ML20235E964
Text
_ _ , - _ - - _ _ _ _ _ _
e UNITEo STATES 52 R
NUCLEAR REGULATORY COMMISSION ADVISORY COM:dTTEE ON REACT 0 WASHINCToN. D. C. 20555 TEGuArDS U.S. A.E.C.
AUU 5 hD
- 8 3 AUG 6 MS a ,
72 6,9,101112,12:3 3 4 4,ti 6 ;
DOCKETS NOS.: 50-219, 50-220, 50-237, 50-245, 50-249, 50-254, 50-259 and 50-260, 50-263, 50-265, 50-271, 50-277, 50-278, 50-293, 50-296, 50-298, 50-321, 50-324 and 50-325, 50-331, 50-336, 50-341, 50-354 and 50-355, 50-356 LICENSEES: Boston Edison Company, Carolina Power 6 Light Company, Commonwealth ,
Edison Company, Detroit Edison Company, Georgia Powcr Company, '
Iowa Electric Light 6 Powcr Company, Jersey Central Power 6 Light ;
Company, Nebraska Public Power District, Niagara Mohawk Power Company, Northeast Nuclear Energy Company, Northern States Power Company, Philadelphia Electric Company, Power Authority
-State of New York, Public Service Electric and Gas, Tennessee Valley Authority, Vermont Yankee Nuc1 car Power Corporation.
FACILITIES: Pilgr.im., Brunswick Units I and 2, DresdG Units 2 and 3, Quad Cities Units 1 and 2, Enrico Fermi Unit 2, llatch Units 1 and 2, Duane Arnold, Oyster Creek, Cooper, Nine Mile Point 1, Millstone Unit 1, Monticello, Peach Bottom Units 2 and 3, i Fitzpatrick,llope Crock Units I and 2, Browns Ferry Units 1, '
2 and 3, Vermont Yankee SU M RY OF MEETING HELD ON JULY 17, 1975 WITil MARK I OWNERS GROUP On July 17, 1975, representatives of the above named utilities (denoted j collectively as " Mark I Owners Group"), their architect-engincors, constructors, l General El.cctric Company, and the NRC staff met in Bethesda, Maryland. '
The purpose of the meeting was to discuss the role of the Mark I Owners Group and their program and schedule for determining (1) the Mark I safety /
relief valve and LOCA dynamic loads, and (2) the impact of those loads on the operating Mark I plants. A list of attendees is enclosed.
Mr. Keenan, as Chairman of the Mark I Owners Group, indicated that the group was established to coordinate the resolution of the generic probeim associated with relief valve and LOCA dynamic loads. He stated the Mark I Owners Group had no legal status for representing each individual utility in licensing actions. Mr. Keenan will be serving as the contact with the NRC and the Mark I owners Group for arranging meetings, informal transmittal of information, etc., but will not be abic to officially speak for the ,
individual licensees.
Representatives of GE, Teledync, and Bechtel made presentations on the status of the short range (completion in September 1975) and long range programs (scheduled completion in the last quarter of 1976) designed to verify the Mark I containment function. Copies of some of the slides presented by the Owner's group are enclosed. A swnmary of the objectives of the short and lon rm ograms are shown on the enclosurcs.
f eo1 05 63 .__
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9
- Meeting Summary .
The owner's group concluded that continued operation of Mark I containment BWR's present no danger to the health 'and safety of the public. At the request of the NRC staff, the licensees agreed to doewnent the technical justification for their conclusion in letters to be submitted by July 31,.
1975.
The NRC staff stated the following:
- 1. The completion of the short term program'should be expedited.
- 2. Design effort, procurement, etc., for any modifications identified during the short range program should be initiated in parallel with the long range program. -
- 3. The earthquake should be included in the short range prcgram as well
~
as the 1ong' range program.
- 4. AnydatafromforeignplantswhichisusedintyeMarkIprogram must be supplied to the NRC for review.
- 5. Mcotings between the.0wner's Group, contractors, and the NRC Structural
- Engineering,' Mechanical Engineering,.and' Containment Systems Branches should be arranged as soon as possibic.
- 6. The schedule for the completion of the long range program (last quarter
-of 1976) should be improved. .
. . 8 4 W. A. Paulson Operating Reactors Branch #3 g Division of Reactor Licensing ;
1
Enclosures:
- 1. List of Attendees
- 2. Slides tc: See next page '
i L__________-__ - - 1
9 ENCLOSURE 1 ATTENDANCE LIST MARK I OWNER'S GROUP 1
AND NRC STAFF Nuclear Regulatory Commission PASNY -
W. Paulson Z. Chilazi L. Shao 4 T. Litchfield A. Gluckmann I. Sihweil. Carolina Power 6 Light Company A. Hafi:
K. Wichman R. Black M. Hartman R. Cdthren
. S. Hou '
P. Chen Boston Edison Company i
l R. Stuart B. Liaw J. Larson J. Guibert e C. DeBevec United Eng. 6 Const. Inc.
D. Caphton M.'Fairtile B. Huselton H. Krug B. Redd S. Burwell R. Mattu Nebraska Public Power District K. Jabbour S. Varga F. Williams J. Glynn R. Kiessel General E1cetric Company H. Chakoff W. Butler
- L. Sobon P. Riehm M. Shirly G. Lear A. James B. Buckley D..Bridenbaugh J. Cutchin C. Anderson Yankee Atomic C. Grimes R. Cudlin Wm. Metevia J. Kudrick W. Webb G. Lainas N. Su E. Porter Electric Power Research Institute L. Beratan ~
W. Loewenstein R. Tedesco A. Adamantiades R. Maccary l
G. Sliter A. Gopalakrishnan i
l
- -------- - - - - - - - - - - - - - - - - - - ----^
1
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Bechtel Power Corporation Vermont Yankee. Nuclear Power Corp.
