ML20028H605
| ML20028H605 | |
| Person / Time | |
|---|---|
| Site: | Peach Bottom  |
| Issue date: | 12/31/1990 |
| From: | Breeding R, Helton J, Jow Hn, Payne A, Shiver A, Laura Smith ARIZONA STATE UNIV., TEMPE, AZ, SANDIA NATIONAL LABORATORIES, SCIENCE APPLICATIONS INTERNATIONAL CORP. (FORMERLY |
| To: | NRC OFFICE OF NUCLEAR REGULATORY RESEARCH (RES) |
| References | |
| CON-FIN-A-1322 NUREG-CR-4551, NUREG-CR-4551-V4R1P2, NUREG-CR-4551P2, SAND86-1309, NUDOCS 9101090470 | |
| Download: ML20028H605 (438) | |
Text
{{#Wiki_filter:- - - - - - - - - - - - - - NUREG/CR-4551 SANDS 6-1309 Vol. 4, Rev.1, Part 2 Eva~ uation 0:? Severe Accicent Risks: : Peach Bottom, Unit 2 Appendices Prepared by A. C. Payne, R. J. Urceding, II.-N. Jow, J. C. liciton, L N. Smith, A. W. Shiser Sandia National Laboratories Operated by Sandia Corporation Prepared for U.S. Nueicar Regulatory Commission Dk A CkOO b77 P PDR
_ . _ __ ____ _ _ ._ _ _ . . _ ..____.-_ _ _ m__._ _ _ _______._ _ 1 AVAllAtllLITY NOTICE ! AvakMty of Reference Matorbis Cited h NRC Pubhcatons Most documents cited b NRC publications vdll be available from ono of the fonowing sources:
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NUREG/CR-4551 SANDS 6-1309 Vol. 4, Rev.1, Part 2 i Evaluation of Severe Accic ent Risxs: Peach Bottom, Unit 2 Appendices -l Manutcript Completed tsecember 1990 Date Published: December 1990 , Prepared by A. C. Payne, R. J. I! reeding, ii.-N. Jow, J. C. lietton', L N. Smith 2, A. W. Shiver Sandia National Laboratories Albuquerque, NM 87185 ( Prepared for ~ Division of Systems Research OITice of Nuclear Regulatory Research , U.S. Nuclear Regulatory Commission Washington, DC 20555 NRC FIN A1228 iArizona State University.Tempc, AZ 2 Science Applications International Corporation, Albuquerque, NM
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i 3 CONTENTS i j . l'M.C i j APPENDIX A: AC01 DENT PROGRESSION EVENT TREP... . . . . A.1 1 i ! A.1 Accident Progression Event Tree.. . ...... . .. A.1-1 ' ] A.1.1 Listing of the Accident Progression Event . Tree............................ ....... . A.1-1 A.1.2 Listing of the Peach Bottom Binner.. .. .. A.1 80 i- A.1.3 Listing of the Peach Bottom Rebinner... /.1 85 A.2 Listing and Doncription of the User Functions... A 1
,. A.2.1 Description of the User-Function for the 1 Peach Bottom APCT.... ........... ........ A.2 1 A . 2. 2. Listinr, of the Peach Bottom APET User i Function......................... .... ... A.2 7 Additional Infortoation for the Accident A.3 d
Progression Analysis............................. A.3 1 , A.3.1 Summary of Plant Information: Peach Bottom . Atomic Power Station............ . ....... A.3 1 A.'3.2 Initializo*lon Questions........ . .... ... A.3 2 { 5 APPENDIX.B: Supporting Information For the Source Term . Analysis..................... ............. .. B.1 1 B;1. Listing of PBS0R............ ......... ... . ..... B .1 1 - B.2 PBSOR Data-File.............. ........ . ...... B.2-1 B I B.3- Sou rc e Te rm Re s ul ts . . . . . . . . . . . . . . . -, . . . . ........ B.3-1 i B.4. Information Used in Source Term Partitioning...... B.4-1 APPENDIX C: SUPPORTING INFORMATION FOR THE CONSEQUENCE ANALYSIS.. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . < . . C.1 L i APPENDIX D: RISK RESULTSc.............................. ... D.1
- l. APPENDIX E SAMPLING INFORMATION, ...,,,,,,,,.., . . . E.1 E.1 LHS Input File.PB2.INP Listing......... , ., .. E.1 1
' E.2 User Distribution Subroutine USRDST,FOR Lit. ting.
E.2-1 1 l 111
._m__=.____.._ _ _ _ . _ . - -
. . - _ ~ . . . . - - _ - . _ . . - . - - - - - - - --_ -.. - - _ -.
I I L l' 1 i i l CONTENTS (Concluded) i l l l j- E.3 Extender Code EXTulS.FOR Listing........ . . E.3-1 3-E.4 Listing of MODEL.F00 ..................... ... E.4 1 l -- E.5 Listing of CPR0BPB.DAT.............. . ..... .... E.5 1 E.6 Listings of ONES.DAT and DISTR.DAT............ .. E.6 1 [ E.7 Listing of AVENAM.DAT...... ........... .. . ... E.7 1 E.8 Listing of Log Normal Random Variate for Seismic 4 llazard Curve........................... . ...... E. 8 1-l E.9 Listing of SEIutS.F0R.......................... .. E.9 1 E.10 Listing of PDS Frequencies for LuiLLOWG.DAT, , LLNulIG.DAT, EPRILOWG.DAT, and EPRIllIG.DAT. . . . . . . E.10 1 1-1 I i ,i -
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I ] 4 i i 3 F1 CURES I i fl.gure Eng,e i A.2-1 Process Used to Determine the Mode of Contaitur.ent Failure for the Fast Pressure Rise Case.. .. .. A.2 5 B.3 1 Exceedance Frequencies for Release Fractions Internal Events (I, Cs, Sr, La).................... B.3 2 B.3 2 Total Release Fractions for Summary APB 3 and 7..... B.3 4 B.4 1 Hoan Early Fatalities vs. I 131 Release (Bq)..... .. B.4-2 D1 Excendance.Frequenc!es for Risk Peach Bottom Internal Initiators.................................... .... D.2 l E1 File Structure Used to Create Final IJIS Sample For Poach Bottom....................................... E.1 L I I V
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_._..______...m_.. _ _ . _ _ _ _ - . _ _ _ _ . _ . - . - - - - - - . . _ _ . _ _ . . _ _ 2 4 ' TABLES l T.uhlt hrt A.2 3 Peach Bottom User Function Description.. .. 4 . . . A.2 2 I B.4-1 Selected MACCS Mean Results for Single Isotope Releases For Peach Bottom........... .... ....... B.4-3 l B.4-2 PARTITION Input File for Peach Bottom Analysis.... B.4 6 l C-1 Detailed Listing of Mean Consequence Results for Internal Initiators.............. ... ........ . C.3 D1 PPJJilS Results for Peach Bottom Internn1 Initiators. D.S ' E1 LDSP Recovery Probabilities. . . . . . . . . . . ......... . E.3 l __ t i l l vil _ . . . - u_ _ ___..._. u . _ . _ . _ . _ . . _ . . . . _ _ . . . _ . _ , - - _ _ _ . _ _ ._. ._ . _. _ ; _.._ . , _ . .. . :2 2;:_w ;. _ .. .,-..,a
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4 a 4 a 1 ? 1 l 0 n o i 1-L 1 4 AP!'ENDIX A f 4 ACCIDENT PROGRESSION EVENT TREE-i. d.. i-.i J .i. i i 4 s i . W t
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- 1 1
! APPENDIX A: ACCIDENT PROGRESSION EVENT TREE A.1 Accident,frorression Event Tree 1 A.1.1 Listing of the Accident Progression Event Tree I This subsection of Appendix A lists the Peach Bottom APET. The 145
- questions in the Peach Bottom APET are listed concisely in Table 2.3-1.
The event tree itself is too large to be depicted graphically and exists 4 only as the computer input which is listed here. 4 The Peach llottom APET used in the accident progression analyses for NUREG- , 1150 is in the form a computer input file. .This file is designed to be easily understood, with mnemonic abbreviations for each branch of every + question. Comments in the APET appear to the right of $s and are ignored by EVNTRE, The structure of the input file is defined in the EVNTRE reference manual, NUREC/CR 5174.'(Ref. A1) ERRORS: L'
- 1. Question 48 Case 1, case currently fails all AC (both offsite and onsite) for seismic cases. Should be seismic LOSP (2/1)
- station l blackout (3/1) not just seismic LOSP. Fails all AC in some cases where onsite may still be_available.
- 2. Question 61, case 1, need to add no early venting (17/ 1). This case
, asks if containment is already at low pressure (i.e., previous containment failure , venting during core degradation, and venting
'before core degradation begins).
- 3. Question 120, case 8. User function should be FUN CCI3 not FUN CCI2. '
[ 4. Question 126, case'3 , 4, and'5. Wet CCI (118/2) should be wet CCI (118/2):and flooded CCI-(118/3)-.- -'
- 5. Question 126,_ case 4. Low metal content is high Zr oxidation'(121/1 +
121/2). Complement of whats in the case structure.
- 6. -Question 126, case 8. Should be 121/-1
- 121/-2 not 87/1
- 121/.1.
That is, high metal content, not high flow and moderate metals, l l 7. . Question-128, casu 1. Should be 127/ 2
- 121/1 not-127/2. I.e., no containment failure and containment _at high pressure.
Re fe rences ' JA1. J. M. Griesmeyer and L. N. Smith," A Reference Manual for the Event Progression Analysis Code (EVNTRE)," -NUREC/CR-5174, SAND 88-1607, Sandia National Laboratories, Albuquerque, NM, September 1989. A.1-1
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5 l ElpSPD t 0.000 -1.000 0.000 C+herwise $ CASE 4: NO V4 LEAK OR WWL OR WR AB%*E WATER LINE OR WWI.2M . . 0.000 0.000 1.000 $ (NOT ENOUGH TIME TO DRAIN). 36 What is the RFV pressure before core damage? S -RIP : RFV IS AT HIGH FRESSUFtE; -cyp RFV FRESSURE IS 3 E3-hip E3-cyp. E3-lop S CYCLING FRCet NEAR SRV RECLOSURE FRESSJRE TO BELCW VENTING 2 1 2 3 S F"ISSURE, -lop - RFV IS AT IfW PRESSffRE. ? ! 7 4 22 26 10 34 S CASE 1: A*WS WITH CONTINUED INJECTION, RER WOFIS AND LEVEL 2 - 9 2 3 2 S LEAK C(P83INE TO KEEP FRESSURE CYCLING NEAR SRV
'TC-CV E2nDep El-R!!R E3-CL2 .S. RECLOSURE PRESSURE.
H 0.605 0.394 0.000 $ AIS CAN FAIL ON HARSH ENVIRC*tMENT. b & 22- 26 10 ' ::4 S CASE 2- ATWS WITH wnwww INJECTICN, RER WORKS AND NO LEAX
# 9' '2 3. 1 S COMBINE TO KEEP FRESSURE CYCLING NEAR SRV RECLOSLTE .;
l TC-CV E2nDep El-RHR .E3nCL 3 FRESSURE. i t36,1,1 "132,1,2 0.000 S ADS CAN FAIL ON HARSH uumdiENT. i 8 1 1 24 5 1 22 - 22 22 S CAST 3: LOCA CR SMALL LEAK AND TRANSIENT l RESULTS IN PPV DEFRESSURIZATION BEFORE CD. i 1' + '2 + 1 + (( 1 +. 3)- ( 5 +- 6 + 7)) S A S1 FCyak .SORV S2/3 TOUV/FTB f IB TW 7 0.000 0.000 1.000 5 25 8 8 26' 44 S CASE 4: Aud FAILED DUE TO DC, RAN!XPf, OFERATOR, OR HIGH 1 + 1 '+ 2 + '(2 -3) CONTAINMENT FRESSURE. E2fDC. ElfADS El-hip E2nDeP nE3-CL3 , 1.000 0.000- 0.000 , 4 '22- 26 '10 34 - S CASE 5: ATWS WITH CONTINUED INJECTION, RHR WORKS. AND LEVEL 9 .2 3 3 S 3 COMBINE TO DECREAS* FRESSURE FROM NEAR SRV r TC-CV E2-.Dep El-RER E3-CL3 S PICLOSUPI PRESS TO IDi YALUE.
!36,1,1 t36,1,2 0.000 '
2 25 34 S CASE 5: RFV REPRESSURIZES ON ADS FAILURE DUE TO HARSH ENVIR09fENT OR HIGE FRESSURE EVEN T30tGB CONTAINML1T I 2 3 S E2nDep E3-CL3 .
$ DEFRESSURIZES. CF AT HIGH FRESS, ABCFJT 3 50 FSIG.
0.716 ~0.000 0.284 i- Otherwise $ CASE 7: REACTOR AT LOW FRESSURE. l . 0.C00 0.000 1.000-37 Will the SP flash following containment vent or ruptu-e? S SatSF . SATURATED FOOL WILL FLASH ON CF: SubcSP : SUBCOCLED 2 SatSP SubcSP 'S SP WILL NOT FLASH CN CF.
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8 22 . 22 - 22- 22 22 22 6 22 3 CASE:1: SUPPREFSION POOL SATURATED IN LCNG TEP21 RHR 7AILURE, 7 + 4 + 6 + '/-+(( 8 + 10) 2) + 9 S LONG TERM STATION BLACKOUT, HIGH FRESSURE ATWS WITH
.. AW SIW .TB TW TC-fDeP."TC-De? El-BPC TC-CV S. HPCI, OR LOW PRESSURE ATHS WITH LPC SEQUENCES.
1.000 0.000 Othe m se S CASE 2: CD BEFORE REACHING SATUitATED POCL. b.000 1.000 38 'Does the LPC systers fail to inject during TC-CV?. S fLPCV LPC INJECTION VALVES FAIL DUE TO CYCLING -LPCV .g 2 E3fLPCV E3-LPCV S LPC DOES NOT FAIL DURING TC-CV. 2 1 2 2. 3 22 36 24 S CASE 1: LPC FAILS DUE TO VALVE CYCLING DURING TC-CV. t 9 2' 2 i TC-CV E3-cyp nPCyEk - y 7.100 0 900 Otherwise S CASE 2: NOT TC-CV OR AT LOW OR HIGH FRESSURE AND NO CYCLING 0.000 1.000 $ OR FIFE BRFAK FAILS LPC. 39 Is HPSW system used in time in'TC-CV? S fBPCV : HPSW IS AVAILABLE BUT NOT STARTED IN TIME TO PREVEM
.2 E3fEPCV E3-HPCV $ CD AFTER LPC FAIlliRE IN TC-CV: -H"CV . HPSW IS STARTED IN
? 2 1 2 S TIME TO PREVENT CD.
- 2 H 2 12 38 S CASE 1: HFSW AVAILABLE AND LPC FAILS DUE TO CYCLING, -f
'3 1 $ OFERATOR NOT LIKELY 10 START HPSW IN TIME.
ElaEPSW E3fLPCV I 0.950 0.050 Otherwise S CASE 2: NOT TC-CV s1 FIPE BREAK FAILS BOTH OR LPC KEEFS-1.000 0.000 S WORKING OR .*V AT HIGH OR Lou IRESSURE AND NO 40 What is the status of low pressure ECC injection before CD? 3 CYCLING. 4 E3fLPC E3rLPC E3aLPC E3-LPC , 2 1 .2 3 4 , i 10 i 6 9 24 34 37 35 38 $ CASE 1: LOW FRESSURE INJECTION HAS FAILED DUE TO FREVIOUS 1 + 1 + (-1 1) +'I + 1 S FAILURE, INDUCED PIPE BREAK, LOSS OF NPSM CR CYCLING ON TC-CV.
~ -
ElfLPC FCyBk nE3nCL SatSP E3-SPD E3fLPCV S 1.000 0.000 0.000 0.000 2 9 ' 29 S CASE 2: LPC RECOVERABLE MAY FAIL BY FIFE BREAK OR HAE H . 5 S r.:smwn,rnl IN REACTOR BUILDING, FIFE BREAK IS .34 ! 2 ' ElrLPC CWWR S , BARSH ENVIR IS .557 FOR BASE CASE. O.672 0.328 0.000 0 000 3 9 31 32 4 S CASE 3: LPC RECOVERABLE MAY FAIL DUE TO HARSH ENVIRtM"EM 2 (2 + 2) S IN Rt2CTOR B*# u ING. ElrLPC E3-DWF E3-WWF 0.503 0.497 0.000 0.000 9-S CASE 4: LPC RECOVERABLE, NO IOK OR LEAK TO R1EFUELING ; 1 j 2 S FLOOR. [ E
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ElaRPSW. E3-LcP nTC E3-LPC E3aLPC E3-COND E3aCOND 1.000' O.000- S CASE 2: HPSW ALREADY RUNNING OR RECOVER OLE OR FAILED OR Otherwise $ RPV AT HIGH PRESSURE. 1.000 0.000 45 What is the status of HPSWF 4 E3fBPSW E3rHPSW E3aHPSW E3-HPSW 1 2 3 4 2-12 S CASE 1: HPSW FAILED ALRE.@Y. 2 12 24 1 + 1' ElfEPSW PCyEk l .0.000 0.000 i 1.000 -0.000 IF CW'R S CASE 2: HPSW FECOVERABLE ON RESTORATION OF AC. 2 12 29 S OCCURES CAN FAIL PIPE CR PCC. FIPE BREAK IS .34-2 5 , EARSH INVIR IS .747 FOR BASE CASE. S Elrf!PSW - CW'R 0.723 0.277 0.000 0.000 32
.S CASE 3: RECOTtABLE HPSW FAILS AM CF OR YENTING DUE TO 3 12 31 S HARSH ESVIRONMENT.
2 (2 + 2) ElrHPSW E3-L%T E3- WT
!4?,2,1 !42,2,2 !a2,2,3 !42.2.4 S CASE 4: HPSW STILL RECOVERAELE WITH NO CF OR CF TO REFUELING
- 1 12 S FLOCR.
# 2 F. rti. '3W '
CX3 ' 900 1.000 0.000 0.000 36 29- S CASE 5: HPSW AVAILABLE FOR f*ANUAL ACTUATION ON LOW PRESSURE. 5 30 12 12 3 IF CWWR OCCURES CAN FAIL PIPE OR NCC. HPSW NOT (1 4 + 3 I) 5 S STARTED ON TC-CV OR OTFEFWISE. E3-hip El-HPSW ElaHPSW E3 '"1P CWWR
!45,2,1 0.000 !45,2,2 .0.000 44 1 29 S CASE 6: CASE 5 CONTI!PJED.
6 12 39 1 5 +l 1 -5) 5 3 (1 TC E3fSEPS nTC CWWR ElaHPSW E3fBPCV
!45,2,1 0.000 !45,2,2 0.000 36 " 3. 32 S CASE 7: AVAILABLE BPSW FAILS AFTER CF OR VENTING DUE TO 6 36 12 12 .
+ 2) S HARSH 2.a v nnwa x . HPSW NCI STARTED ON TC-CV OR (1 4 .+ 3 1) (2 S CTEERWISE.
E3-hip El-HPSW ElaHPSW E3-hip E3-DWF E3- WT
!45,3,1 0.000 !45,3,2 0.000 31 32 $ CASE 9: CASE 7 CONTINU;'.
- 72. 39 .1 44 1 7
(1 5 +- 1 -5)- (2 + 2) 3 ElaHPSW E3fHpr V TC E3fSHPS nTC E3-DWT E3-WT
!45,3,1 0.000 !45,3,2 -0.000 1 44 1 36 S CASE 9: HP W STILL AVAILABLE WITH NO CF OR CF TO REFUELING 8 36 12 12 39 ,
5 + 1 -3 +- 1) S F100R. HPSW NOT STARTED ON TC-CV OR CTHERWISE.' 1 4 + 3 -(1 E3-hip El-MPSW ElaHPSW E3fHPCV TC E3fSEPS nit E3-HIP 0.000 0.000' 1.000 0.000
$ CASE 10: WORKING HPSW FAILS ON CW'R ONLY FIPE BREAK 1 29 CAN CAUSE FAILURE.
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S CR YENTING OR CF OCCURS A D ADS LCFIS. E3-Cr? E3-LoF 0.000 1.030 S CASE 2: FF7 AT E!GH F"RESSUFI, DC FAILIL. Cor AIA W 25 26 48 40 34 5 S FFISSURIZES, AC RESTCRED. AD LEvr* 'J LUIAGE WITH 1 2 2 -1 3 S LFC NOT FAILD. GET AUTO ADS. MAT EAVE ADS . aILt?I E2'DC E2nDeF E4-AC E3fLFC E3-CL3 S DUE TO PARSE E'fVTRONENT (EIGH TD'F AD FFISS).
!35,6,1 !35.6.3 S CASE 3: RF7 AT EIGH FRESSGE, DC FAILED, CONTAIM"ENT 4 25 25 48 34 S FRESSURIZES, AC RESTCRD, AD LEVEL 3 LEaxasE wITg 1 2 2 3 S LPC FAILED. P_8SJAL ADS. PLY FAIL AS IN CASE 3.
E2fDC E2nDeF E4-AC E3-CL3 0.629 C.372 S CASE 4: RF7 AT EIGH FFISSURE DC FAILED. CONTAIM'ENT 2 25 25 S FCISSURIZIS. AC NOT RESTCRED CR AC RESSID EUT NOT 1 2
$ LEVEL 3 LEAKAGE.
