. Kerr to C. Michelson, D. Okrent, H. Lewis, G. Reed, P. Shewmon, "Draf t R; port," dated July 10, 1985.

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I SCHEDULE OF ACRS SUBCOMMITTEE MEETING DATE SUBCOMMITTEE MEETING STAFF ENGR. & MEMBERS SEPTEMBER 4 & 5, 1985 METAL COMPONENTS (IGNE) Shewnon, Ward Cons.: Dillon, Kassner(tent.),

Odette PURPOSE: To review Regulatory Guide 1.99, Rev. 2, and other related concerns, and to discuss the status of the NDT of piping program and HSST program. Also to review steam generators girth weld cracks.

LOCATION: WASHINGTON, DC BACKGROUND:

_ What action is requested; by what date is it needed?

_ ..CRS action is requested by the NRC Staff before Reg. Guide 1.99, Rev. 2, is promulgated.

What will be done at this meeting?

The Subcomittee will review with the NRC Staff Reg. Guide 1.99, Rev. 2, and develop recommendations for ACRS comments.

What would be the consequence of postponing this meeting?

None, except that ACRS coments, if any, will not impact Reg. Guide 1.99, Rev. 2.

PERTINENT PUBLICATIONS AND THEIR AVAILABILITY:

1. Reg. Guide 1.99, Rev. 2 " Effects of Residual Elements on Predicted Radiation Damage to Reactor Vessel Materials, due for ACRS review in August 1985.

2. A status report will be provided with pertinent background infonnation prior to the

. meeting.

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! ) SCHEDULE OF ACRS SUBCOMMITTEE MEETING

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DATE SUBCOMMITTEE MEETING STAFF ENGR. 8 MEMBERS SEPTEMBER 9, 1985 REACTOR OPERATIONS (MAJOR)Ebersole,Kerr, (1:00 P.M.) Michelson, Okrent, Reed, Remick, Ward, Wylie PURPOSE: The Subcommittee will discuss recent operating occurrences.

LOCATION: WASHINGTON, DC BACKGROUND:

What" action is requested; by what date is it needed?

Review recent operating experience, select incidents of significance for presentation to full ACRS during September meeting.

' hat will be done at this meeting?

Review recent operating experience.

What would be the consequence of postponing this meeting?

Would consider events at a later date.

PERTINENT PUBLICATIONS AND THEIR AVAILABILITY:

1. Status Report to be provided, r^x k

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C (v) SCHEDULE OF ACRS SUBCOMMITTEE MEETING DATE SUBC0ptilTTEE MEETING STAFF ENGR. & MEMBERS SEPTEMBER 10, 1985 WESTINGHOUSE WATER (CAPPUCCI)Ebersole, REACTORS Etherington, Michelson, (CLOSED) Shewmon, Ward (part-time)

Cons.: Davis PURPOSE: To begin PDA review of Westinghouse Advanced PWR (RESAR SP/90).

LOCATION: WASHINGTON, DC BACKGROUND:

What action is requested; by what date is it needed?

\ fCRS letter on PDA approval by 11/86.

What will be done at this meeting?

'Begin reviewing design modules.

What would be the consequence of postponing this meeting?

Delay in the completion of ACRS PDA review.

PERTINENT PUBLICATIONS AND THEIR AVAILABILITY:

1. RESAR SP/90 Standard Plant Design (50-601).

SCHEDULE OF ACRS SUBCOMMITTEE MEETING DATE SUBCOMMITTEE MEETING STAFF ENGR. & MEMBERS SEPTEMBER 11, 1995 LONG RANGE PLAN FOR NRC (MAJOR) Carbon, Lewis, Moeller, Remick, Siess.

Wylie PURPOSE: The Subcommittee will continue di ssions n developing comments on a long range plan for the NRC. Topics nder iscu sfon are primarily technical issues related to the regulation of nu(erad power plant safety and safety regulation over the next 5 to 10 yefs.

LOCATION: WASHINGTON, DC BACKGROUND:

What action is requested; by what date % it/needed?

Currently a report addressed to the ommi ion or as input into a parellel OPE effort is expected. The current project 1 st conclude this effort in October 1985, at will be done at this meeti d To be determined.

What would be the conseque W stponing this meeting?

Timeliness of effort w Id b ffected. Would become out of phase with a parallel OPE cffort on LRP.

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PERTINENT PUBLICATIONS AND THEIR AVAILABILITY:

Dr. Carbon's latest review plan for this effort is available.

A Status Report will be prepared before the meeting.

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SCHEDULE OF ACRS SUBCOMMITTEE MEETING DATE SUBCOMMITTEE MEETING STAFF ENGR. & MEMBERS SEPTEMBER 11, 1985 RIVER BEND (SAVIO) Okrent.

Ebersole, Shewmon PURPOSE: To review Gulf States Utilities Company's application for an OL.

LOCATION: WASHINGTON, DC BACKGROUND:

What. action is requested; by what date is it needed?

Issue an ACRS " full power" OL letter; at the September ACRS meeting.

~

What will be done at this meeting?

/ Complete the Subcommittee action on the River Bend OL review in support of an ACRS y]/ review at the September ACRS meeting.

What would b. the consequence of postponing this meeting?

Possible delay of River Bend full power operation.

PERTINENT PUBLICATIONS AND THEIR AVAILABILITY:

NRC Staff SER Supplement to be supplied at the August ACRS meeting.

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I SCHEDULE OF ACRS SUBCOMMITTEE MEETING DATE SUBCOMMITTEE MEETING STAFF ENGR. & MEMBERS SEPTEMBER 17 & 18, 1985 HUMAN FACTORS TOUR (SCHIFFGENS) Wari, (CLOSED) Lewis, Michelson, Moeller, Reed, Remick, Wylie PURPOSE: This will be a tour and examination of ANO-l's emergency procedures (symptom based) and facilities. The Subcomittee wants the opportunity to examine procedures at an operating plant and see how the TMI required backfits such as SPDS interface. Up to a day and a half is expected. ANO-1 is an 850 MWe, B&W PWR.

LOCATION: ANO-1, Russellville, AR (" 50 miles outside of Little Rock, AR)

BACKGROUND:

What action is requested; by what date is it needed?

eview implementation of TMI required backfits. No schedular requirements.

What will be done at this meeting?

Tcur and review facilities.

What would be the consequence of postponing this meeting?

N;ne PERTINENT PUBLICATIONS AND THEIR AVAILABILITY:

1. One copy of ANO-1 Emergency Operating Procedures is available for your inspection at the ACRS Office (ask J. Schiffgens for it).

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\w, ) SCHEDULE OF ACRS SUBCOPWITTEE MEETING DATE SUBCOPMITTEE MEETING STAFF ENGR. & MEMBERS SEPTEMBER 23 & 24, 1985 JOINT STRUCTURAL ENGINEERING (IGNE) Siess, Ebersole, AND SEISMIC DESIGN OF PIPING Mark, Shewmon Cons.: Bender PURPOSE: To review the status of research programs on containment integrity, seismic margins, piping reliability, and other related matters.

LOCATION: WASHINGTON, DC BACKGROUND:

What action is requested; by what date is it needed?

M eeds and f;:eeting is justifications.

requested Thisby information will be needed to plan the Subcommittee for ACRS consnents Chainnan in order to ke x n the Research Program and Budget.

What will be done at this meeting?

Discuss the status of the research program with NRC research and regulatory. staffs in order to plan to provide ACRS coments for research program and budget reviews.

What would be the consequence of postponing this meeting?

• None, except that it is required by law that the ACRS provide Congress and the Commission periodic reports on the NRC Research Program and Budget.

PERTINENT PUBLICATIONS AND THEIR AVAILABILITY:

A status report will be provided with pertinent background infonnation prior to the me2 ting.

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SCHEDULE OF ACRS SUBCOMMITTEE MEETING DATE SUBCOPe!ITTEE MEETING STAFF ENGR. & MEMBERS SEPTEMBER 25, 1985 ADVANCED REACTORS (EL-ZEFTAWY) Mark, Carbon (tent.),Siess PURPOSE: To discuss the proposed policy for regulation of advanced nuclear power plants.

LOCATION: WASHINGTON, DC BACKGROUND:

What action is requested; by what date is it neede_d?

ACRS comments on the proposed policy statement for advanced reactors; 306th ACRS meeting (October 1985).

O( :Tat will be done at this meeting?

Review the revised version of the policy statement. Prepare the report to the full Committee and suggested report to the Comission.

What would be the consequence of postponing this meeting?

Delay of ACRS comments to the Comission.

PERTINENT PUBLICATIONS AND THEIR AVAILABILITY:

1. Letter fm R. Fraley to J. Zerbe, dated 4/15/85.

2. SECY-84-453A - Regulatory policy for advanced reactors, dated 2/26/85.

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1 SCHEDULE OF ACRS SUBCOMMITTEE MEETING DATE SUBCOMMITTEE MEETING STAFF ENGR. A MEMBERS SEPTEMBER 26, 1985 WASTE MANAGEMENT (MERRILL) Moeller, Axtmann, Carbon, Kerr, Mark, Remick, Shewmon Cons.: Carter, Krauskopf, Parker, Steindler PURPOSE: (1) To review SECY-85-147A, Regulatory-Exempt Radiation Levels (de minimis values), July 25, 1985.

(2) To review the DOE proposed position paper (and NRC's coments on it),

Retrievability and Retrieval for a Geologic Repository.

(3) If available by the meeting date, review the Comission's request or directive to the ACRS regarding their role in the HLW Management Geologic Repository Program.

LOCATION: WASHINGTON, DC KGROUND:

What action is requested; by what date is it needed?

Items 1 and 2 recomended by D. Moeller, (I) to update Comittee on de minimis developments in other countries relative to revised 10 CFR 20 (SECY-B5-147), (2) to become better informed on nature of retrievability issues, particularly on NRC and DOE positions on the' subject, (3) develop plans for Subcomittee make-up and role as requested by Comission (request anticipated prior to meeting date).

What will be done at this meeting?

Review subjects named above.

What would be the consequence of postponing this meeting?

Lose advantage of timeliness on all 3 items; but no adverse consequences foreseen.

PERTINENT PUBLICATIONS AND THEIR AVAILABILITY:

1. SECY-85-147A Policy Issue (Infomation) from EDO to the Comissioners,

" Regulatory-Exempt Radiation Levels (dc minimis levels), July 25, 1985

2. Memo for D. Moeller from S. Parry, "SECY-85-197 - NRC/ DOE Staff Meeting on DOE's

. m Proposed Position on Retrievability and Retrieval for a Geologic Repository -

(  ! July 31, 1985," August 7, 1985

\

' 00E Proposed Position Paper, "Retrievability and Retrieval for a Geologic Repository." (need to obtain)

4. NRC coments on (3) (need to obtain)

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v SCHEDULE OF ACRS SUBC09911TTEE MEETING DATE SUBCOMMITTEE MEETING STAFF ENGR. & MDWERS I OCTOBER 8, 1985 RELIABILITY ASSURANCE (MAJOR) Michelson, (VALVES) Kerr, Reed, Ward PURPOSE: To continue discussions on valve reliability. A risk perspective on valve perfomance will be sought. Also to be studied is the importance of valves from a safety standpoint. A discussion with Limitorque Co. is also expected.

LOCATION: WASHINGTON, DC BACKGROUND:

What action is requested; by what date is it needed?

No action has been requested. This is a self-initiated task.

What will be done at this meeting?

is meeting will conclude a series of three meetings designed to explore the topic of f)lve g u reliability.

What would be the consequence of postponing this meeting?

No adverse impact.

PERTINENT PUBLICATIONS AND THEIR AVAILABILITY:

A Status fleport will be issued prior to the meeting.

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SCHEDULE OF ACRS SUBCOMMITTEE MEETING DATE SUBCOMMITTEE MEETING STAFF ENGR. 8 MEMEERS OCTOBER 8, 1985 SITE EVALUATION (MERRILL) Moeller,

1. 2 , ^' n t.

Cons.: Foster, Healy, Parker, Wilson PURPOSE:

As requested by Comissioners Bernthal, Asselstine, and Zech, respectively (at the ACRS meeting with the Comissioners, July 11,1985):

1) To evaluate, from a probabilistic approach, the relative importance of the full range of natural phenomena in terms of their potential impact on emergency planning;

2) To provide comments on the numerical values proposed in a July 5, 1985 memo from the E00 to the Comissioners on this subject, and

3) To develop a strong ,\CRS consensus view on this rulemaking (Consideration of Seismic Events in Emergency Planning).

LOCATION: WASHINGTON, DC What action is requested; by what date is it needed?

See purpose above; date unspecified; recomend meeting be held before September full Committee meeting to be responsive to Comissioners' requests.

What will be done at this meeting?

Review and draf t letter to the Commissioners on the above topics.

What would be the consequence of postponing this meeting?

Non-responsive to Commissioners' request.

PERTINENT PUBLICATIONS AND THEIR AVAILABILITY:

A Status Report will be issued prior to the meeting.

t SCHEDULE OF ACRS SUBCOMM11 TEE MEETING DATE SUBCOMMITTEE MEETING STAFF ENGR. & MEMBERS OCTOBER 9. 1985 COMANCHE PEAK (MERRILL)Okrent.

(TENTATIVE)

Michelson,fto by others besupplemented deteminedj Cons.: To be selected.

PURPOSE: To review SSERs 7 through 11 and provide coments thereon, as requested by Peter B. Bloch, Chairman and Administrative Judge, ASLB.

LOCATION: WASHINGTON, DC BACKGROUND:

What action is requested; by what date is it needed?

See above; no date specified.

See purpose.

What would be the consequence of postponing this reeting?

None if postponed 1 or 2 months.

Note: V. Noonan, Project Engineer, is currently checking with NRC Staff to detemine if meeting is really necessary.

PERTINENT PUBLICATIONS AND THEIR AVAILABILITY:

1. NUREG-0797, Safety Evaluation Report, comanche Peak, Supplements 7 through 11.

2. A Status Report will be issued prior to the meeting.

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SCHEDULE OF ACRS SUBCOMMITTEE MEETING DATE SUBCOMMITTEE MEETING STAFF ENGR. & MEMBERS OCTOBER 9, 1985 REGULATORY ACTIVITIES (DURAISWAMY) Siess, Kerr(part-time),

Moeller (part-time),

Ward,Wylie(part-time)

PURPOSE: To review the following:

1. Regulatory Guide 1.23. Rev 1, " Meteorological Measurement Programs for Nuclear Power Plants" (pre-coment).

2. Proposed Regulatory Guide (Task No, IC 609-5), " Criteria for Power, Instrumentation, and Control Portions of Safety Systems" (post coment) .

3. Regulatory Guide 1.105, Rev. 2. " Instrument Setpoints for Safety-Related Systems."

LOCATION: WASHINGTON, DC BACKGROUND:

hat action is requested; by what date is it needed?

ACRS concurrence in the Staff's posposal to issue item 1 for public comments.

ACRS concurrence in the Regulatory positions of items 2 & 3.

What will be done at this meeting?

See purpose.

What would be the consequence of postponing this meeting?

Delay the issuance of these Guides.

PERTINENT Pil8LICATIONS AND THEIR AVAILABILITY:

The above mentioned Regulatory Guldes will be sent to the Subcomittee as soon as they are made available to the ACRS Office, m

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SCHEDU E OF ACTS SUBCOMMITTEE MEETING DATE SUBCOMMITTEE MEETING STAFF ENGR. & MEMBERS OCTOBER 9, 1985 LONG RANGE PLAN FOR NRC (MAJOR) Carbon, Lewis, Moeller(part-time),

Remick Wyliefpart-time)

PURPOSE: The Subcomittee will continue discussions on developing coments a long range plan for the NRC. Topics under discussion are primarily technical issues related to the regulation of nuclear power plant safety and safety regulation over the next 5 to 10 years.

LOCATION: WASHINGTON, DC BACKGROUND:

What action is requested; by what date is it needed?

Currently a report addressed to the Comission or as input into a parallel OPE effort is expected. The current projection is to conclude this effort in October 1985.

hat wi'l be done at this meeting?

To be determined.

What would be the consequence of postponing this meeting?

Timeliness of effort would be effected. Would become out of phase with a parallel OPE effort on LRP.

PERTINENT PUBLICATIONS AND THEIP AVAILABILITY:

Dr. Carbon's latest review plan f or this effort is available.

A Status Report will be prepared before the meeting.

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i SCHEDULE OF ACRS SUBCOMMITTEE MEETING DATE SUBCOMMITTEE MEETING STAFF ENGR. & MEMBERS OCTOBER 15, 1985 NUCLEAR PLANT CHEMISTRY (ALDERMAN) Axtmann, Etherington, Mark, Shewmon, Reed PURPOSE: To discuss radiation chemistry of aerosol behavior.

LOCATION: WASHINGTON, DC BACKGROUND:

What action is requested; by what date is it needed?

Aerosol behavior can play an important part in fission product release. There are unknowns regarding behavior; no specific date needed.

Briefing by experts.

What would be the consecuence of postponing this meeting?

None Pi_~INENTPUBLICATIONSANDTHEIRAVAILABILITY:

None o

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x SCHEDULE OF ACRS SUBCOMMITTEE MEETING DATE SUBCOMMITTEE MEETING STAFF ENGR. & MEMBERS OCTOBER 18, 1985 JOINT REACTOR RADIOLOGICAL EFFECTS (MERRILL/ ALDERMAN)

AND FIRE PROTECTION Moeller, Axtmann, Carbon, Ebersole, Michelson, Okrent, Reed, Siess, Wylie PURPOSE: To review the increased N-16 radioactivity and fire protection problems in using H2addition to BWRs to reduce IGSCC.

LOCATION: WASHINGTON, DC BACKGROUND:

What action is requested; by what date is it needed?

Comments to Staff. Needed by no specific date.

Review of H 2 addition regarding N-16 activity and fire hazards.

What would be the consequence of postponing this meeting?

Meeting suggested by Committee members. Postponement of meeting will not have any

, serious consequences.

PERTINENT PUBLICATIONS AND THEIR AVAILABILITY:

1. Paper by V. Benaroya regarding subject available.

2. EPRI workshop report on subject, JAJARC #85-WH21 in ACRS office.

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i SCHEDULE OF ACRS SUBCOMMITTEE MEETING

.x DATE SUBCOMMITTEE MEETING STAFF ENGR. 8 MEMBERS NOVEMBER 1, 1985 BEAVER VALLEY 2 (ALDERMAN) Wylie, (0CTOBER 31, 1985, sitevisit) Axtmann, Ebersole, Kerr, Shewmon PURPOSE: To review application for an operating license.

LOCATION: PITTSBURGH, PA BACKGROUND:

What action is requested; by what date is it needed?

Review application for operating license - ACRS letter report; as soon as possible.

What will be done at this meeting?

Visit the site and review readiness for Beaver Valley !! for operating license.

' hat would be the consequence of postponing this meeting?

Slippage of license.

PERTINENT PUBLICATIONS AND THEIR AVAILABILITY:

1. SER to be issued about mid-September.

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i. ) SCHEDULE OF ACRS SUBC0pmITTEE MEETING

'O DATE SUBCOPMITTEE MEETING STAFF ENGR. 8 MEMBERS

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NOVEMBER 6, 1985 QUALIFICATION PROGRAM FOR SAFETY- (CAPPUCCI)Wylie, (tentative)- RELATED EQUIPMENT Ebersole, Michelson, Reed, Siess, Ward  ;

Cons.: Lipinski PURPOSE: To discuss the final resolution and implementation of USI A-46. Seismic Qualification of Equipment in Operating Plants .(post public coment phase).

LOCATION: WASHINGTON, DC i

BACKGROUND:

What action is requested; by what date is it needed?

ACRS coments on the final resolution of USI A-46 following public coments.

OM hat will be done at this meeting?

Review final resolution and implementation plan for USI A-46; prepare report to full Committee and suggested report to the Comission.

What would be the consequence of postponing this meeting?

Delay of ACRS comments to Comission.

PERTINENT PUBLICATIONS AND THEIR AVAILABILITY Letter fm K. Kniel to F. Fraley outlining public comments plans, dated July 23, 1985.

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f SCHEDULE OF ACRS SUBCOMMITTEE MEETING

'J DATE SUBC0petITTEE MEETING STAFF ENGR. 8 MEleERS TO BE DETERMINED HUMAN FACTORS (SCHIFFGENS) Ward, (SEPTEMBER /0CTOBER) Reed, Remick, Wylie Cons.: Ginny PURPOSE: To explore methods for deciding what actions should be automated in nuclear power plant operation.

LOCATION: WASHINGTON, DC BACKGROUND:

What act sn is requested; by what date is it needed?

Mr. Ward asked researchers from the University of Illinois to make a presentation to the Subconnittee.

What will be done at this meeting?

What would be the consequence of postponing this meeting?

No serious consequences from postponement that ! can see.

PERTINENT PUBLICATIONS AND THEIR AVAILABILITY:

None at this time.

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SCHEDULE OF ACRS SUBCOMMITTEE NEETING DATE SUBCOMMITTEE MEETING STAFF EhGR. & HEMBERS TO BE DETERMINED SCRAM SYSTEMS RELIABILITY (BOEHNERT)Kerr, (OCTOBER) Ebersole. Michelson, Okrent, Reed, Ward, Wylie Cons.: Lee, Lipinski, Davis PURPOSE: To continue the review of the ATWS Rule implementation effort.

LOCATION: WASHINGTON, DC BACKGROUNG:

What action is requested; by what date is it needed?

No specific action date.

at will be done at this meeting?

(SeePurposeabove)

What would be the consequence of postponing this meeting?

No significant consequences.

PERTINENT PUBLICATIONS AND THEIR AVAILABILITY:

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SCHEDULE OF ACRS SUBCOMMITTEE MEETING

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DATE SUBCOMMITTEE MEETING STAFF ENGR. & MEMBERS TO BE DETERMINED QUALITY AND QUALITY ASSURANCE (MAJOR)Remick, (October) ,

IN DESIGN AND CONSTRUCTION Michelson Okrent, Reed, Siess, Ward, Wylie PURPOSE: (1) To review the final rule on the, "Important to Safety Issue."

(2) To be briefed on the NRC Quality Assurance Program Implementation Plan.

LOCATION: WASHINGTON, DC -

BACKGROUND:

What action is requested; by what date is it needed?

,-- Approval of rule by IE 0AB; by the August full Committee meeting, i

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~' What will be done at this meeting?

Review final rule in preparation to bring before the full Comittee for comment / approval. Discussion of NRC's Quality Assurance Program Implementation Plan, for information.

What would be the consequence of postponing this meeting?

It could delay issuance of the final rule on Important to Safety Issue.

PERT!NENT PUBLICATIONS AND THEIR AVAILABILITY:

1. " Quality Assurance Program Implementation Plan," SECY-85 65 is available.

2. Current (for public coment) version of the Important to Safety Rule is available.

Revised version and response to public comments expected prior to meeting.

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SCHEDULE OF ACRS SUBCOMMITTEE MEETING DATE SUBCOMMITTEE MEETING STAFF ENGR. & HEMBERS TO BE DETERMINED STANDARD PLANT DESIGN (ALDERMAN)Wylie,Kerr, (OCTOBER) Michelson, Reed, Siess PURPOSE: To be briefed by the NRC Staff on the status of standard plant design.

LOCATION: WASHINGTON, DC BACKGROUND:

What action is requested; by what date is it needed?

To infonn the Subcommittee on what is being done on standard plant design; date is not critical.

What will tie done at this meeting?

Briefing by NRC Staff and possibly outside organizations, hat would be the consequence of postponing this meeting?

None PERTINENT PUBLICATIONS AND THE!R AVAILABILITY:

1. OTA report, "huclear Power Plant Standardf28 tion." (avail.3blenow)

2. Report by R. Cushman, ACRS Fellow, issued m

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) SCHEDULE OF ACRS SUBCOMMITTEE MEETING DATE SUBCOMMITTEE MEETING STAFF ENGR. A MEMBERS TO BE DETERMINED DECAY HEAT REMOVAL SYSTEMS (BOEHNERT) Ward, (early NOVEMBER) Ebersole. Etherington, (tentative) Reed Cons.: Catton, Davis PURPOSE: To continue the review of NRR resolution position for USI A-45.

LOCATION: WASHINGTON, DC BACKGROUND:

What action is requas'ed; by what date is it needed?

N/A What will be done at this meeting?

<~'~: gin review of NRR's proposed resolution position for USI A-45.

What would be the consequence of postponing this meeting?

At this time, given the "spongyness" in the schedule, no definitive answer can be given to this question.

PERTINENT PUBLICATIONS AND THEIR AVAILABILITY:

To be provided when available.

SCHEDULE OF ACRS SUBCOMMITTEE MEETING DATE SUBCOMMITTEE MEETING STAFF ENGR. 8 MEMBERS TO BE DETERMINED ECCS (80EHNERT) Ward, (NOVEMBER) Ebersole, Etherington, Reed Cens.: Catton, Schrock, Sullivan, Th'eofanous Tien PURPOSE: To continue the review of the joint NRC/B&W Owners Group /EPRI/B&W joint IST and related programs.