R. MacDonald T. Keenan D. Ong !
G. Jones Niagara Mohawk Power Corp.
K. Wiedner M. Mosier Public Service Electric 6 Gas Co. T. Dente W. Bauer Philadelphia Electric Company F. Linn R. Lo*gue TELEDYNE H.' Vollmer J. Hayward Jersey Central Power 6 Light Company W. Cooper R. Wolf e Tennessee Valley' Authority Detroit Edison Company C. Thomas J. Carter F. Gregor W. Colbert Commonwealth Edison Company D. Galle Southern Services (Ga. Power Co. )
J. Abel F. Ehrensperger Northern States Power Company G. Neils Iowa Electric Light G Power Co.
L. Root NUTECH R. Keever N. Edwards Aerojet Nucicar Company L. Wheat Northeast Utilities Service Co.
R. Smart R. Werner
. . _ _ - _ _ _ _ ~
. 4
, , , . 2 v. c ENCLOSURE 2 i DETAILED ArEDA ESTil?.TED
. SBEI E E lITf LlL' TIE _. i
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- 3. Di'TA!LO) lTES!],'il,TlG10~ GEiEr?.L L:iECTRIC II S lilH.
LG'Oli'3 CRIC:iUA trilLilry Jea, g, y.l,73 n p . ..
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- 5. DZTAlLED P!ES3fil:T10:10- GE:,':iR'i El CC~ RlC 9')filli, '
/m!T10:UL LD..1! :3 UlriERIA l/R, A, J/lES lff1LIZD) lii S.H.T-TEf?i p;pw:" 1
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6._
4 DETAlLED AGS1DA (C0iff.'D).
EST1f%1ED SUBJECT PRSfiffED 111. Titi:_
- 6. StfT%RY 0: SIGlllFICNff GSiERALELECTIilC 10Mlli, LOADINGS UTILIZED IN .. MR, & .lNi!11...
~
SHORT-EPli PR03P#i .
- 7. STATUS OF SHORT-TErli BE0iTEL 101 MlH, '
PR03P#1 PL M SPECIFIC MR.. C WlED:ER N!ALYSIS .
- 8. liEPBIEllT ASSESS."EliT TELEDYiE 10Mlii, OF SHORT-lEFli PR33R/M DR,W.C03?ER .
STRJCTUP/,L CRllERIA -
- 9. GSER6L lE0!'11 CAL lEVlB1 GEER 4L El.ECTRIC 30 MlU,-
0F LO:'3-KFli PR03RI?i MR,.P.'IN;NI L
10, liEPB'DBff EVALUAT10'10F TELEDYiiE 10MlU, LO:iG-1EFJi PR33P#1 CO:;CEf6'ING DR,W.C03PER
/PPLICN31LlTY OF /GE CODI ,.
CRITERIA ,
11 SU.IMRY BWR GRO'JP CHA1P!%'l 0?Bi
. A, CONCLUS10:5 MR, T. D KEB W1 B. GSEPAL RJESTIO:G S NGER PERIOD l.
a e GENERAL PURPOSE OF MEETlHG WITH NRC 1.
TO EXPLAIN THE EXISTEllCE OF THE 0Wi!ERS GROUP AND IT PURPOSE-PillLOSOPHY A!!D lilTEllT.
2.
DISCUSS THE RELATIONSillPS WHICH HAVE BEEN ESTABL -
AMol'G THE ORGAl11ZATlot1S lHVOLVED.
3.
REVlEW OUR OVERALL APPROACll TO lllE PRO 1E;i /J:D GEi L ,J.L TIRE FRAME
. i 4.
D]SCUSS Ill DEPTH Tile BASIS OF OUR SHORt AilD L0tlG TE PROGRiviS AilD THEIR OBJECTIVES.
5.
SOLlC!T llUCLEAR REGUL ATORY COMMISS10!1 CO.'iciEHl5., SUG A.UD OBTAlll GEi!ERAL APPROVAL OF OUR JUTEilDED 6.
PRESElli CU" POSIT!0N l!! RF. GARD TO TllE liDEOUAC.Y CD'!Thli'NNTS OF DURs TO MA1NTA1H TiiElR lllTEGRlTY THE MOST PROBAi!LE COURSE OF TllE LOCA EVENTS C THE LATEST INFORMAT10H AVAILADLE ON POOL DYllA.'ilC LO AND Al.S0 AGAlliST THE LOADS RESULTlHG FROM RELIEF BLO!!DOWN.
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- 3. EEllEV OULOYEDhll_f4PPPSACH TO THE PROBLEM AND_GI11ERAL TIME FRMiE A. SHORT-TERM PROGRAM j (1) lil-DEPTil PLANT SPECIFIC ..TECfNICAL REVIEW IN PROGRESS (2) COMPLET10i! DATE EARLY SEPTEMBER 1975 (3) BASED ON gel!ERAL ELECTRIC REFEREliCE PLAllT ANALYSIS - THE LOAD DOCU!iElil HAS BEElf TR/4!!SMITTED TO NRC !
(4) li!VOLVES GE, EPRI, TELEDYNE, PECHTEL
- n. LONS-TERM PROGRAM (1) COMBlllAT10!! 0F TESTlllG AND ANALYSIS (2) WILL VERIFY l!!DlVIDUAL 'PLA!!T C0i1TAlNMENT CAPttDLE OF MEEl]NG ESTADLIS!!ED CRlTERlA (3) gel'ERIC MODIFICAT10i! PACKAGE WILL BE DEVELOPED AS NECESSARY (4) TEllTATIVE PROGRAM ESTABL]SHED (5) TEllTATIVE PROGRAM SUBMISSI0li DATE 10 NRC: EARLY SEPT. 75 (G) TENTATIVE PROGRAM COMPLET10!i: LAST QUARTER 1976 (7) IllDIVIDUAL PLANT DESIGN MODIFICATION PACKAGES ,
IF REQUIRED,
.\
e s . - _ . - . _ . . _ - . _ _ _ _ - - - - - - - -
- 6. PES 9ff oiRf0slT10N IN TEEARD TO TH JE@CY 0? BE PAPK 1-
.00$hlENTS TO FAli#Alti TIEIR INTEGRIT/ Arf 1111SI.J1EfiT11 EE0EJEffEE_OF M10fA EVEiHS CD]SJPallHG TEL61ESI
'INF0,MTION AVAll ABLF_0HQ3L DYi!Mlf_ LOADS #1D ALSO Arall!SI
]lEJ,01T1JESULTli;G FRD'i RELIEF VALVF,Jilfh2M ;
E)(ISTlRG R(1 C0:UAllEJUS APE CAP /GLE OF FAllHAlulHG TIElR FUNCTIO:l UNDER /60VE POSTULEfED L0iDll!G AND PPESElU NO DANGER T0 IEALTli A'iD-S/E 0.: TlE PillLlC. THIS JUDGMENT IS BASED 03:
(1) EJERAL ELECTRIC REFEPE!!CE PiltJ!T /MLYSIS
'(2) LOADlil5S USED ll! IMT NMLYSIS NN CG15ERVATIVE, N!D CK1 i
. JUSTJFl/6LY E RED'JCED (3) Sil0RT TERM PROGRAM WILL CONFIRM THIS JUDGMENT.
