Text

Probabilistic Risk Assessment
	ML20052A093
Person / Time
Site:	Limerick
Issue date:	03/23/1982
From:	; PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC
To:	;
Shared Package
ML20052A091	List: ML20052A090; ML20052A093;
References
	NUDOCS 8204270072
	Download: ML20052A093 (190)
	v • d • e

{{#Wiki_filter:- - - - - - - - - - - - - _ - - - - - - - - - _ - - - - - - - i-1 t I 00 w o 4 ENCLOSURE 3 i N i. N o o 9 E l, 9 E e "n w Mu '1 E \\ ~ y LIMERICK GENERATING STATION j u E PROBABILISTIC RISK ASSESSMENT l j I

SU V VLARY OF CHRONOLOGY i f l o MAY 6,1980 - NRC LETTER i o MAY 21,1980 - NRC MEETING - SCOPE OF STUDY 1 l o OCTOBER 2,1980 - NRC MEETING - INTERIM i ,l PROGRESS REPORT o DECEMBER 9,1980 - NRC MEETING - RESULTS f o MARCH 17,1981 - SUBMITTAL OF REPORT 1

I i i ~ I i f i NRC 1.ETTER MAY 6,1980 l 6 I .o CONDUCT A PRELIMINARY RISK ASSESSMENT i t RECOGNIZE WASH 1400 CRITICISMS I ._o i i o COMPLETE EVALUATION IN 120 DAYS s l 1 e l i i

i l i 1 = AGS PR_A SCOPE. r e EARLY FATALITIES FOR ALL BUT EXTERNAL EVENTS e 1970 POPULATION t o SITE DIFFERENCES i o DESIGN DIFFERENCES i j o DATA / METHODOLOGY DIFFERENCES i i I i i

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i t i j t i 'l i l CASES STUDIED i l 1-o WASH 1400 at Limerick Site f l i o Limerick Plant at Limerick Site I l 4 esar is l l l l l 0

t WASH 1400 PLANT AT LIMERICK SITE t 1 l 0 WASH 1400 Probabilities o WASH 1400 Release Fractions i l i i: i i' o WASH 1400 Methods t

!i o Limerick Site Population l
i o

Limerick Site Meteorology i es.1 e l t i j.

L umsricK Preliminary Risk Assessment Site Comparison 1 j io I I L MANCAUSED RISK 100 } TOTAL NATURAL RISK 1000 N 1 i 10,000 FREQUENCY 1 3 (events / year) 100,000 i 1 1 MILLION WASH 1400 BWR AT LIMERICK SITE N 1 l 10 MILLION 100 MILLION ESU MB \\ 1 1 BILLION 1 10 100 1000 10,000 { FATALITIES j .u= = l 4 L 9

i J i l I l ? I l v LIMERICK ^ PLANT AT LIMERICK SITE 4 l l i I ~ o Limerick Plant Features l o Limerick Site Features l i j o Updated Data and Methods l t i 4 i

l .I (LIMERICK ANALYSIS Systems -- Limerick (BWR/4) o o Procedures -- Limerick Projection I .i o Operating Experience Data -- Philadelphia j I 'l Electric Company Where Applicable j o Containment -- Limerick Mark II l; o Sequences -- Limerick Specific i o Containment Analysis -- Limerick Specific ~ o Consequences -- Limerick Site-Specific Risk -- Limerick Specific i i SS4F 4 i ~

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!1 l' ET I S KC IRE Y M G L O I L .T O A D TN O A H L T P E K M C I RE M I L e, 1 l .I' 1 I;l

\\ ACCIDENT INITIATORS i i, t 4 o Plant initially Operating at Power l ( o Plant Safety Challenge Occurs !l i !i - Normal Transient i t - SmalltLoss-of-Coolant Accident i i. i - Large Loss-o'f-Coolant Accident i - Transient Without SCRAM Allows Specific Capability to be Analyzed 354F to ~

I Task I Frequency of. Core Damage F Trequency Task II of Radioactive Task IV Release Consequences of Radioactive Release l Magnitude Task III of i Radioa ct ive j Release l I. l-I i Major Tasks of the Analysis l l l l l I ' ' * * ~ * ' ' * . ~. -. - - -.. _ _..

i u, ~ }' s, i t s. ~ Technical Spec i f ica t ions ~ [ Reliability Operating i Data Procedures k Construction & Functlona1 Quantification of Description __g of Vital. System / Functional Fault Trees Systems- ? 4 ii f Selection & Construction of Identification, Quantification & p D ma e I Accident Sequence Quant if icat ion p of Accident Event Trees Classification of '"9"""' Init lator s Accident Sequences } Success Criteria 1 Determination of Core Damage 4 I Frequency (Task 1) l-i I k I I j 3 ,s. 1& m

1 j l Accident Initiators i i i I I l Transients Loss of I Transients Without Coolant l Scram Accidents Turbine Trip Large ii MSIV'~ Closure Small Loss.of Feedwater Very Small IORV Loss of Offsite Power i Indepth Analysis for Transients Assured .s.,,. i i

w System Level Fault Trees .~ l ~ .l L:.* ? i, 'M*.Nrk e.P fTE.4 N iff s N I s t.

CRD INJECTION e HIGH PRESSURE COOLANT INJECTION i

e REACTOR CORE ISOLATION COOLINO

FEEDWATER l

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CONDENSATE o LOW PRESSURE CORE SPRAY
AUTOMATIC DEPRESSURIZATION e EMERGENCY SERVICE WATER SYSTEM

$~ ~ o POWER: AC (NORMAL), r. i 'O e RES400AL HEAT REMOVAL AC (EMERGENCY), DC l L. 2 - LOW PRESSURE

DIESELS

[ e-COOLANT INJECTION }.I - CONTAINMENT SPR AY . STANDBY LIQUID CONTROL I'.. - RHR 8ERVICE WATER '.i i h. FAULT TREES ARE LOGIC MODELS FOR FUNCTIONS ( ?: F v 1 I l I i. 4 ;.. .] Q-...... _.. ,a, _. t a

GENERIC ACCIDENT SEQUENCES o Melting lBefore Containment Failure ~ i o Melting: After Containment Failure i i o Transient Without SCRAM i o Transient Without SCRAM (Heat i Removal Failure Case) f ~

i Functional "**I""*"" l Description of _4 i Con t a inme'nt Analysis i ~- i Core Damage Frequencies y Construction Identification & Radloactive of Quantification of Release l Containment Release Pathways Frequencies i Event Trees by Accident Class s. h Core tielt Analysis Determination of Radioactive Release Frquency (Task 11) i ~ i I l l I l

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.f i l l ? N. f Decontamination Fac tors I 'g '. ir l Core ~ ? Analysis of Quantification Radioactive l Radioactive L in-Plant t of Release Release Fractions Fractions 7 Content Transport by Pathway & Class 6 t i 1 Determination of Magnitudes of Radioact ive Release (Task III) l i e i i ! I j

P e 4 Demographic Release Release k Data Frequencies Fract ions U i s, SIE* Analysis of Quantification l Risk Analysis a k External og m unes Transport External Effects t t i ? Determination of External Effects (Task IV) i 4 f ~ e l y e

i, Over 1000 l Events c IFour i I j i Event Classes 4 1 Containment ),I 1 &e Failure Paths t s.%. i 0 M Release Fractions t i Off Site r i Risk i 8547 55 e 4 e e

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Sumary of the Accident Sequence Frequencies Leading to Degraded Core conditions Sumed Over All Accident Sequences within a Class. .i ~, 20-* WASM-1400: Approntaste freewency of rr

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1. I i l PURPOSE PROVIDE'A-BRIEF OVERVIEW OF SELECTED SAFETY RELATED SYSTEMS. l DISCUSSION TOPICS 1 i e MAJOR SYSTEM COMPONENTS ( 0 FLOW RATES .g FLOW PATHS GLOGIC - i O e

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SYSTEMS TO BE O!SCUSSED SECCS SYSTEMS .HIGH PRESSURE C00LANT.If1JECTIOf1 (HPCI) - AUTOMATIC DEPRESSUR!ZAT10ft (ADS) - CORE SPRAY (CS) - LOW PRESSURE C00LAfiT IfiJECTIOf4 (LPCI) OREACTOR CORE ISOLAT!0f4 C00LIf1G (RCIC) ORES! DUAL HEAT REMOVAL (RHR) - SUPPRESS!0fi POOL COOL!rJG - C0f4TAlf1MEi4T SPRAY - STEAM C0fDEriS!f1G GRHR SERVICE WATER SSTAfiDBY LIOUID C0f1 TROL l l l l l

E ONUMBER OF ' PUMPS - 1 (TURBINE DRIVEN) OCAPACITY .5,600 GPM l eWATER SOURCE - CONDEtlSATE STORAGE / SUPPRESSION POOL SINITIATION SIGNAL - REACTOR LOW WATER LEVEL (L2.) Da HIGH DRYWELL PRESSURE OPERMISSIVES - f40NE OTRIP SIGNALS - ISOLATION SIGNALS

1. STEAM SUPPLY BREAK A. HIGH TEMPERATURE IN AREA 0F STEAM Lit 4E B. HIGH STEAM FLOW

2. REACTOR LOW PRESSURE

3. HIGH TURBINE EXHAUST DIAPHRAGM PRESSURE TURB!t1E TRIPS

1. REACTOR HIGH WATER LEVEL (L8)

2. HIGH TURBINE EXHAUST PRESSURE

3. LOW PUMP SUCTION PRESSURE

4. OVERSPEED GPOWER SUPPLY - DC l.

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AD.S STOTAL NO. OF SRV S .14-ALL PIPED TO SUPRESSION POOL. e0PERATION-PRESSURd/ MANUAL ~~ eADS FUNCT10t4 - 5 0F 14 VALVES GOPERATION - PRESSURE / MANUAL / AUTOMATIC eAVERAGE CAPACITY - 1,035,000 LaS/HR/ VALVE Siti!TI ATION SIGNAL - REACTOR LOW WATER LEVEL (L1) AND HIGH DRYWELL PRESSURE SPERMISSIVES-}MINUTETIMEDELAYCS OR 1 RHR PUMP RUNN 4 SPOWER SUPPLY - DC i =. y 4,.m,--, e- - -y,- y-.,

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.CS SNUMBER OF PUMPS - 4 ONUMBER OF LOOPS - 2,. eCAPACITY - 3,175 GPM/ PUMP SWATER SOURCE - SUPPRESSION POOL / CONDENSATE STORAGE elNITIATION SIGNAL - REACTOR LOW WATER LEVEL (L1) OR HIGH DRYWELL PRESSURE OPERMISSIVES - REACTOR LOW PRESSURE eTRIP S1GNAL - MANUAL OPOWER SUPPLY - AC a e.. e w- ,m_,.ww w --m p. m,

eftUMBER'0F" PUMPS-k ONUMBER OF LOOPS - 4 OCAPACITY - 10,000 GPM/ PUMP GWATER SOURCE - SUPPRESSION POOL gINITIATION SIGflAL - REACTOR LOW WATER LEVEL (L1) OR HIGH DRYWELL PRESSURE SPERMISSIVES - REACTOR LOW PRESSURE STRIP SIGNAL - MANUAL OPOWER SUPPLY - AC I' l l l l ww.ev-es g, c- ---e.- i---- g-.--s. ,,._,y,. .-my---- ,,-m,- y y -m, - - - - .ww-w-,

~.-. i .._..n E e fiUMEER OF PUMPS - 1 (TUR3If4E DRIVEii) eCAPACITY - o00 GPM gwATER SOURCE - C0f4 dei 4 SATE STORAGE / SUPPRESS 10f4 POOL ' O IfilII ATIOff SIGT4AL .REhrTOR-LOWA/ATER LEVEL..(L2).. G PERMISS IVES - f10f4E eTRIP SIGi4ALS - ISCLATIOfi SIGt1ALS

1. STEAM SUPPLY ?!PE BREAK A. HIGH TEMPERATURE If1 AREA 0F STEAM LINE S. HIGH STEAM FLOW

2. REACTOR LOW PRESSURE

3. HIGH TUR31f4E EXHAUST DI APHRAGM PRESSURE TUR3INE TRIFS

1. REACTOR HIGH WATER LEVEL (L8)

2. HIGH TURSif4E EXHAUST PRESSURE

3. ovsRSPEED

4. LOW PUMP SUCTIOf1 PRESSURE GPOWER SUPPLY - DC t

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~ SUPPRESSION POOL COOLING ' S NUMBER.0F. HEAT EXCHANGER ;' OOPS.:AVAILABLE. .'2.. L 9 NUMBER OF RHR PUMPS AVAILABLE - 4(2 PER LOOP) e FLOW PATHS RHR SIDE: SUPPRESSION POOL TO SUPPRESSION POOL RHR SERVICE WATER SIDE: COOLING TOWER / SPRAY P0f4D TO C00Lif4G TOWER / SPRAY POND e INITIATION SIGNAL - MAf4UAL eTRIP SIGNAL - MANUAL ePOWER SUPPLY - AC R l l l l rv w --, ,4 -. - - -,m.,,- _.,.m ..w-f- ,--,-, *n r-ar e *- -+er' b--- ~ 7 me w v'

l i CONTAINMENT SPRAY 0 NUMBER OF, SPRAY. LOOPS.'AVAILABLE -l:. -DRYWELL-3 - WETWELL - 4 e NUMBER OF RHR PUMPS AVAILABLE - 4 e WATER SOURCE - SUPPRESSION POOL eINITIATION SIGNAL - MANUAL GPERMISSIVES -DRYWELL SPRAY DRYWELL PRESSURE, PLUS LPCI INITIATION SIGNAL PRESENT, PLUS LPCI INJECTION VALVE CLOSED -WETWELL SPRAY NO LPCI INITIATION, OR LPCI INJECTION VALVE CLOSED eTRIP SIGNAL - MANUAL GPOWER SUPPLY - AC l 9 w-

7.-... STEAM CONDENSING SNUMBER OF HEAT EXCHANGER LOOPS AVAILABLE - 2 ef4 UMBER OF RHRSW PUMPS AVAILABLE - 4 (2 PER LOOP) .r - RHR SIDE: STEAM FROM MAIN STEAM VIA HPCI STEAM SUPPLY LINE - RHR SERVICE WATER: SUCTION FROM AND RETURN TO COOLING TOWER / SUPPRESS 10ft POOL - CONDEf4 SATE RETURN TO SUPPRESSION POOL /RCIC PUMP SUCTION elfilTIATION SIGriAL - MANUAL gTRIP SIGNAL - MANUAL OPOWER SUPPLY - AC I i ~ ~~ - - ---.*y----, v g-- y -w-, -.-

