ML20205C834

Minutes of 318th ACRS Meeting on 861009-11 in Washington, Dc.Meeting Agenda & Apps Encl

ML20205C834
Person / Time
Issue date:	03/25/1987
From:	Advisory Committee on Reactor Safeguards
To:	Advisory Committee on Reactor Safeguards
References
ACRS-2468, NUDOCS 8703300274
Download: ML20205C834 (167)

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MINUTES OF THE 318TH ACRS MEETING W ,; M.)

g N%j/i4 d3MW OCTOBER 9-11, 1986 WASHINGTON, D.C. ,Moggg I. Chairman'sReport(0 pen)...................................... 1 II. SeabrookNuclearStation(0 pen)............................... 2 III. Resolution of Outstanding Issues on Clinton Power Station, Unit 1 (0 pen)................................................. 9 IV. Reacto r Ope ra tions (0 pen) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13 A. Loss of Low Pressure Service Water at Oconee Nuclear Station, Unit 2........................................... 14 B. Loss of Offsite Power Test at Hope Creek.................. 15 C. Transient Overloading at Salem-2 Station Transformers..... 18 V. Activities of the Office of Nuclear Material Safety and Safeguards (0 pen)............................................. 20 VI. Backfitting of Regulatory Requirements (0 pen)................. 24 VII. NRC Standard Review Pl an (0 pen) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 27 VIII. Meeting with Executive Director for Operations (0 pen)......... 28 IX. International Operating Experience (0 pen) . . . . . . . . . . . . . . . . . . . . . 30 X. Executive Sessions (0 pen)..................................... 34 A. Subcommittee Assignments.................................. 34

1. Instructions for the Nominating Panel................ 34 B. Reports, Letters and Memoranda............................ 34

1. ACRS Suggestions for an NRC Long Range Plan.......... 34 1

! 2. ACRS Comments on Draft NUREG-1225, " Implementation of NRC Policy on Nuclear Power Plant Standardization"... 34

3. Clinton Nuclear Power Station - Resolution of ACRS C onine n t s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 34 4 IDCOR Documentation Availability and Staff Review.... 34 DESIGNATKD ORIGINAL 8703300274 870325 8 PDR Certified Br_.

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5. Comments on Proposed Revisions to Standard Review Plans 6.5.2 and 6.5.3............................... 35

6. Containment Performance Issues...................... 35 C. Generic Issues

1. . Basis for Nuclear Plant Improvements................ 35

2. Interpretation of NRC Safety Goals in Terms of the Population Doses Associated with Nuclear Power Plant Accidents........................................... 35

3. Discussion of Significant Safety Issues............. 35

4. Authorization of the Backfit Rule................... 36 D. Future Agenda............................................ 36

1. Future Agenda....................................... 36

2. Future Subcommittee Meetings........................ 36 E. Seabrook Station Probabilistic Safety Assessment, Risk Management and Emergency Planning........................ 36 F. Shearon Harris Status Report............................. 36 G. Proposed Revi sion of ACRS Byl aws. . . . . . . . . . . . . . . . . . . . . . . . . 36 H. Reti rement of ACRS Member Eme ri tus . . . . . . . . . . . . . . . . . . . . . . . 37 I. No ti ce of Awa rd s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 37

, Attachments 1, 2, and 3 pertaining to G. Proposed Revision of ACRS Bylaws................................................... 38,39,40 Supplement - Nuclear Plant Security

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4 iii TABLE OF CONTENTS APPENDICES TO MINUTES OF THE 318TH ACRS MEETING OCTOBER 9-11, 1986 A-1 Appendix I Attendees.................................................

A-7 Appendix II Future Agenda.............................................

A-9 Appendix III ACRS Subcommittee meetings................................

Appendix IV Public Service of New Hampshire Presentation on Seabrook A-14 Station...................................................

Appendix V Introduction to Staff Presentation on Clinton Power A-51 Station...................................................

Illinois Power Company Presentation....................... A-54 Appendix VI A-59 Appendix VII QA/QC, Overinspection'and Audit Programs..................

Appendix VIII Clinton Power Station Equipment Seismic Assessment A-65 Program...................................................

A-70 Appendix IX Recent Significant Events.................................

Appendix X Fuel Cycle Facility and Materials Licensing Division, A-102 NMSS......................................................

Division of Safeguards - Security......................... A-106 Appendix XI Appendix XII International Operating Experience - Comments on Chernobyl A-113 Meeting...................................................

Use............... A-118 Appendix XIII Additional Documents Provided for ACRS i

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4 P Wir Participot.. i:. Unde . . CF'R NUCLEAR REGULATORY .

representatise of. foreign governments r.

1153 20.) the Assistant S.cretor) n my . COMMISSION i . , ' consistent with 5 U.S.C. 552b(c)(1). {

>rescribe alternative pri, Marcs to  !

Advisory Committee on Reactor . Mday. Nowmbu 7.19e8 spedite the review p occ. r im any .  ;

.ther good cause wh.r h m:.v 1.- S Safeguards; Revised Meeting Agenda &30 a.m.-#:30 a.m.:NRC Quantitative consistent with arplicable lav.s 1r e .

Safety Cools (Open)-Discun proposed 4 n ace " p Assistant Secretary findr tl.. t goud 3 ;, s a ACRS comments regarding development the c cause exists for not pblish.ng ihr of a methodology forinterpretation of supplements to the Utah State Plan as a F.n Ac 2 , he g 9,y , nii a NRC Quantitative Safety Coa,ls in terms proposed change and making tne of the population doggs associated with Safeguards will hold a meeting on Rigional Admuustrator's approval November 6-8.1986, in Room 1046.1717 nuclear power plant accidents.

effective upon publication for the H Street. NW., Washington DC. Notice R30 ant.-J&15ama Offa ofNuc. lear following reason (s): .

of this meeting was published in the Matersa/Sofety andSafegeds The standards were adopted in Federal Register on October 23.19es.- Activities-Radioactive Waste ccc:rdance with the procedurag Management andDisposal(Open)-

requirements of State law which Thursday Novembers.1ses. Briefing and dfscussion regarding l included public comment and further em a.in.-e45 a.m.:Repo}fofACRS ! activinn of the NRC Omce o%clw public participation would be * "I * I ** * '8" **

• Chairman (Open}--The ACRS Chairman repetitious.This decision is effective will report briefly regarding items of Division of Waste ManagemenL I

November 4.190& current interest to the Committeei.

A45 a.m.-I&30 a.m.:Improvedt.ight.

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(Sec.186 Pub. L ra -598. 64 Stat.1806 [29 - Report and discussion of ACRS U.S.C. es7)l - WaterReactors (Open)-He niembers Signed at Denver. Colorado this ath Day of , of the Committee will discuss proposed , subcommittee review of the implications of the Chernobyl nuclear plant accident September.1988. ACRS comments and recommendations to U.S. nuclear power plants.

Hany C. W

" to the NRC regarding characteristics of 12:30 a.m.-12 30p.m.:SofetyFeatums

'ActistgilegionalAdministmtor.

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,ip d ght.watunacton~ gyyg,,;g, yggy,,,pj,,g ggy ,,y_3y, fe45 a.m.-1245p.m. NRCSofety members will hear and discuss a report 86-24a4e Filed 11-3-46; 8:45 amj Research Progium (Open)-Briefing and

- caos as+as* discussion of matters related to the NRC by representatives of the NRC Staff regardinF safety-related modifications in safety research program including the Paluel Nuclear Plant.

prioritization of research activities. 2:30p.m.-2:00p.m.: Future ACRS specifically those related to fire Activities (Open}-Discuss anticipated pmtechon research for nuclear power ACRS subcommittee activities and l NATIONAL.TRA!:S! ~.TATIO: p ants.

AFETY BOARO proposed items for consideration by the 2:45 p.m.430p.m.: Review of the full Committee.

RegulatoryProcess (Open)-Discuss the 2D0p.m.-4:45p.m.: ACRS Putdic Hearing in Up..:r Dart >y PA. basis for proposed ACRS review of NRC Subcommittee Activities (Open)-

Railroad Acciden Regulatory requirements and related Discuss reports of designated ACRS regulatory processes and procedures. subcommittees regarding the status of In connection wif , invest @ation of 3:45 p.m.-5:45p.m.:Prioritization of - safety.related activities including Phast tha recident involvi- ie collision and GenericIssues (Open)-Members of the I of the NRC Maintenance and dIrailment of South, t rn Committee will discuss proposed Surveillance Program: the NRC Pennsylvania Trant: . tion Authority priorities for a new (fourth) group of inspection and enforcement program:

Single-Car Train No. Its at the 69th unresolved generic issues.. safety technology, philosophy and Street Terminal. Upper Derby. -

~ 145 p.m.4Dop.m.: Nomination of' criteria, and management of ACRS Pennsylvania. on August 23.1986, the ' . Candidates for A CRS CY 1887 Officers - resources. .-

N:ti:nalTransportation SafetyBoarts - (Closed}-ne Members of the ' 4:45p.m.430p.tir.:Nuclearpower will convene a public hearing at 9 a.m. # Committee will discuss the < - - PlantImprovements (Open)-Discuss (loc:1 time) on December 3,1986. at the'V qualifications of candidates for ACRS

• proposed ACRS comments regarding the Ad m's Mark Hotel. City Avenue and - ' officers and membership on the ACRS- ' basis for nuclear power plant M:nument Road. Philadelphia. . - '.*: Planning Subcommittee for CY 1987. Improvements.

Pennsylvania. For more informatio'n This portion of the meeting will be ' 530p.m.430p.m.: Emergency c:ntact Bill Bush. Office of Government' , closed to discuss information the release ' planning (Open}-Discuss me bases for end Public Affairs. National- ~' : '

• of which would represent a clearly NRC/FT.MA Guidelines and Transportation Safety Boird. 800 "~ '.' ~ unwarranted invasion of personal' ' . requirements for emergency planning in the vicinity of nuclear pcwer plants.

. 'Indipendence DC 20594, telephone ' -

Ave (202) SW.,

362-6007. Washington

. 1, 552b(c)(6). /l , "

• privacy . consistent

• with 5 U.S.C. .

Saturday. November s.19es naysmigk,..'~'L *.

&Dop.m.445p.m.: Safety ofNuclear. .

'N. " M. W: .'..~ Po;.".

" &30" 2s.-2%00 Noon:Preparotion of .

.N..-p , -e ' Powerplants implications of ACRS Intemational (Open/ . ACRS hports to the NRC(Open)-

Closed}-Discuss. .

Ocioberas,lesse:. i.e t.- Discuss proposed ACRS reports to the

' ' Meeting on N ' uclear Power Plant Safety.'

! 6 "; -

,? 'i. .' NRC regarding items considered durint

[ Fit Doc.36-34904 Filed 11-3-a18 Portions 45 airil .ofh,t. this session will be closed

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' as necessary to discuss information this meeting and a proposed report

• Provided in confidence by . regarding selection of nuclear power.

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,%,,,,,, y[q "- / Vol. 51 No. 213 / Tuesday, November 4,1938 / Notices 40095 4

plant personnel by use of sptitude t: sting. [ Docket No. 50-4001 ,

- therefoi(5) the FinalEnvironmental -

12p.m.-Jemp m. Ceneric Safety Shearon Harris Nuclear Power Plant,- * . Statement, dated October 1983: (6) the Issues (Open)--Complete discussion of Lltdt 1;lasuance of Facility Operating PartiallnitialDecisions of the Atomic ~

stems considered during this meeting. Ucense Safety and ucensing Board, dated i ' -

February 20. August 20, December it.

Procedures for the conduct of and Notice is hereby given that the U.S.

participation in ACRS meetings were 19as, and April 28.1988. . . - . .

published in the Federal Register on Nuclear Regulatory Commission (the Dese items are available at the October 2,1985 (50 FR 191. In Commission) has issued Fauility Commission's Public Document Room.

accordance with these pro)cedures, Operating oral Ucense No. NPF-53 to . -

1717 H Street. NW., Washington, DC -

Carolinc Power & Ught Company, and 20555, and at the Wake County Public-tr written statements may be presented North Carolina Eastern Municipal Power Ubrary,1313 New Bern Avenue, m

,by members of the public, recordings Agency (the licensees) which authorizes will be permitted only during those Raleigh, North Carolina 27601.-A copy of Portions of the meeting when a operation of the Shearon Harris Nuclear- the Facility Operating Uoense NPF-53 PowerPlant, Unit 1 at reactorcore . - may be obtained upon request's v w transcript is being kept, and questiens powerlevels not in excess of 2775 ' ,

m:y be asked only by members of the addressed to the U.S. Nuclear M-- /

Committee, its consultants, and StafL megawatts thermalin accordance with . Regulatory Connaission, Washington. ; .

the provisions of the license, the r.-

Persons desiring to make oral DC 20555. Attention: Director. Division TechnicalSpecifications, and the ,

of PWR Ucensing A. Copies of the c . '

stztements should notify the ACRS Environmental Protection Plan with a Safety Evaluation Report.and its - U'.

ExIcutive Director as far in advance as condition limiting operation to five .

Prseticable so that appropriate Percent of reactor core power (139 ----- -supplements (NUREG-1038) and the , .3 Final Environmental Statement may be &

crrangements can be made to allow the megawatts thermal). purchased through the U.S. Government necessary time during the meeting for Shearon Harns Nuclear Power Plant, . Printing Office by calling (202) 275-2000 such statements

• U f still* I Unit 1,is a pressurized water reactor- or by writing to the U.S. Covernment picture and television cameras dun."ng located in Wake and Chatham Counties, Printing Office, P.O. Box 37082, t is meeting may be limited to selected North Carolina, approximately to miles r Washington, DC 20013-7082. Copies <

portions of the meeting as determined southwest of Raleigh, North Carolina. may also be purchased from the by the Chairman. Information regardin8 The application for the license . - National Technical Information Se'rvice','

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the time to be set aside for this purpose complies with the standards and U.S. Department of Commerce. 5285 Port may be obtamed by a prepaid telephone requirements of the AtomicEnergy Act Royal Road, Springfield. Virginia 22161.

call to the ACRS Executive Director, R. of1954, as amended (the Act) and the Dated at Bethesda. Maryland, this 24'th day F. Fraley, prior to the meeting. In view of Commission's regulations.The of October.1986.

'he possibility that the schedule for Commission has made appropriate findings as required by the Act and the For the Nuclear Regulatory Commission.

,.CRS meetings may be adjusted by the Commission's regulations in 10 CFR Imster S. Rubenstein.

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ohairman as necessary to facilitate the gj7ce,,,, pwg pfgj,c, pj,,,,,,,,,2 Chapter 1, which are set forth in the . .

conduct of the meeting, persons gjyf,fon ofp w a ife,o, fog a. offic,of license. Prior public notice of the overall .,

plinmng to attned should check with the action involving the proposed issuance NucleorReactorRegulation. ';

ACRS Executive Director if such of an operating license was published in [FR Doc. 86-24805 Filed .11-3-46: &45 am]

raschedulint would result in major the Federal Register on January 27,1982 g w.a coot n m os-=

inconvenience. (47 FR 3898). The power level authorized I have determined in accordance withby this license and the conditions gon;.,t Nos. 50-324 and 50-32s1 ' '

subsection 10(d) Pub. I.92-433 that it is contained therein are encompassed by nicessary to close portions of this that prior notice.

mating as noted above to discuss Carolina Power and Ught Company; e c mmission has determined that Brunswick Steam Electric Plant; ,

privileged and classified informaiton suana o% hense wW not Meu any env mnmen alimpacts .

Mocatbn d LocalMc ht t'ad information the r[eleaseother of whichfrom than those evaluatedforeign in thesources Final ,l hm 5 -U.S--;C. 552b(Y)(1)]m-

9. - '

w u d represent a clearly unwarranted Environmental Statement since the .

.inv:sfon of personal privacy [5 U.S.C. . activity authorized by thelicense is' -Nuclear  ; .. > Notice Regulatory is hereby Commission given(NRC) that the 552b(c)(6H .

encompassed by the'overall action C

• has relocated the local public document Further information agarding topics evaluated in the Final Environmental room (LPDR) for Carolina Power and; 2..

Statement. . ' Light Comp'any's Bmnswick Steam ; , -

t3 be discussed, whether the meeting Pursuant to 10 CFR 51.32, the' ' '. ? Electric Plant from the Brunswfck , .- .

has been Chiirman's cancelled ruling on requestsor forrescheduled the the . Commission has "determined County Library,that Southport. the to the ..e '

~' - granting of relief arid the issuance of the - William Madison RandallLibrary.,, - -

cpportunity to present oral statements exemption included in this license will - ' University of North Carolina at . 4 g.

cnd the time allotted can be obtained by have no,significant impact on the. .' Wilmington.Wilmington, North ,..m._

o pr: paid telephone call to the ACRS environment (51 FR 36329, dated u." M ' -Carolina. ' :

' J 6- - r Ex:cutive Director, Mr. Raymond F. October 9.1966). . _,

Fr:lry (telephone 202/634-3265).

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M." . Members of the public may.now ,,-. I-~

For further details with respect to this

• Inspect and. copy documents and .-.--

betw1en 8:15 A.M. and 5:00 P.M.. action. see (1) Facility Operating License ' corresp,ondence related to.the licensiag M Octobu so.10aa. No.NPF-53:(2) the Commission's Safetyy and operation of the Brunswick Steam John C. Hoyle, Evaluation Report. dated November

. Electric Plant at the William Madison d dvisory Committee Monogement Officer. 1983 (NUREG-1038), and Supplements 1 ' Randall Library. University of North I

' ' Doc.'as-24932 Filed 11-3-88; 8:45 amj through 4: (3) the Final Safety Analysis : , Carolina at Wilmingtonc601 S. College ;

Report and Amendments thereto:(4) the . Road Wilmington.NC,.28403.De nr m ecoes ne 4 s-=,.,,'.... ,>.

t .. . . w . Environmental Report , . - ,and y.: -supplements'

. 4., Library is open ori-.~ the following .:.n . .: >

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/ 'o,'n UNITED STATES

. 8 NUCLEAR REGULATORY COMMISSION ADVISORY COMMITTEE ON REACTOR SAFEGUARDS o

j ,Y WASH WGTON, D. C. 20565 j

Revised: November 4, 1986 SCHEDULE AND OUTLINE FOR DISCUSSION 319TH ACRS MEETING NOVEMBER 6-8, 1986 WASHINGTON, D. C.

Thursday, November 6,1986, Room 1046,1717 H Street, NW, Washington, D.C.

1) 8:30 - 8:45 A.M. Report of ACRS Chairman (Ope )

1.1) OpeningStatement(DAW 1.2) Items of current interest (DAW /RFF)

2) 8:45 - 10:30 A.M. Improved LWRs (0 pen)

TAB 2 ----------- 2.1) Proposed ACRS report to NRC (D0/RKM) 10:30 - 10:45 A.M. BREAK

3) 10:45 - 12:45 P.M. NRC Research Program (0 pen)

TAB 3 ----------- 3.1) Discuss items of mutual interest with Director, RES (CPS /SD) 12:45 - 1:45 P.M. LUNCH

16) 1:45 - 2:45 P.M. Emergency Planning (0 pen)

TAB 16 ---------- 16.1) Discuss the basis for NRC/ FEMA Emergency Planning criteria / guidelines (DWM/EGI)

18) 2:45 - 3:30 P.M Regulatory Requirements, Procedures, and Processes (0 pen)

TAB 18 ---------- 18.1) Discuss proposed additions to list of issues for full Comittee consideration (HWL, et al/GRQ) 3:30 - 3:45 P.M. BREAK

6) 3:45 - 5:45 P.M. Prioritization of Generic Issues (0 pen)

! TAB 6 ----------- 6.1) Discuss ACRS comments regarding proposed I

priorities for new (fourth) group of generic issues (CPS, et al./3D)

7) 5:45 - 6:00 P.M. Election of ACRS'0fficers for CY 1987 (Closed) 7.1) Report of ACRS Nominating ComrHttee (DWM) l and nominations from the floor l

7.2) Nomination of Member at Large for ACRS Planning Subcommittee (DAW /et al/RFF)

(Note: Portions of this session will be closed to discuss informatin the release of which would represent a clearly unwarranted invasion of per-sonal privacy.)

' 4 319th ACRS Meeting Agenda .

Friday, November 7, 1986, Room 1046, 1717 H Street, NW, Washington, D.C.-

9) 8:30 - 9:30 A.M. Discuss proposed ACRS reports on: (0 pen)

TAB 9 -------------- . Interpretation of NRC Quantitative Safety Goals (DWM/EGI)

10) 9:30 - 10:15 A.M. NMSS Activities - Radwaste Management and Disposal (0 pen)

TAB 10 -------------- 10.1) Briefing by R. E. Browning, Director, Div. of Waste Management, NMSS 10:15 - 10:30 A.M. BREAK

11) 10:30 - 11:30 P.M. ACRS Subcomittee Activities (0 pen)

TAB 11 -------------- 11.1) Subcomittee report and discussion regarding the implications of the Chernobyl nuclear plant accident to U.S.

nuclear stations (D0/RPS)

12) 11:30 - 12:30 P.M. Modification of Foreign Nuclear Plant (0 pen)

TAB 12 -------------- 12.1) Briefing and discussion with NRC Staff of safety-related features in the Paluel Nuclear Plant (DAW /MME) 12:30 - 1:30 P.M. LUNCH

13) 1:30 - 2:00 P.M. Future ACRS Activities (0 pen)

TAB ----------------- 13.1) Discuss anticipated ACRS subcomittee activities (MWL)

SEE HANDOUT ---------- 13.2) Discuss proposed items for full Comittee consideration (DAW /RFF)

18) 2:00 - 3:00 P.M. Regulatory Requirements, Procedures and Processes (0 pen) 18.2) Discuss high priority items from members scoring of proposed list of regulatory requirements, procedures and policy issues (HWL, et al/GRQ) 3:00 - 3:15 P.M. BREAK

14) 3:15 - 5:00 P.M. ACRS Subcomittee Activities (0 pen)

Reports of ACRS Subcomittees regarding:

TAB -------- :------ 14.1) 3:15-3:45: Phase I of the NRC Maintenance and Surveillance Program (CYM/FA)

TAB ---------------- 14.2) 3:45-4:00: Activities of the NRC Inspection and Enforcement Program (CYM/

FJR/PAB) f

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319th ACRS Meeting Agenda TAB ----------------- 14.3) 4:00-4:30: Safety Philosophy, Technology and Criteria Subcommittee meeting (Nov. 5, 1986) regarding status of NRC work on steam-generator overfill (00/RPS) 14.4) 4:30-5:00: ACRS Planning Subcommittee Report of meeting on Nov. 5,1986 regarding allocation of ACRS resources

15) 5:00 - 5:45 P.M. Discuss proposed ACRS report on:

TAB 15--------------- 15.1) Basis for Nuclear Power Plant Improvements (JCE/HA)

8) 5:4S - 6:30 P.M. Safety of Nuclear Power Plants (0 pen / Closed) 8.1) Continue discussion of ACRS report on Improved LWRs, if needed 8.2) Discuss implications of ACRS International Meeting on Nuclear Power Plant Safety (DAW, et al/RFF)

(Note: Portions of this session will be closed

! as necessary to discuss information provided in confidence by representatives of foreign govern-mer.ts . )

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319th ACRS Meeting Agenda Saturday, November 8, 1986

17) 8:30 - 11:30 A.M. Discuss proposed ACRS reports on: (0 pen)

(10:30-10:45 - BREAK) 17.1) 8:30-9:15: Priorities for Generic Issues (CPS /SD) 17.2) 9:15-9:45: ImprovedLWRs(D0/RKM) 17.3) 9:45-10:30: Interpretation of Safety Goals (DWM/EGI) 17.4) 10:45-11:30: Basis for Nuclear Plant Improvements (JCE/HA)

19) 11:30 - 12:30 P.M. Generic Issues (0 pen) 19.1) Continue discussion of Emergency Planning and/or Regulatory Requirements, Procedures and Practices, as needed 12:30 - 1:30 P.M. LUNCH

20) 1:30 - 3:00 P.M. Miscellaneous (0 pen / Closed) 20.1) Complete discussion of items considered during this meeting (Note: Portions of this session will be closed as required.)

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dU '3 PROPOSED MINUTES OF: THE 318TH ACRS MEETING I4

- ! ",~~i.i OCTOBER 9-11, 1986 h b.]t 'hhd[.4 d [

The 318th 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, October 9,1986.

[ Note: For a list of attendees, see Appendix I. G. A. Reed and C. P. Siess did not attend the meeting.]

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 treeting was being held in conformance with the Federal Advisory Committee Act and the Government in the Sunshine Act, Public Laws92-463 and 94-409, respectively.

He also noted that requests were made by the Republicans Against Seabrook, Bacus, Myer, and Solomon, and Mr. Robert Walsch (Attorney, Manchester) to make oral presentations during the session on the Seabrook Nuclear Station.

Written statements on Seabrook were submitted by Harmon and Weiss on behalf of the New England Coalition on Nuclear Pollution (intervenor in the Seabrook operating license case) and P. McEachern, Democratic Nominee for Governor of New Hampshire. He also noted that a transcript of sorte of the public portions 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 ACE-Federal Reports, Inc., 444 North Capitol Street, Washington, D.C.20001.)

I. Chairman's Report (0 pen)

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

Chairman Ward made several announcements. He indicated that the Commis-sion granted on September 29, 1986 a low-power operating license to the Clinton nuclear power plant. He noted that the Perry operating license is still in limbo since a stay was issued by the Sixth Circuit Court of Appeals in Cincinnati, Ohio. The Commission granted an FDA for the Gessar II design on September 22, 1986.

Chairman Ward indicated that the planning for the Wingspread Internation-al Conference is proceeding well, although there is no word regarding attendance by representatives of the Soviet Union. W. Kerr is to chair a group to discuss implications of the Chernobyl nuclear plant accident with those foreign representatives who do attend. He noted that C. P.

Siess will receive the Distinguished Service Award for 1986 from the NRC for his outstanding contributions as an ACRS member. A. Newsom, ACRS Assistant Executive Director, has been awarded the Meritorious Award by the NRC for his support of ACRS activities. Chairman Ward also took note of the fact that the ACRS expects to have the same travel funding in Fiscal 1987 as it has had in Fiscal 1986.

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318TH ACRS MEETING 2 II. Seabrook Nuclear Station (0 pen)

[ Note: E. G. I of the meeting.]gne was the Designated ' Federal Official for this portion D. W. Moeller indicated that on September 26, 1986 the ACRS Subcommittees on Severe Accidents and Occupational and Environmental Radiation Protec-tion Systems were briefed on Public Service of New Hampshire's updated probabilistic risk assessment (PRA) for the Seabrcok Nuclear Power Plant, as well as the Seabrook Station Risk Management and Emergency Planning Study. This was planned as an information briefing by the Staff and the Applicant. D. W. Moeller noted sets of written comments submitted by Gordon R. Thompson on behalf of the Attorney General for the Commonwealth of Massachusetts and Ms. Stephanie Markiewicz, by. an October 10, 1986 letter to the ACRS Chainnan from Diane Curran on behalf of the New England Coalition on Nuclear Pollution, and from Paul McEachern, . on September 25, 1986, Democratic nominee for governor of New Hampshire.