(IO LONG-TERii PROGRAM WlLL PROVIDE A' BAS 1S. FOR CONTA1i,'iiE:1f ~ !
DESIG!! MODIFICATION:! PACl' AGES, IF I!ECESSARY, TO Ei!Atl.E SERVICE LIFETlME T0.ESTABLISilED CRITERI A.
(5) RELIEF VALVE BL0h'DOWN LOADS ARE A FATIGUE CO,'!SlDERATION A!1D WILL BE ADEQUATELY ADDRESSED AS PART OF LOUG-TERii PROGRAM. .
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RA1 TEST Gi Q.EVIS TO FAIL 15E ,
80I STPMiER TD.1 CfL0AATED -
PAN SSi IlPACT TESTS Cl 20" DIA. PIPE ' ..
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IFRLSE UWRIED iHL t.tSER TFAT USED w.mt IJiUL 10 ~*
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FuG \B0 CITY SICEFICT1TLY LESS TE'll TE1T'/EID IIR02 dRIL 10 ,
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- LB2. RISE.A','O \3.EITY
. i IGITIC'1. STr1rRC'1/SLYSIS UU2/6ED Ud Q"71ISIE l u -
IIE!TIFIED /EITIEW E< El SfIE-E4 FiIE'JJ1 ISSUED FID3RESS id0.li '1013C JIE 25 .' .
-- R!LFILL fPRIL 10 OI?iIE;T -
HTiii EVAlinTIEl G: 8.QPE!!1InTA .
- IATETA. \B.T, F01 S'S.L .
FOIO TBiT Ill AIDITICl TO F02. SSL 1ATETAL \ SIT 1f3, IJD VB.T -
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PSCTIEl LOG, llE E.. IT AIR E'F' F LIPS 1D:?.'25D /.l.3 l'?.QD IE33 35., a 6 -..- s i ei6
- AFFET GJTSDT!. CCEi olfu.ds .
). . . _ _ _ _ _ _ _ _ _ . - _ _ _ _ _ - -
A-4 ,
SHORT-TERM PROGRAM
SUMMARY
APPROACH A!!D ASSUMPTIO!)S TO BE USED
- 1. C0!! FIRM FOR ALL MARK 1 PLANTS THAT TllE C0!) Tali! MENT FuilCT10:1 IS RETAli:ED GlVEi! THE MOST PROBABLE SE0UEi!CE OF EVEl!TS FOR THE LOCA i
- 2. CRITERIO:! 0F SUCCESS WJLL BE RETE!! TION OF C0:lTAl':7,0lT FUi!CT10'!
- 3. EXAMINE KEY STRUCTUidS AFFECTliiG C0!HAl:EENT FI
- 4. SIG'ilFICANT ITY LOADS l!!LL BE ADDRESSED
- 5. SEllSITIV.lTY STUDIES TO BE USED TO FORM JUDGMD!TS ,
- 6. USE FSAR THERiiODYl!AiilC. ANALYSIS FOR POOL SWELL INPUTS
- 7. R. V. LOADS WILL BE ADDRESSED IN LONG-TERM PROGRAM (FATIGUE)
f - N. - .
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.. MARK .l. - SHORT TERM PROG RAM -
. , ' , . PRESENTATION OUTLINE
.j i
l 0 SCOPE AND OBJECTIVES i- .
+
i O' DESCRIPTION OF STRUCTURES UNDER INVESTIG/iTION
-i O PR$LIMIN/,RY PLANT GROUPING .
I l O STRUCTURAL. EVALUATION - ACTION P0tN.
y O DESIGN CRITERI A (ORIGINAL /PRESENT) .
i 0- ANALYTICAL TECHNIQUES n
O TESTING OF CRITICAL STRUCTURAL ELEMENTS i
O SUinMARY
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MARK I - SHORT TERM PROGRM,'i N. -
7/9/b-PRELIMINARY PLANT GROUPING
<g liARLY PLANTS (BUlLT 1965-1968) ,
- OYSTER CREE'(* .
s "X" STlFFENED VENT / RING HEADER JOH!T L
T. WIN COLUMNS, 0
l 10 VENTS W1TH S' LOPE OF 20
- NINE !,'ilLE FOINT UNIT 1* SPHERICAL VENT / RING HEADER .l0li,'T,
- . D PESrFM !!:"Tc p ?. 3 r T'.' " " '1' ' " " 3 .
- QUAD CITIES LdiTS 1 e 2 lu o r L \ c.. .' .d l- .
MlD-TERM PLANIS'(BUILT 1967-1970)
- MllLST0!'E UNIT 1 .
in0.NTI CE!.l.0 "X'.' STlFFENED VENTlil!NG HEADER J01'.'i, .
- VERGONT YiEEE - TWIN COLUMNS, .:
l'l LG R ii.'i 8 VENTS ..
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LATE PLANTS (BUILT 1969-1973) ,
- FITZPATRIC;(
- COOPER STATION "Y" STIFFENED VENT / RING HEliDER JC::'?,
- DUANE ARN01D ~. 8E\f N~.fdC0LUi'i,!S, H/JJOH UNiiS 1 & 2
- EN _
- FEP.l,'il UNIT 2 . .
- D. :9'.'J.N S F E R R Y
- TH1CKENED VENT / RING HEADER J0lNT, SINGLE COLUinN, S VENTS i
CONCRETE CONTAINMENT 0972) -
E R D.< S V,; C K ...;;i S i & 2
"Y" SY!FFE!:iD VENT / RING M' l .;0'E.120 ...
l T".'1H COLU;ANS, '
3 SVENTS l
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v MARK l' - SHORT TERM PROGRAM
' STRUCTURAL EVALUATION ~
e_
L . ACTION PLAN
~
L ELASTIC ANALYSIS OF BEAM'MODELS OBJECTIVES: o TO AlD PLANT GROUPING AND SELECTION OF CASES
'FOR INELASTIC ANALYSIS e
TO OBTAIN FI RST-CUT. PREDICT!0N OR VENT SYSTD!i
. DYNAMIC RESPONSE ,
1 t
.:2. CRITICAL STRUCTURAL ELEMENT EVALUATION AND TESTS l OBJECTIVES: o TO IDENTIFY CRITICAL ELEUiENTS i
o TO IDENTIFY TESTABl.E ITEMS o
TO ESTABLISH LOAD CARRYING CAPACITY BY TLSTS-e ~
TO DEVELOP STRUCTURAL MODIFICATION SCHEl.iES f' 3. ESTACLISH BOUNDARY CONDITIONS -
OBJECTIVES: o TO PROVIDE 1.lNEAR SPRINGS FOR TASK )
! e TO PROVIDE REFINED SPRING CONSTANTS FOR 'l ASK 4 i
14.