E2'DC E2nDe? 1.000 0.000 S CASE 5: RF7 AT HIGH FrISSURE. DC FAILID, CONTAI. 6 DCES 3 25 49 40 S NOT TRISSURIZE AC RESTORED. AD LPC NCT FAILED. 1 2 -1 CNLY RANDOM FAILURI. S GET AUTO ADS. E2fDC E4-AC E3fLFC 0.001 0.99-2
$ CASE 6: RF7 AT EIGH FRESSURE, DC FAILED, CCNTAINE N' DOES 2 25 48 NOT FRESSURIZE. AC RESTORED, A D LPC FAILED.
S 1 2 CNLY R*.NDCH FAILURE. S PANUAL ADS. E2fDC E4-AC 0 103 0.900 S CASE 7: RFV AT BIGH PRESSUFI, DC FAILED, CCNTAIN'"ENT DCFS 7 1 25 M $ NOT F"-IT.:JURIZE. AC NOT RES'ORED. AUS STAYS
" 1 S FAILED.
E2fDC 1.C03 0.000 S CASE 8: RFV AT HIGH FRESSURE OPERA'CR FAILS TO PAWALLY 3 8 26 34 DEFRESSURIZE, CCNTAIM'ENT FRESSURIZES. LEVEL 3 S 2 2 3 I S IEAKAGE. fd_^SJAL DEFRESSURIZATION AT THIS TIME. El-BiF E2nDeF E3-CL3 0 050 0.350 S CASE 9: RFV AT HIGH FRESSURE. OPERATOR FAILS TO PAW ALLY 2 e 26 DEFRESSURIZE, CCNTAIMdENT FRESSURIZES. NO LE 3 S 2 2 S LUIAGE. ADS NOT PCSSIBLE. EI-HiF E2nDeF 1.000 0.000 S CASE 10: RFV AT HIGH FFISSURE, CFERATOR IAILS TO MPS 31.LY 1 e S DE?RISSURIZE.CONTAI!2* INT DOES NOT FRE*SGIZE. 2 S PANUAL DIFRESSURIZATICN AT THIS TIME. El-HiF 0.200 0.800 $ CASE 11: RFV AT HIGH PRESS"RE, DC CK CO':TAIN E NT ( Otherwise S FRESSURIZES, NO LEVEL 3 LDIAGE. OFFRATOR 1.000 0.000 CAN NOT DEITESSURIZE. S 30 Is thare injection durins cere desredation? S E4-LFIn LOW FRESS"RE INJECTION BY LPC OR CC D CR HPSW; 3 E4-LPIn E4-Cria fanIn S E4-CDin - HIGH FRESSURE INIECTICN BY CRD ONLY: E4nIn - NO 2 1 2 3 S INJECTICN DURING CD. 11 f CASE 1: NO ELT ISLIES INJECTION WORKING. TW SE2 CASE 2. 2 46 22 2 -7
(- rtELT .nTW l7 . i' 1.000 0.000 .~0.000 1' 46 3 CASE 2: NO MELT, TW WITH CRD ONLY.
'2 r35LT 0.000 1.000 0.000 49 43 S CASE 3: RPV AT HIGH PRESSURE, CRD WCPXING, NOT TW..
2 1 3 ; E&nDsP !3-CRD 0.000 1.000 0.000 3 49 43 48 S CASE 4: RPV AT HIGH PRESSURE, CRD RIC,WERABLE AND AC 1' 2 2 S RESTORED. 5 EenDeP E3rCRD E4-AC C.000 0.9~0 0.100 1 49 S CASE 5: RPV AT HIGH PRESSURE, CRD FAILED OR CRD RECChWLE 1 S AND AC NOT RESTORED. E&nDeP 0.000 0.000 1.000 6 40 40 48 40 42 45 S CASE 6: RPV AT IIM PRESSURE, LPC IS AVAILASLE OR LPC IS 3 + 2 2 +- 4 + 4 + 4' S RECOVERABLE A D AC IS RESTORID. AUTO INJECTION. E4-AC E3 'PC E3-COND E3-HPSW S ALSO LPC, COND,' OR HPSW IS ALREADY WORKING.
. E3aLPC E3rLPC 7 1.000 0.000 0.000 43 43 48 $ CASE 7 RPV AT LOW PRESSURE, COND OR HPSW IS AVAILABLE, N 5 42 45: " .(3 + 33 (3 + 2 2) S CRD WORKING CR ECOVERABLE AND AC RESTORED. ' PJLNUAL E3aCCND E3aHPSW, E3-CRD E3rCRD E4-AC S INJECTION. e f50.4,2 !50.4,3 0.000 42 45 42 45 48 .S CASE 8: RPV AT LOW PhESSURE, COND OR HPSW IS AVAILABLE OR 5
(3 + 3 + (2- + . ' 2) 2) S COND OR HPe.A ARE RECOVERABLE AND AC IS RESTORID. CRD E3aCOND E3aHPSW E3rCOND E3rHPSW E4-AC S FAILED CR RECOVERABLE. MANUAL INJECTICN. ,
!50,4,2 'O.000 150,4,3 43 S CASE 9: RPV AT LOW PRESSURE, CRD IS WORKING.
1 3 E3-CRD . 0.000 1.000 0.000 48 $ CASE 10: RPV AT LOW PRESSURE, CRD RECOVERABLE AND AC 2 43 2 2 S RESTORED. E3rCRD E4-AC 0.000 !50,4,2 !$0,4,3 S CASE 11: NO INJECTION,' VESSEL AT LOW PRESSURE.
' Otherwise 0.000 0.000' 1.000 51 What is the status of containment sprays during CD?
4 E4fCS ~ E4rCS E4aCS E4-CS
- 2. 1- 2 3 4 7
0 13 40 9 38 J5 24 . 34 37 S CASE 1: CSS FAILED INITIALLY OR CSS FAILS DUE TO SUBSEQtTENT 1 '+ 1 ' -1 -I + 1 + 1 +-1 1 S LPC FAILURE ( I.E. LPC FAILED BUT NOT INITIALLY). l 4 _. ~ , _ . -. . ... . _ . .
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1.000 0.000 0.000 0.000
$ CASE 2: CSS RECO'JERAELE BUT AC NOT RnwdED FAILS FROM
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-- 1 ' 2 1' 'S S HARSH ENVIRO 5 M -(PIPE INCLUDED) (LPC ORIGINALLY.
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-t GETHRESH 3 11.35 7.91 S.15 S PARSE BASE PRESSURE.
Otherwise S CASE 3:NO CP. NEW BASE PRESSURE IS SUM OF PRESSURE 3 6 8 12 S AT CD AND DELTA PRESS FROM H2. PBasel DP-H2 PBase2 FUN-PVB3 GETHRISH 3 11.35 7.91 5.15 S PARSL BASE PRESSURE'. 83 Does an Alpha Mode event fail ooth the vessel and containment? S Alpha ALPHA MODE f. VENT FAILS THE RPV AND CONTAIh M . 2 Alpha noAlpha $ noAlpha THERE IS NO ALPHA MODE FAILURE. , 2 1 2 3 1 46 . S CASE 1: NO CORE DAMAGE. ,. 2 nMELT 0.0000 1.0000 1 69 $ CASE 2: ALPHA MODE FAILURE OCCURS WITH RPV CRIGINALLY AT 1 S HIGH PRESSURE. ES-Re?
. 0.0010 0.99C3 " Otherwise S CASE 3: ALPHA MODE PAILURE OCCURS WITH RPV ORIGINALLT AT .
b -0.0100 0.9900 S LCM PRESSURE. 84 What frattion of the core participates in core slurnp? 3 HISL MEDSL LOWSL 2 'l 2 3 , t 6 1 83 S CASE 1- ALPHA PODE FAILURE. 1 ALPFA 1.000 0.000 0.000 2 69 72 S CASE 2: RPV AT HIGH PRESSURE CRD INJECTION, AND NO ALPPA i 1 3 $ t0DE FAILURE. ES-rep ES-CRD 0.600 'O.400 0.000 1 69. S CASE 3: RPV AT HIGH PRESSURE, NO INJECTION, AND NO ALPHA 1 S MODE FAILURE. ES-rep 1.000 0.000 0.000 70 71 73 S CASE 4: RPV AT LOW PRESSURE, LPC CR HPSW CR CDS INJrlTION, 3 4 + 4 + 4 , S AND NO ALPHA MODE FAILURE. ES-LPC ES-COND E5-HPSW 1.000 0.000 0.000 1 72 S CASE 5: RPV AT LOW PRESSURE, CRD INJECTION, A'r0 NO ALPHA 3 $ MODE FAILL%E.
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!84,2,1 !84,2,2 !84,2,3 S CASE 6 RF1* AT LOW FPISSURE, NO INJECTION, AD NO ALTHA Otherwise S MODE FAILUPI.
!84,3,1 !84,3,2 184,3,3 IN-VESSEL STEAM EXPLOSION OCCURS; nVesSTx NO S VesSTx 85 Is there a larse in-vessel steam explosion?
$ IN-VESSEL STERE EXPLOSION OCCLTS.
2 VesSTx nVesSTx 2 1 2 8 N CASE 1: No t*ILT, NO STEAM EXFLOSION. 1 46 2 rEELT 0.000 1.000 S CASE 2: AN ALFSA MODE FAILURE HAS OCCURED WHICH IS A LARGE 1 83 S STEAM EXFLOSION. 1 ALPHA 1.C00 0.000 S CASE 3: THE RFV IS AT HICH PRE *.SURE AC TFERE IS A URGE 2 69 84 S AWT OF PATERIAL MOBILE AT CORE SLLHF. 1 1 E5-ReF HISL 0.100 0.900 S CASE 4: THE RFV IS AT HIGH FRESSURE AC THERE IS A VIDILH 2 69 84 S ANT OF PATERIAL POBILE AT CORE SLLHP. 1 2 ES-Ref 7
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!e5,3,1 !85.3,7 S CASE 6: THE RFV IS AT LOW FRESSURE AE THERE IS '
A LARGE 1 84 S A.%'NT OF PATERIAL POSILE AT CORE : 1 HISL 0.260 0.140 S CASE 7: THE RFV IS AT LOW FRESSURE AC TN IS A MEDILH 1 84 S A*0UNT CF PATERIAL PCBILE AT CC92 SLLHF. 2 MEDSL te5,6,1 te5,6,2 S CASE 8: THE RFV IS AT LOW FRESSURE AND THERE IS A LOW Otherwise A*OUNT OF PATERIAL POBILE AT CORE SLLM. S 185,6,1 125,6,2 ALPHA PODE , SE-Btad RFV BOTTCH HEAD FAILS: S SE-Alpha SS Does a large in-vessel steam erphsten fail the vessel? S SE-LgBreh U.RGE HOLE; SE-SmBrch SPALL HOLE; SE-nFall 5 SE-Alp SE-BtHd SE-LABr SE-SmBr SE-nFAI 4 5 S NO FAILUPI OF VESSEL FROM EXFIDSION. 2 1 2 3 3 S CASE 1: VESSEL HAS FAILED BY ALPHA 700E FAILURE. 1 83 1 Alpha 1.000 0.000 0.000 0.000 0.000 2 CASE 2: IN-VESSEL STEAM EXFLOSION BUT NO M.FBA MODE. 1 85
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i l 2 HI*dE -nHIEE i 2 1 2 18 3 46 . 63 89 $ CASE 1: NO CORE PILT OR NO VESSEL BREACH OR ALFHA MODE 2 + 1 + 5 S FAILURE (POOT). nMelt Alpha nBreach 0.000 1,000 S CASE 2: RFV FRESSURE IS LCM AT VB. 1 89 2 ES-nReP 0.000 1.000 4 85 87 89 89 +L E 3: NO IN-VESSEL SIX, LARGE VB, RFV AT HIGH FFISSUPI, 2 1 ( 2 + 3) S AND HIGH MCBILITY AT VB. nvesSTr HiLiqvB EH-fail LaBrch 0.800 0.200 ' 3 85 87 69' S CASE 4: NO IN-VESSEL STX, Sm LL VB, RFV AT HIGH FFISSURE, 2 1 S AND HIGH MOBILITY AT VB. nvesSIr HiLiqvB SeBrch
!93,3,1 !90,3,2 3 85 89 89 3 CASE 5: NO IN-VESSEL SU, LARGE VB, RFV AT HIGH FFISS'JRE, > 2 ( 2 + 3) S AND LOW POSILITY AT VB. *' nvesSTx BH-fail LgBrch & !90,3,1 !99,3,2 2 85 89 S CASE 6: NO IN-VESSEL STX, SMALL VB, RFV AI HIGH FRESSURE, 2 4 S AND LOW tOBILITY AT VB.
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!90,3,1 190,3,2 4 84 89 89 86 S CASE 7: IN-VESSEL STX FAILS VESSEL, LARGE TB, AND A HIGH I ( 2 + 3) -5 S MOUNT MOBILE AT CORE SLUMP.
HISL BH-fall EsBreh nSE-nFAI
!90,3,1 !90,3,2 4 84 89 89 86 S CASE 8: IN-VESSEL STX FAILS VESSEL, !>.RGE VB, AND A MEDILM
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!90,3,1 !90.3.2 4 84 89 89 B5 S CASE 9: IN-VESSEL STX FAILS VESSEL, LARGE VB, AND A LOW 3 ( 2 + 3) -5 S AMOUNT MOBILE AT CORE SLtadP.
LOWSL BH-fail LsBrch nSE-nFAI
!90,3,1 190,3,2 3 84 89 85 $ CASE 10: IN-VESSEL STX FAILS VESSEL, SMALL VB, AND A HIGH 1 4 -5 S A'00NT POSII.E AT CORE SLLMP.
HISL SmBrch nSE-nFAI
!90,3,1 !90,3,2 3 84 89 86 S CASE 11: IN-VESSEL STX FAILS VESSEL. CMALL VD, AND A MEDILM 2 4 -5 S ATOLMT POBILE AT CORE SLUMP.
MEDSL SmBrch nSE-nFAI
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3 4 ~5 S AMOUNT MOBILE AT CORE SLLMP. l LOWSL SmBrch nSE-nFAI ! !90,3,1 190,3,2 3 84 89 . 89 S CASE 13: IN-VESSEL STX, LARGE VB, AND HIGH A*OU!tT WBILE AT
-1 ( .2 + ' 3) S CORE SLLHP.
HISL BH-fail LgBrch 190,3,1 !90,3,2 3 84 89 89 3 CASE 14: IN-VESSEL STX, LARGE VB, AND MEDILH A*OUNT tOBILE AT 2 ( 2 + 3) S CORE SLLEP. MEDSL BH-fail LgBrch
!93,3,1 190,3,2 3 84 89 89 S CASE 15: IN-VESSEL STX, LARGE VB, AND LOW A*OUNT POSILE AT 3 ( 2 + 3) S CCRE SLL*.*P.
LOWSL BB-fall LaBrch
!90,3.1. !90,3,2 2 84 89 3 CASE 16: IN-VESSEL STX, SMALL VB, AND HIGH A*OUNT POSILE AT 1 4 S CORE SLUMP.
HISL SmBrch
!90,3,1 190,3,2 S CASE 17: IN-VESSEI,STX, SMALL VB, AND MEDILM APOUNT MOBILE 7 2 84 89 3 AT N SLLHP.
D 2 4 O MEDSL SmBrch >
!90,3,1 190,3,2 Otherwise S CASE 18: IN-VESSEL STX, .SMALL VB, AND LOW AMT MOBILE
!90,3,1 190,3,2 $ AT CORE SLLHP.
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-3 * ((2 + 3)'* 1 + (2 + 3) 5 1) S SIX AND HIGH tOBILITY AT CORE SLUMP OR A LARGE VB -[*
nES-DDry SE-EtHd SE-LgBr HISL.BB-fail LgBrch SE-nFAI HiLiqVB 'S AND NO IN-VESS STX AND HIGH tOBILITY AT VB.
!85.6.1 185,6,2 2
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1.000 < 1 ! 18 2.38 Bar (34.5 psi) Otherwise S CASE 14: VESSEL BREACH AT LOW PRESSURE INTO A DRY CAVITY, i 1.000
'1 18 0.00 95 What is the peak pedestal pressure at vessel breach 7 S Fed-VSP - PEAK PRESSURE IN PEDESTAL AT VESSEL BREACH.
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97 Does the RP7 pedestal f ail due to pressurization at vessel breach?. S I-PedFP: PEDESTAL FAILS DtTE TO FPISSURE AT V5. ( 2 I-PedFP InPedFP S InPedFP: PEDESTAL DOES NOT FAIL DUE TO IRESSURE AT VB. 5 1- 2 ; 1 19 l Ped-TBF : l THRESH' 1 11.00 S Fressure required to fail pedestal. 98 Does the Drywell fall on pedestal failure? 2 I-DWPFP InDWPFP
- 2. 1 2 ,
2' S CASE 1:' PEDESTAL FAILS ON OVERPRESSURE. ! 2 97 96' t 1 .+ 1
..I-PedFP I-PedFI .
t 0.175 0.825 ; O herwise; 5 CASE 2: NO PEDESTAL FAILUPl. 3.000 1.000 99 What is the structural capacity of DW to impulse loads? [ f y 1 DRYWELL i 3 1 [ Y 1.000 x t.n p -1 i 20 125.000 $ IMPDWF EXTERNAL IMPULSE WHICH RESULTS IN DRYWELL FAILURE. 100 Is the impulse loading to the drywell at VB sufficient to cause failure? 2 InDWFI I-DhTI3 $ InDWFI: DRYWELL DOES NOT FAIL FROM IMPULSE LOADING AT VB. f 6 1 2 S I-DWTI3: IMPULSE LOAD RESULTS IN DW RUFTLHE. 2 r 1 91 S CASE 1: STEAM EXPLOSION OCCUP.S. 5 1 ExSE 2 20 -14 IMPDWF DW-SEILd AND THRISR 1 0.00 S Stea:n explosien. S CASE 2: NC STEAM EXPIDSION, NO IMPULSE LOADING TO DRYWELL Otherwise. 14 S. AT VB. .I 1 i
'DW-SEILd t MD THRESH 1 0.00 .
S DtLtY PARA *"ETRIS FORCE NO LEAKAGE. 101 Does pressurization fell contain:nent at VB? 2 IP NIP 6 .1 2- r 2- . Y I
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i (2 + (3 +- 4) -* 1) -* (1 + 1) S~ HIGH FLOW INJECTION WITH HIGH FLOW MELT. BH-fail LgBrcH SmBrcH -HFME E5-DF1d ES-LPIn 0,380 0,620 7 89 89 . 89 90 69 . 55 55 S CASE 4: DRY OR WIT DRTTELL WITH HIGH FIDW MELT AT HIGH ; l - FRESSURE WITH HIGH METALS. (2 + (3 + 4) * - 1) * (1 *' -1 -2) S BB-fall. LaBrcH SmBrcH HPME E4-Re? nZrOx75 nZrOx50 , 0.790 0.210 4 6 89 39 89 90 69 87 $ CASE 5: DRY OR WET DRYWELL WITH HIGH FLOW MELT AT HIGH (2 + (3 + 4) . * ' 1) * (1
- 1)' S PRESSURE WITH HIGS SUPEREIAT.
EH-fail LgBrcB SmBrcH HM E4-R,P HiLiqVB i
!103.4.1 !103,4,.2 89 89 89 90 87 55 55 3 CASE 6: DRY OR WET DRTWELL WITH HIGH FIDW MELT WITH HIGH l' ~7 (2 -+ (3 + 4)
- 1) * (1 * -1 *
-2) S METALS AD HIGH SUPER!! EAT. [
4 BH-fail LaBrcB SmBrcB HFME HiLiqv3 nZrox75 nZrox50 , z
!103,4.1 1103,4,2 7 89 89 89 90 87 55 55 S CASE 7: DRY OR WET DRW' ELL WITH HIGH F14M MELT AA LOW
- (2 + (3 + 4)
- 1) * (2 * (1 + 2)) S METALS AC IDW SUF""4T. s BH-fail LaBrcH- SmBrcH HIME LoLiqvB Ir0x75 Irox50 0.600 . O. 4 GD -
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- 1) * (2
- 2) S FRESSURE WITH LOW SUPERHEA' =
7 EH-fail ,Lg!rcH SmBrcH HIME LoLiqTB E4r. rep - u !103,7,1 !103,7,2 l 7 89 89 89 90 69 55 55 S CASE 9: DRY OR WET DRW' ELL WITH HIGH FLOW MELT AE LOW
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ES-DFid E5-LPIn 0.320 0.680 Otherwise S CASE 11: DRY OR WET DR W' ELL WITH LCW FLOW MELT. 0.510 0.490
' 104 Is there a leak in the Drywell head after VB7 1t E5nDEF ES-Df!F 2 1 2 2
3 63 102 102 S CASE 1: ALREADY A DWH FAILURE CR ONE OCCURED DURING CD. 2 +7 + 9 . E4-DEF DWHL DWHR 0.000 1.000 Otherwise S CASE 2: NO FAILURE CR NOT IN DWE. 1.000 0.000 i' 105 Is there a leak in the Drywell after YB? 4 E5nDWF ES-DWF ES-DW7 ES-DWC7 2- 1 2 3 4 L L i
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.0.000 0.000 1.000 110 Is ac power not available late?
2 LfAC L-AC 2 1 2 9 S CASE 1: SEISMIC LOSP, NO RECOVERY. 1 2
'1
'El-SLOP 1.000 0.000 S CASE 2: LONG TERM SB WITH DC FAILED AT 3 ERS, VB AT 7.5 ERS, 48 22 4 3
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'O 390 0.610 4 S CASE 3: LONG TERM SB WITH DC FAILED AT 5 ERS, VB AT 9.5 3 48 >22 ERS, AC NOT RECOVERED FRCH 9.5-12.0 ERS GIVEN NOT .,
3 1 6 4 RECOVERED AT 9.5.