D. Okrent asked if the Comittee should take any action. D. W. Moeller stated that there are many studies that the Staff has underway, both in-house and with contractors, that will not be concluded for months. No action is contemplated.

J. Doughty, representing the Seacoast Anti-Pollution League, Portsmouth, N. H., and currently an intervenor in the NRC licensing proceedings relating to Seabrook. Station, discussed the appropriateness of mathemat -

ical models as useful tools and the appropriate application of PRAs. She cited comments by former ACRS members M. Bender and J. J. Ray, which were appended to a September 15, 1982 ACRS report to the NRC, to the effect that PRAs are not useful for assessing public safety risk, since esti-mates will depend upon the judgment.of a few individuals. She contended that a PRA can never be the foundation for a judgment in the direction of reducing the margin of public safety because of the inherent uncertainty in the process of capturing complex physical phenomena through mathemati-cal modeling as well as the uncertainties resulting from natural physical phenomena, such as seismicity or flooding. She asserted that Public Service Company of New Hampshire was making an inappropriate use of their PRA in their justification for shrinking the size of the emergency planning zone (EPZ). She questioned the adequacy of the containment design in light of the current state-of-the-art which does not specifi-cally evaluate containment performance regarding its effect on the public safety. She cited the Seabrook Station's unique siting among a group of barrier islands separated from the U.S. mainland by an inadequate network of roads. She discussed the inadequacy of decontamination and treatment facilities provided by the New Hampshire Radiological Emergency Response Plans and the reduced level of readiness to deal with a radiological emergency at Seabrook that would result from reducing the radius of the EPZ. C. Mark noted that it is from the PRA that one decides that there is a concern regar61ng the basis for the EPZ. He asked if it was the high confidence in the Seabrook PRA that led her to express concern  ;

regarding a possible decrease in the radius of the EPZ. J. Ocughty thought one should look at other considerations besides the PRA as the basis for setting the radius of the EPZ.

318TH ACRS MEETING 3 M. Fallon, Republicans Against Seabrook, shcwed part of a videotape of crowds and traffic jams regarding sumer attendance at Hampton Beach which is in the vicinity of the Seabrook plant and left a copy of the tape for further viewing by the ACRS. B. Montville, a New Hampshire businessman, alleged that the Seabrook Security Plan was inadequate. He cited the fact that the containment and turbine-generator buildings are vulnerable to various type of terrorist attack. Ms. B. Hollingsworth, Hampton Beach, discussed the demographics for Hampton Beach. She noted that on a sunny sumer day there might be as many as 200,000 individuals or bathers. She indicated that it is a family beach and the tendency when it rains is not to evacuate the area but to retreat into cottages and motels. She claimed that the two roads providir. ingress to and egress from the Hampton Beach area are always jamed wita traffic during the sumer months. She questioned the population figures used by Public Service of New Hampshire in its PRA study for emergency planning.

D. Okrent pointed out to the Comittee that in both of the ACRS reports on Seabrook (December 10, 1974 and April 19,1983) the ACRS took special note of the very considerable population, including the transient popula-tion near the site, and of the Committee's inability to review the emergency plan because it had not been developed by the time of the 1983 report. The Committee had been aware that this is a site with a large population along the ocean.

D. W. Moeller indicated that, at the joint meeting, the Subcommittees reviewed the Seabrook Station Risk Management and Emergency Planning Study (RMEPS) and its companion Seabrook Station Emergency Planning Censitivity Study (SSPSA). He noted that much reliance is placed on containment sprays in the probabilistic safety study despite the fact that LERs show that those do not always function properly. A UASH-1400 source term was used in their calculations and an improved version of the CRAC Code which can handle continuous releases. They still use the whole-body and thyroid dose limits of 10 CFR Part 100 rather than the newer ICRP weighted factors which lead to an effective dose equivalent.

The results show that the Seabrook Station's containment gives protection far above that available at other plants. This is because the main rebar in the containment is continuous, going around the openings rather than l terminating at them. The dome rebar is continuous on the same plane for

! each of the four quadrants. Presenters indicated that earthquakes contribute to 87 percent of the abnormal events in the SSPSA, not to 87 l

percent of the risk. Interfacing system LOCAs account for the other 13 l percent. He noted that Public Service of New Hampshire Group applied a i variety of evacuation strategies and showed that evacuation beyond three niles from the plant would not yield much extra benefit. Mention was made of the fact that two diesel generators are installed in the discon-tinued Unit 2, and these two diesels could be of help to Unit 1. He noted that the Staff has identified 10 issues regarding the containment t

structure integrity and the Applicant is to provide additional informa-tion in response to these 10 issues. Note was also taken of the fact I

that M. Corradini, ACRS consultant, raised two concerns at the Subcommit-tee meeting. One involved the containment bypass through means of a steam generator tube rupture for a high-pressure meltdown sequence.

D. Okrent asked if there was any discussion at the Subcommittee meeting of criteria for establishing EPZs, and whether their radii are

318TH ACRS MEETENG 4 appropriate to emergency planning. D. W. Moeller indicated that the Subcommittees simply did not get into the detail of that aspect.

D. Okrent indicated his intent to ask questions of the Staff regarding appropriate criteria for setting emergency planning distances. He indicated his need to understand how uncertainties are introduced into .

the decisionmaking process.

S. Long, NRC, indicated that the Staff received an emergency planning sensitivity study from Public Service of New Hampshire in mid-July 1986.

The study updated their original technical study of 1983. There is still no request for actually shrinking the EPZ or for taking any other regula-tory action. The company has simply asked the Staff to review this study for its technical merits. He explained that NRR has asked the Office of Inspection and Enforcement, which deals heavily with emergency planning and evacuations and emergency preparedness, to work with NRR in deciding what criteria are pertinent and what factors in NUREG-0396 are important.

That information plus three comparisons made by Public Service of New Hampshire are the focus of the Staff's review. Those three comparisons were the risk of early fatalities at Seabrook compared to the WASH-1400 risk level with a 25-mile evacuation, the Commission's Safety Goal, and the probability of exceeding a dose-versus-distance as shown between the Seabrook conclusions and the NUREG-0396 conclusions. The Staff is primarily focusing on calculations for whole-body dose as they have been made over the last 10 years . The Staff will produce thyroid-dose calculations, taking account of the BEIR-III dose-response curves, as well as weighted-organ doses, and transform them into an effective whole-body dose. He indicated that these three dose-response curves have been added to the Brookhaven National Laboratory review contract.

D. Okrent asked what the Staff believes are the qualitative considera-tions in setting emergency procedures and what are the general quantita-tive considerations that would apply. He thought these two questions ought to be considered before exploring the differences between calculat-ing whole-body doses and organ doses. He hoped that the Staff would develop a safety philosophy to guide itself in decisionmaking concerning these matters. He thought the Staff ought to develop generic quantita-tive guidelines before making an ad hoc decision on a specific case such as this.

D. W. Moeller indicated that a sumary which gives a basis for the selection of the 10-mile emergency planning distance from NUREG-0654 was available at the Subcommittee meeting. S. Long admitted that there is an effort in the Commission to reconsider emergency planning criteria techniques. He mentioned Calvert Cliff's request to include the new source cy terms planning to see what benefit that would give in reducing the emergen-burden. He stressed, however, that the Seabrook plant has not made any such request. W. Kerr asked if the Licensee has presented an emergency plan for Staff evaluation. S. Long indicated that they have presented their own emergency plans. There is a New Hampshire State plan and associated components with some deficiencies which are being correct-ed. It has not yet been approved.

No plan has been submitted by the State of Massachusetts and an September 12, 1986 the Governor of Massachusetts announced that there would be no plan submitted for his State.

submitted.W. Kerr asked if there was a specific evacuation zone in the plan S. Long indicated that the EPZ for evacuation planning was

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318TH ACRS MEETING 5 the 10-mile zone. Another zone considering the ingestion dose pathway is 50 miles. He pointed out that there are still contentions to be heard before the Atomic -Safety and Licensing Board (ASLB) on the emergency planning issues offsite, including the evacuation time estimotes. W.

Kerr asked about approval status of the evacuation plans as they exist.

S. Long indicated that he was not sure of the exact status but could get an answer for the Committee. D. A. Ward asked the Staff its opinion as to whether the State of Massachusetts had not submitted a plan because they thought it technically infeasible. S. Long indicated that a public statement argues that the plan is technically infeasible. The Staff believes there is a misunderstanding of the technical requirements.

J. C. Ebersole asked if the Licensee has exhausted every possibility beyond the regulations to reduce the frequency of challenge to the containment and the potential for a core melt. S. Long mentioned opera-tor actions to recover as well as surveillance testing done by the Applicant. J. C. Ebersole asked if all they have done is administrative in character, such as operator procedures. S. Long indicated that they are basing their optimism regarding core melt primarily on the strength of containment and the configuration of their RHR system, as well as improved procedures. Another way they have shown a lower core melt frequency is in essentially refining assumptions and calculational techniques. In answer to another question by J. C. Ebersole, S. Long indicated that there are no significant hardware changes.

D. Okrent asked if the Staff expects the ACRS to have reviewed the emergency plan before the ASLB completes its hearing on an operating license. R. Hernan, NRR, pointed out that the ACRS letter on Seabrook did point out emergency planning as one area that the Comittee wanted to revisit before a licensing recommendation was taken to the Commission.

He was not sure, however, of the timing of the hearings and the final

? presentation to ACRS.

- W. Derrickson, Public Service Company of New Hampshire, presented a project overview, noting that Unit 1 is complete and Unit 2 is 24.1 l

percent complete and on hold with no work proceeding at this time (see l Appendix IV). He noted the high SALP ratings for the plant with the

[ exception of emergency preparedness for the period January 1985 through 1

March 1986. There are no fuel load open items as far as NRR or Region I are concerned, but that there are some ASLB items open. He noted that a Petition 50.57(c) to load fuel and conduct precriticality testing was submitted August 22, 1986.

W. Derrickson indicated that emergency plans were prepared by the State

! of New Hampshire and submitted to FEMA. A graded exercise for New Hampshire only was held in March 1986. There were some findings as a result of the exercise which were submitted to FEMA, and another exercise l will be scheduled late in the winter of 1986. The hearing on emergency l planning by the ASLB is yet to be held. The Governor of Massachusetts announced refusal to submit plans to FEMA on September 20, 1986, M. W. Carbon asked if Public Service of New Hampshire had made a study of evacuation times. A. Torri, Pickard, Lowe & Gerrick, indicated that 1

evacuation times are in the neighborhood of five hours. J. C. Ebersole asked if the evacuation times accommodate individuals walking out of the i

i 318TH ACRS. MEETING 6

, area.- W. Derrickson indicated that the conservative plan developed assumes that all individuals leave by vehicle.

A. Torri referred to three risk management activities undertaken by .the utility. These include the Seabrook Station.Prcbabilistic Safety Assess- '

i ment (SSPSA), completed in December 1983; a risk management and emergency planning study (RMEPS), completed in 1985; and a WASH-l'400 methodology sensitivity study which was a risk management emergency planning study under WASH-1400 source term assumptions, completed in April 1986. A seismic fragility update - was performed based on some results of the 1 review efforts on the NRC's part of the SSPSA. The SSPSA is being entirely updated. HRC review of these documents has consisted of a plant model review of the SSPSA by Lawrence Livermore National Laboratory in 1985 and review of the containment response source terms and consequences

-performed by the Brookhaven National Laboratory documented as NUREG/CR-4540 (February 1986). From these two reviews there has been nothing significant to challenge either the SSPSA or the emergency planning study. A. Torri discussed the scope and coverage of accident sequences in the SSPSA (see Appendix IV). C. Michelson asked if the SSPSA includes external events. He asked if a fire PRA was done. A.

Torri indicated that the SSPSA is a full-scope level-3 analysis .which includes all identifiable initiating events within the plant as well as i outside the plant. ~ Some external initiating events have not been ana-lyzed such as the crash of a truck into transmission lines. A.. Torri indicated that the study includes a full treatment of dependant events including common cause failures at the system level. C. Michelson asked what sort of data base was used on the effect of common causes such as smoke, heat, or flooding of the non-qualified equipment. He asked how Public Service of New Hampshire accounted for systems interactions effects. W. Stilwell, Pickard, Lowe & Gerrick, indicated that a detailed f

spatial interactions study walkdown was done. C. Michelson asked if the .

i study takes account of systems interactions effects between non-qualified and qualified systems. He pointed first to fire as one aspect, seismic 1 effects (chattering of relays on non-qualified equipment), and flooding.

l A. Torri indicated that chattering of relays is part of the fragility update. C. Michelson asked if that is done on non-qualified equipment.

A. Torri agreed to provide a written response.

A. Torri-discussed results in the SSPSA. He indicated that 76 percent of I the contribution ~ to early health effects results from an interfacing l

systems LOCA which is a breach of- the RHR system piping bypassing the containment structure, resulting in an early containment failure and I-release outside containment. The other 24 percent is due to seismic events which result in failure to isolate the containment. P. G. Shewmon asked if the interfacing LOCA is a steam boiler or steam generator tube rupture. A. Torri indicated that it is not a steam generator tube

,! rupture but a rupture of two series check valves which separate the

high-pressure boundary of the RHR system from the low-pressure portion.

The low-pressure portion fails as a consequence of the check valve failure. A. Torri indicated that loss of offsite power was an important

, contributor to latent health effects, while transients were the key 4 contributor to a no health effects group where the containment remains intact. C. Michelson asked about the normal containment ventilation arrangement for the Seabrook Station. A. Torri indicated that the plant

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318TH ACRS MEETING 7 is normally operated with the containment fully isolated. Venting is done for the purpose of maintaining access to the inside of the contain-ment for inspection, and the amount of time it can be opened is monitored and limited by the Technical Specifications. Conclusions reached by Public Service of New Hampshire in 1983 from the SSPSA study was that, with respect to the early health risks, the plant met the early health effects safety goal by a large margin. With respect to latent health risk the NRC safety goal was again met by a very large margin--of the order of 100-1000. The core melp frequency was determined to be a mean value of approximately 4 per reactor year. The corresponding median value would be per 10_2 reactorxyear.

10~

A. Torri discussed certain unique features of the Seabrook containment.

He indicated that the enclosure building is designed to mitigate the TID 14844 design basis leakage (0.1 percent per day), given a source term of that magnitude inside the containment. The unit is constructed of basaltic concrete which precludes almost all noncondensable gas genera-tion and scme of the flamable gas generation (no decomposition of calcium carbonate). The steam generators are the large Westinghouse model F type which doubles the time for steam generator dryout and, for any transient, doubles the time available before the capability of heat removal through the inherent inventory of water in the steam generator is lost. D. A. Ward asked what the doubling meant. A. Torri indicated that steam generator dryout goes from about one to two hours. The earlier Westinghouse steam generator models had about two-thirds of the secondary side water inventory available to boil off before the primary system would begin to heat up (for example, a station blackout). The RHR vaults in this plant are a confined geometry deep-well-type design. There is the inherent geometry of a suppression pool type as cpposed to a large auxiliary building with large volume but no water-retention capability.

C. Michelson asked about the benefit of this feature. A. Torri indicated that the interfacing systems LOCA accident discharges both the primary system and the refueling water storage tank into these RHR vaults. The most likely location of the breach of the RHR system would be under 30 feet of water in this case instead of an open atmospheric release.

Release of radionuclides will be contained or mitigated by passing through that water (decontamination factor).

A. Torri indicated that the RMEPS had the objective of reexamining the technical basis of the 10-mile EPZ (NUREG-0396) on a plant-specific basis. Advances in the analysis methodology for addressing EPZ questions involved the development of dose-versus-distance curves as published in NUREG-0396 but on a plant- and site-specific basis using the SSPSA.

Comparisons were made against certain risk acceptance criteria such as the NUREG-0396 dose-versus-distance curves for 1, 5 , 50 , and 200-rem whole-body doses; the WASH-1400 risk curves for early fatalities and latent cancer fatalities; and the NRC individual and societal risk safety goals. The spatial distribution of residual risk was measured to deter-mine what the residual benefits might be. The RMEPS update of the SSPSA risk model focused on updating the plant model and the source terms. A more detailed look was made involving the scenario of closure of purged penetration valves if they should open during an earthquake. The update considered recovery of long-term sequences if the pressure reached the failure pressure of the containment. Changes were made in the common

318TH ACRS MEETING 8 cause failure model. C.-Michelson asked the length of time assumed that' fire would burn before it was extinguished in the case of fire-induced common cause failures. A. Torri indicated that the time it takes to extinguish a fire is -dependent upon the intensity of the fire and its

~ location. C. Michelson noted that the Seabrook Station is an Appendix R  !

plant and Appendix R fixes were only one-hour fixes. He asked if it is assumed that a fire is always extinguished in one hour, or did the update take account of the fire not being extinguished for three hours? i A. Torri did not believe the assumption was that all fires would be extinguished in one hour. C. Michelson asked how one analyzes a plant for a fire burn longer than one hour. A..Torri indicated that the plant i has two fully separated safety trains. C. Michelson countered that the fire barriers are only good for one hour in certain cases. The implica-tion is that one must extinguish a fire in one hour because that is all the safe shutdown protection provided. A. Torri indicated that a written response would have to be provided.

P. G. Shewmon noted that early containment failure is dominated by prompt I failures from seismic events. He asked what sort of seismic events are postulated to cause the Seabrook Station a problem. A. Torri indicated that the intensity of the earthquake is much greater than a design basis earthquake, essentially .of such a magnitude as to guarantee failure of the equipment. P. G. Shewmon asked what kinds of failures are postulated

to give rise to this prompt failure of the containment. A. Torri stated

that the scenario is a seismic event without loss of offsite power, the

failure of the signals when the purge valves are open, and they would

remain open. P. G. Shewmon failed to see how the failure of closure of the purge valves might cause core damage. A. Torri contended that seismic-induced ATWS events were a significant contributor to earth- '

quake-induced core damage scenarios. C. Michelson wondered what assump-tions were made concerning fires induced by the same earthquake that L would cause a severe accident. W. Stilwell indicated that an answer at a later date would be appropriate. D. Okrent expressed interest in the i

failure of many large pressurized components inside the primary system which could lead to a loss of containment integrity because of the association of missiles with these failures. A. Torri indicated that this, was not considered explicitly as a contributor to containment

! failure.

A. Torri indicated that a peer review group performed a review of the RMEPS and concurred in the principal study findings. It found that the overall risk is very small, the health risk was lower than thought to exist when general EPZ requirements were established, and the risk was i confined to areas very close to the nuclear plant. The peer review group indicated that the Seabrook containment was a major factor in their favorable findings.

I i A. Torri asserted that the early health effect risk is very low even l without inanediate protective action. The benefits of evacuation are very i small in absolute terms and the risk is confined to the first one or two I miles in the vicinity of the plant. The peer review group supported these conclusions as well as those of the WASH-1400 methodology sensitiv-ity study, t

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318TH ACRS MEETING 9 1

D. Okrent thought it relevant to determine the time scale for the Staff review of these studies. S. Long anticipated conclusion.of the Staff's review by the middle of November 1986. It will. depend on conclusions the Staff reaches at that time and any requests made by the Applicant regard-ing regulatory changes. D. A. Ward asked if the Staff intends to issue a written report. S. Long indicated that there will be a report from the Brookhaven National Laboratory and presumably a memorandum covering this-report. D. A. Ward suggested that the Committee wait for the results of the Staff's actions before taking any further action of its own. .

I III. Resolution of Outstanding Issues on Clinton Power Station, Unit 1 (0 pen)

[ Note: P. A. Boehnert was the Designated Federal Official for this portion of ths meeting.]

R. Hernan, NRR, explained that the Staff is close to considering a full-power operating license for the Clinton Station. It has been the practice of the Staff that, whenever the ACRS has written a letter with unresolved items shortly before the full-pcwer license consideration, the Staff would return to the ACRS with an update on resolution of the open items.

B. L. Siegel, NRC, indicated that the Clinton Station was granted a low-power license on September 29, 1986. All confirmatory and outstand-ing issues were resolved before the issuance. The Licensee is loading fuel. In answer to a question by J. C. Mark, B. L. Siegel indicated that the low-power license is a full-power license limited to five percent of the thermal rated power of 2894 Wt.

B. L. Siegel indicated that the March 9,1982 ACRS report identified two plant-specific items for follow-up:

Quality assurance and quality control organization Seismic capability of emergency AC power supplies, DC power sup-

! plies, and small components, such as actuators and instrument lines

, (part of the decay heat removal system)

He indicated that the quality assurance issue was addressed in Appendix M to Supplement 6 of the Staff's SER. The seismic capability issue was also addressed in Section 22 of Supplement 6. It was identified as Item 3 in Section 22 called the " Quick Seismic Assessment Program for Safety-Related Mechanical and Electrical Equipment" (see Appendix V).

D. P. Hall, Illinois Power Company, discussed recent background informa-tion regarding the Clinton Power Station. He mentioned the Independent Design Review initiated by Illinois Power in May 1984 and ccnducted by Bechtel Corpcration. He indicated that the high-pressure core spray systen, the shutdown service water system, and the Class 1 AC power system were specifically evaluated to confirm the quality of the design.

I The Atomic Safety and Licensing Board was established in April of 1985 after a joint stipulation between the intervenors (Perry Alliance and the State Attorney General) and Illinois Power Company. The ioint stipula-tion resolved the last three outstanding issues regarding emergency planning, control room design, and quality assurance. The basis for the l

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318TH ACRS MEETING 10

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stipulation was the Independent Design Review and a Construction Ap-r praisal Team (CAT) review in which the intervenors were afforded the opportunity to participate in the critiques and review of the results of

. both efforts (see Appendix VI).

F . . ~J . Jablonski, NRC Region III, explained that routine and special 1

mechanical and electrical inspections by the NRC and the Licensee in 1981 and 1982 resulted in several stop work orders. The result was an Overin-spection Program and significant changes to the Illinois Power Associates Quality Program. Early in 1981 a number of significant deficiencies were found by NRC inspectors in the areas of fabrication, installation, and inspection of supports and restraints for piping systems. This led to a stop work action by the Licensee. Based on an NRC inspection conducted

, in June 1981 that stop work action was lifted. Early in 1982 significant

- deficiencies were found in electrical construction- activities. The Licensee issued a stop work order for specified electrical activities .and performed a special inspection. An NRC/ Illinois Power management meeting ,

was held in January 1982. The Licensee was asked to review areas other than electrical at that time to determine if problems identified during the inspections were generic to other areas. Based on that review, Illinois Power initiated stop work action in eight additional areas. By 1983 the NRC had assured itself that work was ready to restart and all stop work actions were lifted. In answer to questions by D. A. Ward, .

F. Jablonski explained that, of 33 stop work actions initiated by the Licensee 'or constructor, about one-third of those actions occurred during the 1982-83 construction shutdown. There have been no stop work actions ordered by NRC and no stop work actions are currently in effect. All previously issued stop work actions were satisfactorily resolved by 1986. ,

An overinspection program of completed construction wo d was necessary at I the -Clinton Power Station because the Licensee could not assure the NRC in 1982 that the installed structures and components were free of criti-

. cal defects which could adversely affect operation of safety-related systems. D. P. Hall explained that the overinspection program was instituted to confirm or correct the quality of electrical hanger and train installations and conduit installations on which deficiencies were

. ' discovered during NRC inspection visits. He noted that none of the i- defective attributes found by the Staff was critical to the safe shutdown or safe operation of the plant.

F. J. Jablonski indicated that major changes were made in the quality assurance organization of Illinois Power due to problems identified by j the NRC in January 1982. The quality assurance organization was restruc-

! tured to report to the Executive Vice President and then, subsequently, l to a Vice President in charge of Quality Assurance, Engineering Startup i and Operations. The quality assurance crganization became more function-al and included quality engineering, surveillance, and audits. The staff

size was increased appreciably and the corporate quality assurance was moved to the Clinton site. It assumed many line review responsibilities regarding site work.

, F. J. Jablonski indicated that the licensee conducted 500 audits and the constructor 1,000 audits on the Clinton project. There were third-party j audits such as the joint utility / management audit and INP0 has audited Illinois Power. During the construction and preoperational phases, E

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• 318TH ACRS MEETING 11

% r several inspections and reviews were conducted by the NRC to review the

, Licensee's quality assurance functions, including the stop-work action in l 1982- and the CAT = inspect; ion in 1985. The Licensee's quality assurance L activities were found generally adequate. Recent inspections by the NRC E have identified improper post-maintenance and post-modification testing l of systems following preoperational tests. The Licensee has reviewed all modification packages and has retested as appropriate. There is escalat-l ed enforcement action still under review for this particular problem.

There is also escalated ' enforcement action pending on a Department of Labor discrimination case, as well as enforcement pending on improper floodproofing of access hatches in construction openings.

F. 'J. Jablonski indicated that stop-work actions and overinspection-activities have been completed 'in accordance with NRC requirements and Region III finds that the Licensee's quality assurance for construction -

and preoperational testing is acceptable (see Appendix VII). The Licens-ee has been responsive to correcting previously identified prcblems and the management changes have been generally positive. Management recog-nizes the necessity for continued attention to assure quality operations -

and perfonnance. In addition, an operational readiness inspection that (

will be conducted prior to issuance of a' full-power license will include j a review of quality assurance, as well as plant operations, maintenance, '

and administrative control for the Clinton Power Station. 4 D. P. Hall indicated that Illinois Power set up an equipment seismic assessment program, specifically to address the ACRS March 9, 1982 comment regarding the seismic acceptance of the Clinton design. The Program specifically addresses the residual heat removal system which includes 'the pumps, heat exchanger, valves, piping; the shutdown service water system which includes the pump ' strainers and small piping; and the AC and DC power supplies that support those systems - (i.e. , battery chargers, motor control centers, switchgears, substations, the diesel generators, their control the fuel oil systems, the engines themselves, and the piping).' The panels, equipment seismic assessment program had three phases. The first phase was to verify the adequacy of design calculations for the small bore piping / instrument lines. The review of compliance with design guidelines found some discrepancies which were formally corrected with 50.55(e) submittals to the NRC Staff. The second phase of the -program involved seismic interactions walkdowns/ analyses done in the field by engineers. These field walkdowns identified poten-tial interactions. In addition, the equipment was evaluated for loadings on the basis of the site-specific seismic spectra. If the equipment had been seismically qualified by test, then test response spectra were examined. If the equipment was seismically _ qualified by calculation, the acceleration values used had to be higher than those in the spectrum (conservative) so that the qualification would be within the design envelope. The component critical stresses had to be shown within the code limits for that specific application. The report was completed and submitted to the NRC in October 1985. D. P. Hall indicated that, while some concerns were identified as the process unfolded, Illinois Power is confident that they have a satisfactory seismic design and installation.