LOCATION: PALO ALTO, CA area BACKGROUND: ,

W;1at action is requested; by what date is it needed?

No specific action date -- part of ongoing Subcomittee review of Program, hat will be done at this meeting?

Continue Program review. Key discussion topics will largely be determined ',.ssed on previous Subcomittee meeting in June. Also visit EPRI SRI-supported test facilities.

What would be the consequence of postponing this meeting?

No significant impact vis-a-vis facility visits.

Uncertain on program discussion pending results of June Subcommittee meeting in Alliance, OH.

PERTINENT DOCUMENTS AND THEIR AVAILABILITY:

To be provided on a timely basis.

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SCHEDULE OF ACRS SUBCOMMITTEE MEETING O)

DATE SUBCOMITTEE MEETING STAFF ENGR. & MEM8ERS TO BE DETERMINED kiLIABILITY AND PROBABILISTIC (SAVIO) Okrent Kerr, (FALL) ASSESSMENT Ebersole, Lewis, Mark, (tentative) Michelson, Siess, Ward, Wylie Cons.:

PURPOSE: To review the PRA for Millstone 3 (not an OL critical path item).

LOCATION: To be determined BACKGROUND:

What action is requested; by what date is it needed]_ ,

Review of the Millstone 3 PRA; the meeting is to be scheduled after the completion of

'The NRC Staff's review of the PRA (estimated to be by the end of May 1985),

sahere is no ACRS action date.

What will be done at this meeting?

Review of the Millstone 3 PRA for information.

What would be the consequence of postponing this meeting?

ACRS has stated that this review need not be completed prior to full power operation.

PERTINENT PUBLICATIONS AND THEIR AVAILABILITY:

1. Millstone 3 PRA (distributed).

2. NRC Staff report on the results of the NRC/LLNL review of the Millstone 3 PRA (expected by the end of May 1985).

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APPENDIX IV Vogtle Status Report c 3 or 3l

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l Meeting :; o . - Agenda Iten Handout No.

304 /0 /

wu igne hour 3rerus geoer Authors [, & JgLgy List of Decunents Attached YA WJ [EPQRr 070F03ED SCHEDULE i

STATEMEN7 BY W ( l/?wlfff l

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Instructions to Prectrer From Staff Person

1. Punch holes c ,

2. Paginate attachments g 3. P' a c e copr in file box

b b V0GTLE ELECTRIC GENERATING PLANT UNITS 1 AND 2 STATUS REPORT FOR ACRS MEETING AUGUST 9, 1985

Purpose:

. To review the Vogtle Electric Generating Plant, Units 1 and 2 for operating licenses. An ACRS letter is requested by the NRC Staff. The*

Vogtle Subcommittee visited the plant on July 18 and held a meeting that was open to the public at the Augusta Hilton Hotel on July 18 and 19.

Plant

Description:

Each unit of the Vogtle Electric Generating Plant (Plant Vogtle) is a 1157 MWe (3411 MWt) pressurized water reactor by Westinghouse, similar to the Comanche Peak and SNUPPS units typified by the Wolf Creek plant.

These are four loop units. Plant Vogtle is located on the Savannah River in northern Georgia, near Augusta and opposite the DOE Savannah River Plant. Currently, the closest resident lives about 1.2 miles from the reactor building. The minimum distances to the exclusion area boundary is 3600 feet. The population within five miles of the site

(1980 data) is 1085 and within ten miles in 2560. The site population p i index (SPI) for this site is 0.03 which makes it comparable to Summer

'and Grand Gulf. -

Unit one is approximately 80% complete while unit two is about 50%

complete. Fuel load for Unit one is currently projected for about December 1986. Most of the Class I structures are founded on compacted backfill. Unlike Midland, the utility appears to have monitored the backfill operation and assured adequate compaction, hence unacceptable differential settlement is not anticipated since settlement to date is within expected valves. The Safe Shutdown Earthquake peak acceleration is set at 0.2g with a response spectrum derived from Regulatory Gaide 1.60. The Subcommittee appeared to be satisfied with these values.

IThe plant utilizes natural draft cooling towers for nonnal operation; each unit has, in addition, two 100% capacity Class I mechanical draft cooling towers that serve as the ultimate heat sink. The utility will have more to say about these cooling towers during the meeting.

rThe Subcommittee did not find any technical issues that were disturbing 1 or that would inhibit the issuance of an operating license.

Utility Description '

Georgia Power Company is the principal owner of Plant Vogtle with Oglethorpe Power Corp. and the Municipal Electric Authority of Georgia owning significant fracM ons and the City of Dalton owning a ninor e fraction. The plant will be operated by Georgia Power which also

-(

a operates two other nuclear units (Hatch 1 & 2). Georgia Power Company is a wholly owned subsidiary of the Southern Company. Mr. Ruble Thomas, n-37

Status Rpt/Vogtle 2 August 9, 1985 Vice President of Georgia Power, is the chief coordinator for this ACRS

' (,,} meeting while Mr. Jim Bailey is the Vogtle Licensing Manager. . Georgia Power has instituted a comprehensive Readiness Review Program that is sort of an independent design and construction review. This program represents a significant commitment of resources by the utility and the NRC Staff and both groups have been asked to discuss this program at this meeting.

The Plant Vogtle operating staff currently numbers 432 with an accu-mulated equivalent of over 1,450 years of nuclear operating experience.

The Georgia Power Company has built an elaborate training facility .

separated from the main plant. The facilities are elegant and look more like well equipped college laboratories than apprentice /journyman training grounds. The Subcomittee seemed satisfied that the utility appeared to be conservatively staffed for operation.

There will be a contested hearing for these operating licenses. The Subcomittee heard two oral statements from members of the public. One from a Mr. Tim Johnson that challenged the utility's Readiness Review Program and pointed out that there is a growing list of allegations. He

/, was unable to provide to be very a written specific statement, noat suchthe statement Subcomittee has meeting (and promised yet 8/6/85)been i received. The second speaker from the public was Mr. Lawless who ex-pressed concerns regarding radioactive contamination from the DOE Savannah River ~ Plant and also concerns regarding the assurance of the

integrity of the marl layer, under the Class I Vogtle structures, as a

- n) barrier between plant releases and the Tuscaloosa aquifer and for the h' f modeling of ground water flow rate and direction. A copy of his state-ment is attached. Thus far (8/6/85) no other written statements have i

been received regarding Plant Vogtle.

Conclusion:

The Subcommittee did not identify any substant'ive issues that would require Comittee consideration prior to ACRS approval of a full power license. An ambitious schedule is attached with many items designed to inform the Comittee but which can be dropped if the Comittee believes it has sufficient information.

The Subcommittee Chairman, Mr. Ebersole, will have a draft letter prepared for Comittee consideration.

,b] .

qO '

PROPOSED SCHEDULE FOR DISCUSSION OF THE OPERATING

. LICENSE APPLICATION FOR V0GTLE ELECTRIC GENERATING PLANT, UNITS 1 AND 2 FRIDAY AUGUST 9, 1985 DRAFT 2 i t

i 2:3'O p.m. '~ I. . Subcommittee Chairman's report 15 min.

2:45 p.m. II. Utility Overview 20 min.

A. Principal design features  !

8. Unique design features C. Plant construction status ,

D. Completion schedule  ;

3:05 p.m. III. NRC Evaluation of Construction 15 min.

A. History of construction compliance with NRC requirements B. QA problems and resolutions

~

3:20 p.m. IV. NRC Review of Open Items' 20 min.

A. Equipment Qualification B. Preservice Inspection Program '

C. Containment sump D. Toxic gas evaluation of chemicals E. Generic Letter 83-28 F. Emergency response capability--RG 1.97,

.( 0 Rev. 2

( G. Fire protection items l

H. Safe and alternate shutdown capability ,

I. Training of emergency diesel generator personnel J. Diesel fuel oil storage tank cathodic protection I. Licensee qualifications for o,peration L. Retesting of simulator response ,

(NUREG-0737. Item I.A.2.1)

M. Emergency preparedness N. Human factors engineering items

. 3:40 p.m. V. Utility Position on Selected Open Items 15 min.

3:55 p.m. VI. Utility Management, Philosophy, and .,

Organization 3:55 p.m. A. Philosophy 10 min.

4:05 p.m. B. Organization 15 min.

1. Corporate Organization

2. Project Organization

3. Plant Operations

4. Training

5. Quality Assurance (Construction and Operation)

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t APPENDIX V - PRINCIPLE & UNIQUE DESIGN FEATURES OF PLANT V0GTLE

(.O PRINCIPAL AND UNIQUE DESIGN FEATURES ACRS FULL COMMITTEE MEETING AUGUST 9, 1985 0 ZEN BATUM GENERAL MANAGER PROJECT ENGINEERING V0GTLE PROJECT 0

8-vr - _ __

9 O

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VEGP FACT SHEET e REACTORS:

TYPE FOUR LOOP PRESSURIZED WATER REACTOR .

SIZE 1160 MEGAWATTS PER UNIT

! e NUCLEAR STEAM SUPPLY SYSTEM:

! . WESTINGHOUSE ELECTRIC COMPANY ,,

f e TURBINE GENERATOR:

GENERAL ELECTRIC

e OPERATOR / CONSTRUCTOR

I GEORGIA POWER COMPANY .

t i e ARCHITECT / ENGINEER:

i BECHTEL POWER CORPORATION AND SOUTHERN COMPANY SERVICES ,

e SIZE OF CONSTRUCTION SITE:

l l 3169 ACRES 1 - . -. ..

4 (

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d VEGP o '-

PRINCIPAL DESIGN FEATURES . . .

l ,

1 j NSSS REACTOR WESTINGHOUSE PWR THERMAL POWER (MWt) 3411 ,

PUMP HEAT (MWp) 14 TOTAL THERMA . POWER (MWt) 3425 l 4 NUMBER OF COOLANT LOOPS l

REACTOR COOLANT PUMPS

HORSEPOWER 7000

'g DESIGN FLOW (GPM)

DESIGN PRESSURE (PSIG) 100,600 2485

'k DESIGN TEMPERATURE (8F) 650

\ STEAM GENERATORS **

TYPE 2

MODEL F,SHELL AND U-TUBE  :

HEATTRANSFER AREA (ft ) 55,000 l 5626 i NUMBER OF U-TUBES ,

l PRESSURIZER DESIGN PRESSURE (PSIG) 2485 ,

F 880 DESIGNTEMPERATURg)(8 INTERNAL VOLUME (ft

) 1800 TOTAL RCS FLOW (1g 6 lb/h) 142.1 TOTAL RCSVO UME, INCLUDING PRESSURIZER AND 12,462 SURGE LINE (ft ,,

O O O 1

1 4 ,

VEGP STEAM GENERATOR

. AUXILIARY FEEDWATER NOZZLE i a

4s ADDED TO STEAM GENERATORS TO MINIMlZE EXPOSURE TO CRACKING IN MAIN FEEDWATER LINE DUE TO THERMAL FATIGUE INDUCED BY i

iN STRATIFIED FLOW DURING STARTUP/ HOT STANDBY / LOW POWER l OPERATIONS

[

l e MINIMlZE WATER HAMMER POTENTIAL FLOW THROUGH THE MAIN FEEDWATER N0ZZLE IS INITIATED AFTER REACHING ABOUT 18% REACTOR POWER ,

l

VEGP FEEDWATER PIPING (O

STEAM GENERATORg 8" 16"

\ /

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F LOW DISTRIBUTION SAf f Lt Y Y  !

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TuttSMitt SLOWDOWN PIPE & CONNECT 404 DIVIDER PLATE  :

h COOL ANT CM AMet R PRIMAR Y COOLANT fv022LE D '

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i 1

VEGP ,

MAIN STEAM ISOLATION VALVES (MSIVs?

i  !

l

• TWO VALVES IN SERIES ON EACH STEAM LINE, ONE POWERED FROM EACH TRAIN ,

• PRECLUDES BLOWDOWN OF MORE THAN ONE STEAM GENERATOR  !

• EACH VALVE IS PROVIDED WITH AN ELECTRO-HYDRAULIC ACTUATOR D e EACH MSIV IS QUALIFIED FOR OPERATION IN THE POSTULATED MSLB i

( ENVIRONMENT r

• MSIVs CLOSE IN 5 SECONDS  :

• MSIVs ARE TESTABLE AT POWER BY PARTIAL CLOSURE i

)

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i VEGP l MAIN STEAM SYSTEM j' $^g g. ., R- @== TURBINE STOP VALVES lg l t,

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CONTAllIMENT BOUNDARY l

t i

~ PRINCIPAL DESIGN F URES (Cont'd)

REACTOR

! FUEL ASSEMBLY ARRAY 17x17 NUMBER OF FUEL ASSEMBLIES 193 NUMBER OF FULL-LENGTH RODS 53 j ABSORBER MATERIAL HAFNIUM l FUEL ENRICHMENT (WT%)

! REGION 1 2.10

! REGION 2 2.60

! REGION 3 3.10 l NUMBER OF CORE EXIT THERMOCOUPLES 50 NUMBER OF MOVABLE INCORE DETECTORS PATH 58 lQ

! CONTAINMENT

) TYPE STEEL-LINE0, PRESTRESSED, l POST-TENSIONED, CONCRETE CYLINOER, HEMISPHERICAL 00ME ROOF DESIGN PRESSURE (PSIG) 52 3 2,750,000 NOMINAL FREE VOLUME (ft )

CONCRETE THICKNESS (ft)

VERTICAL WALL -

33/4 DOME 33/4

VEGP '

PRINCIPAL DESIGN FEATURES (Cont'd) ,

ENGINEERiO SAFETY FEATURES EMERGENCY CORE COOLING .

ACCUMULATORS NUMBER 4 OPERATING PRESSURE, MINIMUM (PSIG) .

SOS MINIMUM OPERATING WATER VOLUME, 350 EACH (FT3)

CENTRIFUGAL CHARGING PUMPS NUMBER 2 DESIGN FLOW (GPM) 150 DESIGN PRESSURE (PSIG) 2000 SAFETY INJECTION PUMPS 4 s NUMBER DESIGN FLOW (GPM) 2 425 S DESIGN PRESSURE (PSIG) 1750 N RESIDUAL HEAT REMOVAL PUMPS

\ ~

NUMBER 2 DESIGN FLOW (GPM) 3700 DESIGN PRESSURE (PSIG) SOS CONTAINMENT SPRAY SYSTEM CONTAINMENT SPRAY PUMPS .;

NUMBER 2 TYPE HORIZONTAL CENTRIFUGAL DESIGN FLOWRATE (GPM) 2000 l DESIGN PRESSURE (PSIG) 300 l REFUELING WATER STORAGE TANK **

NUMBER 1 NOMINAL VOLUME 738,000 OPERATING PRESSURE ATMOSPHERIC OPERATING TEMPERATURE ,

AMBIENT,588F MIN.  !

O O -

O VEGP REFUELING WATER STORAGE TANK (RWST)  !

e RWST IS REINFORCED CONCRETE STRUCTURE WITH STAINLESS STEEL LINER e RWST AND TANK PENETRATIONS ARE TORNADO MISSILE-PROOF e NOMINAL RWST VOLUME OF 730,000 GALLONS ACCOMMODATES:

h -

OPERATING BAND DELIVERING A MINIMUM OF 300,000 GALLONS TO CONTAINMENT l

SEMI-AUTOMATIC SWITCHOVER VALVE AND PUMP MANIPULATIONS SINGLE FAILURE INSTRUMENT INACCURAClES

. m. ,.

vroe U 2

i PRINCIPAL DESIGN FEATURES (Cont'd)

COOLING WATER SYSTEMS NSCW PUMPS NUMBER S TYPE VERTICAL, CENTRIFUGAL DESIGN FLOW, EACH (GPM) 8800 DESIGN PRESSURE (PSIG) 100 ,

, TRANSFER PUMPS

! NUMBER 2(100% EACH) i . TYPE VERTICAL, CENTRIFUGAL DESIGN FLOW, EACH (GPM) 500 4 DESIGN PRESSURE (PSIG) 50

! NSCWTOWER **

NUMBER 2 PER UNIT FAN DATA QUANTITY (PER TOWER) 4 i SPEED (RPM) 164 l 535,000 j AIRFLOW (CFM/ FAN)

WET BULB TEMPERATURE ('F) 82 .

DRIFT LOSS) PERCENT) 0.01 COMPONENT COOLING WATER PUNS 'l NUMBER $ (50 PERCENT EACH)

TYPE HORIZONTAL CENTRIFUG AL i DESIGN FLOW (GPM, EACH) 5000 DESIGN PRESSURE (PSIG) 70 f

s -

O O ;

n

, VEGP '

NSCW TOWERS j o PLANT ULTIMATE HEAT SINK

• TWO INDEPENDENT TRAINS PER UNIT e ONE TRAIN OPERATES DURING ALL MODES OF OPERATION hb

• WATER VOLUME IN THE BASINS MEETS THE VOLUME REQUIREMENTS OF REGULATORY GUIDE 1.27 1

• EACH TRAIN CONSISTS OF THE FOLLOWING:

DNE FORCED DRAFT COOLING TOWER WITH FOUR CELLS

) ONE COOLING TOWER BASIN (WATER STORAGE SUPPLY) i THREE 5B% CAPACITY PUMPS

.ONE TRANSFER PUMP

i O

VEGP NSCW COOLING TOWERS 1 A AND 1B GENERAL ARRANGEMENT i

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VEGP O

TYPICAL CROSS SECTION OF NSCW TOWER

. n -

p DISCHARGE EL.250'-11' 1 I i w

BOTTOM OF ELIMINATORS T

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e SPRAY HEADER j _ _ f, EL.236'-5'

-- w - - w - - w - - w - - - - -

SPLASH RING E L. 232'4"

/ _ _

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! TOP OF BOTTOM AIR INTAKE OF E L. 230* 3" AIR FILL INTAKE EL. 230'-9' a  %

% EL. 217'-9' HIGH WATER LEVEL m

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VEGP

! PRINCIPAL DESIGN FEATURES (Cont'd) ,

l l-AUXILIARY FEEDWATER SYSTEMS 1

AU.XILIARY FEEDWATER PUMPS .  ;

MOTOR - DRIVEN NUMBER 2 TYPE HORIZONTAL CENTRIFUGAL

h DESIGN FLOW (GPM, EACH) 630

~

j DESIGN PRESSURE (PSIG) 1515 l TURBINE - DRIVEN NUMBER 1 l TYPE HORIZONTAL CENTRIFUGAL DESIGN FLOW (GPM) 1175 DESIGN PRESSURE (PSIG) 1525 i

l 0 -

O -

~

Q

.- VEGP l PRINCIPAL DESIGN FEATURES (Cont'd) . .

l

! NATURAL DRAFT COOLING TOWER QUANTITY -

1 PER UNIT APPROXIMATE HEIGHT (FT) 540 DESIGN WET BULB TEMPERATURE 78 DESIGN RELATIVE HUMIDITY (%) 51 DESIGN INLET TEMPERATURE ('F) 122 DESIGN OUTLET TEMPERATURE (8F) 89 1

4 DESIGN FLOWRATE (GPM) 509,600 A

y TURBINE - GENERATOR MANUFACTURER GENERAL ELECTRIC j

TUR8INE HIGH PRESSURE 1 l LOW PRESSURE 3 OPERATING SPEED (RPM) 1800 GENERATOR GENERATOR RATED OUTPUT (kW) 1,156,622 VOLTAGE (V) 25,000

- . _ _ _ . - _ _ _ . _ . , ,- - . - - - - - - . , , - _ . _ , . . . - . . . - , _ . - - . , . n- --_._..-- .-

0-l _

1 .

VEGP 1 PRINCIPAL DESIGN FEATURES (Cont'd)

! i RA0 WASTE VOLUME REOUCTION SOLIDIFICATION SYSTEM **

VOLUME REDUCTION SYSTEM

VENDOR AEROJET ENERGY CONVERSION COMPANY l A

TECHNOLOGY FLUIDIZED BED COMBUSTION l{6 FLUIDIZED BED CALCINATION ,

SOLIDIFICATION SYSTEM VENDOR STOCK EQUIPMENT COMPANY TECHNOLOGY POLYMER SOLIDIFICATION FOR DRY PRODUCT CEMENT SOLIDIFICATION 8

L F

6' .

O o.

l j -

j VEGP l CIRCUMSTANCES LEADING TO SELECTION j OF VOLUME REDUCTION OPTION i

l e UNCERTAINTY OF BURIAL SITE AVAILABILITY NATIONWIDE IN THE FUTURE i4 e '

"\ DUE TO POSSIBLE BURIAL SITE UNAVAILABILITY VOLUME REDUCT10N CAPABILITY WAS ADDED i

X .

j e MINIMlZE SHIPMENT OF WASTE WITH FREE WATER I

e INCREASED TRANSPORTATION AND BURIAL COSTS ASSOCIATED WITH GREATER VOLUMES OF WASTE l IF NOT VOLUME REDUCED i

i i

I l

., T, i

V0GTLE ELECTRIC GENERATING PLANT PROJECT PERCENT COWLETE 85-1 CONSTRUCTION SCHEDULE PERIOD ENDING 07/21/85 ACTUAL e

UNIT 1 & COM EN APPROXIE TELY 82 UNIT 2 & COM WN APPROXIMRTELY 48 O

l 3

- = = _ _ .- . _ - - - - _ - - - - __

- - ~ . .

k.

MILESTONE SCHEDULE COMPLETED MILESTONES MILESTONE SCHEDULED COMPLETED ENERGIZATION MAR 1985 FEB 1985 RV AVAIL FOR FLUSH MAY 1985 MAY 1985

. SCHEDULED MILESTONES

' MILESTONE SCHEDULED SECONDARY HYDRO DEC 1985 TURBINE ON TURNING GEAR DEC 1985 PRIMARY HYDR 0 JAN 1986 INITIAL CONDENSER VACUUM FEB 1986 HOT FUNCTIONAL MAY .986 ILRT JUNE 1986 FUEL LOAD DEC 1986 COMMERCIAL OPERATION JUNE 1987 f

v .

1. dd ._ _ _ _ _ . __

APPENDIX VI - NRC REGIONAL ~

EVALUATION OF CONSTRUCTION .

1 9

3 i

4  :

1

{ ,

PRESENTATION TO ACRS i ON V0GTLE 1 OPERATING LICENSE  !

!. i AUGUST 9, 1985

• REGIONAL EVALUATION OF CONSTRUCTION -

?

1' i,. -

+

i.

1 t

i i

i t

I By: M. V. Sinkule Projects Section Chief l

1 Region II, NRC '

l.

I 1 j i

i i

1

]

1 1

l l

1 1

i-m ', L 4 6,4-,%4, m. ae , moven,_ _

A47 _ - - - - - - - -w 1

,. .. . , _ _ . . , . _ , . . _ - . . _ . . _ . . _ - _ . . . _ _ _ _ _ . _ _ _ _ . - - . . . . - . . . . . . _ _ = -

I s

I

-i l- l 1

f l

i i i -

L l 4  :

t l- OUTLINE *

< i i

• 1 I

. l

! INTRODUCTION i

. t i  !

REGIONAL EVALUATION OF CONSTRUCTION I i

I HISTORY OF NRC INSPECTIONS ,

4

}

t l- OPEN CONSTRUCTION ITEMS f f

{-  !

SYSTEMATIC ASSESSMENT OF LICENSEE PERFORMANCE  !

I t i

CONCLUSIONS c ,

e i r t I

i t i l i i

h l- i r .:

1  !

i. l t

i 2 e f h i i

}

s I

!, - I

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

]

l l

l l

O f

INSPECTION PROGRAM REVIEW 0F SAFETY-RELATED ACTIVITIES -

REVIEW CONSISTS OF O

PROGRAM / PROCEDURES TO CONTROL ACTIVITY (PROCEDURE REVIEW) .

VERIFICATION THAT PROGRAMS / PROCEDURES HAVE BEEN IMPLEMENTED (OBSERVATION ) '

I VERIFICATION THAT PROGRAM ACTIVITY HAS BEEN DOCUMENTED (RECORDS REVIEW)

RESIDENTS SPECIALISTS SPECIAL TEAM INSPECTIONS f

O .

1. 49 __

..~,, -..-

d .4.

L .

i 1  :

{ STAFF HOURS EXPENDED ON V0GTLE l

YEAR NO. OF INSPECTIONS

. NRC STAFF HOURS  ;

4 1975 1976 2 l 9 i

! 1977 5 164 1

1978 10 262 I '

4 1979 19 281 .

j 1980 16 178 a

- 1981 15 i

530 i 1982 29 1246 t

! 1983 24 1654 1 1984 37 1578 +

i 1985 23 1828 4

i

!- V0GTLE 1 TOTAL - 8010 '

i V0GTLE 2 TOTAL - 4522 i  ;

j V0GTLE SITE 12,532 HRS.