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E4fAC " TB El-DCS .) 0.374 0.625 S CASE 4: IDNG TERM SB WITH DC FAILED AT 7 ERS, VB AT 11.5 1 3 48 22 4
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$ RECOVERED AT 11.5.
7 E4fAC TB El-DC7 l 0.364 0.636
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E4fAC . TB El-DCS 0.355 0.645 4 S CASE 6: -LONG TERM SB WITH DC FAILED AT >12 ERS, EPCI FAILURE 3 48 22 3 . DRIVES, EPCI FAILS AT 10 ERS, VB AT 15.5 ERS, AC NOT 1 -6 7 S RECOVETJdD FRCH 15.5-18 ERS GIVEN NOT REC AT 15.5. E4fAC =TB El-DC>12 0.: . 0.648 4 S CASE 7: SHORT TERM STATION BLACKDUT WITH DC POWER. VB AT 2 48 3.6 ERS, AC NOT -REC 07ERED FRm 3.6-6.1 ERS GIVEN 3 1 2 S NOT anwur.nili AT 3.6 ERS.
,E4fAC El-DC 0.473 0.527 4
$ CASE 8: SHORT TERM STATION BLACK 0CT WITBOUT DC POWER. DC 2 48 LOST AT T-0, VB AT 3.6 ERS. MCCI AT 6.1 HRS.
1 S. 1 SEIFT CURVE'1.1 ERS,; in:.xtrunE, NO R MY FROM S E4fAC ElfDC S 2.5-5 ERS GIVEN NOT RECC"ERED AT 2.5 liRS 0.519 0.481 S CASE 9: AC RECOVERED FRTVIOUSLY OR NEVER FAILED. Otherwise ' 0.000 --1.000 111 What is the status of low pressure ECC after.vessei breach? e2 EfLPC- L-LPC
'2 1 -- 2 11 109 110 89 S CASE 1: LPC ALRFADY FAILED OR AC NOT RESTORED CR SP DRAIMED .
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i a i _ y t t 2 S NO CF OR NO CF TO R3. ; ElrRHR 0.100- 0.000 3 9 62 102 S CASE 5: RER WORKING, AND LIC FAILED INITIALLY. CATASTROPHIC 1 (5 + 5) S. WETWELL RUPTURE. ElfLPC CWWR CWWR
!40,2,1 140,2,2 5 9 105 105 106 106 S CASE 6: RER WORKING, AND LPC FAILED INITIALLY. CF TO RB.
1 (2 + 4 +2 + 4) ElfLPC ES-DWF ES-DW9V .E5-WWF E5-WW8V
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0.000 1.000 115 Do containoent sprays operate following vessel breach? 2 nLCS LCS' 2 1 2 3 6 $1 111 9 38 110' 114 S CASE 1: CSS FAILED OR AC NOT RESTORED OR LPC FAILED. 1' + 1 ' -1 -1 + 1 -+ 1 E4fCS' LfLPC nElfLPC nL3fLPCV LfAC .nLRER k
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3 E4eCS E4rCS. 0.000 1.000 Otherwise S CASE 3: CSS ALREADY WORKING. 0.000 1.900 e 115 Is service water sprayed following vessel breach?- 2 nLSWS LSWS 2- 1 2 2' 3 113 115 110 S CASE 1: RPSW FAILED R INJECTING INTO VESSEL OR AC FAILED
-2 + 2 + 1 S AND NOT RES*JRED CR CSS WORKING.
nLaEPSW LCS LfAC 1.000. 0.000
$ CASE 2: HPSW f; i1LABLE, MATJAL ALIGNMENT REQJIRED.
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- 89 111 112. 113 115 116. S CASE 1: NO VB IMPLIES INJECTION INTO VESSEL CR CSS. H N ,
6 5 + 2 + 3 + 3 +'2 + 2 S - COND, OR LPC WORKING.' nBreach L-LPC L-COND L-HPSW LCS *SWS 0.000 0.000 1.000 3 112- 113 - 72 3 CASE 2: COND OR HPSW AVAILABLE AND CRP WJRK7NG. MANUAL
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s ,c ,oe m v i suc.iam 5:en aic>a s cowviu es ,o: n s. et6 J.'& m Z ?,'f,' 'L T,?,i' '" BIBLIOGRAPHIC DATA SHEET w m4 o , . w .,, , .m . ,,,, .. NURit/CR-45Si g, F"""'" Vo1. 4, Rev. 1, Iatt 2 Evaluntion of Severe Accident Rishs: Ibach Ibt t om, Uni t
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SPICOUICOS pegg ,b e y l 14[0
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A1322 s sono,a s ,,,so,estos-Arthur C. Payne, Jr. Lanny '.i. Smith ** Technical i Roger J. Breeding Ann W. Shiver . Ilont-Nian Jov t ' ' ' ' 00 ' O ' ' ' l ' " " Jon C. Helton*
- Arizona State University
** Science Applications International Corp.
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> a e,*o~a, .: anos a e s:,,.oo.,.c,* . uv i is o aoo a i 5s m .,c ,,,,,, o.,, . c,, , .. ..,... y ,..,,..,,c-~ae,~a.,,,,..........,,,........
Sandia National Laboratories Albuquerque, NM 87185 3, ,, .. .=< ,,,r , . n e c o.. . o ... ., a,,,a v 4 = ... . ..... c -~...r 5 sec,,hs,c,,a e,,,.,,...o is :, o n c a N.r anos . h av i Asc a oon i ss m n e c .. Division of Systems Research Office of Nuclear Regulatory Research U. S. Nuclear Regulatory Commission Washington, DC 20$$$ n ssu sivistan satis n s m u t , w,,,, ,. .., In support of the Nuclear Regulatory Commission's (NRC's) assessment of the risk from severe accidents et commercial nuclear power plants in the U.S. reported in NUREG-1150, the Severe Accident Risk Reduction Program (SARRP) has completed a revised calculation of the risk to the general public from severe accidents at the Peach bottom Atomic Fover Station, Unit 2. This power plant, located in southeastern Pennsylvania, is operated by the Philadelphia Electric Company. The emphasis in this risk analysis was not on determining a "so-called" point estimate of risk. Rather, it was to determine the distribution of risk, and to discover the uncertainties that account for the breadth of this distribution. Off-site risk initiated by events both internal and external to the power station were assessed. h $avwoaosonsca noes 6.. .,,,,,....... ,,,,,,, , m ., ,= ,,. , ia .. .. 46..m o m vi,a Probabilistic Risk Assessment, Reactor Safety, Sever Accidents, Unlimited g .,, ,n, , , , , Peach Bottom Containment Analysis, Accident Progression Analysis. - Source Term Analysis, Consequence Analysis, NUREG-1150.
"**'{1assified Unclassified a suvet* ossac,ts y Perct hat c..v m oes
L 1 i j NUREG/CR-4551 ' l SAND 86-1309 l Vol. 4, Rev.1, Part 1 i j 3 Evaluation 0:? Severe Accic.ent Risks: Peaca Bottom, Unit 2 Main Report i Prepared by A. C. Payne, R. J. Urceding, ii.-N. Jow, J. C. Ilciton, L N. Smith, A. W. Shiver Sandia National Laboratories 1 Operated by l Sandia Corporation l l Prepared for l U.S. Nuclear Regulatory Commission l l l 9101090481 901231 PDR ADOCK 05000277 . P pan l
. . - . .~. . ~.- . ~ _ . - . .----- . - - .-- -c. .- - - ----- --- --
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NUltliG/ Cit-4551 SAND 86-1309 Vol. 4, Rev.1, l' art 1 Evaluation of Severe Accident Risis: Peach Bottom, Unit 2 Main Report Manuscript Completed: December 1990 Date Published: December 1990 Prepared by A. C. Payne, R. J. Ilrceding, i1.-N. Jow, J. C liciton', L N. Smith 2, A. W. Shiver Sandia National laboratories Albuquerque,NM 87185 Prepared for Division of Systems Research Omcc of Nuclear Regulatory Research U.S. Nuclear Regulatory Conunission Washington, DC 20555 NRC FIN A1228
' Arizona State University,Tempe, AZ PScience Applications International Corp 3 ration, Albuquerque, NM r.- -r,,, # ,.--- - , -
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ABSTRACT In support of the 11uclear Regulatory Commission's assessment of the risk from severe accidents at commercial nuclear power plants in the - U . S . reported in NUREG 1150, . the Severe Accident Risk Reduction Program (SARRP) has completed a revised calculation of the risk to the general public from the operation of the peach Bottom Atomic Power Station, Unit 2. This power plant, located in southeastern Pennsylvania, is operated by the Philadelphia Elcetric Company (PECO). The emphasis in this risk analysis was not on determining a 'so called' point estimate of risk. Rather, it was to determine the distribution of risk, and to determine the fundamental parameters or phenomena whose uncertainties account for the breadth of this distribution. The offsite risk from internal initiating events was found to be quite' low with respect to the safety goals. For internal initiators, the offsite risk is dominated by long term station blackout type accidents (loss of all AC power) in which AC power is never recovered and ATWS (failure to scram) accidents in which injection works untti it fails from high suppression pool temperatures or harsh environments in the reactor building af ter containment venting or failure. The low values for risk can be attributed to the low core damage frequency, the good emergency response, and plant features that reduce the potential source term. The offsite risk from fire initiators is also low with respect to the safety goals but higher than internal events. The fire accidents havs 1cce recovery potential than the interna'ly initiated accidents and have a higher core damage frequency. The fits accidents are dominated by sequences that are equivalent to short and long term station blackouts. The seismic results are even higher than the firo results because of the higher initiating event frequency and significantly reduced recovery potential. The risk is above or close to the safety goal for early fatalities and within a factor of 100 of the latent cancer goal. Given that core damage occurs, it appears quite likely that the containment will fail during the accident . Considerabic uncertainty is associated with the risk estimates produced ,in this analysis. - Safety Internal Fire Seismic Analysis Goni Analysis 6Dnivsts _LLNL EPRI i Individual Early Fatality 5.0E 07 4.7E-11 4.8E 10 1.6E 06 5.3E 08 Mean Risk 0 1 Mt. 2.4E 10 1.7E 09 4.3E 06 1.8E 07 95% Individual Latent Cancer
-Fatality Risk 2.0E 06 4.3E-10 2,4E-09 3.4E 07 1.1E 08 Hean 0 10 Mt. 9.1E-10 8.1E 09 6.4E 07 3.0E 08 95%
iii/iv
1 1-1 l CONTENTS i i EE.C p i ABSTRACT.................. . .............................. 111/1v FOREW0RD................................................... xxix ACRONYMS AND INITIALISMS................................... xxxiii AC KNOW LE DC EM E NT S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . xxxv
SUMMARY
.......-.,............................................ S.1 f
; S.1 Introduction...................................... S,1 J
S.2 Overview of Peach Bot. tom Atomic Power Stat.lon, Unit 2........................................... S.2 S.3 Description of the Integrated Risk Analysis....... S.4 , . S.4 Results of the Accident Frequency Analysis........ S.7 S.4.1 Internal Initiators........................ S7 S.4.2 Fire Initiators............................ S.14 S.4.3 Seismic Initiators........................, S.15 , S.5 Accident Progression Analysis....................,. S.16 . S.S.1 Description of the Accident Progrnssion Analysis.................................. S.16 S.S.2 Posults of the Accident Progression Analysis S.16 S.S.2.1 Internal Initiatora................... S.20 S.5.2.2 Fire Initiators....................... S.21 S.5.2.3 Seismic initiators.................... S.21 S.S.2.4 Clobal Insights....................... S.22 S.5,2.5 Core Damsgo Arrest.................... S.22 S.S.2.6 Early Containment Failure............. S.24 S.6 Source Term Analysis.............................. S.25 S.6.1 Description of'the Source Term Analysis.... S'.25 S.6.2 Results of the Source Term Analysis. . . . . . . . S.26 S.7 Consequenco Analysis...................,.......... S,29 S 7,1 Description of the Consequence Analysis.... S.29 S 7.2 Results of the Consequence Analysis........ S.36 L S 8_ Integrated Risk Analysis.......................... S.63 S.8.1 Determination of Risk.......... ........... S.43 S 8.2 Results of the Risk Analysis............... S.43 S 8.3 Important Contributors to Risk.............. S.56 S.8.4 Important Contributors to Uncertainty in Risk...................... .... .. . S.76 S.9 References.............................. . . . S 77 i , V
-,_...,.---,._.~.a., . _ _ . . . - . - - _ _ _ _ . , . - _ .
Contents (Continued). ) DEC
- 1. INTRODUCTION........................................... 1.1 1.1 Background and Objectives of NUREG.1150. , ....... 1.1 i
1.2 Overview of Peach Bottom Atomic Power Station, Unit 2................................... ....... 1.3 1.3 Chnnges Since the Draft Report........... ........ 1.4 1.4 Structure of the Analysis....................... . 1.8 1.5 Organization of this Report....................... 1.17 1.6 References........................................ 1.17 2.- ANALYSIS OF Tile ACCIDENT PROGRESSION . . . . . . . . . . . . . . . . . . 2.1 ; 2.1 Plant Features Important to Accident Progression at Peach Bottom.................................. 2.1 2.1.1 The Pet ch Bottom Primary Containment l 2.1-Structure................................. 2.1.2 The Reactor Pedestal Cavity......... ...... 2.2 2.1.3 The Containment llent Removal System........ 2.3 l 2.1,4 The Automatic Depressuritation System...... 2.3 2.1.$ The Primary Containment Venting System..... 2.4 2.1.6 The Reactor Building Design............. .. 2.5 . 2.2 Interface with the Core Damage Frequency Analysis.................................. ...... 2.6 2.2.1 Definition of Plant Damage States. ..... .. 2.6 7 2.2.2 Plant Damage State Frequencies............, 2.11 i 2.2.2.1 Internal Plant Damage States.......... 2.22 2.2.2.2 Fire Plant-Damage States.............. 2'.24 2.2.2.3 Seismic Plant Damage-States........... 2.25 j- 2.2.3_ liigh Level Crouping of Plant Damage States. 2.27 2.2.4 _ Variables Sampled in_the Accident Frequency Analysis....... . . . . . . . . 2.29 2.3 Description of the Accident Progression
- Event Tree................... ............... . 2.29 2.3.1 _ overview of the Accident Progression Event Tree................................ -2.29 L :2.3.2 Overview of the Accident Progression Event Tree Quantification....... ., ... . 2.39 2.3.3 -Variables Sampled in the Accident Progression Analysis. ........... 2.49 L
vi i l l j
I Contents (Continued) Eutt 2.4 Description of the Accident Progression Bins.. .. 2.68 l 2.4.1 Description of the Bin Characteristics. ... 2.69-2.4.2 Robinning.................................. 2.78. 2.4.3 Reduced Bins for Presentation.. ...... .... 2.84 2.$ Results of the Accident Progression Analysis. .... 2.87 2.5.1 Results for Internal Initiators............ 2.87 l 2.5.1.1 Results for PDS Group 1 10CA. . . . . . . . 2.87 2.5.1.2 Results for PDS Group 2 i Fast Transient.............. ...... 2.88 ! 2.5.1.3 Results for PDS Group 3 l
- Fast Transient................ .... 2.90 2.5.1.4 Results for PDS Group 4 Fast SBO....- 2.93 2.5.1.5 Results for PDS Group 5 Slow SBO.... 2. 9 5-- ,
2.5.1.6 Results for PDS Group 6 Fast ATWS... 2.97. ! 2.5.1.7 Results for PDS Group 7 - ATWS CV..... 2.97
-2.5.1.8_ Results for PDS Group 8 ATWS CV..... 2.99 2 5.1.9 _Results for PDS Group 9 - ATVS CV..... 2.101 2.5.1.10 Core Damage Arrest, Avoidance of VB. , 2.101 2.5.1'11
. Early Containment Failure............ 2.ls /
2,5.1 12'
. Summary.............................. 2.108 2.5.2 Sensitivity Analyses for Internal Initiators............................. . 2.114 2.5.2.1- No Drywell Shell Mc1tthrough............... 2.114 2.5.3 Results for Fire initiators......... ...... 2.117 2.5.3.1 -Results for PDS Group 1 Fast Transient........,............ 2.117 2.5.3.2 Results for PDS Group 2 Slov SB0.... 2.119 2.5.3.3 -Results for PDS Group 3 - Slow SB0.... 2.119 2 5.3.4 Results for PDS Group 4 Transient CV...... ................ 2.122 2.5.3.5 Coro;Damago Arrest, Avoidance of VB... 2.124 2.5.3.6 Early Containment Failure............. 2.124 2.5,3.7 . Summary.... . . ......... ...... ... . 2.124 2.5.4 ' Sensitivity Analyses for Fire Initiators... 2.128-2.5.4.1 No Drywell Shell Meltthrough.......... 2.128 2.5.5 'Results for Seismic Initiators.............. 2.128 !
2.5.5.1~ Results for PDS Group EQ 1-- FSB-RPV.. 2.131 l 2.5.5.2 Results for_PDS Group EQ 2 : FSB LLOCA........................... 2.131 ? 2.5.5.3 Results for PDS Group EQ 3 FSB LLOCA..........,...... ,,...., 2,134~ 2.5.5.4 Results for PDS Group EQ 4 Slow SB0................ .. . . .. 2.136 vii 1 , . - ~n, c -,n-,-_,.,n,,,,, .,n--,..,,n--,,n,a,-~--..--_...-. -- . _ _ . . - _ --
1 Contents (Continued) l' ate 2.5.5.5 Results for PDS Group EQ 5
- . - Fast SB0........................... 2.136 3
2.5.5.6 Results for PDS Group EQ 6
- FSB IL0CA.......................... 2.139 2.5.5.7 Results for PDS Group EQ 7 FSB 1/SLDCA........................ 2.141 2.5.5.8 Core Damage Arrest. Avoidance of VB. . . 2.141 R2.5.5.9 Early Containment Failure....... ..... 2.141 2.5.5.10, Summary.............................. 2.143 2.5.6 Sensitivity Analyses for Seismic i Initiators................................ 2.143 2.5.6.1 No Drywell Shell Meltthrough...... ... 2.143 2.5.6.2 No CFs at the Start due to RPV
. Support Failures......... ........... 2.15,0 2.6 Insights From the Accident Progression Analysis... 2.151 4-2.7 References........................................ 2.156 ,
-3. RADIOLOGICAL SOURCE TERM ANALYSIS...................... 3.1 3.1 Peach Bottom Features Important to the Source Term Analysis..................................... 3.1
] 3.2' Description of the PBSOR Code..............1...... 3.3 3.2.1 Ove rview of the Parame tric Model . . . . . . . . . . . 3.4 L 3.2.2 Description of PBS0R........... ........... 3.4 3.2.3 ~ Variables. Sampled in the Source Term Analysis...................... 3.9 3.3 Results of Source Term Analysis................... 3.15 3.3.1 Results for Internal Initiators............ 3' 16 3.3.1.1 Results for PDS 1: LOCA............... 3,16 3.3.1.2 Results for PDS 2: Fast Transient. . 3 . 2 0_. 3.3.1.3 Results for PDS 3: Fast Transient..... -3.21 f 3.3.1-4 Results for PDS 4: Fast SB0...........
. 3.21 3.3.1.5 .Results for PDS 5: Slow SBO..... .... 3.26 3.3.1'6. Results f or PDS 6 : ' Fas t ATWS. . . . . . . ... 3.26 3.3.1.7 Results for PDS 7: ATWS rV.. ......... 3.31 3.3.1.8 Results for PDS 8: ATUL CV,........... 3.31 E 3.3.1.9~ Results for PDS 9: A;WS CV............ 3.36 3.3.1.10- Results for Generalized Accident Progression Bins.... ............ .. 3.36 3.3,1,11 Summary............. . . ..... ..... 3.50 3.3.1 12 Sensitivity Analysis Lesults... .....
. 3.50 i-viii
.u - - a ,.w _ ...... _ - - . .a .. - _ . _ . - _ . _ . , _ . _ . . . _ _ . _ _ . _ . _ . _ . _ - _ . . . . - - _
4 Contents (Continued) i , faf.c 3 3.3.2 Results for Fire Initiators............... 3.50 3.3.2.1 Results for PDS 1: Fast Transient..... 3.50 1 3.3.2.2 Results for PDS 2: Slow SB0........... 3.54 3.3.2.3 Results for PDS 3: Slow SBO... .... . 3,54 3.3.2.4 Results for PDS 4: Transient CV....... 3.59 3.3.2.5 Results for Generalized Accident Progression Bins.................. .. 3.59 3.3.2.6 Summary............................... 3.64 3.3.2.7 Sensitivity Analysis Results.......... 3.64 3.3.3 Results for Seismic Initiators............. 3.64 , 3.3.3.1 Results for PDS 1: FSB RPV............ 3.64 3.3.3.2 Results for PDS 2: FSB LLOCA.......... 3.148 c 3.3.3.3 Results for PDS 3: FSB LLOCA.......... 3.148' 3.3.3.4 Results for PDS 4: Slow SB0........... 3.151 3.3.3.5 Results for PDS 5: Fast SB0........... 3,15'1 3.3.3.6 Results for PDS 6: FS B I LOCA . . . . . . . . . . 3.155 4 3.3.3.7 Results for PDS 7: PSB 1/SLOCA........ 3.155 3.3.3.8 Results for Generalized Accident Progression Bins..................... 3.155 '
- 3.3.3.9 Summary............................... 3,158 d
3.3.3.10 Sensitivity Analysis Results......... 3.158 -
~
3.4 Partitioning of the Source Terms for the ' Consequence Analysis............................. 3.162 3.4'1 Results for Internal Initiators............ 3.162 3.4.2 Sensitivity Analysis for Internal Initiators................................ 3.175 , 3.4.3 Results for Fire Initiators................ 3.175 3.4.4 Sensitivity Analysis for Fire Initiators... 3.182 3.4.5 Results for Seismic Initiators: LLNL Ha z a r d Cu rv e . . . . . . . . . . . . . . . . . . . . . . . . . . . . . , 3.182 3.4.6 Sensitivity Analysis for Seismic Initiators: LLNL Hazard Curve . . . . . . . . . . . . . 3'.208 3.4.7 Results for Seismic. Initiators: EPRI
. Hazard curve.............................. 3.208.