1 RHE 3ERVICE WATER 9 SHARED SYSTEM ?, 9 TWO LOOPS, EACH C0f;SISTli4G OF: - 1 HEAT EXCHAfiGER FR0h EACH UillT - 2 PUMPS 9 CAPACITY - 9,000 GPM/ PUMP SWATER SOURCE - SPRAY P0llD - C00LIi1G TOWER SI(4ITIAT10f; SIGlJAL - MAflUAL S PERiilSS IVES - fiOf1E S TR IP S IGfiAL - f4AfiUAL - HIGH RADIATIOf1

POWER SUPPLY - AC i

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. ~. STANDBY LIQUID CONTROL 1 9 NUMBER.OF PUMPS - 3 g. SCAPACITY - 43 GPM/ PUMP 4 INITIAT10t1 SIGt4AL - REACTOR HIGH PRESSURE OR REACTOR LOW WATER LEVEL (L2) SPERMISSIVES - APRM f40T DOWN SCALE - TWO Mit4UTE TIME DELAY STRIP SIGr4AL - STORAGE TANK LOW LEVEL GPOWER SUPPLY - AC i ~

LIMERICK PRIMARY CONTAINMENT ULTIMATE PRESSURE CAPACITY INVESTIGATION l l SUBTASKS

1.. CONTAINMENT. RESPONSE TO PRESSURE BEYOND DESIGN

PRESSURE ~

~ ' ' - ~- A. SEMI-INFINITE CYLINDER CALCULATIONS B. FINITE ELEMENT ANALYSIS 2. REFUELING HEAD AND HATCH PRESSURIZATION RESPONSE A. REFUELING HEAD B. HATCHES 3. PRIMARY BOUNDARY VALVES PRESSURE RESPONSE A. BUTTERFLY VALVES B. GATE, GLOBE & CHECK VALVES 4. WRITE REPORT l i

l t ~. - -. l t l @ RPV MOVASLE SHIELDING PLUSS 2*l ELLIPTICAL HEAD "/ /+ 1 EL. 348 '- 4" -REMOVABLE BOLTED COVER 5 -lo" 1.0 '- / Ih, ~' EL.325'- 8' l. 4: e/ 36'-4' I.D. TRUCTURAL ~ EL.312 ',8' ./ SEPARATION ),.1/. .j ;. * '. '. m. STEEL LINER PLATE ANCHORED TO G'2'.I CONCRETE. }?:' 24 '- 7' ' .i PLATF. V-l EL.2G5'- o't .'/

' % /

GUlPMENT HATCH 7 ,i 'J, W PERSONNEL EQUIPMW'(g L K IF HATCH DOOR P / PuTF. .,A,- HATCH 9ts'3%'I.O. ? +' EL237'-ll' / p w i a r t l: ET DEFLECTOR PLATES fa DOWNCOMER yS 4'-( to'- 3'_ T VENT PlPES ~ /OOL WATER LEVEL v ~_ N VENT PIPE G,, g _ 87 ' slVI.D. l-RESTRAINT 515 TEM = FL 181'- 11' / Y

. -v: .u.

.:..c; --s'- o* SECONCARY REACTOR BL CONTAINMENT WALL WALL (SECONDARY CONTAINMENT) VERTICAL SECTIO N. CONTAINMENT GENERAL ARRANGEMENT l

==e=- e..

i e 1 t eu m um I

s i

I l l e 1 e M = I, 4 me g 5.0 g tsE Deuel 8.e 9.e 4.e ee 98.9 88.0 5, 994 luWel 3, espro I A e m W INI. W ""*'* pe,,, e a te.,m eo w IM. W e. t ese. g

p.---

M M-E S PRESSURE AND TEMPERATURE TIME HISTORY

t ". ~ _ _ ' _ _ ' l 4 I i r g 2 l l I G o l H m l v \\ a E F I ~ l I b L W g l 2 , a ~ l l H e .: 4 . z. I w E g { 7 - a t, S t 3 2 ~ 4 Ew \\ 1 2 ) i I w-e + I M O w I D W W w - sy E I A

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a a i

1 s

em J 9 I i i I 8 tle8t M 9 e

H M%
e

.e e q ,,.e e, m--___- -~%-._.._ s--___. r-__

~_, i &#Samog NOW9ee

M.e e

G 1 e g 1 i e o a F 2 ( r o l F g W D s L 1 \\ , -}, ^I l w EiI w } m D 3__ 2 ll 4 } -s m g r 1 1 I I I w W 8""* e + 0 0 Z< I a:C x w O x g D ~ e l ow i C 24,l = Ei g J. 3-3 e e q g 3 a g e a ItM musas k e l ' - - - ~ - -.....,, _ _ l ~~ -.. ~. _,

I e ~ t ....s..e i samme a t se e i l e I EE8 -

l..

E, \\l \\ t ~ se e4 I e a e a a n's me se me *=^ ,e ne we we me fue timanos 3.e im mo w.es as sgamees=,. tem.Suommune se, tem. eA w same e-mmeena. >=u'r' "'** '*** . fe a W MO"" 9 ,,, y g DRAFT p_ u 1 g. PRESSURE AND TEMPERATURE TIME HISTORY 9

.t TA BLE I l MATERI AL PROPERTIES e.~ MATERIAL COMPRESSIVE YlELD ULTIMATE MODULUS OF Poisson REMARK STRESS STRESS STRESS ELASTICITY RATIO CCNCRETE 400aOpsi 3G054 KSI 0 17 l 1 l RENFORCE-G800Copsi 100000.Opsi 29000.0 KSI O.3o ASTM AGIS MENT GRADE i UNER PLATE fA000.O psi 290004KSI O.30 ASTM A 285 GRADE A FIREBOX GUALITY i

CONTAINMENT ANALYSIS FOR EXTREME PRESSURE SEMI-INFINITE CYLINDER ANALYSIS 1. NEGLECT RESTRAINT OF BASE SLAB AND DIAPHRAGM . SLAB . a.. 2. USE "CECAP" COMPUTER PROGRAM - LINEAR ELASTIC ANALYSIS 3. MEMBRANELOADCALCULATEDBYE.R. & .PR T 2T 4. MATERIAL PROPERTIES FROM AS-BUILT RECORDS AT MID HEIGHT OF WETWELL 5. INCLUDE REBAR AND LINER PLATE 6. FAILURE CRITERIA - YIELD STRESS OF ALL STEEL COMPONENTS

1 l l R= 52S" 1 l C A

'z-DL g

a \\_ .PR ( i Ov DIAGONAL STEEL As = 0.8590 Wm' / \\ y MERIDlONAL STEEL

g -l As=0.5GO3 m2/m l

8 i 0 f HOOP STEEL z As = 0.6855 in /m / / LINER PLATE p'4, t = 72* i 'I SEMI, INFINITE ANALYTICAL MODEL l j

l

.l 'j eme m -ene t

TABLE 2 @ RESULT - SEMI-INFlNITE CYLINDER AN A Lysis STRESS STR AIN _a PRESSURE Ksi INCH /INcHe m HOOP MER. DI AG. HOOP MER. DIA 120 psi 68.00 l 1. 6 0 - 69.63 Al73 . 401. 2287 RESULT - FIN IT E ELEMENT AN ALYSIS STRESS STR AIN NCH/ INCH xIci+' /in KSI in EM mmN fy = GS Ksl dy =ER234sxio / SHEAR DIAG HOOP MER. [MAGj$ HEAR HOOP M t 7hM8 24.5 3 6.67 53.13 879 12G4 1832 1574 1332 120 psi MID-HEEHT 45.G4 38.64 Ms 47.30 41.10 Z2.55 iG31 15 91 820 ygve 3 7ss sz.o. .+.s5 ise8 i7eo z22e este 201i 150 psi Mio+1ElsHT G7.24 se.ss NsE 55.00 53.09 29.Gl 190Z 18 31 1020 ,_e m

CONTAINMENT ANALYSIS FOR EXTREME PRESSURE \\ FINITE ELEMENT CONTAINMENT ANALYSIS 1. IDEALIZE ENTIRE CONTAINMENT 2. USE "FINEL" COMPUTER PROGRAM WITH IDENTICAL MODEL AS USED IN INITIAL DESIGN

a. -

3. APPLY UNIFORM INTERNAL PRESSURE IN WETWELL AND DRYWELL 4. USE AS-BUILT MATERIAL PROPERTIES 5. "FINEL" COMPUTER PROGRAM IS CAPABLE OF NON-LINEAR ANALYSIS - CONCRETE CRACKING AND STEEL YIELDING ARE INCLUDED IN CODE 6. FAILURE CRITERIA - YIELD STRESS OF ALL STEEL COMPONENTS AT A SECTION ..1

26000-K Sz = 2.0" SR=05" y I \\ 2400.0-Idz=1.Z" R=0 20' \\ l

goo
o.

o I 120 PSI - PRES 5URE I 150 PSI - Pf9 ESSURE I V 1 z000.o ( I \\ f- .\\*\\ '\\g00.0 r \\ i \\ 1600.o l b6z=l.20" l ll4R=0.bl* I400.O -. /I l6 / z=0.75" I200. O -- l l fa = 0 55" / / )'h \\ l000. o - h. T / /N s N 800.O N f N s N l .J 1 sz = 0.44' 000.O /7 f a =. i.se" ~

ll:

gzgo.27" 400.O-- b / = l 00" ,11. 200.o -. [ M.. I I I I I I A ' -- 0 0 0 o O O O FINITE ELEMENT MODEL WITH DISPLACEMENTS AT 120 & 150 PSIG -,.--y

l l l nRPV I i m ]\\ EMOVABLE HEAD-WALL f =-.==h( JUNCTION J.- i. i 1 1 F i I l DRYWELL. i ~l G 'g- \\ l l (.i ab./ [ l \\ I \\ t s APHRAGM DIAPHRAGM SLAS-WALL SLAS JUNCTION 's i ' \\ CONTAINMENT' WALL WETWELL WALL MIDHEIGHT ( WET WALL % \\ PEDESTAL _ N BASE SLABN , WALL SLAB g*/, JUNCTION ] i-t l LOCATION OF CRITICAL SECTION

1 Com W'2" FEN 7-c it18 DIAG SARS / / -M 18 VERT. SARS - l DIAPHRAGM SLAS VERTICAL BAR ANCHORAGES - sD DIAPHRAhM SLAS D

Eri e

l.- ( \\ pI b Y LINER PLATE e i f2'// HOOP BARS (K

- W ;*

DIAPHRAGM SLAB-WALL JUNCTION --4 4 1

18 VERT. BARS r

i SHEAR TIES HOOP BARS I 'l x i l g. g. .i -4 WALL-MIDHEIGHT DETAll OF CRITICAL SECTION l l l e -r- ,. - + - - +, _ _ _.,

G'_ 0" - - + - rHOOP BARS ii['ih VERTICAL BARS BASE SLAB Fi CONTAINMENT WALL l pi$ 9 DlA6. TIES 7 O MAD MAT CONCRETE i 2 /) pLL SLAB JUNCTION L l DETAIL OF CRITICAL SECTION 1 \\ i 1 l l l

PRIMARY B0UNDARY VALVE PRESSURE CAPACITY INVESTIGATION 1. IDENTIFY POTENTIAL VENT PATHS 2. IDENTIFY VALVE TYPES IN VENT PATHS A. BUTTERFLY VALVES B. GATE VALVES C. GLOBE VALVES D. CHECK VALVES 3. DETERMINE CAPACITY OF VALVE TYPES CONCLUSIONS 1. NO VALVE FAILURES BELOW 300 PSIG 2. LEAKAGE NOT PREDICTED BELOW PRESSURE OF 140 PSIG. e

4

i..... <,

.x........;.. -, 4 l BUTTERFLY VALVES a,- a.., .: = I I 3 TO SGTS Y Y D DUCTING s CONTAINMENT s t l 1 1 TYPICAL POTENTIAL VALVE LEAKAGE PATH N w w

REFUELING HEAD AND HATCHES PRESSURE CAPACITY (CHICAGO BRIDGE & IR0iG REFUELING HEAD & HATCHES 1. LINEAR ELASTIC SHELL ANALYSIS 2. PRES'SURE ' ANd TEMPERATliR$ i.0 ads ? INCL.dDED T 3. BOUNDARY CONDITIONS FROM BECHTEL "FINEL" ANALYSIS 4. ANALYSIS PRESSURES AS GIVEN BY BECHTEL "FINEL" RESULTS CONCLUSIONS 1. GENERAL MEMBRANE STRESSES BELOW YEILD AT 120 PSIG 2. NO FAILURES PREDICTED

..eee. e.,,,, ,w. f \\ SECTION @ CONTAINMENT REMOVABLE ,f seAo saett SECTION V / s socrs 1 g e c,n o w g 0.s.- ., y _ / / / l b: O' ( SECTI ON @ l \\ \\ \\ SECTION @ e 4/ \\ l \\ \\ \\ / -SECTION @ i t f CONTAINMENT NALL \\ i a REMOVABLE HEAD WALL JUNCTION I

2 ) 1 t TABLE 3 l N z g o MAXIMUM STRESS INTENSITY P KSt m m v g y TEMPERATURE 340*F TEMPERATURE 70*F l 30.o 23.8 ' 05 2

5. S 79 3

c 75 93 t 4 G.S 47 i 5

9. 2 91 G

lB. O IS O I I4 0 4I,yo '5 L

3. 7 II.2 4

3 c.6 IG.O Oo 4 G3 G.9 ~ 5 9.I I20 6 24 0 24 0 6 STRESS AT CRITICAL SECTIONS SbOWN ON FIG. 9 REFERENCE : CB & I ENCLOSURE - - - - ~ * '~~ ~ ~

l l' ., u : l- [ CONCLUSIONS ULTIMATE PRESSURE FOR S.TRUCTURAL INTEGRITY 140 PSIG L t FAILURE L VERTICAL CRACK AT WETWELL MIDHEIGHT e I l - J t - o l e)

i l j! DISCUSSION OF RESULTS l l i4 l o Site Differences l 1 li o Plant Design Differences o Data Differences l 0 Methodology Differences l t t i g* l,

t ACCIDENT SEQUENCE DEFINITION i 4 BASES FOR ANALYSES SCOPE / GROUND RULES (SECTION 1.5) DEFINITION OF TERMS SUCCESS CRITERIA-Ia. e' .. EVENT TREE" DEVELOPMENT. , '~ 't ~ L: "'- - ' - i - # ' ? hl:- i SOURCES OF RADIOACTIVITY ACCIDENT INITIATORS CLASSES OF ACCIDENT SEQUENCES 1 e RESULTS