F. J. Remick asked if any of their concerns were of any significance.

D. P. Hall indicated that nothing was found that would preclude safe shutdown of the plant. J. C. Mark noted that the Clinton Plant is more

318TH ACRS MEETING 12 similar to one particular operating plant than to others. D. P. Hall indicated that there are four BWR-6 Mark IIIs which are similar to each other: Grand Gulf, River Bend, Perry, and Clinton. 'J. C. Mark asked if

.the differences between Clinton and the other three make the situation better or worse. - D. P. Hall indicated that Clinton has profited from the experience with the ~ other three plants simply because Clinton was the last in the chain and has incorporated many design-corrective features.

He' noted that Clinton is the only one of 34 plants that contracted with

'GE for a solidastate instrumented reactor protection system. Control room displays on the cabinets are television scopes. J. C. Ebersole i noted the susceptibility of solid-state equipment to fairly small differ-ences in, or excessive, ambient temperature, certainly to the. invasion of-the influence of fires. He asked if the system had an overtemperature protection system in anticipation of malfunction because of high tempera-ture. D. P. Hall indicated that plant personnel are more concerned about voltage and current spikes than temperature fluctuations. J. C. Mark asked if any provisions have been made against lightning or to address electromagnetic pulse. D. P. Hall indicated that nothing special has been done different from plants with relays. He iterated more persistent worries regarding voltage spikes from cycling of components.

J. C. Ebersole asked if the Clinton operators know the direction of the spurious reading of the safety circuits when one gradually raises tempera-ture to the point of failure. D. P. Hall thought it a good question but indicated that he did not know the answer and it would depend on which component fails first.

W. Kerr indicated that the ACRS has advocated systematic examination of operating plants to look for outliers. He asked if, c'uring the number of inspections and reinspections that were made, any items were found in terms of their safety or reliability significance. D. P. Hall indicated that nothing was found that compromised the plant's safety maigin. He did think that the inspection programs were helpful in many ways. He noted that the Construction Appraisal Team gave credit for the high quality of construction in great measure because of the overinspection program.

A. Lee, NRC, discussed the three-phase equipment seismic assessment program. He indicated that the first phase examined the adequacy of the design methods used for small bore piping. Phase two examined the as-built equipment configurations of the decay heat removal and emergency power supply systems which may be susceptible to earthquake damage. The last phase evaluated the ability of equipment in these systems to with-stand an earthquake of the form predicted by the revised response spectra which were developed using the elastic half-space approach for soil structure interaction analysis. C. Michelson wondered why the seismic program was done in the first place. A. Lee explained that this equip-ment was originally seismically qualified using a finite-element method of analysis. The spectra (specification) changed. He explained that the Staff reviewed the Licensee's report (letter of October 14, 1985), and also reviewed the small bore piping procedure, selected design calcula-tions, and the design of small tap lines and their qualification method.

On January 16, 1986, the Staff performed a plant site audit on the equipment seismic assessment program phase one results. The Staff also perfonned a site walkdown audit of several areas reviewed by the ,

1

m 318TH ACRS MEETING 13 w

Licensee's nuclear safety engineering department. The Staff found sufficient assurance that the small bore piping in the decay heat removal and emergency power systems has sufficient capability to withstand an earthquake. In conjunction with the Staff's site walkdown audit, seismic interactions and interferences were identified. The Staff has verified that this kind of problem has been properly resolved.

The Staff also performed a detailed review of the reassessment of equip-ment seismic capability. For most of the equipment which was qualified by test, the test response spectra were found either to envelop the revised response spectra at the entire frequency range or to envelop the revised response spectra for all but the low-frequency range.

C. Michelson asked how Illinois Power knows that some of the other equipment, for which the ACRS asked that seismic design be checked, is acceptable. A. Lee indicated that the Staff has performed a seismic qualification review for the entire spectrum of safety-related electrical or mechanical equipment at Clinton and has concluded that it is accept-able. For pieces of equipment that were qualified by analysis, the accelerations attributable to the revised response spectra were generally found to be less than the accelerations from the procurement specifica-tions response spectra. In the exceptional cases where revised response spectra exceeded procurement response spectra accelerations, the equip-ment stresses due to the revised response spectra were calculated and found to be within material allowable limits. C. Michelson noted that some of the equipment examined must have involved control devices or electrical contacts. He asked if contact chatter effects were examined after the change in spectra. D. P. Hall had a qualified "yes" answer.

D. A. Ward noted that the Committee's comments in its letter were that the issues raised be resolved to the satisfaction of the Staff.

R. Hernan indicated that the Committee's comments were resolved to the Staff's satisfaction as noted in the Supplemental SER. D. A. Ward asked if tho Staff is in need of a letter from the Committee. R. Hernan indicated that a letter from the Committee is not necessary at this time unless there is substantial disagreement with the Staff's documented resolutions.

IV. Reactor Operations (0 pen)

[ Note: H. Alderman was the Designated Federal Official for this portion of the meeting.]

J. Rosenthal, IE, presented the events analysis selection process tow being applied to ACRS review of recent significant events in lieu of subcommittee meetings. He indicated that in the last two months 680 immediate notification reports (10 CFR 50.72 reports), as wel' r daily reports from each of the regions, were reviewed. Based on that informa-tion, IE selects about one-quarter of those reports for follow-up. Based upon follow-up activity, three to five events per week are presented to NRR/IE management. Forty candidate events (two- to three-page suwlary of each event) were presented to J. C. Ebersole and 12 were jointly selected from which a final selection was made for ACRS full Committee presentation (see Appendix IX).

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318TH ACRS MEETING' 14

-2 A. Loss of Low Pressure Service Water at Oconee Nuclear Station, Unit 2 l H. Bailey, IE, indicated that on October 1,1986 at Oconee Nuclear l Station, Unit 2 a~ loss of low pressure service water suction cooling n occurred. This is significant because it is a loss of-the ultimate L heat sink and a design deficiency that may be applicable to other L plants. The system design was discussed, and it was noticed that-the low pressure service water only takes suction from the condenser circulating water main. The pump in this. system is at a high point and a siphon is maintained in the system as long as the component cooling water pump is running. If there is a blackout condition the system is designed to maintain a siphon so that water will drain,'by gravity, through to the discharge. In the. blackout condition, low

pressure service water pumps would be taking suction from the j i component cooling water header, from this gravity flow. .

4 i The low pressure service water pumps on Unit 2 lost suction after about one hour of operation following a load shed test with

condenser circulating ~ water pumps idled. There was a loss of flow and the operators recognized that there was a problem somewhere in the system. Suction .was restored after a cross-connect valve to Unit 1 was opened. Duke personnel recognized that there was a problem with the system and, since the -system was safety related with safety-related heat exchangers feeding decay beat removal components and bearings' for the auxiliary feedwater pump turbine, a decision was made to shut down Units 1 and 2. Plant ~ personnel did not know whether the low pressure service wate- system was opera-

tional. After reaching cold shutdown, it was ~oecided to leave the i low pressure service water pumps on Unit 1 and then to do a normal i

gravity drain test through the main condenser on Unit 2. This nonnal gravity drain test failed after some length of time. Even

-without the suction from the. low pressure service water pumps it

! became apparent that the _ siphon from the high point was not being maintained for some reason. After some investigation,. it~ was

} determined that there were flange leaks on the Unit 2 component cooling water pumps. These were sealed and the pumps then passed the component cooling water gravity flow ' test. Note was taken of l the fact that the low level of Keowee Lake was a' contributing factor to the problem since the component cooling water pump has a flange at - a high point in the Keowee hydro emergency power dam setup. .

C. Michelson cited tb.e fact that if the condenser circulating water pump is not running the plant depends upon the siphon to bring the cold water from the Keowee Lake reservoir to the intake point for L the pump. The Coninittee discussed the fact that the Oconee-2 plant

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emergency pcwer is provided by a hydro scurce of electric power rather than emergency diesels. J. Rosenthal indicated that the event was added to the list for ACRS consideration as a significant l

event because there was a potential loss of the ultimate heat sink

=

for'the plant. He noted that the Staff will be exploring facets of this event over the next several weeks. The important feature of i this event was reliance on gravity feed to provide a water source to the service water. W. Kerr asked if there is only a problem if one l

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318TH ACRS MEETItiG 15 has a station blackout. H. Bailey indicated that there is only a problem if one does not have power to the component cooling water pump. As long as the component cooling water pump is running you have suction and there does not appear to be a problem.

J. C. Ebersole asked if the component cooling water pumps have offsite power in the case of a station blackout. H. Bailey indi-cated that the component cooling water pumps can be fed from two different paths from the Keowee hydro system. Loss of offsite power does not include loss of power from the dam.- The Committee dis-cussed the seismic capability of the Keowee dam. H. Bailey noted that the Keowee hydro dam has the same function as the emergency diesels in most plants and is seismically qualified. Plant batter-ies are also seismically qualified. J. C. Ebersole suggested that there is definitely a design deficiency with such an indirect supply of water to the low pressure service water pumps. N. Rutherford, Duke -Power, indicated that the low pressure service water pumps do not depend on a siphon or vacuum system--they depend upon running a component cooling water pump and there are 12 pumps at Oconee, any one of which could supply water to the suction of all the low pressure service water pumps. J. C. Ebersole introduced the scenar-io of a condenser neck failure as a result of a seismic event and resultant flood. He asked if the plants could tolerate the result-ing flood while trying to maintain suction in the low pressure service water pumps. N. Rutherford indicated that this was one of the reasons why a safe shutdown facility was added to Oconee; it provides an independent means of safely shutting down the station in the event of a flood. J. C. Ebersole suggested possible redesign of the plant because of this postulated loss of service water.

J. Rosenthal noted that the design basis of the plant is that one should be able to gravity feed the circulating water system.

W. Kerr asked if the Staff thought this situation high-risk.

J. Rosenthal indicated that it is until the Staff fully understands the situation. He also noted that the loss of offsite power, or station blackout, is a very unlikely event. N. Rutherford pointed out that blackcut is loss of offsite power and also' loss of the Keowee dan, and Keowee has proved to be a highly reliable source of power over the years in relation to diesel generators. H. W. Lewis asked how long the Oconee units will be shut down. N. Rutherford indicated that Duke Power will meet with the Staff on Tuesday, October 14, to discuss the results of the siphon test with Duke Power's improved seal on the pumps to shcw that they indeed meet the design basis. E. Jordan pointed cut that the units were shut down voluntarily by the utility. It was not a shutdown order by the NRC.

N. Rutherford indicated that Duke Power intends to restart Unit 2 at the first opportunity. Units 1 and 3 are having additional work done and will be down for a bit longer.

B. Loss of Offsite Power Test at Hope Creek J. Wiggins, NRC Region I, indicated that 24 unexpected problems were identified at the outset of a loss of offsite power startup test conducted at Hope Creek on September 11, 1986. The test was con-

ducted from about 20 percent power. The test was aborted after

!~ about five minutes by the Licensee due to prcblems with one of the L

318TH ACRS MEETING 16 diesel generators and due to concerns about increasing drywell pressure caused by loss of several auxiliary systems in the plant.

A subsequent test was performed on September 19 while the plant was shut down. This second test was done cold and 19 additional prob-lems were identified. The number and nature of.the problems identi-fied during these two tests raised the level of Staff concern, both at Headquarters and the Region, about the adequacy of the facility design, construction, and testing program.

J. Wiggins indicated that the Staff classified the causes of the 41 problems identified into four areas. These were as follows: 1)

Bailey (solid mance powerstate supply logic failures, 2) selection), 3)Regulatory Guide minor design and1.97 confor-equipment problems, and 4) preoperational test weaknesses. Region I dis-patched an Augmented Inspection Team (AIT) to analyze the results of the loss of power testing and to assess the safety significance of the problems found. The Staff's principal concern with safety significance involves the reliability of the Bailey solid state logic modules for control of safety-related systems.

J. Wiggins indicated that Regulatory Guide 1.97 conformance problems basically involved power supply selection. -

The as-built design method drawings but not the design conformed to FSAR statements about power supplies for the safety relief valve acoustic monitoring system, level indication, temperature recording, and source range /

intermediate range monitor drives. Power supplies to these items were supposed to be uninterruptable power. It turned out that they were not. They were on non-IE power which went away during the loss of power test. In answer to a question by C. Michelson, J. Wiggins indicated that they were not connected to the right source. In answer to an inquiry by C. J. Wylie, J. Wiggins indicated that the safety instrumentation has backup power, but the non-safety instru-mentation does not necessarily remain energized on a loss of power event or loss of power to the 1E buses. The regulatory etnformance issue was identified as a regulatory problem and not a significant safety problem.

J. Wiggins indicated that the preoperational test program was conducted as described in the FSAR. Some weaknesses were identified in testing of non-safety-related systems but these tests conformed to Regulatory Guide 1.68 requirements as well as the FSAR. The preoperational loss of power test only involves loss of power to Class 1E buses; there is no precperational loss of instrument air test. He indicated that the failure of instrument air was traceable to a design problem in the interface between the air system and another system that provides cooling to it. That particular inter-face was tested during the precperational test program, but was done dry. If the test had been done as a wet test the problem that turned out to be a timing problem would have occurred when one set of valves fully closed before another set started opening.

J. C. Ebersale noted that the preoperational tests are individual tests and not really an integrated switching of all systems at once.

J. Wiggins agreed that most of the preoperational program is indi-vidual tests. He did indicate, however, that there are two

318TH ACRS MEETING 17 integrated tests that typically occur on a preoperational program at a'BWR. One is a preoperational-loss of offsite power test which, if done, would have been broad enough to ~ identify all of these problems in the' preoperational program and not subsequent to it. The Hope Creek preoperational loss of offsite power . test only involved the simultaneous loss of power to the Class 1E buses. The non-1E buses maintained power. The Staff' identifies this as a weakness in-their preoperational program. The_ AIT concluded that a broader loss of offsite power test involving all offsite power supplies to all buses could have identified these problems early.

. 4 C. Ebersole asked if there are -'any generic aspects to this situation. J. Wiggins indicated that it does apply to other BWRs,

, _ one in particular. that did the same type of tests. Some other facilities have run a preoperational loss of instrument air test as

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recommended in the Regulatory Guide. This utility took exception to that ' Regulatory Guide position in. its FSAR and points to that exception which was not challenged or overturned by the Staff.

C. Michelson asked the longest time that this plant has done a loss of 'offsite power test. J. Wiggins indicated the second test ran

. 20-30 minutes, long enough for the air system to bleed down. Valve

problems due to the -loss of instrument air started to occur around 18 minutes into the test.

J. Wiggins indicated that the Bailey solid state logic system is used extensively for centrol of the balance-of-plant systems at Hope i

Creek. The balance of plant is both safety-related and non-safety-related. The Comittee discussed the temperature specification of the Bailey central memory chips. J. Wigg%s indicated that the reliability of the Bailey system is a concern that was recognized by the Staff in its operating review. The Staff recognized that this

, Bailey system, as implemented at Hope Creek, has manual and automat-ic functions relying on more common circuitry than one typically i would see in a relay logic system. As a result of these concerns,

- the Staff proposed a license requirement for a reliability program with a report issued prior to the restart from the first refueling outage. W. Kerr asked the nature- of the reliability program re-quired of this Licensee. J. Wiggins indicated that the Licensee i must accumulate data and statistics about in-plant performance of these modules. The Comittee discussed the environmental qualifica-tion of the Bailey circuit boards.

J. Wiggins mentioned construction problems with the Bailey modules.

These involved staple jumpers which are not soldered, but are removable and movable. These staples are moved to make a particular buffer accept a particular voltage level. Also mentioned was the manually programable field programable logic array (FPLA). There were some programming errors identified as a result of the testing

in the loss of power test. Those programing errors, however, appear to be inconsequential and in sections of the chip not used to

, perform the function expected during the loss of offsite power.

l J. Wiggins discussed the adequacy of functional testing of the

Bailey modules, noting that the limited M situ testing is a 0

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318TH ACRS MEETING .18 shortcoming. Only about 50 percent of the Bailey modules get tested in an 18-month cycle in a limited bench test where all functions are not checked. In addition,- the test devices require realignment of l the little staple jumpers to complete. the board continuity test.

The Licensee does not have the capability on the bench to check the

,' boards with the- jumper arrangement that should be there before it-goes into the system. This raises de possibility of configuration problems with the modules. This is a human problem, and there is no check of continuity in the staple arrangement.

1 D. Okrent expressed concern that either temperature or human error.

could lead to a loss of much of the safety-related equipment in the L Hope Creek plant; it is a point of vulnerability. J. Wiggins agreed

and noted that this concern resulted in the reliability program

. requirements .that NRR prescribed. .J. Wiggins indicated that reli-ability would improve as the Licensee moves toward an improved in i '

situ test device which includes an interim effects of the capability' i to completely bench test the logical functions, all possible combi-

nations, that should give the Staff an extra measure of confidence

- about the ability of the system to operate reliably. C. Michelson indicated that the real concern is the failure modes and effects of this circuit board caused by - such things as local fires,. smoke, dust, and electromagnetic discharges. He suggested that this is r.ot being explored. J. Wiggins indicated that, while he could not

, address Appendix R considerations, these issues should be investi-gated by the Staff. C. Michelson- applauded the nature of the

. discussion on solid state control and indicated that a' subcommittee meeting on the whole subject of solid state control . is planned in '

the near future. J. Wiggins explained that the Staff does not believe that there is a fundamental base line weakness .in the Hope Creek design or design control program, preoperational test or

- surveillances, operator or test . engineer knowledge or QA/QC. Most 2

of the events discussed can be treated as isolated cases. While there is some connection between them, they do not appear to repre-

. sent a fundamental defect in the Licensee's program. He indicated that discussions with the Licensee are still in process regarding the-Bailey concerns.

C. ~ Transient Overloading at Salem-2 Station Transformers L. Bettenhausen, Region I, indicated that on August 26, 1986, in the process of trouble shcoting a steam generator level, an instrumenta-tion team grounded a vital instrument bus. The direct result was a reactor trip and, shortly thereafter, a safety injection. Approxi-

!- mately a minute later a blackout condition (loss of offsite power to the vital buses only) was sensed at Salem. Normal power was always available. The vital buses stripped the safety injection loads and resequenced the loads into the mode called " blackout plus safety injection." Treating the event as though there was a loss of 2

offsite pcwer, and the safety injection resulting from this initial personnel error, the operators at this point responded in accord 4 with their Emergency Operating Procedures. The operators noted a

, loss of component cooling to the reactor cooling pumps because of the resequencing. They made a conscious decision not covered in the

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4 318TH ACRS~ MEETING 19 m ,

procedures to trip the . reactor coolant pumps. They deliberately i entered natural- circulation, and stayed in natural circulation for about 26 minutes. As a result of equipment losses, the reactor coolant system pressure was controlled only'by.the pressurizer PORV. a Strictly following procedures, the operators reset safety injection l in order to begin to' restore equipment, , valves, somt punps, and other components. They restored component cooling by procedure, terminating safety injection about 21 minutes into the event, Abcut.

- 30 minutes later they restored the reactor coolant pumps which restored control ~over pressurizer pressure in the vital buses to offsite power in a proper sequence..

L. Bettenhausen discussed the station blackout signal that occurred following. transfers of. group buses from the station auxiliary . power transformer to the station power transformer. Due to low voltage,

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vital . buses transferred from one station power transformer to the other, back again, and then stripped their loads on the backout -

signal. One result of this incident is that the Licensee will do a comprehensive design . study of their AC distribution system.

J. C. Ebersole suggested that this is a case where there was load stripping and overloading of transformers. L. Bettenhausen indicat-ed that the station power transformers under these transient condi-tions ended up with undervoltage conditions on the vital bus. The transfers were taking place because'of the degraded grid undervolt-age protection relays sensing a voltage condition just below'the 91

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percent- setpoint for long enough to cause this process to start.

C. J. Wylie asked why the voltage was low. L. Bettenhausen indicat-ed that the transformers were overloaoed in the transient sense.

J. C. Ebersole suggested that they were attempting to execute starts-on loads ~ they could not carry. C. J. Wylie noted that _this was because they. transferred all the auxiliaries to that transformer.

L. Bettenhausen agreed that it was an unprogrammed load. He sug-gested 'that over r period of time loads have been added to the station's electrical system. The fast transfer would be with added' loads and the particular rack-up of the configuration of the plant at this point resulted in sensed undervoltage. J. C. Ebersole suggested that the moral was adding of loads and not examining transient starting problems. C. J. Wylie suggested that the trans-p formers may have been out of phase during the transfer.

! J. Rosenthal indicated that the Staff has recently seen a number of AC distribution failures at various plants. He cited an event at Zion and another at Turkey Point which the Staff had planned to discuss. The Staff plans to survey a few other plants and the ultimate outcome should be a generic letter or bulletin which would require response on reliability of the distribution systems to transients. W. Kerr suggested that if there is phase slippage during the transfer there will be a problem regardless of whether the transformers are overloaded. J. Rcrenthal indicated that the review of the AC distribution system at Hope Creek will include dynamic effects. W. Kerr suggested that it would be useful to know if this was a load problem or a switching problem.

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318TH ACRS MEETING 20 C. Michelson expressed concern regarding the new alternate process of presenting these operating events to the full Committee. He suggested that the time normally allotted to these -events is not sufficient to both hear about the event and and also try to under-stand it. He thought the Committee should either just hear about an event, not necessarily fully understanding its ramifications, or reduce the number of events. He indicated that he would prefer hearing about more events and not attempt to understand them com-pletely. H. W. Lewis thought 15 minutes per event would.be suffi-cient to both hear about and understand an event provided that the Staff speakers recognize that the objective is to convey their understanding of the event instead of a very detailed survey of the sequence of occurrences during the event. The Committee must also avoid any attempts to redesign particular plants. F.J. Remick stated his preference for more events, rather than a very few. He was concerned, however, that the Committee would have to have the discipline to refrain from detailed arguments concerning each event.

V. Activities of the Office of Nuclear Material Safety and Safeguards (0 pen)

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

J. Davis, Director, NMSS, mentioned Chairman Zech's September 18, 1986 memorandum to the ACRS listing those technical issues for which ACRS advice would be of particular benefit to the Commission. He noted that none of the areas specifically mentioned by Chairman Zech involve NMSS, but also pointed out the existence of written agreements between the ACRS and NMSS. He wanted to make certain that both the ACRS and NMSS continue to proceed under these agreements. He reminded the Committee of the fact that NMSS has the license issuing responsibility' for all of the domestic health and safety licenses issued by the NRC other than for power plants!

NMSS has the technical responsibility for all domestic reactor safeguards activity under the NRC, including nuclear power plants. NMSS does the Federal licensing from the front-end of the fuel cycle after mining of the are to the back-end of the cycle, including overview of DOE's demon-stration reprocessing plant at West Valley. The front-end of the fuel cycle includes commercial fuel material processing, commercial fuel fabrication and non-reactor use of radioactive and nuclear materials.

NMSS also has responsibility for the regulation of waste disposal and transportation. Under the Nuclear Waste Policy Act, NMSS is responsible for licensing the nation's high-level waste repository which is to be sited, constructed, and operated by D0E. He noted that NMSS does not have complete programmatic responsibility for all the NRC non-reactor licensees. He indicated that his office conducts approximately 5,800 licensing actions a year, both within Headquarters and in the Regions.

In the current Fiscal Year 1988 budget forecast MMSS will have about nine percent of the NRC's dollars and about eleven percent of.its personnel.

J. Davis indicated that NMSS functions are divided into three Divisions:

Fuel Cycle Facility and Nuclear Material Safety, Safeguards, and Waste

, Management. These programs are somewhat independent although there is some ccordination at times.

318TH ACRS MEETING 21 R. Cunningham, Division Director of the NMSS Division of Fuel Cycle Facility and Nuclear Material Safety, noted the Committee's activities in two particular programs in which there has been an interface between the ACRS and NMSS. These are the Spent Fuel Storage Program and the Transpor-tation Program area. He explained that most nuclear power plants current-ly have capacity for spent fuel storage into the early 1990's. A few will run to full core recovery capacity before that time. Oconee and Millstone-2 are examples and they are considering dry storage or red consolidation, or some combination of both, to solve the problem.

R. Cunningham indicated that the Staff has received a number of applica-tions for both the use of dry casks to store fuel and for the review of dry-cask storage cask designs. The first license for dry-cask storage was issued to VEPC0 in June 1986 for its NNS-5 cask made of nodular iron.

, It will hold 21 spent fuel elements (see Appendix X). The second license issued was to Robinson-2 (Carolina Power and Light) for a dry concrete module storage cask which will hold six PWR elements.

. Proposed rule revisions to 10 CFR Part 72 were discussed. Part 72 is being amended to make it clear that the licensing of the monitored retrievable storage (MRS) facility is a part of the Nuclear Waste Policy Act. The proposed rule change was issued in May 1986 and should be final, and in place, in early 1987. The rule changes will have little effect on onsite nuclear plant storage. Included in the rule itself will be QA requirements in the emergency plans which are now referenced in the existing rule as Appendix B and Appendix C to Part 50. The rule changes with regard to the MRS will include provisions for storage of commercial high-level waste. They establish a licensing period of 40 years, and an opportunity for a second discretionary hearing which is necessary because of the time involved when one issues a license that would authorize construction and cperation and the actual operation of the facility. P. G. Shewmon asked if there are temperature limits on the dry-cask storage. L. Rouse, NMSS, indicated that the casks are designed to 370*C, believed to be a fairly comfortable figure for long-term storage. H. Etherington asked the incentives and obligations of utili-ties with respect to onsite storage and ridding themselves of the spent fuel. R. Cunningham indicated that, until there is a location to which the utilities can transfer the spent fuel, they must manage their own fuel. One of their options has been to expand the present spent fuel pools that they have by re-racking or rod consolidaticn. Another option

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is dry-cask storage at the reactor site. Still another option is avail-l able. If the utility has more than one nuclear plant it can transfer i

fuel among the various nuclear plants which it owns. This last option has not been very desirable because it generates a lot of public inter-vention. The most popular option has been for utilities to re-rack their pools as far as they can. D. W. Moeller suggested that mest utilities have an agreement with 00E that 00E has to take their fuel by a certain date. R. Browning, NMSS, indicated that under the terms of the Nuclear Waste Policy Act DOE has to take the fuel from the utilities by 1998.

The current arrangements are contracts with the utilities.

R. Cunningham discussed the status of the MRS facility. He indicated that 00E was prepared to submit its proposal to Congress to build an MRS facility and obtain funding when the State of Tennessee sued based on the

318TH ACRS MEETING 22-method of site selection that DOE followed for the MRS. Tennessee blocked DOE from submitting the proposal to Congress. The matter is presently is in the Appeals Court and a decision is expected soon.

Nevertheless, DOE has no money in its 1987 budget for MRS work and the future of the MRS is uncertain. D. W. Moeller asked the NRC's legisla-tively-mandated responsibilities with regard to the MRS. R. Cunningham indicated that the NRC will license the facility.

R. Cunningham mentioned planned future revisions to 10 CFR Part 72.

These involve generic certification of casks for dry spent fuel storage.

W. Kerr noted that the Staff licenses a design rather than a cask, and R.

Cunningham agreed. R. Cunningham also indicated that the Staff is preparing a general license which will specify the conditions under which a nuclear plant can use the certified cask for dry fuel storage.