I ,

i i ,

i. i r

i .

4 4

i 5 I

h.

---nnr nn. -- _ . - ~ , . - -.

C 70 - - - , - - - -

O ENFORCEMENT HISTORY SEVERITY SEVERITY DEVIATION LEVEL- LEVEL FROM YEAR 4 5 COMMITMENT 1977 INF e

1978 DEF 1979 7 1 1 1980 2 1 1981 16 1982 3 4 1 1953 4 5 .

1934 11 8 1985 12 4 40 40 2 V0GTLE 1 TOTAL = 82 s

O N-1/

- - - ~ -  !

3 1

t t

4 i '

i OPEN TIMES I i >

J.

=,

" f

+

4

)

4 REGIONAL TRACKING SYSTEM ,

i 1

l

i. - I 4

CONSTRUCTION DEFICIENCIES .

k 'f I

< - i i

NON-COMPLIANCE ITEMS 1

l. ' -

{

OTHER OPEN ITEMS  !

t 4 .

< UNRESOLVED ITEMS  !

l 3 '

i

• INSPECTOR FOLLOW UP ITEMS j-  !

ALLEGATIONS  !

IE BULLETINS f i

f 1

1

+

P 1

1 I

f i

h.

\ j i

1 1

i ,

l i

eo >

t

E - . .

() SYSTEMATIC ASSESSMENT OF LICENSEE PERFORMANCE APRIL 1, 1979 - AUGUST 31, 1980 GPC RESPONSIVE TO NRC FINDINGS HEALTHY ATTITUDE TOWARD NRC REGULATIONS PROMPT ACTION ON CORRECTING SIGNIFICANT ITEMS NO CHANGE IN RECOMMENDED INSPECTION EFFORT JULY 1, 1980 - JUNE 30, 1981 NO STRONG OR WEAK AREAS IDENTIFIED v NO CHANGE RECOMMENDEO IN INSPECTION EFFORT JULY 1, 1981 - OCTOBER 31, 1982 MAJOR STRENGTHS - CONTAINMENT AND OTHER SAFETY RELATED STRUCTURES

- LICENSING ACTIVITIES NON-COMPLIANCE ITEMS WERE NOT INDICATIVE OF PROGRAMMATIC BREAK DOWN NOVEMBER 1, 1982 - OCTOBER 31, 1983 LICENSEE CONTINUES TO IMPLEMENT A VIGOROUS CONSTRUCTION PROJECT MANAGEMENT EFFORT WITH WELL QUALIFIE0 AND EXPERIENCED PERSONNEL MAJOR STRENGTHS - QA PROGRAM

- SAFETY RELATE 0 COMPONENTS NON-COMPLIANCE ITEMS WERE NOT INDICATIVE OF PROGRAM BREAKDOWN V 71

CONCLUSIONS REGION II CONSTRUCTION INSPECTION AND ENFORCEMENT PROGRAMS ARE ON SCHEDULE REGION II RESOURCES AVAILABLE TO COMPLETE THE INSPECTION AND ENFORCEMENT PROGRAM e -

AS WE PERCEIVE THE SITUATION TODAY, REGION II FOLLOW UP OF OPEN ITEMS WILL NOT DELAY ISSUANCE OF AN OPERATING LICENSE LICENSEE MANAGEMENT 'IS RESPONSIVE TO NRC FINDINGS LICENSEE PERFORMS THOROUGH EVALUATIONS, TAKES PROPER CORRECTIVE ACTION, AND PROPERLY REPORTS DEFICIENCIES THAT ARE IDENTIFIED LICENSEE MANAGEMENT HAS IMPLEMENTED ADEQUATE PROGRAMS TO ENSURE PLANT IS DESIGNED AND CONSTRUCTED IN A QUALITY MANNER n

V -

BASED ON THE INSPECTION AND ENFORCEMENT ACTIVITIES TO DATE, THE REGIONAL STAFF BELIEVES AN OPERATING LICENSE RECOMMENDATION WILL BE FORTH COMING FROM REGION 11 1

O w

APPENDIX VII - NRC LICENSING REVIEW OF PLANT V0GTLE O -

NRR STAFF PRESENTATION TO THE i

, ACRS i l

! l

SUBJECT:

- ~

Vogtle Electric Generating Plant. Units 1 and 2 DATE: August 9, 1985

{ l i , PRESENTER: Melanie A. Miller

. PRESENTER'S TITLE / BRANCH /DIV: Project Manager Licensing Branch flo. 4 Division of Licensing PRESENTER'S NRC TEL. NO.: 492-4259 SUBCOMMITTEE:

p ... . . . . ..... . . ...... .

i O LICENSING OVERVIEW i

1

CONSTRUCTION PERMIT ISSUED JUNE 28, 1974 i

FSAR DOCKETED SEPTEMBER 16, 1983 i

ENVIRONMEllTAL REPORT DOCKETED NOVEMBER 28, 1983

! DES ISSUED OCTOBER 1984 .

i- FES ISSUED MARCH 1985 l.

SER ISSUED JUNE 1985 l

14 OPEN ITEMS ,

l 50 CONFIRMATORY ITEMS 11 LICENSE CONDITIONS i i ,

!O 4

//~74

v V0GTLE OPEN ITEMS (1) EQUIPMENT QUALIFICATION (2) PRESERVICE INSPECTION PROGRAM (3) CONTAINMENT SUMP (4) T0XIC GAS EVALUATION OF CHEMICALS (5) GENERIC LETTER 83-28

~

( 6 )- EMERGENCY RESPONSE CAPABILITY -- RG 1.97, REV. 2 (7) FIRE PROTECTION ITEMS (8) SAFE AND ALTERNATE SHUTDOWN CAPABILITY (3) TRAINING Or EMCROENCY DIESCL GENERATOR PCRSONNEL (10) DIESEL FUEL OIL STORAGE TANK CATHODIC PROTECTION (11) LICENSEE QUALIFICTAIONS FOR OPERATION (12) RETESTING CT SIMULATOR RCSiONS (NUREC-0737, ! TEM .A.2.1)

(13) EMERGENCY PREPAREDNESS (14) HUMAN FACTORS ENGINEERING ITEMS O e

1. 77

+ , . . . .

' 't .

OPEN ITEM 1 EQUIPMENT QUALIFICATION ENVIRONMENTAL EQUIPMENT QUALIFICATION SEISMIC EQUIPMENT QUALIFICATION PUMP AND VALVE OPERABILITY PRE-AUDIT MEETING FOR SEISMIC AND PUMP.AND VALVE AREAS SCHEDULED FOR EARLY SEPTEMBER O

TYPICAL OPEN ITEM AT SER ISSUANCE FOR ALL PLANTS DUE TO THE TIMING OF EQUIPMENT INSTAlt.ATION AND THE AUDITS e

TENTATIVE AUDIT SCHEDULE:

ENVIRONMENTAL -- JANUARY 1986 SEISMIC AND PUMP AND VALVE -- OCTOBER 1985 O

or

c O

.NJ OPEN ITEM 2 PRESERVICE INSPECTION PROGRAM ISSUES CAST STAINLESS STEEL WELD ULTRASONIC EXAMINATION PROCEDURE STAFF REVIEW OF METHOD OF COMPLIANCE WITH RG 1.150 SUBMITTAL AND EVALUATION OF RELIEF REQUESTS O

STATUS 1

, STAFF PRESENTLY REVIEWING WELD EXAMINATION PROCEDURE AND RG 1.150 COMPLIANCE l

i APPLICANT PLANS TO SUBMIT RELIEF REQUESTS BY MARCH 1986 FOR STAFF REVIEW t

nn

1 O

OPEN ITEM 3 CONTAINMENT SUMP ISSUE: THE APPLICANT HAS NOT MET THE GUIDELINES OF RG 1,82, REVISION 0, NOR SHOWN THAT UNACCEPTABLE BLOCKAGE CONDITIONS WOULD NOT OCCUR RESOLUTION: THE APPLICANT HAS REANALYZED THE DEBRIS BLOCKAGE POTENTIAL AND POSSIBLE IMPACT ON NPSH MARGINS IN THE POST-LOCA PERIOD O THE STAFF AND APPLICANT MET TO DISCUSS THIS REANALYSIS ON JULY 31 THE APPLICANT'S ANALYSIS APPEARS TO BE APPROPRIATELY CONSERVATIVE A SUBMITTAL FROM THE APPLICANT IS ANTICIPATED IN MID-AUGUST 1985 O -

. to

O OPEN ITEM 4 T0XIC GAS EVALUATION OF CHEMICALS ISSUE: THE APPLICANT HAS NOT COMPLIED WITH THE GUIDELINES

. . . OF RG 1.95 NOR PROVIDED ADEQUATE JUSTIFICATION FOR NOT DOING S0.

~

STATUS: THE STAFF AND APPLICANT MET ON JULY 31 TO DISCUSS THIS ISSUE.

THE STAFF AND APPLICANT DISCUSSED AND AGREED UPON THE APPLICANT'S APPROACH, THE APPLICANT PLANS TO SUBMIT A REANALYSIS IN EARLY SEPTEMBER.

THE STAFF AND APPLICANT WILL CONTINUE TO DISCUSS THIS ISSUE.

O

&//

O OPEN ITEM 5 GENERIC LETTER 83-28 ISSUE: THE APPLICANT OWES THE STAFF INFORMATION ON 6 ITEMS IN THE GENERIC LETTER INCLUDING THE 13 PLANT-SPECIFIC QUESTIONS RELATED TO THE AUTO-MATIC REACTOR TRIP USING SHUNT COIL TRIP ATTACHMENT, THE STAFF HAS THE APPLICANT'S RESPONSES ON 4 ITEMS UNDER REVIEW.

O RESOLUTION: THE STAFF WILL CONTINUE TO DISCUSS THIS ISSUE WITH THE APPLICANT,

~

O k-/r

! OPEN ITEM 6 EMERGENCY RESPONSE CAPABILIT'i

RG 1.97, REVISION 2

-)

ISSUE i

r .. THE APPLICANT HAS NOT FULLY COMPLIED WITH RG 1.97, i REVISION 2, NOR PROVIDED ADEQUATE JUSTIFICATION FOR '

NOT DOING S0.

STATUS j THE APPLICANT HAS PROVIDED ADDITIONAL INFORMATION i BY LETTER DATED JUNE 20, 1985, WHICH IS CURRENTLY UNDER STAFF REVIEW, i - .

i lO

,k

O OPEN ITEM 7 FIRE PROTECTION ITEMS ISSUES FIRE D0 ORS AND DAMPERS

. THE APPLICANT HAS NOT JUSTIFIED THE USE OF SPECIAL-PURPOSE DOORS WHEN USED IN RATED FIRE BARRIERS.

THE APPLICANT HAS NOT PROVIDED JUSTIFICATION FOR OVERSIZE FIRE DAMPERS USED FOR HVAC PENETRATIONS OF FIRE BARRIERS O

POWER SUPPLIES FOR VENTILATION THE APPLICANT NEEDS TO DEMONSTRATE THAT BOTH TRAINS OF VENTILATION FOR SAFETY-RELATED AREAS WILL NOT BE DISABLED BY A SINGLE FIRE STATUS THE APPLICANT PLANS TO SUBMIT INFORMATION IN LATE SEPTEMBER O

ser .

O OPEN ITEM 8 SAFE AND ALTERNATE SHUTDOWN CAPABILITY ISSUE: THE APPLICANT NEEDS TO DEMONSTRATE COMPLIANCE WITH

- GDC 3 BY MEETING BRANCH TECHNICAL POSITION 9.5-1, POSITIONS C,5.s AND C.S c STATUS: THE APPLICANT PLANS TO SUBMIT ITS ANALYSIS IN LATE SEPTEMBER 1985

. M -

i l

^

( R-/f

(

() OPEN ITEM 10 DIESEL FUEL OIL STORAGE TANK CATHODIC PROTECTION ISSUE: APPLICANT DOES NOT PROPOSE TO PROVIDE CATHODIC PROTECTION FOR BURIED FUEL OIL STORAGE TANKS AND ASSOCIATED PIPING

~

THE APPLICANT HAS STATED THAT CORROSION BY ELECTROLYSIS WILL NOT OCCUR

_. BECAUSE OF THE S0ll CONDITIONS AND WATER TABLE AT V0GTLE O

• C0AL TAR EP0XY COATING WILL BE ADEQUATE PROTECTION RESOLUTION: THE STAFF REQUESTED THE APPLICANT TO PROVIDE SUPPORTING DATA. THE APPLICANT RECENTLY COMPLETED A TESTING PROGRAM, STATUS: THE APPLICANT PLANS TO SUBMIT ITS REPORT IN AUGUST 1985.

O

. nn

1 OPEN ITEM 11 LICENSEE QUALIFICATIONS FOR OPERATIONS ISSUES PLANT REVIEW BOARD INVESTIGATION OF VIOLATION OF .

TECHNICAL SPECIFICATIONS

~

~ *

.. CORPORATE REPORTING LEVEL FOR INDEPENDENT SAFETY ENGINEERING GROUP T00 LOW INDEPENDENT REVIEW BY QUALIFIED PERSONNEL OF PROCEDURES AND PROGRAMS THAT AFFECT NUCLEAR SAFETY O

• VISIT TO CORPORATE OFFICE RESOLUJION: THE STAFF IS AWAITING DOCUMENTATION TO RESOLVE THE FIRST 3 ITEMS.

THE CORPORATE OFFICE VISIT IS TENTATIVELY PLANNED FOR THE FALL OF 1985, O

1. 7

c O

OPEN ITEM 13 EMERGENCY PREPAREDNESS ISSUE THE APPLICANT HAS NOT FULLY COMPLIED WITH THE REQUIREMENTS OF 10 CFR 50.47 AND APPENDIX E OF

_. 10 CFR 50.

STATUS O

THE STAFF IS REVIEWING THE LATEST REVISION TO THE V0GTLE EMERGENCY PLAN.

THE APPLICANT IS REVIEWING STAFF COMMENTS ON ITS EMERGENCY PLAN.

9 O

vir

O OPEN ITEM 14 HUMAN FACTORS ENGINEERING ITEMS .

• DETAILED CONTROL ROOM DESIGN REVIEW

• IN-PROGRESS AUDIT CONDUCTED JULY 10-12, 1985

SUMMARY

REPORT FROM APPLICANT SCHEDULED MARCH 1986

• SAFETY PARAMETER DISPLAY SYSTEM

SUMMARY

REPORT FROM APPLICANT SCHEDULED SEPTEMBER 1985 e

O on

APPENDIX VIII - V0GTLE PROJECT ORGANIZATION l PROJECT ORGANIZATION i ACRS C0fftITTEE EETING l

AU6UST 9, 1985 t

+

, c 4

I VICEPRESIDENTbDhhh6ENERALMANAGER GEORGIA POWER COPFANY f'9A - --_ .

m a m n' -

ks V0 GILE PR0KCT ORGMIZATION VOGTLE ELECTRIC GENERATING PLANT UNITS 1 AND 2

........................1

.- ..3. .

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I

APPENDIX IX - MANAGEMENT PHILOSOPHY AT GEORGIA POWER COMPANY MANAGEMENT PHILOSOPHY ACRS WASHINGTON, D.C.

AUGUST 9, 1985 O J. T. BECKHAM, JR.

VICE-PRESIDENT AND GENERAL MANAGER NUCLEAR OPERATIONS GEORGIA POWER COMPANY l

l

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p c o p ao n e o o me o v o

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L Management Philosophy (continued) i F. To cany out a quality assurance program for the purpose of reviewing l and inspecting plant construction and operation.

1 I G. To keep personnel radiation exposures as low as practical and below I 1-NRC limits. ,

H. To maintain an adequate staff of nuclear plant personnel who are well l

. % qualified in their r==-+2.';;

jobs.

\

j' % L To provide ongoing training for plant personnel in their job functions, in h3 safety, in health physics, and in other joineisted matters as necessary.  !

J. lb keep nuclear generating plant discharges to the environment as low i j as practical and below regulatory limits.

i K.1b cooperate with the institute of Nuclear Power Operations (INPO) in i- order to achieve the highest standards of excellence for construction '

L and operation of nuclear generating plants.

4 1

i 4

i

i Ten Rules of Plant Operation '

1. procedure.

Know the procedure system well enough to be sure you are using the right

2. Follow procedures.

3. In an undefined condition, place equipment in a safe condition per l: Procedure.

!. 4. Remember that in an emergency a licensed SRO can change the intent of

{! a procedure and to protect to theprevent core. injuries, to prevent excessive equipment damage, L  % 5. Uphold your individual responsibilities under a# circumstances as defined by procedure.

6. Develop a questioning attitude but execute procedures correctly.

[

! 7. Place asfety in every circumstance above continuity of operation.

8. If it's not c' overed by a procedure, let safety dictate the action. Also see #1.

j

9. When you think of better ways to do things initiate procedure changes.

[ 10. Make adherence to procedures a way of life.

i i

I i n 2

i

~ THE BIG PICTURE 4

DESIGN PROCESS AND CONSTRUCT i

h OPERATE PROCESS

!  % F e: g 4

E

! E i D MONITOR PROCESS B A

C

K l

l CORRECT PROCESS .

l

APPENDIX X - GEORGIA POWER CORPORATE ORGANIZATION

.g \

l -

! CORPORATE ORGANIZATION t

1 ACRS WASHINGTON, D.C.  ;

1 AUGUST 9, 1985 i

4 l J. T. BECKHAM, JR.  ;

l VICE-PRESIDENT AND GENERAL MANAGER

~

NUCLEAR OPERATIONS r

i I

l t

l GEORGIA POWER COMPANY t

}

• f a

1 e

P97. . .

V General Office Organization Chairman of the Board t

and 5 Chief Executive Officer R. W. Scherer President

% J. H. Miller Jr.

Executive Vice Executive Senior Executive President and Vice President Vice President Vice President General Manager Power Supply Division Operations Finance H. G. Baker Jr. R. J. Kelly J. A. Gantt Jr. W. Y. Jobe

(s

,: e, ,

i-

}

Power Supply i

1 Executive I Vice President 4

R. J. Kelly 1-

k Senior Vice Senior Vice Senior Vice l

\

President President P 1d "' Nuclear Operations Fossil & Hydro Power l .

R. 9n ay J. P. O,Reilly G.F. Head i

4 Vice President Vice President &

Engineering & General Manager Construction Services Quality Assurance R. H. Pinson P. D. Rice l

a O

i:

Nuclear Operations i General Office Staffing r .

4 Sr.V.P.

Nuclear Oper.

J. P. O'Reilly i

~

Sr.V. Exec.Sec.

L Beazley N (Vacant) ,

4 i

Manager Manager V.P. & Gen. Mgr. Mgr. Nuclear Engr.

Managw Nuclear i

i Nuclear Training Nuclear Support Nuclear Oper. & Chief Nuc. Engr. P r ormance j J. J. Badgett T. McHenry J. T. Beckham L. T. Gucwa M. y i

i i

l 1

i

_A s Q\

\ "

~J

'" Nuclear Operations o,ng,,i,'ug,, General Office Staffing Nuclear Oper.

Asst Sr.Sec.

I I Gen. Manager Gen. Manager HATCH VOGTLE OPERATIONS OPERATIONS l l l l Mgr. Mgr. Maint. Mgr. Project s' Engineering & Modification Engineering Red ical N Liaison Safety O

N Sec. H Sec. M Sec. Il Sec. Il l I I I Hatch Engr. Proj. Mgr. Radiological Nuc. Emer.

Sr. Nuc. Engr. Engr.Ser.1 Safety Plan Supv.

Liaison Vogtle Engr. g,,g,g, (SCS) LI n 3 ,p ,,,g,,,

Secretary planning) St. Nuc. Engr. Sr. Reg. Spec.

( )

Engr. (SCS) 1. Health Phy.

(SCS) Sr. Nuc. Engr. 2.

SC Engr. (SCS) (maint. & Sr. Nuclear g,y 3"Nuc. Engr. Engr. S[upv St. Engr. Nuclear Nuclear Engr.

Assoc. Engineer

O n

[P;, '

Safety Review Board I

i

{ Executive i Vice President Power Supply Sr. Vice President Hatch SRB Vogtle SRB O at ns

', bg Chairman Chairman N Vice-Chairman Vice-Chairman g and and x 7 members Vice President 7 members

} y & General Manager Nuclear

[ Operations fI General Manager General Manager Nuclear Nuclear .

Operations Operations l Hatch Vogtle

APPENDIX XI - V0GTLE PLANT OPERATION ORGANIZATION AND TRAINING i

1 PLANT OPERATION ORGANIZATION

. AND TRAINING  !

1 I

N 4

q

ACRS i WASHINGTON, D.C.

AUGUST 9, 1985  ;

1

( .- .

! . GEORGE BOCKHOLD GENERAL MANAGER .

I V0GTLE NUCLEAR OPERATIONS  !

l t

GEORGIA POWER COMPANY  !

l l  :

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QUALIFICATIONS '

KEY MANAGERS AND DEPARTMENT HEADS (12)

• ALL WITH B.S. DEGREES

• 5 WITH ADVANCED DEGREES

• 108 YEARS COMBINED NUCLEAR EXPERIENCE CURRENT TOTAL-DEPARTMENT STAFFING NUCLEAR EXPERIENCE MANAGERS 3 39 REGULATORY COMPLIANCE 15 70 ENGINEERING 68 279 MAINTENANCE 159 314

~

~

HEALTH PHYSICS 12~ 70 CHEMISTRY 12 39 QUALITY CONTROL 20 100

- TRAINING 27 134 OPERATIONS SUPT 1 10 OPERATIONS SUPERVISORS 7 80 OPERATIONS ENGINEERS 5 24 l SHIFT SUPERVISORS 16 74 l SHIFT TECHNICAL ADVISORS 11 62 REACTOR OPERATORS 57 138 I

OPERATORS 19 18 TOTAL 432 1A51

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1 EDUCATION OF NUCLEAR OPERATIONS PERSONNEL DEPARTMENT BACHELORS DEGREES ADVANCED DEGREES ENGINEERING 68 7 OPERATIONS 42 2 REGULATORY COMPLIANCE 12 1 HEALTH PHYSICS / CHEM. 18 6 MAINTENANCE 12 -

QUALITY CONTROL 3 -

TRAINING 15 5 MANAGERS 3 -

TOTAL 173 21 OF THE 173 BACHELOR'S DEGREES, 127 ARE ENGINEERING DEGREES AND 32 MORE ARE TECHNICAL OR SCIENCE RELATED.

OF THE 21 ADVANCED DEGREES, 4 ARE DOCTORATES, 17 ARE ENGINEERING OR TECHNICAL RELATED.

(

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OPERATIONS DEPARTMENT QUALIFICATION HIGHLIGHTS STAFF

SUMMARY

SUPERVISORS (24) 10 with previous PWR licensed experience 4 with previous BWR licensed experience 13 with > 6 months hot participation experience 15 with NRC-SR0 instructor certification VEGP 18 hold B.S. degrees in Engineering or related science SHIFT TECHNICAL ADVISORS (11) 6 with BWR STA experience (Plant Hatch) 3 with > 6 months hot participation experience

~

3 with NRC-SR0 instructor certification on VEGP All hold B.S. degrees in Engineering or related science REACTOR OPERATORS (57)

I with previous BWR licensed experience 21 with >3 months hot participation experience ON-SHIFT EXPERIENCE PROGRAM Industry Working Group report to NRC Commissioner by J. H. Miller Jr. (Feb.1984) and resultant generic-letter 84-16 Extensive hot participation experience program (Farley, V.C. Sunmer and Sequoyah)

( \.

On-shift experience at VEGP during Initial Test Program R'/o '1 ._

MAINTENANCE DEPARTMENT EXPERIENCE

( HIGHLIGHTS STAFF

SUMMARY

SUPERINTENDENT (1)

B.S. Electrical Engineering '

10 years nuclear experience Participated in construction / testing / start-up of Hatch SUPERVISION (32) 138 years nuclear experience 348 years maintenance related experience MAINTENANCE ENGINEERING (8)

All with B.S. degrees _in Engineering

( -

20 years nuclear experience -

MECHANICS / ELECTRICIANS (79) 82 years nuclear experience Over 500 years of power plant maintenance experience All are Journeyman INSTRUMENT & CONTROL' TECHNICIANS (35)

All have A.A. degree or 2 year technical degree 47 years nuclear experience RW

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• Training And Qualification

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GE NE RAL MANAGE R OUALITY ASSUR ANCE DEPUTV GE NE RAL MANAGER DUALITV ASSUR ANCE I

I OA SPECI AL PROMCTS SR ASSISSI AN I SECRf f ARV 1 I I OA VOGit t H A TCH g gGig, g RegG O A M AN AGE R QA MANAGE R SUPPOR T MANAGEst x

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! QA Management. . . . . . . . . Top Seven Managers Have Over 120 Years '

i Nuclear Experience Vogtle OA Staff. . . . . . . . . 23 Technical - 60% With 4 Year Degree Vogtle QA Manegor. . . . . . .Over 15 Years Vogtle And Hatch Nuclear QA Experience l4

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Ib Vogtle QA Site Manager. . .Over 11 Years Vogtle And Hatch Nuclear (Construction) QA Experience Vogtle QA Site Manager. . .Over 15 Years Vogtle And Hatch Nuclear (Operations) QA Experience (10 Years Operations QA) l Vogtle QA Staff. . . . . . . . . .Over 140 Years Vogtle And Hatch Nuclear Experience OA/QC Experience

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TRAINING PROGRAM. ~j

• Formal 5 Phase Program 4

• 6-9 Months On The Job Training 3

• Written Examinations '

• Oral Board Examinations ,

• Special Training

d O O  ;

i PREVENTION ORIENTED QUALITY ASSURANCE

• Procedures Development

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• Contractor Programs k

• Operations Programs l.