SUMMARY

CORE MELT FREQUENCY .i 9 e

N w

m e.e... . me. - o. . e mm.m.. asy w. -...... r g

~ 1 SCOPE i LIMERICK GENERATING STATION PRA S e PERFORM A WASH-1400 TYPE ANALYSIS TO CALCULATE THE RISK TO THE PUBLIC ASSOCIATED WITH THE OPERATION OF.THE LIMERICK. GENE, RATING STA.TI0N.,.. e ASSESS THE MAJOR CONTRIBUTORS TO POSTULATED CORE MELT IDENTIFIED IN WASH-1400 PLUS OTHER RELATED SE0.UENCES, e DO NOT EVALUATE POSSIBLE CONTRIBUTORS WHICH WERE DISMISSED IN WASH-1400 AS LOW PROBABILITY EVENTS, I.E., EXTERNAL EFFECTS, SABOTAGE, FIRES, NON-CORE RELATED EVENTS, e USE PLANT SPECIFIC SYSTEM CONFIGURATIONS AND DATA. WHERE AVAILABLE. e USE SITE SPECIFIC POPULATION AND METEOROLOGY FOR CONSEQUENCE CALCULATIONS, e PERFORM WITHIN 120 DAYS e, ew e e ee emmewen-e - - ow eme w=* ,i-- -g

e=e

= t

? t SUCCESS .= SAFE SilVIDOWN + SYSTEM CONTAINMENT. RELEASE OF INITAITOR CONSEQUENCES = DEMANDS + FAILURES RADIONUCLIDES z 2 IN-CONTAIN-t HENT { INTERACTIONS + i. 1 FAILURE t j V v TASK I L TASK III TASK II i ^ j flow of Information for Final Quantitative Evaluation

TASK I TASK II TASK III REPORT i 1: l l ji l' I I l' I 5 i i [ LOCAL RELEASE OF RADIO-1 I. I NUCLIDES I l., l I

I 1:

l l l g U g g b I IN-PLANT RAD 10AERSOL TRANSPORT .I i i i If I 2 8 9 10 15 16 SELECTION OF ACCIDENT CONTAINMENT PROBABILITY & CONSEQUENCE OVERALL RISK & RISK IN PER-ACCIDENT SEQUENCE / RESPONSE & MAGNITURE OF + MODEL RADIOLOGIC + SPECTIME WITil INITIATORS CONTAINMENT RADIONUCLIDE RADIONUCLIDE RISK ASSESS OTilER RISKS i EVENT TREES BEHAVIOR RELEASE JL JL a h 4 4 O 3 7 11 12 13 PROB. DEFINIT. PROBABILITY OF CONTAINMENT DISPERSION POPULATION HEALTil EFFECTS o SYS DESCRIP. - m-SYSTEM FAILURE

RESPONSE

MODEL DISTRIBUTION & PROPERTY o SUC. CRITER. ANALYSIS DAMAGE o GROUND RULES l h JL i 4 14

0MPONENT EVALUATION i

FAILURE RATE MODEL i

)ATA i

Simplified Task Diagram for Probabilistic Risk Analysis. (This diagram 4 May Also be Used as a Map to Direct The Reviewer to The Appropriate PRA I Report Section for Detailed Discussion.) l

9:. d OVERVIEW 0F ACCIDENT SEQUENCE PROBABILITIES e ATTEMPT TO MAINTAIN THE SAME CRITERIA FOR SUCCESSFUL SYSTEM OPERATION AS WE PERCEIVE WAS USED IN WASH-1400 e N0 HER0IC ACTIONS c. e- -LITTLE.CREDII. GIVEN F.0R?.0PERATOR. ACTION,-WITHIN. '... 30 MINUTES e CHANGES ONLY IN THE FOLLOWING DIFFERENTIATION IN THE TYPES OF ACCIDENT INITIATORS e SYSTEM SUCCESS CRITERIA e NEW FEATURES (I.E. ATWS ALTERNATIVE 3A) e NEW DATA (COMPONENTS, MAINTENANCE, DIESELS, OFFSITE POWER) e USE SAME TECHNIQUE AS WASH-1400: e EVENT / FAULT TREE METHODOLOGY M*

- e e-w m

-+-y-p. -- m ,,,y-,

n GROUNDRULES: INITIAL CONDITIONS AND ASSUMPTIONS TRANSIENTS 8 ACCIDENT INITIATORS TO BE CONSIDERED LOCAS J ATWS 0 END POINT OF THE ANALYSIS: 20 HOURS -- HOT SHUTDOWN .. l , ' ~ 7...,;. ...g .n. 4. SAME AS WASH-1400) ADDRESS l 9 METHODOLOGY TO BE USED', CRITICISMS POINTED OUT BY LEWIS COMMITTEE 8 PLANT CONFIGURATION: LIMERICK e SYSTEM INCLUDED: ALL SYSTEMS NO OPERATOR INTERVENTION TO 9 OPERATOR INTERACTION PREVENT PROPER SYSTEM i OPERATION 9 SYSTEMATIC FAILURE CAUSES S COMPONENT FAILURE RATE DATA: LATEST AVAILABLE O MAINTENANCE AND TEST DATA: GENERAL ELECTRIC t i 8 SUCCESS CRITERIA: GENERAL ELECTRIC i { I O UNCERTAINTIES t t

{' .t FLOW 0F !!iF0??ATION g, i GROU'iD RULES l c,., - Q.. ACCIDENT N Ck III AIO" EVENT TREES P,RgggBILITY& DEES') FUNCTIONS l s T RIA i i i i l 1 SYSTEr.S I I COMPONENTS I I I I I PAINTENANCE FAILURE OPERATOR Cor.50N r.0DE UNAVAllABILITY PROBABILITY INTERACTIONS DEPENDENCY a 6 4 o t 1 I Al n A 4 uPERATING N [Q PERIENCE O p FD#

'*W 9

6*4 6h "- p@h 4 p y

_a.__. 1 s k .e s i / i g v -,j .i r POTENT!AL ACCIDENT INITIATORS. g. 5 's. . >. r ','L... 1, j* .g .5 7 d NOR=AL CONTROLLED TRAN5!ENTS LOCA ATVS SHUTDOWNS C5 LOCA ATwS h s v ANTICIPATED RARE t TRAns!ENTS TRANS!E"T5 1 e t O AT RT ? .l ' I f J (/ /' s f Simplified Hierarchy of Possible Initiating Events for a Typical PRA g g M d 4