R. Cunningham reminded the Committee that the ACRS reviewed the NMSS Transportation Program in 1982. Committee recommendations cited the fragmentation of the entire transportation system regarding DOE and even within the NRC. The Staff responded by centralizing management of the program. A program area plan was developed. The plan had a safety element, an environmental element, an emergency response element, a safeguards element, and a prograziatic ccordination element.

R. Cunningham explained that a modal study was done to examine the response of casks in real-world accidents as part of the safety element.

The modal study cataloged severe accidents. Then a cask was developed or designed that just met performance standards associated with these accidents. An analysis followed that studied the loadings which would be experienced by the various severe accidents. This work, which is being done by the Lawrence Liven 11 ore National Laboratory, should conclude that the analysis that was done in the Staff's Environmental Impact Statement (considered as a basis for accepting the Staff's perfon11ance standards) is valid. W. Kerr asked if the Staff believes that the public's accep-tance of the transportation of radioactive materials will increase as a result of this study. R. Cunningham indicated that questicns usually come up in the course of hearings and meetings. Members of the public look at engineering standards and have a difficult time bridging the gap between engineered performance standards and real-world accidents. He thought this study should help to answer those kinds of questions.

R. Cunningham indicated that a low specific activity (LSA) rulemaking is being considered by the Staff. When LSA criteria were first established the material principally of interest was yellow cake. More recently nuclear power plant spent resins have been thought to fall under the definitien of LSA. Some of these spent resins have fairly high radiation levels associated with them, even though they fall within the concentra-i tion limits of LSA material. The Staff, as well as the IAEA, is con-sidering, in addition to the concentration limit, a radiation limit measured at 3 meters unshielded, as part of the definition of LSA materi-al. He noted that the U.S. Department of Transportation (D0T) is funding a study at the Sandia National Laboratories which examines the economic impact of a transportation accident involving spent resins. W. Kerr asked about the Staff's risk goals with regard to the transportation of j LSA material. R. Cunningham indicated that at this stage the Staff is

'318TH ACRS MEETING 23 attempting to quantify the economic consequences of an accident involving 1.SA spent resins. He implied that the effect on the public would be substantial. W. Kerr asked if the Staff is going to make a transporta-tion decision based upcn risk or upon public concern. R. Cunningham thought it might be a combination of both. W. Kerr expressed an interest in how the Staff- expects to quantitatively gauge public concern.

R. Cunningham admitted that it is a subjective matter that will have to be a policy decision mde by the Commission. R. Cunningham stressed that the first thing the Staff will try to do is establish the risk and then establish economic impact. If public concern is a factor the Staff will consider it after the economic impact and risk are determined. D. Okrent asked if it is a statistical economic impact or a per-accident impact.

R. Cunningham indicated that it is a D0T-sponsored study which looks at an accident in various locations and tries to set potential consequences.

D. Okrent noted that all too often hazardous materials are involved in transportation accidents. He asked, who is looking at the overall question? R. Cunningham indicated that the Congressional Office of Technology Assessment recently released a report on hazardous material transportation, including nuclear material transportation. One of the conclusions of that study was that the area of nuclear spent fuel trans-portation is by far safer than transportation of other hazardous mater-ials. D. Mausshardt, NMSS, offered to provide the Connittee with infor-mation on the D0T's study to which the NRC is jointly connitted only from a technical risk standpoint. He indicated that the demographic and social impacts are being handled by DOT. H. Etherington asked if the study looked at the consequences of a collision between a gasoline truck and a vehicle transporting spent fuel. D. Mausshardt indicated that the study looked at that very question, taking the case of an actual spent fuel shipment with two propane tanks on either side of the load (worst type of environment). It was found that the spent fuel cask had a high probability of survivability without any failure. F. J. Remick asked about spent resins and a gasoline truck fire. D. Mausshardt noted that spent resins are shipped in less than high integrity containers. This issue is being examined since the Staff is concerned.

R. Cunningham indicated that there is a proposal to use some non-specifi-cation materials as structural components in spent fuel casks. Nodular cast iron is one example. Borated stainless steel and uranium are others. There are substantial questions regarding this practice since some of the properties of these materials appear inferior to conventional stainless, and there are a number of questions about reproducibility of certain materials. J. C. Ebersole asked if the Staff monitors, or controls the quality of, transport vehic,les. R. Cunningham indicated ,

that the Staff regulates the cask. The safety of the transport vehicle comes under D0T. Nevertheless, transport vehicles with radioactive loads tend to be subject to much more stringent inspection to be sure that they meet standards. Transport vehicles nonnally do not have special require-ments over and above those that would be required of other trucks carry-ing heavy loads, but there does tend to be much more inspection of the trucks carrying radioactive materials. F. J. Remick asked if the Staff has any plans, regarding fuel cycle facilities, of pulling out of Part 50 where it is confused in the area of licensing of specialized facilities (especially reprocessing). R. Cunningham noted that there are

S 318TH ACRS fiEETING 24 expressions of interest in the area of private enrichment and the Staff is relooking at separating enrichment from Part 50.

R. Burnett, Director, Division of Safeguards, NMSS, noted the Comittee's particular interest in the area of safeguards for control rooms. He sug-gested that this is an outgrowth of the Insider Rule which NMSS discussed with the Committee in February 1986.- There was a rather controversial area in the insider package regarding psychological assesscent and behavioral observations to aid in establishing the trustworthiness of personnel that were employed at power reactors (see Appendix XI). He mentioned the NUMARC initiative in lieu of a formal rule where the industry pursuing a voluntary comitment to very similar requiremer.ts.

The industry's negotiation sections will be voted on within the next 45 days by the NUMARC steering comittee. If these sections are accepted they will be forwarded to the Comission with an accompanying policy statement which will later be issued as a formal document. It is an extensive set of guidelines which establishes background clearance requirements, psychological analysis requirements, as well as behavioral observation requirements. The industry is also committed to incorporat-ing them in safeguards plans that are required. The Comittee then went to closed session to discuss the issue of protection against truck bombs.

The material on this subject will be contained in a supplement to the Minutes.

VI. Backfittina of Regulatory Requirements (0 pen)

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

S. Crockett, Office of the General Counsel (0GC), noted the ACRS request for an opinion on how backfitting might apply to systems interactions and revised ECCS evalution modeling. He mentioned a memorandum from OGC to the ACRS with a preliminary opinion. He indicated that they were working under constraints in that the Comission was unable, because of a post-ponement, to present oral arguments before the D.C. Circuit Court of Appeals in the backfitting litigation brought by the Union of Concerned Scientists (challenging the rule in court). The Court has postponed the oral argument and the Commission has taken a rather ambiguous stand on the matter. As a result, the OGC opinion is somewhat preliminary.

S. Crockett explained that the principal conclusion, as mentioned in its August memorandum to the ACRS, was that probably the Backfit Rule did not i apply to the Staff's proposed resolution of systems interactions. OGC thought a related rule, the so-called Information Request Rule, probably did apply. Both rules were revised at the same time; both are to some extent (inaccurately) cost-benefit analysis rules. The Backfit Rule requires that the cost of the proposed action be justified by a substan-tial increase in safety, or at least a reasonable expection of a certain substantial increase in safety. The so-called information request rule is a lesser standard which requires that the cost of the information request be justified by the potential safety significance of the informa-tion sought. The two rules, if properly applied, make sense. In answer to a question by J. C. Mark, S. Crockett indicated that the standard of the Backfit Rule is that the cost of the backfit is justified by a

r 318TH ACRS MEETING 25 substantial increase in safety. He assumed that sensible persons would construe those words sensibly and that a sensible backfit would not be blocked by those rules or those words. According to the Rule, he did not think that the Backfit Rule applies in the case of systems interactions.

He argued that it would be almost impossible to justify by reasonable expectation of a substantial increase in safety. However, it would be possible to justify the costs of a request for further analysis by an argument about the potential safety significance of the information sought. One can learn more without having to apply the high standards of the Backfit Rule. The test of potential safety significance would be a matter for technical judgment. J. C. Ebersole complained that the history of the nuclear industry has been the application of mountains of complex patchwork to fix problems. He saw the study of systems interac-tions a way to identify unreliabilities and extended vulnerabilities that could be fixed. S. Crockett indicated that the Backfit Rule could only be applied by arguing that the increment in safety is significant enough to justify the cost.

D. Okrent suggested that an applicant for a construction permit has promised to build a safe plant and he is not supposed to do only what is in the regulations. If there was not a regulation requiring him to do systems interactions studies and he built his plant so as to miss many systens interactions which the Staff has since identified, he suggested that that is inadequate design on the part of the utility and not in the province of a backfit. These systems interactions should have been dealt with properly the first time. S. Crockett indicated that if the plant is not as safe as one thought it was when the license was granted the backfit rule does not apply on its own terms. You have a case of undue risk either short term or long term. In either case, one takes costs into consideration in deciding the most efficient way to bring the plant back to a level of adequate protection. Costs must be justified by a substantial increase in protection. D. Okrent argued that the Backfit Rule has nothing to do with the situation where the designer himself was deficient in what he did. The Backfit Rule applies if the Staff wants to change something from what was supposedly a safe position to another one.

S. Crockett contended that if the plant does not meet the regulations, regardless of whose fault it is, the Backfit Rule does not apply. If, however, the plant is as safe as everyone hoped it would be and it was somehow a condition of the license and the Staff is seeking more, then the Backfit Rule does apply. Whether there is a situation of undue risk in any particular case is a matter for those with technical judgment to argue among themselves. W. Kerr expressed concern that the ACRS dces not know about systems interactions that have not yet been identified. The Committee wants applicants to look for these possible interactions.

S. Crockett intended that this is not a backfit situation, but fits more appropriately under the information request rule. Some members agreed with S. Crockett that it is a case of whether there is a significant effect on safety. D. Okrent thought that " safety significance" was an ill-defined terminology, as has been pointed out by Commissioner Assel-stine. D. A. Ward indicated that W. Kerr has made the key distinction in that systems interactions is a process for attempting to identify design deficiencies. He agreed with S. Crockett that one does not get involved with a backfit until it is dealing with a particular design deficiency.

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318TH ACRS MEETING 26 9

It is no longer a systems interactions question at that point, but some particular designer hardware situation.

C. Michelson indicated that the revised Appendix rs requirements allow a licensee to propose a model for Staff review instead of meeting Staff prescriptions. S. Crockett postulated a disagreement in the course of the review of the redel. One possible situation is where the disagree-ment is simply about whether the model is sufficient. Another possibil-ity is where the Staff is not certain whether the model is sufficient or whether the range of uncertainty associated with the model is too great.

To compensate for that uncertainty, the Staff asks for some kind of surveillance or, possibly, an experiment. This is again a case for the infomation request rule. If the case is where the Staff is asking for some permanent change in procedures that will allow for surveillance, or looking for a permanent solution to the uncertainty, where the Staff is not certain what to do but asks the applicant to do something, then the question is, could the licensee invoke the backfit rule? He noted that the Staff is not obliged to do a backfit analysis before it disagrees with the licensee. What is more important is that the licensee cannot simply put the backfit rule into effect by disagreeing. What can happen i" an invoking of the appeal process under the NRC manual chapter on plant-specific backfitting. The matter would go through levels of appeal and on to the ED0's desk for final decision. The Staff would then say that it is not a backfit and, as a result, the Staff would not have to go through the analysis and apply the bottom line cost justification stan-dard in the Backfit Rule. This would be for resolution by the Staff at its highest levels. This does not rule out forcing immediate application of the Backfit Rule, but one could at least raise the question within the Staff for resolution at higher levels, whether in fact the Backfit Rule ought to apply.

C. Michelson suggested the situation in which the NRC tells the licensee he will have to do something. The licensee then says that he will not do it unless the Staff justifies it. Then the Staff will be forced to prepare a cost-benefit analysis. S. Crockett stated that it is not that simple for the applicant to say that it is a backfit--it is the Staff's decision whether it is a backfit. C. Michelson cited a case where a

licensee believes that the amendment process is being used by the Staff 4 to impose a backfit. In this process where the licensee has requested an amendment, the licensee may invoke the backfit rule under 50.109.

S. Crockett indicated that C. Michelson was quoting the Staff manual chapter. C. Michelson noted that the manual chapter is the Staff's internal guidance on how to implement the Rule. It has no legal stature.

S. Crockett indicated that there has been no effort to nake it defensible under the Rule. It does not make practical sense to allow an applicant to force application of the Rule. C. Michelson agreed. C. Michelson cited a particular case where an ECCS model submitted by an applicant, when reviewed by the Staff, is found not to have enough experimental evidence to support a particular feature of the model. S. Crockett explained that the Backfit Rule will be applied, but only after the licensee, or applicant, persuades the Staff that it ought to be applied.

Until the licensee, or applicant, takes the Staff to court, it is the Staff's judgment as to whether it applies.

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. 318TH ACRS MEETfNG '27 1

VII. NRC Standard Review Plan (0 pen)

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

D. W. Moeller explained that a joint meeting of the Severe Accident and Nuclear Plant Chemistry Subcommittees was held on September-24, .1986 to discuss the proposed revisions to the NRC Standard Review Plan (SRP) regarding the reduction of radioactive iodine in containment atmospheres.

He indicated that the Subcommittees were to review proposed revisions to Sections 6.5.2 and 6.5.3. One Section pertains to the timing or whether to add caustic to the PWR -spray in a fission product cleanup system and the other to the effectiveness of the suppression pool as a scrubber in BWRs. The Subcommittee discussed the matter with the NRC Staff. The Staff stated that they were moving ahead with these revisions but needed concurrence of the ACRS. Because of budgetary constraints, the Subcom-mittee sent the proposed revisions to two ACRS censultants, Melvin W.

First and Ronald R. Bellamy, both air cleaning experts.

D. W. Moeller cited three ccncerns that he had with the revisions.

First, he indicated that he could not find anywhere that the Staff really had investigated air cleaning practices in foreign countries and factored that into their deliberations. Second, in searching through both of the proposed revisions, he indicated finding only one reference to the proceedings of the Nuclear Air Cleaning Conferences which are the land-mark sources on this subject. Third, he indicated that he could not understand the three options the Staff was proposing in the SRP -for containment sprays for PURs. D. W. Moeller indicated that three review-ers of the proposed revisions had reservations concerning the issuance of the revised SRPs at this time. He proposed sending a short letter to the ED0 asking that the Staff withhold issuance of the revised SRPs because of sufficient concerns raised by the ACRS consultants.

P. G. Shewmon noted that the ACRS consultants did not pick up what he thought was the main justification for taking out the sodium hydroxide from the PWR sprays, inadvertent use. J. Read, NRC, indicated that what the ACRS was reviewing was the third revision of the related Standard Review Plan. He indicated that the main burden of the Staff's argument was that there is no evidence that very high pH is needed at any time to assure removal of iodine in any chemical form. There is considerable evidence to indicate that there would be organic iodide evolution if the pH were acid, and considerable evidence that if the iodine level is low enough that the pH is entirely unrelated to spray effectiveness.

P. G. Shewmon noted that one will continue to have salts in the sump, which will raise the pH of the fluid before it is recycled through the sprays. With successive runs, the pH will not rise as it would in the absence of these salts. J. Read agreed. He indicated that the Staff proposes to leave that as a solution to each individual licensee. The Staff suggested the use of trisodium phosphate or other alkaline salt in the sump as probably the most cost-effective way of achieving sufficient rise in pH. The licensees are more or less required to adjust the pH, but how they do it is left up to them. C. Michelson asked if the SRP requires that a licensee do a safety evaluation if they do not wish to change their spray arrangement. Can they leave the arrangement as they already have it in the plant? J. Read indicated that that is one option.

318TH ACRS MEETING 28 If the licensee elects to leave its spray system as it is, the SRP does not require a safety evaluation. R. Hernan pointed out that the SRP is a document the Staff uses to review new applications. There is no backfit calendar in the SRP. D. W. Moeller requested forwarding the ACRS censul-tant comments to the EDO. D. A. Ward objected to sending these comments as a Committee report and the Comittee decided not to send a letter.

The Comittee agreed that ACRS Executive Director R. F. Fraley shculd send the consultant comments, which are in the public record, directly to the ED0 and the Staff. D. A. Ward suggested writing a separate letter in which the Committee indicates that it agrees with the general approach but does not think the SRP revisions as written are ready for issuance.

The Comittee agreed.

VIII. Meeting With Executive Director for Operations (0 pen)

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

D. A. Ward mentioned the formation of the ACPS Subcommittee on Regional and It programs, noting the rescheduling of its first meeting to Decem-ber 2,1986 in Chicago. He asked V. Stello, the EDO, to comment on the ACRS interface with the NRC Regional Offices. V. Stello indicated his concern regarding two recent moves by the ACRS. The first was the December 2 meeting just mentioned and the second, the request to have ACRS involvement in actual NRC plant inspections. He asked whether this was a new departure for the ACRS and wondered what the ACRS expects to accomplish by the new actions. He noted Staff briefings on numerous occasiors in the past on what activities occur in the regions. He also noted that the ACRS has been intimately involved in regional activities as a result of ACRS subcomittee meetings, such as joint visits to sites and interactions with licensees. J. C. Ebersole suggested that the Comittee would like to explore whether there is suffici nt quality and regulatory uniformity from region to region now that the field offices have bcen broken up into regicnal administrations. V. Stello indicated that this breakup has been in process for many years. J. C. Ebersole noted, however, that control has been focused at headquarters until the recent change. The uniformity of regulatory control was enhanced by the fact that it was centralized. V. Stello suggested that an answer to this question might possibly require a special briefing.

D. Okrent brought up the Committee's interest in the quality assurance issue which arose during a Committee visit to Oyster Creek. He suggested that the Committee does not need to have a specific faulty area in mind if it thinks it would be useful to lot k at an area or wish to gain insight by having a specific meeting wita a regional office. V. Stello brought up the need for preparation in the region if the Committee wished to discuss a specific issue at such a subcommittee meeting at a regional office. V. Stello suggested that if the Committee needs information on a particular site or particular utility the best advice would be to go to that site to explore that issue. He thought ACRS input would be extreme-ly useful to setting up a particular inspection during which Committee members wish to participate.

318TH ACRS MEETING 29 F. .J. Remick indicated that a specific agenda had already been prepared

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4 for tht Chicago meeting, but hc was not sure what the EDO was requesting regarding the inspection issue. V. Stello indicated that he was informed-that the Committee wished to have members accompany the Staff on a fire

. inspection of a nuclear facility. D. A. Ward indicated that this request-was entirely disconnected from the December 2 regional meeting.

C. Michelson suggested that the regional visits were intended for educa-tional purposes as well as to establish better communication with region-al office personnel. W.-Kerr indicated that it was his perception that, with the larger number of operating' reactors that now exist, the regicns have become more active in making decisions about ouestions that arise in -

connection with operating plants. He thought that the ACRS ought to be up to date on regional activities with respect to operating plants. He expressed his interest in exploring the issues of authority and responsi-F bility being given to the regions. V. Stello pointed out that regional authority regarding the issuarice of amendments such as ordering a plant shutdown hre unchanged from what they were 15 years ago. D. W. Moeller' noted his recent visits to two different plants with INP0 on INP0 evalua-tions. He expressed he interest in going with an NRC IE inspection group to learn what they do and how they do it. V. Stello still ex-

pressed confusion as to what the ACRS wished to accomplish, and he indicated his need to authorize the allocation of resources to be consis-

, ' tent with the ACRS participation. D. A. Ward referred the EDO to the Connittee's letter. of July ' 21, indicating that it was a reasonable

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statement of the Committee's intentions, i

D. Okrent indicated that some who have tried have not found a good correlation between- SALP reports and other kinds of performance indica-tors when they have. looked at a limited class of plants. He suggested that there may be a difference in how the regions do ratings. He ex-

> pressed an interest in how much inspectors at plants, as well as the STA

.and the SR0s, really know about management of severe core accidents.

V. Stello suggested that these questions were generic in nature. He did not deny that there are some differences with regard to SALP ratings

region by region, but this is also a generic question. He thought that one would have to go to the sites to find out what the STA and SR0s as well as resident inspectors know about severe core accidents, rather than 90 to the regional offices. He again suggested that ACRS questions are

.nore suitable to generic briefings by IE which is charged with these i responsibilities. -He also invited ACRS members to resident meetings i which occur on a quarterly basis at the regions. Committee members could raise the types of questions previously mentioned at these gatherings, r C. Michelsen saw little differences between this proposal and three or 4 four ACRS menbers going to a region to. talk around the table.

1 J. Keppler, Region III Administrator, 'noted that there were certain controversial issues in the agenda for the December 2 meeting. He <

! suggested that there might be a spectrum of views on certain subjects i mentioned. V. Stello reminded the Connittee that it should interface

with J. Keppler so that the proper individuals would be present at this particular meeting. On a generic basis, he suggested that the Committee go directly through the ED0's office in the future regarding regional contacts, i

4

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• C 318TH ACRS MEETING 30 IX. International Operating Experience (0 pen)

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

W. Kerr called the Committee's attention to his coments on the 'IAEA'

meeting in August 1986 on the Chernobyl accident (see Appendix.XII). He indicated that the USSR ' team headed by V. A. Legasov was extremely .

. competent and, during its extensive presentation, did not hesitate to depart occasionally from their written report. On the second day there were presentations by other USSR delegates on specific topics. Questions ,

. were solicited and about 600 questions presented. These were culled by

an expert group to a manageable number of about '20-30 questions. He

. thought it a very fruitful exchange in terms of literally discussing 1 items that were important and interesting. There were several smaller

working group sessions. .One, in particular, involved medical people

interacting with the USSR medical and radiation effects experts.

! M. Eisenbud and. M. Goldman were both . convinced that the Russians were -

!~ overestimating the number of cancer casualties predictable from the

cesium released at Chernobyl. All agreed that these initial estimates from the Soviets were based on a worst case which could reasonably be lower by a factor of 10. He noted that the Soviet estimates for

! whole-body dose appeared to be averaged over the population of European USSR. American' recalculations estimate the dose to be about double the background radiation expected. W. Kerr indicated that Soviet estimates

of exposure of the population .within the 30-kilometer radius from the plant averaged about 3.3 rems. He noted' that M. Eisenbud was skeptical

, of this ' number because of the uncertainty associated with it. It was i noted that 'some firefighters during the Chernobyl accident received doses ,

i above 1000 rems whole-body. J. C. Mark asked what were the largest i offsite doses reported. W. Kerr indicated that the largest numbers were 25-30 rem whole-body. He noted that no one offsite was hospitalized.

The Comittee discussed various dose estimates and total radicactivity l released.

. W. Kerr reported a discussion at the IAEA meeting about the particular

i. Chernobyl design and the fact that the Soviets are still building similar i plants and have plans for building some larger ones. It was noted that i the Soviets claim that they did not have the technology to build a large e pressure vessel when graphite-moderated reactors were designed, but they

do have this technology now since they have built PWRs. He mentioned the L advantage the Soviets have gained from the elimination of the need for steam generators. He also mentioned the fact that they knew about the i positive void coefficient at the Chernobyl plant. Mentien was also made i of the much better online experience the Soviets have over the U.S. in

part because of online fuel replacement. He speculated that this design

! evolved because of its use in weapons material production. Because cf L the positive void coefficient the operators have local control of control rods despite the fact that there is no differentiation between control i and safety systems. Because of the possibility of local instability on

! actual power oscillations within the reactor core, the control system has

! to be fairly sophisticated. The control system has safety functions and i at certain points one would insert control rods and bring the power level i down. Apparently, the Soviet philosophy was influenced by their ultimate f goal to keep the plants on line, producing power. W. Kerr noted that at i

L,__-~_-__._,__.,_. ,,__- - ,--_ ,._ _ _ _ _ ___ _ _._ _ _.,- c_ _ _.,_ _

318TH ACRS MEETING 31 one point in the discussions someone asked if there were any precursors to the Chernobyl accident and the answer was no.

H. W. Lewis indicated interest in the procedures during the loss-of-power test which precipitated the Chernobyl accident. He expressed skepticism l about the effectiveness of procedural instructions when all of the safety systems were shut down at the Cherncbyl plant. W. Kerr indicated that it was his impression that the instructions to the operators were relayed by individuals in charge of the tests. He did not know who was responsible for making decisions, such as pulling out all of the control rods, but the normal operators must have taken the action. H. W. Lewis asked why the operators obeyed those instructions. W. Kerr indicated that the Soviets, as well as the team of operators at the plant, were as puzzled as everyone else. Perhaps the operators had difficulty believing that this accident was credible. It was obvious that the test procedure violated operating procedures regarding the shutoff of safety systems.

The Soviets said that the review of the test procedure by the station manager was not thorough enough. The test had been performed once before with the plant shut down. Perhaps those in charge did not think that running the test at 20 percent power would be much different.

The Committee briefly discussed the course of the accident. W. Kerr noted that the Chernobyl reactor was prompt critical by about half a dollar in reactivity, where a dollar is about 0.5 percent Ak. He noted that the Soviets do not have very much data about the accident, in part because most of their instrumentation (used for control) is very slow response, subpower detectors. They do not have many fission detectors for fast response detection to handle what happened. The Soviets do not have a gcod idea of what happened to core power density. There is some question whether the core acted as a single unit or as many prompt-critical separate units. The Comittee discussed the efforts of fire-fighters who knew of the dangers and, nevertheless, made the sacrifice.

W. Kerr mentioned the fact that Chernobyl, Unit 3 was kept on line while the fire burned at Unit 4. Ventilation systems continued to operate and, as a result, Unit 3 was contaminated. Extinguishing of the fire in Unit 4 was necessary to save Unit 3. When the fire was under control, the Soviets considered evacuating the yard, Chernobyl, and then the 30-kilo-meter radius around the plant. The principal objective of putting the fire out was to avoid damaging Unit 3.

The Committee discussed the reasons for putting out the fire at Cher-nobyl, Unit 4. Allowing the fire to continue to burn would keep the temperature down to the temperature of t'urning graphite; with the fire extinguished the temperature is likely to rise because of the inability to dispose of decay heat frcm the core. The Soviets decided to extin-guish the fire because of the large amounts of aerosols being released from the reactor building. H. W. Lewis asked why the Soviets dumped lead at the beginning. W. Kerr indicated that boron carbide was first durrped and, then, dolomite, just for cooling purposes. The lead was for cooling but would also dissolve a significant number of the fission products, as well as providing shielding. A large amount of sand and clay was dumped after the lead.

l .

H. W. Lewis was impressed by the speed with which the Soviets responded to the Chernobyl accident. W. Kerr speculated that they may have had experience with an accident like this before, especially with regard to medical treatment. The marshalling of medical resources was particularly remarkable. None of the firefighters had radiation badges. Doses were estimatad based upon bloed examination and the fluorescence excited in the fibers of their clothing.

J. C. Ebersole mentioned the fact that there was a second explosion. He wondered if this was due to the dropping of the control rods that were available. W. Kerr indicated that a number of individuals raised ques-tions about the second explosion. The Soviets were unable to explain the second explosion, and this appeared to be a low-priority item for the U.S. at the time. One hypothesis was that the first explosion was l probably caused by energy input to the reactor core, and the second j explosion due to hydrogen. H. W. Lewis wondered whether there was enough time for hydrogen to accumulate after the first explosion.