3,4.8 Sensitivity Analysis for Seismic Initiators: EPRI Hazard Curve............. 3.248 3.5 Insights from the Source Term Analysis............ 3.248 3.6 References........................................ 3.248
- 4. CONSEQUENCE ANALYSIS............................ . .... 4.1 4.1 Description of the Consequence Analysis.... .. ... 4.1 ix
i I 1 s i Contents (Continued) L&Le l l
*.2 MACCS Input for Peach Bottom...................... 4.2 l l
l 4.3 Result s of MACCS Consequence Calculations. . . ..... 4.6 l- 4.3.1 Resulta for Internal Initiators. . . . . . . . . . . . 4.6-4.3.2 Results for rire Initiators................ 4.S 4.3.3 Results for seismic Initiators: LLNL Hazard l, Curve..................................... 4.17 l 4.3.4 Results for seismic Initiators: EPRI Hotard l 1 Curve..................................... 4.17 I 4.3.5 Results for Seismic Sensitivities......... 4.17 [ 4.3.5.1 No CFs at the Start due to RPV Support -' 1 s Failures: LLNL Hazard curve. ........ 4.17 4.3.5.2 Normal Evacuation Speed for EPRI Low 5 PCA.................................. 4.40-l
- 4.4 ReferenCc8................................ 3 ....... 4.40 i
i 5. R I S K RESU LTS FOR P EACH B0TTOM . . . . . . . . . . . . . . . . . . . . . . . . . . 5.1 5.1 Resul ts o f Risk Calcula tions . . . . . . . . . . . . . . . . . . . . . . 5.2 5.1.1 Risk Results for Internal Initiators. . . . . . 5.2 ; [ 5.1.2 Risk Results for Fire Initiators........... 5.11 i 5.1.3 Risk Results for Seismic Initiators: LLNL Hazard Curve.............................. 5.18 ) $.1.4 Risk Results for Seismic Initiators: EPRI Har.ard Curve...............'............... 5.25 5.2 Contributors to Risk...............,.............. 5.32 5.2.1 Contributors to Risk for Internal- . ! Initiators............................... 5.35 5.2.2 Contributors to Risk for Fire Initiators... 5.46 l 5.2.3 Contributors to Risk for Seismic Initiators -5.53 c 5.3 Contributors to Uncertainty-in Risk............... 5.6f e 5.3.1 Contributors to Uncertainty in-Risk for Internal Initiators....................... 5.67 5.3.2 Contributors to Uncertainty =in risk for-Fire Initiators.....-...................... 5.76 5.3.3 Contributors to Uncertainty in Rir.k for Seismic Initiators........................ 5.84 5.4 Sensitivity Study Results......................... 5.92
-5.4.1 Sensitivity Results For LLNL Seismic Analysis Vith No Early Containment Failure -
5.92 5.4.2 Sensitivity Results For EPRI Seismic Analysis With Normal Evacuation Speed For Lo w P C A . . . . . . . . . . . . . . . . . . ........ ... ... 5.92 s X
- - . _ _ . _ . - - . _ . ~ . _ - - . - _ _ _ . . . - , . - -
_ m.__.- . _ _ _ - . _ _ _ _ _ _ _ _ _ . _ _ . _ . - i-4 1 4 1 1 !. Contents (Concluded) W I Pare l 4
- 6. INSIGHTS AND CONCLUSIONS......... ........... ......... 6.1
! 6.1 Core Damage Arrest................... ............ 6,1 1 6 .1.1 ' Internal Initiators........................ 6.1- ! 6.1.2 Fire Initiators............................ 6.2 6.1.3 Seismic Initiators......................... b.2 ' i 6.2 Early Containment Failure................... ..... 6.2
- 6.2.1 Internal Initiators........................ 6.2
!~ 6.2.2 Fire Initiators............................ 6.3 l 6.2.3 Seismic Initiators......................... 6.4 6.3 Results of Accident Progression Analysis.......... 6.4 j' 6.3.1 Internal Initiators........................ 6.4. 6.3.2 Fire Initiators............................ 6.5
~
6.3.3 Seismic Initiators......................... 6.6' L 6.3.4 Overall Insights for the Accident . ) Progression Analysis...................... 6.6 i' 6.4 Accident Progression Results for Sensitivity Analyses......................................... 6.7 6,4.1 Internal Initiators No trywell Shell .
-t Me1tthrough............................... 6.7 l 6.4.2 Fire Initiators No Drywell Shell i He1tthrough............................... 6.7
.6.4.3 Seismic Initiators......................... 6.8 6.4.3.1 No Drywell Shell Meltthrough........... 6.8 6.4.3.2 No CFs-at the Start due to RPV Support Failures.............................. 6.9 6.5 Source Term Analysis Results...................... 6.9
-6.6 Risk Results...................................... 6,10 6.6.1 Absolute.Value of Risk..................... 6.10.
6.6.1.1 Internal-Initiators.................... 6.10
- 6.6.1.2 Fire _ Initiators........................ 6.11 6.6.1..' Soismic Initiators LLNL Hazard Curve. 6.12 d
6.6,1.4 Seismic Initiators EPRI Hazard Curve. 6.13- .
-6.6.2 Uncertainty in Risk...... ........... ..... 6.13 ]
6.7 Sensitivity Study Results......................... 6.14 6.7.1 Sensitivity Results For LLNL Seismic 1
-Analysis With No Early Containment Failure 6.14 '
6.7.2 Sensitivity twsuits For EPRI Seismic Analysis With Normal Evacuation Speed For Low PGA........ ..................... ... 6,15 6.8 Comparison to Safety Goals. ..................... 6.15- 1 l xi
. . - . _ . _ , _ . _ , _ . _ . ._ .~ - _ .. . _ _ _ . _ ._ _ _ _ __ _ .-_. _ . _ ._ . _ . _ _ _ . _ . _ _ . _ _ . _ . _ __
FICURES F1rure- Eaft 1 Back End Documentation.............................. xxxi { 1 S-1 Cross Section of the Peach Bottom Containment....... S.3 ! S.2 Overview of the Inte6 rated Plant Analysis Performed i in NUREG 1150 of Risk Analysis..................... $.5 S3 Conditional Probability of Collapsed APBs for ) I Collapsed PDS Groups........... ................. S.17 S-4 Conditional Probability of Core . .ge /rrest for Collapsed PDS Groups............ ................. S.18 S5 conditional Probability of Early Containment Failure j for-Collapsed PDS Croups........................... S.19 S6 Peach Bottom: Total Internal Source Term CCDF....... S.27
-S 7 Peach Bottom: Total Fire Source Term CCDF........... S.30
- S.8 Peach Bottom: Total LLNL Seismic Source Term CCDP. . . S.32 S9 Peach Bottom: Total EPRI Seismic Source Term CCDP... S.34' S 10 Consequence CCDFs for Internal Source Term Croups... S.37 S-11 Consequence CCDFs for Fire Source Term Croups....... S.38 S 12 Consequence CCDFs for LLNL Low PGA Source Term Groups ............................................ S.39 S 13 Consequence CCDFs for LLNL High PGA Source Term Croups ............................................ S.40 S 14 Consequence CCDFs for EPRI Low PCA Source Term Croups ...................................... ...., S.41 S 15 Consequence CCDFu for EPRI High PCA Source Term Groups ................ ........................... S.42-S 16 Peach Bottom: Internal Events Risk-CCDFs,........ .. S.44 S+17. Peach Bottom: Fire Risk CCDFs,...................... S.47 S 18 Peach Bottom: LLNL Hazard curve Risk CCDFs.......... S,50 S 19 Peach Bottom: EPRI Hazard Curve Risk CCDFs....,..... S.53 !
S 20 Peach Bottom:-Internal Events Mean Risk Distributions...................................... S.57 S 21 -Peach Bottom:-Fire Mean Risk Distributions.......... S.58 S 22 Peach Bottom: LLNL Hazard Curve Mean Risk 4. Distributions...................................... S.59 l S 23 Peach Bottom: EPRI Hazard Curve Mean Rick -
-Distributions...................................... S.60 S 24- Peach Bottom PDSs for Internal Initiators: Percent ;
Contribution to Risk......................-......... S.68 1
-S 25 - Peach Bottom PDSs for Fire Initiators: Percent Contribution to Risk............................... S.69-S 26- Peach' Bottom PDSs for LLNL Seismic Initiators:
Percent Contribution to Risk.......... ............ S,70-S 27 Peach Bottom PDSs for EPRI Seismic Initiators: 1 Percent Contribution to Risk...........,........... S,71 xiii
- - . . . - - - . - - - - - - ~ - - . . . - - - . - - - - _ _ . ~ - -
J ] J j FIGURES (Continued) j l Firure l'at.c . I s l i S.28 Peach Bottom Summary Accident Progression Bins for l Internal Initiators: Percent Contribution to Risk.. S 72 S 29 Peach Bottom Summary Accident Progression Bins for Fire Initiators: Percent Contribution to Risk...... S.73
- S 30 Peach Bottom Summary Accident Progression Bins for i LLNL Seismic initiators
- Percent Contribution i to Risk............................................ S.74 S-31 Peach Bottom Summary Accident Progression Bins for L EPRI Seismic Initiators: Percent Contribution
! to Risk......................... ... .............. S.75 ' 1: 1 3 1 11 Section of the Peach Bottom Containment............. 1.5 j- -1 2 Overview of Integrated Plant Analysis in NUREC 1150. 1.9 13 Example Risk CCDP..... ............................. 1.16, . 2.5 1 Conditional Probability of Core Damage Arrest for ! Internal PDSs...................................... 2.105 L 2.5 2- Conditional-Probability of Core Damage Arrest for l Collapsed PDS Croups..'............................. 2.106 2.5 3 Conditional Probability of Early containment Failure for Internal PDSs.......................... 2.109 < 2.5 4 Conditional Probability of Early. containment Failure for Collapsed PDS Creups................... 2.110 ,
~2._5 5 -Conditional Probability of Cc'11apsed APBs for 1
Internal PDSs,..................................... 2.111
.2.5 6 Conditional Probability of Collapsed APBs for F Collapsed PDS Croups............,.................. 2.112 2.5 7 Conditional Probability of Core Damage Arrest for Fire PDSs......... ....,........................... 2.125-2.5 8 -Conditional Probability of Early containment Failure for Fire PDSs.............................. 2.126 2.5 9 Conditional Probability of Collapsed APBs for .
Fire PDSs.............. .......... ................ 2.127 2.5 10 Conditionni Probability of Early Containment Failure for Seismic PDSs, LLNL and EPR1,.......... 2.146
;2.5 11 Conditionni Probability of Collapsed APBs for Seismic PDSs, LLNL and EPRI...................,,,, 2.148 3.2 1- Blood Flow Diagram for PBS0R........................ 3,10-3,3 1 Peach Bottont PDS 1 LOCA, Source Term CCDF......., 3.19 3.3 2 Peach. Bottom: PDS 2 Fast Transient, Source Term.
CCDP.............................................. 3.24 3.3 3 i'ench Bottom: PDS 3 Fast Transient, Source Term CCDP................................. .... . .. 3,25 xiv
FIGURES (Continued) Firure l'act 3.3 4 Peach Bottom: PDS 4 Fast SBO, Source Term CCDP. . . 3.29 3.3 5 Peach Bottom: PDS 5 Slow SBO, Source Term CCDF.... 3.30 3.3-6 Peach Bottom: PDS 6 - Fast ATWS, Source Term CCDF. . . 3.34 3.3 7 Peach Bottom: PDS 7 ATWS CV, Source Term CCDF..... 3.35 3.3 8 Peach Bottom: PDS 8 ATWS CV, Source Term CCDF..... 3.39 3.3+9 Peach Bottom: PDS 9 ATWS CV, Source Term CCDF .... 3.40 j' 3.3 10 Peach Bottom: Generalized APB 1, Source Term CCDF: Core Damage, VB>200 Psi, Early Wetwell Failure.. ' M.41 3.3 11 Peach Bottom: Generalized APB 2, Source Term CCDF' Core Damage, VB<200 Psi, Early Wetwell Failure..... 3.43
.3.3 12 Peach Bottom: Generalited APB 3, Source Term CCDF: j Core Damage, VB>200 Psi, Early Drywell-Failure..... 3.44 i 3.3 13 Peach Bottom: Generalized APB 4, Source Term CCDF:
Core Damage, VB<200 Psi, Early Drywell Failure..... 3.45-l 3.3 14 Peach Bottom: Generalized APB 5, Source Term CCDF: 4 Core Damage,_VB Late Wetwell Failure..... ......... 3.46 3.3 15 Peach Bottom: Generalized-APB 6, Source Term CCDF: ! Core Dsmage,.VB, Late Drywell Failure.............. 3.47 j 3,3 16 Peach Bottom: Generalized _APB 7, Source Term CCDF: 1- Core Damage, VB, Venting........................... 3.48 3.3 17 Peach Bottom: Generalized APB 8, Source Term CCDF: Core Damage, VB, No Containment Failure............ 3.49 3.3 18 Peach ~ Bottom: Generalized APB 9, Source Term CCDF: l No Vessel Breach.......................... ........ 3.51
.3.3 19 Peach Bottom: Total Internal Source Term CCDP....... 3.52 3.3 20 Peach Bottom: Fire PDS 1 Fast Transient, Source Term CCDP.......................................... 3.57
-3.3 21 Peach Bottom: Fire PDS 2 - Slow SBO, Source Term CCDF............................................... 3.58 3.3 22 Pec.ch Bottom: Fire PDS 3 Slow SBO, Source Term CCDF............................................... 3.62 3.3-23 Peach Bottom: Fire PDS 4 ' Transient CV,- Source Term CCDP.. ................. ..................... 3.63
'3,3 24 Peach _ Bottom: Fire Generalized APB 1, Source Term CCDF: .
LCore Damage, VB>200 Psi, Early Wetwell Failure..... 3.65 3.3 25 Peach Bottom: Fire Generalized APB 2.LSource Term CCDF: Core Damage, VB<200 Psi, Early Wetwell Failure...... 3.66 3.3+26 Peach Bottom: Fire Generalized APB 3, Source Term CCDF: Core Damage, VB>200-Psi, Early Drywell Failure..... 3.67 3;3 27 Peach Bottom: Fire Generalized APB 4, Source Term CCDP: Core Damage, VB<200 Psi, Early Drywell Failure., .. 3.68
'3.3 28 Peach Bottom:. Fire Generalized APB 5. Source Term CCDF:
1 Core Damage, VB-Late Wetvell Failure............... 3.69 3.3 29 Peach Bottom: Fire Generalized APBL6,_ Source Term CCDF: Core Damage, VB, Late Drywell Failure.............. 3.70 3,3 30 Peach Bottom: Fire' Generalized APB 7, Source Term CCDP: Core Damage, VB. Venting...................... . 3.71 ~ xv
__ _ ____.-_._ - -__m__ 1 1 4 J: FIGURES (Continued) ) a- 1 i i E1EEE Eatt , ! l i 3.3 31 Peacn Bottom: Fire Generalized APB 8, Sour;c Term CCDF: Core Damage, VB, No Containment-Failure. .......... 3,72-l . 3.3 32 Peach Bottom: Fire Generalized AFB 9, Sou: cc Term CCDF. l_ No Vessel Breach........................ .......... 3,73 f 3.3 33 Peach Bottom: Total Fire Source Term CCDF........... 3.74 1 3.3 34 Peach Bottom: LLNL Seismic PDS 1 FSB RT / Hi- PGA, Source Term CCDP....................... ........... 3.76
- i. 3,3 35 Peach Bottom: LLNL Seismic PDS 2 FSB.L OCA Hi j PGA, Source Term CCDF.................. ........... 3.77
[ 3.3 36 Peach Bottom: LUNL Seismic PDS 3 FSB 1sOCA Hi ! PCA~, Source Term CCDP.............................. 3.78 ! 3.3 37 Teach Bottom: LURL Seismic PDS 4 Slow SB0 Hi ! PGA Source Term CCDF.............. .. ............ 3.79 3.3 38 Peach bottom: LLNL Soismic PDS 5 Fast SB0 Hi PCA, . Source Term CCDF. .................... ............ 3.80. ' l 3.3 39 Peach Bottom: LLNL Seismic PDS 6 FSB ILOCA Hi 1 PGA. Source Term CCDF.............................. 3.81 !- 3.3 40 Peach Bottom: LLNL Seismic PDS 7 FSB I/SLOCA Hi -! ! PGA, Source Term CCDF............................... 3.82 1 3.3 4'i Peach bottom: LLNL Seismic Cencralized APB 1 Hi ' PGA, Source Term CCDF: Core Damage, VB>200-Psi,
- Early Wetvell Failure.............................. 3.83 i 3.3 42 reach Bottom
- LLNL Seismic Generalized AFB 2 Hi L PGA, Source Term CCDF: Core Damage, VB<200 Psi,
[ _ Early Wetwell Failure.............................. 3.84 3.3 43 Peach Bottom: LLNL Seismic Generalized APB 3 H1 i PGA, Source Term CCDF: Core Damage, VB>200 Psi, Early Drywell Failure...................,.......... 3.85 3,3-44 Peach Bottom: LLNL Seismic Generalized APB 4 i Hi ! PGA, Source Term CCDF: Core Damage,-VB<200 Psi, Early Drywell Failure.......... . ................... 3.86 1- -3.3 a5 Peach Bottom: LLNL Seismic Generalized APB 5 Hi' j PGA, Source Term CCDF: Core Damage, VB_ Late Wetvell Failure.................................... J 87. 3.3 46 Peach Bottom: LLNL= Seismic Generalized APB 6 Hi h 'PGA, Source Term CCDF: Core Damage,- VB, Late
- Drywell Failure.................................... 3.88-3.3 47 Peach Bottom
- LLNL Seismic Generalized APB 7 Hi
. PGA, Source Term CCCF: Core Damage, VB, Venting.... 3.89 l! 3.3 48 Peach Bottom: 12XL Seismic Generalized- APB 8' Hi PGA, Source Term CCDF: Cote Damage .VB, No - Containment. Failure..,....................... ...... 3.90
- - 3.3449 Peach Bottom
- LLNL Seismic Hi PGA Source Term CCDF 3.91 3.3 50 Peach Bottom: LUNL Seismic PDS 1 FSB RPV - Low PGA, I
Source Term CCDP... .......................... . .. 3.93-3.3 51 Peach Bottom:- LLNL Seismic PDS 2 - FSB LLOCA Low l PCA. Source Term CCDF......... . ..............., . 3.94 l-4 xvi
l l FIGURES (Continued) Figure Par.e 3.3 52 Peach Bottom: LLNL Seismic PDS 3 PSB LLOCA - Low PGA, Source Term CCDP.............................. 3.95 3.3 53 Peach Bottom: LLNL Seismic PDS 4 Slow SB0 Low PGA, Source Term CCDF.............................. 3.96 3.3 54 Peach Bottom: LLNL Seismic PDS 5 Fast SB0 - Low PGA, Source Term CCDP................................... 3.97 3.3-55 Peach Bottom: LLNL Seismic PDS 6 FSB ILOCA Low a PGA, Source Term CCDP.............................. 3.98 3.3 56 Peach Bottom: LLNL Seismic PDS 7 FSB 1/SLOCA Low PGA, Source Term CCDP.............................. 3.99 3,3 57 Peach Bottom: LLNL Seise.ic Cencralized APB 1 - Low PGA, Source Term CCDF: Core Damage, VB>200 Psi, Early Wetwell Failure.............................. 3.100 ' 3.3 58 Peach Bottom: LLNL Seismic Generalized APB 2 Low PGA, Source Term CCDP: Core Damage, VB<200 Psi, . Early Wetwell Failure.............................. 3.101 3.3 59 Peach Bottom: L11Hs Seismic Generalized APB 3 Low PGA, Source Term CCDF: Core Damage, VB>200 Psi, Early Drywell Failure.................. ........... 3.102 3.3 60 Peach Bottom: LLNL Seismic Generalized APB 4 Low l PGA, Source Term CCDF: Core Damage, VB<200 Psi, Early Drywell Failure.............................. 3.103
.3.3 61 Peach Bottom: LLNL Seismic Generalized APB 5 Low PGA, Source Term CCDF: Core Damage, VB Late Wetwell Failure.................................... 3.104 3.3 62 Peach Bottom: LLNL Seismic Generalized APB 6 Low PGA, Source Term CCDF: Core Damage, VB, Late i
. Drywell Failure................................-.... 3.105 ( j 3.3 63 Peach Bottom: LLNL Seismic Generalized APB 7 Low
'PGA, Source Term CCDF: Core Damage, VB, Venting.... 3.106 3.3 64 Peach: Bottom: LLNL Seismic Generalized APB 8 Low PGA,~ Source. Term CCDF: Core Damage, VB, No Containment Failure................................ 3.107 3.3 65 Peach Bottom: LLNL Seismic - Low PGA Source Term CCDP J.108 3.3 66 Peach Bottom: LLNL Seismic - Hi 6 Low PGA Source Term CCDF.......................................... 3.110 3.3 67 Peach Bottom: EPRI Seismic PDS 1 FSB RPV Hi PGA, Source Term CCDP................................... 3.112 3.3 68 Peach Bottom: EPRI Seismic PDS 2 FSB LLOCA - Hi PGA,-Source Term CCDP............... ........... .. 3.113 3,3 69 Peach Bottom: EPRI Seismic PDS 3 FSB LLOCA Hi L _ . _
PGA, Source Term CCDF.............. .............., 3.114 3.3 70 Peach Br. tom: EPRI Seismic PDS 4 Slow SB0 Hi PGA, Source Term CCDP.............................. 3.115 3.3 71 Peach Bottom: EPRI Seismic PDS 5 -Fast SB0 - Hi PGA, Source Term CCDP................................... 3.116 3.3 72 Peach Bottom: EPRI Seismic PDS 6 - FSB ILOCA Hi PGA, Source Term CCDF.................... ......... 3.117 i I xvii I i i
.1
-----.a-.---. - - . - - . - . = , - . - - . . - - --
,.~_.a _ _._ _ ._ _ _ -- . _ _ _ _ ._...__ _ _ _ ._ _ _ _ _ __._ _ _ . _ __. _ 1 FIGURES (Continued) l El r.u r.e ran j 3.3-73 Peach Bottom: EPRI Seismic PDS 7 FSB I/SLOCA Hi POA, Source Term CCDP......................... . . . . 3.118 3.3 74 Peach Bottom: EPRI Sci.mic Generalized APB 1 Hi PGA, Source Term CCDF: Core Damage, VB>200 Psi, Early Wetwell Failure...........-................... 3.119 2 3.3-75 Peach Bottom: EPRI Seismic Generalized APB 2 Hi PGA, Source Term CCDF: Core Damage, VB<200 Psi, Early Wetwe11 Failure.............................. 3.120 3.3 76 Peach Bottom: EPRI Seismic Cencralized APB 3 Hi PGA, Source Term CCDF: Core Damage, VB>200 Psi, Early-Drywell Failure.............................. 3.121 3.3 77 Peach Bottom: EPRI Seismic Ceneralized APB 4 Hi PGA, Source Term CCDF: Core Damage, VB(200 Psi, Early Drywell Failure.............................. 3.122' 3.3 78 Peach Bottom: EPRI Seismic Generalized APB $ Hi
- PGA, Source Term CCDF: Core Damage, VB Late
- Wetvell Failure.................................... 3.123 3.3 79 Peach Bottom: EPRI Seismic Generalized APB 6 Hi PGA, Source Term CCDP: Core Damage, VB, Late Drywell Failure................,................... 3.124 L
3.3 80 Peach Bottom: EPRI Seismic Generalized APB 7 Hi PGA, Source Term CCDF: Core Damage, VB, Venting.... 3.125 3.3 81 Peach Bottom: EPRI Seismic Generalized APB 8 Hi PGA, Source Term CCDF: Core Damage, VB, No Containment Failure......................... . . . . . . 3.126 3.3 82 Peach Bottom: EPRI Seismic Hi PGA Source Term CCDF 3.127 3.3 83-Peach Bottom: EPRI Seismic PDS 1 FSB RPV Low PGA, i Source Term CCDF................................... 3.129 3.3-84' Peach Bottom: EPRI Seismic PDS 2 FSB LLOCA Low PGA, Source Term CCDP.............................. 3.-130 3.3-85 Peach Bottom: EPRI Seismic PDS 3 -FSB LLOCA Low PGA, Source Term CCDF............................... '3.131
-3.3 86 Peach Bottom: EPRI Seismic PDS 4 Slow SB0 Low ..