' INELASTIC ANALYSIS OF BEliM MODELS OBJECTIVES: o TO OBTAIN PRELIMINARY ASSESSMENT OF CONTAINMENT INTEG RITY >
o TO PROV!DE INPUT FOR TASK 6
- 5. FINITE ELiMENT SHELL ANALYSIS OF VENT SYSTEM (PERFORM IF NECESS!',RY 0BJEC1IVLS: o TO ASSESS SIGNi :CANCE OF SHELL PESPO!)SE.
o TO PROVIDE INPUi .OR TASh 6 t #
. 4 L6 . DETAlLED FINITE ELEMENT AN/iYS1S Oi hCAL AREAS OBJECTIVE: o TO MAKE FINAL ASSESSMENT OF C0ilTA!!.'MEi.'T .
INTEG RlTY (VENT SYSTEM, SUPPORTS, AND CDU .W.'.R!:.
^7. REPORT
DESIGN CRITERI A MARK I - SHORT TERM PROGRAM 7/7/75 V
R ORIGINAL DESIGN PER SAR PRESE
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LOAD CASES ' i.0AD CASES (A) DL + P + T ,
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~
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j LOADS:
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t, z TENTATIVE- ,
MARK I LONG-TERM TECHNICAL PROGRAM OBJECTIVES ,
THROUGH DETAILED HYDRAULIC A!1D STRUCTURAL A!!ALYSIS AND TESTING flEEDED OR Tl1 ROUGH FIXES IF !!EEDED, ENSURE Tl.iAT-THE ColiTAlilME!!TS- !
ARE ADEOUh.TE FOR 40-YEAR LIFE BASED 0.'1 CRITERIA ACCEPTABLE 10 HRC I ACT10:l plt.ilS LOADJj"i TO lj_E _ Col!SIDERED o LOAl)1NGS C0:!SIDERED li! C0ill. ORIGINAL DESIGil o REllEF VALVE AIR DISCllARGE LOADS o LOCA POOL SWELL A!1D LOCA-RELATED LOADS STRUCTURAL o DETAILED STRUCTURAL Ai!ALYSIS TO CRITERIA o TESTING Al!D/0R MODIF1 CAT 10!iS, IF !1EEDED 1
CRITERIA ESTABLISHMEllT i t
o DEVELOP STRUCTURAL DESIGli CRITERIA FOR ALL SIGlilFICANT .
LOAD COMBli!ATIO:!S !
o REVIEW WIT!! ASME CODE C0i911TTEE 8 NRC
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Pif."TS ~
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. EfFJMI!E FATIGUE LIFE C: CRITICAL TCRUS-f?FAS liiROJGH A!% LYSIS OF DATA ,
.2, CCNSID31 CCE0JIIVE R. V. ACit!; TION EST -
. :(4 0 1975)
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,YES G3SJRED) 2 1, ClD3Gil!3 LQN)S .
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- 1. CollTil!llf.D OPr. RAT 10i10F MlH CONTAINMENT BWRs PRESE UilDUE RISKS TO THE HEALTH AND' SAFETY OF THE PUBLIC' .
- 2. OUR Sl10RT AND LONG TERM PROGRAMS WIL'L PROVIDE THE FOLLOWil1G RESULTS: .
.SJJORT._IERii.PROGR6M (A-) THE SHORT TERM PROGRAM WILL PROVIDE II1 CREASED ASSURAiiCE OF TiiE ADEOU/'.CY OF MlH CONTAlliMEl;TS TO MAINTAll! TilEIR FUNCTION AGAll!ST THE MOST PROBABLE COURSE OF THE LOCA l EVElff ColislDf.hil!G TiiE Li,YEST iiii Didi i10i; Gi; i vl DYN6M!C LOADS (n) ll!DIVIDUAL PiltllT GEOMETRY WILL BE EVALUATED. ,
1 (c) IF AS A RESULT OF TlilS EVALUAT10i!, A!!Y PLAi1T IS DETERMl!!ED TO REQUIRE MODlFICAT10!!, THEN SUCl1 MODIFICATION WILL EE DESIGNED AliD ll! STALLED BY THE AFFECTED PLAliT
.LQJE_IE.PELPROERBB (A) T!! ROUGH A COMDlilAT10i! 0F. T.ESTI.tJG AllD At1ALYSIS, Ti1E LO!!G TERM PROGRAM WILL PROVIDE THE BASIS FOR A DESIGN MODIFICATI0i!
PACKAGE, IF REQUIRED, WHICH CAN BE UTILIZED BY EACH PLA!)T SO
'lWVOLVED, TO CORRECT DESIGN WEAK!lESSES TO THE EXTENT
. IiECESSARY TO EHABLE EACH CONTAlHME!lT TO MEET A SET OF ESTABLISHED CRITERIA THROUGHOUT ITS DESIGN LIFETIME, UNDER THE LOADIllG CREATED BY A LOCA EVEllT, OR COMBillATION OF RELlEF VALVE BLOWDOW!!S. ,
i
.e., .
- 3. -
-SCHEDULE STATUS:
~ ' '
'(A) SHORT TERM PROGRAM .
(1)' 111 PROGRESS (2) DUE FOR COMPLETI,0il Af1D SUBMISSION TO NRC:
EARLY SEPTEMBER 1975- .