H. W. Lewis asked what W. Kerr thought the most important lesson for the l U.S. from this accident. W. Kerr thcught it was a well-trained, well-disciplined operating staff at a nuclear plant from management on down.

I l W. Kerr indicated that the Soviets are considering increasing the U-235 enrichment from 2.0 to 2.4 percent as a fix. F. J. Remick asked if this was so that they could have more control rods in the core. W. Kerr explained that the increase in enrichment would reduce the positive void l coefficient because of a greater absorption of neutrons in the fissionable material. F. J. Remick speculated on addition of more U-238 s to increase the Doppler coefficient. W. Kerr agreed that one could do that and it would necessitate more control rods and decrease the positive l

coefficient. Nevertheless, this has a negative influence on neutron economy and power generation. Another reason the Soviets are putting more absorber rods in the core is because, with the slow drive speed they use, one would have a more rapid insertion of negative reactivity, j F. J. Remick pointed out that in order to remain critical, if one uses j more control rods, one would have to have more U-235. D. Okrent and l J. C. Ebersole noted that BWRs in the U.S. tend to be autocatalytic in that collapse of voids causes a positive reactivity insertion.

J. C. Ebersole wondered if this might be one of the lessons to be learned from the Chernobyl accident. W. Kerr disagreed that BWRs are autocata-lytic.

When a BWR trips, you introduce positive reactivity until you collapse the voids, at which point the introduction of positive reactivi-ty would stop. D. Okrent pointed out that in a BWR one wants to lose reactivity as the power goes up. This is what tends to yield stable operation, stable under natural circulation. He agreed that the BWR scenario they were discussing was not autocatalytic in the usual sense.

The usual definition of the term is, if power goes up reactivity goes up.

He did point cut that there are power reactors that nave the capability for a super-prompt reactivity transient in the U.S. In fact, this is what an ATWS is all about on a PWR. J. C. Mark noted that the reactors are still not autocatalytic in the usual sense. O. Okrent agreed.

J. C. Ebersole indicated that they are autocatalytic in the sense that the progression of an ATWS further increases the pressure even thnugh the safety valves have blown open. That increases reactivity and finally

T

+

, 318TH ACRS MEETING 33 causes ' loss of the reactor vessel. H. W. Lewis thought the i.ec.n 'outo-catalytic" was being used incorrectly since it has a chemical meaning and actually refers to an instability. The basic problem'with graphite as a moderator is that it moderates without absorbing unless one increases the enrichment. W. Kerr noted that it was said that the graphite in the Chernobyl reactor actually has a positive reactivity coefficient.

W. Kerr indicated that the Soviets are considering the use of some dedicated safety rods as a departure from their current design. The Committee discussed the functioning of the control-rod-control system on the Chernobyl reactor. D. A. Ward noted that the automatic control system does not function well at very low powers, as the operators have to keep juggling the rods.

H. W. Lewis concluded that the Chernobyl accident is really a limiting accident since it was a very large core and a large fraction of the core inventory was dumped into the biosphere. D. A. Ward agreed that that was an important conclusion, the fact that the accident was not exactly a cataclysmic event.

D. Okrent -indicated that .the NRC Staff is trying to understand what happened at Chernobyl and to draw implications from the accident.

R. Hernan indicated that a Commission paper on the subject is due by December 31, 1986. D. Okrent mentioned that a national laboratory report may be out'by the end of October 1986. He suggested that the Committee obtain a copy of the NRC expert task force report which would be out in December, and schedule another session of the full Committee to discuss this matter. H. W. Lewis suggested that, in considering the implications of Chernobyl, the ACRS should consider also the DOE reactors. He thcught that DOE ought to ask the ACRS for help in its review of the DOE reac-tors. D. Okrent noted that the ACRS once did review DOE production reactors but decided to end the practice in the early 1970's.

H. W. Lewis pointed out that, after the split of the AEC in 1976, the ACRS did not have the right to review the DOE reactors.

H. W. Lewis noted that one of the indications of the Chernobyl accident is that Chernobyl was a low-probability accident--if there is a lesson, it is that one must go beyond and deeper than just the hardware problems thrt are usually discussed by the ACRS. D. Okrent asked if there are any other suggestions from members for possible implications to be examined during the subconinittee's first meeting on November 5, 1986. He en-couraged as many members as possible to attend this meeting, i

F. J. Remick suggested investigation of the fire protection aspects for

j. U.S. utilities, such as the protection of ~ firefighters. J. C. Mark cautioned that aspects of the Chernobyl accident ought to be put in proper prospective so that the end result is not a vast number of differ-ent and possibly irrelevant prescriptions for plants. H. W. Lewis thought the misunderstandings about the nature of the explosion which occurred during the Chernobyl accident ought to be investigated. It was originally called a steam explosion. D. Okrent noted that there are still some individuals in the U.S. who think it may have been a steam explosion although not autocatalytic in the usual sense. D. A. Ward stated that it was just an explosive thermal expansion.

,- 318TH ACRS MEETING 34

-X. Executive Sessions (0 pen)

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

A. Subcomittee Assignments

1. Instructions for the Nominating Panel The nominating panel for ACRS officers for CY-1987 (D. W.

Moeller, Chairman) was asked to consider for the office of ACRS chairman the member "best able to run the Ccmittee," not necessarily the most senior individual. It was noted that nominations will also be accepted from the floor.

B. Reports, Letters, and Memoranda

1. ACRS Suogestions for an NRC Long Range Plan The Ccemittee prepared a report to the Comissioners of its thoughts and recomendations on a Long Range Plan for the Nuclear Regulatory Comission.

2. ACRS Coments on Draft NUREG-1225, "Im)lementation of NRC Policy on Nuclear Power Plant Standardizatnon" The Committee prepared a report to the Comissioners of its review of the draft NUREG-1225, " Implementation of NRC Policy on Nuclear Power Plant Standardization." The ACRS would like to be kept informed regarding this matter.

3. Clinton Nuclear Power Station - Resolution of ACRS Coments The Comittee prepared a memorandum to the EDO which indicates that the ACRS is satisfied with the resolution of ACRS concerns in its letter of March 9,1982 to Chairman Palladino concerning Illinois Power Company's request for an operating license for the Clinton Nuclear Pcwer Station.

4 IDCOR Documentation Availability and Staff Review The Committee is currently reviewing the NRR Implementation Plan for the Comission's Severe accident Policy Statement. An integral part of the review is the IDCOR Methodology for Individual Plant Examinations. The ACRS has provided a memo-randum to the ED0 asking that all IDCOR documents be provided to the ACRS in a regular and timely manner, that the NRC Staff should review the IDCOR Modular Accident Analysis Program (MAAP), and that all IDCGR documents should be placed in a retrievable file available to all interested parties.

e 318TH ACRS NEETING 35

5. Comments on Proposed Revisions to Standard Review Plans 6.5.2 and 6.5.3 The Committee prepared a memorandum to the EDO regarding its discussion of portions of the NRR Implementation Plan for the Severe Accident Policy Statement, regarding proposed revisions to Standarc Review Plan (SRP) Sections 6.5.2, Containment Spray as a Fission Product Cleanup System, and 6.5.3, Fission Product Control Systems and Structures. The ACRS would like an oppor-tunity to review these revisions prior to issuance in final form.

The Comittee also approved a memorandum from the ACRS Execu-tive Director to the ED0 to transmit comments on the draft revisions of the SRPs by two air cleaning experts who consulted for the ACRS Subcommittees on Severe (Class 9) Accidents and Nuclear Plant Chemistry and by Dr. lioeller.

6. Containment Performance Issues The Comittee approved a memorandum from the ACRS Executive Director to the ED0 expressing ACRS interest in further par-ticipation in the development of a proposed generic letter regarding Mark I containments to accomodate severe accidents.

C. Generic Issues

1. Basis for Standardized Nuclear Plant Improvements J. C. Ebersole introduced a proposed report to supplement the Committee's report of August 12, 1987, subject: Proposed NRC Standard!zation Policy Statement and the Staff-Proposed Imple-mentation Plan. Further discussion of this report, which deals in " standard" withimprovementsinsafety(November for the 319th ACRS meeting 1986) plants,wasscheduled

2. Interpretation of NRC Safety Goals in Terms of the Population Doses Associateo with Nuclear Power Plant Accidents D. W. Moeller introduced and the Committee discussed a draf t report regarding a cuantitative interpretation of the NRC's safety goals in terms of population doses that might be associ-ated with major nuclear power plant accidents. Further dis-cussion of this subject is scheduled for the 319th ACRS meet-ing.

3. Discussion of Sionificant Safety Issues During the discussion of future ACRS activities H. W. Lewis discussed the need for enother Comittee brainstorming session on significant safety issues confronting the nuclear industry.

D. Okrent suggested that the topic of emergency planning critoria guidelines be added to the items for discussion at such a meeting.

.' 318TH ACRS HEETING 36

4. Authorization of the Backfit Rule The Committee asked that the ACRS Executive Director request a formal written opinion from the NRC General Counsel regarding the authority of licensees to invoke the Backfit Rule.

D. Future Agenda

1. Future Agenda The Comittee agreed on tentative agenda items for the 319th ACRS Meeting, November 6-8, 1986 (see Appendix II).

2. Future Subcommittee Activities  !

A schedule of future subcommittee activities was distributed to members (see Appendix III).

E. Seabrook Station Probabilistic Safety Assessment, Risk Management and Emergency Planning The Comittee heard presentations on and discussed the Seabrook Station Risk Management and Emergency Planning Study (RMEPS) and its companion Seabrook Station Emergency Planning Sensitivity Study (SSPSA) performed by the licensee with the ultimate goal of reducing the emergency planning zone from 10 miles to a 1-2 mile radius. The ACRS heard several oral and written statements by members of the public regarding emergency plans in the vicinity of the site. The Committee decided to take no action until the NRC Staff has finished and reported to the ACRS on its review of the RMEPS and SSPSA.

F. Shearon Harris Status Report R. Hernan, NRR, requested time on the December agenda (320th ACRS meeting) to present a status report update regarding the January 16, 1984 ACRS Letter to be documented in SSER No. 4 (10/86). The Committee decided that it would like the Staff to forward a written reply regarding resolution of the issues discussed in the January 16, 1984 ACRS report before it considers a Staff briefing on Shearon Harris.

G. Proposed Revision of ACRS Bylaws The Committee approved Proposed Amendment No. 3 to the ACRS Bylaws as revised which pertains to Section VIII -- Election of Officers.

This Amendment provides for one-year terms by the Chairman and Vice-Chairman with a note that they can be reelected for a second sequential term if the Committee desires. A recall provision was also included. Proposed Amendment 3 (Alt) was not approved by the members. The specific previsions of Proposed Amendment No. 3 are attached as Attachment 1.

i i The Ccmmittee approved Proposed Amendment No. 4 to the ACRS Bylaws,

! regarding a new section XV entitled " Planning Committee" with the l " preceding Chairman" deleted as a member of the Planning Committee l

~

i g 318TH ACRS MEETING ~ 37 s*

w and a' limit of four years on the length of service of any member was added. (See Attachment 2).

A proposed revision to adjust quorum requirements near the end of ACRS meetings was discussed and dismissed (Proposed Amendment No.

2). A proposed amendment to Section IV Conduct of Meetings regard-ing use of written proxies of members to complete a quorum for specific decisions (Proposed Amendment No. 5) was also turned down.

A revised wording proposed by OGC of a " Proposed Addition to Section IX -- Conduct of Members of the ACRS Bylaws" (Proposed Amendment No.

1) regarding meetings of individual ACRS members with individual Commissioners was approved by the Comittee (see Attachment 3).

H. Retirement of ACRS Member Emeritus H. Etherington announced that he will retire from his position as ACRS Member Emeritus, effective December 31, 1986. He has requested that the ACRS Staff discontinue sending him documents other than ACRS memoranda.

I. Notice of Awards Chairman Ward in his report to the Committee at the beginning of the 318th ACRS meeting announced the award to ACRS member C. P. Siess of the NRC's Distinguished Service Award and the award of the Meritori-ous Service Award to ACRS Assistant Executive Director A. L. Newsom for their outstanding contributions to and support of ACRS activi-ties.

The 318th ACRS Meeting was adjourned at 2:25 p.m., Saturday, October 11, 1986.

, 318TH ACRS MEETING 38 o

ATTACHMENT 1 PROPOSED AMENDMENT TO SECTION VIII -- ELECTION OF 0FFICERS A. The Committee Chairman and Vice-Chairman shall be elected to serve for one year commencing on January 1 and ending on December 31.

Either or both may be reelected to serve no more than one addition-al consecutive one year term. Either or both are subject to recall by a vote of two thirds of the Committee members. The motion for recall shall be made, seconded and discussed during one meeting and voted upon at the next meeting.

B. In the event.the Chairman is unable or unavailable to carry out his duties for a limited period, the Vice-Chairman shall act as Chair-man.

C. In the event that either the Chairman or the Vice-Chairman (or both) is (are) unable to continue to serve, a special election will be held to fill the position (s) for the remainder of his (their) term (s).

D. Reaular Elections The Chairman shall appoint a nominating Committee of three members during September of each year. The names of the members of the nominating comittee shall be announced at the September meeting or if no meeting is scheduled for that month in writing to the ACRS.

The nominating comittee will report its slate for Chairman and Vice-Chairman at the beginning of the next-to-last regularly sched-uled meeting for the year. Nominations for officers will also be accepted from the floor by members of the committee. The nomina-tion and election of an ACRS Member (s) at large to fill vacancies on the ACRS Planning Committee will also occur at this meeting.

Nominations will be made from the floor by the Committee members at large. The member (s) selection for this position will serve for the following calendar year.

The Committee Chairman and Vice-Chairman for the following year shall be elected during the last regularly scheduled meeting of the

. yea r.

E. Special Elections In the event that a special election is required under the terms of C. above, the Chairman or Vice-Chairman (whichever office is not at stake in the election) shall select a nominating committee of three members. If both are at stake, the next most recent available ACRS Chairman shall select the nominating committee and shall act as ACRS Chairman until the election is completed. The nominating committee will report its slate to the ACRS at the next regularly scheduled meeting. Officers will then be elected at the next regularly scheduled meeting. Nominations may be made from the floor.

318TH ACRS MEETING. 39 o-ATTACHMENT 2 Proposed Amendment No. 4 Add a new section XV. Planning Committee A. A subcommittee, known as the Planning Committee, shall be established to discuss the priority of ACRS work, to assign appropriate resources and to recommend to the ACRS both near-and long-term goals. The Planning Committee will be composed of:

1. the current ACRS Chairman;

2. the Vice-Chairman;

3. an ACRS Member (elected by the ACRS Members to serve concurrent term with the Chairman /Vice-Chairman); and

4. the Executive Director;

5. No ACRS Member shall serve more than four years.

• -318TH ACRS MEETING 40

.o ATTACHMENT 3

" PROPOSED ADDITION TO SECTION IX -- CONDUCT OF HEMBERS OF THE ACRS BYLAWS "Though not explicitly constrained to do so by the enabling legislation, the Comittee has historically functioned as a collegial body, focusing the Merbers' disparate views into a comon position. For this reason it is inappropriate for an individual Member to attempt to interpret Committee reports, recommendations, or actions except as authorized by the Comittee (seeSectionV.D.).

" Individual 'Hembers are always free, as individuals, to interact and communicate with individual Commissioners. This channel will nonnally, but not always, be opened by the relevant Comissioner seeking information, and it should always be clear that the Merrber is not representing the Committee, but is functioning as an independent expert. Such contacts, where substantive, should be noted to the Chairman or to the Executive Director."

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i SUPPLEMENT - NUCLEAR PLANT SECURITY ]

DRETION 3 Due to Proprietary Information

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APPENDIX I ATTENDEES 9 NRC ATTENDEES 318TH ACRS MEETING Thursday. October 9, 1986 0FFICE OF NUCLEAR REACTOR REGULATION Ronald W. Hernan David H. Moran B. Siegel D. Terao A. Lee W. R. Butler C. R.. Van Mel REGION III F. Jablonski, DRS .

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PUBLIC ATTENDEES 318TH ACRS MEETIliG Thursday, October 9, 1986 J. Trotler, NUS -

L. Rib, LIVRA P. F. Collins, KMC, Inc.

Eve Fotoroulos, SERCH Licensing, Bechtel L. Connor, DSA -

P. F. Riehm, KMC J. E. Kan, Stone & Webster Electric Corp.

R. E. Schaffstall, XMC. Inc.

M. Levy, Nuclear Control Inst.

Reyner, A.C.S.

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INVITED ATTENDEES 318TH ACRS MEETING Thursday, October 9, 1968 ILLIN0IS POWER COMPANY D. L. Holtzscher A. J. Hable P. J. Telthorst

• J. H. Greene R. E. Campbell D. P. Hall W. Conner G. Edgar T. A. Spangenberg A. L. Ruv J. Geier GENERAL ELECTRIC D. K. Henrie J. E. Leatherman ,

SARGENT A LUNDY A. K. Singh D. K. Schopfer H. M. Spoka

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['\i t NRC ATTENDEES V 318TH ACRS MEETING Friday, October 10, 1986 0FFICEOFNUCLEARREACTORRep0LATION R. W. Hernan H. Pastis B. S. Marcus M. Haughey K. S. West H. Reponen 0FFICE OF NUCLEAR ftATERIAL SAFETY & SAFEGUARDS J. R. Cook W. C. Walker C. E. MacDowann M. Dunkelman 0FFICE OF INSPECTION & ENFORCEMENT E. W. Weiss R. J. Jolliffe O J. Rosenthal

(- K. P. Wolley H. A. Bailey J. C. Stewart OFFICE OF STATE PROGRAMS F. Young J. T. Wiggins, Region I L. H. Bettenhanar, Region I T. Peebles, Region II A-4

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4 I-INVITED ATTENDEES

, 318TH ACRS MEETING 1

i

_ Friday, October 10, 1986 i New Hampshire Yankee i J. Grillo J. Kyte R. White J. Vinlentis

{ K. L. Kiper

B. Derrickson i

3 Yankee Atomic Electric Company

J. W.Stacy D. Maidrand 4

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{~h f \ PUBLIC ATTENDEES 3187H,ACRS MEETING Friday. October 10, 1986 P. F. Riehm, KMC N. Rutherford, Duke Power F. Stetson, NUS CORP.

M. Wetterhahn, C8W S. Letourneau, Bechtel D. Airozo, McGraw-Hill H. M. Fontecilla, Virginia Power R. E. Schaffstall, KMC S. Bogan, Potomac News Rep. Beverly Hollingsworth, State of New Hampshire M. Fallon, Sun Valley, Republicans J. Doughty, SAPL R. Gardner, Stone & Webster B. Montuilue. Hampton, N.H.

H. Brown, ABC News ABC News J. Lunaq,PLS A. Torre D. Johnson, ABC news -

P. Branch, Mass. State Office -

S. Markiewicz, Mass. State Office R. Toland, United Engineers K. Unnerstall, Newman & Holtzinger L. Correia. Energy & Connerce R. Lee, WNEV-TV D. D'Arrigo. NIRS M. C. Keegan, Cong. Marroules L. J. Toth, Gasser Associates R.Sweeney, Self -

L. M. Cuolo, Fried Frank K. Woodard, PLG D. W. Stillwell, PLG L. Finkelstein, Newman & Holtzinger S. Feldman, Potomac News P. Glass Ottaway, News Serv.

K. Silberugh, NECNP

~

APPENDIX II

• FUTURE AGENDA APPENDIX A b FUTURE AGENDA NOVEMBER ACRS MEETING i

Selection of Nuclear Power Plant Personnel -- Proposed ACRS 3/4 hour .

report regarding use of aptitude testing in selection of nuclear power plant personnel Proposed Revision of NRC Regulatory Guides -- ACRS coments i hour Meeting with Director, RES -- Discuss items of mutual interest 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> including:

- Consideration of risk reduction potential in the prioritization of research activities

- Safety research activities and priorities in connection with nuclear plant fire protection provisions i

Activities of NMSS -- Briefing by R. Browr.ing, Director of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Division of Waste Management, NMSS, regarding items of mutual interest in the area of waste management Meeting with NRC Commissioners -- I hour Potential topics to discuss:

- ACRS comments regarding NRC Long Range Planning

- Additional guidance and discussion of the role that Chairman Zech and the other Commissioners would like l the Comittee to fill on the five year plan l - New ACRS members  ;

Containment Performance Design Objectives -- ACRS coments 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> Systems Interactions -- ACRS coments regarding proposed 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />  :

, resolution of ACRS May 13, 1986 coments on proposed  !

resolution of US! A-17, Systems Interactions in Nuclear  ;

Power Plants

_ Safety-Related Modifications in Paluel Nuclear Power Plant -- I hour i

Discuss matters of general interest regarding the safety implications of these changes Wingspread International Conference on Reactor Safety -- i Discuss implications of information provided/ discussed at the j Wingspread International meeting Nomination of Officers for CY-1987_ -- Report of the nominating

panel ,

I V

l

~

/O 318TH ACTIONS AND AGREEMENTS Y '

Improved 1.WRs -- ACRS report to NRC regarding characteristics 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> i of improved LWRs Prioritiration of Generic Issues -- ACR$ coments on the 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> adequacy of the NRC 5taff - proposed priority rankings of a fourth group of generic issues j Regulatory Guides -- ACRS comments regarding proposed revision Deferred to  :

of Regulatory Guide 1.99, Revision 2, Radiation Damage February i February General Design Criterion 4 -- ACRS comments on the broad scope Deferred to t rule regarding pipe failure in primary coolant systems January BWR Pipe Crack Guidance -- ACRS comments regarding incorporation Deferred to i of public coments into NUREG-0313 Revision 2 regarding BWR January / I pipe cracking February Severe Accident Research -- Resolution of issues related to Deferred to severe accident policy implementation, status report by RES January / '

February GE - Advanced BWR -- Briefing by NRC Staff regarding ground Deferred to rules for review by the Staff December ,

NRC Staff Briefing on the Implications of Chernoby1 Accident Deferred to December ACRS Subcommittee Activities

• Maintenance practices and procedures -- Phase I i hour of the NRC Maintenance and Surycillance Program Plan

• IE Programs -- Activities of the NRC Office of i hour Inspection and Enforcement  ;

• Safety philosophy, Technology and criteria regarding: 3/4 hour

- Chernobyl implications

- Status of the NRC Staff's work on steam generator overfill ,

• Waste Management -- Report of review of selected radio- i hour i active waste management topics with NMSS and RES l t

• Planning -- Allocation of ACRS resources t

F8 l

REYtstc II %

O ACRS SUBCOMMITTEE MEETINGS APPEllDIX III Q ACRS SUDCOMt1ITTEE t1EETINGS Wingspread International Conference (Closed), October 19-23, 1986, Racine, WI (McCreless). Representatives from the ACRS, RSK, GPR, and Japan will eichange information on nuclear reactor safety.

Waste Management, October 30 and 31, 1986, 1717 H Street NW, Washington, DC (Merrill), 8:30 A.M., Room 1046. The Subcommittee will review the T611owing radioactive waste management topics: (1)theBWIP(Panford) site, including health and safety issues, (2) the Staff's review of DOE's ,

five final Environmental Assessments (3) assessing compliance with the EPA Standard, (4) rulemaking to confom Part 60 to EPA Standard, (5) states' implementation of LLRWPAA, (6) alternatives to shallow land burial. (7) safety assessment of alternatives to shallow land burial, and (8) status of NRC waste package corrosion program. Lodging will be announced later.

Attendance by the following is anticipated:

Dr. Moeller Mr. Till Dr. Carbon (tent.) Dr. Orth Dr. Kerr Dr. Parker Dr.Remick(Oct.31only)

Dr. Shewmon 5, 1986, 1717 H OSafetyPhilosophy, 5treet,(NW,1) continue Technology,andCriteria, November Washington, its reviewDC (Savio),

of US! A-17. "9:00 A.M.,

Systems The Subcomittee Room 1046.

Interaction in Nuclear will:

Power Plants " (2) review the status of the NRC work on steam generator overfill,(3}discussthestatusoftheNRCStaff'sdevelopmentofaSafety Goal Policy Imple.nentation Plan, and (4) discuss the implications of the Chernobyl Accident. Lodging will be announced later. Attendance by the following is anticipated:

Dr. Okrent Mr. Michelson Dr. Carbon Mr. Ward Dr. Lewis Mr. Wylie Planning Subcommittee (Closed), November 5, 1986, 1717 11 Street, NW, Washington, DC (Fraley), to convene after__SPTC w atino Room 1010. The Subcomittee will discuss allocation of ACh$ resources to various subcomittae activities. The Subcomittee will also consider thtt memo from Chairman Zech dtd. September 18, 1986 ra items on which the Comittec is requested to concentrate. Proposed ACRS participation in hRC review of EPRI Improved Standard Plant Guide, plans fo" review of the ABWR and the Westinghouse APWR and ACRS review of the WPPSS 1 reactor plant as replacement for Hanford N-Reacter will also be diccussed. Lodging will be announced later. Attendance by the follow 4g is enticipated:

Mr. Ward Dr. Lewis Mr. Eborsele I'

E

O 319th ACRS Meeting, November 6-8, 1986, Washington, DC, Room 1046.

Extreme External Phenomena November 20, 1986, 1717 H Street, NW, Washington, DC (Savio), 8:30 A.M., Room 1046. The Subcomittee will continue its review of the Diablo Canyon long-term seismic program and the NRC Staff's Seismic Safety Program. Lodging will be announced later.

Attendance by the following is anticipated:

Dr. Stess Dr. Moeller Mr. Etherington Mr. Wylie Dr. Lewis Saent Fuel Storace, November 21, 1986, 1717 H Street NW, Washington. DC (Merrill), 8:30 A.M., Room 1046. The Subcommittee will continue its review of10CFRPart72andMonitoredRetrievableStorage(MRS). . Lodging will be announced later. Attendance by the following is anticipated:

Dr. Siess Dr. Remick Dr. Kerr Dr. Shewmon Dr. Moeller RegionalOperations, December 2,1986, Chicago,IL(Boehnert). The Subcom-mittee will begin its review of activities which are under the control of

( the NRC Regional Offices. This meeting will focus on the activities of the Recion !!! Office. Lodging will be announced later. Attendance by the following is anticipated:

Dr. Remick Mr. Ward Mr. Michelson Mr. Wylie l Mr. Reed (tent.)

SafetyResearchProgram(Closed), December 10, 1986, 1717 H Street, NW, Washington,0C (Duratswamy), 8:30 A.M., Room 1046. The Subcomittee will discuss the following and gather information for use by the ACRS in its l preparation of the annual report to the Congress on the NRC Safety Research Program and budget for FY 1988: (1)proposedNRCSafetyResearchProgram and budget for FY 1988 (2) areliminary OMB Mark and the impact of the  !