PGA, Source Term CCDP.............................. 3.132 3.3 87 Peach Bottom: EPRI Seismic PDS S - Fast SB0 - Low PGA, Source Term CCDF................................... 3.133 3.3 88-Peach Bottom: EPRI-Seismic PDS 6 - FSB ILOCA Low . PCA, Source Term CCDP.............................. 3.134 3.3 89 Peach Bottom: EPRI Seismic PDS 7 - FSB-I/SLDCA Low PGA, Source Term CCDF.............................. 3.135 3.3 90 Peach Bottom: EPRI Seismic Generalized AFB l' Low PGA, Source Term CCDF Core Damage, VB>200 Psi, Early Wetvell Failure.............................. 3.136 3.3-91 Peach Bottom: EPRI Seismic Generalized APB 2 Low PGA, Source Term CCDF: Core Damage, VB<200 Psi, Early Wetwell Failure........................... . 3.137 xviii pe gy g' y" "--
-m-%cpTg. 9 9 7e
i i l y FIGURES (Continued) 1 4 1.ituI.c l'unt ! 3.3 92 Peach Bottom: EPRI Seisnde Cencralized APB 3 Low l PGA, Source Term CCDF: Core Damage, VB>200 Psi, {- Early Drywell Fa11ure............................. 3.138
- 3.3+93 Peach Bottom
- EPRI Seismic Generalized AFB 4 Low
[ PGA, Source Term CCDP: Core Damage, VB<200 Psi, Early Drywell railure.............................. 3.139 l 3.3 94-Peach Bottom: EPRI Seismic Generalized APB $ Low ] PCA. Source Term CCDF: Core Damage, VB Late l Wetvell Failure.................................... 3.140 a 3.3 95 Peach Bottom: EPRI Seismic Generalized APB 6 Low l PCA, Source Term CCDF: Core Damage, VB, Late j' Drywell Failure.................................... 3.141 2- 3.3 96 Peach-Bottom: EPRI Seismic Generalized APB 7 Low 4
'PGA, Source Term CCDF: Core Damage, VB, Venting.... 3.142 ;
3.3 97 Peach Bottom: EPRI Seismic Generalized APB B Low - 1 'PGA, Source Term CCDF: Core Damage, VB,- No - {" Containment Failure................................ 3.143 l 3.3 98 Peach Bottom: EPRI Seismic ' Low PGA Source Term CCDF 3.144 ) ! 3.3 99 Peach Bottom: EPRI Seismic 1116 Low PGA Souren ' l Term CCDP.......................................... 3.146
- 3.4 1 Distribution Of Non zero Early and Chronic l!calth
, Effect Weights for Internal Initiators.......... .. 3.168
! 3.4 2 Distribution Of Non zeru Early_and Chronic llealth Effect Weights for Fire Initiators................. 3.181 L 3.4a3 Distribution of Non zero Early and Chronic Health Effect Weights for Seismic Initiators LLNL l- High PGA........................................... 3,206-1 -3,4 4- Distribution Of Non zero Early and Chronic llealth Effect Weights for Internal Initiators - LLNL Low PGA............................................ 3.207 3.4 5 Distribution of Non zero Early and Chronic !!calth , Effect Weights for Seismic Initiators - EPRI- , li i gh P G A . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . 3.246 3.4 6 Distribution Of Non zero Early and Chronic Health Effect Weights for Internal Initiators EPRI Low PCA............................................ 3.247 l . 4.3;1 Consequence CCDFs for Internal Source Torm Groups,._, 4.15 4.3 2 Consequenco CCDFs for Fire Source Term Groups....... 4.16 4.3 3' Consequence CCDFs-for Seismic =LLNL Low PGA Source Term Groups....... ... ..................... 4.36
'4.3 4 Consequence CCDFs.for Seismic LLNL High PGA.........
Source Term Groups......... . . . . . . . . ............... 4.37
. 4.3 5 Consequence CCDFs for Seismic EPRI Low-PGA Source Term Groups................ .............. . 4.38 ,
4.3-6 Consequence CCDFs for Seismic EPRI High PGA......... 1 Source Term Groups..... ......... . . . . . . . . . . . . 4.39 xtx 1
l'ICURES (Concluded) 11Euri' l'EE.t 5.1 1 Peach bottom: Internal Events Risk CCDFs............ 5.3
- 5.1 2 Peach Bottom: Internal Events Mean Risk Distributions 5.9 5.1 3 Peach Bottom: Fire Risk CCDPs....................... 5.12, j 5.1 4 Peach Bottom: Fire Mean Risk Distributions.......... 5.17 ,
5.1 5 Peach Bottom: L1RL llazard Curve Risk CCDFs . . . . . . . . . . 5.19 5.1 6 Peach Bottom: LIRL llatard Curve Mean Risk Distributions 5.24 l 5.1+7 Peach Bottom: EPRI Ilazard Curvo Risk CCDFs . . . . . . . . . . 5.26 5.1 8 Peach Bottom: EPRI liarard Curve Mean Rish Distilbutions 5.31 5.2 1 Peach Bottom Summary PDS Croups for Internal Initiators: Percent Contribution to Risk........... 5.40 5.2 2 Peach Pottom PDSs for Internal Initiators: Percent Contribution to Risk............................... 5.41' 5.2 3 Peach Bottom Summary Accident Progression liins for Internal Initiators: Percent Contribution to Risk.. 5.42 5.2 4 Peach Bottom PD$s for-Pire Initiators: Percent i Contribution to Risk............................... 5.49
- $.2 5 Peach Bottom Summary Accident Progression Bins for i
Fire Initiators: Percent Contribution to Rish...... 5.50 5.2 6 Peach Bottom PDSs for LLNL Seismic Initiators: Percent 1 Contribution to R1sk............................... 5.56 i 3 5.2 7 Peach Bottom PDSs for EPRI Seismic Initiators: Percent ] Con t ribu tion t o ki rk. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.57 5.2*8 Peach Bottom Summary Accident Progression Bins for LLNL Seismic Initiators: Percent Contribution to Rish 5.63 5.2 9 = Peach Bottom Summary Accident Progression Bins for EPRI Seismic Initiators: Percent Contribution to Risk 5.64 1 XX
)
i. j d i 5
- i. TABLES 4
i Table f.uc
-1 NUREG 1150 Analysis Documentation................... xxxii S1 Plant Damage State Frequet.cies All Analyses . . . . . . . S.8 S2 Fractional PDS Contritetions (in percent) to Annual -
Risk at Peach Bottom Due to Internal Initiators.... S.62 S3 Fractional PDS Contributions (in percent) to Annual Risk at Peach Bottom Due to Fire Initiators........ S.64 S4 Frc tional PDS Contributions (in percent) to Annual Risk at Peach Bottum Due to Seismic-Initiators..... S.66 ,- 2.2 1 Peach Bottom-Plant Damage State Characteristics..... 2.7 2.2-2 Plant Damage States for Peach Bottom - Internal, Fire, Seismic...................................... 2.12 2.2-3 Plant Damage State Comparison Internal, Fire, Seismic .......................................... 2.16 2.2 4 Relationship Between PDSs and Collapsed PDS Croups for Internal Events................................ 2.28 j '2.2 5 Variables Sampled in the Accident Frequency
; -Analysis Internal, External...................... 2.30 2,3 1 Questions in the Peach Bottom APET.................. 2.40
-2.3 2. ' Peach Bottom APET Quantification Summary............ 2.48 2,3 3 Variables Sampled in the Accident Progression 1 Analysis........................................... 2.50 2.4 1 Description of Peach Bottom APB Characteristics -
Binner .. ......................................... 2.70 2.4 2 Description of Peach Bottom APB Characteristics - Reb 1nner........................................... 2.80 2,4-3 Description of Reduced APB Characteristics.......... 2.85 2.5 1 Results of the Accident Progression Analysis-for Peach Bottom. Internal Initiators PDS 1 1DCA. . . 2'.89 2.5 2 Results of the Accident Progression Analysis for-Peach Bottom, Internal Initiators'- PDS 2 Fast Transient.......................................... 2.91 2.5 3 Results of the Accident Progression Analysis for Peach Bottom Internal Initiators - PDS 3 Fast Transient, ........................................ 2.92 2.5 4 Results of the Accident Progression Analysis for Peach Bottom,-Internal Initiators - PDS 4 -Fast SB0................................................ 2.94 l 2,5 5 1Results of the Accident Progression Analysis for i Peach Bottom. Internal Initiators - PDS 5 S1,w- ' SBO... ....................................... .. 2.96 xxt
TABLES (Continued) i I Inhlc far.n 2.5 6 Results of the Accident Progression Analysis for 4 Peach Bottom, Internal Initiators PDS 6 Fast I l ATWS............................................... 2.98 2.5 7 Results of the Accident Progression Analysis for Peach Bottom. Internal Initiatort - PDS 7 ATWS CV 2.100 2.5 8 Results of the Accident Progression Analysis for Peach Bottom,_ Internal Initiators PDS 8 ATWS CV 2.102 2.5 9 Results of the Accident Progression Analysis for Peach Bottom, Internal Initiators PDS 9 ATWS CV 2.103 2.5 10 Peach Bottom Internal PDS - Containment Failure at or Before Vessel Breach (Early) Sensitivity case: No Drywell Moltthrough. .................... 2.115 2.5 11 Peach Bottom Internal PDS Containment Failure
-at or Before Vessel Breach (Early) Base Case:
_ Drywell Heltthrough A11 owed........................ 2.116 ^ 2.5 12 Results of the Accident Progression Analysis for Peach Bottom, Fire Initiators PDS 1 Fast Transient.......................................... 2.118 2.5 13 Results of the Accident Progression Analysis for Peach Bottom, Fire Initiators PDS 2 Slow SBO... 2.120 2.5 14 Results of the Accident Progression Analysis for Peach Bottom, Fire Initiators
- PDS 3 - Slow SBO... 2,121 2.5-15 Results of the Accident Progression Analysis for Peach Bottom, Fire Initiators PDS 4 Transient CV 2.123 2.5 16 Peach Bottom Fire PDS , Containment Failure at or Before Vessel Breach (Early) Sensitivity case: No Drywell Me1tthrough................................ 2.129 2.5 17 Feach Bottom Fire PDS - Containment Failure at or Before Vessel Breach (Early) Base Case: Drywell Meltthrough A11 owed................................ 2.130 2.5 18 Results of the Acciderr 'ogression Analysis for Peach Bottom, Seis.r: a intors PDS 1 - P:3B RPV..-. ,,....... ... . ......................... 2.132 2.5-19 Results'of the Accident L.7gression Analysis for .
Peach Bottom, Seismic Initiators - PDS 2 - FSB LL0CA............................................. 2.133 l
- 2.5-20 Results of the Accident Progression Analysis for ;
Peach Bottom. Seismic Initiators PDS 3 FSB
.LLOCA.............................................. 2.135
- 2.5 21 Results of the Accident Progression Analysis for Peach Bottom, Seismic Initiators PDS 4 Slow SB0................................................ 2.137
- 2.5 22 Results of the Accident Progression Analysis for Peach Bottom Seismic initiators 1- PDS Fast SBO... ............................................ 2.138 2.5 23 Results of the Accident Progression Analysis for Peach Bottom.. Seismic Initiators PDS 6 - FSB ILOCA.........................,. ... ....... ...... 2.140 xxii
- . . ,.. -4.- _ _ . _ - , - _ _ _ - . - - . . _ _ _ . . . _ _ . - _ . . _ . . - . _ - - _ _ . . .-.--. .,_.- . -_ _ _ - . . . _ . , _ - - . -
L -
- . - - - - - - = -
4 4
. TABLES (Continued)-
Table Pars 52iSi 24fResults of the Accident Progression Analysis for-Peach Bottom,1 Seismic' Initiators -PDS 7 FSB -
- I /S LDCA . . .. . . .. . . . . . . . . x ............................. 2.142 2.5 25 Peach Bottom Seismic PDS Containment Failure at or -
ep - Before Vessel Breach (Early) Sensitivity case: LNo
- D ryw e l l- _ M e 1 t th r o u gh . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.144-2.5 26 Peach Bottom Seismic PDS Containment Failuro at or Before Vesselfbreach (Early)= Bass Case: Drywell
_ _ Mel t thtre gh 'A11 owe d . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.145
..2.5 272 Peach Bottom Seismic PDS Containment Failure at or Before Vessel Breach (Early) Ccmparison: Ini tial g0 - Containment Failure vs No Initial Containment Failure............................................ 2.152 (6 , . 2.5-28 Results of the A0cident Progression. Analysis for Peach-Bottom, Seismic Initiators PDS 1 - FSB
__RPV, No Initial-Containment Failure...... .......... 2.153
'2.5<29_Results of.the Accident' Progression Analysis-for l 4 - - Peach- Bottom,: Seismic Initiators - PDS 2 - FSB LLOCA,-No Initial Containment Failure..... ...... . 2.154-g' -2,5-30 Results-of the Accident Progression Analysis for Peach. Bottom,: Seismic-Initiators --PDS 3 FSB LIDCA, No Initial Containment Failure. , ........... 2.155 3 1352I15? Isotopes in Each Radionuclide Release Class......=... 3.5
'3.2-2_-Variables Sampled in the Source Term-Analysis....... 3.11 -
131. 1 : Mean Source Terms for Peach Bottom Internal n -Initiators - PDS 1 L0CA..-..... ..................
. 3.17 3EP- 3.3-2 Mean-Source Terms:for. Peach Bottom Internal-
= Initiators-- PDS 2. ; Fast-Transient................ 3.22
;3.3-3' Mean Source Terms'for Peach Bottom Internal
' Initiators -LPDS 3 Fast Transient.-........... .. 3.23
~3,3-4 Moan Source. Terms for-Peach Bottom Internal .
P LInitiators --PDS 4 -1 Fast SB0...................... 3.27 4(7 3.3 Mean Source Terms.for: Peach-Bottom-Internal "i Initiators - PDS 5 Plow SB0...................... _3.28_ 3.3 6 Mean-Source Terms for. reach Bottom' Internal
, , _ _ Initiators - PDS 6 Fast ATVS..................... 3.32 Lb i l 3.? ;7 Mean Source-Terms for Peach Bottom Internal
$[ .
In i t i a t o r s - PDS 7 - ATW S (31. . . . . . . . . . . . . . . . . . . . . . . 3.33
?3.3;S: Mean Source Terms for Peach Bottom' Internal
, Initiators -PDS 8 ATWS CV....................... 3.37 3.3-9, Mean Source Terms for Peach Bottom Internal -
", Initiators PDS 9 LATWS CV............... ....... 3.38-3.3 10~Mean Source Terms-for Peach Bottom Fire a Initiators PDS Fast Transient................ 3.55 a ,
xxiii at ,
} _.