(n) 'LONG TERM PROGRAM (1) TENTATIVE PROGRAM CREATED (2) FINAL PROGRAM DECIS10il: MID-AUGUST 1975 (3) Sul: MISSION OF PROGRAM TO I!RC: EARLY SEPTEMBER 1975 (10 IOLTATIVE PROGRAM COMPLETION: LAST OUARTER 1976 (5) SUDMISSIDH OF ll!DIVIDUAL PLAi!T DL' sigil MODIFICA110!1 PACKAGES /ASWARRANTED, i
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, 2,. .i-VALUE-1. . ACT ANALYSIS OF SEVERE ACCIDEh. PREVEb7 ION AND HITICATION SYSTEMS l
A. S. Benjamin, S. W. Hatch, D. R. Strip, P. R. Bennett, D. D. Drayer, and V. L. Behr Sandia National Laboratories -
Albuquerque, New Mexico 87185, U.S.A.
ABSTRACT As part of Sandia's Severe Accident Risk Reduction (SARR) Program for the U.S. Nuclear . Regulatory Commission, we have been investigating the cost-benefit' tradeoffs of filtered-vented containment (rvC) systems, hydrogen contro1 ~ systems, alternate decay heat removal systems, and a variety. of other reactor modifications designed to reduce the risk from severe accidents. ~ We have been appraising our; results in terms of the 1 l
'Commis'sion's proposed safety goals and in terms of the true costs of reactor accidents.
]
I Some of our preliminary results to date indicate that (1) TVC sys- !
tems are potentially cost-effectiv,e for many BWRs, but apparently not cost-effective for most PWRs; (2) the opposite is true for hydrogen con-trol systems; and (3) the most effective safety approaches include combi-nations of both preve~ntive and mitigative features. Further analyses are i being performed 'to fully investigate the sensitivity of these results to uncertainties.
INTRODUCTION I Sandia National Laboratories is supporting the NRC severe accident rulemaking activitics through a number of technical investigations that broadly address the phenomenological, systems, and human aspects of degraded core accidents. The inte- ,
gration of results from these programs to provide a technical base for regulatory decisions is the responsibility of a program entitled, " Severe Accident Risk Reduc-tion (SARR) Program." Basically, the objectives of the SARR program are to provide
- an assessment of the values and impacts of a set of degraded core prevention and ,
mitigation features,- summarized in Table I, and to assist NRC in formulating rules.
The safety approaches being considered include not only the individual features listed in Table I, but also all plausible combinations of these features, applied both to existing reactors and to new (design-stage) reactors. To quantify these tradeoffs, we have been evaluating the potential reduction in accident risks against the retrofitting and/or implementation . costs, and have been appraising our results in terms of the Commission's proposed safety goals and in terms of the true costs of reactor accidents.
The SARR program is being conducted in two phases, the first of which is scheduled for completion around September 1983 and the second two years later.
Existing value-impact studies at Sandia on filtered-vented containment systems
[1,2), alternate decay heat removal systems [3), and molten core retention devices (4) are being incorporated into the SARR program during the first phase.
i
,, Tchic I. . % graded Core 'Saf ety Features
- 1er Consideration Candidate terrevesents options q
J. Adut4onal Containaant seat annovel a. Active versus Passive
- 2. Contalament atmosphere Particulate capture
- 3. Containment atmosphare has Removal a. Filtered Versus Unfiltered
- b. Iow Flow Versus sigh Flow i
- d. Zacreased containment k rgins a. Deliberate Ignition
- b. 2ncreased Pressure capability
- c. Pressure suppression Features .j
- 5. Combustible ces control _ a. Deliberate 2pnition
- b. Inerting (Prior / Post Accident) ce seat sinks (Fogs /Foans)
- 6. core Setention Devices a. Dry versus Wet
- b. .
Active Versus Passive cooling or No cooling
- 7. Nissileshields a.- ' Gas Detonations or steam Erplosions
- b. Vessel Thermal shock
- 6. SWR Containment spray system
- 9. PWR Primary System Depressurisation 4. Automatic Versus knual
, . b. Additional Relief Capacity
- c. Pressure suppression Features
- d. Radioactivity Removal systems
- 10. Add-on Decay seat Removal systems 4. algh Pressure Vers,us Imv Pressure
- b. open loop Versus closed 1 mop
- c.
- Primary systen versus secondary system
- 11. specific Prevention concepts a. Zaproved Drain or valve Design
- b. Zaproved hintenance Procedures
- c. Zaproved control logic
- d. Daduction of F-n Mode Dependencies We are conducting Phase I of the value-impact assessment in two parts. In.the first part, we obtain a preliminary rating of the candidate safety approaches for each of several existing baseline reactors, which are listed in Table II. The risks from these reactors were originally analyzed in the Reactor Safety Study [5] and the Reactor Safety Study Methodology Applications Program (6}. The rating is based on the following steps:
Step 1:- Initial screening of safety approaches. A preliminary analysis is made for each baseline reactor to estimate the risk reduction poten-tial of the candidate safety approaches and to screen out those approaches which provide no possibility of a significant risk reduc-tion benefit.
Step 2: Initial screening of uncertainties. The phenomenological, system, and human response uncertainties are prioritized according to their rela-tive importance in the risk reduction evaluation, and the sensitivity of the results to these uncertainties is asnessed.
I
. _ _ _ _ _ _ _ - _ _ _ _ _ _ __ __ _ D
Table I. dCacription of Basell e Recetora l
l Baseline Reactor Reactor Type MBs supplier Containment Type Grand Culf BWR Ceneral Electric Mark'III Unit 1 Peach Bottom BWR Central Electric Mark 2 Unit 2 Oconee PWR. sabcock and Wilcom Large, Dry Unit 3 Calvert cliffs PWR Combustion Engineering Large, Dry l
, Unit 2 1 surry PWR Westinghouse Large, Dry Unit 2 Substuospheric
)
sequoyah PWR Westinghouse 2ee Condenser Unit 1 Step 3: Development of cost-bonefit measures. Procedures are developed to enable us to evaluate the risk reduction results in terms of monett.ry values. -
Step 4: Initial quantification of cost-benefit measures. Based on available estimates of construction cost and reactor downtime, the cost of retrofitting the proposed safety approaches into the baseline reactors is compared to the value of the risk averted. -
The analyses and results to be described in this paper pertain to Steps 2 through
- 4. (A separate paper presents results from Step 1 [7].)
In the second part of Phase I, we shift from plant-specific analyses to more generic analyses, and include a more detailed evaluation of costs and feasibility, both for existing plants and for new plants. The specific steps are as follows:
Step 5: Determination of generic plant categories. Results 'from NRC's Acci-dent Sequence Evaluation Program (ASEP) [8] are used to group reactors into generic categories.