• Lessons Learned

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O O O QUALITY CONCERN PROGRAM ACRS COMMITTEE MEETING L

% AUGUST 9,1985  :

L9 i%

x P.D. RICE es j VICE PRESIDENT & GENERAL MANAGER, E!

QUALITY ASSURANCE is 37 8

4 GEORGIA POWER COMPANY 51 a1 3

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t I VOGTLE QUALITY CONCERN PROGRAM

• History jd

• Organization l4 iL lt

• Objectives .

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• Program i

• Results 1 1

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VOGTLE QUALITY CONCERN PROGRAM 1

ORGANIZATION VICE PRESIDENT &

VICE PRESIDENT &

GENERAL MANAGER PROJECT GENERAL MANAGER QUALITY ASSURANCE

% P.D. RICE D.O. FOSTER QCP SCREENING I

COMMITTEE

!W '

i QUALITY CONCERN . OCP MANAGER , STEERING i-i' COMMITTEE L.B. GLENN QCP INVESTIGATION QCP ACTIVITIES CORPORATE REVIEW QCP STAFF COMMITTEE b . - _ _ - _ _

VOGTLE QUALITY CONCERN PROGRAM OBJECTIVES '

4 1

• Identify Concerns x -

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• Provide Feedback s

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VOGTLE QUALITY CONCERN PROGRAM i

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l RECEIVE CONCERN .

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READINESS REVIEW PROGRAM i

ACRS COMMITTEE MEETING AUGUST 9,1985 4

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GEORGIA POWER COMPANY  !

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READINESS REVIEW PROGRAM i

• ORGANIZATION .

READINESS REVIEW PROGRAM PROCESS ik

• RESULTS OF SELF-ASSESSMENT ACTIVITIES RN

• LESSONS LEARNED M

• CONCLUSION i

I

i l VOGTLE READINESS REVIEW OBJECTIVES

• Provide Georgia Power Company Added Assurance That Vogtle Facilities And Staff Are Ready For 4 Operation

k l

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• Provide A Systematic And Interactive Mechanism For The Nuclear Regulatory Commission's Independent Review, inspection And Acceptance Of Vogtle Work On A Phased And Scheduled Basis

p. A h

- U READINESS REVIEW ORGANIZATION VOGTLE PROJECT MANAGEMENT BOARD R.W. SCHERER-CHAIRMAN E

V.P. AND P.G.M.

VOGTLE I -

READINESS REVIEW BOARD READINESS REVIEW  !

% PROGRAM MGR. TECHNICAL EXPERTS QUALITY INDEPENDENT ASSURANCE REVIEW OUP Stone & Webster DESIGN CONSTRUCTION OPERATIONS GENERAL DISCIPLINE MGR. DISCIPLINE MGR. DISCIPLINE MGR. APPENDICES MGR.

0 0 D I

f READINESS REVIEW PROCESS SCOPE OF READINESS REVIEW MODULES SELF-ASSESSMENT PROBLEM IDENTIFICATION AND CORRECTION Q

4

• MODULE PREPARATION o t GEORGIA POWER COMPANY MANAGEMENT REVIEW AND APPROVAL '

NRC REVIEW AND ACCEPTANCE

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PLANT VOGTLE GENERIC FUNCTIONS

• civil

• MECHANICAL 4

M .

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• ELECTRICAL N i

• READINESS FOR OPERATION l

• A m O' O O MODULES i

civil ELECTRICAL CONCRETE /REBAR/CADWELDS RACEWAYS l STRUCTURAL STEEL / EMBEDS / WELDING CABLES / TERMINATIONS BACKFILL / COATINGS / POST TENSIONING EQUIPMENT HANGERS / SUPPORTS -

MECHANICAL READINESS FOR OPERATION NSSS INITIAL TEST PROGRAM PIPE HANGERS / SUPPORTS OPERATIONS ORGAN. & ADMIN. i

~

PIPING / VALVES / PUMPS OPERATIONS TRAINING & QUAL.

HVAC/ FIRE PROTECTION PLANT OPERATIONS ,

INSTRUMENTATION & CONTROLS RAD. PROTECTION & CHEMISTRY EMERGENCY PREPAREDNESS OPERATIONS TECHNICAL SUPPORT PLANT MAINTENANCE ,

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{! GENERAL APPENDICES A. ORGANIZATION B. DESIGN CONTROL C. PROCUREMENT l

D. DOCUMENT CONTROL E. MATERIAL CONTROL F. INSPECTOR QUALIFICATION / CERTIFICATION

! G. MEASURING AND TEST EQUIPMENT 4

g H.NONCONFORMANCES

1. PROJECT QUALITY ASSURANCE ORGANIZATION . .

I J. EQUIPMENT QUALIFICATION  !

K. CONSTRUCTION COMPLETION -

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n n SELF-ASSESSMENT IDENTIFY COMMITMENTS DEVELOP VERIFICATION PLANS k

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4 OPERATIONS VERIFICATION PERFORM INDEPENDENT DESIGN REVIEW i

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O O o PROBLEM IDENTIFICATION AND CORRECTION FORMALLY IDENTIFY DEFICIENCIES AND AREAS OF POTENTIAL COLLECTIVE SIGNIFICANCE

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• REPORTABILITY EVALUATION PROJECT EVALUATION PROJECT CORRECTIVE ACTION READINESS REVIEW EVALUATION AND ACCEPTANCE OF ACTIONS -

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i a- 0 O MODULE PREPARATION

INTRODUCTION .

ORGANIZATION AND DIVISION OF RESPONSIBILITIES

• COMMITMENTS PROGRAM DESCRIPTION

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• AUDITS AND SPECIAL INVESTIGATIONS PROGRAM VERIFICATION INDEPENDENT DESIGN REVIEW ASSESSMENT e

,_ __ ___ ,_ y.__. - -. -r--- - ------- - '**~P'- --~'" ' ' ' ' " ' " " '

MANAGEMENT REVIEW AND ACCEPTANCE READINESS REVIEW BOARD ACTIONS SCOPE OF MODULE VERIFICATION PLANS F!NDINGS AND CORRECTIVE ACTIONS N OVERALL ASSESSMENT W

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• PROJECT MANAGEMENT ACTIONS

• REVIEW

=

ACCEPTANCE SUBMITTAL TO NRC i

~

O O O 1

NRC REVIEW

• NRC REVIEW, INSPECTION AND VERIFICATION l NRC IDENTIFIES TO GPC ANY AREAS OF CONCERN IN -

ACCORDANCE WITH THE CURRENT POLICY AND PROCEDURE FOR ENFORCEMENT ACTIONS (10 CFR 2,

% APPENDIX C)

{ GPC MAKES AN IN-DEPTH INVESTIGATION INTO THE AREA g OF CONCERN, POS01BLY RELATED ITEMS, POTENTIAL PROGRAMMATIC PRC*lLEMS, AND CORRECTS ALL PROBLEMS INCLUDING iHE ROOT CAUSE -

i NRC ACCEPTS THE SCOPE OF WORK COVERED BY THE READINESS REVIEW MODULE i

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i READINESS REVIEW PROGRAM ORGANIZATION READINESS REVIEW PROGRAM PROCESS RESULTS OF SELF-ASSESSMENT ACTIVITIES LESSONS LEARNED f

CONCLUSION

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l EXAMPLES OF READINESS REVIEW FINDINGS PROBLEMS WITH RETRIEVABILITY OF DOCUMENTATION WEAKNESSES IN INITIAL TEST PROGRAM PROCEDURES i DEFICIENT CONTROLS ASSOCIATED WITH FIELD CHANGES S i

$ WEAKNESSES IN CERTAIN DESIGN CALCULATIONS l

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6 0 _

Q MODULE 1 - REINFORCED CONCRETE STRUCTURES READINESS REVIEW FINDING 36 OF APPROXIMATELY 4000 QUALITY CONTROL INSPECTION DOCUMENTS COULD NOT BE RETRIEVED PROJECT RESPONSE

• FURTHER INVESTIGATION LOCATED SOME OF THE MISFILED DOCUMENTS, OTHERS COULD NOT BE LOCATED

• MISSING DOCUMENTS WERE CATEGORIZED AS EITHER PRIMARY (NO BACKUP)

OR SECONDARY (SUPPLEMENTARY TO A PRIMARY DOCUMENT)

PRIMARY DOCUMENTS WERE FOUND FOR ALL MISSING SECONDARY DOCUMENTS, THUS SUBSTANTIATING THE ADEOUACY OF THE WORK k

• D HARDWARE ITEMS ASSOCIATED WITH MISSING PRIMARY DOCUMENTS WERE EVALUATED AND ACCEPTED USING EXISTING ALTERNATE QUALITY DOCUMENTS OR FIELD INSPECTION READINESS REVIEW EVALUATION

• THE SMALL PERCENTAGE OF MISSING RECORDS COMBINED WITH THE PROJECT ENGINEERING EVALUATION LEADS TO THE CONCLUSIC'l THAT THE REINFORCED CONCRETE STRUCTURES ARE ACCEPTABLE

• SUFFICIENT OBJECTIVE EVIDENCE EXIST TO VERIFY ACCEPTABILITY OF THE HARDWARE O

e

n -

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O O O MODULE 4 - MECHANICAL EQUIPMENT AND PIPING READINESS REVIEW FINDING THE MINIMUM SEPARATION BETWEEN HOT PlPE AND RACEWAYS WAS NOT MAINTAINED PROJECT RESPONSE THE PIPING INSTALLATION SPECIFICATION DID NOT GIVE CRITERIA LIMITING THE h SEPARATION DISTANCE BETWEEN HOT PIPE AND RACEWAYS

• THE PIPING SPECIFICATION AND INSTALLATION PROCEDURES WILL BE REVISED TO ADDRESS SEPARATION PREVIOUSLY INSTALLED PIPING WILL BE INSPECTED TO VERIFY CONFORMANCE TO THE NEW CRITERIA FUTURE PIPE INSTALLATION WILL CONFORM TO THE ADDED SPECIFICATION REQUIREMENTS READINESS REVIEW EVALUATION THE CORRECTIVE ACTIONS TAKEN WILL ENSURE PAST AND FUTURE PIPE INSTALLATION IS ACCEPTABLE O

d O O  :

LESSONS LEARNED NEED TO ADJUST LEVEL OF READINESS REVIEW STARTUP EFFORT NEED TO MODIFY SCOPE OF WORK COVERED BY CERTAIN MODULES NEED TO DEVELOP GENERAL APPENDICES NEED TO ADJUST THE INDEPENDENT DESIGN REVIEW PROCESS NEED TO ADJUST PROGRAM ADMINISTRATIVE

! AND DOCUMENTATION PRACTICES s

t I. .

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S S E T N N I F E O E B 0 I S

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APPENDIX XV - NRC PARTICIPATION IN V0GTLE READINESS REVIEW NRC PARTICIPATION IN VCGTLE READINESS REVIEW NRC PARTICIPATION ON A TRIAL BASIS TO DETERMINE IF THE REGULATORY PROCESS CAN BE IMPROVED.

NRC* DECISION TO PARTICIPATE WAS BASED ON AN ANALYSIS OF THE PROS AND CONS OF THE PILOT PROGRAM. (SECY-85-122)

PROS NRC PARTICIPATION REPRESENTS A POSITIVE STEP BY NRC IN ATTEMPTING NEW AND ALTERNATIVE APPROACHES TO THE LICENSING PROCESS.

A SUCCESSFUL PROGRAM WILL DEMONSTRATE PREDICTABILITY AND STABILITY IN THE LICENSING PROCESS.

NRC PARTICIPATION REPRESENTS A PRUDENT C000tITMENT OF RESOURCES EARLY IN THE PROCESS TO MINIMIZE THE RISK OF A LARGER COP 911TMENT OF RESOURCES LATER.

PILOT PROGRAM CAN BE CARRIED OUT WITHIN EXISTING REGULATIONS.

NRC PARTICIPATION DIRECTLY RESPONDS TO CONGRESSIONAL -

REQUESTS.

IMPROVE CONTINUITY OF GPC/NRC COMMUNICATIONS AND DIRECTIONS.

CONS IMPACT ON NRC RESOURCES.

TIMING OF THE PILOT PROGRAM IS NOT OPTIONAL.

9 9

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APPENDIX XVI - UT EXAMINATION OF CAST STAINLESS STEEL PIPING O '

UT EXAMINATION OF CAST STAINLESS STEEL PIPING ACRS FULL COMMITTEE MEETING ..

AUGUST 9, 1985 T. N. EPPS/ -

MANAGER, INSPECTION, TESTING,ANDENGINEERING SOUTHERN COMPANY SERVICES i

4 a

e e

1. - yt

TO) v UT EXAMINATION OF CAST STAINLESS STEEL PIPING o

9 o Reactor coolant piping on Vogtle is centrifugally casc stainless steel with statically cast stainless steel fittings.

o Calibration block from Vogtle piping material.

(I, o Used pitch / catch transducers that would yield effective results. '

o Ultrasonic procedure written for this testing purpose, i

o Demonstrated technique to NRC personnel.

o Conducted ultrasonic examination of 100% of circumferential and branch connection S.S. welds in the Vogtle RCS and evaluated results.

o Examinations demonstrated to NRC inspector and compared favorably with his verification examinations. (See NRC IEE Report No. 50-424/85-25 and 50-425/85-24, page 6).

t

1. 4V7

- . - . . . . . . ~ . . .

. APPENDIX XVII - SAN ONOFRE PLANT OVERVIEW i  :

SAN ONOFRE NUCLEAR GENERATING STATION, UNIT 1

PLANT OVERVIEW 4

1. PLANT HISTORY j II. PLANT CHARACTERISTICS

! III. MAJOR OUTAGES AND BACKFITS IV. UNIT CAPACITY FACTORS f

4 i

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PLANT HISTORY OWNERS:

SOUTHERN CALIFORNIA EDISON - 80%

SAN DIEGO GAS AND ELECTRIC - 20%

OPERATOR:

SOUTHERN CALIFORNIA EDISON LOCATION:

5 MILES SOUTH OF SAN CLEMENTE, CA.

AEC APPLICATION FILED n FEBRUARY 1, 1963 CONSTRUCTION STARTED JULY 15, 1964 PROVISIONAL DPR-13 ISSUED OPERATING LICENSE MARCH 27, 1967 .

COMMERCIAL OPERATION JANUARY 1, 1968 RECORD 218 OPERATION CONSECUTIVE DAYS OF JANUARY 26, 1976' O

1 #

e 4

0

+4f _ - . _

PLANT CHARACTERISTICS LICENSED POWER RATING: 1347 MWT NSSS: 3 LOOP WESTINGHOUSE ARCHITECT / ENGINEER / CONSTRUCTOR: BECHTELPOWERCORPORATf0N

~

FUEL: STAINLESS STEEL CLAD URANIUM DIOXIDE ULTIMATE HEAT SINK: PACIFIC OCEAN EMERGENCY AC POWER SYSTEM: REDUNDANT DIESEL GENERATORS EMERGENCY CORE COOLING SYSTEM: REDUNDANT " LOW PRESSURE" SAFETY INJECTION TRAINS REDUNDANT HLGR PRESSURE CHARGING PUMPS CORE AVERAGE TEMPERATURE: 550 0 p RCS OPERATING PRESSURE: 2050 PSIG O

O so

MAJOR QUTAGES AND BACKFITS STANDRY POWFR ADDITION! 1976 1977 0

INSTALLATION OF REDUNDANT HIGH CAPACITY DIESEL GENERATORS (6 MW)

SPHFRF FNCLOSURF PROJFCT! 1976-1977 0 CONCRETE ENCLOSURE OF CONTAINMENT SPHERE INSTALLED DUE TO REDUCTION IN EXTENT OF EXCLUSION AREA BOUNDARY STFAM GFNFRATOR SLFFVfMG PROJFCT! 1980-1991 0 LEAK LIMITING SLEEVES INSTALLED IN APPROXIMATELY 6500 0F 11,400 TUBES TO MITIGATE EFFECTS OF CORROSION TMI MODfFICATIONS! 1982-1984 , . . , .

.O PASS,. RADIATION MONITORING, HYDROGEN MONITORING AND CONTROL, EMERGENCY RESPONSE FACILITIES, RCS VENTS, PLANT SHIELDING, AUXILIARY FEEDWATER SYSTEM UPGRADES, ETC.

SFISMIC MODIFICATIONS! 1997-1984 0 SYSTEMS UPGRADED TO 0.67G HOUSNER CRITERIA TO ATTAIN AND MAINTAIN A SAFE HOT STANDBY CONDITION O

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AW/

'.-. - . . _ _ _ _ . - _ _ - - _ . . - . . _ - _ - . - _ . _ - . . , _ . . - - __- -__---._,-----__-J.- '

O UNIT CAPACITY FACTORS CUMULATIVE CAPACITY FACTOR PRIOR TO 1980 STEAM GENERATOR OUTAGE: 72% ,

CUMULATIVE CAPACITY FACTOR TO DATE: 53.1 l

CAPACITY FACTOR FROM NOVEMBER 27, 1984 RETURN-TO-SERVICE THROUGH JULY 31, 1985: 81%

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SONGS UNIT 1 CAPACITY FACTORS

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. sttttttttttttn{ st 1968 1969 1970 19 71 1972 1973 1974 1975 1976 1977 1978 1979 1980 1981 1982 1983 1984 1985 Year of Operation a

- APPENDIX XVIII - NRC PRESENTATION ON )

SEP INTEGRATED ASSESSMENT - SAN ONOFRE

.r h

\

NRR STAFF PRESENTATION TO THE  ;

ACRS

SUBJECT:

SEP INTEGRATED ASSESSMENT - SAN ONOFRE NUCLEAR GENERATING STATION UNIT 1

• DATE: AUGUST 9, 1985 PRESENTER: E. MCKENNA PRESENTER'S TITLE /DIV: SENIOR PROJECT MANAGER DIVISION OF LICENSING PRESENTER'S NRC TEL. NO.: 49-27468 O

k O ,

() SYSTEMATIC EVALUATION PROGRAM SAN ON0FRE UNIT 1 SEP

SUMMARY

TOPICS DELETED COMPLIANCE WITH CURRENT CRITERIA INTEGRATED ASSESSMENT ISSUES APPLICATION OF PRA SEISMIC UPGRADE STATUS INTEGRATED ASSESMENT RESULTS FURTHER EVALUATION REQUIRED ISSUES OF DISAGREEMENT HARDWARE MODIFICATIONS O

PROCEDURAL OR TECH SPEC CHANGES NO ACTION REQUIRED CONCLUSION O

g ytir'  !

1 _ _ . - . - - - - _ __ . .

_ . . _ J

SilMMARY PHASE II TOPICS - 137 GENERIC TOPICS DELETED 24  !

PLANT SPECIFIC TOPICS DELETED 24 TOPICS REVIEWED 89 TOPICS ACCEPTABLE 53 INTEGRATED ASSESSMENT TOPICS 36 ISSUES - TOTAL 86 I

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DISPOSITION OF ISSUES e

CATEGORY

NUMBER OF ISSUES NO ACTION 4]

HARDWARE 5

PROCEDURES OR TECHNICAL SPECIFICATIONS 12

FURTHER EVALUATION 26 i - -

U N R E S O L V.E. D

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, l GEWEF.IC TOPICS DELETED ap # m1, UH , or Topic No. SEP Title SEP No. TMI, U51, or SEP Title rh

[ 11 2.8 Onsite Meteorological Measurements TM1 !!.F.3 Instrumentation for Monitoring Accident

\v Program Conditions TM! III.A.1 Improve Licensee Emergency Preparedness -

Short Ters

!! 2.0 . Availability of Meteorological TM! II.F.3 Instrumentation for Monitorin'g Accident -

Data in the Control Room Conditions TMI !!!.A.1 leprove Licensee Emergency Preparedness -

Short Ters TM1 1.D.1 Control Roon Design Reviews

!!I 8.0 Core Supports and Fuel Integrity USI 4-2 Asyneetric Blowdown Loads on Reactor Primary Coolant Systee

!!!-9 Support Integrity U5! A-12 Fracture Toughness of Steam Generator and Reactor Coolant Pumo Supports i U5! A* 7 Mark 1 Containment Long-Term Progras 051 A 24 Enviro # mental Qualification of Safety-Related Eouipment U51 A 45 seismic Qualification of Equipment in operating Plants SEP !!!-6 Selssic Design Considerations  ;

SEP V-1 Compliance With Codes and Standards (10 CFR Part

50. Section 50.55a) 111 11 Component Integeity U51 A-46

• 5elseic Qualification of Equipment in Operating Plants U$1 A 2 Asymmetric Blowdown Loads on Reactor Primary Coolant SEP !!! 6 Seismic Design Considerations

!!! 12 Environmental Qcalifications of USI A-24 Qualification of Safety Related Equipment W-3 Overpressurization Protection U51 A 26 Reactor vessel Pressure Transient Protection V-4 Piping and Safe-End Integrity us! A 42 Pipe Cracks in Boiling water Reactors V-8 Steam Generator Integrity U$1 A-3, Westinghouse, Combustion Engineering, and

/,m\ A-4, A-5 Babcock and Wilcon Steam Generator Tube Integrity

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V-13 Water Hammer U51 A-1 Waterhammer VI 2. A Pressure-Suppression-Type BWR 051 A 7 Mark 1 Containment Long-Term Progree Contairments VI 2.8 Subcompartment Analysis U5! A-2 Asyumetric Blowdown Loads on Reactor Primary Coolant Systee

'v ! 5 Combustible Gas Control ,,, TM! !!.8.7 Analysis of Hydrogen Control U51 A 48 Hydrogen Control Measures and Effects of Hydrogen Burns on Safety Eouipment v!-7.E Energency Core Cooling Systee Sume U51 A-43 Containment Emergency sumo Reliability Design and Test for Recirculation -

Mode Effectiveness

. VI 8 Control Room Macitability TMI !!!.D.3.4 Control Room Habitability Requirements VII-4 Effects of Fa :gre in Monsafety- U51 A-47 Safety lupilcations of Control Systes .

Related Systeers on Selected U51 A 17 Systees Interactions in Nuclear Power Plants Faninoered Safetv Features Wit-5 Instruments for Monitoring Radia- TMI II.F.1 Additional Accident Monitoring Instrumentation tion and P*ocess variables During TMI !!.F.2 Identification of and Recovery From Conditions Accidents Leading to Inadeovate Core Cooling TM! !!.F.3 Instruments for Monitoring Accident Concitions IX-2 Overhead Handling Systees (Cranes) USI A-36 Control of Heavy Loads Meer Spent Fuel Pool X Auxiliary Feedwater Systee TMI II.E.1.1 Auxiliary Feessater System Evaluation XI!!-1 Conduct of Operations TMI I.C.6 Procedures for Verification of Correct Performance of Operating Activities TMI III.A.1 Improve Licensee Emergency Preparedness -

Short Tere

[q TMI !!!.A.2 leproving Licensee Emergency Preparedness -

Long Ters k XV 21 Spent Fuel Cask Drop Accident USI A 36 Control of Heavy Loads hear Spent Fuel Pool XV-22 Anticipated Transients Without Scram U5! A 9 Anticipated Transients Without Scram XV-23 Multiple Tube Falltres in Steam U51 A-3 Westinghouse, Combustion Engineering, and Generators A4,A-$ and Wilcox Steam Generator Tune Integetty U51 A 9 Anticipated Transler.t Without Scram XV-24 Loss of All AC Power U5! A-44 , Station Blackout Y[

PLANT SP.ECIFIC TOPICS DELETED kicNo. SEP Tit 12 NI Reason Ftr Delztfon af Trple 11-4.E Dam Integrity 11/16/79 Not applicable to site. i j/' N !!!-3.8 Structural and Other Consequences 11/16/79 g ) (e.g.. Flooding of Safety-Related Not applicable to site because site does Nj Equiprent in Basements) of Failure not have a system whose function is to of Underdrain Systems louer the groundwater table.