s

.: j -) r ~r f e se..,p-me-p e,.~ ^

~~~ ~ ~ ~ - -~ VSTEMS TAOULATEO~AS ~A FUNC11UN ~uf at.t.tuuis nitainsuro - 5 ~ SUCCESS CRITERIA

ACCIDENT INITIATOR Coolant Injection Containment Heat Removal'g i

large LOCA: 1 of 4 Core Spray Pumps Ncrmal Heat Removal 2 OR OR Steam Break >.2ft 1 Core Spray Subsystem 1 RHR-2 Liquid Break >.lft (2 pumps) OR Containment Overpressure Relief , -. ~. Small LOCA: HPCI' Steam Break-OR 2 Same .016' to.08 ft Large LOCA. Liquid Break 2 .004 to.1 ft Items Above Small-Saall LOCA HFCI OR RCIC OR f Feedwa ter Same OR ( 2 of 4 CSIS-OR t l < ADS AND 1 of 4 LpCI l l OR ~ I Condensate ( Normal Heat Removal Transient ** OR Same 1 RHR l OR

enta'inment Overoressure Relie-Normal Heat Remov.a1 10Ry OR Same._

1 RHR OR

ontainment Overpressure Reliei 1

l Transient + SORV Same Same

Success Criteria 'are the minimum number of systems required to 0

maintain the fuel temperature below 2200 F.

- Includes all the observed transients from operating experience data

.....e 1ADW ,w

a

,.( s F 1 a.,.

~

Table 1.3 i I

SillViAlly of LGS CAPA8illTY FOR ATHS HITIGATI0li l

(Alteniate 3A Modifications) ~' Transient ?-' failed Systems or functions Initiator 1 SLC 1 SLC i 1 SLC 1 1 SLC + fW t fW + llPCI FW HSIV llPT l . PuliP fit i 1 1111R . 2 fulR RCIC llPCI LEVEL B RiltillACK LEVEL 1 l r ItCIC TRIP TRIP O TilillilitE lillP 'A A A A A A il A A N 14SIV Clusi1RE 'A A A COR A A A A N 3 LOSS Of OffSITE 'A A A C0ll A A N A A A P0tlEll ^ 'lilADVElt1Elif OPfli A A A COR A il N A A A ItEl.lEF val.VC A: acceptable id:not acceptable - C0!!: Contairunent Overpressure llelief i b l

} I. SOURCES OF RADIONUCLIDE RELEASE fe NORMAL EMISSIONS DURING PLANT OPERATIONS e UF! USUAL OCCURRENCES . ESTIMATED { ~T0 BE ,e . SPENT..FU.EL,.STO. RAG,E. POOL. ACCIDENTS;...; SMALL ~, CONTRIBUTION e TRANSP0 RATION ACCIDENTS TO RISK IN WASH-1400 e OFF GAS SYSTEM ACCIDENTS e REFUELING ACCIDENT e REACTOR CORE DURING/0R IMMEDIATELY FOLLOWING POWER OPERATION s*. - e. . - + - . - ~ * -s h

i f DEFINITION OF ACCIDENT SEQUENCE CLASSES -^- e DECOUPLE'THE PROBABILISTIC' EVALUATION OF" "' 7'" I ACCIDENT SEQUENCES FROM THE DETERMINISTIC IN-CONTAINMENT CALCULATIONS OF RELEASE FRACTIONS, e ELIMINATE SMOOTHING AMONG RELEASE CATEGORIES (SEE LEWIS COMMITTEE REPORT) e I e a l

GENERIC ACCIDENT SEQUENCE TYPES i GENERIC ACCIDENT SEQUENCE CORE CONTAINMENT EXAMPLE DESIGNATOR CONDITION CONDITION SEQUENCE CLASS I e CONTROL RODS INSERTED 0 LOW PRESSURE T QUV p AT INITIATION OF 0 POWER LEVEL AT 30 MIN. DEGRADED CORE POST SCRAM LEVEL CONDITION CLASS II e CONTROL RODS INSERTED 0 CONTAINMENT FAILED TM AT INITIATION OF 0 POWER AT 20-30 HOUR DEGRADED CORE POST SCRAM LEVEL CONDITION L CLASS III 0 ATWS 0 CONTAINMENT AT TCU pM ELEVATED PRESSURE O POWER AT APPROX. 30% PRIOR TO CORE TCC pg2 LEVEL AT BEGINNING DEGRADED CONDITION (Mode 113) 0F CORE DEGRADED CONDITION CLASS IV 0 ATWS 0 CONTAINMENT FAILED TCC pg2 AT INITIATION OF 0 POWER AT APPROX. 30% DEGRADED CORE (Mode 1) LEVEL : AT BEGINNING CONDITION OF CORE DEGRADED CONDITION e e O b es-O i

m e TW W-a g-rg r.em.e4,m aw& er ea. O h INCLUDED IN THE LIMERICK PRA ARE AREAS OF POTENTIALLY HIGHER RISK THAN ASSESSED'IN WASH-1400 1. HIGHER PROBABILITY FOR ATWS 2. HGIHER RADIONUCLIDE RELEASE FRACTIONS FOR ATWS COMPARED WITH OTHER ACCIDENT SCENARIOS. 3. LOSS OF 0FFSITE POWER MAY BE A SIGNIFICANT CONTRIBUTOR TO CORE MELT FREQUENCY e 4. FAILURE TO PROVIDE COOLANT INJECTION MAY BE A SIGNIFICANT CONTRIBUTOR TO CORE MELT FREQUENCY = 9 0 we. - i e c .e-. .mw* ...as, k

-4 10 WASH-1400: Approximate frequency of decraded core condition -5 10 -6 10 10-7 ~ E / -8 10 / 10 ' CLASS I CLASS II CLASS III CLASS IV Loss of Coolant loss of Heat ATWS ATWS Removal Capabilit Inventory Makeup From Containment +y +Makes use of containment overpressure relief e e+ -..e .w

QUANTITATIVE TREATMENT OF DEPENDENCY e DETAILED PAULT TREES INCLUDE SYSTEM INTERFACES AND INTERDEPENDENCIES SUCH AS COMMON LOGIC, SENSORS, AND ELECTRIC POWER e ENVIRONMENTAL DEGRADATION OF EQUIPMENT FOR IDENTIFIED CASES OF EXCEEDING DESIGN SPECIFICATIONS, FOR EXAMPLE: 1. CONTAINMENT LEAKAGE INTO REACTOR i BUILDING (E.G., TW) (CONSERVATIVE, SAME AS WASH-1400) 2. FAILURE OF ROOM COOLING 3. ADVERSE IN-CONTAINMENT ENVIRONMENT e DIESEL DEPENDENCY MODEL BASED UPON OPERATING EXPERIENCE W/ DIESELS e MAINTENANCE DEPENDENCY MODEL e COMMON-MODE OPERATOR ERROR INCLUDED FOR 1. MISCALIBRATION OF INSTRUMENTATION 2. MANUAL SYSTEM INITIATION ~ t

t SYSTEM INTERFACES INCORPORATED INCLUDE: e ELECTRIC POWER - AC & DC EMERGENCY BUSES u o ROOM COOLING (EMERGENCY SERVICE WATER) e INITIATION LOGIC - COMMON SENSORS AND LOGIC (RX LEVEL OR HI DRYWELL) AMONG SYSTEMS e COMMON SUCTION AND DISCHARGE LINES e COMMON WATER SOURCES BOOLEAN COMBINATION OF SYSTEMS REQUIRED TO ENSURE THAT COMMON INTERFACES ARE PROPERLY ACCOUNTED FOR IN QUANTIFICATION 4 I =

t TECHNICAL SPECIFICATIONS i [ 8 DEPENDENCIES OF SYSTEMS DURING SYSTEM MAINTENANCE I OUTAGES WHILE THE PLANT IS OPERATING 8 USE PEACH BOTTOM TECHNICAL SPECIFICATIONS 8 ' EXAMPLE: RCIC IN MAINTENANCE t' HPCI MUST BE ON-LINE i l-i 9 e I I n

~ -. .s.--- -.. --.a s l a w w .~.p s ales a .a (step) sogg satu es 3 sa0 attea.ttf

4. art j

n futete g -I 3413eg ..a. .es.teg e pitei Setete t .. ~ m. e e 3eisgas w a - y m act." letto.3 u. n..= sua.ut e nue3, .em e es.ieassag

e....

a m. (seats ~-

f. tut e.e g e.o

.e W =*

Z 1345

~ 1. = se se a o -.- U 3 .a 7*. w

e SCT s,.

4 m -) .e.* M Z 31:3 e m A= e Q = ,e C.

4. art e*

w stgeasdo =. 2 3 (seets e e. =. H ease. 4 3 -e e CC .assiegas I*. .e w e. O tuttee; .e La I^ : see.ateau.3 r fututrwes ,. - e o . swiegas I i ~.a H 1:41

2" sujete.es

2.

I cc w

e..s ess,as j. c. = de44s o .s.2 gs.s g o Isa .e a .e n. ,10 ~O e..setegas

g.

seu s ases = a. g ".= i * .-e toepumpeg ~. o = ges.4a et o . p.stan.: 3 e. y e.eteds t e z ~ .e. r ~ x .1 i s. ~; e 3 u J u - g .y. 8.s. 1.s.. 2 3. e. 4 O as 9. O G e S 9 4 .e. y g g A 4 2 s u .a. > a "[ ,3 ].., .=. 1 .s I 3 s.3 d 5 .e a. 3 - s.. .e y y .4 e.e e 3 e, e e. 3 p 3 s e e 3 e O a.*A 8*. e. g gg g W$ .e 4 a. we e e .4 A d 4 3 y e u. 3 e. ,,m, ,o,g.,. e, 3 e gg Q .e g ,y se y "T n e'. 9 3 p. =* p. 9e

s..

3. e U

U K.6 Q .em a 4 j

3 g

E W. J e. e 4.0 .S. .#. Q ug 4.8 d D. g G p. e 9 W eog. e o f a

I COMMON-MODE TREATMENT IN WASH-1400 1. DURING EVENT TREE CONSTRUCTION e INCORPORATION OF FUNCTIONAL DEPENDENCIES BETWEEN SYSTEMS IN THE ACCIDENT SEQUENCES. e DEVELOPMENT OF ACCIDENT SEQUENCES INCLUDING CONTAINMENT FAILURE MODE DEFINITIONS WHICH INCORPORATE SYSTEM AND ACCIDENT INTERDEPENDENCIES. m ewe.--

- - e, e

-==*ww. e eee -ew-*w-- -**-m

- ,-N r--

w~ ,..+, p c.-

.,.,,e. e 1 COMMON-MODE TREATMENT IN WASH-1400 2. DURING FAULT TREE CONSTRUCTION a RESOLUTION OF FAILURES TO A LEVEL SUCH THAT COMMON SYSTEM HARDWARE WILL BE IDENTIFIED. e FAULT TREE CONSTRUCTION WHICH IDENTIFY HUMAN INTERFACES, TEST AND MAINTENANCE INTERF..CES, AND OTHER INTERFACES OF POTENTIAL DEPENDENCY. ,m. ,e oo>-

mee-o

'

.....M

i COMMON-MODE TREATMENT IN WASH-1400 i I 3. DURING FAULT TREE QUANTIFICA-TION L e PRACTICAL DATA UTILIZATION, WHICH INCORPORATES t UNCERTAINTIES AND VARIATIONS, u e QUANTIFICATION FORMULAS WHICH INCORPORATE DEPENDENCIES AND CONTRIBUTIONS DUE TO HUMAN l ERROR, TEST AND MAINTENANCE, AND ACCIDENT RELATED ENVIRONMENTS, l l l l e MATHEMATICAL TECHNIQUES INVOLVING BOUNDING l l CALCULATIONS AND ERROR PROPAGATION CALCULATIONS, WHICH SERVE TO DETERMINE THE SIGNIFICANCE OF j POSSIBLE DEPENDENCIES AND SERVE TO INCORPORATE i RESULTING UNCERTAINTIES, 4 t

~' = .....i'- = - .... = - -.. - 3. i COMMON-MODE TREATMENT IN WASH-1400 4. DURING EVENT TREE QUANTIFICATION IDENTIFICATION OF COMPONENTS COMMON TO MORE THAN e O!1E SYSTEM BY BOOLEAN ALGEBRA TECHNIQUES. QUANTIFICATION FORMULAS'WHICH INCORPORATE COUPLINGS e AND DEPENDENCIES ACROSS SYSTEMS DUE TO HUMAN ERROR, T & M, AND ACCIDENT ENVIRONMENTS. l l GROUPING OF. ACCIDENT SEQUENCES OF SIMILAR OUTCOME e AND IDENTIFICATION OF THE DOMINANT ACCIDENT l SEQUENCES USING DISCRIMINATION AND BOUNDING l TECHNIQUES. I l.

i COMMON-MODE TREATMENT IN WASH-1400 5.. SPECIAL ENGINEERING INVESTIGATIONS i l e INVESTIGATION OF SPECIAL, SUSCEPTIBLE ACCIDENT SEQUENCES TO DETERMINE ANY REMAINING POSSIBLE .I COMMON MODES INCLUDING THOSE DUE TO EXTERNAL l EVENTS AND COMMON COMPONENT SENSITIVITIES. I ) e A SPECIAL DESIGN ADEQUACY TASK TO INVESTIGATE i; COMMON MODE FAILURES RESULTING FROM EARTHQUAKES, OTHER EXTERNAL FORCES, AND POST ACCIDENT I ENVIRONMENTS. I ( t I FINAL CHECKS ON THE FAULT TREE AND EVENT TREE e MODELS FOR MODEL ACCURACY AND CONSISTENCY. l l i l l 1. e l 3 i 1 . ~. -.... +

e . m. -~* .i (.* ( o -i ' LEVEL OF DETAIL IN FAULT TREES i e FAULT TREE INPUT LEVEL ITEMS s i . GENERIC FAULT TREES e I ~ EXAMPLE OF COMPONENT LEVEL INPUT e i i i f 1 i k l P i 2 I k. t r I s l \\ l i r h w,,m w,,-~vm-,. - - - em m we -o, now.r~,w-~ -es r--,.,e- _ v,-m er.,w ev-,n.mw,,-~~-w-x -vv--, w-4 V'

i s RESOLUTION OF USEFUL OPERATING EXPERIENCE DATA APPEARS TO BE AT THE COMPONENT LEVEL e COMPONENT LEVEL DATA USsD FOR THE LIMERICK QUANTIFICATION INCLUDE: MANUAL VALVES MOTOR.0PERATED VALVES PUMPS TURBINES INSTRUMENTATION SENSORS / SWITCHES DIESELS

0FFSITE POWER (INITIATOR)*

MAINTENANCE

OPERATOR ERROR (IN GENERAL)*

1 s SPECIAL TOPICS e' e - am,e,e ,,s .,-w ~ - -

i ,1, TUR5!!!C FAILS 70 STAAT CR RUM .TU...cv. on f I 6 1 s Fcarica *Arca!At et$ ten stAupaCE!NG P!t5CNNEL tance ca CONTAr! NATION (RROR FAILURES FAULTY 790Ctcuat$ 0 0 0 0 B(A41NG $ HAFT / COUPLING TUi8!NE BLADE FAILUR[ FAILURE FAILURE O O O Figure 3.16. Generic Fault Tree Model Displaying the Potential Failure Modes of the ?.trbine (RCIC or HPCI) I w n-

5 t [ I [ l l A PtrW FAILS /p TO Staaf 'U" .pn..oug y l I I I e t 'I'j'" M $ dart /CNPL!w CI5!;N SIAL /Pacats 7 rggg ( Esaa imuuS rmust cm.nari. ,,ct,,,, O O O O O PJW Clacuti BatAtta ca g(Asig Ltaaf

Odat FAILCD Als/va7Ca