OMB-proposed reductions on tie continuing and proposed research contracts. .

and(3)RESresponsestoACRSrecommendationscontainedinitsJune11, 1986 report to the Commission. Lodging will be announced later. Atten-dance by the following is anticipated:

Dr. Siess Dr. Okrent r Dr. Carbon Dr. Remick (

Dr. Kerr Dr. Shewmun Dr. Mark Mr. Ward Mr. Micholson Mr. Wylie [

i O Dr. Moeller (p.m. only) I A -lO i

r

(')

NJ 320th ACRS Meeting, December 11-13, 1986, Washington. DC, Room 1046.

Decay Heat Removal Systems, December 17, 1986, 1717 H Street, NW, Washington, DC (Boehnert), 8:30 A.M., Room 1046. The Subcomittee will Lodging continue its review of the NRR Resolution A.ttendance Position by the following for USI A-45.

is anticipated:

will be announced later.

Mr. Wylie Mr. Ward Dr. Catton Mr. Ebersole Mr. Davis Dr. Kerr Mr. Michelson Instrumentation and Control Systems _, December 18. 1986, 1717 H Street, NW, Washington, DC (El-Zeftawy), 8:30 A.M., Roon 1046. The Subcomittee will discuss the effect of adverse conditions such as high temperature Attendance by theon solid-state components in nuclear power plants.

following is anticipated:

Dr. Moeller Mr. Ebersole Mr. Wylie Dr. Kerr Mr. Michelson SafetyResearchprogram(closed), January 7, 1987, 1717 H Street, NW, Washington,_0C (Duratswamy), 8:30 A.M., Room 1046. The Subcomittee will discuss the following and gather updated information for use by the ACRS in its presaration of the annual report to the Cong'ess on the NRC Safety (1 Researc1 Program and Budget for FY 1980: 10,1986;(2)

NRC Safety Research Program and Budget since the December final OMB mark and the impact of the OMA-proposed reductions on the con-tinuing and proposed research contracts; and (3) Draf t 0 of the ACRS report to the Congress. Lodging will be announced later. Attendance by the following is anticipated:

Dr. Okrent Dr. Siess Dr. Remick Dr. Carbon Dr. Shewmon Dr. Kerr Mr. Ward Dr. Mark Mr. Wylie Mr. Michelson Dr. Moeller 321st ACR$ Meeting, January 8 10, 1987, Washington. DC, Room 1046.

Regional Operations January 20, 1987, 1717 H Street, NW, Washington, DC Illochnert), 0:30 ATM,, Room 1046. The Subcomittee will continue its review of the activities of the Office of Inspection and Enforcement.

Attendance by the following is anti-Lodging will be announced later.

cipated:

Mr. Reed Dr. Remick Mr. Ward Mr. Michelson Mr. Wylie Dr. Moeller

/ -Il

O O p-,

4 v

Standardization of Nuclear Facilities, January 21, 1987, 1717 H Street, NW, The Subcommittee will Washington, DC, (Alderman), 8:30 A.W , Room 1046.

review NRC evaluation of Charter I ("Overall Requirements") of the EPRI Advanced Light Water Reactor Program. Lodging will be announced later.

Attendance by the following is anticipated:

Mr. Michelson Mr. Wylie Mr. Reed Mr. Ebersole Albuquercue,NM(Igne).

Structural Enoineering, January 21 and 22, 1987, The Subcomittee will review containment integrity and Category I struc-tures, programs, and test facilities. Lodging will be announced later.

Attendance by the following is anticipated:

Dr. Okrent Dr. Siess Dr. Shewmen Dr. Carbon Mr. Bender Mr. Ebersole Dr. Pickel Dr. Kerr Seabrook/ Occupational and Environmental Protection Systems / Severe Accidents (Closedl, Date to be determined (mid-November) Washington,_DC"(Major /

/o ') Igne/ Houston). The Subcommittees will continue the review of Public

\_/ Service of New Hampshire'sThe updated probabilistic risk asse NRC Staff is expected to the emergency planning zone for the site.

complete their review and formulate a position on this case by the end of October 1986.

Attendance by the following is anticipated:

Dr. Remick Dr. Kerr Dr. Corradini Dr. Mark Dr. Moeller AC/DC Power Systems Reliability, Date to be detemined (December),

The Subcomittee will review the proposed Washington, DC (El-Zeftawy). Attendance by the following is anticipated:

5tation Blackout rule.

Dr. Lewis

• Dr. Kerr Mr. Wylie Mr. Ebersole Eharleston,SCor(Asii}ington.OC(Bochnert). The Subcommittee wi Naval Reactnes Attendance by the follow-thThavalleactnr Koored Training Ship Project.

ing is anticipated:

Dr. Remick Dr. Kerr Mr. Ward Dr. Lewis

[v)

' A-/ L

h

• 1,

[ Metal Components, Date to be determined (January), Washington, DC (Igne).

The Subcomittee will review

(1) hear a status report of the Whipjet program (application of broad scope GDC-4 criteria) as applied to lead l plant Beaver Valley Unit 2; and (2) review public coments on NUREG-0313,

Revision 2(longrangefixforBllR-IGSCCproblems)perACRSletter.

1 Attendance by the following is anticipated:

) Dr. Shewmon Mr. Ward Mr. Etherington Mr. Bender

! Dr. Lewis Mr. Rodabaugh i Mr. Michelson i

1 AdvancedReactorDesigns,Datetobedetermined(January),Washintton,DC i (El-Zeftavy). The Subcomittee will review DOE advanced non-LWR cesigns

! regarding the use of proven technology and standardization. Attendance

by the following is anticipated

1 l Dr. Carbon Dr. Kerr 1

Dr. Mark Dr. Siess i _Seabrook Unit 1, Date to be determined (fall / winter), Washington, DC (Major). The Subcomittee will review the application for a full power l operating license for Seabrook Unit 1. Attendance by the following is anticipated:

i Dr. Kerr Dr. Moeller j Dr. Lewis Mr. titchelson I

l I

i i

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O O O

~

SEABROOK STATION PROBABILISTIC SAFETY ASSESSMENT AND EMERGENCY PLANNING STUDY .

A 1

-+

E2 0;P Presentation to @M ADVISORY' COMMITTEE ON REACTOR SAFEGUARDS $$ ,

Washington, D.C. 83 3 October 10,1986 8 ", E

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) Agenda '

i e Pro.ect Overview

! . O SALP

e Low Power Testing .

l4 9 Emergency Planning ,

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1 Project Overview 1

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l e Unit 2 24.1% Complete 7

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O SALP FUNCTIONAL AREA RATING (1/1/85 - 3/31/86)

Construction 1 reoperational Testing 1 ire Protection And Housekeeping 1 Operational Readiness 1 Emergency Preparedness 2 Assurance Of Quality 1 Licensing 1

&/f 4

u l

O Low Power Testing e No Fuel Load Open items With NRR Or Region 1 e ASLB (Low Power License)

- Open items O Equipment Qualification Time Duration O Detailed Control Room Design Review- SPDS O Emergency Action Levels Hearing Held 9/29 - 10/3 -

In Portsmouth, N.H.

l

. - Petition 50.57 (c) To Load Fuel And Conduct Precriticality Testing Submitted 8/22/86 l

O A-19 l- .

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l l

lI i

Emergency Planning i G Graded Exercise Held 3/86 (N.H. Only)

O e Risk Management And Emergency Planning Study Submitted To NRC i

i -7/21/86 l e Revision 2 Of N.H. Plan Submitted 9/86 l e Gov.Of Mass. Announced RefusalTo

! Submit Plans To FEMA- 9/20/86 i

l  :

1 I

lO i A -AO 1

j. i

O O O PRESENTATION OUTLINE e SSPSA (1983) OVERVIEW

• RMEPS (1985) OVERVIEW e WASH-1400 METHODOLOGY SENSITIVITY STUDY 1  !

h e TECHNICAL RESULTS e TECHNICAL BASES FOR RESULTS  :

e CONCLUSIONS i

O O O 1

SSPSA RISK MODEL DEVELOPMENT .-

l 1

l l e SSPSA COMPLETED DECEMBER 1983 (PLG-0300) e TECHNICAL SPECIFICATION AUGUST 1985 (PLG-0431) '

UPDATE i

l e RMEPS UPDATE DECEMBER 1985 (PLG-0432) '

i kN e SENSITIVITY STUDY UPDATE APRIL 1986 (PLG-0465) f e FRAGILITY UPDATE JUNE 1986 (SMA-12911.01) e SSPSA UPDATE IN PROGRESS e NRC REVIEW l - PLANT MODEL REVIEW - LAWRENCE LIVERMORE (1985) l PSNH RESPONSE TO PLANT MODEL REVIEW (MAY 1986) i REVIEW OF CONTAINMENT RESPONSE, SOURCE TERMS, AND CONSEQUENCES - BROOKHAVEN (BNL), NUREG/CR-4540 l

(FEBRUARY 1986)

O O O SSPSA SCOPE AND COVERAGE OF ACCIDENT SEQUENCES

e COMPREHENSIVE COVERAGE OF ACCIDENT SEQUENCES

! - 58 DISTINCT INITIATING EVENT CATEGORIES l - 39 PLANT DAMAGE STATES (" BINS")

- 14 RELEASE CATEGORIES

] - 16 MODULARIZED EVENT TREES j

l e FULL TREATMENT OF DEPENDENT EVENTS

- COMMON CAUSE FAILURES (SYSTEM ~ LEVEL)

I - EXTERNAL EVENTS

> - INTERNAL PLANT HAZARDS j

4 0

g

- EXPLICIT MODELING OF FUNCTIONAL DEPENDENCIES e PLANT-SPECIFIC AND ENHANCED CONTAINMENT MODEL ,

( - ASSESSMENT OF CONTAINMENT FAILURE MODES

- QUANTIFICATION OF SOURCE TERM UNCERTAINTIES ,,

- ENHANCED METHODOLOGY e SITE-SPECIFIC CONSEQUENCE MODEL

- MULTIPUFF RELEASE TREATMENT

- ACTUAL SITE CHARACTERISTICS

- QUANTIFICATION OF UNCERTAINTY

e O O PRINCIPAL RISK CONTRIBUTORS O

IN THE SSPSA (1983) .

Accident Containment Response - Group Group Fraction of Sequence Group Contributing Contribution Frequency Total Core Damage Initiating Events Percent (mean values) Frequency Group I Early Containment Failure 2.4 x 10-6 per .01 Early liealth - Interfacing LOCA 76 Reactor Year Effects - Seismic 24 or Once in TOU 410,000 Reactor Years Group II Delayed Containment Failure 1.7 x 10-4 per .73 Latent Health - Loss of Offsite Power 40 Reactor Year E f fects - Transients 19 'or Once in

- Fi res 15 6,000 Reactor

- Seismic 15 Years

- Others 11

. TOV

)J 4

Group III Containment Intact No Health - Transients 57 6.0 x 10-5 per .26 E f fects - SLOCA 29 Reactor Year

- Others 14. or Once in

'TUU 17,000 Reactor Years Total 2.3 x 10-4 per 1.00 Reactor Year or Once in 4,300 Reactor Years

1 . .

l SSPSA CONCLUSIONS (1983) 1 l e EARLY HEALTH RISK

{ - NRC SAFETY GOAL MET WITH LARGE MARGINS l - INTERFACING LOCA DOMINATES

! e LATENT HEALTH RISK i

- NRC SAFETY GOAL MET WITH VERY LARGE MARGINS hi

- SUPPORT SYSTEM FAILURES DOMINATE

! )J l e CORE MELT FREQUENCY 2 x 10-4/ REACTOR YEAR i

e CONTAINMENT EFFECTIVENESS i

- PRIMARY CONTAINMENT VERY STRONG

- EARLY FAILURE UNLIKELY

- LONG TIME FOR OVERPRESSURE

['i r ; '." .,,- -

O l Unique Features e Ultimate Containment Strength e Secondary Containment Enclosure Building e Two Service Water Systems -

e Basaltic Concrete 1 9 Steam Generator Secondary Water Inventory 9 RH R Vaults O

O A -Rb

j. .

i O O O l

RMEPS OBJECTIVES e REEXAMINE TECHNICAL BASIS OF THE 10-MILE EPZ j (NUREG-0396) ON A PLANT-SPECIFIC BASIS l e DEVELOP AN ENHANCED PRA METHODOLOGY FOR j ESTABLISHING A PLANT AND SITE-SPECIFIC EPZ I e APPLY THIS METHODOLOGY TO SEABROOK STATION i

l 1 - UPDATE SSPSA RISK MODEL (1983 - 1985) l $ - DETERMINE RISK IMPACT OF EMERGENCY PLAN OPTIONS l e ADDRESS UNCERTAINTIES AND SENSITIVITIES l

l e PROVIDE DOCUMENTATION AND PEER REVIEW l

i -

! O O O

, ENHANCED METHODOLOGY FOR EPZ DETERMINATION i

j e DEVELOP NUREG-0396 RISK OF DOSE VERSUS DISTANCE

) -

CURVES BASED ON PLANT / SITE-SPECIFIC RISK MODEL i .

l e CHARACTERIZE TOTAL POTENTIAL FOR RISK REDUCTION l BY PROTECTIVE ACTION AS RISK WITH NO EVACUATION

! e QUANTIFY SPATIAL DISTRIBUTION OF NONEVACUATION RISK

! w e CALCULATE ACTU1L RISK REDUCTION FOR PROTECTIVE

! L ACTION STRATEGIES

• MILE EVACUATION MILE EVACUATION MILE EVACUATION MILE EVACUATION AND SHELTERING OUT TO 10 MILES e EVALUATE UNCERTAINTIES AND SENSITIVITIES e COMPARE RESULTS WITH ALL AVAILABLE RISK ACCEPTANCE CRITERIA

! O O O RISK ACCEPTANCE CRITERIA UTILIZED ,

e NUREG-0396 DOSE VERSUS DISTANCE CURVES FOR 1,5, 50, AND 200-REM WHOLE-BODY DOSES i

t e WASH-1400 RISK CURVES FOR EARLY FATALITIES AND LATENT CANCER FATALITIES (MEAN AND MEDIAN RESULTS) i l b L e NRC INDIVIDUAL AND SOCIETAL RISK SAFETY GOALS t l e SPATIAL DISTRIBUTION OF RESIDUAL RISK I

i i

l O O O RMEPS UPDATE OF SSPSA RISK MODEL j e UPDATED SSPSA PLANT MODEL

- ENHANCED V-SEQUENCE MODEL ENHANCED SElSA/IIC ANALYSIS

- CONTAINMENT RECOVERY MODEL

! - ENHANCED TREATMENT OF COMMON CAUSE FAILURES l

I l T e UPDATED SSPSA SOURCE TERMS a

l - EXISTING SSPSA SOURCE TERMS i

l INCORPORATED SOME ZION IDCOR RESULTS j - PERFORMED SEABROOK/ ZION DESIGN COMPARISON i

- DEVELOPED SOME SEABROOK RESULTS WITH MAAP i

j - REASSESSED UNCERTAINTIES

! - EXAMINED SENSITIVITIES

O g o.

g O

RMEPS RESULTS (1-MILE EVACUATION)

Accident Containment Response - Group Group Fraction of Sequence Group Contributing . Contribution Frequency Total Core Damage Initiating Events Pe rcent (mean values) Frequency Group I Early Containment Failure 3.5-7 per Reactor .001 Early Health -

Interfacing LOCA 13 Year or Once in Ef fects -

Seismic 87 2[114000 Reactor APC and TMLL >l Years 100 Group II Delayed Containment Failure 1.6-4 per Reactor .58 Latent Health - Loss of Offsite Power 10 Year or Once in E f fects - Transients 43 G;t)00 Reactor

- Fires 16 Years 1 - Seismic 17 i

s

- Others 14 TU6 Group III Containment Intact lio Health - LOSP 46 1.2-4 per Reactor 42 Effects - Transients 27 Year or Once in

- SLOCA 15 I W Reactor Others 12 Years TU0 fl0TE: Exponential notation is indicated Total 2.8-4 per Reactor 1.00 in abbreviated form; Year or Once in i.e., 4.4-8 = 4.4 x 10-8, '

4p300 Reactor Years

i O O -O RMEPS KEY RESULTS e EARLY HEALTH RISK WITH NO EVACUATION IS:  ;

- LESS THAN WASH-1400 WITH 25-MILE EVACUATION i

j - MEETS NRC SAFETY GOAL WITH WIDE MARGIN l - CONFINED TO AREA CLOSE TO THE SITE  ;

I  %

l d, o VERY SMALL RISK REDUCTION BY ANY EVACUATION i i

t' e ALL NUREG-0396 DOSE VERSUS DISTANCE CRITERIA j SATISFIED AT 1 MILE OR LESS i e LATENT HEALTH RISK INSENSITIVE TO ASSUMPTIONS j REGARDING EVACUATION i

I

0 -

0 0 EMERGENCY PLANNING SENSITIVITY STUDY METHODOLOGY .

• PURPOSE: DETERMINE IMPORTANCE OF SOURCE TERMS VERSUS PLANT-SPECIFIC FEATURES AND ENHANCED PRA TECHNOLOGY .

• APPROACH: RMEPS CALCULATIONS REDONE USING:

i Os W SH-1400 SOURCE TERM METHODOLOGY

- BEST ESTIMATE ASSUMPTIONS ON ALL OTHER UNCERTAIN RISK PARAMETERS i

MEDIAN RISK OF dc EARLY HEALTH EFFECTS FOR DIFFERENT EVACUATION DISTANCES 10-3

_4 _

LEGEND

--- =^""e4 R K^o "

WASH-1400 SOURCE

$ TERM METHODOLOGY g (MEDIAN RESULTS)

WASH-1400 d

o 10-5 -

(MEDIAN RESULTS) li

~ ' '

O I iMEDIATE PROTECTIVE O $ 10-6 -

N N

[ EVA ATION \

w N N

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E 10-7 - \ {

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10-8 -

2 - MILE \

f EVACUATION \ {

RMEPS RESULTS N \

OFF SCALE N \

10-9 f I I ' I \ I 0 2 3 4 10 5 1

10 10 10 10 10 EARLY FATALITIES O '

A-3%

THE BENEFITS OF RISK REDUCTION BY O EVACUATIO.N OR SHELTERING ARE:

l e VERY SMALL DUE TO VERY LOW INHERENT PLANT RISK e FULLY REALIZED BY CLOSE-IN EVACUATION e NOT NEEDED TO MEET NRC SAFETY GOALS 10*2 a g a g ND CU Rt AUL IPLIED 30-3 _ S TE B U DA _

w E e cc 5 10 4 - -

EJ i

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2  ; ~

10-5 4 s s O u F

- ~

10'0 l

UAT! f4RhiTH SHELTERING TO 10 MILES X

10~7 0 2 4 6 8 to O- EVACUATION DISTANCE (MILES) h 'b

OCOMPARISON WITH NUk-0396 200-REM AND O 50-REM WHOLE BODY DOSE PLOTS FOR NO IMMEDIATE PROTECTIVE ACTIONS

i. . . . . . . . . , . . . . . . . . , , . . . . . . . .

~ ~

~ ~

NUR EG-039G

$$ _ ----- SENSITIVITY o$

a ~

STUDY yh . .. ... . .. . . . .. .. R M E PS R ESU LTS

< _ (200 REM CURVE OFF _

@D co SCALE)

O us sg oi r 8% -  :

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85 I J c.. i

_\ l j r.1 200 REM

, kl 50 REM g ogi [i ,, ,,it i i i e i i iil i ie i iivi 1 10 100 1,000 DIS,TANCE (MILES)

O O O.

COMPARISON OF INDIVIDUAL RISK WITH  ;

BACKGROUND AND SAFETY GOAL VALUES l 10-2 Y

O BACKGROUND ACCIDENTAL FATALITY RISK 10 (5 FATALITIES PER 10.000 POPULATION PER YEAR)

=

h b 104 -

t 4

h B _

g $ 10-5 W m 2

SAFETY GOAL (.001 TIMES Q Q BACKGROUND RISK) 8 THIS STUDY FOR 4 10-6 - SEABROOK STATION 0

WITH NO IMMEDIATE 3 PROTECTIVE ACTIONS 5

k 10-7 o"

fff j WITH 1 MILE l

i g EVACUATION Y RMEPS RESULTS g 10'8 -

WITH NO IMMEDIATE '

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PROTECTIVE ACTIONS 4 \

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t l

i i O O O PEER REVIEW GROUP t

e ROBERT BUDNITZ, CHAIRMAN, FUTURE RESOURCES ASSOCIATES, INC.

e DAVID ALDRICH, SCIENCE APPLICATIONS INCORPORATED e JOSEPH HENDRIE, CONSULTANT y e NORMAN RASMUSSEN, MASSACHUSETTS INSTITUTE OF .

w TECHNOLOGY

= t l e ROBERT RITZMAN, ELECTRIC POWER RESEARCH INSTITUTE l e WILLIAM STRATTON, CONSULTANT e RICHARD WILSON, HARVARD UNIVERSITY t

_ . - - - _I

i  ;

! O O O PEER REVIEW FINDINGS i

e CONCURRED WITH PRINCIPAL STUDY FINDINGS -

- OVERALL OFFSITE RISKS VERY SMALL  ;

~

, - EARLY HEALTH RISK LOWER THAN THOUGHT TO EXIST l WHEN GENERAL EPZ ESTABLISHED I

l

- EARLY HEALTH RISK CONFINED TO AREAS VERY CLOSE TO j REACTOR e CONCLUSION ROBUST EVEN IN LIGHT OF UNCERTAINTIES e BELIEVE THE "BEST ESTIMATE" PROBABLY OVER-ESTIMATES ACTUAL CONSEQUENCES -

e SEABROOK CONTAINMENT MAJOR FACTOR 1

4

O .O O FAVORABLE RESULTS DUE TO

• CONTAINMENT EFFECTIVENESS i

1 l

• ENHANCED V-SEQUENCE MODEL i

! I e SOURCE TERMS i

4 i

I i

! O Safety soy Results: O CONTAINMENT EFFECTIVENESS '

j i

(Percent of Core Damage Frequency) '

i I

i  !

l 66% 99 % 99.9 %  !

i i

34 % 1% .1 %

3 WASH-1400 (1975) SEABROOK STATION (1983) SEABROOK STATION IMPROVED LOCA OUISIDE t CONTAINMENT MODEL (1985) .

i

~ EARLY DEGRADED CONTAINMENT CONTAINMENT 0R i FAILURE CONTAINMENTINTACT l

i

\

O O O 1

CONTAINMENT FAILURE TYPES 1  !

i j A. SMALL LEAK (0.02 SQUARE INCHES TO l j 6 SQUARE INCHES) j PRESSURE RISE CONTINUES i i

B. LOCAL FAILURE (6 SQUARE INCHES TO b 60 SQUARE INCHES)_

i PRESSURE RISE STOPS

?3 EXTENDED RELEASE (>1 HOUR)  !

C. GROSS FAILURE (>60 SQUARE INCHES)

RAPID CONTAINMENT BLOWDOWN (<1 HOUR) i l

l l

I

O O O LOCAL CONTAINMENT FAILURE MODES CONSIDERED e FLU!D SYSTEMS PENETRATION e HIGH ENERGY PENETRATION I i

e FUEL TRANSFER TUBE e ELECTRICAL PENETRATION

> e PURGE LINE PENETRATION  ;

g i

• PURGE VALVE SEALS '

e EQUIPMENT HATCH e PERSONNEL LOCK  !

e OTHER PENETRATIONS l

e LINER TEARING e WELD IMPERFECTIONS l I

rm CONTAINMENT FAILURE MODES AND TYPE Median Lognormal Fallure Median Failure Leak Fallure Standard Mode Pressure A

Type Deviation (psfa) 8 Structural Failure Modes Cylinder Wall Hoon 231 Larges C .12 Dome Hoop or Merf dfonal 238 Larges C .12 Wall Merfdfonal 296 Largea C .12 Base Slab Shear 338 Largea C .23 Base Slab Flexure 415 Largea C .25 Wall Shear at Base 423 Largea C .30 g Lncel Failure Modes Feedwater Penetration 194 Self-Regulatingh 8 0.5 Flue Head Feedwater Pf pe Crushing 231 Self-Regulatingb 3 .12 Fuel Transfer Tube > 260C Sel f-Regulatingb 8 d Bellows Penetrations X-25, X-26, 181 0.5 Square Inch A 0.16 X-27 Each All Otherse > 231C Sel f-Regulatingh 8 d aMuch larger than 0.5 square foot, b

leak area is self-adjusting to stop pressure rise.

C Probability of failure is less than 50% at ultimate wall hoop capacity.

drailure pressure model not lognormal.

e composite estimate of liner adhesion, microcracks, weld faults, equionent hatch, other mechanical penetrations, and electrical penetrations.

O COMPOSITE CONTAINMENT FAILURE PROBABILITY DISTRIBUTIONS FOR TYPE B (LEAK) FAILURE, TYPE C (GROSS)

FAILURE, AND TOTAL FAILURE i_ ,,............--

ggygua$ - -

a;w -

. /

y /, TYPE B (LE AK)

. / / FAILURE I

/

/ /

/ WET SEQUENCES

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14 16o 12 200 22. 24 26. o PR E SSU R E (PSIAI A-W ,

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! O O O i ENHANCED TREATMENT OF INTERFACING ~

4 SYSTEMS LOCA t

I e MORE COMPLETE MODELING OF VALVE FAILURE MODES e NEW DATA ON CHECK VALVE FAILURES VERSUS LEAK SIZE

! e MORE REALISTIC TREATMENT OF DYNAMIC PRESSURE PULSE

{

e EXPLICIT MODELING OF RHR RELIEF VALVES D e QUANTIFICATION OF RHR PIPING FRAGILITIES TO

$ OVERPRESSURE e MODELING OF RHR PUMP SEAL LEAKAGE e OPERATOR ACTIONS TO PREVENT MELT CONSIDERED  !

e THERMAL HYDRAULIC AND SOURCE TERM FACTORS MODELED USING MAAP e UNCERTAINTIES QUANTIFIED 1

O O O  !

INTERFACING SYSTEMS LOCA KEY RESULTS t

FREQUENCY (PER REACTOR YEAR) .

UPDATED EVENT SSPSA ANALYSIS k+ VALVE RUPTURES, LOCA 1.8 x 10-s 7.8 x 10-s J VALVE RUPTURES, LOCA, 1.8 x 10-8 3.1 x 10-7 CONTAINMENT BYPASS VALVE RUPTURES, LOCA, 1.8 x 10-s 4.1 x 10-8 CONTAINMENT BYPASS, l MELT l

l

~

C) SOURCE TERMS FOR SEABROOK STATION c ,

}  ; i ALL ACCIDENT SEQUENCES ,

IDENTIFIED IN SEABROOK PRA v .' -

',A '

13 RELEASE CATEGORIES TO >

COVER ALL SEQUENCES IN SSPSA #' ,

l_) ^

^

r 6 RELEASE CATEGORIES WITH SIGNIFICANT -

RISK CONTRIBUTION IN SSPSA -

^

.i- ,. ! ,

~

12 SOURCE TE S FOR RMEPS

- 6 BEST ESTIMATE '

- 6 CONSERVATIVE ESTIMATE '

s

/

2 SOURCE TERMS SOUP.CE TERMS DEVELO?ED FROM l

FROM IDCOR (MAAP) SSPSA USING WASH 1400 (CO.9RAL)

METHODOLOGY AND CGRRECTICNS FOH IDCOR AND NRC SOURCE TEHM v _

ESEARCH p#j 2 SOURCE TERMS CALCULATED FOR SEABROOK l

p USING IDCOR (MAAP)

Q METHODOLOGY g

9

- ---.---n, , - - -_n. - . _ - - - . . - , - . _ _ _ _ , . -.n---,.