!!!-7.A Inservice Inspection, including 05/07/81 Prestressed Concrete Contafruments Not applicable to this facility's design.

with Efther Grouted or Upgr'euted Tendons 111-7.C Delanination of Prestressed 11/16/79 Not applicable to this facility's design.

Concrete Containment Structures

!!!-8.8 Control Rod Drive Mechanism 10/01/80 Integrity tot applicable to Pressurized Water Reactors (PWRs).

!!!-10.C Surveillance Requirements on SWR 11/16/79 Not applicable to Ms.

Recirculation Pumps and Discharge Valves IV-3 SWR Jet Pure Operating Indications 05/07/ Not applicable to PWRs.

V-1 Compliance With Codes and Standards 11/27/81 Reviewed under Inservice Inspection / Inservice Test Program.

V-2 Applicability of Code Cases

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11/16/79 Not applicable to Ms.

V-9 Reactor Cor'e Isolation Cooling System (BWR) 11/16/79 Not applicable to PWRS.

V-12.A Water Purity of SWR Primary Coolant 11/16/79 Not applicable to PWRs.

V!-2.C Ice Condenser Containsent 11/16/79 Not applicable to this unit's containment cesign.

] VI-7.A.2 Upper Plenum Injection 05/07/81 mot applicable to this facility's design.

VI-7.A.4 Core Spray Nozzle Effectiveness 05/07/81 Not appitcable to PWRs.

VI-7.C.3 Effect of PWR Loop 1 solation Valve Closure During a Loss-of-Coolant

!!/16/79 Not applicable to this facility's design.

Accident on Emergency Core Cooling System Perforsunce VI-7.F Accumulator 1 solation Valves 11/16/79 Not applicable to tnis facility's design.

Power and Control System Design 8

VI-9 Main Steae Line Isolation Seal System (BWR) 11/16/79 Not applicable to PWRs.

V11 7 Acceptability of Swing Bus Design 11/16/79 Not applicable to PWRs.

on BWR-4 Plants II-1 Appendia 1 12/04/81 Sefng resolved under generic activity A-02, "Appenois !." and 8 35, ' Confirmation of Appendis ! Models." (See " Basis for Deletion" in Appendia A under Topic Il-1.)

Il-2 Radiological (Effluent and Process) 12/04/81 Seing resolved under generic a tivity A-02 Monitoring Systems "Appendis !.* (See " Basis for Deletion" in Appendia A under Topic Il-2.)

IV-!! Inadvertent Loading and Operation 10/01/80 Not applicaole to Ms.

of a Fuel Assembly in an Improps-position (SWR)

IV-13 Spectrum of Rad Drop Accidents (SWR) 11/16/79 Not applicable to PWRs.

IV-18 Radiological Conseovences.cf Main 10/01/80 Not applicable to Ms.

Steam Line Failure Outside

[g) Containment fl Technical Specifications 11/05/80 Will be addressed after crupletion of the integrated assesssent.

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TOPICS WHICH MEET CURRENT CRITERIA l l OR ARE ACCEPTABLE ON "ANOTHER [

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TOPIC TITLE II-1.A Exclusion Area Authority and Control II-1.B Population Distribution II-2.A Severe Weather Phenomena II-2.C Atmospheric Transport and Diffusion Characteristics for Accident Analysis II-3.A Hydrologic Description II-3.B.1 Capability of Operating Plant To Cope With Design-Basis Flooding Conditions II-3.C Safety-Related Water Supply (Ultimate Heat Sink [ UHS))

II-4* Geology and Seismology i

. II-4.A* -Tectonic Province II-4.B Proximity of Capable Tectonic Structures in Plant Vicinity II-4.C* Historial Seismicity Within 200 Miles of Plant II-4.D Stability of Slopes III-4.B Turbine Missiles III-4.C Internally Generated Missiles III-4.0 Site-Proximity Missiles (Including Aircraft)

III-8.C Irradiation Damage, Use of Sensitized Stainless Steel, and Fatigue Resistance IV-1.A Operation With Less Than All Loops In Service 4

V-6 Reactor Vessel Integrity V-7 Reactor Coolant Pump Overspeed V-10.B Residual Heat Removal System Reliability VI-2.0 Mass and Energy Release for Postulated Pipe Break Inside Containment VI-3 Containment Pressure and Heat Removal Capability VI-6 Containment Leak Testing VI-7.A.1 Emergency Core Cooling System Reevaluation to Account for Increased Reactor Yessel Upper-Head Temperature VI-7.A.3 Emergency Core Cooling System Actuation System

. .- dW/ - .- - - . _

TOPIC TITLE VI-7.C Emergency Core Cooling System (ECCS) Single-Failure Criterion and Requirements for Locking Out Power to Valves Including Independence of Interlocks on ECCS Valves VI-7.C.1 Appendix K--Electrical Instrumentation and Control Re-Reviews VI-7.D Long-Term Cooling Passive Failures (e.g., Flooding of Redundant Components) ,

VI-10.B Shared Engineered Safety Features, Onsite Emergency Power and Service Systems for Multiple Unit Facilities VII-1.B Trip Uncertainty and Setpoint Analysis Review of Operating Data Base VII-2 Engineered Safety Features System Control Logic and Design VII-6 Frequency Decay

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VIII-2 Onsite Emergency Power System (Diesel Generator)

VIII-3.A Station Battery Capacity Tast Requirements IX-1 Fuel Storage IX-4 Boron Addition System (PWR)

XIII-2 Safeguards / Industrial Security XV-3 Loss of External Load, Turbine Trip, Loss of Condenser Yacuum, Closure of Main Steam Isolation Valve (BWR), and Steam Pressure Regulator Failure (Closed)

XV-4 Loss of Nonemergency AC Power to the Station Auxiliaries XV-5 Loss of Normal Feedwater Flow XV-6 Feedwater System Pipe Breaks Inside and Outside Containment (PWR)

XV-7 Reactor Coolant Pump Rotor Seizure and Reactor Coolant Pump Shaft Break XV-8 Control Rod Misoperation (System Malfunction or Operator Error)

XV-9 Startup of an inactive Loop or Recirculation Loop at an Incorrect Temperature, and Flow Controller Malfunction Causing an Increase in BWR Core Flow Rate XV-10 Chemical and Volume Control System Malfunction That Results in a Decrease in Boron Concentration in the Reactor Coolant (PWR)

XV-12 Spectrum of Rod Ejection Accidents (PWR)

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( XV-14 Inadvertent Operation of Emergency Core Cooling System and Chemical and Volume Control System Malfunction That Increases Reactor Coolant Inventory XV-15 Inadvertent Opening of a PWR Pressurizer Safety / Relief Valve or a BWR Safety / Relief Valve XV-16 Radiological Consequences of Failure of Small Line Carrying Primary Coolant Outside Containment XV-17 Radiological Consequences of Steam Generator Tube Failure (PWR)

XV-19 Loss-of-Coolant Accidents Resulting From Spectrum of Postulated Piping Breaks Within the Reactor Coolant Pressure Boundary XV-20 Radiological Consequences of Fuel-Damaging Accidents (Inside and Outside Containment)

. XVII -Operational Quality Assurance Program 1

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7s TOPIC ISSUE MODIFICATION id II-4,F SETTLEMENT - BACKFILL S0ILS GRADE BEAMS UNDER 480V ROOM AND AUXILIARY FEEDWATER PUMPS III-6 SEISMIC REEVALUATION UPGRADES TO SAFETY-RELATED STRUCTURES EQUIPMENT PIPING AND SUPPORTS NEEDED FOR HOT STANDBY CONDITION FOR 0.67 MODIFIED HOUSNER' SPECTRUM EARTHOUAKE V-10.A ' RADIATION MONITORING OF TS ON COMPONENT COOLING CCW WATER RADIATION MONITOR V-11.A SAFETY INJECTION SYSTEM INDEPENDENT VERIFICATION O OF VALVE POSITIONS AFTER TESTING VI-I CONTAINMENT C0ATINGS INSPECT. ION AND REPAIR OF PAINTS INSIDE CONTAINMENT VI-4 CONTAINMENT ISOLATION - TS TO MAINTAIN PURGE VALVE PURGE LINE CLOSED DURING OPERATION VII-3 AUXILIARY FEEDWATER INSTALLATION OF NEW TANK AND PARALLEL SUCTION PATHS VIII-1.A VOLTAGE MONITORING MODIFICATION OF TAP SETTINGS IX-3 SALT WATER COOLING INSTALLATION OF CHECK DISCHARGE VALVES, REMOVAL OF AIR

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TOPIC ISSUE MODIFICATION IX-3 SWC WATER SUPPLY TSUNAMI GATES REMOVED i

IX-5 VENTILATION SYSTEMS INSTALLATION OF NEW VENTILATION SYSTEM IN Q

l- SWITCHGEAR AND CAPLE SPREADING ROOMS

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1 ISSUES ADDRESSED BY STAFF PRA TOPIC TITLE RANK 1 III-8.A LOOSE PARTS MONITORING LOW 111-10.A -THERMAL OVERLOAD PROTECTION FOR LOW MOTORS OF MOTOR-0PERATED VALVES IV-2 REACTIVITY CONTROL SYSTEM FAILURES LOW V-10 A RESIDUAL HEAT REMOVAL SYSTEM HEAT LOW EXCHANGER TUBE FAILURES V-11.A REQUIREMENTS FOR ISOLATION HIGH-AND

LOW PRESSURE-SYSTEMS CHEMICAL AND VOLUME CONTROL LOW s

SAFETY INJECTION MEDIUM 2 N -

LONG-TERM RECIRCULATION LOW V.11.B RESIDUAL HEAT REMOVAL SYSTEM LOW INTERLOCK REQUIREMENTS VI-4 CONTAINMENT ISOLATION SYSTEM LOW VI-7.B ESF SWITCH 0VER FROM INJECTION MEDIUM 2 TO RECIRCULATION VI-7.C.2 FAILURE MODES ANALYSIS - ECCS VOLUME CONTROL TANK ISOLATION MEDIUM CONTROL POWER FOR FLOW CONTROL MEDIUM VALVES g,/g

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v ' ')' TITLE RANK I VI-10.A TESTING OF REACTOR TRIP SYSTEMS AND LOW ENGINEERED SAFETY FEATURES, INCLUDING RESPONSE TIME TESTING VII-1.A ISOLATION OF REACTOR PROTECTION SYSTEM . LOW FROM NON-SAFETY SYSTEMS VII-3 SYSTEMS REQUIRED FOR SAFE SHUTDOWN LOW (CCW SURGE TANK LEVEL)

VIII-3.B DC POWER BUS MONITORING AND ANNUNCIATION MEDIUM VIII-4 ELECTRICAL PENETRATIONS LOW IX-3 STATION SERVICE AND COOLING WATER SYSTEMS

,e (j)' PASSIVE FAILURES

, LOW TSUNAMI GATE CLOSURE LOW IX-5 VENTILATION SYSTEMS REACTOR AUXILIARY BUILDING LOW 480V SWITCHGEAR LOW HIGH} [

4160V SWITCHGEAR/ CABLE SPREADING ROOM BATTERY AND INVERTER ROOMS H I G H ) ' " ,j XV SPECTRUM 0F MAIN STEAM LINE BREAKS LOW I RANKED AS LOW, MEDIUM OR HIGH IMPACT OF TOPIC DIFFERENCE ON RISK FROM THE PLANT IN CONTRACTOR REPORT.

2 RANKING WAS AFFECTED BY STAFF'S REVIEW.

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O FURTHER EVALUATION IPSAR TOPIC ISSUE SECTION II-1.C T0XIC GAS MONITORS 4.1,3 II-4.F SETTLEMENT OF FOUNDATIONS 4.4.1 To 4.4.7 III-2 WIND AND TORNADO LOADINGS 4.5 (LOAD COPSINATIONS) 4.12.1 III-3.A GROUNDWATER HYDROSTATIC LOAD 4.2 4.6.1 y ROOF PONDING 4.2 4.6.2 III-5.A EFFECTS OF PIPE BREAK INSIDE 4.9 CONTAINMENT 4.18.1.1 III-5.B PIPE BREAK OUTSIDE CONTAINMENT 4.10

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l IPSAR TOPIC ISSUE SECTION III-6 SEISMIC DESIGN CONSIDERATIO.NS 4.11 4.12.1 4.23.5 4.23.6 III-7.B LOAD COMBINATIONS FOR CONTAINMENT 4.12.2 ,

VI-7.C.2 FAILURE MODE ANALYSIS (ECCS SYSTEMS) 4.25.4 4.32.2 4.32.5 VIII-1.A VOLTAGE MONITORING PROGRAM 4.29

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SAN ON0FRE UNIT 1 SEISMIC RE-EVAL.UATION 0.67G MODIFIED-HOUSNER GROUND RESPONSE SPECTRUM RETURN TO SERVICE (RTS) PROGRAM (COMPLETE 11/21/84)

HOT-STANDBY CAPABILITY STAFF AUDITS AND CONFIRMATORY ANALYSES CONTINGENT RECISSION OF SUSPENSION LONG-TERM SERVICE (LTS) PROGRAM (ONGOING)

COLD SHUTDOWN CAPABILITY ACCIDENT MITIGATION CAPABILITY IMPROVED ANALYSIS TECHNIQUES IMPROVED ACCEPTANCE CRITERIA RTS CONFIRMATORY EVALUATIONS STAFF AUDITS AND CONFIRMATORY ANALYSES MODIFICATION SCHEDULE - ONE CYCLE O

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l III-4.A TORNADO MISSILE PROTECTION 4.8 VI-7.B ESF SWITCH 0VER - AUTOMATIC TEP.MINATION 4.?h '

0F ECCS FLOW AND BACKUP LEVEL INDICATION O

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TOPIC ISSUE SECTION VII-3. MOTOR-DRIVEN AUXILIARY FEEDWATER TRAIN 4.28.3 XV-2 4.36 VIII-1.A UNDERVOLTAGE LOGIC AND TS 4.29 IX-3/IX-6 FIRE PROTECTION DEDICATED SHUTDOWN 4.32.2

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PROCEDURES AND TECHNICAL SPECIFICATIONS IPSAR TOPIC ISSUE SECTION III-3.C INSERVICE INSPECTION PROGRAM FOR WATER 4.7 CONTROL STRUCTURES V-5 TECHNICAL SPECIFICATIONS FOR OPERABILITY 4.18.1.2 1

0F LEAK DETECTION SYSTEMS SEISMIC QUALIFICATION OR PROCEDURES 4.18 I.3 V-11.B RESIDUAL HEAT REMOVAL - OVERPRESSURE 4.21.2 PROTECTION TS VI -I 4

O VI-4 CONTAINMENT C0ATINGS - INSPECTION PROGRAM SEQUENCER TEST PROCEDURES 4.22 4.23.1.2 W

LOCKING DEVICES / PROCEDURES FOR REFUELING4.23.7.1 WATER LINE BRANCHES PROCEDURES FOR ISOLATING MAIN STEAM 4.23.7.2 DRAIN LINES VI-10.A TECHNICAL SPECIFICATIONS FOR RPS 4.26.1 CHANNELS TESTING TECHNICAL SPECIFICATIONS FOR CONTAINMENT4.26.2 SPRAY ACTUATION LOGIC IX-5 TEMPERATURE MONITORING PROGRAM - CABLE 4.33.2 SPREADING ROOM AND 480V ROOM O' PROCEDURES FOR HYDROGEN DISPERSION / ROOM4.33.3 COOLING FOR BATTERY R00M k'Qf .

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O F0 ACTION IPSAR TOPIC ISSUE SECTION II .1.C OVERPRESSURE FROM EXPLOSIONS 4.1.1 FREQUENCY OF SHIPMENTS 4.1.2 III-7.D CONTAINMENT STRUCTURAL !NTEGRITY TESTS 4.13 III-8.A LOOSE PARTS MONITORING 4.14 III-10.A THERMAL OVERLOAD PROTECTION FOR 4.15 MOTORS OF MOTOR-OPERATED VALVES III-10.B PUMP FLYWHEEL INTEGRITY 4.16 IV-2 REACTIVITY CONTROL SYSTEMS 4.17 V-5 TESTABILITY - LEAKAGE DETECTION 4.]8.1.4 l

INTERSYSTEM LEAKAGE 4.18.2 V-10.A RADIATION MONITORING 4.19.3 SAMPLING 4.19.2 TESTING OF RECIPCULATION HEAT EXCHANGEP 4.]9.3 V-13.A CHARGING PUMP DISCHARGE 4.20.1.3 LETDOWN 4.2n.).2 SAFETY INJECTION 4.20.2

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VI-4 PURGE LINES b M 4.23.1.1 ANNUNCIATOR WINDOW DESIGN 4.23.1.3 AUTOMATIC LOADING OF DIESEL 4.23.1.4

. GENERATOR FANS OVERRIDE CAPABILITY OF SAMPLE LINES 4.23.1.5 4.23.2 VALVE TYPE 4.23.3

.,w-VALVE LOCATION 4.23.4 SPARE PENETRATIONS 4.23.8 AIRLOCKS AND HATCHES 4.23.9 VI-7.C.2 REDUNDANT VALVE FOR VOLUME CONTROL 4.25.]

TANK ISOLATION CONTROL POWER TO FCV-1]I5 D, E, F 4.25.2 HDT-LEG RECIRCULATION 4.25.3 VI-10.A TESTING OF SUPPORT SYSTEMS 4.26.3 O

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' IPSAR TOPIC ISSUE SECTION VII-1.A REMOTE METERS AND RECORDEPS 4.27.1 DATA LOGGER 4.27.2 FEEDWATER CONTROL 4.27.3 VII-3 COMPONENT COOLING WATER 4.28.1 SEISMIC CATEGORY I WATER SUPPLY 4.28.2 FOR AUXILIARY FEEDWATER VIII-3.B DC POWER BUS MONITORING AND 4.30 ANNUNCIATION O VIII-4 ELECTRICAL PENETRATIONS 4.3]

IX-3 COMPONENT COOLING WATER (CCW) 4.32.1 SYSTEM TEMPERATURE SWC SUPPLY FAILURE 4.32.4 LOSS OF BEARING FLUSH 4.32.6 IX-5 REACTOR AUXILIARY BUILDING 4.33.1 1

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-Os INDUSTRIAL AND MILITARY FACILITIES IPSAR SECTION 4.1 OVERPRESSURE FROM EXPLOSIONS ON HIGHWAY OF RAILWAY FREQUENCY OF SHIPMENT '

T0XIC GAS MONITORING FOR CONTROL P.00M PRESENT CP HVAC IS SINGLE TRAIN SYSTEM, NOT SEISMICALLY QUALIFIED, NO T0XIC GAS MONITORING O .

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l II-3.B, HYDROLOGIC ENGINEERING, FLOOD PROT 5CTION REQUIREMENTS, O

5 II-3.B.1 EFFECTS OF HIGH WATER LEVEL ON STRUCTURES II-3.C 1

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IPSAP SECTION 4,2 ROOF PONDING UNDER PMP - VENTILATION AND FUEL STORAGE BUILDING SHORT-TERM HYDROSTATIC LOAD - GROUND WATER LEVEL AT GRADE INFORMATION ON PONDING SUBMITTED - UNDER STAFF REVIEW O  :

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BACXFILL SOILS UNDER SOME STRUCTURES AND COMPONENTS l EFFECTS DURING EARTHOUAKES CONSIDERED IN SEISMIC PREVALUATION (IPSAR SECTION 4.11)

STXFF E' VALUATING LICENSEE'S ASSESSMENT OF SETTLEMENT

III-), QUALITY GROUP CLASSIFICATION IPSAR SECTION 4.4 PADIOGRAPHY OF WELD JOINTS PRESSURE VESSEL FATIGUE ANALYSIS FRACTURE TOUGHNESS EFFECT UNDER CYCLIC LOADS OF GROSS DISCONTINUITIES ON USAGE FACTOR FOR PIPING

.- VALVE STRESS LIMITS PUMP DESIGN STANDAPDS STF.ESS RECUIREMENTS FOR STORAGE TANXS O*LICENSEETOEVALUATETHESEISSUES,AND CORRECTIVE MEASUPES O

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III-2, WIND AND TORNADO LOAI}ING/ TORNADO MISSILES III-4.A  %, ,

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, STAFF POSITION IS THAT LICENSEE SHOULD PROVIDE A SINGLE PROTECTED TRAIN FOR SAFE SHUTDOWN CAPABILITY IN THE EVENT OF A TORNADO (MISSILES) '

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III-6, SEISMIC DESIGN CONSIDERATIONS O IPSAR SECTION 4,11 RETUPN TO SERVICE PROGRAM EST4BLISH ACCEPTANCE CRITERIA AND METHODOLOGY FOR LONG-TEFM OPERATION IMPLEMENT MODIFICATIONS AS NECESSARY FOR COLD SHUTDOWN AND ACCIDEN7 MITIGATICF SYSTEMS 8

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7

III -7,B, DESIGN CHANGES, LOADS AND LOAD COMBINATIONS O IPSAR SECTION 4.12 CONSIDER APFROPRIATE LOAD COMBINATIONS IN ASSESSMENTS OF WIND AND TORNADOS, PIPE BREAK AND SEISMIC EVENT FORMSLBPLb5 EARTH 0VAKE,HIGHSTRESSINSAND-FILLED TRANSITION ZONE, POSES A CONCERN FOR POTENTIAL BUCKLING OF CONTAINMENT STAFF IS REVIEWING LICENSEE RESPONSE ON BUCKLING O

e 4

0 O

syo

III-7 D, CONTAINMENT STRUCTURAL INTEGRITY TEST IPSAR SECTION 4.13 .

CRITERIA SPECIFIES FACTOR OF 1.1 PRESSURE / PEAK CALCULATED PRESSUPE TEST PRESSURE WAS 53.4 PSIG LOCA PRESSURE IS 49.4 PSIG MSLB PRESSURE IS 53.3 PSIG

q. . ISSUE C0ySIDERED CLOSED a

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O III-8,A, LOOSE PARTS MONITORING -

IPSAR SECTION 4.14 LOOSE PARTS MONITORING PROGRAM NOT IMPLEMENTED NO SAFETY-RELATED INCIDENTS HAVE OCCURRED AS A RESULT OF LOOSE PARTS MOST LOOSE PARTS CAN BE FOUND.DURING REFUELING PRA RESULTS SHOW LOW SIGNIFICANCE ISSUE CLOSED O

t 9

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i  !

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III-10.A, THER. MAL OVERLOAD i

i I IPSAP SECTION h.15 i  !

1 TWO VALVES IN COMPONENT COOLING WATER SYSTEM PAVE MOVs W!TFOUT i BYPASS OF THERMAL OVERLOADS +

1 I ONE OF TWO IS NORMALLY OPEN l ONCE OPENED, NOT REQUIRED TO CYCLE 1

• t ACCESSIBLE FOR MANUAL OPERATION s

4 l

f

_- ISSUE CLOSED i i e +

] I i I

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e k -/f6

III-10.B, PUMP FLYWHEEL INTEGRITY O IPSAR SECTION 4.16 ISSUE WAS NEED TO INCLUDE FLYWHEEL TESTING IN INSERVICE INSPECTION PROGRAM OR TECHNI Al SPECIFICATION INSPECTIONS HAVE BEEN PERFORMED ON FREQUENCY SPECIFIED TEST PROCEDURE SATISFIES R.G. 1.14 GUIDANCE ISSUE CLOSED e

O O .

A/f7

IV -2, REACTIVITY CONTROL SYSTEM -

O iPSAR SECTION 4.17 RCD POTIONS OCCUR AS RESULT OF SINGLE FAILURE i

- GROUP, BANK OR TWO BANK MOVEMENT  ;

i

- GROUP OR BANK MAY.NOT Pt0VE

- ROD, GROUP OR BANK MAY FALL INTO CORE '

TWO BANK WITHDRAWAL HAD NOT BEEN EXPLICITLY CONSIDEPED

' REACTIVITY INSERTION RATE WITHIN PANGE CONSIDERED IN POD WITHDRAWAL ANALYSIS '

6 ISSUE RESOLVED G

I t

P O

9

+

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e . .

8 G

! i i

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V-5, REACTOR COOLANT PRESSURE BOUNDARY (RCPB) LEAKAGE DETECTION i

IPSAR SECTION 4.18  !

SEVERAL SYSTEMS PROVIDED, INCLUDING THE THREE RECOMMENDED IN R.G. 1.45 LEAK DETECTION SENSITIVITY.NOT 1 GPM IN ONE HOUR i TS ON OPERABILITY TO BE PROVIDED SEISMIC QUALIFICATION OR PROCEDURES ESTA5LIS$NEEDEDSENSITIVITYINCONJUNCTIONWITHFRACTUPE MECHANICS EVALUATIONS FOR PIPE BREAK ASSESSMENT j i

4 i  !

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_ . = = _ _ _ _ _ _ ____

$Vff _ - - - - .