..t Aa ggttpgALS IOUNO LCCAL ELECistCAL FA!Lyng pgtt;gc5 F A'JL T5 O O O O O O Elttedted not bened sa data. l j, Figure 3.15. Generic Fault Tree for Pt=ps, Including *.he Dominant Failure Modes of Pump Failure Cbserved in Operation i i l i l i t 1 - ee er ..oa ,m- ^

I .. ~ l 1 i i i r i r 4 4( e e f f f Ci me l 1 i I l l l l l l

) -a _e. --+---en._. .em=um.-. -e-. i 1 i 1 1>l Ssetr \\

=
- et N

e as 4 mm - .wwww Lu c-e d zI l

&e

^

da W-m Nh-+@ m ae me

m

-e--ei e w 4 O 1 % s.,- 9 ?t 1

O e e e-o - -===-M*else-- N wwee .,m, ee .,e, e-es ,e-1 l l l I psq.g r 10 ?I i l l l l l l

i s DISCUSSION OF SELECTED ACCIDENT SEQUENCES T QUX CLASS I T T QWCO) CLASS II F TCC CLASS IV pg2 1 QUANTIFICATION OVERVIEW e RESULTS 8 .--...e.e

v CLASS I T QUX o T a TURBINE TRIP INITIATOR WITH SUBSEQUENT FAILURs TO"-} PROVIDE ADEQUATE.COOLANTSAKEUPCAPABILITY A CONTRIBUTING SEQUENCE TO CLASS I ' -s. e - It!ITIATOR: TURBI'NE TRIP (T ) T LOSSOFHIGHPRESSURECOOLANTijAKEdR e FEEDWATER UNAVAILABLE (Q) HPCI-AND RC' VAILABLE (U)^ e INABILITY TO DEPRESSURIZE RAPfDLY " I - MANUAL DEPRESSURIZA, TION (X) s k. Wr-- s 9 4 \\ w + 'e i - ..ey

%y

+

i ,I,. e i e 5 =, a a ~ ~

3 =..~'-

~ - ~ w _.,. s ~ e e e . ~ a, a. a ,a ..,. a., -= ~ w w .a .a. a = J t o = 2

o e=-

, *. 4 ~ t. 2 '.o. l e_. .e w-o ..o l s. a E

7. ~.5

-7.,. 5 -.t *'.- e- -t=5 T... ~. .-a E 75 E i. ? -aa ~ . -~ e s. c M ~., .t j 2

v. 2 a e ty 5

.z b& & 5 a* E E E l' E 's L-8 P-M M M M M W M M M M M M g':I r. ~, s-2 r. jg 3 -e l' a ? r a a I O W .5 3-a. M.** 8'* e. g u. d 8 O g .1 m .r 'e-2 -4 e ~ g I = g 1 .S h-3--) v. w.- a. [ ga-au O h> ~~ =.J a, ~. s

2

' 's., e .Z.- r, pg 99 = .a + D tsfO' p b ..e pp b .E. l .~ i 8, I 94 s e.4 t S a. %e- -*J i> i =. s

O E

8 8'* 6 %w. ..8 2 E.8

=

.E. :: -e

s.-

i l t w x-a t -e v. 1 q eI '.: e = f") een E ~~# e l 3 p er..4 3 a* e e. a e" bs

m..-

p a i i C l l l / ".. ) $. 1 O O MM WM .g a1 -a-

CLASS II T 0W(0) p MSIV CLOSURE INITIATOR COUPLED WITH A LONG-TERM PROBLEM WITH THE POWER CONVERSION SYSTEM. l l A CONTRIBUTING SEQUENCE TO CLASS II e INITIATOR: MSIV CLOSURE (T ) p e FEEDWATER UNAVAILABLE SHORT TERM (Q) e '- CONTAINMENT HEAT TREMOVAL (W) INADEQUATE t RHR 4 RCIC IN THE STEAM CONDENSING MODE POWER CONVERSION SYSTEM l. l-l:. t i t l. l-m

9 I i .a. ~ .= -a v s.- .x 5 3 J ..~ .= ~ ? a - .2 . 'o. .,'e . o 'o 'o =* : =g g*- 5 *. 17.* 1.. = f.

g 3.

t, - -a . y -a a 2

l;l 2

E' l' E E

s 52 s

'a & & 6 a 2

3 a

a. r-.e .n- .m- .s-v. T 2

g 2g s

o b o e g m a

ty c

g _o -4 .m T y c 'o o sw a. c.s - * ~. c

s

C

1; u

j4 g o -2 ~ mc ~' a 3 - o- 'a = s2 m s.I ww -w.. s =a 2 o o s. 6. cr t 5e e. 7 2 w a 2 2 C Q i

== Mo

~

m 2.. ' t. 2 C N - ~ . t. = l o c~ 13: uw i ~- ~ l ,1 : - L a l .e - i o 1 x

2..:

t,2.-

,~ m o i I 3 - m 5. 33 e6 g -~n g-l --2. ::. w-w m. i

3 g.

N 3. -g W C EO-6 3 w o.. 5* m t I I l 3-26 1 i

4 CLASS IV i 2 T CC p g2 MSIV CLOSURE INITIATOR WITH SUBSEQUENT FAILURE TO INSERT CONTROL RODS ON DEMAND DOMINANT SEQUENCE FOR CLASS IV* 2 e INITIATOR - MSIV CLOSURE (Tp) OR TURBINE TRIPS LEADING TO ISOLATION EVENTS e MECHANICAL COMMON-MODE FAILURE OF ALL (C ) g CONTROL RODS TO INSERT e FAILURE OF BOTH SLC PUMPS TO PROVIDE (C } 2 BORON INJECTION e HPCI REMAINS ON IHROUGHOUT DURATION (U) 0F TRANS!ENT 4

COMPARABLE WASH-1400 SEQUENCE'IS TC

-*"'*h*N=u"O* +=e = ,.,.w.

TEllTATIVE REQUIREMEi1TS TO BE IMPOSED BY llRC ALTERNATE 2A ALTERNATE 3A ALTERNATE 4A ARI ARI ARI SCPAM DISCHARGE SCRAM DISCHARGE SCRAM DISCHARGE VOLUME MODIFICA-VOLUME MODIFICATI0i1 VOLUME MODIFICATI0ft TION RPT RPT RPT-REDUCED VESSEL REDUCED VESSEL REDUCED VESSEL ISOLATION ISOLATI0f! ISOLATI0fl PERlilT FW RUNBACK PEPJili FW rut?3ACK PERiilT FW RUiiBACK CLOSE CONTAI;iMENT CLOSE C0tlTAli!MEllT ISOLATI0i1 VALVES OR ISOLATI0i! VALVES OR FUEL FAILURES FUEL FAILURES AUTOMATIC SLC 85 GPii AUTOMATIC SLC 300 - 400 GPM OPTIMIZATION STUDY IF 4A IS NOT PPACTICAL -w

e o m, . = = n e 6, 9 e I e I4-e 9 t 3 3 w S - a M ao " O m. = .e b s e. m - Q y o ae .e ~. I"3 -= w ua kpI l e 7& E" 2 "u 9 Xw=3* a -g 3 ** a=- w e w if ( ** g c e~ M== w ..=. m e as. e a= a --a g "3 T. o e = s o e n 2" 2..", ba4 < L ~. e.e. 8 D id hQ se u

== g *a. GQ s.s ao be = W f.m. M M -e ad.e to M as 2 C a.'*d e a g . tee w w w w N. N N w5

A

=. -e N N ox is e e m a m C. g', S

i j

b 4 me 0

A 09

.= 8 5C* e g i 8 2 12 = u: m 12 20 a w= <a - 3 4 .s= e.= W g e c = s. ~ 3O 6 8 53 g ~ , o a = o U sa 8 e >= v g T,,e,, O C I ,M Y ? s au 2 . I.:. 3

a. 2

= o.- w =w e > IJ 3 o=

gm se 9

AA a a. ,e. -a .e i o4 I 4. -4 g 1 U h. es. ek c o

a m

n. g. w = -* y .. n 3. y a _E 6., 4 2 =. ~ A 5 2., Q e -== g *8 8" E M e et. .h. -M 2" ~ w i j 1 i { 1 3-59

4e.eN.ae.$

m__

4 ' r o .i. g se s r 5 m .e ~ 3.a. g .a. \\ . a. a $,a,.. w = 29 .s.35 .--.,,.,.,,e .a a,e, n a e. w

o. 3 -

e. 5..- 3-s g w -w o. 22 ?.,g,.,. s_ ~. ?. s u*.s s =. ~ ~- 1 . ? *= .?.I.

z :. a-

_l -i s~ -..g ~_o ..o o a o. o 2 i. : e t. .o- .o p.i 33

5.

1 *.2 3 3 3 " a 5

g g,3. *. *. *.

3, :-

1 g, *.s a.* ~,

5 i "' 13

.'i Y".~ w1 *.4 7 ?*i12234 L a ai

.:4 2.;

y-c. = ~ e o f 2 8e ~~~~~~~I. 3 s e. a. a 3 S.i.a f,,e".t. t 1 1c 4 ;. z, ~ ~ ~ - 1 m.'.s. lt "a k Y *z z.-.z ".z%, T T 4 x, % T "r s 1 1.y ~ - w w - .I,t i o w w w s e w w ~~ o o a c 6 e 3 aa 2-g y q vs ~ c g a s = c. s- -s ~ o m ~ a- = o. a s

== a. o. c o 50 w b 'o 'a .D 2 5. _5, a 0 = =

- 5 "*

%ggw -2 a g ~ ~ "I = 2 m 'a b 'a b b 2.- ib;3;, 3.CC=. = = ~ ~ i"2 ~ ~ u W l ae 5 w 2 2 .o. = x .#t a o 2 u g*= 2 a o g 8 s =5s I uo

n -

s O

.e..n s 22w c: a

- 2 8

I l =3 W a a es 5:8 b-u a w c.o a ~ n ~ O O 4 E wu ~ w 5 1 .:n 5 ~ s e. g m. p = w en 3 = c. I u s.. - s . =. S ~. m .r, w -w o. ~~ w o. o. 3 50 4 1

.=- RESULTS a COMPARISON OF LIMERICK VS. ' DASH-1400 FOR POINT ESTIMATE OF CORE MELT FREQUENCY 3

SUMMARY

OF CONTRIBUTING SEQUENCES BY CLASS e OVERVIEW OF REGIME OF INVESTIGATION. (AREA 0F FORECASTING) O .m .a -en

.. _ ~... 8 l hA'#.1400 L M ttcg S.5 *A g 2 / i f'NV f Wf g TV LC55 Cr CMa Tc,*y " [VIWS CTFSITE 77.t2 TC AT*d $ t5r i CORE MELT FREQUENCY (PER REACTOR YEAR) Figure 3.5.4 Comparison af the Contributing Accident Sequences to the Calculated Frequency of Core Melt from NASH-1400 and the Limerick Analysis. 9 ..y. .7 + r -.-.9

i i i ~-

{

l -5 e.I - i. }:e.i. I 5 %.%. %-N s -_=%- e N.=-. il I E b%%%%N l 3s I N%wsNwwwww ":f l ,I o v !l LNN%Dwh> 53 2 ~~ 3,. n. =l = siri kxxggggggs f:t z-q y

q 4:e EBal E

l%%%N%%%N E ai j N =5 a g e L%%%%%%%s ! g 8 o l C. .l = LNN% % % %% %V !,_? E l m ISN%%%%%%%%N1jE 5 5 3 uu l 1 k%wgg%N l_ y a C z-1 8 Tb C I L% % % % % %' s Ii i 1 o O a 7-b I\\\\N%%%MNWN 2 0 '5 o r b%DDDDDOYN [- 5 1 =a p_________________________________ I LA%%%%%%%%N E,i i l m- = x O .e \\\\\\NN\\\\NNNNNNN N bbo c l -r u = l a i g kh%%%%%%%%N 8_ c bY%YNW%hWWNgMW) f-f I i i l%N%%%%%%%%%%%N 8 B. 3 I l t t t e T T i ? ? E 2 2 2 2 2 (seaA ac peag aag) A wambaJJ V

'd v. I WASH-1400: Approximate frequency of decraded core condition -5 10 x -0 10 m V = v 10_7' g h / ?] 5 8-b 10'O - ~ 10 CLASS I CLASS II CLASS III CLASS IV Loss of Coolant loss of Heat ATWS ATWS Inventory Makeup Removal Capability Frem Containment + +t'akes use of containment overpressure relief O 9 4 g - .. ew e, e = .w

.e ~ ~. .s I .l 3 CURRENT USA EXPERIENCE SASE _g 7 E 10-2 3 a:g LIMITING ENVELOPE u 5 10-3 DESIGfi BASIS REGION s [ 5:s 4 3 10 i 2 SAFETY MARGIN REGION 10-5 _ l I RESEARCH REGION i l 10-6 i ~ AFPEN0lX l NO IDENTIFI ABLE SITE EMERGENCY l i PUS L!C INJU RY PLAN CONSEQUENCES Interi: e.ap of quantified safety regions l 9 9 4

o .I Ua eoC G W CJ G = v =6 { GJ = C 3 c CJ C== m 3=6 g-- y

a C" 3 ed %

C C wC 23 U? m O C vuo .g a 6 m e c=.cw y 6.= a o .a ~E 2 2 X,3" a 4 = c u O T cJ a a a = = = =d y gg 6-g W W 4 eC a g M M a C U w

'8 o

t t t m<ve% e CU ECC .) -O I,=,3-- c a C. -a U = 3 13 4,,. =- as{

]*

g e. m >mu c l_ eeen- $wg bg o 3 o% -a c. u. o 3 *. a a-I*a o =.= o= o I "'. $ 2 3 o e a ou 2 EI = :=. -o = = = - pwa - ~*E2C e - 3' cum 5 e. a e = .ie a Obme u

= = =

I oh %~* C >= 0

C o

t g-

=.

a:- o p =:

um . = =*$.- a 9*=*g9

j a=-

w e c.

C N M C* 323 = as " o e g c m, -.

- h n *$

3-G 7 -~ -m. m : _= %.a e = W 3 C a. ~, * * "6 e"* - p g aq. = = J

I e 3 D **

v 2

C4->

=

* *. d Y

dIZ 3

2. 3 6 6d M

-a--- - s 3 b"3 i _I e w w

=

O 41 as 21 83 ~e=w amo $1 5*% .=. 2 2, 3 2

3."

= = -2462 a-- m 48 f - W es a Q== e 8 . dup g e e e b a a ce = w g I 3 g "-= 3 3 -=ge< 2' I z -w 3e4t ew .Q 4 g

e e

4 %d 1 1 I 1 9 O - -. + - - - - . ~ c

J BRIDGE TREE EVENT SEQUENCES, IMPACT SEQUENCE FAILURE MODE IMPACT TIME FRAME NONE OK NA MODE 1 COR FAILS DELAY CORE MELT 27 HOURS MODE 2 COR FAILS DELAYED CORE MELT 27 HOURS MODE 3-COOLANT MAKEUP CORE MELT (SIMILAR TO 2-10 HOURS TQUV) MODE 3/4* COR FAILS OPEN CORE MELT (DIRECT 2-10 HOURS RELEASE) MODE 5 LONG TERM MAKE-POTENTIAL DIRECT RELEASE 2-10 HOURS UP FAILS AND FROM CONTAINMENT FOLLOW ' CONTAINMENT ING CORE MELT INTEGRITY FAILS MODE 4 IS TREATED THE SAME AS MODE 3 e =o.e-e.- 4-.- _%.w e

~ 8 e t t. Table 3.5.9 EXNiPLE SMARY OF TP TIPE EVENT SEQUENCES WHICH ARE PROCESSED THRCUG'i THE BRIDGE TREE, FIGURE 3.4.12, TO CETE.94INE THEIR SEQUENCE Ct.ASSIFICATION l 'CU~iM rit:i,0CT (7!R 3DC".2 ftA2) f yeg$Ngf cantsuta is m os3 xs .c.t tie.? CE

ce!

NE limCZ :DC:2 ?palt T t,t et/ts3 g l C*t35!!}C:.A:3 !!! ! ::A13 Y ":::t t ' (2 a13-4) g,,,gg.7 5.da13*I CI 1/3 (2 810"$)

eilli'lle g

rA W3 (1.3:13 " } 3 1.I.813', c;C 3/4 (t sto'5) y 5.5s10 11 4 @t1 (Z at0*d) 3.5s10 = y g,3,gg 4 W E t/3 (2 a t34) aegligthte W3 (1.3 10*#1 3.sst0*IO g,3,gg.11 W 3/4 (1 a10 5) -c:t 1 (2 sto*3) 3.:sto*# %>w g,3,gg-5 . W 1/3 (2 sto*I) segt (gt s t e 4 W3 (1.3:10*.4)

2. h10

,3,3g.33 W /4 (I atQ.5y CE1 (Z s134) 1.h10** g,3,gg TC t/3 (2 a10*I) segl'fgtste F St3 (1.3 s t 3"l 1.2.10*IO C E :/a (t utg*5) 5.!a10-12

X:t 1 (2 a134) l1.; sic'I f

"4 ggg4 CC t/3 (Z aLQ-53 I,,g ,;, g, g, g g iJ -Oct 3 (1.) t0'#1

.5st3*I g