Step 6: Risk benchmarking. The risk is reevaluated for each generic plant type, based on updated estimates of accident frequency and consequences.
Step 7: Cost and feasibility assessment and evaluation of competing risks.
For those safety approaches which appear attractive af ter Step 4, conceptual designs are developed, the overall feasibility of incor-porating these designs into existing or new reactors is assessed, potential system interactions are evaluated, and cost estimates are obtained.
Step 8: Reevaluation of Part I results. The risk reduction potentials, sensitivity to uncertainties, and cost-benefit tradeoffs are reevaluated for each generic plant type.
In this paper, we s 1 describe out cost-benefit mi odology and will present .
exar.ple results for certain specific cases (namely, cases involving containment venting and hydrogen control). Ac a preface to the presentation that follows, it is important to mention that the analyses are ongoing and that the results are preliminary. This is particularly the case in our analysis of the influence of uncertainties on the cost-benefit tradeoffs, in our evaluation of the financial aspects of reactor risk, and in our assessment of the effectiveness of certain safety approaches for preventing certain containment failure modes. Further, we have not yet specifically considered accidents initiated by acts of sabotage or by external events such as earthquakes, winds, or airplane crashes. In the Reactor Safety Study, external events were excluded as "not contributing significantly to reactor accident risks." Later studies have shown, however, that external events may be significant contributors for certain plants [ 9,10) . For all these reasons, the results to be presented should be considered preliminary findings that may be subject to change upon further analysis.
COST-BENEFIT MEASURES The measures we use to characterize the benefit of various safety approaches are as follows:
(1) The amount of risk averted (expressed 'in terms of man-rem averted over the life of the plant).
(2) The total accident cost averted (including all offsite and onsite costs).
(3) The cost of offsite effects averted (including offsite health effects and property damage).
The basis for using averted population dose (man-rem) as a measure of benefit stems from the NRC proposed safety goals [11] . Included in the Commission proposal are specific guidelines for acceptable limits of core melting frequency, prompt fatality risks, and latent cancer fatality risks. Also included is a cost-benefit guideline, which is stated as follows: "The benefit of an incremental reduction of risk below the numerical guidelines for societal mortality risks should be compared with the associated costs on the basis of $1,000 per man-rem averted." Subsequent safety goals proposed by the Atomic Industrial Forum [12] and supported by the American Nuclear Society [13] have agreed to the principle of evaluating the benefit in terms of averted population dose but have argued for a figure of $100 per man-rem.
The rationale for using actual accident costs in the benefit determination is I that it allows cost-benefit analyses to be performed on a dollar-for-dollar basis.
on the one hand, it avoids the argument over how the balance between costs and bene-fits should be fixed, since the break-even point is clearly defined in terms of I
equal costs. On the other hand, it shifts the argument to the question of how financial risk should be evaluated. Because there is a rather strong sentiment in I
the nuclear power industry against using total accident costs as a basis for safety regulation, we evaluate the cost-benefit tradeoffs in terms of both the total finan-cial risk (offsite plus onsite) and the offsite financial risk alone.
A cethod for determining accident costs is described elsewhere [14,15) . In the following paragraphs, we shall summarlze some of the key features of the methodology.
Cost of Health Effects To assign a dollar value to various health effects, we impute a social percep-tion of the worth of a life from the expenditures that society is willing to make in order to prevent a death. This approach leads to widely varying values on human lives, from the low tens of thousands of dollars for some cancer prevention tests '
and highway maintenance, to hundrgds of thousands of dollars per life for some auto l
saf ety f eaturer, to mi)* *ons of dollars per life for se mine refety and radio-logical stendards.
for purpotee of the cost-benefit examples to be presented in thin paper, we I place a value of $1 million on early fatalities and $100,000 on early injuries and latent cancer fatalities. The higher figure is larger than most values for traffic saf ety programs or equipment which are used to prevent prompt deaths, comparable to the early death as defined for our purposes. In addition, it is in the range of the imputed life values based on other considerations such as aircraf t safety. The lower figure is in the range (although slightly larger) of imputed life values based on various medical treatments or screening techniques, mostly related to cancers, which are comparable to the delayed deaths caused by radiological accidents.
An analysis of the sensitivity of the total financial risk to the values assigned to health effects has been made [14).
The conclusion is made that for most reactors, the cost of health cleanup, and replacement power. effects is low compared to the cost of property damage, Property Damage Cbsts The economic costs resulting from lost wages, relocation expenses, decontami-nation of property, lost property, and interdiction of land and farm crops are calculated by the CRAC2 computer code (16).
The costs are evaluated on the basis of statewide ' averaged land use and land value data, and utilize the actual population distribution surrounding the specific reactor site.
I Power Replacement and Cleanup Costs Estimates of replacement power costs are based on preliminary results from an ongoing research project at Argonne National Laboratory. In this method replacement power costs are estimated on the basis of the cost of replacement fuels and power availability for various regions that span the United States and part of Canada.
The dominant factor in determining these costs is the relative proportion of oil-fired backup plants versus economical alternative sources. ,The cost of replace-ment power is estimated to be Co = (0.286 x R + 0.086) $ millions per MW year where purchases. R is the fraction of replacement energy by oil-fired or noneconomical power The values of R vary from .05 to .95, with an average of .41.
Cleanup costs for reactor accidents are diffleult to estimate due to a lack of experience and data.
For the initial stages of this project we are using a value of 100 million dollars per year for ten years, which represents the cost of early decommissioning or cleanup and repair, depending upon the severity of the accident.
Discounting It is standard practice in economic analyses to use present value discounting of future income or losses to provide a basis of comparison for economic events that occur at different times, or over. a period of time. In this study we have used formulae based on continuous discounting (14,15), at a rate of 4 percent per annum.
Other Costs There are a number of other costs which have not yet been included in our analyses.