V-10,B, RESIDUAL HEAT REPOVAL SYSTEM HEAT EXCHANGER TUBE FAILURES IPSAR SECTION h,19 SAMPLING COVERED BY PROCEDURES RADIATION. MONITOR PROVIDED FOR CCW SYSTEM RECIRCULATION HEAT EXCHANGER TESTING .

ISSUE CLOSED 6

O O .

+w

V-11.A, REQUIREMENTS OF ISOLATION OF HIGH AND LOW PRESSURE SYSTEMS IPSAR SECTION 4.20 CHARGING PUMP SUCTION LINE OVERPRESSUPIZATION

- CHECK VALVE TESTING

- SEVERAL FAILUPES REQUIRED LETDOWN ISOLATION

_- SAFETY INJECTION SYSTEM

- MOV AND CHECK VALVE IN SERIES

- CHECK VALVE HAS TS LEAKAGE LIMIT

- PIPING DESIGN PRESSURE IS 1400 PSIG LONG-TERM RECIRCULATION

- SEVERAL VALVES IN SERIES

- CHARGING PUMP SUCTION (SEE ABOVE)

/

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V-11.B, RESIDUAL HEAT REMOVAL INTERLOCK REQUIREMENTS IPSAP SECTION 4.21 TWO MOVs IN SERIES ON BOTH SUCTION AFD DISCHARGE

~

ONE SET WITH PRESSURE INTERLOCK '

ADMINISTRATIVE CONTR01.S PROVIDED RHR SYSTEM TOTALLY INSIDE CONTAINMENT TS FOP OVERPRESSURE PROTECTION TO BE IN OPERATION

.- WHENEVER.RHR SYSTEM IS a

t 6

0 .

t ll t

O sn

VI-1, ORGANIC MATERIALS IPSAR SECTION 4.22 INSPECTION AND REPAIR (PAINT TOUCHUP) DURING PAST GUTAGE PERIODIC REINSPECTION PROGRAM TO BE INSTITUTED 4

9 6 S

O 4

6 f

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4 O

VI-4, CONTAINMENT ISOLATION SYSTEM IPSAR SECTION 4.23 PURGE LINES LOCKED CLOSED DURING MODES 1 THROUGH 4 PROCEDURE CHANGES FOR SEQUENCER TESTING TO BE MADE .

CONTAINMENT ISOLATION CONFIGURATIONS

- BOTH VALVES OUTSIDE CONTAINMENT

- CHECK VALVE INSTEAD OF AUTOMATIC VALVE

- SEISMIC QUALIFICATION FROCEDURES/ LOCKING DEVICES FOR TEST, INSTRUMENT CCNNECTIONS PROCEDURES TO ISOLATE MAINSTEAM DRAIN LINES IF FEQUIRED 9

9 0

0 9

-f*

., ._ - - . . - . . . = - - -

.,,.n -.+------.-_.>-.:-. _ - --___ - _ . -

VI-7.B,ESFSWITCHOVERFRSMINJECTIONTORECIRCULATIONMODE IPSAR SECTION 4.24 ,

TANK INSTRUMENTATION IS LEVEL ALARM AND LEVEL INDICATION

,MANUAL SWITCHOVER TWO-PART PROCEDURE  ;

- REDUCED INJECTION

- RECIRCULATION TIME FOR OPERATOR ACTION IS SHORT IF FEEDWATER FLOW NOT REDUCED PROMPTLY STAFF POSITION AUTOMATIC TERMINATION OF SAFETY INJECTION FLOW ,

BACKUP TO RWST LEVEL INSTRUMENT

.2 en s-r

a VI-7.C.2, FAILURE MODES ANALYSIS - ECCS IPSAR SFCTION 4.25

" INTERIM MODIFICATIONS" .

- MOV-1100C ,

- NITROGEN SUPPLY FOR FLOW CONTROL VALVE

- HOT LEG. RECIRCULATION i

OTHER RECOMMENDATIONS

- ENVIRONMENTAL QUALIFI. CATION

- SEPARATION i  !

O i

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1

VI-10.A, TESTING 0F REACTOR TRIP SYSTEM AND ENGINEERED SAFETY FEATURES, INCLUDING RESPONSES TIME TESTING IPSAR SECTION U 26 TS FOR RPS CHANNEL TESTING, CALIBRATION (MANUAL, STARTUP RATE,. SAFETY INJECTION)

TS FOR CONTAINMENT SPRAY ACTUATION LOGIC ALREADY INCLUDE SUFFICIENT TS TO ENSURE OPERABILITY OF SUPPORT SYSTEMS o

O s-7

' - ' ^ - - ~

^.

VII-1.A, ISOLATION OF REACTOR PROTECTION SYSTEM FROM NON-SAFETY SYSTEMS, INCLUDING QUALIFICATION OF ISCLATION DEVICES

_IPSAR SECTION 4,27 REMOTE METERS DATA LOGGER FEEDWATER CONTROL i

FAILURE MODES ASSESSED TO SHOW SUFFICIENT CHANNELS AVAILABLE TO PROTECT PLANT ISSUk PE50LVED O

i

// M r ,

VII-3, SYSTEMS REQUIRED FOR SAFE SHUTDOWN O IPSAR SECTION 4,28 COMPONENT COOLING WATER SUPGE TANK LEVEL INDICATION

- LOCAL GAUGE CPECKED ONCE PER SHIFT AUXILIARY FEEDWATER SYSTEM MODIFICATIONS .

- NEW STORAGE TANK i.

- NEW PARALLEL SUCTION LINES TO PUMPS THIRD (MOTOR-DRIVEN) AUXILIARY FEEDWATER PUMP TO P.E INSTALLED O

s 6

O

1. . a7

VIII-1.A, ADEQUACY OF STATION ELECTRICAL VOLTAGE DISTPIBUTION O IPSAR SECTION U 29 BUS.UNDERVOLTAGE PROTECTION

- COINCIDENCE LOGIC i - TS GRID OVERVOLTAGE

- TRANSFORMER TAP SETTINGS OPTIMIZED

-VOLTAG5MONITORINGPROGRAMTOCONFIRMAPPROPRIATE SETTINGS O

r-r O

A. No

- - _ - _ _ _ _ . _ . - . . . . = . - . _ . - . - - - - - -

. ._. - - - . - -. . .. _ _ . . . . - . .- - _ . = _ _ _ - . - . -

i VIII-3.B, DC POWER BUS SYSTEM MONITORING AND ANNUNCIATION

~

O IPSAR SECTION 4,3 l I CONTROL ROOM TROUBLE ALARMS WITH LOCAL MONITORS ACCEPTABLE INSPECTION / TEST PROGRAM FOR SATTERIES ISSUE RESOLVED i

i i

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._ -.---._ _ _ - .. . ......= - - -. --. .--- =

= _ _ . . _ . . - . _ _ _ . __.. .

i 4

VIII-4, ELECTRICAL PENETRATIONS f

10510 SECT!Dr t,3 -

i LOW VOLTAGE PENETRATIONS EXCEED DESIGN TEMPERATURE IN EVENT OF LOCA FAULT CURRENT PLUS FAILURE OF PRIFARY CIRCUIT INTERRUPT DEVICE SEALED CANISTER PENETRATIONS LOW P.!SK CONTPIPUTION 1

j ISSUE FESOLVED I

a e

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f f -tT-- ---wew cemmw---.m,,- -w-rw---c - , ,-- , ,-3--s,. e,- ww- - - - - -_.0-w-+,w---rvw---w,ewww-tv--w,v <w-m.m.,-wwew w -g-- w- --ere- -,-m-------g-

._ .__.,4 -. .

O IX-3,STATIONSERVICEANDC0blINGWATERSYSTEMS IPSAR SECTION 4.32 PASSIVE FAILURES NOT SIGNIFICANT CONTRIBUTOR TO SYSTEM UNAVAILABILITY SALT WATER COOLING SYSTEM EXPERIENCE CHECK VALVES INSTALLED AIR OPERATORS REMOVED TSUNAMI GATES REMOVED RECOMMEND QUANTITATIVE RELIABILITY ANALYSIS i

l l

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IX-5,' VENTILATION SYSTEMS O

l i

I PSAR SECTION 4,33 REACTOP AUXILIAPY BUILDING CABLE SPREADING /4160V SWITCHGEAR AND UE0V SWITCHGEAP ROOMS

- NEW VENTILATION SYSTEMS

- TEMPERATURE MONITORING PROGRAM / PROCEDURES FOR LOSS OF VENTILATICN

._- BATTERY /[NVERTER ROOMS

- HYDROGEN DISPEPSION/ ROOM COOLING PROCFDUPES O

I,

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• 1 i i -

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i: -IX-6, FIPE PROTECTION

\

j IPSAR SECTION 4,% .

. 1,

,- t t-DEDICATED:POWEP SUPPLY; TPANSFER SWITCHES l l-i

• ' STUDY:OF CABLING (SEE SECTIOP 4,9) 1 -

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!: . l i XV-1, INCPEASE IN FEEDWATER FLOW f t

.IPSAR SECTION 4.35

{

• -i 1: EVAll' ATE POTENTIAL FOR STEAM GENERATOP OVERFILL -l e ,

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-__.__.v-.-__-.7.-.-

XV-2, SPECTRUM OF STEAM SYSTEM PIPING FAILURES INSIDE AFD '

4 OUTSIDE CONTAINPENT IPSAR SECTION 4.36 l

• ~

INSTALL ADDITIONAL MOTOR-DRIVEN AUXILIARY FEEDWATER PUMP I

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APPENDIX XVIX - GESSAR II SEVERE ACC ISS. PRESENTATION TO THE ACRS O

~

GESSAR 11 SEVERE ACCIDENT ISSUES A PRESENTATION TO THE ADVISORY COMMITTEE ON REACTOR SAFEGUARDS O .

WASHINGTON, D.C.

GENERAL ELECTRIC COMPANY AUGUST 8, 1985 O

n vr _

(~) HYDROGEN ISSUES

'v/

o RATE AND AMOUNT GENERATION RATES VARY FROM 0.4 TO 1.6 g LB /SEC 1300-2300 LB g TOTAL IN-VESSEL HYDR 0 GEN ONLY EN0 UGH 0XYGEN TO SUPPORT COMBUSTION OF 2480 Nj s HYDROGEN (+ 67 PERCENT OF ACTIVE CLAD MW o HYDROGEN DETONATIONS INSIGNIFICANT RISK REDUCTION FOR ADDITIONAL HYDR 0 G CONTROL (BASED ON PRA RESULTS WITH DETONATIONS)

CURRENT UNDERSTANDING--LOW LIKELIHOOD OF DETONATI IN MARK III $*

RISK EVEN LOWER THAN ORIGINAL PRA RESULTS SER SHOWS NO RISK REDUCTION FOR HYDR 0 GEN CONTRO INTERNAL EVENTS, FACTOR OF 2 FOR SEISMIC RISK (BASED ON

( DRYWELL FAILURE BY LOCAL DETONATIONS, GE ANALYSES DISAGREE) o GE COMMITMENT:

PROVIDE A HYDROGEN CONTROL SYSTEM CONSISTENT WITH OUTCOME OF HC0G PROGRAM AND NRC REVIEW NRC REQUIRING DIVERSE POWER SUPPLY FOR IGNITERS (BEY HC0G POSITION OF POWER FROM EDG)

GE FINDS NO TECHNICAL JUSTIFICATION FOR DIVERSE POWE SOURCE .

- cc J + urrs o GE POSITION:

HYDROGEN CONTROL UNNECESSARY--ABSOLUTE RISK ALREADY L N0 JUSTIFICATION FOR IGNITER SYSTEM ON COST-BENEFIT BASIS O

i t

O EFFECT OF STANDING FLAMES ON SEALS e ISSUE: CAN STANDING FLAMES FROM HYDROGEN DEGRADE DRYWELL SEALS LEADING TO P0OL BYPASS?

e ASSESSMENT:

DRYWELL EQUIPMENT HATCH HAS A 5 FOOT CONCRETE

, SHIELD PLUG

]()

j PERSONNEL AIRLOCKS ARE DOUBLE SUBMARINE D0 ORS WITH CEMENT Si!IELD PLUG ON WETWELL SIDE 4

ELECTRICAL PENETRATIONS ARE 5 FOOT LONG AND POTTED WITH A PORTLAND CEMENT MIXTURE p e CONCLUSION:

NO EFFECT OF STANDING FLAMES ON DRYWELL SEALS 3

DAH

. - . . = . - ..-.- .

Yff

r ABLATION OF RPV PEDESTAL o

PEDESTAL IS A STEEL-CONCRETE COMPOSITE CONSTRUCTION TWO CONCENTRIC STEEL SHELLS CONNECTED WITH STEEL SHEAR TIES CONCRETE FILLED BETWEEN THE SHELLS o

EVALUATED SUPPORT CAPABILITY AFTER ABLATION ASSUME LOSS OF 1.4M 0F CONCRETE ASSUME ONLY SUPPORT IS OUTER STEEL SHELL ASSUME OUTER SHELL TEMPERATURE IS 1100*F o RESULTS LOADS ON OUTER SHELL WEIGHT OF RPV 2300 KIPS i

O WEIGHT OF SHIELD WALL + EQPT 2700 KIPS WEIGHT OF PEDESTAL TOTAL

_1100 KIPS 6100 KIPS COMPRESSION IN STEEL SHELL = 3.4 KSI YIELD STRENGTH OF STEEL AT 1100*F = 21 KSI o CONCLUSIONS PEDESTAL WILL CARRY LOADS - SUBSTANTIAL MARGIN NO LOSS OF PEDESTAL, DRYWELL OR CONTAINMENT STRUCTURAL l

INTEGRITY .

l t

. - , s ,-. .

APPENDIX XX - GESSAR II PRA REVIEW DETAILED DISCUSSION OF HYDROGEN by T. PRATT O .

GESSAR-II PRA REVIEW

' DETAILED DISCUSSION OF HYDROGEN PRESENTED BY TREVORfR6TT .

BROOKHAVEN NATIONAL LABORATORY

-. UPTON, NEW YORK 11973 PRESENTED TO THE ACRS AUGUST 5, 1985 BROOKHAVEN NATIONAL LABORATORY l} g)l

- A5500ATED UNIVER$1 TIES, INC.(llli M L) J 8

HYDROGEN ASSESSMENTS 6ESSAR II PRA REVIEW:

BASED ON FULL CORE MELTDOWN ACCIDENTS INITIAL SUBMITTAL INCLUDES NO PROVISION FOR H 2 CONTROLDU$1NGSEVER{ ACCIDENTS

- CONSEQUENTLY, VERY HIGH PROBABILITY OF EARLY

~~ O' CONTAINMENT FAILURE AND SIGNIFICANT PROBABILITY OF EARLY LOSS OF DRYWELL INTEGRITY (VIA DETONATION)

CONTAINMENT EVENT TREES IN SUPPLEMENT NO. 2 SER (NUREG-0979) DO NOT CONSIDER Hz CONTROL l

IMPACT OF H2 CONTROL ADDRESSED IN SUPPLEMENT NO. 4 TO SER e .

l O' BROOKHAVEN NATIONAL LABORATORY l} g)l A5500ATED UNIVER$1 TIES, INC.(llll

HYDROGEN ASSESSMENTS (CONT.)

l HC06/NRC INTERACTIONS: .

DEALS WITH DEGRADED CORE ACCIDENTS (CORE REMAINS IN-VESSEL)

AIM IS TO MAINTAIN CONTAINMENT AND DRYWELL INTEGRITY.,B,Y Hz CONTROL., .

RATES AND AMOUNT OF H2 GENERATION ARE IMPORTANT p( / FOR DESIGN OF H2 CONTROL DEVICE ISSUES RELATED TO DELIBERATE IGNITION:

OPTIMUM IGNITION SOURCES TYPE OF POWER SOURCE LIMITATIONS OF IGNITION SOURCES EFFECT OF STANDING FLAMES l

O BROOKHAVEN NATIONAL LABORATORY l} g)l ASSOCIATED UNIVER$1 TIES, INC.(E Lll

[> > j -

_,  : .. . ---.l.________.._-. __

HYDROGEN DETONATIONS

- H 2 GENERATION 4

H2 DISTRIBUTION ,

, .~~ . . . . .

POTENTIAL FOR DETONATIONS:

IGNITION SOURCE DDT - )g% %1-bT MAGNITUDE OF SHDCK LOAD I

RESPONSE OF STRUCTURES TO LOADS BROOKHAVEN NATIONAL LABORATORY l} g)l A5500ATED UNIVERSITIES, INC.(llll ku4 .__ _ - _ _ . _ _ . _ _ _ _ _ _ _ _ _ _ _ _

H, GENERATION .

RATE OF RELEASE FROM PRIMARY SYSTEM ,

TOTAL AMOUNT GENERATED t

GE AND BNL APPROACHES SIMILAR (BASED ON MARCH CODE)

.v' e. , , , ,

GE AND BNL PERFORMED SENSITIVITY STUDIES TO ASSESS IMPACT OF UNCERTAINTIES FOR

REFERENCE:

M/::S Zn IN CLADDING 72,000 La MASS Zn IN BOXES 64,000 LB POTENTIAL H2 MASS 6,200 La l

1,700 ts H2 PRODUCES 20 VOLUME PERCENT IN CONTAINMENT i

O BROOKHAVEN NATIONAL LAB' ORATORY l} g)l A5500ATED UNIVERSITIES, INC.(11,1I

.. -. . . - _ _ - $ '_**

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H, DISTRIBUTION i

i GE CALCULATED DISTRIBUTION USING IN-HOUSE CODE r

i

. . .~ , . . . .

BNL DISTRIBUTION BASED OR IN' HOUSE CODE AND HECTRE i ' CALCULATIONS AT SNL EXAMPLE: 50 LB/ MIN FOR 27 MINUTES (1350 La TOTAL H RELEASE) i l

\

i i

BROOKHAVEN NATIONAL LABORATORY l} g)l A5500ATED UNIVERSITIES, INC.(llli N - _ . . - - . - . . . - - - -

i l

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H2 - 50. LB/ MIN,FOR 26.667 MIN .

c d .

E. . ............;........................-.--.--.-

o +

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8 oc .

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0.0 O. 1$.0 05.0 TIMC - MINUTCS.

35.0 4$.0 9 ZONES 55.0 E0.0

QUANTIFICATION OF CET .

H2 IGNITION: ,

DELIBERATE IGNITION DEVICE NOT INCLUDED HIGH PROBABILITY OF IGNITION AT TIME OF POWER RESTORATION ,

.: H2 FLAMMABILIT.Y., LIMITS: ,

>4 VOLUME PERCENT UPWARD PROPAGATION

>9 VOLUME PERCENT DOWNWARD PROPAGATION H2 DETONABILITY LIMIT

>18 VOLUME PERCENT

• PROBABILITY OF H2 EVENTS BASED ON PROBABILITY OF POWER RESTORATION AND FRACTION OF TIME LIMITS ARE EXCEEDED l

BROOKHAVEN NATIONAL LABORATORY l} g)l A5500ATED UNIVERSITIES, INC.(Illl

-x 7.

O COMPARISON OF CONDITIONAL PROBABILITIES FOR H, PHENOMENA (CLASS'I TRANSIENT WITH LOOP - POWER RESTORED Lt. 7MC di. ; . PRIOR-T0 VESSEL FAILURE) .

.fiE BL GLOBAL DETONATION 01 00

~ .. - - .  :

6LOBAL COMBUSTION 03 0 66~

. LOCAL DETONATIDs 03 0 08 LOCAL COMBUSTION 03 0 26 1

BROOKHAVEN NATIONAL LABORATORY l} g)l ASSOCIATED UNIVERSITIES, INC.(Illl L

. #~M j

DYNAMIC LOADS RESULTING FROM H, PHENOMENA .

I

~

_5 = -

GE AND BNL CALCULATE SIMILAR LOADS FOR H 2 DEFLAGRATIONS _

GE ASSESSMENT OF H2 DETONATIONS p -_.

PEAK P . .= 17 (STRUCTURE PARALLEL TO WAVE)

INITIAL '

11 '

PEAK

- P =

41 7 (STRUCTURE PERPENDICULAR TO WAVE)

INITIAL WAVE TREATED AS EQUIVALENT TRIANGULAR PULSE NRC ASSESSMENT OF H2 DETONATIONS BASED ON CALCULATIONS AT SNL USING CSQ CODE BROOKHAVEN NATIONAL LABORATORY l} g)l A5500ATED UNIVER5mES, INC.(llll

. k-w f ..

O b RESPONSE OF STRUCTURES -

' ~

!*t*! O g g : -- . . _ _ . .

, f rui t  ::

DYNAMIC SHOCK LOADS CONVERTED INTT EQUVALENI SIATIf . .

LOADS BY DYNAMIC LOAD FACTORS ,

= . v ...

11.' > di .  ::

EQUIVALENT STATIC LOADS.CDMPARED AGAINST CAPACITIES OF STRUCTURES

~.. .

.. .~.

4. . .. .

GE CONCLUDED:

L DETONATIONS FAIL CONTAINMENT

. . .:n -e , : . - . : .. t . ..... .

GLOBAL DETONATIONS:FAlt!'DRYWELt ROOF UNDERWATER

.- NRC CONCLUDED:  : :,e , - -

DETONATIONS FAIL CONTAINMENT POTENTIAL FOR FAILURE OF DRYWELL WALL AS WELL AS ROOF UNDERWATEP.

i i

BROOKHAVEN KAil0NAL LABORATORY l} lj l A5500ATED UNIVERSITIES, INC.(Illl l

j _

. A M80

IlYDROGEN BilRN EVENT TREE FOR R' 'ATION OF POWER BEFORE CORE StilMP t .

GLOBAL EVENT DETONATION DRYWELLBREAdi FAILIIRE IN -

' REMARKS (SPATIAL ntST.) (CONCENTRATION) (STRESS ANALYSIS) DRYWELL llEAD DRYWELL WALL FAIL 00

.ES 10 $

NO FAIL Y

t is in FAIL OllENCllED

'h+

5 N -

q, 8 5 FAIL NO OllENCH

\

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34 24 2 NO Fall

, 76 NO FAIL BROOKHAVEN NATIONAL LABORATORY l) g)[

A5500ATED UNIVER5mES, INC.(I til

Table 15.1

• Conditional consequences predicted by the staff for internally initiated events and probability of .

occurrence with and without UPPS, per reactor year Release' Early E'arly Probability - .

category" Laten't Person-fatality injury fatality rems 1-T- L3 0 w/o UPPS w/UPP5 0__' 40

_ 7 x E5"" 3 x E-6

+

1-T-E3 0 9 x E-7 0.0005 200 3 x E6 8 x E-6

. 1 x E-6 1-T-12Q 0 3 200 3 x E6- 1 x E-5 2-T-83 0 1 x E-6 0 3'00 5 x E6

' ATWS 0 4 x E85 - 4 x E-7,- '

i.

1 400 ,

6 x E6 3 x E-6 3 x E-S

• 1-T-I2 0 .. 6 500 s

8 x E6 3 x E-6

• 1-5B-R 3 x E 0.006 19 s 604 - '

9 x E6 1 x E-9 1.x E-9 .

"See definitions in Table 15.15.

""7 x E5 = 7 x 105.. .

m Notes: ,

(1) -

All_ conditional'mean consequences were calculated using the upper range BNL source term values described in SSER 2.

5d2)

The calculations assumed the Shippingport site, with public evacuati'on within 10 miles and relocation 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after plume cassage.

(3)

! .Mean consequences were computed over 91 different weather conditions. ,

i i

l 9

i .

- ----,_e.,-<+a--en-e- ,----s - - - - - -

O nestacearion

.i- . 76LdBAL '- NO IGNITORS .

'- FAILURE OF METWELL SEAL 4

, ASSUMED UNIT PROBABILITY

- SMALL PROBABILITY OF RPV PIPE BREACH OR DRYWELL SEAL F,AILURE

' J LEADINGTOElORE2 RELEASE [ - - - - -

t, OTHERWISE 6 RELEASE

- ABOUT FACTOR OF 3 IN 2

PERSON.-REM CON 4FfiUENCES LOCAL - WETWELL SEAL MAY FAIL

- SMALL PROBABILITY OF RPV PIPE BREACH OR DRYWELL SEAL FAILURE

'?

4 LEADING TO E OR 8 RELEASE OTHERWISE L3 RELEASE

,,, - ABOUT FACTOR OF 4 IN PERSON-REM CONSEQUENCES -

1

.I

. I h"NO

LOCAL DETONATIONS 1-SB-El PORTRAYS DRYWELL AND WETWELL EARLY FAILURE CREDIT FOR PRIMARY SYSTEM RETENTION AND POOL

. SCRUBBING OF YOLATILES PERSON-REM CONSEQUENCES ABOUT AN ORDER OF MAGNITUDE GREATER THAN 1-T-L3 - - -

1 - T - 12, 1 - T - 120 PORTRAYS *.DRYWELL HEAD FAILURE DUh TO DETONATION SHOCK LOAIF- -

i t*

ON FAILURE LOCATION-l.7 l

4

1. -LL?Y. .