9 .- 2 x 1-3 .

3-.

SEABROOK STATIONiSOURCEJERMS

[

CONTAINMENT SOUlicE kEF$1 DEVELOPMENi

[~ x ACCIDENT RELEASE CATEGORY FAILURE TYPE RMFPS. ^ ' USITIVITY A B C SSPSA . BEST ESTIMATE CONSERVATIVE STUDY I S1 EARLY GROSS FAILURE VMT RSS- MODIFIED RSS RSS METHODOLOGY SSPSA METHODOLOGY METHODOLGY S2 EARLY INCREASED LEAKAGE VMT (LOP) LOP RSS SSPSA SSPSA RSS i WITH LATE OVERPRESSURE METHODOLOGY METHODOLOGY

+ UNCERTAINTY S3 LATE OVERPRESSURE OR LATE (LOP) LOP RSS IDCOR SSPSA RSS b BASEMAT MELT-THROUGH METHODOLOGY (ZION-MAAP) METHODOLOGY t + UNCERTAINTY 4

S5 CONTAINMENT INTEGRITY - - -

RSS RSS RSS RSS MAINTAINED METHODOLOGY METHODOLOGY METHODOLOGY METHODOLOGY

+ ENCLOSURE NO ENCLOSURE l SS CONTAINMENT PURGE -

At -

RSS IDCOR SSPSA RSS l ISOLATION FAILURE t=0 METHODOLOGY (ZION-MAAP) METHODOLOGY

+ UNCERTAINTY S7 CONTAINMENT BYPASS BYP - -

ASSIGNED SEABROOK-MAAP SEABROOK-MAAP RSS VIA RHR PUMP SEALS RMEPS TO S6 + POOL NO POOL METHODOLOGY

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t O O O I

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e EARLY HEALTH RISK VERY LOW EVEN WITHOUT l IMMEDIATE PROTECTIVE ACTIONS i

! e BENEFITS OF EVACUATION VERY SMALL AND CONFINED

! TO AREA CLOSE IN TO SITE i i l D e PEER REVIEW GROUP CONCURS WITH RMEPS AND l SENSITIVITY STUDY CONCLUSIONS l

e SEABROOK RISK MANAGEMENT ACTIVITIES ARE l

i

' CONTINUING l .

_ ~;'.<

NRR STAFF PRESENTATION TO THE O ACRS APPENDIX V INTRODUCTION TO STAFF PRESENTATION ON CLINTON POWER STATION

SUBJECT:

11 ARCH 9,1982 ACRS REPORT ON CLINTON POWER STATION THAT IDENTIFIED TWO PLANT SPECIFIC ISSUES ON WHICH COMMITTEE REQUESTED TO BE KEPT INFORMED.

DATE: OCTOBER 9,1986 PRESENTER: BYRON L. SIEGEL PRESENT5R'S TITLE / BRANCH /DIV: LICENSING PROJECT MANAGER PROJECT DIRECTORATE NO. 4 DIVISION OF B0ILING WATER REACTORS PRESENTER'S NRC TEL. NO.: (301) 492-9474 .

SUBCOMMITTEE:

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( )THEMARCH9,1982ACRSREPORTONCLINTONPOWERSTATION, UNIT 1 IDENTIFIED THE'FOLLOWING PLANT SPECIFIC' ISSUES ON WHICH THE COMMITTEE REQUESTED TO BE KEPT INFORMED:

4 QUALITY ASSURANCE AND QUALITY CONTROL ORGANIZATION (SSER 6, APPENDIX M - QUALITY ASSURANCE PROGRAM)

SEISMIC CAPABILITY OF EMERGENCY AC POWER SUPPLIES, DC POWER SUPPLIES, AND SMALL COMPONENTS SUCH AS ACTUATORS AND INSTRUMENT LINES PART OF DECAY HEAT REMOVAL SYSTEM (SSER 6, SECTION 22, ITEM 3 - SEISMIC ASSESSMENT PROGRAM FOR SAFETY RELATED MECHANICAL AND ELECTRICAL EQUIPMENT)

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r STATUS CLINTON UNIT 1 STATUS

SUMMARY

LOW POWER LICENSE ISSUED SEPTEMBER 29, 1986 ALL CONFIRMATORY AND OUTSTANDING ISSUES RESOLVED AT ISSUANCE OF LOW POWER LICENSE l.ICENSEE LOADING CORE APPR0XIMATELY HALF THE FUEL BUNDLES HAVE BEEN LOADED O* PARTIAL CORE SHUTDOWN MARGIN TEST PERFORMED AFTER 144 BUNDLES WERE LOADED INTO CORE IN SPIRAL LOADING PATTERN INITIAL CRITICALITY SCHEDULED FOR NOVEMBER 20, 1986 EARLIEST DATE FOR FULL POWER LICENSE: MID-DECEMBER, 1986 O s-ss

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ILLINOIS POWER COMPANY NUCLEAR PROGRAM L

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, OCTOBER 9,1986 3

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O BAcxaReuNo - et&@N P@WER STATIOC 950 MWE GE, BWR-6, MK III CONTAINMENT

• INDEPENDENT DESIGN REVIEW REPORT - JANUARY 1985

• ACTIVATED NUCLEAR REVIEW AND AUDIT GROUP - JANUARi 1985

• ASLB DISESTABLISHED - FEBRUARY 1985

• COMPLETED REACTOR VESSEL COLD HYDRO - FEBRUARY 1985

• CONSTRUCTION APPRAISAL TEAM REVIEW COMPLETE - JUNE 1985 4

4

• COMPLETED REACTOR PLANT HOT OPERATIONS - JULY 1985

• SALP REPORT - NOVEMBER 1985

• ASME N-5 CERTIFICATION PROGRAM COMPLETE - APRIL 1986

• OPERATING LICENSE - SEPTEMBER 1986 10/86

O seismic issQ STATUS O i

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• ACRS (1982) - SPECIFIC ATTENTION TO DECAY HEAT

REMOVAL SYSTEM i

EMERGENCY AC POWER SUPPLIES DC POWER SUPPLIES SMALL COMPONENTS INSTRUMENT LINES I

, y

• CPS DECAY HEAT REMOVAL SYSTEM

~

l RESIDUAL HEAT REMOVAL l -

SHUTDOWN SERVICE WATER ASSOCIATED AC/DC POWER SYSTEMS -

• EQUIPMENT SEISMIC ASSESSMENT PROGRAM .

DEVELOPED IN RESPONSE TO ACRS REQUEST

~

10/86

t O EQUIPMENT SEISMIPASSESSMENT n

PROGRLsd l

• THREE PHASE PROGRAM i

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• SMALL BORE PIPING / INSTRUMENT LINE DESIGN ADEQUACY

- A/E DESIGNED PIPING IN RHR, SHUTDOWN SERVICE WATER, DIESEL GENERATOR FUEL OIL AND AIR START

- VERIFIED ADEQUACY OF DESIGN AND CALCULATIONS l

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• SEISMIC INTERACTION WALKDOWNS/ ANALYSIS 4 . FIELD WALKDOWNS N - INTERACTIONS OF PIPING, VALVES,' CONDUlT, TRAYS, ETC ll l

- ADEQUACY OF RESTRAINTS

{ - EQUIPMENT SUSCEPTIBLE TO DAMAGE

!

• EQUIPMENT EVALUATION

- VENDOR SUPPLIED COMPONENTS

- DEMONSTRATE ABILITY TO SURVIVE EARTHQUAKE (SITE SPECIFIC SPECTRA)

- EVALUATE EQUIPMENT STRESS LEVELS i

10/86 i

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• EQUIPMENT SEISMIC ASSESSMENT PROGRAM REPORT

! TO NRC - OCTOBER 1985 l

• SSER #6 - ACRS CONCERNS SATISFACTORILY ADDRESSED

>

• CONFIDENT THAT SEISMIC DESIGN AND INSTALLATION OF h SYSTEMS REVIEWED ARE ADEQUATE 1

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APPENDIX VII QA/QC, 0VERIflSPECTI0tl AND

- AUDIT PRO ~. RAMS gR STAFF PRESENTATION TO THE ACRS

SUBJECT:

QA/QC, 0VERINSPECTION AND AUDIT PROGRAMS EATE: OCTOBER 9, 1986

'PM ER: F. J. JABLONSKI PRESENTER'S TITLE / BRANCH /DIV: CHIEF, QUALITY ASSURANCE PROGRAfiS SECT 10i1 ENGINEERING BRANCH DIVISION OF REACTOR SAFETY REGION 111 FRESENTER'S NRC TEL NO.: FTS 3CS-5555

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• 1982 - MAJOR QA CHANGES

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• LOCATION 1973/1986 - AUDITS

• LICENSEE 500 A

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• THIRD PARTY

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• NRC INSPECTIONS - TEST & STARTUP

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• OVERINSPECTION COPPLETED AP0 ACCEPTABLE i

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1. APPENDIX VIII

, CLINTON POWER STATION EQUIPMENT SEISMIC ASSESSMENT PROGRAll NRR STAFF PRESENTATION TO THE O ACRS

SUBJECT:

CLINTON P0'.'ER STATION EQUIPMENT SEISlilC ASSESS!iENT PROGRA". .

DATE: OCTOBER 9,198G P ENTER: 03. ARNDLD LEE ,.

PRESENT$R'S TITLE / BRANCH /DIV: ItECHANICAL Et'GINEER EllGINEERItiG BPNICH DIVISI0'l 0F BUR LICE!! SING PRESENTER'S NRC TEL. NO.: 492-7305 /

SUBCOMMITTEE:

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/  ; EQUIPMENT SEISMIC ASSESSMENT THE EQUIPMENT SEISMIC ASSESSMENT PROGRAM (ESAP) WAS DIVIDED INTO THREE PHASES:

PHASE I EXAMINED THE ADE0VACY OF THE DESIGN METHODS USED FOR SMALL BORE PIPING PHASE II EXAMINED THE AS-BUILT EQUIPMENT CONFIGURATIONS OF THE DECAY HEAT REMOVAL AND EMERGENCY POWER SUPPLY SYSTEMS FOR SEISMIC CONCERNS pPHASEIIIEVALUATEDTHEABILITYOFEQUIPMENTINTHESESYSTEMS V

TO WITHSTAND AN EARTHOUAKE OF THE FORM PREDICTED BY THE REVISED RESPONSE SPECTRA, WHICH WERE DEVELOPED l

USING THE ELASTIC-HALF-SPACE APPROACH FOR S0IL-STRUCTURE INTERACTION ANALYSIS.

1 A-U l

I NRC STAFF'S APPROACH THE STAFF REVIEWED THE ESAP REPORT PROVIDED IN THE APPLICANT'S I f

LETTER OF OCTOBER 14, 1985.

THE STAFF ALSO REVIEWED THE SMALL BORE PIPING PROCEDURE, SELECTED !

DESIGN CALCULATIONS, AND THE DESIGN OF SMALL TAP LINES AND THEIR i QUALIFICATION METHOD,  !

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~

ON JANUARY 16, 1986, THE STAFF PERFORMED A PLANT SITE AUDIT ON U HE ESAP PHASE I PROGRAM RESULTS. THE STAFF ALSO PERFORMED A SITE WALKDOWN AUDIT OF SEVERAL AREAS REVIEWED BY THE LICENSEE'S NUCLEAR SAFETY ENGINEERING DEPARTMENT TEAM,  ;

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I l DETAILED ASSESSMENT OF EQUIPMENT SEISMIC CAPABILITY HAS BEEN l REVIEWED BY THE STAFF.

l l FOR MOST EQUIPMENT QUALIFIED BY TEST, THE TEST RESPONSE l RESPONSE SPECTRA SATISFACTORILY ENVELOPED THE REVISED RESPONSE SPECTRA.

FOR THOSE PIECES OF EQUIPMENT 00ALIFIED BY ANALYSIS, THE i

ACCELERATIONS DUE 10 THE REVISED RESPONSE SPECTRA WERE GENERALLY FOUND TO BE LESS THAN THE ACCELERATIONS FROM THE PROCUREMENT SPECIFICATION RESPONSE SPECTRA. IN SOME EXCEPTION CASES (REVISED RESPONSE SPECTRA EXCEEDED PROCURE-MENT RESPONSE SPECTRA ACCELERATIONS) THE EQUIPMENT STRESSES DUE TO THE REVISED RESPONSE SPECTRA WERE CALCULATED AND FOUND TO BE WITHIN MATERIAL ALLOWABLE LIMITS.

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LJ ) CONCLUSIONS THE STAFF HAS COMPLETED THE REVIEW 0F THE ASSESSMENT PROGRAM AND FOUND THE CONCERN EXPRESSED BY THE ACRS IN ITS LETTER OF MARCH 9, 1982, HAS BEEN ADDRESSED IN A MANNER SATISFACTORY TO THE STAFF, AS AN ADDED ASSURANCE, THE STAFF REVIEW 0F THE CLINTON SEISMIC QUALIFICATION PROGRAM FOR ALL SAFETY-RELATED EQUIPMENT HAS BEEN COMPLETED AND THE RESULTS DOCUMENTED IN SSER N0, 7, O

O ur

APPENDIX IX RECEllT SIGilIFICAtlT EVENTS AGENDA FOR ACRS MEETING

'~

Friday October 10, 1986 10:15 a.m.

Room 1046 H-Street Washington, D.C.

RECENT SIGNIFICANT EVENTS Presenter / Office Mephone Page INTRODUCTION - (NEW SELECTION PROCESS) J. Rosentha1/IE 24193 g

Date Plant Event

/ 107I/86 Oconee Loss of Low Pressure Service Water H. Bailey /IE 29006 h

/ 9/11/86 Hope Creek Loss of Offsite Power Test J. Wiggins/ REG I /O 488-1198 ELECTRICAL PROBLEMS

/26/86 Salem 2 Overloading of Station Power Transformers L. Bettenhausen/ REG I 488-1291

/p C^9/21/86 Zion 2 Degraded Voltage - ESF Bus Loss of Power J. Rosentha1/IE 24193

.1/

C8/15/86 Turkey Design Deficiency in g Point 3, 4 Emergency Power System 9/22/86 Duane Arnold Class 1E Station Battery E. Weiss /IE .2.f Problem 29005 C^No detailed presentation. Only a brief overview of these events and their relation to the Salem 2 event.

A-fo

EVENTS ANALYSIS AND SELECTION PROCESS (J. ROSENTHAL, IE) 680 IMMEDIATE NOTIFICATIONS (850.72's) PLUS DAILY REPORT ITEMS SCREENED IN 2 MONTHS EVENT FOLLOWUP (ADDITIONAL INFORMATION AND EVAL.UATION) BY IE/NRR ENGINEERS WHO REPORT BACK TO MORNING CONFERENCE CALL ON ABOUT A QUARTER OF THESE EVENTS

• BASED ON EVENT FOLLOWUP, 3 TO 5 PRESENTED TO NPP/IE MANAGEMENT EACH WEEK 40 CANDIDATE EVENTS PRESENTED TO J. EBERS0LE 12 JOINTLY SEl.ECTED FOR PRESENTATION TO J. EBEPsnl.E O* FINAL SELECTION FOR ACRS O h -71 l

/ \

N._/

LIST OF RECENTLY ISSUED IE 8ULLETINS Bulletin Date of No. Subject Issue issued to 86-03: Potential Falture Of Multiple --- All facilities holding ECCS PUMPS Due To Single Failure an OL or CP.

Of Air-0perated Valve in Minimum Flow Recirculation Line 86-02 Static "0" Ring Differential 7/18/86 All power reactor Pressure Switches facilities holding an OL or CP 86-01 Minimum Flow Logic Problems 5/23/86 All GE BWR facilities That Could Disable RHR Pumps holding an OL or CP o

O' A - 7.2-3

6 f LIST OF RECENTLY ISSUED i

( IE INFORMATION NOTICES 1986  !

Information Date of  :

Notice No. Subject Issuance Issued to i 86 64 Deficiencies In Upgrade 8/14/86 All power reactor  !

Programs For Plant Emergency facilities holding Operating Procedures an OL or CP 86-65 Malfunctions Of ITT Barton 8/14/86 All power reactor Model 580 Series Switches facilities holding During Requalification Testing an OL or CP 86-66 Potential For Failure Of 8/15/86 All power reactor Replacement AC Coils Supplied facilities holding By The Westinghouse Electric an OL or CP Corporation For Use In Class 1E Motor Starters and Contractors 86-67 Portable Motsture/ Density 8/15/86 All NRC licensees Gauges: Recent Incidents And authorized to possess, Connon Violations Of Require- use, transport, and ments For Use, Transportation, store sealed sources And Storage in portable gauges 86 68 Stuck Control Rod All BWR facilities holding an OL or CP 86-69 Spurious System Isolations 8/18/86 All GE BWR facilities Caused By The Panalarm Model holding an OL or CP 86 Thermocouple Monitor 86-70 Potential Failure Of All 8/18/86 All power reactor '

Emergency Diesel Generators , facilities holding an OL or CP ,

86-71 Recent Identified Problems 8/19/86 All power reactor i With Limitorque Motor facilities holding L Operators an OL or CP i

86-72 Failure 17-7 PH Stainless 8/19/86 All power reactor Steel $srings In Valcor facilities holding '

Valves Que To Hydrogen an OL or CP i Embrittlement '

86 73 Recent Emergency Olesel 8/20/86 All power reactor facilities holding Generator Problems an OL or CP d OL = Operating License CP = Construction Permit A - 7J I

(~N LIST OF RECENTLY ISSUED

) IE INFORMATION NOTICES 1986 Information Date of Notice No. Subject Issuance Issued to 86-74 Reduction Of Reactor Coolant 8/20/86 All 8WR facilities inventory Because Of Misalign- holding an OL or CP ment Of RHR Valves 86-75 Incorrect Maintenance 8/21/86 All power reactor Procedure On Traversing facilities holding Incore Probe Lines an OL or CP 86-76 Problems Noted In Control 8/28/86 All power reactor Room Emergency Ventilation facilities holding Systems an OL or CP 86-77 Computer Program Error Report 8/28/86 All power reactor Handling facilities holding an OL or CP and nuclear fuel manu-facturing facilities 86-78 Scram Solenoid Pilot Valve 9/2/86 All BWR facilities (SSPV) Rebuild Kit Problems holding an OL or CP 86-79 Degradation Or loss Of 9/2/86 All power reactor Charging Systems At PWR Nuclear facilities holding Power Plants Using Swing-Pump an OL or CP Designs 86 80 Unit Startup With Degraded 9/12/86 All power reactor High Pressure Safety Injection facilities holding System an OL or CP 86-81 Broken Inner External C1csure 9/15/86 All power reactor Springs On Atwood & Morrill facilities holding Main Steam Isolation Valves an OL or CP 86 82 Failures Of Scram Discharge 9/16/86 All power reactor Volume Vent And Drain Yalves facilities holding an OL or CP 86 83 Underground Pathways Into 9/19/86 All power reactor Protected Areas, Vital Areas, facilities holding Material Access Areas, And an OL or CP Controlled Access Areas O

b' 6L = Operating License CP = Construction Permit

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n OCONEE 2 - LOSS OF LOW PRESSURE SERVICE WATER (LPSW)

OCTOBER 1, 1986 - (H. BAILEY, IE) l l

PROBLEM: LPSW COOLING FAILED ON UNIT 2 l SIGNIFICANCE: LOSS OF ULTIMATE HEAT SINK. DESIGN DEFICIENCY MAY j BE APPLICABLE TO OTHER PLANTS l

l CIRCUMSTANCES:

LPSW PUMPS ON UNIT 2 LOST SUCTION AFTER ABOUT 1 HOUR OF OPERATION l FOLLOWING LOAD SHED TEST WITH CONDENSER CIRCULATING (CCW) WATER PUMPS IDLED.

SUCTION REGAINED AFTER CROSS CONNECT TO UNIT 1 WAS OPENED.

CCW GRAVITY FLOW TEST ON UNIT 2 FAILED. LPSW SUCTION FROM UNIT 1.

CCW USES SIPHON FOR FLOW OVER HIGH POINT, ULTIMATE HEAT SINK ON l UNITS 1AND2DECLAREDIjiOPERABLEPERT/S. UNITS SUBSEQUENTLY TAKEN TO COLD SHUTDOWN.

FLANGE LEAKS ON UNIT 2 CCW PUMPS SEALED. PASSED CCW GRAVITY FLOW TEST.

KE0 WEE LAKE LEVEL FOR TEST A CONCERN. CURRENT LEVEL 785 FT.

NORMAL LAKE LEVEL 800 FT. MINIMUM TECH SPEC LEVEL OF 775 FT.

FOR KE0 WEE HYDRO EMERGENCY POWER. CCW PUMP HAS A FLANGE AT HIGH POINT IN SYSTEM.

I FOLLOWUP:

REGION II HAS LEAD FOR RESTART I

• LICENSEE TO DEMONSTRATE OPERABILITY OF LPSW SYSTEM PRIOR TO RESTART n

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PROBLEM: A LOSS OF 0FFSITE POWER STARTUP TEST WAS CONDUCTED.ON SEPTEMBER 11, 1986 WITH 24 UNEXPECTED PROBLEMS IDENTIFIED. THIS TEST WAS ABORTED AFTER ABOUT FIVE MINUTES A SUBSEQUENT TEST PERFORMED ON SEPTEMBER 19, 1986 WITH THE PLANT SHUTDOWN RESULTED IN 17 ADDITIONAL PROBLEMS IDENTIFIED.

THE NUMBER OF PROBLEMS IDENTIFIED DURING THESE TWO TESTS RAISED THE LEVEL OF STAFF CONCERN OVER THE ADEQUACY OF THE FACILITY DESIGN, ,

CONSTRUCTION AND TESTING PROGRAMS.

CAUSES: 1. BAILEY SOLID STATE LOGIC FAILURES s 2. -REGULATORY GUIDE 1.97 CONFORMANCE (POWER SUPPLY SELECTION)

3. MINOR DESIGN AND EQUIPMENT PROBLEMS

4. PRE 0PERATIONAL TEST WEAKNESSES A go SLIDE 2 ll

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SAFETY SIGNIFICANCE:

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I l CONCERN REGARDING THE RELIABILITY OF THE BAILEY SSLM SYSTEM i

FOR CONTROL OF SAFETY-RELATED SYSTEMS I.

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CIRCUMSTANCES DATE EVENT

1. SEPT. 11, 1986 FIRST LOSS OF 0FFSITE POWER EVENT CONDUCTED FROM ABOUT 20% POWER.

TEST TERMINATED AFTER ABOUT 5 MINUTES. 24 TEST OBSERVATIONS MADE BY THE LICENSEE'S TESTING CREW.

2. SEPT. 19, 1986 CORRECTIVE ACTIONS TO FIRST LOP TEST FINDINGS COMPLETED AND A LOP TEST WHILE SHUTDOWN WAS RUN. ,

- 17 TEST OBSERVATIONS MADE

3. SEPT. 24, 1986 REGION I ISSUES CONFIRMATORY ACTION LETTER TO LICENSEE

4. SEPT. 24, 1986 REGION I AIT DISPATCHED TO ANALYZE THE RESULTS OF LOP TESTING AND TO ASSESS THE SAFETY SIGNIFICANCE OF THE PROBLEMS FOUND.

i

5. OCT. 2, 1986 AN ADDITIONAL LOP CONDUCTED

6. OCT. 3, 1986 AIT EXITS

7. OCT. 6, 1986 CAL REVISED AND EXTENDED O SLIDE 4 A-8 2--

/.4

AIT CONCLUSIONS i

EVENT CAUSES l

I .

EQUIPMENT FAILURES (6) - MOSTLY BAILEY RELATED PREOPERATIONAL TESTING DEFICIENCIES (5)

OBSERVATION ERROR (1)

NON-PROBLEM (9) - WORKED AS DESIGNED DESIGN / DESIGN CONTROL (9) - MOSTLY MINOR EXCEPT FOR RG 1.97

! PROCEDURE ADEQUACY'(4)  !

CONSTRUCTION (3) i

! OPERATOR ERROR (1) ,

i.

j SECURITY (1) i TRAINING (2)

INDETERMINATE (5)

SLIDE 5 A-83

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t SIGNIFICANT CONCLUSIONS:

1. BAILEY MODULES

2. REGULATORY GUIDE 1.97 CONFORMANCE

3. PRE 0PERATIONAL TESTING -

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O RG 1.97 CONFORMANCE (POWER SUPPLY SELECTION)

AS BUILT DESIGN WAS NOT IN AGREEMENT WITH FSAR FOR:

SAFETY RELIEF VALAVE ACOUSTIC MONITORING SYSTEM LEVEL INDICATOR

, O TEMPERATURE RECORDER.

SOURCE RANGE / INTERMEDIATE RANGE MONITOR DRIVES i

O f.gg SLIDE 7

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PREOPERATIONAL-TESTING

.THE PREOPERATIONAL TEST PROGRAM WAS CONDUCTED AS DESCRIBED IN THE FSAR AND THE OL SER SOME WEAKNESSES WERE IDENTIFIED IN TESTING OF NON-SAFETY RELATED SYSTEMS BUT THESE TESTS CONFORMED TO FSAR AND RG 1.68 REQUIREMENTS.

i PRE 0PERATIONAL LOP ONLY INVOLVED LOSS OF POWER TO CLASS IE BUSES NO PRE 0PERATIONAL LOSS OF INSTRUMENT AIR TEST

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' BAILEY MODULES RELIABILITY b

CONSTRUCTION PROBLEMS JUMPER PLACEMENT FPLA PROGRAMMING O --

ADEQUACY OF FUNCTIONAL TESTING FUNCTIONAL TESTING ONE PER 18 MOS BENCH TESTING ALL FUNCTIONS NOT CHECKED TEST METHODS REQUIRED PARTIAL RECONFIGURATION OF THE MODULES  :

i O A-" SmeE e

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i REV. 3 SALEM 2 - TRANSIENT OVERLOADING OF SALEM 2 STATION TRANSFORMERS AUGUST 26, 1986 (L. BETTENHAUSEN, R1)

PROBLEM: AS ELECTRICAL LOADS WERE ADDED, THEIR EFFECTS IN ELECTRICAL TRANSIENT SITUATIONS WERE NOT CONSIDERED. OFFSITE ELECTRIC POWER RELIABILITY WAS DEGRADED, AND IN THIS EVENT FAILED TO PROVIDE POWER TO EMERGENCY BUSES.

SIGNIFICANCEi EVENT POINTED TO NEED FOR DETAILED TRANSIENT ANALYSES AS ELECTRICAL CHANGES ARE MADE..