Examples are (1) the cost of medical care for injuries and illnesses sustained, (2) the cost of litigation and settlements pursuant to the accident, and (3) various secondary future analyses.
business costs. We will be exploring these other costs in I
4 1 ETTECTS OT UNCER/AINTIES a I A key element of the risk reduction ovaluation is the traatment of uncertain- {
ties. Existing risk asses =ments are currently thought to be highly dependent upon l uncertainties in the following areas:
(1) Extent of fission product release from fuel and physicochemical form, (2) Modeling of fission product retention in primary and containment systems, (3) Severity of steam explosions, hydrogen burns, and other pressure spiking .
phenomena, (4) Survivability of safety systems under severe accident conditions, (5) Accounting of human interf aces. '
To account for these uncertainties, we first define quantitative upper and lower bounds for the parameters involved. In selecting these bounds, we try to re-flect the spectrum of expert opinion as defined, on the one hand, by the Reactor safety Study [5] and subsequent risk assessments (6), and on the other hand, by the evolving industry position [9,10,17,18). Table III illustrates this bounding of un-certainties for one of the reactors analyzed in this study [19) .
Table III. , Uncertainty Bounds Used for Peach Bottom Analysis source of Uncertainty Conservative sound Wonconservative sound
- 1. Containment failure 0.01 c.
probability from in-vessel steam explosion
- 2. Rationale for con = Large steam explosion small steam emplosions tainment failure in reactor cavity with in reactor cavity with from ex-vessel 50% of debris dispersed up to ist of debris core-water inter = to suppression pool disper actions sien p, ool sed to suppres=
- 3. Zodine release form 22 and 0.7% CB32 Cs2 and 0 24 CH3I
- 5. Pa r ti c u l a t e retention MARCH / CORRAL Calculation M A RC H/COR RA L in primary system calculation plus and containment for additional other accident types Dr = 100-2000
- 6. Effect of containment All water lines fail Lines that bypass failure loads on torus may survive water delivery systems {
S. Recoverability of Common modes and plugged containment cooling All faults recoverable valves nonrecoverable (MRT = 19 hours2.199074e-4 days <br />0.00528 hours <br />3.141534e-5 weeks <br />7.2295e-6 months <br /> ) 1 syste.s j
- 9. Recoverab111ty of RPS nonrecoverable )
reactor protection RPs recoverable {
sy= ten (MRT = 2 5 houral '
1
__u
,. 2n the risk reduc' n evaluation, we calculate tw ,ets of risks, one based on the conservative assumption set and the other based on the nonconservative set, and we take the difference betw2en the two to represent the overall risk uncertainty due '
to phenomenological and system response unknowns. To date, we have performed an un-certainties analysis for the two boiling water reactors but not for the four pres-surized water reactors. Our results will reflect this state of development.
COST-BENEFIT RESULTS Figure 1 and Table IV illustrate some sample results of our cost-benefit calculations for safety approaches involving filtered-vented containment (TVC) systems and hydrogen control systems. (FVC systems are intended to prevent l containment failure from gradual overpressurization, whereas hydrogen control sys- 1 tems are designed to prevent containment failure resulting from hydrogen burning. )
i The vertical axes depict the value of man-rem averted (at $1000 per man-rem) or the value of accident costs averted (both total costs and offsite costs) for the ,
various FVC options and hydrogen control options considered. The estimated retro-fitting costs are overlayed in the form of dashed horir.ontal lines. In this format, an TVC or hydrogen control option would be considered cost effective if the bar depicting its value lay predominantly above the' dashed line depicting its cost, and would be considered not cost effective if the opposite were true.
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i * .
Table IV. Cases Considered for Cost-Benwelt Comparison casa cont 4 ara.ent Conf 1pration Modifjestion Praor to Modification A As analysed in ' original PAA - low-Volume Containment Vent 4-15,000 ft3/ min.)
~
3 As analysed An original PAA high-Volume containment Vent (=300,000 f t 3/ min. )
C With hydrogen control and specific preventive times (see below) tow-Volume containment vent't-15,000 ft3/ min.)
D . Wath hydrogen. control and specific preventive fines Righ-volume containment Vent (-300,000 f t 3/ min. )
E As analysed in original PAA Wydrogen Control system F With specific preventive fixes only sydrogen Control system '
2 identification of Preventive *Fixea' for cases e, D, and F 1.
Zaproved design or maintenance of the low pressure injection system check valves to reduce the i probability of direct containment bypass (surry, sequoyah. and oconee) {
2.
Zaproved maintenance of upper-to-lower compartment drains to reduce the probability of common mode failures of emergency core cooling and containment spray recirculation systems (sequoyah) 3.
Zaproved control logic to prevent the actuation of containment spray recirculation in the event of an empty containment sump (surry) d.
hoa.ified auxiliary feedwater system, based on improvements to be implemented by November 1983 (Calvert Cliffe) .
Zaprovea reactor protection- system, based on General Electric's proposed Atws "Fim 3A' (Peach Sottom and Crand Cult) i As mentioned earlier, high and low uncertainty bounds on the benefit measures ;
have been estimated for the BWRs. The top and bottom of the bars in Figure 1 depict the difference between these bounds. On the other hand, only the high bounds have been estimated so far for the PWRs, and so the bars in these cases are left incom-plete.
Cases A, B, end E consider the cost-benefit attributes of FVC and hydrogen con-trol systems applied to the baseline reactors as analyzed in the original PRAs
[5,6), whereas Cases C, D, and F consider the incremental effects of adding these systems to containments that have already undergone some modification. These modi-fications consist of certain preventive " fixes" designed to reduce specific vulnera-bilities that were identified in the original PRAs (see Table IV). In Cases C and D, a hydrogen control system is also assumed to have been implemented prior to the addition of containment venting.
1 All add-on systems were assumed to have a failure rate of 0.01 on demand.
We assumed that the FVC systems cannot react fast enough to reduce the risk of containment3failure from hydrogen burns, but that-a high-volume venting strategy
(~300,000 ft / min) can prevent containment overpressurization both from antici-pated transients without scram (ATWS) and from ex-vessel " steam spikes" (i.e.,
pressurization caused by the quenching of core debris in water in the reactor cavity l or on the containment floor). 'the assumption regarding steam spikes appears to be l
{
important only for Calvert Cliffs, where the reactor cavity is designed to retain (
large amounts of water.
l I
Par.1's for Retrofitting sts '
Schematics of the containment venting designs used for cost estimation are shown in Figure 2.
For the BWR Mark I containment (Peach Bottom), we tapped off from two existing personnel access penetrations (one 30 inches in diameter and the ,
l other 24 inches in diameter) located above the suppression pool in the wetwell (
torus.