TABLES (Continued) ; Tnble faf,c 3.3 11 Mean Source Terms for Peach Bottom Fire Initiators -PDS 2 Slow SB0...................... 3.56-3.3 12 Mean Source Terms for Peach Bottom Fire Slow SB0...................... 3.60 Initiators PDS 3 13.3 13 Mean Source Terms for Peach Bottom Fire Initiators PDS'4 - Transient CV.................. 3.61 3.3 14 Mean Source Terms for Peach Bottom Seismic Initiators - PDS 1 - FSB RPV....................... 3.149 3,3 15 Mean Source Terms for-Peach Bottom Sciamic
-Initiators - PDS 2 FSB LL0CA..................... 3.150 3.3 16 Hean Source Terms for Peach Bottom seismic Initiators - PDS 3 - FSB LL0CA................... 3.152 3.3-17 Mean Source Terms for Peach Bottom Seismic Initiators - PDS 4 - Slow SB0................ ..... 3,153'
_ 3.3 18 Mean Source Terms for Peach _ Bottom Seismic Initiators PDS 5 - Fast SB0...................... 3.154
'3.3 19 Mean Source Terms for Peach Bottom Seismic Initiators . PDS 6 - FSB IL0CA..................... 3,156 3.3 20'Mean Source Terms for Peach Bottom Seismic Ini tia tors - PDS 7 FSB I/SL00A. . . . . . . . . . . . . . . . . . . 3.157 3.3 21 Mean Source Terms for Peach Bottom Seismic Initiators - PDS 1 - FSB RFV No CF at T-0. . . . . . . . 3.159 t 3.3 22-Mean Source Terms for Peach Bottom Seismic Initiators FFS 2 - FSB LLOCA No CF at T-0....., 3.160
-3.3 23-Mean Source Terms for Peach Bottom Seismic Initiators' .PDS 3 - FSB LLOCA - No CF at T-0...... 3.161 3.4-1 Summary of Early and Chronic Health Effect Weights for Internal Initiators............................ 3.164 13.4-2 Distribution of Source Terms with Non zero Early Fatality and Chronic Fatality Weights for Internal
, Initiators........... ............................. 3.165 3,4 3 ' Distribution of Source. Terms with Zero Ecrly ,
-Fatality Weight and Non zero Chronic Fatality Weight _-
-fo6 ernal Initiators. ......................... 3.170-3.4-4. Mean Sc.rce Terms Resulting from Partitioning for p -Intertal Initiators - Peach Bottom................. 3.171 l- ,3.4 5 Summary of Early and Chronic Health- Effect Weights 4 for Fire Initiators................................ 3.176 3.4-6 -Distribution of Source Terms with Non zero Early Fatality and Chronic Fatality Weights for' Fire
-Initiators.......... ......................,. ..... 3.177
-3.4 7 Distribution of Source Terms with Zero Early Fatality Weight and Non-zero Chronic Fatality Weight for' Fire Initiators............................... 3.180 3.4 8- Mean Source Terms Resulting from Partitioning for Fire Initiators. Peach Bottom............... . ... 3.183 3.4 9 Summary of Early and Chronic Health Effect Weights for Seismic Initiators LLNL High PCA. ......... . 3.188 xxiv s
TABLES (Continued) Inhlt Pare 3.4 10 Summary of Early and Chronic Health Effect Weights for Seismic Initiutors L1NL Low PCA. . . . . . . . . . . . . . 3,189 3.4-11 Distribution of Source Terns with Non zero Early Fatality and Chronic Fatality Weights for Seismic Initiators - LLNL High PGA..................... 3.190 3,4 12 Distribution of Source Terms with Zero Early Fatality Weight and Non zero Chronic Fatality Weight for Seismic Initiators - LLNL High PGA.,,,...... 3.193 3.4 13 Distribution of Source Terms with Non zero Early Fatality and Chronic Fatality Weights for Seismic Initiators - LLNL Low ICA....................... 3.194 3.4 14 Distribution of Source Terms with Zero Early Fatality Weight and Non zero Chronic Fatality Weight for Seismic Initiators . LLNL Low PGA . . . . . . .... 3.197'
,. 3.4 15 Mean Source Terms Resulting from Partitioning for Seismic Initiators - LLNL High PGA. . . . . . ......... 3,198 3.4-16 Mean Source Terms Resulting from Partitioning for Seismic Initiators - LLNL Low PGA,... ........ .. 3.202 3.4 17-Summary of Early and Chronic Health Effect Weights for Seismic Initiators LLNL High PGA No CF atT-0........................................ ... 3,209 3.4-18 Summary of Early and Chronic Health Effect Weights for Seismic Initiators LLNL - Low PGA No GF at T-0................ . ..... ...................... 3.210 3.4-19 Distribucion of Source Terms with Non-zero Early Fatality and Chronic Fatality Weights for Seismic Initiators -- LLNL High PGA No CF at T-0...... 3.211 3.4-20 Distribution of Source Terms with Zora Early Fatality Weight and Non-zero Chronic Fatality Weight for Seismic Initiators -- LLNL - High PCA No CF at T-0 ...........,..........................,..... 3.214 3.4-21 Distribution of Source Terms with Non zero Early Fatality and Chronic Fatality Welfhts for Seismic Ini tiators - - LLNL - Low PGA - No C f a t T-0. . . . . . . . 3',215 3.4-22 Distribution of Source Terms with Zero Early Fatality Weight and Non zero Chronic Fatality Weight for Seismic Initiators - LLNL Low PGA No CF at T-0 ........................... ................ 3.218 3,4 23 Mean Source Terms Resulting from Partitioning for Seismic Initiators - LLNL High PGA No CF at T-0... ..........................., ........,...... 3.219 3,4-24 Mean Source Terms Resulting from Partitioning for Seismic Initiators - LLNL - Low PGA - No CF at T-0............., ........... .............. ... ., 3.223 3.4-25 Summary of Early and Chronic Health Effect Weights for Selsmic Initiators - EPRI Hi gh PGA. . . . . . ,,,,, 3.228 3.4-26 Summary of Early and Chronic Health Effect Weights f r.r Seismic Initiators EPRI Low PGA. . . , .. .... . 3,229 s
XXV
-TABLES (Continued)'
Table Engg 3.4-27 Distribution of Source Terms _with Non zero Early Fatality and Chronic Fatality Weights for Seismic -i Initiators .- EPRI - High PGA....................., 3.230 3.4 28 Distribution of Source Terms with Zero Early Fatality Weight and Non zero Chronic Fatality Weight for Seismic Initiators - EPRI - High PGA.., ..... 3.233 3.4-29 Distribution of Source Terms with Non zero Early Fatality and Chronic Fatality Weights for Seismic Ini ti a to rs EPRI Low PCA . . . . . . . . . . . . . . . . . . . . .. 3.234 3.4 30 Distribution of Source Terms with Zero Early Fatality Weight and Non-zero Chronic Fatality Weight for Seismic Initiaturs -- EPRI - Low PGA........... 3.237 3.4-31 Mean Source Terms Resulting from Partitioning for Seismic Initiators -- EPRI High PGA............. . 3.238' 3.4-32 Mean Source Terms Resulting from Partitioning for EPRI Low PGA. . . . . . . . . . . . . . . . . 3.24'2 Seismic Initiators - i 4.1-1 Definition of Consequence Analysis Results.......... 4.3 4.2 1 Site Specific Input Data for Peach Bottom RACCS Calculations...... .......................... .....
~
4.4 4.2-2 Population at Different Radii From the Plant........ 4.5 4.2-3 ShieldinS Factors used for Peach Bottom MACCS Calculations....................................... 4.7
- 4. 3 -l' Mean Consequence Results for Internal Initiators.... 4.9 4.3-2 Mean Consequence Results'for Fire Initiators........ 4.12 4.3 3 Mean Consequence Results for Seismic Initiators LLNL Hazard Distribution - High PGA. . . . . . . . . . . . . . . . 4.18 4.3-4 Mean Consequence-Results for Seismic Initiators .
LLNL Hazard Distribution - Low PGA................. 4.21 ! 4,3-5 Mean Consequence Results for Seismic Initiators EPRI Hazard Distribution - High PGA................ 4'. 2 4, 4.3 Mean Consequence Results for Seismic Initiators EPRI Hazard Distribution Low PGA................ 4.27 4.3-7' Mean Consequence Results for Seismic 1 Initiators LLNL Hazard Distribution High PGA - No CF-at T-0, 4.30 4.3-8 _Mean Consequence Results for Seismic Initiators LLNL Hazard Distribution Low PGA - No aCF at T-0. . 4.33 4i3-9 -Mean Consequence Results-for Seismic Initiators EPRI Hazard Distribution - Low PGA - Normal Evacuation......................................... 4.41 5,1 1 Distributions for Annual Risk at Peach Bottom due to Internal Initiators. .......... ............. .. . 5.8 5.1-2 Distributions for Annual Risk at Peach Bottom due to Fire Initiators... ..... .. .. ....... .... .... 5.16 xxvi
1 i 1 TABLES (Continued) Tobig her 5.1 3 Distributions for Annual Risk at Peach Bottom due to Seismic Initiators - LLNL Hazard Distributions. . . . 5.23-5.1-4 Distributions for Annual Risk at Peach Bottom due to Seismic Initiators - EPRI Hazard Distributions..... 5. 1 5.1-5 Distributions for Annual Risk at Peach Bottom due to Seismic Initiators............... .. .............. 5.33 5.2 1 Fractional Summary PDS Contributions (in percent) to Annual Risk at Peach Bottom due to Internal Initiators.......... ......................... .... 5.36 5.2-2 Fractional PDS Contributions (in percent) to Annual Risk at Peach Bottom due to Internal Initiators.... 5.38 5.2-3 Fractional APB Contributions (in percent) to Annual Risk at Peach Bottom due to Internal Initiators.... 5.39 5.2 4. Fractional PDS Contributions (in percent) to Annual Risk at Peach Bottom due to Fire Initiators........ 5.47 5.2 5 Fractional AFB Contributions (in percent) to Annual Risk at Peach Bottom due to Fire Initiators........ 5.48 5.2 6 Fractional PDS Contributions (in percent) to Annual Risk at Peach Bottom due to Seismic Initiators - LIEL Haza rd Dis t ributions . . . . . . . . . . . . . . . . . . . . . . . . . . 5.54 5,2-7 Fractional PDS Contributions (in percent) to Annual Risk at Peach Bottom due to Seismic Initiators - EPRI Hazard Distributions.......................... 5.55 5.2 8 Fractional APB Contributions (in percent) to Annual Risk at Peach Bottom due to Seismic Initiators... . 5.59 5.2 9 Fractional PDS Contributions (in percent) to Annual Risk at Peach Bottom due to Seismic Initiators.... 5.60 5.2-10 Fractional Contributions -(in percent) from Hi and Low PDSs to Annual Risk at Peach Bottom due to Seismic Initiators................................, 5.62 5.3 1 Regression Results for Peach Bottom Intcenal Initiators............-............................. S 6% 5.3-2 Regression Results for Peach Bottom Fire Initiators............................ ............ 5.77 5.3-3 Regression Results for Peach Bottom Seismic Initiators........................................ 5.85 5.4 1 Distributions for Annual Risk at Peach Bottom due to Seismic Initiators - LLNL Hazard Distributions - No CF at T-0....................................... ..
- 5.93 5.4-2 Fractional PDS Contributions (in percent) to Annual Risk at Peach Bottom due to Seismic Initiators --
LLNL Hazard Distributions - No CF at T-0..... .... 5.94 5.4-3 Distributions for' Annual Risk at Peach Bottom due to Seismic Initiators - EPRI Hazard Distributions - Normal Evacuation for Low PCA. .. ..... ... . ... 5.95 l. i I xxvil
- .. -. ~ ..-. .- . .. -. .-. _ ~ . - - ... . - . . . - - . - . - . . . . _ . . . - -
l TABLES (Concluded) Tabig I' age
' 5.4 4 Fractional PDS Contributions-(in percent) to Annual-Risk at Peach Bottom due to seismic Initiators -
EPRI Ilazard Distributions Normal Evacuation at Low PGA.. ......................................... 5.96.
- 6 1- Comparison with Safety Goals........................ 6.16 i
i p
'I l
)
l i l xxviii
- + - , , , , g-w- 4yn.,,,,.y- 4, nww e ~-m +
4 FOREWORD This is one of numerous documents that. support the preparation of the final NUREG 1150 : document by the NRC Office of Nuclear Regulatory Research. Figure 1 illustrates the documentation of the accident progression, source term, consequence, and risk analyses, The direct supporting documents for the first draft of NUREG 1150 and for the revised draf t of NUREG 1150 are given in Table 1. They were produced by the three interfacing programs that performed the work: the Accident Sequence Evaluation Program (ASEP), the Severe Accident Risk Reduction Program (SARRP), and the PRA Phenomenology and Risk Uncertainty Evaluation Program (PRUEP). The 7. ion volumes were written by Brookhaven National Laboratory and Idaho National Engineering Laboratory. The Accident Frequency Analysis, and its constituent analyses, such as the Systems Analysis and . the Initiating Event ' Analysis , are reported in-NUREC/CR 4550. Ori Sinally, NUREG/CR 4550 was published wi thou t" the designation " Draft for Comment." Thus, the current revision of NUREG/CR-4550 is designated Revision 1. The label Revision 1 is used consist'ently on all volumes, including Volume 2 which was not part of the original documentation. .NUREG/CR-4551 was originally _ published as a " Draft for Comment", While the current version could have been issued without a revision indication, all volumes of NUREG/CR-4551 have been designated Revision 1 for consistency with NUREG/CR 4550. The material contained 'in NUREG/CR 4700 in the original documentation is now contained in NUREG/CR 4551; NUREG/CR-4700 is not being revised. The contents of the volumes in both NUREG/CR 4550 and NUREC/CR 4551 have been altered, In both documents now, Volume 1 describes the methods utilized in the analyses, "alume 2 presents the clicitation of expert judgment, Volume 4 concerns the analyses for Peach Bottom and so on. In addition to NUREG/CR 4550 and NUREG/CR 4551, there are several other reports published in association with NUREG 1150 that explain the methods used, document'the' computer codes that implement these methods, or present the results of calculations performed to obtain information specifically for this project. These reports include: . NUREG/CR-5032, SANDB7 2428, "Modeling Time to Recovery and Initiating Event Frequency for Loss of Off-site Power Incidents at Nuclear Power Plants," R. L. Iman and S. C. llo ra , Sandia National Laboratories, Albuquerque, NM, January 1988. NUREG/CR-4840, SAND 88 3102, " Procedures for the External Event Core Damage Frequency Analyses for NUREG-1150," M. P. Bohn and J. A. Lambright, Sandia National Laboratories, Albuquerque, _ NM, December 1990. I o XXIX l
NUREC/CR 5174, SAND 88 1607,- -J , M. Griesmeyer _ and L. N. Smith, "A Reference Manual-- for the Event Progression and Analysis Code (EVNTRE) " Sandia -National Laboratories,' Albuquerque, NM, September 1989. NUREC/CR-5380t,-SAND 88 2988, S. J. liiggins, "A User's Manual fort the-Post Processing Program PSTEVNT," Sandla National Laboratories , Albuquerque, NM, November 1989.
- NUREG/CR 5360, SAND 89 0943, ll.-N. Jow, W. B, Murfin, and J- . D.
Johnson, "XSOR Codes User's- Manual," Sandia National Laboratories, Albuquerque, NM, 1991. NUREG/CR-4624,- BM1 2139, R. S. Denning et al., "Radionuclide Release Calculations for Selected Severe Accident Scenarios," Volumes I-V, Batte11e's Columbus' Division, Columbus,-Oll, 1986. NUREG/CR-5062, BMI 2160, M. T. Leonard et al., " Supplemental Radionuclide - Release Calculations for Selected Severe Accident ' Scenarios,"- Battelle Columbus Division, Columbus, 011, 1988, NUREG/CR 5331, SAND 89 0072, S, E. Dingman -et al., "MELCOR Analyses - for Accident Progression . Issues," Sandia National Laboratories, Albuquerque, NM,-1990. NUREG/CR 5253, SAND 88-2940, R. L. Iman, J, C, llelton, and J. D. Johnson, " PARTITION: A Program for - De fining the Source _ Term / Consequence Analysis Interfaces in the NUREG 1150 Probabilistic i Risk . Assessments , User's Guide," Sandia National Laboratories, Albuquerque, NM, May 1990. 1 XXX
r SUPPORT DOCUMENTS TO NUREG-1150 ; t
/ I NUREG-1150
/(NRC Staff} >
p EVALUATION OF SEVERE ACCIDENT RISKS NUREGICR-4551 005 "^'0"'"*'*"*"" Y "*C" * " *H > NUREG/CR-4550 "" vet. s vos2 5'"."3 vd vos.4 ""voi. s 8"*""s.C vo s " v"oi. 7
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i } s f 1 . NUREG-1150 Analysis Documentation ( . Table 1. l } Orieinal Documentation
.NUREG/CR-4551 NUREG/CR-4700 NUREG/CR-4550' Containment Event Analysis Analysis of Core Damase Prequency Evaluation of Severe Accident Risks -
From Internal' Events. and the Potential for Risk Reduction ..for Potential Severe Acciden;s Vol. 1 Surry Unit 1 'Vol. 1 Surry Unit 1 Vol. 1 Methodology Sunrnary (Not Published) 2 Sequoyah Unit 1_ '2 Sequoyah Unit 1-
.2 .
3 Surry Unit 1 3' Peach Bottom Unit 2 3 ' Peach Bottom Unit 2 4 Peach Bottom Unit 2 4 Grand Golf Unit 1 4 Grand Gulf Unit 1 ,
'5 Sequoyah Unit 1 6 . Grand Gulf Unit 1 j 7 Zion Unit 1 x Revised Documentation NUREG/CR-4550. . Fev.1, Analysis of Core Damage Frequency NUREG/CR-4551, Rev. 1. Eval. of Severe Accident Risks
[ Vol. 1 Part 1. Methodology; Part 2. Appendices
' vol. I Methodology 2 Part 1 Erport Judsment Elicit Expert Panel 2 Part 1 In-Vessel Issues ,
f Part 2 Expert Judgment Elicit. Project Staff Part 2 Containment Loads and MCCI Issues Part 3 Structural Issues Part & Source Term Issues I Part S Supporting Calculations Part 6 Other issues Part 7 MACCS Input Part 1'Surry Unit 1 Internal Events' 3 Part 1 Surry Analysis and Results 3 Part 2 Surry Unit 1 Internal Events App. Part 2 Surry Appendices Fart 3 Surry Erternal Events-4 Part 1 Peach Bottom Analysis and Results
'4 Part 1 Peach Bottom Unit 2 Internal Events Part 2 Peach Bottom Appendices i Part 2 Peach Bottom Unit 2 Int. Events App. [
Part 3 ' Peach Bottom Unit 2 External Events 5 Part 1 Sequoyah Analysis and Results }
". Part 1 Sequoyah Unit 1 Internal. Events Part 2 _ Sequoyah Unit 1 Internal Events App. Part 2 Sequoyah Appendices ,
6 Part 1 Grand Gulf Analysis and Results 6 Part 1- Grand Gulf Unit 1 Internal Events' Part 2 Grand Gulf Appendices Part 2 Grand Gnif Unit 1 Internal Events App. 7 Part 1 Zion Analysis and Results . 7 Zion Unit 1 Internal Events [ Fart 2 Appendices
'[
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ACRONYMS AND INITIALISMS ADS automatic depressurization' system
- APB accident progression bin APET accident progression event tree .
ASEP accident sequence evaluation program ATVS anticipated transient without scram BAP bottom of active fuel BNL Brookhaven National Laboratory BWR boiling water reactor CCF- common cause failure CCI core concrete interaction CCDP complementary-cumulative distribution function CDF cumulative distribution function CF containment failure CFW chronic fatality weight CHR containment heat removal CRD- control rod drive system CSS containment spray system CST condensate storage tank CST condensate storage tank DCH- direct containment heating DG diesel generator ECCS emergency core cooling system EP early fatality EFW - carly fatality weight E0P emergency operating procedures EPRI Electric Power Research Institute CVSE -ex vessel' steam explosion FSAR final safety analysis report HEP human error probability
=HPCS high pressure core spray system HPME high pressure melt ejection
-HRA -human reliability analysis
'INEL -Idaho National Engineering Laboratory IVSE in-vessel steam explosion LCF latent cancer fatalities LHS iatin hypercube-sampling
.LOCA- loss of-coolant accident xxxiii l
_ . _ . ~, . . .. - _ . ~ - ,-. IASP loss of offsite power - LPCI low- pressure coolant injection system
- LPCS- low-pressure core 1 spray system MCDP mean core _ damage frequency MDP motor driven pump
~MOV motor operated valve MSIV main stream isolation valve Nuclear Regulatory Commission
~
NRC PCS power conversion system PDS plant damage state PRA probabilistic risk analysis PWR pressurized water. reactor RCIC reactor core isolation- cooling system RCS reactor coolant-system RHR residual heat removal RPS reactor protection system
-RPV reactor pressure-vessel RSS Reactor Safety Study SB station blackout SBO- station blackout SERC . steam explosion review group SLC . standby liquid control SNL Sandia National Laboratories SORV--stuck open relief valve
=
SPC- suppression pool cooling system SRV safety relief valve TAFf . top of active fuel-TDP turbine driven pump TEMAC_ Top Event Matrix Analysis Code VB vessel breach ' xxxiv
_. . . . - . . - ~ - - - - -- - - -. . . . . . - - -, ACKNOWLEDCEMENTS Ve wish to thank the many people who worked in various capacities to support this analysis. E. Corham Bergeron (SNL), who was the program manager and provided many helpful suggestions in methods and techniques. F. T. Ilarper (SNL), who provided the day to day leadership of the project and worked wherever help was needed. C. N. Amos (SAIC, formerly Technadyne) for his many helpful suggestions which smoothed the transition from the first draft analysis to the current revision. The consequence analysis team of J. L. Sprung (SNL), J. D. Johnson (Applied Physics, formerly SAIC), and D. I. Chanin (Technadyne) who performed the MACCS analysis. R. L. Iman (SNL) for his work in designing the overall computational strategy and the codes to be used in implementing that strategy and J. D. Johnson (Applied Physics, formerly SAIC) for constructing some of those codes. S. J. Higgins (SNL) and A. H. Bahena (Comision Nacional De Seguridad Nuclear Y Salvaguardias, Mexico)~ for performing detailed quality control reviews of the various codes, t E. Dingman (SNL) for the mary computer calculations that she performed in support of this analysis and - for her help in suggesting ways to model various aspects of the accident progression in the APET, C. J. Shaffer (SEA) for the MELCOR caluulations he performed for some Peach Bottom accident progressions. We wish to thank the other plant analysts, T. D. Brown (SNL), and G. J. Gregory (SNL), for their many helpful suggestions and the work that all the plant ' analysts performed together in order to make sure that everyone succeeded in this effort. We wish to thank the Quality Control team (K. D. Ber6eron (SNL), G. J. Boyd (SAROS), D. R. Bradley (SNL), R. S. Denning (BMI), S. E. Dingman (SNL), J . E. Kelly (SNL) , D. M. Kunsman (SNL) , J . Lehner (BNL), S. R. Lewis (SAROS), and D. W. Pyatt (NRC)) for their efforts in reviewing the various parts of
-the analysis 'and their constructive suggestions for improving the' overall quality of the analysis. In particular, we would like to thank . them for their review of the Peach Bottom APET and its user functions.
Wo wish - to thank the- Level I Peach Bottom analysts for their w rk. In particular, - A. M. Kolaczkowski (SAIC), W. R. Cramond (SNL), and T. W. Wheeler (SNL) for their efforts to make the interface between the Level I internal events analysis and Level II analyses work efficiently. Also, M. P. Bohn'ana J. A. Lambright for their work on the - seismic and fire interface. Finally, we wish to thank the NRC for their funding and support of this - project. In particular, we wish to thank M. A. Cunningham. J. A. Murphy, and P. K. - Niyogi for - their program and management support. 4 XXXV l , -
, . - .- _ .. - . . . . - - .- - . .~ . .
SUMMARY
S.1 Introduction The United States Nuclear Regulatory Commission (NRC) has recently completed a maj or study to provide a current characterization of severo accident risks from-light water reactors (LVRs). This characterization is derived from integrated risk analyses of fivo plants. The summary of this study, NUREG-1150,1 has been issued as a second draft for comment. The risk assessments on which NUREG-1150 is based can generally be characterized as consisting of four ' analysis steps, an integration step, and an uncertainty analysis step:
- 1. Accident frequency analysis: the determination of the like-lihood and nature of accidents that result in the onset of core damage.