STATUS OF HCOG CONSIDERATIONS NRC STAFF POSITION ON ACCEPTABLE HYDROGEN RELEASE HISTORIES DEFINED IN LETTER FROM BERNER0 TO HOBBS, DATED JUNE 24, 1985

- CASE A: 150 GPM STARTED 3100S AFTER SCRAM

- CASE B: 5000 GPM FLOW

- CASE C: CASE A FOLLOWED BY 0 1 1.s/s H2 UNTIL 75% MWE

_. O ABOVE WILL BE USED FOR 1/4 SCALE TEST PROGRAM HCOG TEST PROGRAM TO CONFIRM ADEQUACY OF DELIBERATE IGNITION HCOG TEST PROGRAM WILL NOT TEST FOR OPTIMUM IGNITION SOURCES BROOKHAVEN NAll0NAL LABORATORY l} g)l A5500ATED UNIVER$lTIES, INC.(1 ElI

, f~.2.3 f _ _

EFFECT OF STANDING WETWELL HYDP.0 GEN FLAMES 1

DETAILS PROVIDED BY DR PARCZEWSKI (NRC) IN APPENDIX A TO NUREG-1037 (CPWG REPORT)

HEATFLUXES.PROVIDEDBY[CLW6(NUREG-1079)

~

SEAL TEMPERATURES ARE SIGNIFICANTLY ELEVATED BUT REMAIN BELOW-FAILURE LATER IN ACCIDENT HIGH DRYWELL TEMPERATURES DURING CORE / CONCRETE INTERACTIONS MAY CAUSE SEALS TO EXCEED FAILURE LIMIT i

l BROOKHAVEN NATIONAL LABORATORY l} g)l

\ . A5500ATED UNIVERSITIES, INC.(8 til l

$US -. -- . -

F APPENDIX XXI - GESSAR II PRA REVIEW EFFECT OF INTEGRITY by T.APratt CORE MELT ON VESSEL ,

l t

1 SESSAR-II PRA REVIEW- - -

EFFECT OF A CORE MELT ON VESSEL SUPPORT INTEGRITY PRESENTED BY t

TREV0 H RATT BROOKHAVEN NATIONAL LABORATORY O UPTON, NEW YORK 11973 PRESENTED TO THE ACRS

. AUGUST 5, 1985

, BROOKHAVEN NATIONAL LABORATORY l} g)l

. A5500ATED UNIVERSITIES, INC.(ll.ll

. /J- .L J 7 _ __ _.__ _ _ _

O TOPICS l $i '

ABLATION OF SUPPORT ci N  :

T. . ~. -- -  :

SIGNIFICANCE OF LOSS OF CONTAINMENT INTEGRITY FOLLOWING SUPPORT FAILURE

. ~

EFFECT OF CONTAINMENT VENTING

~.- .. _.

l- -

BROOKHAVEN NATIONAL LAB 0RATORYl} g)l A5500ATED UNIVERSITIES, INC.(llll f edd f

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t GESSAR II SSER 2 .- 15-44

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gLATIONRATES 1

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j g=23 }@0NCRETE)(I ABLATION - I INITIAL) + k. ABLATIOf d j=10-20w/cM2 (LOWER CAVITY) y=12-3 w/cM2 (SURROUNDING) g = 2.5 p/cM , 3C = 1. J/,gM/K, A.* 240 J/Gn g = ABLATION RATE = 10-20 cM/HR (LOWER CAVITY)

INITIAL , __ ,

= 12-3 cM/HR (SURROUNDING)

M 10 HRS % 120 CM AXIAL ~

,, t 140 cM R,ADIAL, PEDESTAL INTEGRITY DOUBTFUL ERMAL GRADIENT fT= x8Tfg -

I. .

1 HR

• " ' ~

T INITI AL = 300K

+ 25 cM .

20 HR T INITIAL _ _ _ __ _

-< 60 cM  ;

O- .

h O fl

l

\

\

SIGNIFICANCE OF LOSS OF VESSEL SUPPORT MEASURE EFFECT RELATI'!E TO RISK ESTIMATES IN TABLE 15 9 0F SUPPLEMENT 11 TO SER (NUREG-0979) 4 EARLY LOSS OF CONTAINMENT INTEGRITY T. .~. -- .

LATE CONTAINMENT FAILURES (L2, L3) BECOME EARLY FAILURE (12, 13)

EARLY LOSS OF CONTAINMENT INTEGRITY PLUS LOSS OF DRYWELL INTEGRITY COMPLETE POOL SCRUBBING SEQUENCES (E3, 13, L3, B3) BECOME PARTIAL POOL SCRUBBING SEQUENCES (E2, 12)

BROOKHAVEN Nail 0NAL LAB 0RATORY[} l} l A5500ATED UNIVERSITIES, INC.(llll

.. [YJ

Table 15.1 Conditional consequences predicted by the staff for internally initiated events and probability of occurrence with and without UPPS, per reactor year Release Early 'Early Latent Probability

. category

• Person-fatality ' injury fatality rems ~

w/o UPPS w/UPPS 1-T-L3 0 0 40 7 x E5** 3 x E-6 9 x E-7 1-T-E3 0 0.0005 200 3 x E6 8 x E-6 1 x E-6 1-T-I2Q 0 3 200 3 x E6 1 x E-5 1 x E-6 2-T-R3 0 0 300 5 x E6 4 x E-6 4 x E-7 ATWS 0 1 400 6 x E6

. . 3 x E-6 3 x E-6 1-T-12 0' ' 6 500 8 x E6 3 x E-6 3 x E-7 1-SB-El 0.006 10 600 , 9 x E6 1 x E-9 1 x E-9

,See definitions in Table 15.15. .

• • 7 x E5 = 7 x 105 *'

00 2 ~

Notes:

h

-- ()

(1)

All conditional mean consequences were calculated using the upper range BNL source tem values. described in SSER 2.

(2) The calculations a~ssumed the Shippingport site, with public evacuation within 10 miles and relocation 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after plume passage.

(3) Mean consequences were computed over 91 different weather conditions.

G b '

GESSAR II SSER 4 15-31 hYf '

o g Tabla 15.9 Pubile risk from int:rnal events (person-ress per unit per year) fcr GESSAR II base case g and with design modifications m

O ~*

Unlimited u,

generator M

" Base UPPS and cand UPPS Unlimited GESSAR case perfect 10-hour

• DC Unilmited Iand perfect generator w/o Perfect with hydrogen battery charger UPPS and generator ihydrogen and UPPS Release

• UPSS Ha control UPPS control capacity generator igniters end UPPS ' control igniters 1-T-E2 3 -

0.5 -

1 , 1 .

0.3 U- -

1 1-T-E2Q 1 -

0.2 -

0.4 0.3 '

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- ?

4 1-T-E3 23 -

4 -

10 8

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9 7

. 1 - -

g 1-T-12Q 31 -

4 -

12 10 -

1 - -

b 1-T-I3 12 -

2 -

5 4 N 2 0.5 -

0.5 1-T-l.2 - - - - - -

k- -

% 1-T-L3 2 22 0.6 3 1 1

0. 6 - 0.5 2 0.5 1-SB-El 0.01 0.01 0.01 0.01 0.01 0.01 0.01 0.01 0.01 0.01 I; T-83 20 20 2 2 20 20 2 2 2 2 ATWS 18 18 IS 18 18 18 18 18 18 18 T,otal 131 59 33 23 76 68 31 25 22 25

• See Table 15.15 f- a description of the release categories.

e 9

O SIGNIFICANCE OF LOSS OF VESSEL SUPPORT (CONT.) ,

EARLY LOSS OF CONTAINMENT INTEGRITY:

GESSAR W/0 UPPS: 131 PERSON-REM PER YEAR

,. WITH EARLY LOSS: 139 PERSON-REM PER YEAR .

. .~.  : . .

EARLY LOSS OF CONTAINMENT INTEGRITY PLUS LOSS OF DRYWELL INTEGRITY:

GESSAR W/0 UPPS: 131 PERSON-REM PER YEAR WITH EARLY LOSS: 227 PERSON-REM PER YEAR O BROOKHAVEN NATIONAL LAB,0RATORYl} g)l A5500ATED UNIVERSITIES, INC.(lll t k3Y(

EFFECT OF CONTAINMENT VENTING

- "CLEAW" VENTING:

- ATTEMPTS TO MITIGATE CLASS 2 AND ATWS SEQUENCES

- MEASURE EFFECT RELATIVE T0 RISK ESTIMATES IN SER

- CLASS 2 S$QUENCES S'IENIFICANTLY REDUCED BY UPPS

~

- ABILITY TO MITIGATE ATWS BY VENTING UNCERTAIN

- VENTING AFTER CORE DAMAGE:

MINIMAL IMPACT ON EARLY H2 PHENOMENA HENCE Hz CONTROL NEEDED EVEN WITH VENTING BROOKHAVEN Nail 0NAL LABORATORY l} l3 l A5500ATED UNMRSITIES, INC.(llli

.- $*  ?

.y, g r"*fW -. ;- ,

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USI A-46 ACRS PRESENTATION AUGUST 8, 1985 SUNIARY OF USI A-46 PROGRAM T. Y. CHANG N. R. ANDERSON

~*

PROPOSED RESOLUTION, SCOPE AND BASIS g IMPLEMENTATION REQUIREMENTS STATUS OF ONG0ING SQUG/EPRI ACTIVITIES 5%

J. THOMAS (SQUG) ~A-3m 2 ANCHORAGE GUIDELINES .

EU' TEST DATA BASE DEVELOPMENT

$5 '

RELAY REVIEW PROCEDURE s

S006 GENERIC IPFLEMENTATION PLAN Es

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g- - - - - - -

O O O i

i 1

i BACKGROUND 1

l

• SEIMSIC SAFETY MARGIN IN OPERATING PLANT EQUIPMENT MAY VARY CONSIDERAILY

!

• SEISMIC QUALIFICATION OF EQUIPMENT IN OPERATING PLANTS NEEDS TO BE REASSESSED i

i l

• PROBABLY NOT PRACTICAL TO SEISMICALLY QUALIFY OPERATING PLANT EQUIPMENT USING i CURRENT CRITERIA

( -

!w%

• NEED TO DEFINE ALTERNATIVE METHODS i

1

• TASK A-f6l DE!,IGNATED AS USI IN DECEMBER 1980 i

l l i

i

t.

U V

( ,

)

SEISMIC QUALIFIg TION OF EQUIPMENT USING SEISMIC EXPERIENCE DATA STAFF ESTABLISHED FEASIBILITY OF USING EXPERIENCE DATA (LLNL STUDY)

SOUG CONDUCTED PIL0T PROGRAM 10 COLLECT AND EVALUATE SEISMIC EXPERIENCE DATA (SEPTEMBER 82)

ADDITIONAL. EXPERIENCE DATA COLLECTED FOR C0ALINGA, MORGAN HILL, CHILE EARTHOUAKES b

• S FORMED JUNE 1983, JOINTLY SELECTED BY SOUG AND NRC

\

k g

• SSRAP ISSUED REPORT IN JANUARY 1985 SSRAP DEVELOPED RULES FOR USE OF DATA FevJN5"b7 NRC STAFF PARTICIPATED IN DATA EVALUATION AND CLOSELY MONITORED SOUG/SSRAP EFFORTS PROPOSED STAFF POSITION BASED ON USE OF SEISMIC EXPERIENCE

Y~~ ,

O O '

O  !

SCOPE OF PILOT PROGRAM

' GATHERED AND DOCUMENTED EQUIPMENT AND EARTHOUAKE PERFORMANCE DATA FOR EIGHT CLASSES OF EQUIPMENT MOTOR CONTROL CENTERS LOW-VOLTAGE (480 v.) SWITCHGEAR '

METAL-CLAD (2.4 To 4KV) SWITCHGEAR UNIT. SUBSTATION TRANSFORMERS MOTOR-0PERATED VAI.VES AIR-0PERATED VALVES ll0RIZONTAL PUNPS AND MOTORS VERTICAL PUMPS AND MOTORS

• INVESTIGATED DATA ON e 3000 ITEMS OF EQUIPMENT IN CONVENTIONAL (NON-NUCLEAR) PLANTS

i.

SSRAP CONCLUSIONS FOR 8 EQUIPMENT CLASSES

• EQUIPMENT INSTALLED IN NUCLEAR POWER PLANTS IS GENERALLY SIMILAR AND AT LEAST AS RUGGED AS THAT INSTALLED IN CONVENTIONAL POWER PLANTS k

• THIS EQUIPMENT, FHEN PROPERLY ANCHORED AND WITH SOME RESERVATIONS, HAS AN INHERENT k '

SEISMIC RUGGEDNESS AND HAS A DEMONSTRATED CAPABILITY TO WITHSTAND SUESTANTIAL SEISMIC MOTION WITHOUT STRUCTURAL DAMAGE

• FUNCTIONALITY AFTER THE STRONG SHAKING HAS ENDEC HAS ALSO BEEN DEMONSTRATED, BUT THE ABSENCE OF RELAY CHATTER DURING STFORG SHAKING HAS NOT BEEN DEMONSTRATED

~

O O -

O

.t EQUIPMENT BEYOND 8 CLASSES NO REQUIREMENT FOR COLLECTING ADDITIONAL SEISMIC EXPERIENCE DATA ,

BASIS FOR SEISMIC ADEQUACY MUST BE DOCUMENTED FOR EACH EQUIPMENT TYPE, THIS CAN BE PROVIDED h BY: l i

<t kk -

VERIFICATION EQUIPMENT EXISTS IN DATA BASE PLANTS i TEST DATA CURRENTLY BEING COLLECTED BY EPRI/SQUG l  :

i t

l

Ar g g .

g e 4

9 4

THREE CONCERNS h 1. EQUIPMENT ANCHORAGES

(,

4 2. nEuv maoiuTv

3. OUTLIERS i

PROPOSED RESOLUTION .

OPERATING PLANTS DEVELOP EQUIPMENT LIST PERFORM WALK THROUGH INSPECTION .

)$ -

VERIFY ANCHORAGES k

q -

VERIFY FUNCTIONALITY OF EQUIPMENT (RELAYS)

D -

IDENTIFY a ADDRESS DEFICIENCIES AND OUTLIERS a

NEW LICENSEES NO REQUIREMENTS IMPLEMENT BY GENERIC LETTER 9

e _

O O O~

SCOPE OF SEISMIC ADEQUACY REVIEW ASSUMPTIONS SSE DOES NOT CAUSE LOCA LOCA DOES NOT OCCUR SIMULTANEOUSLY WITH OR DURING SSE 0FFSITE POWER WILL BE LOST DURING OR FOLLOWING SSE MAINTAIN HOT SHUTDOWN FOR A MINIMUM 0F 72 HOURS.

N 4

• EQUIPMENT SCOPE k -

ACTIVE ELECTRICAL AND MECHANICAL COMPONENTS INCLUDING INSTRUMENTATION AND CONTROLS NEEDED TO ACHIEVE AND MAINTAIN HOT SHUTDOWN

- ANCHORAGES ON TANKS, HEAT EXCHANGERS REQUIRED TO ACHIEVE AND MAINTAIN HOT SHUTD0WN

- NO REQUIREMENT T0 (1) REVIEW MASONRY WALLS, (2) REVIEW SOME AUX FEED SYSTEMS (3)

INSPECT RCS PIPING (4) REVIEW SEISMIC INTERACTION ITEMS PLANTS AFFECTED OPERATING PLANTS NOT REVIEWED TO CURRENT CRITERIA AS DOCUMENTED BY SER'S. ABOUT 49 SITES, 72 UNITS. SEP PLANTS WILL BE REVIEWED FOR FUNCTIONAL CAPABILITY ONLY

o o ~.

o i IMPLEMENTATION REQUIREMENTS

[ DEVELOP EQUIPMENT LIST VERIFY ENVELOPE OF SITE FREE FIELD SPECTRA BY APPROPRIATE BOUNDING SPECTRA I

i i

WALK-THROUGH INSPECTION _

ANCHORAGE REVIEW i -

IDENTIFICATION AND REVIEW 0F " DEFICIENCIES" AND "0UTLIERS"

! x.

IDENTIFY ALL EQUIPMENT THAT MUST FUNCTION DURING STRONG SHAKING -

j -

RELAYS ARE MAJOR CONCERN 1

{ REVIEW 0F EQUIPMENT UNIQUE TO NUCLEAR PLANTS REPLACEMENT PARTS

\

i i

)

I

~

y.

RELAY REVIEW GUIDELINES NRC GENERAL REVIEW GUIDELINES IDENTIFY ALL RELAYS ASSOCIATED WITH EQUIPMENT NEEDED TO BRING PLANT TO HOT SHUTDOWN RELAYS WHICH MUST FUNCTION DURING STRONG SHAKING: 3 VERIFY WITH TEST DATA REPLACE WITH QUALIFIED RELAYS QUALIFY BY icST b -

RELAYS WHICH MUST FUNCTION AFTER STRONG SHAKING:

VERIFY, REPLACE OR QUALIFY AS ABOVE OR LICENSEE SHOW CHATTER OR CHANGE OF_ STATE DOES NOT AFFECT PLANT SHUTDOWN RELAY VERIFICATION CAN BE DEFERRED UNTIL TEST DATA BASE COMPLETE

---

_a

t.

RELAY REVIEW (CONTINUED)

SQUG DEVELOPING REVIEW PROCEDURE IDENTIFICATION OF RELAYS TO BE EVALUATED DEFINITION OF FUNCTIONALITY REQUIREMENTS DEVELOPMENT OF EVALUATION PROCEDURES REVIEW BY NRC STAFF AND SSRAP ~

CHILEAN EARTHQUAKE CONFIRMS NEED TO REVIEW RELAYS

(

• SCOPE OF RELAY REVIEW

, TYPICAL BWR (DRESDEN/LASALLE) 1000/1200 RELAYS G/8 RELAY TYPES

, TYPICAL CE PWR (CALVERT CLIFFS)

• 1100 RELAYS

• 6 RELAY TYPES, 25-30 MANUFACTURES TYPICAL B&W PWR (OCOMEE)

• 750/900 GENERAL PURPOSE / INCLUDING PROTECTIVE RELAYS

• 25 MANUFACTURES

RELAY REVIEW (CONTINUED) '

TYPICAL.W PWR (Zion)

• 1100 RELAYS ,

• 7 RELAY TYPES w

O

_ __ _ _ - _ . _ . . . ~ _ . _ _ _ _ _ _ _ . ~_ . _ _ . _ _ . _ _ _ . _ _ _ _ _

APPENDIX XXIII - SQUG PRESENTATION SQJG CONCLUS10P6 SEISMIC ESISTANCE OF STN0ARD POWER PUWT EQUIRENT, 401 F90PERLY

1. DORED, WS VERIFIED DLRING TE PILOT PROGRN1.

EXPLICIT, SEISMIC QUALIFICAT10N OF THIS EQUIRER IS ET JUSTlFIED.

SEISMIC QUALIFICATION IS 10T A SimlFICNR SAFETY CDNCDN, TEREFOPE, FURTER ACTION IS PCT EQJIPED.

O O

O 4w-

, gm t.x_/)

SUBJECT:

QUESTIONS FOR OPERATOR INTERVIEWS DURING TH IN CHILE

1. What was plant status prior to earthquake?

• What are

2. Does the plant have any special earthquake procedures?

they? Are copies available?

3. During strong motion in the event did automatic action If auto- of the plant systems take place? What were these automatic actions?

In the absence of 4.u tc-What action did not take place, should it have? action (based on a "

What alarms were ittitiated? or manual)?

action did he take?Did the plant respond properly (auto infomation? Were

4. Af ter the strong motion was over, what was the Wereplant status?

systems any auto-actions which were needed taking place?Was the operator resetting action?

to normal?If so, what action did he take and did the plant respond a it was supposed to?

O 9 Was off-site power lost?

(v)

' 8 8

Was auxiliary power lost?

Was diesel power available?

e Was d.c. power available and was load shedding required?

e Did power distribution (internal) respond to loss of power or to relay chatter?

5. In determining failures or damage to equipment: were there any misoperations or malfunctions of equipment of a mechanical Were ,

nat

?

(Were there breaker trips which were not elect e What type of equipment? Type of relay?

8 9

Any damage due to induced improper Maintained system 8 Any problems with momentary contacts on switches?

contacts? Problems?

8 Any mercury switches?

9 9

Any system change of state not attributabl Were there problems with cables or cable terminations?

6.

Were there structural failures which affected systems function? Inst Large pipes 2 1/2" small? Pipe supports?

air? Other degraded Were there reduced or increased flows in cooling systems?

4 I

O)

(

b 7,

functions?

CRT's?

8. Any damage to Control Boards?

k 'o&.L

1 l

9. What worked that wasn't expected to work?

What failed that was expected to work?

10. What people related problems were experienced? Access to tools, procedures, damage control equipment, connunications, etc.

11. What secondary events occurred? Fires? Spills? * -

12. Any problems with equipment in operation? Cranes? Portable equipment? Maintenance in progress?

13. Is there' seismic monitoring equisment at the site?

14. For problems encountered at power plants, were they considered a

" systems" problem or an " operation" problem?

'9 Steam cycle?

8 Condensate?

8 Feedwater?

8 Power?

O O

4 su

a Questions related to switchyard functions:

1. Have there been any design changes to prevent vibrations from triggering fault pressure relays when there has been no system damage that would require action?

2.

Are there any special switching arrangements developed for earthquake response? Any line isolation provisions?

3. Are there any special provisions for starting if off-site power is lost due to the earthquake? (black start)

4. Were any problems encountered in synchronizing the plant with the transmission system after the earthquake?

5.

Were there any degradations in the relaying communication system?

What type system - microwave or carrier? Any special procedures related to degraded comunications?

6. Does dispatch system exist? Did it create any problems?

4 O

e O

O O

+m

, SQUG PROGRAM OUTLINE

1. SCREEN ESSENTIAL EQUIPMENT LIST .

O COVERED IN SQUG PROGRAM ,

0 OTHER DATA AVAILABLE (EXPERIENCE, TEST)

"O ENGINEERING JUDGENT

2. DOCUMENT SEISMIC ROGGEDESS OF EQUIPMENT 0

ASSIGN RUGGEDNESS LEVELS WICH CAN BE ' JUSTIFIED 0

IDENTIFY EXCEPTIONS /Ytt.NERA8ILITIES FOR EACH EQUIPMENT CLASS O O DEFINE DATA NEEDS. IF ANY

3. C0frLETE/REYlEW EMI PROGAAMS, DEVELOP ANCHORAGE litSPECTION EWIDEllES o AllCHORAGE o

TEST DATA ASSIMILATION O

Jar

. 4. DEVELOP SIMPLIFIED APPROACH FOR DETERMINING REQUIRED SEISMIC RUGGEDNESS IN NUCLEAR PLANTS 0 ELEVATIONS LESS THAN 40 FEET 0 HIGHER ELEVATIONS

5. ATTEMPT TO LIMIT SCOPE OF RELAY FUNCTIONALITY REQUIREMENTS ON GENERIC BASIS 9

6. DEVELOP PLANT WALK-THROUGH GUIDELINES AND TEAM

7. PERFORM ' TEST' WALK-THROUGH

, 8. DEVELOP PLANS FOR SQUG MEMBER IWLEENTATION 0 ,>

SEMINARS 0 GENERIC SOUG TEAM APPROACH 0 SSRAP/NRC AUDIT

c SQUG ACTIVITIES PROMPTED BY ACRS/ STAFF CONCERNS

1. DEFINITION OF GENERIC EQUIPMENT REQUIRED TO ACHIEVE SAFE SHUTDOWN.

2. DEVELOP RATIONALE TO ASSURE SEISMIC RUGGEDNESS OF EQUIPMENT BEYOND THE 8 CLASSES DEFINED IN THE PILOT PROGRAM.

3. FUNCTIONALITY DURING STRONG-MOTION (PRIMARILY RELAYS)

4. S006'S GENERIC IMPLEMENTATION PLAN

. EQUIPMENT SCREENING ,

. PLANT WALKDOWN PROCEDURES O . AUDIT FUNCTION O

1. w7

I APPENDIX XXIV -

SUMMARY

OF THE EFFECTS 0F THE GREAT CHILE EARTHQUAKE OF 1985 i f 1 u 1 i b Meeting No. Agenda Item Handout No.

3, l c7.0 c-

) s',\ W, '-I l ()E T1.fc (f E(-(7) ON $l( CCM,i C M 6 E - M.le. . t or 1%5 Authors .