m t :/a (1 sir 5) i.utr" 4 i m u. ! 5.utr' Ilasir' ! !.:,tet: i t .It =st :e .se. ca t ce .,e

a e,e t :r,, c:au s s a in=ia. ne s.,e ic.e e.. :.:. - ci 2,.i.es ce :,e ou initutor c= ia.i2 ee aiiare u -,..e e.c f,s ::n:u,.e-t.

r., cis :artin'ae :ase care *s ass. nee ta te suff ttiet ist:ac:1,ity -viesse< : ee :sanimet

==s:cere :s :ause ce C 2.al.es :s se tatariscae1 closet. Be targa set -setae '.r.A seneme.s een ::nti-:via st e=:tly s "* ass !! ans :s not : ass =r;uga ce artage tree. "sse 5 v'uts are :mtr*:vtion :s 'Cm W Mjp mitpMe MWu J m mee 3/4 oalatim a we I taclases re made 2 (i.e. s% 1. P-sce 21 f attur-s staca cese as.e :M sue av. alt tat *ve effut sn ::etaarnent sna ac::: ant seQbeacet. - Be sace :ssigrat:es gt es in :ais unie are.w secueace sesignat rs f-se rigure 3.5.5 saa are fsr-ies sy ee areauct af : e event arneoastlites assac':stee.ita tse secuence usicastart. For enamole, tse are:sottity of seeuences gestgrated, ode 1/2 ts esiculates as

e arseuct of :.9e :cotant11ttes of t-eae 1) a (soce 3 given

.,oce 11 : (not -ode 51 eni of.aics is taken from Tiele 3.!:33 l

e av neer of suca se:vences is toen multtelted tt+es tats ;rseuct

t. :ete-,,ae :3e..iue ino.a in t*e aeo.e :a.ie.

~ Se ac=1:ent seeueace :r :astit:1es anseer*N ta :ais asarmie unie are artves for sesveaces *nt:1stes a:ove !!! se.ee. Se.aiwes assear--g *= astes 3.5.4. 3.5.5, and 3.5.5 are t3e so af :rtesf eats l t si tt s tre "ese 411 sows. L f ? \\ l i j 2-107 l l + -..a

i + i 4 LOSS OF 0FFSITE POWER e LOSS OF OFFSITE POWER INITIATOR (T ) E e FAILURE TO RECOVER OFFSITE POWER FOR 4 HOURS e LOSS OF ALL DIESELS e FAILURE TO RECOVER DIESELS FOR 4 HOURS 0 DEPLETION OF STATION BATTERIES i 4 i t i i e s j r-i --4..--, , _. _ _ ~. _,. _.. _. _, _,. _, __t.

i I I ,1 .k I m Zo s -o F-w a-o w <w Mu M <~ o r >= F-a w ce Z e a ww o rZ He N >- w o w I - o ce c. O s= Z w ><. r >=. o,- os w oo ve - o ce az ce c. u >-zwm = = = u p # N ouz o <zo i-o ce o c. m a-rs ce u w u- = < >- o - .o-a c ce o cr e o u - o w ce c. w 5-e- - = cr o z >- a u c >- z w = wi-w <zz-o c. =< ce w ei-o_2m w H ce < > >- 2 o ze o w oo u W i-- w Z - Z / m e-a c. z <-o -m z e<- az / = u w w,- 00 c. O wecoco / u o u _a >- u ce aa H / mo g OO / w -M o c: u ce ce / mm za ow / 3 W A ~3 u ce c. w mz a r o ce o e a w-oo< ce as __d / A < upe / m -H

c. - >

e ce mm =2< / o o mo= / =

c. H A

/ w u_ 2: m ~ o / s / O c: w / vE / D / m / / / /o' m ~ / w /- m 'w I / m a ~ / / E I I o o o. o o o

o 100X3YlE NOI1Y15 3*I
E3M0d DV 'E3A033E 01 DNII7IV3 ONY 'DNIS07 30 ADN30b333 3AI1V738

-,m.- .~

i QUANTITATIVE EVALUATION OF Tile TIME PilASES OF Tile LOSS OF OffSITE POWER ACCIDENT SEQUENCE I t COMMON-MODE FAILURE TO HIGH LOW FAILURE OF TOTAL TIME PilASE ACCIDENT DIESEL RECOVER PRESSURE PRESSURE DM FRQUENCY PilASE - 0F ACCIDENT INITIATOR GENERATOR OFFSITE SYSTEMS SYSTEMS GENERATOR (Per Reactor SEQUENCE T FAILURE E POWER *t U Yt REPAIR Year) PROBABILITY 8 ttt 1 0- 2 hours 5.3x10-2 .66 8x10 t 1.08x10-3 1.0 3 x10-7 -3tt II 2-4 hours 5.3x10-2 .35 .15* t 1.08x10-3 .66 2.0x10-6 i III 4-10 hours 5.3x10-2 .158 1.0** t 1.08x107 .47 4.2x10-6 3 IV 10-72 hours 5.3x10-2 .01 1.0** t 1.08x10-3 .2 1.1x10-7 Probability of requiring ventilation of HPCI and RCIC rooms coupled with the probability of the operators establishing a natural circulation ventilation path for these rooms. Conditional probability of successful operation of RCIC using manual control with no ~ power (DC or_AC) for-times greater than 4 hours. I t Because of the redundancy of the available low pressure pumps the dominant contributor to 4 I the low pressure systems during a loss of offsite power is the comon-mode failure of all the emergency diesels, tt No AC power required for llPC1/RCIC operation during the initial 2 hours following the loss of of fsite power. l

t Probability of recovery of offsite power is derived from the data analysis performed in l

j Appendix A. l u i ttt 45 Min, used as the discrete assessed tine of recovery. I t 1

t s

w w. *. - +

me.. * ... - =... ..w.. . ~ m.., 52-C O N .e -.=== n I e e 3 c,. w

== h b f E E us es4 Q 3 3 ** e#* 86 G

== e +

== een -y

~ e m

g e =, 3 e e 3 80 em AW Q eI - se e O me enas-w T F*9

=

'.$D E en = % e 33 em em 9 gg - ens f 9 4 "C"' b an.s e

em se-4

.., 8 m .,. R

I 2".

O b f"* e== 4 666 dD. O e-o.E 4 4 9 2 g ur - at O m +

8 "O

e W% 89b D 8 8"o 3 m 4 m er e V g

== 5 O 4 saem 84 - 4 4s 2 D . =a g yg% a er as O

t-e

== m e C o -g"e

==# 3 84 I 9 =* y m .5 R2 w e O Tm- =

. yvf gr.

4oc & G4 muum m P g3 \\ =e A O M 4 m e== m ie an p ya F

== 9 FW

a.m 8

M 4" m h Q 4 8 se sum em ( F se -- 5 3m gps e p

1. fFG 3 8*

-%O* em 4 M N 82W R Em@ e .M. s e (m -M w em W E. D} C m* o .o.n - o 9 9 e.se e e M., 3 4 E g St t#b E &==*. e 8 p 4 4 g % e.o. n = g@ me es $he f* 8"9 99 FR 8R P Wh P M Sh Pt fM Pt e

M 2

9 9 9

p C

C C C E D e.8* - EemC-S C C C e"' E

- a sg o

as ag gi e M ge 9 m 4 E e OC me as e e 3rs e 's esp

== M I 9, g an O Q d'b

e se E

4.'9 ~. = w,. e .. =,. E, c.,. m,. -= w. E. 0,,. gE C e 3 m a a O 3 e a y== 3 3 m _ =- .i C em E W se = C"" 9 =* ** *

== a- .* eC em y, a p K gg en en su en yt -p e>4 s= % a.m em 9 O C Oe 9 Oe C=

== 9 em .. M e-CP *

r-O' 9

.e 9 at O, 9 3 486 aut F the P4 s - sum O MO g etJ -f 9 w w ,b8% 9 'e f5 O aun h aug n n .**ame me = .'t m e as

=

-.=-. .a** a a .e* EDDM2 er-- = $m-pm .s. re#, w = a#-

088

= m a e em e w w 0 4 88 US = w e w t y 84 D w 0 f 6 M es O ew em .m> m = en==sen me w es* Mg g C 1 . p e an* e 4 e -~..m--.s

p r.ii.a;.: G:.3 (as! 0F RICvt27 'J Rf Cvt27 0F '21*:1ttfF AI CYt2f 0F uts 1C';;!9G

1751T1 0185;*E OFF51TI Oi851*I 0FFl!TI 88t!I;;2(

U IL W N O CT I.%"AfE3 ;

M.G 77.G

- 4u M.t
77.t2 575 T S iiiiL Sl%DCI Sl%DCI'

117

MOR 3 73 2 *R3. f *3 a ts, a ;3 ;3 mains f3 is 4./o ::ct'.3

Utu st? tit U* OAT:2 n :3Ast(ITY l

1 l i, l i!) l (4) l (133 l (241 l 1 1 d, i V i 7 G Y t31 t g

lZlU 01 g

I OE ?g(2)uvg . g ?g;l:W.2 3 'l3 T (8) 01 g .. g(aluv... 01 l g f !4)U 01 g 4 f ;*)W:2 2 s13 g f (13) t g f (13)U

K 3

g 1 T tictuv 01 I -g g i-- r::::w;,..uio

g Tl:8) g f ( 41U CC e

g l T (24)uv. I g Og r,::.>2, :.mr' 1 @ :"us 3* sac.1 trs**f e 1 ta "$;7 c!:sure 'as tistan since it ei*tt:1. elf is as *5IV O!45=rt.aen sffstte a n-er e t Ms t:m. Ficure 3.4.ab Less of Offsite.:cwer Transient Even: Tree (T! e-Fhasec Cecian: Injection) As seen in the time-chased event t: se and Table 3.4.1, the time l

erieds of hicnes: ;rctability of inadecuate cccian: infection are the ;eriods 2 - a hcurs and 4 - 10 hcurs.

9 2-33

s 2. Tne HPC! and ACIC systams recuire cum; recm cooling if there is a loss of offsita power for greater than 2 hours, or battary charging for long-tere loss of offsita pcwer. (Neither of these appear to have been included in the WASH-la00 model.) 3. Tne anticipated maintanance unavailability on diesel generators may be significant1y' different t::an that assumed in WASH-la00. t:ss y ste:vur u at::vur r at::vur :r

FFl!TE CFF51TE OFF5ITE CFF1!rt IDM mu ou ma nu eu::ar:a

3 ::u.a

.4 m .is ans er: c as D DE g a 4 25 g ;$ <xlAS 't s E ib r,2 ce l tg Figure 3.t.2c Time phased Event Tree for Cal'culating Containment Heat Recoval Capability Fo11cwing a 1.oss of Offsita Fewer 3.4.1.5 Inadvertant Open S/R Valve Transient (See Figure 3.a.5) Exami ation of the WASH-lac 0 analysis, and a review of new n c:arating data, has revealed an accident initiator ;reviously c:nsidered unim:ortant may result in a group of ac:fdent secuencas which c:ntribura

o calculatad risk. Tnis initiator is the Inadver ant.0:ening of Safety Relief valves (ICRV) during full ; wer aceration.

3-25 t

e

5. '

am=~ O =~= w O 9 W e e e = po w .e .e e. = n e e 4 .e e . - a 88 **a =5 3 w w 3 5 -5 .ca LA m e a.h 9 6 O ds'lO. e e dEh a. 3 .m'a 'oT 'a .o ao 4 . ~., .'a 'e ~ e 'a-

~

~

- U o

o .e. g 's t, 8- =.. =. e w a = 4 . =. e e 8 .. a.t =. .e. d m. g e a g = o = e e 2 =d d e e = =. r. g 4 8m e - o =. ee Im.e

- 9W g*

e. 8'8 #8 8'**EF 5'E WN ~ 48 p 84 84 El" I t.d. e.m. O D W 4 f .E. a a 5 3 3 3 .A. e. 4 3 3 3 =. m o 3 3 g O O 4 e. am em o s. 4 b E w d M e=E E E ad e8 s8 s.E o.E M p.E sh 5 mm 8m em se en am -3.. ~. .~. ~. W., o-o e o o o go 3 e e e 8 e e g ag ud O +3 W = w e. o ,o w - de o.o 2 e 2 e O di an. em m. w a .m< se O 'o 6, = * * .e. E -2 = a gaf M gu2 o s ww- ' S et e = l al ~ .a

=

m me i: 'S "2 d I 5 e = 3 ~ W I c.2 o 4 4 .k. M b8 C

8 EP 3

8"" WS ~ e3 ed 3 O e, "g e, .= 48 8 w g I as

3"

.2 j g.m. q

g C

~. = a e C ~3 I 5" I** if 3. N o-w 46 ""w

*D

=3 Y.,.e. I. 2 m

- 2

.o es e4 et g 3 .a. .e ,= y .& =..= w 3 y 1 i I 3-M E

1 t I 1 e. a e 3=.33-I a a a 7 ;;:..g e -= j ::

4 4 : 2

f

-c 1<s.: w.

: ~ =

a a r a.a 3 =w o 4 w w 3 s.- s.- - -e ~., f= ?.. . = 2. *. 'o 2 8 .3 a

1,":

2. 's g -

.g

-sf .g c 3 :.g . - = = a

{

=

27e = g ~ ~ 5 - ~ 6 a a . ~. e .a. a - ~ s,ec 8 o 4 w y- =3 5 I. .E .E L. I s-5 * * * * ' ' " * " a Q a a a a e s a a.- c a p. 3

=,

9 C 2.,. g ay = 6 e i ag 4 Se I Y e a. au e w.= 3 .a r g 0 4 m 2 .f. Y U 25 I =: ?. a. 33 3 ~ M. I 34 w ~. 23 m ~ m .e~ T I C .8 3 w.s f .u.6 =. g S w= 3 8 i4 e. = - Q e a. 4 Pt e,.d 9 A. e h =

I"

? T. 'u-a w S w- .a I ,e,, m w "j 6 4 a e-

* =. 3

= C e I. e. b.

- 3e

-I. =. 4 P e. t b c.e -w m. E 4 I =. g 8 e. O a s 9 m. U N5. " a. l .E 4-6 e.== a m 4 3 p ( b 4 l l I i l l [ e 3-37 v- - -, r-,

I 6 t i e. 1 J a large Reactor (aergency Coolant Contalement Sequence istimates benerallied C s LOCA $ cram Coolant tecircu-Heat Bemove Designator Sequence + p Injection lation Fre6 ability Degraded Cc.re Condittees ] A C C I J j A OE I i i 4 a 10'4 AJ 1.6 a 10'I C1A55 II* ' u s A 5 a 10'4 N Al 2.0 a 10'I CLA55 Ill l 4 a 10 5 a 10'8 2.0 m 10'i g CuS5 lit j l a le AC e* 4.0 m le Class IV* d 1 'Iransfer to Bridge Iree inappropr6 ate la this case