VITAL BUS PROTECTION SCHEME PERMITTED NUMEROUS POWER SOURCE TRANSFERS.

EVENT RESULTED IN NATURAL CIRCULATION COOLING.

CIRCUMSTANCES:

LAST TRANSIENT ANALYSIS OF STATION POWER TRANSFORMER (SPT) LOADS WAS PERFORMED IN 1981 IN RESPONSE TO NRC DEGRADED GRID AND VOLTAGE ADEQUACY CONCERNS.

LOADS WERE SUBSEQUENTLY ADDED TO ALL SPTs AT SALEM.

A UNIT 1 CONDENSATE PUMP ALSO WAS TEMPORARILY CONNECTED TO UNIT 2 AUXILIARY POWER.

AUGUST 26, 1986 EVENT

- TECHNICIAN INADVERTENTLY GROUNDED VITAL INSTRUMENT BUS "C" CAUSING FALSE LOW RCS FLOW AND SG PRESSURE SIGNALS.

- REACTOR / TURBINE TRIP FROM 100%, STEAM DUMP, AND SI FROM COINCIDENT HIGH STM FLOW AND LOW SG PRESSURE

- EDGs A AND C STARTED: EDG B OUT FOR MAINTENANCE.

- STATION BLACX0VT SIGNAL OCCURRED FOLLOWING TRANSFERS OF GROUP BUSES; DUE TO LOW VOLTAGE, VITAL BUSES TRANSFERRED FROM ONE SPT TO THE OTHER, BACK AGAIN, AND THEN STRIPPED THEIR LOADS ON THE BACK0VT SIGNAL.

- SAFETY INJECTION WITH BLACK 0UT LOADS WERE SEQUENCED ONTO THEIR BUSES.

- SI WITH STATION BLACK 0UT SIGNALS CAUSED COMPONENT COOLING WATER (CCW)

PUMPS TO BE SHED AS DESIGNED, SINCE SENSED CONDITION SHOULD HAVE RESULTED i IN RCP TRIP.

- RCPs WERE MANUALLY TRIPPED, NATURAL CIRCULATION WAS ESTABLISHED, AND A PRESSURIZER PORV OPENED NUMEROUS TIMES TO CONTROL PRESSURE.

A -8B 19

- CCW PUMPS AND RCPs WERE RESTARTED AND VITAL BUSES WERE TRANSFERRED TO O SPT.

- PLANT CONDITIONS STABILIZED (EVENT LASTED APPR0X. 26 MINUTES).

- ELECTRICAL AND REACTOR COOLANT SYSTEMS RESTORED AND PLANT IN MODE 3 (HOT SHUTDOWN) WITHIN 3 HOURS OF EVENT.

CORRECTIVE khTIONS (SHORT TERM)

LICENSEE REDUCED SPT LOADS CONSISTENT WITH 1981 TRANSIENT ANALYSIS. BOTH UNITS RESTARTED AT DERATED POWER.

OPERATING PROCEDURES WERE REVISED BASED ON REDUCTION IN AVAILABLE EQUIPMENT.

GROUP BUSES WHICH WERE ON AUXILIARY TRANSFORMERS ARE NOW POWERED FROM SPTs TO AVOID FAST BUS TRANSFER AND RESULTING TRANSIENT.

CORRECTIVE ACTIONS (LONG TERM)

COMPREHENSIVE DESIGN ANALYSIS OF AC ELECTRICAL DISTRIBUTION SYSTEM.

MODIFICATIONS, IF NEEDED, FOR RETURN TO FULL POWER WITH GROUP BUSES ON AUXILIARY POWER TRANSFORMER.

MODIFICATIONS TO PRECLUDE. REPEATED HUNTING FOR POWER BY VITAL BUSES.

O NRC RESPONSE INFORMATION NOTICE WILL BE ISSUED.

SAMPLING TO DETERMINE WHETHER OTHER PLANTS HAVE INCREASED LOADS.

PREPARATION OF A TI AND AN INSPECTION MODULE OR A GENERIC LETTER.

1 A -99 20

_.. -_ - L

. ZION 2 - DEGRADED V0LTAGE - ESF BUS g( 3 LOSS OF POWER

'V SEPTEMBER 21, 1986 - (J. STEWART, IE)

PROBLEM:

DEGRADED VOLTAGE RELAYS ACTUATED DIESEL START AND ESF BUS LOAD SHED VOLTAGE REC 0VERED DURING DIESEL STARTUP AND RESET DEGRADED VOLTAGE RELAYS, DIESEL STOPPED, ESF LOADS STILL DISCONNECTED CAUSE: POTENTIAL DESIGN OVERSIGHT, PERSONNEL ERROR, ELECTRICAL RELAY CHATTER SIGNIFICANCE:

P0TENTIAL LOSS OF ALL ESF BUSES, TEMPORARILY, DUE TO COMMON MODE DEGRADED VOLTAGE SMALL TIME WINDOW (APPR0XIMATELY 10 SEC) THAT LOP COULD INITIATE POSSIBLE INADVERTENT VIOLATION OF IEEE 279 COMPLETION OF SAFETY FUNCTIQ1L _

SCUSSION:

9/21/86 GRID FLUCTUATIONS RESULT IN DEGRADED VOLTAGE CONDITIONS SENSED AT ESF BUSSES DEGRADED VOLTAGE RELAYS ACTUATE AND INITIATE 5-MINUTE TIMER OPERATORS FAILED TO FOLLOW PROCEDURE BY NOT SHEDDING NON-ESSENTIAL LOADS TIMER TIMED OUT, EDG STARTED, ALL ESF LOADS ON AFFECTED BUS WERE SHED DEGRADED VOLTAGE RELAYS RESET, EDG SHUTS DOWN BEFORE LOCKING ON ESF BUS, ESF LOADS REMAIN DISCONNECTED l ZION OPERATORS NOT FULLY AWARE OF DESIGN DIESEL ENGINE START CIRCUITRY KNOWN DEFECTIVE FOLLOWUP:

NRR WILL REVIEW ZION SPECIFICS AND POTENTIAL DEGRADED VOLTAGE GENERIC ASPECTS, IE WILL ASSIST f

• REGION FOLLOWING PLANT ACTION V

A - 90 i 2/

9

3

'N.-

TURKEY POINT 3, 4 - DESIGN DEFICIENCY IN EMERGENCY POWER SYSTEM AUGUST 15, 1986 (A ADemo, NRR)

PROBLEM: IF OVERCURRENT RELAY IN TIE-BREAKER ACTUATES, ONE OR BOTH EMERGENCY DIESEL GENERATORS MAY BE LOST SIGNIFICANCE:

STATION DESIGN PROVIDES ONLY TWO EDGs, SHARED BETWEEN TWO NUCLEAR UNITS, UNDER CERTAIN CONDITIONS, SINGLE FAILURE COULD CAUSE LOSS OF ALL FEEDER BREAKERS TO BOTH REDUNDANT DIVISIONS OF ELECTRIC POWER SYSTEM, INCLUDING BOTH EDGs SAME FAILURE COULD ALSO PREVENT ACCESS TO BACKUP " CRANKING" DIESELS LOW PROBABILITf EVENT ONE OF MANY DEFICIENCIES DISCOVERED IN EMERGENCY POWER SYSTEM IRCUMSTANCES:

ORIGINAL LICENSED DESIGN INCLUDED CAPABILITY TO CROSSTIE REDUNDANT EMERGENCY BUSES WITHIN EACH UNIT, TO BE ABLE TO FEED BOTH DIVISIONS FROM ONE EDG OVERCURRENT PROTECTION FEATURES INCLUDED FOR " ADD-ON" BUS OPERATION OF OVERCURRENT RELAY CAUSES TRIPPING OF ALL FEEDERS T0 " ADD-0N" BUS, AND TRIPPING OF EDG IF SAFETY INJECTION SIGNAL IS PRESENT CONNECTION TO FOSSIL PLANT " CRANKING" DIESELS WAS MADE BETWEEN TIE BREAKERS IN 1983 UPGRADE WEAKNESS DISCOVERED DURING MAJOR SYSTEMS EVALUATION CROSSTIE BREAKERS EXPECTED TO BE USED INFREQUENTLY, I.E., IF ONE EDG INOPERABLE, ACCESS T0 " CRANKING" DIESELS FOLLOWUP:

LICENSEE JUSTIFIED CONTINUED OPERATION " TEMPORARILY" BY LIFTING LEADS FOR RELAYS, UNDER 50.59 NRR/ PROJECT MANAGER TO INITIATE NRC TECHNICAL REVIEW

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DUANE ARNOLD - CLASS lE STATION BATTERY PROBLEM SEPTEMBER 22, 1986 (ERIC WEISS, IE) 7 ROBLEM: EXTENSIVE EROSION INSIDE BATTERY NEAR TOP 0F THE PLATES BETWEEN THE PLATES AND SUPPORTS IDENTIFIED BY VENDOR AFTER LICENSEE NOTICED WHITE SEDIMENT IN ELECTROLYTE.

CAUSE: TEMPERATURE ACCELERATED AGING SIGNIFICANCE:

BATTERIES MEET ALL TECHNICAL SPECIFICATION SURVEILLANCE REQUIREMENTS VENDOR CAN PROVIDE NO CRITERIA TO LICENSEE FOR FUTURE USE IN IDENTIFYlHG PROBLEM BECAUSE IT REQUIRES SKILL AND EXPERIENCE OF QUALIFIED BATTERY TECHNICIAN BATTERY SUPPLIES POWER TO HPCI AND SOME ISOLATION VALVES SEISMIC EVENT COULD DISABLE BATTERY AT TIME WHEN DC POWER IS NEEDED (SANDIA TESTS IDENTIFY THE PLATE TO TERMINAL CONNECTION AS THE SEISMICALLY MOST VULNERABLE PORTION OF STATION BATTERIES)

IRCUMSTANCES:

PLANT AT 20% POWER LICENSEE PERFORMS VISUAL INSPECTION IN RESPONSE TO INFORMATION NOTICE 86-37 (MAY 16, 1986) WHICH IDENTIFIED EROSION PROBLEM AT RANCHO SECO (MARCH 1986)

LICENSEE NOTICES WHITE SUBSTANCE IN ELECTROLYTE OF 250 VOLT BATTERY AND REQUESTS ASSISTANCE OF BATTERY VENDOR

! VENDOR REPRESENTATIVE NOTICES THREE PLATES SEPARATED IN ONE CELL AND 50% EROSION OF PLATE TO BUS-BAR CONNECTION IN 11 OTHER CELLS l CELLS WERE INSTALLED IN 1971 AND ARE NO LONGER MANUFACTURED l 125 VOLT BATTERIES WHICH HANDLE BREAKER MOVEMENT DO NOT EXHIBIT l CORROSION VENDOR RECOMMENDS REMOVING ONE CELL FROM SERVICE IMMEDIATELY AND l REPLACING OTHER CELLS IN NEAR TERM LICENSEE REQUESTS CONTRACT ASSISTANCE TO TEST 11 CELLS LICENSEE NOTICES 2 OTHER CELLS WITH DEGRADATION l

n h -93

_d __

DUANE ARNOLD - CLASS 1E STATION BATTERY PROBLEM SEPTEMBER 22, 1985 (ERIC WEISS, lE)

RCUMSTANCES, (CON'T.)

IE CONTACTS VENDOR AND DISCUSSES SEISMIC VULNERABILITY -

INFORMATION RELAYED TO LICENSEE VIA RESIDENT INSPECTOR LICENSEE' PERFORMS SEISMIC ANALYSIS AND CONCLUDES 14 CELLS MUST BE REMOVED FROM SERVICE - PLANS REPLACEMENT WITH LEAD-CALCIUM VENDOR EXAMINES ROOT CAUSE AND POTENTIAL FOR PART 21 REPORT VENDOR BOARD CONCLUDES PROBLEM CAUSED BY ACCELERATED AGING DUE TO DIFFERENTIAL TEMP AND NOT A MANUFACTURING DEFECT VEND 0R SAYS PROBLEM IS CONFINED TO PLANTE' DESIGN NO OTHER PLANTS ARE THOUGHT TO HAVE PLANTE' DESIGN MADE BY THIS VENDOR FOLLOWUP IE EVALUATING GENERIC CONCERNS AND CONSIDERING SUPPLEMENTARY INFORMATION NOTICE o

U (3

\J p-94 a

> DUANE ARNOLD - CLASS lE STATION BATTERY PROBLEM SEPTEMBER 22, 1986 (ERIC WEISS, IE)

CUMSTANCES, (CON'T,) l' IE CONTACTS VENDOR AND DISCUSSES SEISMIC VULNERABILITY -

~ INFORMATION RELAYED TO LICENSEE VIA RESIDENT INSPECTOR l

LICENSEE' PERFORMS SEISMIC ANALYSIS AND CONCLUDES 14 CELLS MUST BE REMOVED FROM SERVICE - PLANS REPLACEMENT WITH LEAD-CALCIUM ,

VENDOR EXAMINES ROOT CAUSE AND POTENTIAL FOR PART 21 REPORT l VENDOR BOARD CONCLUDES PROBLEM CAUSED BY ACCELERATED AGING DUE TO DIFFERENTIAL TEMP AND NOT A MANUFACTURING DEFECT VENDOR SAYS PROBLEM IS CONFINED TO PLANTE' DESIGN NO OTHER PLANTS ARE THOUGHT TO HAVE PLANTE' DESIGN MADE BY THIS VENDOR FOLLOWUP IE EVALUATING GENERIC CONCERNS AND CONSIDERING SUPPLEMENTARY INFORMATION NOTICE O

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TRANSIENT OVERLOADING OF SALEM 2 STATION TRANSFORMERS AUGUST 26, 1986

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L. H. BETTENHAUSEN Region I FTS 488-1291 A - 98

0 SALEM 2 EVENT OF 8/26/86 BRIEF SEQUENCE OF EVENTS

-Vital instrument bus grou nd; reactor trip and safety injection

-Plant equipment responds as expected

-1 minute la te r, vital buses sense blackout; normal power available

-Vital buses strip a nd re seq u e nce loads

%, -Ope ra to rs note loss of component cooling to RCP's

-Trip RCP's and enter natural circulation

-Natu ra l circula tion established: RCS pressure control by PORV

-Safety injection reset; operators begin to restore equipment

-Component cooling restored; safety injection flow terminated

-RCP restored; RCS pressure restored

-Vit al buses returned to offsite power l . . . - . . - - - - . - . . - . . - . . _ . _ _ . . -.. _

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t I

l l

. 1 APPEllDIX XII b -

INTERNATIONAL OPERATING EXPERIENCE COMMENTS ON CHERN0DYL MEETING m.

Date: 9. September, 1986

[{W)'t s Memo to:

From:

ACRS Members W. Kerr gr./

Subject:

- Some Comments on the Chernobyl Meeting .};}

INTRODUCTION .y6 4 These-comments are meant to supplement the material in the the official report'provided by the Soviets and translated by the IAEA. There is some overlap,.but some of what is here came from asides made by the Soviet representative during their i ni ti al' presentations, and some was provided during the: working sessions in response to questions.

The first day's presentation was made by Dr V. A. Legasov, (First Deputy Director,.I. V. Kurchatov Atomic Energy Institute) who was in charge of the recovery team, and was the team leader for the delegation. During the working sessions on reactor behavior Dr.

A. A. Abagayan (Ministry of Energy and Electrification, Nuclear Power Station Institute), directed the group which responded to questions.

THE " EXPERIMENT" '

The RBMK reactors at Chernobyl,are on a preventive maintenance

' [ \ schedule which calls for a shutdown once each year. Every fourth

\ year there is an extensive program called " capital" maintenance.

The program for the other annual shutdowns is called a " medium-range" program. The shutdown during which this experiment was to be run was one of the " medium-range" programs.

The limiting design basis LOCA for this plant requires that emergency cooling be available in a f ew seconds af ter a break occurs. Thus if electrically driven pumps are used to supply emergency cooling, they must either have offsite power avai l abl e, or an almost immediately available source of emergency electric power. Since their diesels require about 2-3 minutes to start and pick up load, they have designed a combination ECCS , part of which provides a short term, but almost immediate11y available water supply, and a second segment, for longer term cooling, using diesels, if offsite power has been lost. The short term system consists of two identical water storage tanks, under pressure, connected to the reactor coolant system by motor operated valves supplied by a battery bank. As an additional source of early cooling they were exploring the use of the stored kinetic energy

'in the turbine generator, after turbine trip, to provide electric power to run some of their main coolant pumps for a period of about 45 seconds. An experiment in 1982 had indicated that, with the system tested at that time, the voltage decrease during rundown was unacceptable. They had subsequently redesigned the voltage regulator to correct the problem. The experiment was to

! test this system.

p s, s .

h/13 l n

(j%lV7

.j j PRE-TRANSIENT SCENARIO V

Since the experiment involved tripping the turbine, .thus, it was tkNd -

assumed, decoupling the reactor from the experiment, no Mr?'

particular attention was given to reactor behavior in experiment'~ 7 planning, it having been concluded that this was an experiment.d ts 5:

involving only electrical equipment. (However Abagayan insisted during the workshop sessions, in response to a question, that the station manager was responsible for a safety review, and that one had been performed. It simply was inadequate.) The experiment was said to have been under the supervision of an electrical engineer, the same one who had directed the 1982 experiment.

The reactor, once the turbine valves had been closed, could have been shut down. However in order to make it possible to re-run the experiment almost immediately if it was unsuccessful on the first try, it was planned to drop the power, but not to shut the reactor down. (Probably, if the reactor had been completely shutdown, xenon poisoning would have prevented an immediate restart. If the experiment was not completed during this maintenance period, it could not be scheduled for another year.)

The experiment was to be run with,only one turbine generator on line. Hence on the afternoon of April 25th, (*1400) one turbine

? ' ? y-ws generator was taken off line, and reactor power was reduced to

, ] I about 1600 MWth. However when the station personnel were about N- / to begin the experiment they were directed by the system load dispatcher to delay it until that night, because the electric power was needed by the system. The experiment was postponed.

In preparation for the experiment the ECCS system was disabled.

It remained disabled. (Legasov, in his presentation, said that it was disabled "in accordance with regulations". Abagayan, in later comments, remarked that it was turned off "in order to avoid introduction of cold water into the fuel channels". Later comments seemed to indicate that it was disabled in violation of regulations. It is possible that what was meant--remember we were listening to a translation--was that the disabling was contrary to regulations, but in accordance with the test procedures. In any event the f act that the ECCS was unavailable seems to have had no significance except perhaps as an indication of a somewhat casual attitude toward the deliberate disabling of safety systems.)

By that night, when the experiment was to begin, the reactor had been operating at roughly 50% power for about 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />. In preparation for the experiment the operators attempted to drop the power to about 700-1000 MWth and stabili=e it there. They were unable to achieve stabilization at this level. (Legasov commented that the auto control system would not " synchronize" at this level. It may be that they had used an improper set point.)

',, ]f s Eventually the power dropped to about 30 MWth. (Xenon buildup was

);3/s_, occurring) However by manually withdrawing control rods, the operators eventually were able to raise the power to about 200 MWth. In the process, however, only 6-8 control rods remained in position in the core. (Normally a minimum of 30 rods

/% -//4

i

D so that if 5 [jm)e to be kept in a high worth positionthe in the ratecore, of increase of is 3 emergency shutdown is required, negative reactivity is adequate to withtakethecare of any explicit anticipated approval of This the In some situations, transient.

otation manager, as few as 15 rods can be in this position. During is the minimum permissible under their " tech specs").Legasov comment his presentation, installed to prevent reactor why some interlock system was not operation with f ewerwas thanthatthe2530years rods ago, in a high when worth these position.

plants The reason, he said, were designed, it was thought that humans were more reliable than the hardware then available.

two additional cooling Just before turbine trip was to occur,one in each cooling circuit. (Normally pumps were turned on, For the experiment there are 3 in operation in each circuit. and four were to be connected to thepower site trippedinturbine order togenerator, supply cooling four were to be tied to offpoint the coolant flow rate became this for the reactor.) At (the translator's words) and cavitation may

" dangerously have occurred high"(possibly introducing bubbles into the coolant) .

in the steam separator began to fluctuate.

At this time the level This would normally have produced an emergency shutdown, but this shutdown system had been disabled in order to conduct the At this stage only 6-8 rods were in the core.

) ~gexperiment.

4 )

V TRANSIENT INITIATION Turbine trip occurred at 01:23:04 (April 26th). Since athis two would normally produce emergency shutdown, turbine trip will but this shut down system normally have shut down the reactor, ,

had also been disabled for the test.

At 01:23:40 an order was given Power immediately began to rise. The rods did not go in all the to way, scram the reactor manually.and the operator dropped (Thethem under gravity by dis Soviets servos. However the rod insertion was too late.

full power operating calculate void coefficient at normal They estimate that at the conditions as positive 2 E-4/% void.

core conditions that existed at the 1.5 beginning times of the transient this.) Almost the positive coefficient was about followed, in about four simultaneously an enplosion occurred, seconds, by another explosion.

Calculations, made by the Russians, (there are f ew data available because of lack oforappropriate instrumentation, destruction possible diversion of the normal of data storage instrumentation, indicate a rise computer facility to the experiment 110-120 times being run.)

the normal full power in in power to a value some y n about 4 seconds.

O EARLY DAMAGE .

Observers outside saw release of glowing particles About 30above the fires were reactor building at the first explosion.

A -#C 3

. ~ _ _

\ .

O I- started. Soviet analyses conclude that energy input into the

[v sT fuel produced by the first excursion was at least 300 calories

\s / per gram. (During one of the working sessions No awa, from * ,

Japan, commented that 200 cal per gram produces molten UO2. That if the energy input was more than about 400 cal per gram, vaporization was likely.) This, they conclude, led to rupture of fuel channel s and release of steam, providing enough energy to lift the massive top plate above the reactor through which all the channel outlet pipes pass. The plate was said to be lifted off onto an incline, from whence it slid, severing all the channel outlet pipes. The second explosion (there is no universally accepted cause--there are conjectures which include

[13 a second power surge, [2] a hydrogeq explosion C33 no explosion, an echo from the first) is thought by the Russians to have blown chunks of graphite from the top of the core, to have produced additional fuel damage, to have produced the destruction of the top of the reactor building, and to have started the graphite fire.

FIRE FIGHTING First efforts were made at trying to control the fires, with priority given to protecting Unit 3. They concentrated on the

" machine room",.the cable channels, and the oil supplies. They

, . , used "mostly water". Fire fighting required going to the top of'

' p-~s the machine room and to the top of the reactor building of Unit

( j( ) 3, and required lifting equipment to those locations. General y_ / Kimstach, who is in charge of fire protection, praised the courage of the fire fighters. Legasov commented that the fire fighters understood the danger they faced. Most of those who died, he said, died of overexposure incurred during fire fighting.

RESPONSE FROM MOSCOW After the accident occurred, members of an emergency team,

• previously constituted for this purpose, were notified and assembled in Moscow about 2:30 AM. They flew immediately to Chernobyl, arriving about 8:00 AM. They had been told that the accident was serious, but were nevertheless surprised at the extent of the damage. They immediately began recovery plans.

EVACUATION Because of heat generation and the consequent plume rise, early measurements in Pripyat were misleading in terms of damage to the reactor. However as the graphite fire continued, with continuing releases of fuel and fission products, measurements in Pripyat made it clear that evacuation would be necessary. The decision to evacuate was made about 9:00 PM on the 26th (?).

,f s Unfortunately, because of the fallout distributien and the plume t

i dynamics, the original emergency plans could not be used.

13'(_,/ However Pripyat (population *49,000) was evacuated in about 2 1/2 hours, followed by Chernobyl (popul ation *12,500), and then those remaining within a 30 km radius of the reactor (total population, ,

h -//h 4

1 .,

N, .

'~'N including Pripyat and Chernobyl, *100,000. At some places in the

~( )reportthenumber evacuated is given as 135,000). ,,

p/- t.. .

RECOVERY by d..

wg 3 After evacuation, attention was focussed on the reactor. The T. .

graphite was burning "which limited core temperature to that of burning graphite". The dilemma--try to put out the fire, thus risking a core temperature rise, or let the graphite continue to burn, thus keeping the temperature down. It was decided to try to extinguish the fire because of the large aerosol release that was being observed (In response to questions the amount of graphite burned was estimated to be about 250 tons, about 10% of the total). It was determined that the fission process had ceased by measurements made on iodine isotopes, and also by inability to detect any neutron flux.

First, to guard against recriticality, about 40 tons of baron carbide were dropped on the top of the reactor from helicopters.

Then about 800 tons of dolomite were dropped to provide cooling and filtering of aerosols. Next 2400 tons of lead were dropped to melt and dissolve fission products, and to provide shielding which made it possible to get closer to the damaged reactor.

(Prior to this could not get closer than about 200 meters.)

r< ', Eventually about 1800 tons of sand and clay were dropped to rs provide additional filtering of aerosols. (Total, about 5000

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Itons). Most of this was in place by May 2.

By May 6 the Soviets estimate that about 50 million curies of activity had been released (*3.5% of the core inventory).

(Others think that as much as 20% may have been released.) At this point the release rate dropped by many orders of magnitude.

NOTE: This is a preliminary and partial report. It probably-contains errors in fact and in interpretation. I plan to supplement and to review it for accuracy.

wk 8-X-86 (Rev 2)

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APPENDIX XIII ADDITIONAL DOCUf1ENTS PROVIDED FOR ACRS' USE 318TH

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ADDITIONAL DOCUMENTS PROVIDED FOR ACRS' USE

1. Letter, R. Naegelin, Director, Swiss Federal Nuclear Safety Inspectorate to R. D. Hauber, Assistant Director for International Cooperation, Office of International Programs, U.S. NRC, ACRS Request for a Description of the Pressure Suppression Containment Systems used in Swiss-BWRs, September 16, 1986

2. Letter, R. Wilson, Department of Physics, Harvard University, to The Honorable M. Dukakis, Governor of State of Massachusetts, Comments upon the 3ress release of the honorable Michael Dukakis, Governor of flassaclusetts September 20, 1986, September 21, 1986

3. Letter, N. Mavroules, Member of Congress, to D. A. Ward, ACRS Chairman, In the matter of Seabrook Station probabilistic safety analysis, October 10, 1986 4 Letter, P. McEachern, Democratic Nominee-for Governor of New Hampshire to W. Kerr, ACRS member, Press release on Seabrook Emergency Planning Zone, September 25, 1986 O

(' 5. Letter, Diane Curran for Harmon & Weiss to D. A. Ward, ACRS Chairman, ACRS Review of Seabrook Probabilistic Risk Assessment, October 10, 1986

6. News Release from the Office of Governor Michael S. Dukakis, Dukakis will not submit Seabrook evacuation plans; says emergency planning cannot adequately protect public, September 20, 1986

7. Testimony to ACR1 October 10, 1986, prepared by Institute for Resource and Security Studies, Some Comments on Recent Studies Sponsored by Public Service Company of New Hampshire, Regarding Emergency Planning at the Seabrook Nuclear Plant (Gordon R.

Thompson on behalf of The Attorney General, Commonwealth of Massachusetts), October 10, 1986 O

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