- 2. Accident progression analysis: an investigation of the core damage process, both within the reactor vessel before it fails and in the containment afterwards, and the resultant impact on the containment.
- 3. Source term analysis: an estimation of the radionuclide transport within the reactor coolant system and the containment, and the magnitude of the subsequent releases to the environment.
- 4. Consequence analysis; the calculation of the offsite consequences, primarily in terms of health effects in the general population.
- 5. Risk integration: the assembly of the outputs of the previous tasks into an overall expression of risk.
- 6. Uncertainty analysis: the propagation of the uncertaintic's .
in the . initiating events, 'f ailure events , accident progression branching ratios and parameters, and source term parameters through the first three analyses above, and the
. determination of which of these uncertainties contributes the most to the uncertainty in risk.
This volume' presents the details of the last five of the six steps listed
'above for the Peach Bottom Atomic Power Station, Unit 2. The first step is
- described in NUREG/CR-4550.2 S.1 l l 1
_ _ _ _ _ _ . . _ __. ~ _ _ _ _ _ _ . _ . _ _ _ _ _ _ _ _ _ _ _ _ . - - _ . _ _ __ . l I
-S.2- Overview of Pench Bottam Atomic Power Station. Unit 2 l
The Peach Bottom Atomic Power Station, Unit 2 is operated by Philadelphia Electric Company (PECO) and is located on the west shore of Conowingo Pond ) in southeastern Pennsylvania, York County. The plant is 38 miles northwest j of Baltimore, Maryland, and 63 miles west southwest of Philadelphia, Pennsylvania. The nuclear reactor of Peach Bottom Unit 2 is a 3293 MWL BWR-4 boiling I water reactor (BVR) designed and supplied by General Electric Company. Unit 2, constructed by Bechtel Corporation, began commercial operation in July 1974 Peach Bottom has four diesel generators (DCs) shared betvcen the two units that are used to supply emergency AC power in the event that offsite power from the grid is lost. Tha DGs supply AC power to four trains of emergency systems for each unit simultaneously. In the event of an accident, there are several systems that can supply coolant injection to the core. 'Two systems are available to provide high pressure coolant injection: the.high pressure coolant injection system (HPCI) and the reactor core isolation cooling system (RCIC). Both systems use turbico driven pumps with steam obtained from the reactor pressure vessel (RPV) and can only be used when the vessel pressure is high enough to run the turbines. Both the low pressure core spray system (hPCS) and the low pressure coolant inj ection system (LPCI) -(which is a mode of the residual heat removal system (RHR)) can provide coolant injection to the reactor vessel during accidents in which the system pressure is low, .Both systems use motor driven pumps and have two loops with two pumps in each loop. Additional systems that can be used as primary sources of coolant, in special cases, are the main feedwater system (FV) and the condensate system (CDS). For additional backup sources of coolant injection the high pressure service water system (HPSW), the control rod drive system (CRD), and the firewater system (DFW) can be used in some circumstances. To allow any of the low pressure injection systems to supply coolant to the vessel, either a break in the primary system has had to occur of sufficient size to depressurize the RPV or the automatic - depressucization system (ADS) is used depressurize the reactor vessel. This system (ADS) uses five relief valves to lirect the vessel steam to the suppression pool (as backup another six relic'f valves or the ADS valves may be opened manually). The Peach Bottom containment is a Mark I BWR containment. The containment ! consists of a light bulb shaped steel pressure vessel forming the drywell which is connected to a toroidal shaped steel pressure vessel forming the suppression chamber (votwell). In the Mark I design the reactor pressure vessel is housed in the drywell. The drywell and the wetwell communicate through passive vents (downcomers) in the suppression pool. Figure S1 shows a section through the Peach Bottom containment. During an accident, steam from the vessel is directed through the safety / relief valvea and is discharged - through a sparger into the suppression pool. The steam is condensed in the pool and any noncondensibic gases pass through the pool l S.2
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into the wetwell atmosphere, Vacnum breakers allow any overpressure in the wetwell to be relieved back Lnto the drywell _ to keep the pressure difference less than 2 psig. Sim Q rly, any steam and noncondensible gases released into the drywell are anted into the suppression pool through the downcomers. The design pressure of the Peach Bottom containment is 56 psig (487 KPa) and the free volume of the containment is 307,000 cubic feet.- To suppress the pressure in the contal vnent during an accident, two trains of containment sprays are located in the Peach Bottom containment. The containment spray system is one mode of the residual heat removal system (RHR). In the event that the Rl!R system fails to suppress the pressure in the containment, the containment can be vented. To reduce the potential of a severe hydrogen combustion event during an accident, the containment is inerted with nitrogen. S.3 Description of the Integrated Risk Annivsis Risk is determined by - combining - the results of four constituent analyses: the accident frequency, accident progression, source term, and consequence analyses. Uncertainty in risk is determined by assigning distributions to important variables, generating a sample from these variables, and propagating _ each observation of the sample through the entire analysis. The sample for Peach Bottom consisted of 20) obs~rvations involving variables from the first three constituent anclyses. The risk analysis synthesizes the results of the four constituent analyses to produce measures of offsite risk and the uncertainty in that risk, This process is depicted in Figure S-2. This figure shows, in the boxes, the computer codes utilized. The interfaces between constituent analyses are shown between the boxes. A mathematical summary of the process, us a matrix representation, is given in Section 1.4 of this volume. The accident frequency analysis uses event tree and fault tree techniques to investigate the manner in which various initiating evencs can lead to core amage and the frequency of various types of accidents. Experimental data, past observational data, and modeling results-are combined to produce frequency estimates for the minimal cut sets that Icad to core damage. A minimal cut set is a-unique combination of initiating event and in'dividual hardware or operator failures. The minimal cut sets are grouped into plant damage states (PDSs), where all minimal cut sets in a PDS provide a similar set of initial conditions for the_ subsequent accident progression analysis (e.g., similar system successes and failures) . Thus, the PDSs form the interface between the accident frequency analysis and _the accident , progression analysis. The outcomo of the accident frequency analysis is a frequency for each PDS for each observation in the sample. The accident progression analysis uses large, complex event trees to determine .the possible ways in which an accident might evolve from each plant damage state. The definition of each plant damage state provides enough information to define the initial conditions _for the accident S.4
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progression event tree (APET) analysis. Past observations, experimental data, mechanistic code calculations, and expert judgment were used in the development of the model for accident progression that is embodied in the APET and in the selection of the branch probabilities and parameter values used in the APET. Due to the large number of questions in the Peach Bottom APET and the fact that many of these questions have more than two outcomes, there are far too many paths through the APET to permit their individual consideration in subsequent source term and consequence analysis. Therefore, the paths through the trees are grouped into accident progression bins (APBs), where each bin is a group of paths through the event tree that define a similar set of conditions for source term analysis. The properties of each accident progression bin define the initial conditions for the estimation of a source term. The result of the accident progression _ analysis is a probability for each APB, conditional on the occurrence of a PDS, for each observation in the sample. A source term is calculated for each APB with a non zero conditional probability for each observation in the sample by PBSOR, a fast running parametric computer code. PBSOR is not a detailed mechanistic model; it is not designed to model the. fission product transport, physics, and chemistry from first principles. Instead, PBSOR integrates the results of many detailed codes and the conclusions of many experts. Most of the parameters that calculate-fission product release fractions in PBSOR are sampled from distributions provided by an expert panel. Because of the large number of APBs, use of a fast executing code like PBSOR is necessary. The number of APBs for which source terms are calculated is so large that it is not computationally practical to perform a consequence calculation for every source term. As a result, the source terms had to be combined into source term groups. Each source term group is a collection of source terms that result in similar consequences. The process of determining which AP3s go to which source term group is called partitioning. This process considers the potential of each source term group to cause early fatalities and latent cancer fatalities. The result of the source term calculation and subsequent partitioning is that each APB for each observation is assigned to a source term group. A consequence analysis is performed for each source term group, ge'norating both mean consequences and distributions of consequences. As each APB is assigned to a source term group, the consequences are known for each AFB of each observation in the sample. The frequency of each PDS for each observation is known from the accident frequency analysis, and the conditional probability of each APB is determined for each PDS group for
-each observation in the accident progression analysis. Thus, for each APB of each observation in the sample, both frequency and consequences are determined._ The risk analysis assembles and analyzes all these separate estimates of offsite risk.
S.6
S,4 Belmits of the Accident Freounnev Analysis The accident frequency analysis for Peach Bottom is documented elsewhere.2 This section only summarizes the results of the accident frequency analyses since they form the starting point for the analyses that are covered in this' volume. Table S-1 (a f) lists four summary measures of the core damage frequency distributions for Peach Bottom for the 9 internal, 4 fire, and 7 scismic PDSs used in the analysis. The four summary measures are the mean, and the 5th, 50th (median) and 95th percentiles and are based on an IJIS sample of size 1000 from the Level I analysis. S.4.1 Internal Initiators PDS is composed of two accident sequences: the first is a large LOCA followed by immediate failure of all injection; the second is a medium LOCA with initial llPCI success but almost immediate failure as the vessel depressurizes below itPCI working pressure, all other injection has failed. Early core damage results, CRD and containment heat removal are working.
-Venting is available.
PDSs 2 and 3 are fast transients - and are composed of four sequences consisting of a transient initiator followed by two stuck open SRVs (the equivalent of an intermediate LOCA) . IIPCI works initially but fails when the vessel depressurines below llPCI working pressure; all other injection has failed and early core damage results. In PDS 2, CRD and containment heat removal are working and steam is directed through the SRVs to the suppression pool, Venting is available. PDS 3 is similar to PDS 2 except that containment heat removal is not working and CRD may not be working for some subgroups (CRD is assumed to be working since the cut setc where it.is not are negligible contributors). PDSs 4 and 5 are station blackouts. PDS 4 is a short term station blackout with DC power failed, It consists of two sequences: one with a stuck open SRV and one without, Early core damage results from the immediate loss of all inj e c tion. Venting is possible if AC power is restored (manual venting is possible if AC is not restored'but considered unlikely). PDS 5 is a long-term station blackout. It is composed of three sequences, one of which has;a stuck open SRV, liigh pressure injection is initially ' working, AC power is not - recovered and either: 1) the batteries deplete, resulting in inj ection ' failure, reclosure_of the ADS valves, and repressurization of the RPV (in those cases where - an SRV is not stuck open), followed by boiloff of the' primary coolant and core damage or 2) IIPCI and RCIC fail on
.high suppression pool temperature or high containment pressure, respectively, followed by boiloff and core damage at low RPV pressure (since if DC has not failed, ADS would still be possible, or an SRV is stuck open). The containment is at high pressure but less than or equal to the saturation pressure corresponding to the temperature at which IIPCI will fail _(i.e., about-40 psig at the start of core damage), PDS 5 is one of i the two dominant internal initiator PDSs.
S.7 l
. .- . - -. . - -.- .-.- - . - - ~ - - . - -
l l
. Table S la Plant Damage State Frequencies - Internal Events )
l Plant. Damage. Core Damage Frequency (1/yr) % TCD State 5% Median Mean 95% Freq,* PDS1 LOCA 2.5E 09 4,4E 08 2.6E-07 7.8E 07 5.8 PDS2 Fast Transient 1.1E-09 3,0E-08 2.2E-07 8.1E-07 4.9 PDS3 Fast Transient 5.9E 11 1.2E 09 6.1E-09 2,7E 08 0.1 PDS4 Fast SB0 3.5E 09 5.0E-08 2.1E 07 7.1E 07 4.7 PDS5 Slow SB0 3.5E 08 4.0E 07 1,9E 06 4,8E 06 42,0' PDS6 Fast ATWS 3.2E 09 5,9E 08 3,0E 07 1.1E 06 6.7 PDS7 ATWS CV 1.2E-09 2.3E 08 1.1E 07 3.8E 07 2,4 PDS8 ATWS CV 1.8E-08 2,9E-07 1.5E 06 5,6E 06 33.0 PDS9 ATWS CV 4.3E-10 1.0E 08 4.4E-08 1,6E 1,0 Total .3,5E-07 1,9E 06 4.5E 06 1,3E 05 100.0
- FCMCD, fractional contribution to the mean core damage frequency based on an LHS sample of 1000, S.8
' Table S-lb Plant Damage State Frequencies - Fire .
1 I Plant Damage Core Damage Frequency (1/yr) % TCD State 5% Median Mean 95% Freq.* POS1 Fast Transient 8.3E-08 2.0E 06 6.8E-06 2.4E-05 34.0 PDS2 Slow SB0 6.8E-09 3.3E 06 5.9E 06 2.1E 05 30.0 PDS3 Slow SB0 2.1E 09 6.5E 07 5.7E 06 2.3E 05 29.0 PDS4 Transient CV 9.5E 10 3.9E 07 1.1E-06 4.2E-06 5.5 Total 1.1E 06 1.2E-05 2.0E 05 6.4E 05 100.0
- FCMCD, fractional contribution to the mean core damage frequency based on an Ills sample of 1000.
S.9
Table S 1c Plant Damage State Frequencies Seismic IIIG, LLNL Plant
- Damage- Core Damage Frequency (1/yr) % TCD
-State 5% Median Mean 95% Freq,* PDS1 FSB RPV 4.7E-10 1.1E 07 7.2E-06 1.4E 05 9.6 PbS2 FSB LLOCA 6.9E 10 4.8E 07 1.4E 05 6.1E 05 18,6 PDS3 FSB LLOCA 1.9E 11 7.7E 08 2.8E-06 2.0E 05 3.7 PDS4 Slow SB0 4.1E 09 6.6E-07 1.7E 05 4.0E-05 22.6 PDS5 Fast SB0 7.7E-ll 4.2E-08 1.8E-06 5,3E 06 2.4~ PDS6 FSB ILOCA 1.9E-10 1.6E 07 3.9E 06 2.1E-05 5.2 PDS7.FSB I/SLDCA 1.6E 10 5,2E-08 1.4E 06 6.1E-05 1.9
!!IG 200 3.3E 08 2.8E-06 4.8E-05 2.8E-04 64.0
- FCMCD, fractional contribution to the mean core damage frequency based on an UlS sample of 1000.
1 S.10 l
p.,...,_.. - . , . - 41 Table S-1d Plant Damage State Frequencies Seismic WWG, LuiL Plant Damage Core Damage Frequency (1/yr) % TCD State 5% Median Mean 95% Freq.* PDS1 FSB RPV 1.0E-10 2.4E 08 1.6E 06 3.1E-06 2.1 PDS2 FSB LLOCA 1.4E-10 9.8E 08 2.9E 06 1.2E-05 3.9 PDS3 FSB LLOCA 1,7E 12 6.7E 09 2.4E-07 1.7E-06 0.3 FDS4 Slow SB0 5.0E 09 8.0E 07 2.0E 05 4.9E-05 26.6 PDSS Fast SB0 6.3E-11 3.4E 08 1.4E 06 4.3E 06 1. 8' ' PDS6 FSB ILOCA 3.6E-11 3.1E 08 7.5E-07 4.0E 06 1.0 PDS7 FSB I/SLOCA 2.2E-11 7.1E-09 1.9E-07 8.3E 07 0.3 LOWG 200 1.4E 08 1.5E 06 2.7E-05 1.0E 04 36.0
- FCMCD, fractional contribution to the mean core damage frequency based on an UlS sample of 1000.
4 S.11
)
I l l Table S le Plant Damage State Frequencies Seismic llIG EPRI Plant Damage Core Damage Frequency (1/yr) % TCD State 5% Median Mean 95% Freq.* PDS1 FSB RPV 7.2E-11 1.7E 08 2.5E 07 1.0E-06 7.9 PDS2 FSB LLOCA 1.5E 10 6.2E-08 5.0E 07 2.0E 06 15.9 PDS3 FSB LLOCA 3.0E 12 1.3E-08 1.2E 07 6.2E 07 3.8 PDS4 Slow SB0 2.4E 09 9.6E-08 6.3E 07 1.8E 06 20.0 ! PDS5 Fast SB0 1.4E-11 4.6E 09 9.1E 08 3.4E-07 2.9~ PDS6 FSB ILOCA 6.2E 11- 1.7E 08 1.5E 07 6.2E-07 4.8 PDS7 FSB I/SLOCA 2.6E 11 6.7E 09 6.1E-08 2.0E-07 1.9 lIIG 200 1.1E 08- 3.6E 07 1.8E-06 8.6E-06 57.2
- FCMCD, fractional contribution to the mean core damage frequency based on an LHS sample of 1000.
S.12
l Table S-1f Plant Damage State Frequencies Seismic LOWG, EPRI Plant Damage Core Damage Frequency (1/yr) % TCD State. 5% Median Mean 95% Fregi
- PDS1 FSB RPV 2.3E-11 5.3E 09 7.9E-08 3.2E 07 2.5 PDS2 FSB LIDCA 4.1E-11 1.6E-08 1.3E 07 5.3E-07 4.1 PDS3 FSB LLOCA 3.7E-13 1.6E-09 1.5E 08 7.7E 08 0.5 PDS4 Slow SB0 3.8E 09 1.5E 07 9.8E-07 2.8E-06 31.0 PDSS Fast SB0 1.5E 11 5.1E 09 1.0E 07 3.8E-07 3.2' PDS6 FSB ILOCA 1.5E 11 4.2E-09 3.7E-08 1.6E-07 1.1 PDS7 FSB I/SLOCA 4.5E 12 1,2E 09 1.1E-08 3.6E-08 0.4 LOWG 200 6.9E-09 2.7E-07 1.4E-06 5.0E 06 42.8
- FCMCD, fractional contribution to the mean cora damage frequency based on an LilS sample of 1000.
S.13
PDSs 6, 7, 8, and 9 are all ATWS sequences. PDS 6 is an ATWS with SLC working. HPCI works and the vessel is not manually depressurized. Injection fails on high suppression pool temperature and early core damage ensues. Venting is available. PDS 7 is an ATWS with failure of SLC, the initiator is a stuck open SRV. High pressure injection fails on high suppression pool temperature and the reactor either is: 1) not manually depressurized or 2) the operator depressurizes and uses low pressure injection systems until either the injection valves fail due to excessive cycling or the containment fails or is vented and the injection systems fail- due to harsh environments in - the reactor building or loss of NPSH (condensate can not supply enough water since the CST can only supply about 800 gpm to the condenser, condensate can only last a few minutes) . Early core damage ensues in case 1 and late core damage in case 2. Venting will not take place before core damage if the operator does not depressurize; but, it may, if he goes to low pressure systems. RHR and CSS are working and the containment pressure vill begin to drop in case 1 or will level off at the venting or SRV reclosure pressure in case 2. PDS 8 is an ATWS sequence with loss of an AC bus or PCS followed by failure to scram. Everything else is the same as PDS 7. PDS 8 is the other dominant PDS for internal initiators. PDS 9 is an ATWS with failure of SLC, the initiator is T1 (LOSP); however, other AC is available. Otherwise, this PDS is the same as PDS 8. PDSs 5 and 8 are the dominant contributors to the core damage frequency. S.4.2' Fire Initiators PDS 1-is a fast transient and is composed of three fire scenarios, two in the control room and one in the cable spreading room. Power is available but remote control of the systems has been lost and auto actuation has failed due to the fire. No injection is available and early core damage ensues. PDSs 2 and 3 are slow station blackouts. PDS 2 is composed of eight fire scenarios in dif ferent emergency switchgear rooms (2A, 2B, 2C, 2D, 3A, 3B, 3C, and 3D). All lead to a fire induced LOSP followed by a random loss of emergency service water due to valve failure resulting in an early loss of all AC power and station blackout. HPCI will work until it fails on battery depletion or high suppression pool temperature and late core damage will ensue. PDS 3 is composed of eight fire scenarios in different switchgear rooms (2A, 2B, 20,-2D, 3A,3B, 3C, and 3D). All lead to fire induced LOSP followed by a random loss of emergency service water from DG failure to run resulting in a delayed station blackout. HPCI will work
- until failure on high suppression pool temperature and late core damage will ensue.
PDS 4 is a core vulnerable transient and is composed of two fire scenarios in emergency switchgear room 2C. The fires result in LOSP with failure of PCS, venting, and failure of most RHR trains. Random failures complete the failure of containment heat removal. The HPCI and LPCI systems succeed but S.14
core damage results when HPCI fails on high suppression pool temperature and LPCI fails when the SRVs reclose on high containment pressure. PDSs 1, 2, and 3 all contribute equally to the core damage frequency. S.4.3 Seismic Initiators PDS 1 is composed of one sequence with a seismically induced LOSP followed by RPV rupture. All injection is lost as a result of the initiator and early core damage ensues. The core damage estimate does not depend on any other consideration; but, for the Level II/III analysis, the status of the containment systems needs to be determined. Onsite AC could be available but the failure probability of a DC is also high in this scenario, we assess,ed that enough onsite AC would be available to vent the containment; but, not enough tc operate the containment heat removal systems. Early containment failure occurs as a result of the seismic event. PDSs 2 and 3 are both fast station blackouts with concomitant Large thCAs. PDS 2 is composed of one sequence with a seismically induced LOSP followed l by a loss of all onsite AC leading to a station blackout. A large 1DCA is l also induced by the seismic event resulting in high pressure inj ection l failure (only steam-driven systems are available and these fail on low pressure in the RPV) and early core damage results. Early containment
-failure occurs as a rs sult of the seismic event. PDS 3 is the same as PDS 2 except that DC power has also_ failed. This has no effect on accident progression since all systems have failed anyway.