- ladependence assumed betmeen the control ved lasertlen system and the LOCA ble do=n forces t in additten only mechanical failures la the control red system af fect tAls sequence. AEl will reduce the probability of electrical failures la the RPS to e aegilglble value.

Figure 3.4.6a Limerick Large LOCA Event Tree f l m _. ___, ______i +.

t \\ \\ e g ~3a. a .N. 33 I x' A c '. = -_- i a S i .? m.z ..e = =d - 1' - .o o ..a. ,3 a 2 -&*:I, -e 3 El 3 2, 3 as 8 3 1 v3 E y. =- fr 5a a 5 & :. ?::: N O N M M W d W h Y Y 1. 'a 'a n -2 m m a w 3 U C = L. z 33= = .S. c U> e a W 55 a. 5' ' l ~7 um E

- *2 3

.5 33 3 3-=

J 3

E .a e y 2

== 9-e J E = e ~ r te. d I-l 5.'E J-a .wa <w w 3 G W .o .a. e E . 9 j 5 m. ?. e e -e 3 al m a w I 1 I l s 1.4C. j .. -. =.. 1-

O .I s. m SW O e e ese as a. .e. O m W J = = m. .o og e

A 2 2 2Cs -

e m a s..3 . 8 e ~* * = ';~ 2. 'a ~. ~~ .a 5y w e 3 3 3 = wew e S h ee d 8 h .N M* A .3

3

!= E =~ =~ :s F ,,.~ a a m tA s,me an t 7 <~s a a we. 3 ,d e Iw--g a 6 >= s. 4.s C e y 3 'o. o se y

M3 tJ C

d e as 2-a

2 3

= = g-T d U - E F 2 ~ e -++ a .v z- -2

e. um m

O M g

1. -

y &. 2 m3 @u "W. w a = v 2a -IE M .s - S. 6, g I a, s ~- 3 s w .11 l 1 l l e e e 4*

=wl

]

e l l luustat telP 3 i IAAn58141 SALAM(E IuaalNC CONot h5t R MilVs 5(CONDARY (V(Ni IA(E luaalME f(10 BIPA55 HEAI SINK ainAls Com!AIMP(Ni TRIP WAIER Ortes 5(qu(nCE Con 5(quthCE IRA 15fte T Q A W 0 E i fl ere 3.4.88 i il ulth typass 9 (3.4 TE T1 =lth Orpass ggp n 3,4,s. no Contalement ID lil5V Closure figure 3.4.Se j (.la) TW Ti utthout sypass tegure 3.4.9e i 10 2 TWE It without e pasi figure 3.4.ta s ta b .95 1.0 Th0 Milf Closure ft pre 3.4.9e O (.04) Fi =" 3 4 9* t 1A in Without sypass-10'3 1Af 11 Without Bypast itgun 3.4.9e 3 94 feo Contalsment l.0 TA0 ft51V Closure figure 3.4.98 gpy IQ II Without)ypa ss* Figure 3.4.9e .05 IQE 11 WI thout type ts" too Contalanent IltwM 3 4 I' I.0 TQO M5lv Closure figure 3.4.9* (.20) I [

All fusblne Irlps for which bypass to the Condenser is not functional, are considered to be equivalent to Mity 1

Closure teents. n0lt This event tree la evaluated Figure 3.4.7 Event Tree Diagram of Accident Sequences Following ess in, si,si e tuoi e iri, t 'e ** 'c' - a Turbine Trip Initiator. 'etwe4 Sr * '.*'The ese of the Is la progress I tree is to discriminate between events leading to isolation sad I those fer whlth the EgnGenser reselns evallab.e e e a.

.m + - s s. .+ s. 's 7 4 b I h e N \\- g \\ e s + s.. . s s et 3 s. .T. 4 1 / 4, + ac w a 5 4 4 g :eo- -a ~ ,g- -# $,.g *

4 38 4,.

c i g a = w w oL 2egd-2 2 V O 3"2 f, a 8 i -u u oe w w 8 U.c= .= 3 ga w 1aa% a e. a.,,, e -.>TI .o. .o. - e o MC n = 2-e o< m= 5 4 gh c = = M sw d 4 N aeeg 4 < c-4 4 .A t.3 3 'w e seg ~ M 61 T, 8'* e m eu =.s K =E a = s C 4 4 as '".e W / w w w 3 . I s-. .o N 6 9 5 +' 9-D 8 $me O W g M L e j ~_ o3 ,e >r

c. w y

n . no O et '4 4 a 01 o g EC 4 m em d +- e. L 3 ~ t \\ Cn o e u m 8""* ( t88 e.=* g 0 g g E T q a

r* p-
G w

M -i-p e .i , I a L&. l

C

.i-l o 3 U g 2 A C 5

E ae r

w C3 T U.* n ..e. GJ c= >W ' ~ 0 g WM = W e 6 l '\\ '. w t ,E. E 4 i> M a{ W e CQ O. w W 6 l e g-t I E E 3 M e* E' 56< = 3 g as ggg a + L g 2 3 = a ) 8 o, .e r 'N i e l i I I 3-52 i i

m m + j 5 a 's o - 4 4. a.e o - _. =. = - a =-.= . =. = > = gs<*5 = . 5 =.n u* + ,.,. e e # .. + # +.. + +. + + _3 + + + a + + . m + + + +++. .~3 +< .n a 6 w w w w w w w w w w w w - = -=wgg, e > <- .a - w w w w. a 3 g ~., o _-. m..a. c. e o e e._. _e a - e a. o .o -=- g g : 2 = ~. u 1 o 4 -a - s a _a .e a o o - a a a.. =w>,. .. < a a o-a- .c m 4333 u 6. l 3-1 y y r = ~. d a-a E < _ _ <c, -- e a a-s a s ..s. o

=

1 a.. .3 a s 3- ~ = ~ ~~ 2.o E r =

=

= ~=

. ~

- ~ ~ ~~~ ~ ~ - - - ~ X "z % i s a a, a % %" 2

Di:

44444%w% 4 *r *z en = a t t12 z% w w w w w w w w c _w_w w w w w w w w-. c. _ w w_w w w-E 2 O I -o ws ~_ u aa 2 m 2 o v = c =, o ~ s a.= 2 = e w = q o m E, T o a _ s_g=o -

== o a w o c s . a- = 2 .t 3 2jde l ~ n v u , sol o 'o- 'o. 'o 'o o .%[h t m s l~.- 2 ~ ~

<=

_~ ~ t 2 -w o a a w w 3 e _= 3 w =, -Yb3 3

o n.

~.. m e Wj o g e s wg - a. o 3-u a e- = o >d E s 0%o O zga s L m m +^ m6 c.o. 2 e W= 3 ** 8 9* e c; - <i- 'o oa L-6 >-. c a 5-2 c c. at ~ e-O -g > L-W i u.J W = ~ 5 w a o c 5*$ + w 2 = w u a e r1 e C

5. _.

n 'o L 4- + = s z.

e.

m. cn oS.o --ww w u "J3 'I

i i fC l g 1 l!aITIATam PeEYEWT!3t n!T!r.ATI3R SA!:EE DEE MI6 M Ff s. 3.4.10m Ffg. 3.4.1.3 Ffg. 3.4.14 IMITIATCM APPRCE; MATE ' GENERALtlED LC15 0F RPS . RP5 REC!RC PRCE.45!LITY CLAS5 CF 0FF317Y (CHMICA4. ELECTRICAt. PLw AAI SE'JUEMCE LD O. PC57t' LATED PCW3t TA!P (PER DE~.M CED AEACTot CCRE C0a0!TICn3 'EW C. CG) 7' c,, E g TRM5 TENT PREVCtTIC# r 1 O I O I T 3 g Cg cc 1 I 2.10-5 l (' 5.3110-2 I to-2 T 3Ct T2.uSFE2 FI EaE 3.4.100 g g I (1.s10~3) l I 1:10-5 3 Tgg TruSFEx f! Att 3.4.12 (5.3:10' ) g I I

For tre loss of offsita.se=er initiator, efectrical faults leading to a fatture ta scram my be virsally rers for loss of offstte ocner inef dents. %ver. Since no eetailed evaluation was been perfoM ta verify tnis assartion electrical aPs r ttures are tac 1voed Pece 'or a

cweleteces s. They are a sea 11 contrtNtion ts t 4 ever:11 omant1 tty of desenced enre conditions. 1 Figure 3.4.10a Event Tree Diagram of Accident Sequences Folicwing a Loss of Offsit? Power Initiator I i 1 3-63

i 8 o a - .w . o-m ~ = a v m t . gg .. - e e e e. 3,. s a m + .n

' +

.o e + - m e I +. * *. =*,..* +.* s .s .q .o..,. .a + + e s... .n +.. a. ... a + = 3 O e .,n e. e a e. aca,3- .s - w w - w w w w w w- = .s. J e. .. w w- - w w w a = w w y g w w ~ 4 ..g -., g _,.- en - a -. ."'s -_a ., g 3 ._o =-

e...3 t--

c . a .,..a a.o ~- 2 .a. o o=.a -o ~ ea a a ~.~...,,,.s .o, , o -. m.o, a.=-

e e.

a. = -. o w o = . o a.,.. .a,..,3 a ~3 a a. s. a. . =. o. oo G a_- 6 e o 4 ti i - ~~sg~ .g a ;5 ,a. a ~ s s e C z % %.u-l N. N. y s a g, a =7 er .a.a g o P i 2- _2.. 2 2. 9 E s'2 ca g, w s.m .n m .m. se-z r. r s z w w. w = m.s.3 .t e e . t.,.=s *, .,,,w w w. w w m =- m ta 4m = *.=.** Q y e C 0 'r 5 .M c ~g p g M,,,. o I .E. = V. ze -g w o-2 a.

C U

e as U o' ggw N ITY .P M" .... 5 gp g"*3 w ** - e,-- o -o o o ~

- 2 O

4 e 3 e

o.*

e N w = =. N g 0 .s. .a. ** e e ,.e .e a 4 :; - "; o. 'o_ e'o-g 9 u a t.S O g c,..T,, N .e e

4g -.

e.# 2 g e W O 0 b g .O a .e L' o m. $M m- = ~.m B~ ,g %em 3 t'3 t. QQ s l g* - e.: g ..'*.O. J a. [ y,",. .n==M E l Cb I A >w O "4

t J C3 x

y ,.n

.O.

rO. m - =, e 4 g4. 9 a C n a f w .n q eg w 2 5 n o .a. o %'8 k. .a o Q**

1..

W- .S = w A e.4 e y y -? k .W 5.. N N. a s. 3 =#- .4.s. = 2 a re m, .g ww w a JD 2

1.. A f.

* =

em e . a ,..g..

~ l J l l 1 1 s b 5 s ~ .o 3 =. =- e 1 a ~ s,1. a .a a'_._ 1 W " W s I w,s_ - a = -- - b' 1 J. 8 8 g =r b ~ ~ ~ ~ -.2t c 1 3 ;- .s. ?. 5 .a 22- _2 .e.s. a 4-o-...*W: -3 3, 4 * -.i :s s'@* .C oo :.s s o 4- - a s, o a

h8.*t

~*t n 3, 3 x =. s s cs s s' e-n.- E f* f 2 e. .3 3 = .c. ,2 't 'k g w

- *- *- *- *,_ *, *.,., =,

3 w i: s s n o

z c z.

w w w w w,, w w no W d -e a. o.:; G s U = s.- C j a2 =

2 C*

.z. - a .s

a..=

s m o-a e $.l~ - -o-o . s a.R.E = o 7 s o b. ~ a = m Q a6..38~ M ~ ~ a. E..eO e. 8 W q./1 O eO oud. .p .=.3 "." 1._ . s a ~ a

E 4 ows a

4 4 e w D C O-w r3 w s I V. g e.e .M. W C3 ~- 5 ~ 5 G b.$. 8 .a O wi 3 E -a b a _._e

gg T -.
g-

= rJ E.I I $. 3 n .I Q f. 2" 22= e* s, o Y*g -o-g6 & e.g O g uT 4 Cu we y -. 8"De' p-E W.e. 3 4n .o g 8'. 5 5. h t 4 w a 1-3 *i.s g 3 2 g s am j M g [. b = E 4 e w &_ g-. .'1 ~ 3-66 m-

.6 5 =. .g ~ ~ . :=. E =. =, = 3 I ~. =- =. = =- _. _ 2 ... =. = = g g = =,,, =, g C. .a. .n. - =- =- ~ 3 =- .-. =.. - 3 .= ~ w ~ e e o o o = e = c e o o u e. o n l o. =. o. =. o o C. ? o. a .C, = .~. C g . 3 ~ 2 2 O 38 = m aa o h, A A so o e e t A e e b e

,h g

3e.s

j

= q W M W .sD 8 s.h .g,,g e* .e

- - g em. s

-e e Q 6 3 as e x N 'O's s Ig e.e ae M-K g .6 ..3 e A .e 4 8 h.h.* e .wa .,.h W 4 e U.3 M m g g g,. .e g_ -g m 4 .M = Y 0% mm g d = s* ae S e W M S. 3 8". k 'E 88 n G .p. g W h 6 d k =. w e .O 2 6 ei. b W .i De a W pg E E k m 3 w=.& W a. .x.e .M.

.

W h"g e ( .,, e 'w n O'* x Y =-e _ 6 M T ** e= - A x e a E Q 3 ", e =. O u .=. .f g 5m.l> 3 W h .a.e g O - .S. m O W e e M . g f W g .E 48 .O. e m d. g .3 m 6 'O-W .S. O. w a E m .e. e .,..=.& M g e =- W Z m-3- = T,. =*- 48 =~ e ed 6,y 6 He M g E .A S O S = - v - = 8 M& U

d.

g=3 W%

O = - 3 acw ~3 3,. e. e W 6 8* W M .=* T =. g m e .h. em e. e 3 .e. en e==.e - =

o_

a E.* h,g e e =.* g W e. W e m,., M e O O W e.s e.d .J. 48 "8 gg E 9 M = .m. g 48 6 E

W e.d

.s. 4.8 .i. b" 48 15 g 4.$ .*5.e 3 g M i* 8

.i.
gg

,4 M = O e== m O

" e

=, =.* O, 4"".E .S I,.=3 s= W 2_ 84E*9 'M' M ib., e. . q a.mg .m M 8W W g 4.P 9.lk N j# .e E 3 II

I I s 'g l INITIATOR PREYEwiten m!T!AAT!cm 541W TEE H .CONTAMMT Nt! Fig. 3.4.11b Fig. 3.4.13 pg gy In!T!ATOR RPS RPS R10!RC WRC1ME GE N ! ZED 10ay (CHANICAL [LECTRICAL P.9 AAI $[WEMCE PRC5A8!LITT CLA55 of TRIP LEVEL P057ULATED (P(Ra(ACTC2 O(WED TEAR) CORE C0901 TIM in C, C K @ r TaAss! EXT c PREVEMTI0ft .I h i g I T C CK g g I 2210-5 I I .O?m I t T Ca 25m HM 3.4.Mb g g l (1.4:10*8) I I l 1x10-5 T'g 25}' R M E RI 3.4.11b g I (7:10 ) I I I i All motes art in Tante 3.4.fa. Figure 3.4.11a Event Tree Diagram of Accident Sequences Following an 10RV Initiator 4

I e

L 3-55 '"'w=v w't-4 y

4 t SYSTEM UNAVAILABILITY e FAULT TREE / EVENT TREE DATA INPUT ~ e METHODS OF EVALUATING THE FAULT TREE e TABLE OF COMPARIS0N WITH WASH-1400 i e DISCUSSION OF HPCI FAULT TREE SUBSEQUENT STARTES MAINTENANCE UNAVAILABILITY ELECTRIC POWER e FAULT TREE VERIFICATION: COMPARISON OF FAULT TREE RESULTS VERSUS OPERATING EXPERIENCE DATA

~ ~ --..- - - -

1 i FAULT TREE INPUT DATA THE NEXT SET OF DATA TO BE DISCUSSED IS THE FAULT TREE INPUT DATA. THESE DATA TAKE ON A WIDE VARIETY OF FORMS AND INCLUDE THOSE DISCUSSED BELOW. e TRANSIENT INITIATORS e COMPONENT FAILURE RATE e HUMAN ERROR RATE e MAINTENANCE OUTAGE '0F SAFETY SYSTEMS e-0FFSITE POWER UNAVAILABILITY e LOCA INITIATORS n