< 1.' E \ ', n b 0 List of Documents Attached r

l l . a C Y-'

Instructions to Pr epa rer From Staff Person

1. Punch holen e

2. Paginate attachments . (ffhdC-(,l

3. Place conv in ffle box m j

l

-i l

SUMMARY

OF THE EFFECTS OF THE i

GREAT CHILE EARTHQUAKE OF 1985 l l

- l em, --

l j .. . _ 4

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I l Damage to Feedwafer Pipe Supports l Pipe was not damaged j

( 1.as Ventanas Power Plant t l Chlte Earthquake, March 1985 .

I Magnitude 7.3 l I

Presentation to: l l

SQUG and NRC 1 Williamsburg, VA August i and 2,1985 Peter I. Yanev & Paul D. Smith k '*2$ f -

4

• O V THE MAGNITUDE 7.8 CHILE EARTHQUAKE OF MARCH 3,1985 OCCURRED IN ONE OF THE WORLD'S MOST SEISMICALLY ACTIVE AREAS l% l _! l/ ,

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' The dark areas on thu ma in-l s =- ' ' /, dicate the dutribution an den.

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1 tity the 42,000 earthquakes

1985 EPICENTE8 ft .Jrecor

. _j. from 1961throughout to 1970.

• f

• the world MAGNITUDE 7.8; LARGEST AFTERSHOCK 7.2 AFFECTED AREA: 20,000 SQUARE MILES POPULATION: 6 MILLION DEATHS: 180 HOMELESS: 250,000 DAMAGE: $1.8 DILLION DWELLINGS DESTROYED: 45,000 gN

.p U SQUG SENT TWO TEAMS TO INVESTIGATE THE EARTHQUAKE RECONNAISSANCE TEAM (MARCH 7-14,1985)

PAUL D. SMITH - EQE DAVID L. McCORMICK - EQE RENE W. LUFT - SG&H LORING A. WYLLIE, JR. (SSRAP); PART-TIME C\

INVESTIGATIVE TEAM (MAY 16-24, 1985)

PETER I. YANEV - EQE , , ,

j',%,,f DENNIS K. OSTROM - SQUG/SCE JAMES E. THOMAS - SQUG/ DUKE DAVID L. McCORMICK - EQE STEPHEN HOM - EQE MARY D. MUSULIN - EQE/EPRI ANSHEL J. SCHIFF - SSRAP

/

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() THE AFFECTED REGION IS THE MOST DENSELY POPULATED

    .v AREA IN CHILE; ABOUT 200 MILES BY 100 MILES s                i          t   '

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() THE AFFECTED REGION IS THE MOST DENSELY POPULATED AREA IN CHILE: ABOUT 200 MILES BY 100 MILES n- n. ,,. ,,. ,,,

                                                                                                   /                        .

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                                                                                '" Wk                                         e e ense tv mateJones. ,

e sites (0S vtLOS N' up - LaeNI#tifv Op Csatt

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                                                                \  ~         f 5* * ' L'M ese.s+wn, oa tas seemsta=4                      pa w                %

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                                                                                          ,,g, nataosuo          /

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               }              as awa'J             I                     UI*A* A              9                 \

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4. Pata tt asu $Jae l Safe pl*% Ps00 i . 1 AT LEAST 31 STRONG MOTION RECORDS WERE MADE THROUGHOUT THE AFFECTED AREA O V E k->73

       ,o

( )

SUMMARY

OF STRONGER PEAK ACCELERATION RECORDS (G) LOCATION H H V AVE H

1. PAPUDO 0.13 0.47 0.11 0.30

2. ZAPALLAR 0.28 0.26 0.18 0.27

., 3. LAS VENTANAS P.P. 0.18 0.18 0.14 0.18

4. SAN FELIPE 0.35 0.43 0.21 0.39

5. LLAYLLAY 0.34 0.49 -- 0.41 (m '

6. SAN PEDRO (S.S.) 0.60 0.57 0.38 0.58

7. VINA DEL MAR 0.23 0.36 0.20 0.30

8. VALPARAISO (1) 0.16 0.29 -- 0.22

9. VALPARAISO (2) 0.17 0.19 0.1 't. 0.18

10. PELDEHUE 0.64 -- -- 0.64

11. QUINTAY 0.20 0.18 0.13 0.19

12. SANTIAGO 0.11 0.11 0.07 0.11 i

13. LLOLLEO 0.43 0.67 0.86 0.55

14. MEllPILLA 0.67 0.68 0.34 0.67

15. RAPEL P.P. 0.31 0.14 0.11 0.23

16. PICHILEMU 0.43 0.23 0.14 0.33

(D

'u

                                  /> -> 7/

                /

m \ em U TYPICAL STRONGE] RECORDS E VE]Y LONG DURATIONS (v')

SAN FELIPE (H = 0.35 AND 0.43; V = 0.21) (~ rice nar

    .0,'%' & V M A W W lY N,ri,k );**y ,,; ,f[,Y. 3E              i ,} ll*T-Y/YM,"f/47e'4W/fMN*MdWIN/b/ f)b d             (

i .c  : -c_' :. .l:. c.w4.:.aa~ 9mv.%'uyW.=,y,*h,9.7;4f.cs:%/+,w,~mme-e:+mw.:==, : % g ,

   'W=+*L..+,~%+A~W/$.'[,0\,Wh,i,,t'!s**,';'t\,
                                                         , j, ',E \,$k,h? l$fQ.1690%W(M.*;%'#MW'.QffM,tC's                     _

0 0 5 10 NR" "# '15 - 20 . 25 - 30 SEC __L.......,_......_,_......,_......,... _ __ 4 4 k

                              $ & f.* & & W m :%.-                       "vvl.e d w w .'

f

       \                         -
                                               ~

m :_-_ + m e - = -

                                                                                =_z- _ - _ _ = _   .

d.%v,4,%%+.%/e ,+,v ,ws.we-

                              . . .     ,35.. . .. . ,40
                                                                               .,4 5              5,0 SEC            j j

h.#' 8).U

                                                                                                                      / \

w ONE OF THE STRONGEST RECO S IS AT LLOLLEO / H= 0.43g AND 0.67g: V = 0.86g

              . :Q%.,:::.'.6f:.':'-l!?,W:,m.~vNr!?M '<.4lGl't,'d ll'jlidp'5gA:(.'i ?'::'*l,'i& '1; y~:::.~;, ~;::;-
                                         .. e.-~e.~;yy7.;v.~.w q:ay_svll,1_flll&i,Q,.'$'l.l0g,Wlk;f,%\N
              .,.;:;:i,i:lJf\cwjtl.<lwwtw; ppg.).,y.57.I,pf,tgt,\ffgi-                               '

O 5 10 15 20 I 25 SEC r _ .c_ - n - - - :- - u -- - - - - _ r- _ - -.- -- - .r :- ---.- - _-- h b '.>ls':%l;@n4y+s1:+wn~~myw .~- 910L' ..'idlN1.Ki.U,'.'"r},\'.'%'sN$&,w.wMswMee:.~ i f : *- @ l l ll. q \ l ll M y f v : n'i y h w n w ." w N v-30 35 40 45 50 SEC

                     ---------c_______~_____,__._-__,.___.

sahD.

                                                              .j O    WE VISITED ABOUT 40 SITES THROUGHOUT THE STRICKEN AREA, INCLUDING:

6 POWER PLANTS - 3 SUBSTATIONS - 2 VERY LARGE REFINERIES 1 CHEMICAL PLANT 3 WATER TREATMENT PLANTS 5 COMMERCIAL FACILITIES THE 6 POWER PLANTS HAVE 13 UNITS LAS VENTANAS (LV) 2 UNITS LAGUNA VERDE 2 UNITS (cf

                                              /0)

RAPEL (HYDRO) 5 UNITS LAS VENTANAS COPPER 1 UNIT CON CON REFINERY 1 UNIT RENCA 2 UNITS ms

                              #-a n

i ? SQUG/EQE HAS NOW INVESTIGATED THREE DESTRUCTIVE EARTHQUAKES AFTER THE COMPLETION OF THE PILOT PROGRAM o MORGAN HILL,1984, M=6.2 SMALL MAGNITUDE

          -   HIGH ACCELERATIONS
          -   LIMITED DAMAGE
          -   LIMITED FACILITIES o  COALINGA,1983, M=6.7
           -  MODERATE MAGNITUDE
           -  HIGH ACCELERATIONS OVER LIMITED AREA SEVERE DAMAGE
           -  MANY FACILITIES
           -  STRONG AFTERSHOCKS, UP TO M=6.0 o   CHILE,1985, M=7.8
           -  GREAT MAGNITUDE HIGH ACCELERATIONS OVER A LARGE AREA MANY FACILITIES AND POWER PLANTS
           -  STRONG AFTERSHOCKS, UP TO M=7.2, WITH HIGH ACCELERATIONS nos
                             +-27r

APPENDIX XXV - NUCLEAR REGULATORY MAINTENANCE & SURVEILLANCE PROGRAM NUCLEAR REGULATORY COMMISSION MAINTENANCE AND SURVEILLANCE PROGRAM . PROGRAM MANAGER DR. HAROLD R. B00HER, CHIEF ' LICENSEE QUALIFICATIONS BRANCH DIVISION OF HUMAN FACTORS SAFETY /NRR GREGORY C. CWAliflA, SECTION LEADER MAlf4TENANCE/ SURVEILLANCE SECTION 4 O k -4 7 f

i . k .I ' f

t. I t-i

i. .

2 i

   ?

l i 4 t 1 OUTLINE  ; i .! i OBJECTIVES & SCOPE i CURRENT STATUS I l PROGRAM

SUMMARY

                ~

9 PHASE I PROJECTS f SURVEY PROGRESS .  : i I i I  ! l e l lO i l N-M

    ' (~x_-)                   MAINTENANCE PROGRAM DEVELOPMENT NOV 83   NRC MAINTENANCE WORKSHOP MAINTENANCE INDICATOR PILOT STUDY INITIATED JAN 84   COMMISSION POLICY AND PLANNING GUIDANCE MAY 84   US/ JAPANESE STUDY PART I COMPLETED ACRS BRIEFED (MAINTENANCE SUBCOMMITTEE AND FULL COMMITTEE JUN 84   DRAFT PLAN TO NUMARC ASME BRIEFED PUC BRIEFED JUL 84   IEEE BRIEFED (3      SEP 84   ANSI BRIEFED
       'G             '

OCT 84 UPDATED PLAN TO NUMARC ACRS - MAINTENANCE SUBCOMMITTEE - JAPANESE STUDY BRIEFING DEC 84 MAINTENANCE INDICATOR PILOT STUDY COMPLETED

;              JAN 85   PLAN PHASE I APPROVED NUMARC PROPOSED INDICATORS NRC INDICATOR TASK FORCE FORMED MAR 85   REVISED MSPP (NUMARC COMMENTS) f               APR 85   SUBMITTED TO COMMISSION (SECY-85-129)

MAY 85 SURVEY PROJECT INITIATED (') JUN 85 COMPLETED REGIONAL BRIEFING / COORDINATION ACRS - MAINTENANCE SUBCOMMITTEE

     . - . . .           _ -____--_ - _ ___ _ __ _ _ _ _ _ k_ _"d N

O G - NRC SPECIFIC OBJECTIVES DETERMINE EFFECTIVENESS OF CURRENT MAINTENANCE PROGRAMS IDENTIFY PRACTICES WHICH REDUCE HUMAN ERROR RATE IN PERFORMANCE OF MAINTENANCE . IMPROVE EFFECTIVENESS OF MAINTENANCE PROGRAMS IN ASSURING k OPERABILITY OF SAFETY SYSTEMS REDUCE UNNECESSARY AND UNANTICIPATED RADIOLOGICAL EXPOSURE TO MAINTENANCE PERSONNEL DETERMINE REGULATORY APPROACH TO ASSURE EFFECTIVE MAINTENANCE PERFORMANCE O

                           ~

in

c __.. . .. MSPP SCOPE ALL ASPECTS REQUIRED TO CARRY OllT A SYSTEMATIC MAINTENANCE PROGRAM SURVEILLANCE AND TEST ACTIVITIES EQUIPMENT REMOVAL FROM/ RETURN TO SERVICE POST-MAINTENANCE TESTING MAINTENANCE MANAGEMENT / ADMIN. CONTROL PERSONNEL SELECTION, QUALIFICATIONS, TRAINING PROCEDURES DOCUMENTATION j THOSE COMPONENTS WHICH AFFECT PERFORMANCE OF SAFETY SYSTEMS l lO j - 4 aV2

              ._..__.__ _ _ _ __ __. _ __. . _ _ _ _ _ _ . _ _ _ _ _ . . . _ _ _                         _ - . .. ~___
    - . . _                           s                                  ,.

> l l 1

IDENTIFIED PROBLEMS I

1  : i e i  ! j- t i

1. MAINTENANCE PERFORMANCE .

i 6 L

2. FAILURES DUE TO IMPROPER PERFORMANCE .

l l l , t i ! 3. MAINTENANCE /0PERATIONS INTERFACE 1  ; I i I l

4. CHALLENGES TO SAFETY SYSTEMS ,

4 i

i. l t 5. OCCUPATIONAL EXPOSURES 1

i t 4 j

i. ,

I.

7 _ . _- . . i 4 1 2' 1 , 3 2 i 1- j 1 , i i j STRATEGY l

;                                                                                                                                        i i-                                                                                                                                        ;

{ BROAD SCOPE t 4 i 1-1 FOCUS ON TECHNICAL ISSUES i  ! i, i j i i I-USE PHASED APPROACH - IOl INTEGRATE STAFF ACTIVITIES i j l ) A 4 . -I COORDINATE INDUSTRY INITIATIVES l f b I' t i s.

l t

h I  ! l

.t i l I

I k 'O~  ! I i l- - l M -

l l l FIGURE 3.1 PSOEMM Ple&5ES PSOERM 3 1RITIATIOR/ % YEAR 1 YEAR 2 , TEAR 3 M TER 1 2 3 4 2 3 4 i 1 1 2 3 4 Phase I Survey sa. Evaluation A A i EDO Aeview AA I 8 Phase.< Pv. !!. i Identification

i i

       .                   Ispacts lA                                                .A 1

A, T E brier J k .

                                                                                          ,                                                    e                               ,

men u ru l l 1

                      ==c                         -                 A                    A                                                     A                            A                       l
                      -                                                   A              I,                                                    i amours                                                                                                                ,

190 A A A A ! mr A i i i ( _ _ _ ,c.,_,_-- , - - - - - - - - . - - - - - - - - - ' ' - - ' " - ' ' - ' - - ~ ' - - - - ' - - -- '

A O . PHASE i PROJECTS

1. SURVEY OF CURRENT PRACTICES

2. MAINTENANCE PERFORMANCE INDICATORS e

3. MONITOR INDUSTRY ACTIVITIES

4. PARTICIPATE IN STANDARDS GROUP

~

5. PROGRAM INTEGRATION (NRC AND INDUSTRY)

6. ANALYSIS OF JAPANESE /U.S. Mt.INTENANCE PROGRAMS

7. MAINTENANCE PERSONNEL QUALIFICATIONS

8. H.F. IN IN-SERVICE INSPECTION -

9. HUMAN ERROR IN EVENTS INVOLVING WRONG UNIT OR WRONG TRAIN (GENERIC ISSUE 102)

O sm

     , _ , = . ..    . _.

a 4 . SURVEY OF CURRENT MAINTENANCE PRACTICES 1\ r i

                                                                                                                   .i TASK 1      -

DATA COLLECTION AND ASSESSMENT i , i TASK 2 - SALEM PREVENTIVE MAINTENANCE PROGRAM l

TASK 3 QUESTIONNAIRE i

;                           TASK 4      -

MAINTENANCE REVIEW PROTOCOL

)

l TASK 5 - SITE SURVEYS L 1 TASK 6 -

SUMMARY

REPORT i

• k

   .                                                                                                                 6 4

I t i f l '! L t Ii I i i 4 i ! l

A-a/t

a " E 2

                                  .          i V

M .s O V # 11 ti i"

                                      'i 't 11 ik 1

t1  ?; z *i 11 h. s W J 3 2 c W "a 1 Z I U

        *                            $.;                     a
                                     -if                     *t E
                           , ,       II

{!

        <        li       II I           '
       =

u cA i 1.\ !] sli 5

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                    \

[-

                                                    "*       }#

e O. - W I ej i N k h 15 IE -

                               .h 5                                       ,

e  : ,3 <o.

                          ,,l sa
                                    $k 1

u {!

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al I - r

• 2

       "                                           }-
                                               =
                                               <   11
                                    .          7't 11 E                 1
                 ?                     i
                                   .S k

1 3 ga

( '

2. MAINTENANCE PERFORMANCE INDICATORS SCOPE:

                            *' MONITOR INDUSTRY INDICATORS

• DEVELOP NRC INDICATORS IF WARRANTED I

i MILESTONES: PILOT STUDY, PNL, JAN - APR 1984 ENDORSEMENT OF NUMARC INDICATORS, MAY 1985 [} , VERIFICATION OF INDUSTRY DATA DECISION ON NEED FOR NRC INDICATORS i l i l t O ( /}-.1.16

' C} NUMARC - MAINTENANCE PERFORMANCE INDICATORS I. DATA AVAILABLE THROUGH EXISTING SOURCES

1. UNIT FORCED OUTAGE RATE

2. UNIT EQUIVALENT AVAILABILITY

3. N0. 0F UNPLANNED AUTOMATIC SCRAMS DUE TO MAINTENANCE II. DATA TO BE OBTAINED FROM UTILITIES

1. TOTAL RADIATION RATE PER UNIT DUE TO MAINTENANCE i

  )      2.. OVERTIME WORKED BY MAINTENANCE PERSONNEL

3. LOST TIME ACCIDENT RATE N0. FOR MAINTENANCE PERSONNEL

4. AMOUNT OF OUTSTANDING NON-0UTAGE CM WORK

5. RATIO 0F HIGHEST PRIORITY NON-0UTAGE CM WORK REQUESTS

6. PM ITEMS OVERDUE

7. RATIO OF PM TO CM III. UNDER DEVELOPMENT

1. PERCENTAGE OF MAINTENANCE REWORK

2. SAFETY SYSTEM AVAILABILITY O

V .

                             &-4/

l - i

3 .' MONITOR INDUSTRY ACTIVITIES  !

SCOPE: i SUMMARIZE FINDINGS

ASSESS APPLICABILITY DOCUMENT ACHIEVEMENTS MILESTONES

;

I

\                                                                                                                                ,

i C0 ORDINATION WITH NUMARC  ; JUNE, OCTOBER 1984 ( i JAN, FEBRUARY 1985 l ATTEND EPRI SEMINARS, MARCH, APRIL 1985 l - l REVIEW EPRI/MIT REPORT, JUNE 1985 ( i l l I i O l M M/h I

i OL> -

                                                                                                  .          i

4. PARTICIPATE IN STANDARD GROUPS SCOPE:

ENC 0URAGE INDUSTRY INITIATIVES e PROVIDE NRC CONTRIBUTIONS MILESTONES: i ANS 3,9  ! SEPTEMBER 1984 - SUBCOMMITTEE FORMED O MARCH 1985 OUTLINE CIRCULATED i 1 FALL 1985 - NUMARC DRAFT DUE ASME OPERATIONS AND MAINTENANCE COMMITTEE APRIL 1985 - NO PM STANDARD SPRING 1985 - SEMINARS COMPONENT STANDARDS DEVELOPMENT CONTINUES IEEE W.G. 3,3 -

                                                                                 " PRACTICES" 1

OCTOBER 1984 - W.G. FORMED SPRING 1985 - SCOPE DEFINITION 1

j' I i i 1 i !O . i i 5. PROGRAM INTEGRATION  ; i i SCOPE: i i

IDENTIFY RELATED PROGRAt1S  !

       .                                           ItiTERPRET CONTENTS AND SCHEDULES                                                l 4

PREVENT OVERLAP l

PRESENT UNIFIED POSTURE l , . L i MILESTONES: i t

t. -

!

• L a

PROGRESS REPORTS, FINAL REPORTS ARE BEING  ! T i REVIEWED  : c i P i i t 1 t }  ! i t i i t l l I I .  !

,                                                                                                                                   i i
                                                -                                                                                   i i

1 }

• NRC PROGRAMS RELATED TO MAINTENANCE RESPONSIBLE ORGANIZATION QUALITY ASSURANCE PROGRAM, R.G. 1.33 IE SYSTEMS IMPORTANT TO SAFETY IE

       ' SAFETY IMPLICATIONS OF CONTROL SYSTEMS (USI A-47)                                         NRR COMPREHENSIVE REEVALUATION OF STANDARD TECHNICAL SPECIFICATIONS                            RES SURVEILLANCE AND TEST REQUIREMENTS O     (ECCS OUTAGE CRITERIA)                              RES          ,

NUCLEAR PLANT AGING RESEARCH RES EFFECTIVENESS OF INDUSTRY ALARA PROGRAMS NRR EQUIPMENT QUALIFICATION - R.G. 1.89 NRR RELIABILITY RESEARCH RES IMPROVING OUALITY IE TRAINING RULE - SECTION 306 WASTE ACT NRR O

                                  & .21f

6. ANALYSIS OF JAPANESE /U.S. MAINTENANCE PROGRAMS SCOPE:

COMPARE OPERATING EXPERIENCE COMPARE MAINTENANCE REQUIREMENTS ANALYZE ORGANIZATION 4ND MANAGEMENT ! MILESTONES: APRIL 1984 - FROGRAM INITIATION I MAY 1985 - PROJECT COMPLETED iO JULY 1985 - NUREG/CR-3883 AND 3883P PUBLISHED I l 5 l lO l

( '

7. NAINTENANCE PERSONNEL QUALIFICATIONS NEED TO DETERMINE THE KSAs REQUIRED FOR MAINTENANCE JOB TASKS T0, IDENTIFY THE RELEVANT SOURCES OF THE KSAs IN TERMS OF O' -

EDUCATION, TRAINING, AND APPRENTICESHIP PROGRAMS TO CONDUCT AN ANALYSIS OF APPLICABLE INDUSTRY GUIDELINES AND STANDARDS AGAINST JOB RELEVANT KSAs AND SOURCES O 4997

4m e'f- w ash M E t 1 ,1 l . i

8. HUMAN FACTORS OF IN-SERVICE INSPECTIONS ,

i i i 3 L i SCOPE: t IDENTIFY HUMAN ERROR POTENTIAL i I  ! MILESTONES: l { i [ j MAY 1985 - PROJECT INITIATION

.                                                                                                                    l

[

JULY 1985 l

DATA COLLECTION i (  ; t j DECEMBER 1985 - FINAL REPORT i b I [ [ I t i t i I t i i i i I i t

l.  ?

l

1._ _ . 9.- HUMAN ERROR IN WRONG UNIT / WRONG TRAIN EVENTS SCOPE: IDENTIFY PROBLEMS ANALYZE ROOT CAUSES DEFINE ACTIONS FOR RESOLUTIONS MILESTONES: JANUARY 1984 - AE0D REPORT MAY-JULY 1984 - AEOD/NRR COORDINATION JUNE-DECEPSER 1985 - SITE VISITS - JANUARY 1986 - FINAL REPORT O ser

( PROGRESS NRC MAINTENANCE SECTION FORMED TASK FORCE REVIEW NUMARC DRAFT INDICATORS PARTICIPATION NEW STANDARDS EFFORTS L

• REVIEW JAPANESE MAINTENANCE /0PERATIONS EXPERIENCE

       ~

PARTICIPATION IE INSPECTION FEEDBACKNRCREGIONAL/RE51DENTiNSPECTORS INDUSTRY STANDARDS EFFORT INITIATED (IEEE, ANS) ANALYZE MAINTENANCE-RELATED CONTRIBUTION TO EVENTS REVIEW / REVISE INP0 PLANT EVALUATION OBJECTIVES / CRITERIA IN MAINTENANCE AREA DEVELOP PRELIMINARY PERFORMANCE IND.ICATORS REVIEW INP0 DRAFT MAINTENANCE GUIDELINES EVALUATE STATE OF MAINTENANCE IN INDUSTRY A U a -acro

I l I T l ! i i  ! l i i t 1 i i e i i . L 1 e i v. - ,, l

                                                 ,.f  /                             >

l 6 , e f la e #jg jgs y <op ca <y a n,. L l l 4 i }

• i i t

h t I i \ l i > i t f 4 I I l l 1 1 , 1

• i f .

1 i I F l i

n, 5 APPENDIX XXVII - ADDITIONAL DOCUMENTS PROVIDED FOR ACRS' USE ADDITIONAL DOCUMENTS PROVIDED FOR ACRS' USE

1. Memorandum, R. F. Fraley to ACRS Members, ACRS Review of Alternate Proposals for NTSB-Type Review of Nuclear Facility Accidents, August 9,1985

2. Report, W. R. Stratton to Dixy Lee Ray, Chairman, U.S. AEC, Report on Alvin W. Vogtle Nuclear Plant, Units 1,2,3, and 4 April 16,1974 a

N

                                        #-Jo/}}