By choosing to vent from above the pool, we were able to use the existing I suppression pool as a fission product senibber, thereby eliminating the need for any add-on filtration capability. The additions, therefore, consisted mainly of the j
valving necessary. to open and close the vent paths and a main vent line routed from '
the torus to the turbine building roof, where we chose to exhaust the effluent.
Provisions were included to inert the vent line prior to normal operation, since the Mark I containment is also preinerted.
The cost of the BWR Mark I containment venting system was estimated by Holmes and Narver, Inc., (an architect-engineering firm subcontracted by Sandia) to be I about $1 2 million for a high-volume system (~300,000 f t 3/ min. ) and about $0 9 million for a low-volume system (~15,000 f t /3 min. ) . These costs are quoted in l 1980 dollars and include a 15 percent contingency and a 6 percent fee. The system was designed to be seismic category is i.e., it was afforded the same level of sels-mic protection as the containment.
Reactor downtime required for retrofitting was estimated to be about 15 days. This estimate was based on the architect-engineer's assessment that the reactor would have to be shut down only for construction within the reactor containment building and not for construction in the turbine building.
It was judged that the retrofitting could be accomplished during a normal refueling outage. ,
I A containment venting system has not been specifically designed for the BWR Mark III containment (Grand G,tif ) . We are currently assuming that the characteris-tics of the system and costs would be similar to Peach Bottom.
For the PWRs, we designed a BWR-style water-filled suppression pool as a compo-nent of the venting system outside containment. (We could have,used a submerged gravel scrubber in place of a suppression pools we estimate that the costs and filtration capabilities would have been similar.) Although the pool had its own (a) BWR Peasa.T
.f comoast comiaamasset p s v j v 3
....... u i
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Figure 2. Schematic Showing the Flow Paths for Two Containment Venting Systems' (a ) BWR Kark I containment; (b) PWR Iarge Dry Containment (Not to Scale.)
h22t exchanger, it was signed according to a criteric that in th? event of c losz l of ac power, the water chould remain subcooled for at lesst 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> (so es to remain ef fective as a fission product scrubber). This criterion set the pool water centent at cbout one million gallons, the size of a BWR suppression pool.
The cost of the PWR containment venting system was estimated by Holmes and Harver to be about $16 million. The entire system was designed to be seismic cate- 4 gory 1. Eecause the size of the suppression pool was determined by its capacity to retain heat rather than by the venting flow rate, the low-volume and high-volume options were comparable in cort. Since most of the construction occurred far from the reactor containment building, the reactor downtime was assessed to be short and to be easily accommodated during a normal refueling outage.
The cost estimates for hydrogen control systems are based on two concepts deve-loped by General Electric Company, one a deliberate ignition system and the other a CO2 post-accident containment inerting system [20). Although these designs were developed for Mark III containment boiling water reactors, we are currently assuming that the costs for pressurf ted water reactors would be similar. (The Mark I con-tainment BWRs are preinert 24 and do not require any additional hydrogen control).
The deliberate ignition system includes a separate power source and electrical system. It has recently been suggested that the effectiveness of deliberate igni-tion systems may be enhanced by modifying the containment spray system in the following two ways: (1) replacing the nozzles with emitters that discharge fog-sized droplets, and (2) eliminating common mode dependencies with the emergency core cooling systems (especially for BWRs) . Such modifications have not been in-cluded in the cost estimate depicted in Figure 1(d). i l
DISCUSSION AND CONCLUSIONS I
As mentioned in the introduction, the results shown in Figure 1 are preliminary and the analyses are ongoing. In particular, the following caveats need to be reiterated: '
(1) External events (earthquakes, winds, etc.) have not been included in the risk assessments, nor has the risk of sabotage. l (2) The effects of uncertainties have been only partially evaluated.
(3) Assumptions have been made regarding the effectiveness (or lack of effec-tiveness) of containment venting systems under certain scenarios. l (4) Several factors have not yet been considered in the evaluation of finan-cial risks from reactor accidents.
Keeping these caveats in mind, the following observation can be made from the results in Figure 1:
First, the estimated value of accident prevention and/or mitigation appears to be much higher when NRC's proposed $1,000 per man-rem is used as a cost-benefit mea-cure than when either total or offsite accident cost is used. The guildeline of
$1,000 per man-rem should, therefore, not be construed as; representing a surrogate for actual accident costs.
Second, large differences can be observed among the three large dry PWR con-tainments. In pa rt , this is due to the relatively larger level of risk eva3uated in the RSSMAP study for Calvert Cliffs than for Oconee or Surry. Also, the potential benefit of filtered-vented containment systems for these reactors is governed by factors not related to containment type, such as the relative independence of
, ; t r ' i : .s O tt : s; w :a ./. a:c c . . . L .. . . ..t i . t. . , tHa.p no asn coo: ( r r> ) , the reliability of onsite a >ower, and the design of the 2 ,
ctor cavity.
Third, based on man-rem averted or total accident costs, containment venting appears to be potentially cost ef f ective for the BWRs evaluated. The potential cost 4 effectiveness is predicated, however, on a simple, inexpensive containment venting l
- 4. sign that uses only the existing suppression pool as a scrubber. A highly effi- '
cient reduction add-on filtration system would increase the cost without improving the risk benefit.
1 Fourth, with the possible exception of Calvert Cliffs, containment venting systems do not appear to be cost effective for the PWRs evaluated, regardless of the cost-benefit measure used. 7he potential benefit for Calvert Cliffs may be illu-cory, because it is dependent upon some highly uncertain assumptions regarding the phenomenology of ex-vessel " steam spikes".
Fif th, hydrogen control systems appear to be potentially cost effective for the PWRs, but are apparently not cost effective for the BWR evaluated. Hydrogen control for PWRs would be especially cost effective if it could be shown that a simple deli-berate ignition system is capable of preventing containment failure from hydrogen burns with a high degree of reliability. The reliability and effectiveness of deli-barate ignition systems is currently under investigation [21).
The cost-benefit examples presented in this paper pertain to safety approaches involving the addition of containment venting, hydrogen control, and specific pre-vantive fixes to existing reactor containments. A wide variety of other safety approaches is also being considered (see Table I), and results from these analyses will be documented at a later time. Future reporting will also include updated iterations of the results presented here.
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)
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1 I
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)
_ _ _ - - - - - _ - - _ - - _ - - - - - - - - - -- -- - - - - - - - --- - - - - - - - - - - - - - - - - ~ = - - - - ~ - - - - - - - - - - -
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