PDSs 4 and 5 are station blackouts. PDS 4 is a short term station blackout and is composed of one sequence with a seismically induced LOSP followed by loss of all AC leading to station blackout. HPCI succeeds until battery-depletion or high suppression pool temperature results in HPCI failure and late core damage. PDS 5 is a long-term station blackout and is composed of two sequences, one with a stuck open SRV and one without. Both sequences hwe a seismically induced 1DSP followed by a loss of all AC resulting in station blackout, High pressure injection fails initially upon Radwaste/ Turbine buildhg failure and early core damage ensues. PDSs 6 and 7 are both faut station blackouts with concomitant Inte'rmediate or Small LOCA. PDS 6 is composed of one sequence with a seismically induced LOSP, failure of onsite AC due to cooling water failure, and a seismically induced intermediate LOCA. HPCI works until primary pressure drops below working pressure and early core damage ensues. PDS 7 is composed of two sequences both with a seis,mically induced LOSP followed by a loss of onsite AC resulting in station blackout. A seismically induced
' intermediate or small 'OCA occurs and high pressure injection fails when RPV pressure drops below the systems working pressures resulting in early core damage.
PDS 5 contributes about half the core damage frequency and PDS 2 about a J quarter of the core damage frequency. q l 1 S.15
S.5 6ttidtt1L_hrltrninD_htmL1th S.5.1 Description of the Accident Progrer,sion Analysis The accident progression analysis 1. perf ormed by incans of a large and detailed event tree, the accident p ogression event tree (APET). This event tree forms a high Icvel model of the accident progression, including the response of the containment to the loads placed upon it. The APET is not. meant to be a substitute for detailed, inechanistic computer situulation codes. Rather, it is a framework for integrating the results of these codes together with e xpe r irne nt al results and expert judgment. The detailed, rocchanistic codes require too much comput er time t- ha run for all the possible accident progression paths. Further, no sin vallable code treats all the irtpo r tant phenomena in a complete and ths,..., manner that is acceptabic to all those knowledgeabic in the field. Therefore, the results f rotu these codes, as interpreted by experts, are summatized in an event tree. The resulting APET can be evaluated quickly by computer, so that the full diversity of possible accident progressions can be considered and the uncertainty in the many phenomenn involved can be included. } The APET treats the progression of the accident from the unset of core damage through the core concrete interectian (CCI). It accounts for the various events that way lead to the release of fission products due to the accident. The Peach Bot tom APET consists of 145 questions, most of vbich have recre than two branches. Five time periods are considered in the tree. The recovery of ofit.ite power is considered both before vessel failure as well as after vessel failure. The iussibility of arresting the core degradation prt..ess before failure of the vessel is expitettly considered. Coro damage arrest may occur following the recovery of offsite power or when depressurization of the RPV allows injection by a low pressure injection syste n that previously could not function with the RPV at high pressure. Containment failure l ai considered before vessel breach, around the time of vessel breach and late in the accident. The dominant events that can cause containment failure are drywell moltthrough and the accumulat hn of steam and/or noncondensibles in th containment. The APET is s c. large and compicx that it cannot be presented graphically and must be evaluated by computer. A computer code, EVNTRE, has , been 3 written for this purpose. In addit ton to evaluating the APET, EVNTRE, sorts the myriad possibic paths through the tree into a manageabic number of outcomes, denoted accident. progression bins (APBs). S.5.2 Results of the Accident Progression Analysis Results of the accident progression analysis at Peneb Ecttom are summarized in Figures S 3, S 4, at.d S 5. Figure S 3 shows the mean distribution among the summary accident progression bins for the summary PDS groups. Technically, this figure displays the mean probability of a summary APB conditional on the occurrence of a PDS 3roup. Since only mean values are shown, Figure S 3 gives no indication of the range of valuer encountered. 5 S.16
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i The distributions of the expected conditional probability for core damage j arrest for a given s.ummary PDS group are shown in Figure S-4 Similarly, the distributions of the expected conditional probability for early containment failure (CF) for a given summary PDS group are displayed in . Figure S 5. Early CF means CF before or around the time of vessel breach (VB). Figure S 3 indicates the mean probability of the porsible outcomes of the accident progression analysis. The width of each box in the figure } indicates how likely each accident progression outcome is for each type of accident. S.S.2.1 Internal Initiators Because the Level I analysis did not resolve some of the ATWS sequences all the way to core damage, the ATWS Croup has a probability of 2.4% of no core damage. These involve sequences where low pressure injection is being, used te cool the core and injection does not fail from severe environments or injection valve cycling. In the Level I analysis, these were conservatively assumed to go to core damage. l The LDSP group is composed of two PDSs representing a short term station l blackout with no DC power (PDS 4) and a long term station blackout (PDS 5). j These two PDSs are 46.7% of the mean core damage frequency and PDS 5 is 90% j of the group frequency so that its characteristics dominate. There is a ' O.112 probability of recovering AC power during core degradation and arresting core damage. The high probability of early drywell failure (0.569) is mostly from dryvell shell meltthrough. The dominant APBs for this group have no recovery of AC power and the vessel breach occurs at high RPV pressure. The next highest APBs have AC recovery but no core damage arrest and vessel breach occurs at low RPV pressure. In either case, drywell failure by meltthrough is the dominant containment failure mechanism (although the relative probability is lower in the AC recovered , cases because the drywell can be flooded by containment sprays). If drywell meltthrough does not occur then there is still some probability of failure by overpressure , venting, or pedestal failure. In 12.1% of the cases, AC power is recovered, vessel breach occurs, and the sprays provide sufficient heat removal and reduced CCI to prevent containment ' failure altogether. The LOCA group is composed only of PDS 1 representing 5.8% of the mean core damage frequency. In order to get core damage all injection had to fail and there is no possibility of recovering injection; therefore, core damage
, arrest is not possible. There are no high pressure RPV vessel breach scenarios because of the 1DCA depressurizing the vessel. Since the drywell is flooded by water from the vessel, drywell meltthrough is less likely in this case (only 0.36) . There is some probability of overpressure failure or venting; but, the availability of containment heat removal in this sequence results in a _ high probability of no containment failure at all (0.536).
S.20
1 The ATWS group is composed of four PDSs (PDSs 6, 7, 8, 9). This group is 43.1% of the core damage frequency. PDS 8 is 77% of the group frequency, PDS 6 is 16%, PDS 7 is 61, and PDS 9 is 2t. Since PDSs 7, 8, and 9 are almost the same, 85% of this group is repretented by PDS 8. PDSs 7, 8, and 9 were not resolved all the way to core da nage in the Level I analysis and there is a group average of 2.4% no core damage. All the PDSs have some chance of recovery of injection during core damage and arrestin 6 V088cl breach. The group average is 9.1%. If vessel breach is not avoidedi inost J accident progression bins (about 75%) will have containtment venting before core damage (PDS 7, 8, and 9). Drywell incitthrough can still occur, mainly in cases were the RPV is at high pressure at vessel breach (about 50% of the time usually concurrent with wetwell venting). The Transient group is composed of two PDSs (PDS 2 and 3). This group is 5% of the core damage frequency and PDS 2 is 98% of the group f requency. PDS 2 is very similar to the LDCA group with contaitunent heat removal ! working but no injection recovery. PDS 3 does not have containment heat removal but does have some possibility of recovering injection. It can be seen that there is a small possibility of core damage arrest (1.4%) for the group. The rest is identical to the IDCA group and for the same reasons. The frequency weighted average results are about equally weighted between the LDSP and ATWS groups which are dominated by PDS 5 and 8, respectively. For accidents which proceed to core damage and vessel breach, there is still a significant probability that the core debria will be cooled by an overlying pool of water and either no CCI will occur or the CCI releases will be scrubbed through the water. S.S.2.2 Fire' Initiators The fire PDSs are dominated by scenarios (66%) that do not allow for the recovery of injection or containment heat removal (CilR) and they look much like short or long term station blackout sequences. The impossibility of recovering injection or CllR, however, means that the containment failure probability will be very high from overpressuro related events since the base pressure in containment can not be reduced before vessel breach and __long term containment failure from overpressure can not be mitigated. For the fire initiated PDSs, only in PDS 1 is there a significant probability of being able to cool the core debris by adding water and thereby preventing CCI. S.S.2.3 Seismic Initiators
- The seismic PDSs are dominated by scenarios (100%) that do not allow for the recovery of injection or containment heat removal (CllR) and they look much like short or long term station blackout sequences. The impossibility of recovering injection or CllR, however, means that the contaitunent failure probability will be very high from overpressure related events since the l
l l S.21 l
4 I 1 l base pressure in containment can not be reduced bef ore vessel breach and
; long term containment failure from overpressure can not be mitigated.
- Por the scismically initiated PDSs, no PDS has a significant probability of
> being abic to cool the core debris by adding water and thereby preventing CC1. All have a dry CCI with only a possibility in some cases of an initial layer of water from a 1.0CA or CRD Icakage.
S.S.2.4 Clobal Insights There are significant differences between the internal events results and ? the external events results. Both of the external events had a much lower probability (if any at all) for recovering injection during core damage and for havin6 continuous water flow onto the debris in the cavity and drywell. These two differences imply that the external events PDSs will, in general, ; have a higher probability of early containment failure, a higher probability of drywell moitthrough, that ultimately the containment will almost certainly fail by some mechanism, and that core damage arrest 'will not be likely. The external events PDSs are mainly like short term station blackout sequences with no recovery of AC power and can have compounding events, such as LOCAs, in addition. In the sensitivity analysis performed for no drywell shell meltthrough, removing the possibility of drywell meltthrough will decrease the probability of early containment failure but not as much as would seem to be possible from its calculated frequency because of the fact that multiple failure modes are possible and if one does not occur than another will. Also the probability of containment failure at some time in the accident is not much affected since the probability of _the late failure modes will increase to compensate for climinating drywell meltthrough. For internal events, the total containment failure probability decreases from 0.82 to 0.70; for fire events, it decreases from 0.84 to 0.78; and, for seismic events, it does not change from 1.0, S.S.2.5 Core Damage Arrest Figure S 4 shows the conditional probability of core damage arrest for the PDS summary groups. That is, given that the PDS group occurs-wha't is the probability of core damage arrest. Jnternni Initiators l For the LOSP collapsed PDS group, the probability of core damage arrest is driven directly by the conditional probability of recovering AC power between the time core damage starts and vessel breach occurs. Because of the many available injection systems, injection into the RPV is possible in most cases immediately af ter A0 is restored. While the probability of recovering AC power is high (0.9) in PDS 4, the probability of recovery in PDS 5 is only 0.37 (for long term station blackout, the probability of recovering AC power within the time window of core damage is about 1/3 that I S.22
i i 1 1 of the short term case) and it is the dominant PDS. Since t;.o probability of core damage arrest is about 25% given injection is restored, the average for this collapsed PDS group is only .112. Many factors must be considered in determining if core damage arrest is possible even if injection is I restored. In particular, six major factors were considered in the APET. 2 First, the timing of the injection recovery with respect to the time j between the start of core damage and vessel breach. Second, the fraction of core participating in core slump. Third, the probability of in vessel a steam explosions. Fourth, the amount of core debris which is mobile in the i lower plenum. Fifth, depending upon the accident secnario, the RPV pressure may also be a factor and, sixth, the probability of the core going a recritical during reflood. All of these contribute to our estimate of the fraction of time injection recovery can result in core damage arrest. l For the LOCA collapsed PDS group, injection is not recoverable in the dominant PDSs. If injection was recoverable core damage - would in most cases not even have occurred. The possibility of core damage arrest is, j therefore, zero. In the ATWS collapsed PDS group, inj ec tion recovery depends upon the j
; conditions allowing the operator to be able to depressurize and then that 1 he does it. PDS 8 dominates this PDS group. In PDS 8, inj ec tion is !
recovered with a probability of 0.33 and core damage arrest is 0.1. In the -l other PDSs the probability of core damage arrest is the same or lower, so . I that the overall probability for this collapsed PDS group is 0.09. In the transient collapsed PDS group, injection is recoverable in one of the PDSs but the other is like the LOCA PDS and injection can not be recovered. The frequency of the PDS where inj ection is not recovered dominates and the probability of core damage arrest for transients is only 0.014 Operator error dominates the recovery probability.
. It must be remembered that core damage arrest does not necessarily mean that there will be no radionuclide releases during the accident. Both hydrogen and radionuclides are released to the containment during the core damage process through the SRVs to the suppression pool. In the aajority of the cases , the release is small because, when injection is restored, -
containment heat removal is also restored and, if the mass of hydrogen released is small, containment pressure remains low. This implies radionuclides get releated only through the nominal containment leakage paths. However, in some cases, either a large amount of non condensibles are generated and containment venting is required or containment heat removal is not restored and venting or containment failure occurs. Fire Initiators For the dominant PDSs in the fire analysis, only PDS 1 has a possibility of recovering injection af ter core damage has begun, For PDS 2 to 4, the failure of injection in a non recoverable manner was necessary to get core l S.23
~- . _ . . . ___ __ _- _
i damage in the first place. The average conditional probability for core
, damage arrest for all the fire PDSs together is 0.078.
Seismie Initiators For the dominant c0Ss in the seismic analysis, no PDS has a possibility of recovering injection af ter core damage has begun. Damage from the seism was assessed to be non recoverable for off-site power within the time frame of interest. Recovery of onsite power from none seismic failures in order to prevent core damage was allowed in the Level I analyses; but no further credit was taken in the accident progression analysis because the failures were either easy to recover and so would have been recovered before core damage took place or so difficult that recovery within the time frame of interest was negligible. S.S.2.6 Early Containment Failure Figure S 5 shows the conditional probability distribution for early CF at Peach Bottom for the PDS summary groups. The probability distributiois displayed in this figure are conditional on core damage and vessel breac). That is, the probability of early CF is conditional on the accident proceeding to core damage and then on to vessel breach.
-Internal Initintora The early fatality risk depends strongly on the probability of early containment failure (CT). Early contairaent failure includes both failures that occur before vessel breach and those that occur at or shortly af ter vessel breach. The Peach Bottom containment is a relatively strong containment with the suppression pool being able to absorb large amounts of energy if not released to quickly. The design pressure is 56 psig; but, after evaluation by the experts, an assessed ruean failure pressure of 150 psig was determined. Because of its high failure pressure combined with its energy absorbing capabilities in the suppression pool, the containment is unlikely to fail early from overpressure in nost accidents. The containment has a significant probebility of early overpressure failure only in those sequences where containment heat removal and venting are failed or inadequate (ATWS) and the suppression pool becomes s a'tura ted .
This can result in a significant base pressure before core damage begins and then the pressure increase from hydrogen generation during core damage or events at vessel breach can result in peak containment pressures in the failure range, for non ATWS sequences, early containment failure is most likely to occur from drywell meltthrough and in ATWS sequences to occur from wetwell venting before core damage (drywell moltthrough is the second most likely). S.24 I i ! L_ -_ _ ._. _ _ .__ _ _ _ __ _ _ _ _. - 1
s i Fire initiators For fire initiated events, the probability of early containment failure is high. This is driven by the nature of the dominant PDSs, most of which do not have AC power or injection. This leads to a high probability of drywell meltthrough since the dry.tell will, at most, only have the water in the reacter cavity surop and this is the most favorabic condition for drywell meltthrough. Seismic Initiators For scismically initiated events, the probability of early containment failure is high (70% or greater). This is driven by the nature of the seismic event which does not allow AC power recovery and the characteristics of the dominant PDSs which do not have any continuing injection or containment heat removal. This leads to a high probability of drywell meltthrough since the drywell vill, at most, only have the water in the reactor cavity sump or on the drywell floor and this is the most favorable condition for drywell meltthrough (i.e. as opposed to having,some continuous supply of covering water) . S6 Source Term Analysis S.6.1 Description of the Source Term Analysis The source term for a Si ven bin consists of the release fractions for the nine radionuclide classes for the early release and for the late release, and additional information about the timing of the releases, the energy associated with the releases, and the height of the releases. It comprises the information required for the calculation of consequences in the succeeding analysis. A source term is calculated for each AFB for each observation in the sample. The nine radionuclide classes are: inert gases, iodine, cesium, tellurium, strontium, ruthenium, lanthanum, cerium, and barium, The source term analysis is performed by a relatively small computer code: PBSOR. The purpose of this code is agi to calculate the behavior of the fission products from their chemical and physical properties and 'the, flow and temperature conditions in the reactor and the containment. Instead, PBSOR provides a means of incorporating into the analysis the results of the more detailed codes that do consider these quantities. This approach is needed because the detailed codes require too many computer resources to be able to compute source terms for the numerous accident progression bins and the 200 observations that result from the sampling approach used in NUREG 1150. PBSOR is a fast-runnire. parametric computer code used to calculate the source terms for each APB to' aach observation for Peach Bottom. -As there are typically about a 450 bins-tot each observation, and 200 observations in the sample, the need for a source term calculation method that requires S.25
i i ] few computer resources for ota evaluation is obvious. PBSOR provides a . framework for synthesizing ths results of experiments and mechanistic codes, as interpreted by experts :n the field. The reason for " filtering" the detailed code results throt.gh the experts is that no code availabic
- treats all the phenomena in a inancer generally acceptabic to those knowledgeable in the field. Thus, the e,.perts are used to extend the code results in areas where the codes are deficient and to judge the applicability of the model predictions. They also factor in the latest experimental results and modify the code results in areas where the codes are known or suspected of oversimplifying. Since the majority of the parameters used to compute the source term are derived from distributions determined by an expert panel, thu dependence of PBSOR on various detailed codes reficcts the preferences of the experts on the panel, i It is not possibic to perform a separate consequence calculation for each of the approximately 93,000 source terms computed for the Peach Bottom integrated risk analysis. Therefore, the interface between the source term analysis and the consequence analysis is formed by grouping the source i terms into a much smaller number of source term groups. These groups are defined - so that the source terms within them have similar health effect weights, and a single consequence calculation is performed for the mean source term for each group. This grouping of the source terms is performed with the PARTITION program, and the process is referred to as
- " partitioning".
The partitioning process involves the following steps: definition of an early health effect weight (EH) for each source term, definition of a chronic health effect weight (Cll) for each source term, subdivision (partitioning) of the source terms on the basis of EH and CH, a further subdivision.on the basis of the time the evacuation starts relative to the start of the release, and calculation of frequency weighted mean source terms. . The result of the partitioning process is that the source term for each accident progression bin is assigned to a source term group. In the risk computations, each accident progression bin is represented by the mean source term for the group to which it is assigned, ' and the consequences _ calculated for that mean source term. . S.6.2 Results of the Source Term Analysis When all the internally-initiated accidents at Peach Bottom are considered together, the plots shown in Figure S-6 are obtained. These plots show four statistical measures of the 200 curves (one for each observation in the sample) that give the frequencies with which release fractions are exceeded. Figurc S 6 summarizes the complementary cumulativo distribution functions (CCDFs) for all of the radionuclide groups except for the nobel gases. _ The _ mean frequency of exceeding a release fraction of 0.10 for I and Cs is on the order of 10 6/ year-and for Te and Sr it is on the order of S.26
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10 7/ year. The mean frequency of exceeding a release fraction of 0.01 for the La radionuclide class is on the order of 10 8/ year. Similar results are displayed in Figure S-7, S 8, and S 9 for the fire, LLNL seismic hazard curve, and the EPRI hazard curve, respectively. S.7 Consecuence Analysis S.7.1 Description of the Consequence Analysis offsite consequences are calculated with MACCS for each of the source term groups defined in the partitioning process. MACCS tracks the dispersion of the radioactive material in the atmosphere from the plant and computes its deposition on the ground. MACCS then calculates the effects of this radioactivity on the population and the environment. Doses and the ensuing health effects from 60 radionuclides are computed for the following pathways: immersion or cloudshine, inhalation from the plume, groundshine, deposition on the skin, inhalation of resuspended ground contamination, ingestion of contaminated water and ingestion of contaminated food. MACCS treats atmospheric dispersion by the use of multiple, straight-line caussian plumes. Each plume can have a different direction, duration, and initial radionuclide concentration. Cross vind dispersion is treated by a
.culti step function. Dry and wet deposition are treated as independent processes. The weather variability is treated by means of a stratified sampling process.
For early exposure, the following pathways are considered: immersion or cloudshine, inhalation from the plume, groundshine, deposition on the skin, and inhalation of resuspended ground contamination. For the long-term exposure, MACCS considers following four pathways: groundshine, inhalation of resuspended ground contamination, ingestion of contaminated water and ingestion of contaminated food. The direct exposure pathways, groundshine, and inhalation of resuspended ground contamination, produce doses in the population living in the area surrounding the plant. The indirect exposure pathways, ingestion of contaminated water and food, produce doses in those who ingest food or water emanating from tho' area around the accident site. The contamination of water bodies is estimated for the washoff of 4and-deposited material as well as direct deposition. The food pathway model includes direct deposition onto the crop species and uptake from the soil. Both short-term and long term mitigative measures are modeled in MACCS, Short term actions include evacuation, sheltering and emer6ency relocation out of the emergency planing zone. Long term actions include relocation and restrictions on land use and crops. Relocation and land decontamination, interdiction, and condemnation are based on projected long term doses from groundshine and the inhalation of resuspended radioactivity. The ~ disposal of agricultural products and the removal of farmland from crop production are based on ground contamination criteria. S 29
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't Peach Bottom: Total EPRI Seismic Source Term CCDF
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