e ( s . SOURCES OF COMPONENT FAILURE RATE DATA e-PLANT-OR COMPONENT-SPECIFIC DATA (E.G., LOSS OF-OFFSITE POWER), e NRC GENEP,IC DATA (PUMPS (16), VALVES (1Z), DIESELS (18), AND~ HUMAN ERRORS (19), e IEEE-500_ (ELECTRONIC COMPONENTS, (20), (NOT USED), e VENDOR DATA, ^ e THE RACTOR SAFETY STUDY (2_1), AND e NPRDS (NOT USED). d 0 7 a e 8 i 4 ye r Esy - - + * -'+Ptr*- gN-F r-F- c +- -y w e---?--- 1 e---- - e w-* - - - - - = - - ,e

s i' 1 MODEL INPUT DATA i SOURCE OF INPUT DATA ITEM SWAIN PMJ OR PEACH BOTTOM GE NRC/EG&G WASH-1400 GUTTMAN OPERATING EXPERIENCE Components Pumps X J Turbines X Valves X + Instrumentation X X- _ Systems Diesel s X X X HPCI X Maintenance X X Offsite Povier X Human Error Probabilities X X I 9 I o e .e 's.- a s ,.,e.,..-, , - -,... ~, - -n- -r --m.

P Frequency (Per Reactor Year) EPRI Survey of 12 BWRs BWR OP. EXP. TRANSIENT cl All Years Exclude Year 1 GE Assessment 2 Po er_ MSIV Closure 1.34 .57 .35 1.08 Closure of all MSIVs (4) 0.67 0.19 0.13 1.00 Turbine Trip Without Bypass (5) 0.00 0.00 0.00 0.01 l Loss of Condenser (8) 0.67 0.38 0.22 0.067 Turbine Trip 7.62 4.23 2.95 3.98 Partial Closure of HSIVs (6,7) 0.12 0.14 0.12 0.20 Turbine Trip with Bypass 3.88' 1.98 1.21 1.33 (3,13,30,33,34,35,36,37) i Startup of Idle Recirculation 0.38 0.08 0.09 0.25 Loop Pressure Regula tor Failure '(9,10) 0.43 0.35 0.31 0.67 Inadvertent Opening of Bypass (12) 0.04 0.05 ~ 0.00 0.00 Rod Withdrawal (27,28,29) 0.14 0.14 0.06 0.10 Disturbanc'e of Feedwater 1.39 0.65 0.53 0.68 (20,21.23,24,25,26) Electric Load Rejection (1,2) 1.04 ,0.70 0.63 0.75 loss of Offsit5 Power (31,32) .16 .11 .14 .38 i Inadvertent Open Relief Valve (11) .20 .08 .03* .06 Loss 'of Feedwater (22) .27 .16 .06 .70 i TOTAL 9.43 5.04 3.51 6.2

Modifies to.07 based upon NUREG-0626.

? \\

SUMMARY

OF TRANSIENT INITIATOR FREQUENCIES - Total Transient . Source Initiator Frequency 1 WASH-1270 10 WASH-1400 6 to 8 NUREG-0460 3.5(AWS) - 5.04WR) EPRI 6.2 GE Evaluation k e 6 a 1 6 l e e v-g ww- -,e &7-y-y9y g-4 ,wg.<g.-- g g-y-p u,--, .g -.m, e.- g yawwm-

.----+no-

-.e.

-e-t e e 1, .4 e h, me me

== em e== ee g e 99 Ef J e't w ee m em b e e o e e e u O O.e O.O. O O. z .e .e .se w . =.e E e ew en N #w em 8"t 9 5 99 er 3 rg e't e EP m Pg g e e e 8 o e .O .O eO O a ee oo e= = .e=

== wt wt e se -e e a w s's A dr e e e y er eeee ee e ten Ed a m b b b b bbzE g b A b ,.3 e K2%% %R w % w eS b O m e@ me es * *

tS
tS C

di OO--e De .O 'O

==% 00 -O = eo== m a a em a me e W m ae we e ow as o e vt n no o g e e =-e .e ee ow ew

== e-e e e o D e.e en --DO O-n. ~ g IIII I II I INN IIII NN O U O O O. og e=t.O. 4 b m et #9 #9 p's e pg - M c=t s's #=t.- em est w g ~ e Q ~~ z I.A.I N b b bb b a e soww w a w a so n s www wm ( .E.C eM e. e. e. P".*e. e e er e.m e. n#.e en. O.,. s't e. e. me e e e. O. O O O O. 9 O O O 0 0 0.e 0000e= em D O. eg . em.e.e .me e .e se - se.a P4.=e PS se .e sa a = m = = m er .=e em o .ee se amma em em aus se ese e ese em. M D ~ ~ ~ ~ s e e ~ l.& 2 LA E E J e.-< e, O .O -b l -e, e e. b

- C.

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b. I b,

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- es W

er edes e8 Sr & O ep W 3 l 8.e OuOU b b b b b b be8 et g & E. E E e-e s 3 se se .em W quD & W auD= ese me3g en a 3 3 em e = 3 3 3 3 3 e-.3 om dum es aus es emeeem We g g O 4, ee S e 3 e ee 3 ee em 3 epe6 b tm. E b W bb tum bbM tem b me 5 h G. C T

==e8 en O en w ed p,. e e an e e= e b O e-A meee e v g y er .= w Cm e= e b 9 e g 9 e W ep 3 99

== 9. w en

== er 3 en b aum me ea se O es m me b O er er e et e er er C e ese e. e te 3 >es b > W b e e .== = U > O ** O C

- e en h p e= D b er QEe

=, E > * = 4C Q > W5 O as O e f 9 e te O& e O Ce

== Cb e a a-*

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.J t9 h. WD eeD at sC 44 a tas L9 > Y e

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___. /. \\ \\ COMPARISON OF PRINCIPAL COMPONENT FAILURE PROBABILITIES FROM THREE DATA SOURCES (As Used for input to the Fault Tree Model. These Probabilities of Component Failure are Assessed for the Duration of the Event Analyzed) Component WASH-1400 GE NRC -3 -3 -3 Pump 2 x 10 2.8 x 10 2 x 10 Vaives -3 -3 -3 NCFC* 1.25 x 10 0.6 x 10 3 x 10 NCF0* 1.25 x 10-4 0.6 x 10 1 x 10-3 -4 -2 -2 Diesels 3 x 10 6 x 10 ADS Relief Valve 1.25 x 10 3 x 10 5 x 10-3 -5 -2

NCFC - Normally Closed Fails Closed NCF0 - Normally Closed Fails Open
- Data Taken in 1968 during a time period in which a generic relief valve problem existed. The problem was subsequently corrected.

o 8 m4,. a ,-e.,-

q COMPONENT DATA VARIATIONS NOT INCLUDED e VARIABILITY AMONG COMPONENTS, E.G., MOTOR-OPERATED VALVES (SIZE, APPLICATION, QUALIFICATION REQUIREMENTS, ENVIRONMENT), e VARIABILITY AfiONG f1ANUFACTURERS (CONSIDERED IN SOME CASES), e VARIABILITY AMONG COMPONENT MODELS (CONSIDERED IN SOME CASES), e VARIABILITY IN METHOD OF INSTALLATION, AND e VARIABILITY IN COMPONENT AGE. 4 0 T*-** - ew em. m em - -%,4m- ..w

. - - -... y \\ MAINTENANCE UNAVAILABILITY e USE GE ASSESSMENT BASED UPON BWR GENERAL OPERATING EXPERIENCE AND PEACH BOTTOM SITE-SPECIFIC DATA e COMBINE IN PROPER LOGIC NOTE THEsE IS A SIGNIFICANT DIFFERENCE IN SYSTEM LEVEL MAINTENANCE UNAVAILABILITY BETWEEN CURRENT OPERATING EXPERIENCE AND THAT ASSUMED IN WASH-1400, HOWEVER, BECAUSE THE LCOs REQUIRE A MINIMUM COMPLEMENT OF SYSTEMS AVAILABlE AT ALL TIMES THE NET IMPACT OF THE CHANGE ON CALCULATED CORE MELT FREQUENCY IS SMALL. e 6 4 9 e u

\\ \\\\ 5 s, \\ Primary System Pipe on-ar g v. s s u.. _____l_ tusn UW ////////// ,d u r,.. All Pipe

g*

,(((((([ All Sizes EU" ' ///// / Na $ N Win? M $l%M j .I 2.-' ~,.......... u.,........,,i......., Comparison of Evaluated Rupture Probabilities for Pipe to Estimate fluclear Power Plant Pipe Rupture Probabilities LOCA SEf1SITIVE PIPIf4G OtiLY A Estimate from WASH-1400 based upon comercial nucledr operating exp. O Current estimate based WASH-1400 Assessed Value upon comercial nuclear (Table III 9) operating exp. (BWRs) e Median I f -l A o I ~ ////////// e i i i -6 -2 -1 10-7 10 10-5 10-4 10-3 10 10 Rupture Probability (Ruptures / Plant Year) Estimates of LOCA Initiated by A large Pipe Break

DESIGN AGENCY FMEA 0F SYSTEM I I l -s l I / N 1 i t GROUNDRULES i i i \\ 1 I N I I DIRECT i ~ EVALUATION - I Y l I GIVEN: CONSTRUCT i I I I SYSTEM LOGIC DIAGRAM 'i l i I I I i N FINAL i NUMERICAL i y MINIMAL z EVALUAR10N OF CUT SETS CUT SETS s / 1 l i g GROUNDRULES I s I I I 1 s l l I ' j I j 1 e SYS. DESCRIPT. I I I e FSAR l l l e O

P. PROCEDURE

S i i e MAINT. PR0CED QUALITATIVE QUANTITATIVE 9 i t Fault Tree Evaluation: Quantitative or Qualitative 4 g

COMPARISON OF SYSTEM LEVEL FAILURE PROBABIlli is (PO!NT ESTIMATE) / g LIMERICK PEACH BOTTU1T (MEAN) (MF.DI AN) SCRAM SYSTEM 3x10-5 1x10-5 HIGH PRESSURE INJECT 10N 8x10-2 9.8x10-2 HPCI RCIC 3x10-2 . 8x10-2 FW 2x10 0.3* 2x10-2 .2* DEPRESSURIZATION 5x10-3 m Low PRESSURE INJECTION LPCI 1.7x10-3) -- 1.5x10-2 3 CS 5.2x10-3 q'9x10-5 ?2x10-4 CONDENSATE 1.0 1.0 4x10-(AUTOMATED) STANDBY LloVID CONTROL 0.1 EMERGENCY SER/ ICE WATER 2x10-4 1x10-4 ~ SERVICE WATER 2.3x10-4 4x10-4 CONTAINMENT HEAT REMOVAL RHR 4.0x10-4 2.3x10-4 PCS 5x10-3* 7x10-3" RCIC Ifl THE STEAM COND. MODE .1 NA SEQUENCE DEPENDENT ESTIMATED BASED UPON OPERATING EXPERIENCE INFORMATION t TAKEN FR0ti GE OPERATING EXPERIENCE INFORMATION

i i t i ( f ~ BENCHMARK COMPARISON 0F BWR/4 FAULT TREE MODEL AGAINST AVAILABLE DATA (RCIC AtJD HPCI Of4LY) O e O S T-

O-. ~ -. O._. t e I } e 9 Ii SU."J1ARY OF HPCI/RCIC FIELD DATA REPORTED 'IC GEllERAL ELECTRIC AS OF 10/24/79 HPCI RCIC-PZANT ATIC4 PTS TO START FAILURES TO START ATTC1 PTS TO START FAILURES TO START 1 A 99 1 96 0 B 96 1 88 3 C 100 11 100 4 D 149 26 125 17 D 160 23 270 31 E 105 4 92 7 E 106 4 100 10 F 151 1 ~184 2 TOTAL 966 71 1055 74 D 8 9 9 o' e y e e e 9 L

i 6 SUMARY OF FAULT TREE CALCULATED FAILURE PROBABILITIES FOR START AND RUN COMPARED WITH AVAILABLE DATA FAULT TREE SYSTEM MODEL CALCULATION SURVEY HPCI 4,6x10-2 7.3x10 ' RCIC 4,Sx10-2 7,0x10-2 8 o ,-e

ESTIMATE THE COMBINED HIGH PRESSURE l SVSTEM PERFORMANCE BASED UPON THE FAILURE PROBABILITY HIGH PRESSURE SYSTEM UNRELIABILiiY 2.3x10 3/ DEMAND O (HPCI&RClb+CRD) ~ 9 NUMBER OF DEMANDS ON HIGH PRESSURE SYSTEMS 27/YR +, LOSS OF RY = 16/YR + LOSS OF OFFSITE PWR = 04/YR + LOSS OF aux, PWR = 79/YR + MSIV CLOSuaE = 67/YR + LOSS OF CONDENSER = 1.93 D/ YEAR = 189 YEARS O NUMBER OF BWR YEARS OF EXPERIENCE ~ PROBABILITY OF A SINGLE OCCURRENCE .84 \\ OF RCIC AND HPCI BEING UNAVAILABLE ~ / SIMULTANEOUSLY UPON DEMAND 9 I e .m m---m w

c-- HUMAN ERROR PROBABILITY t OVERVIEW: 1. PLANT SPECIFIC OPERATING AND MAINTENANCE PROCEDURES DiD NOT EXIST FOR LIMERICK DURING PRA PREPARATION 2. GE/BWR EMERGENCY GUIDELINES WERE USED FOR MANUAL OPERATOR RESPONSE DURING ACCIDENT CONDITIONS 3. NEED FOR OPERATOR RESPONSE FOR AUTOMATIC SYSTEM IyITIATIONISTAKENTOBECONTINGENTUPONAUTOMATIC SYSTEM FAILURES (CONSERVATIVE ASSUMPTION). 4.

SUMMARY

OF HEPs USED IN THE LGS PRA: REQUIRED ACTION HEP REF OPENING REMOTE MANUAL VALVES 0.9 EST. AUTO. SAFETY SYSTEM BACKUP 0.1 WASH-1400 INITIATION (30 MIN HPCI, RCIC, LPCI, CS) RHR INITIATION (15 MIN.) 0.01 SWAIN (ATWS SEQUENCES) DEPRESSURIZATION (30 MIN.) 0.002 SWAIN VALVE ALIGNMENT DURINC MAINT. 0.0001 SWAIN RHR INITIATION (20 HRS..b 6x10-5 NUS

8 i l 1.00 l .90 - 1.0-Ist MIN g .9 AT 5 MIN .1 AT 30 MIN .01 AT 2 HR o I d <5 E ce 8 l n. E oW N_.25 a w AUTCN.ATIC SYSTIN.5 WORXING OK. .10 I 120 . 01 1 5 10 30 60 TIMI (MiNtTTES ATER LARM t.0CA) l l 1 I Estimated Human Performance After a large LOCA f C-2 %v- -.u

"iw

7 g I i ESTIMATED PROBABILITIES OF FAILURE TO INITIATE AFWS* i At End Of X Minute Situation without Dedicated Operator Regular Shift Supervisor Total Failure' Opera tor HEPs HEPs (Conditional) 5 .05 .05 15 .01 1.0 (HD)** .01 30 .005 4 (MD)*** .002 60 No change No change No change

Lower and upper uncertainty bounds of a factor of 10 are assigned to each estimate in the " total" columns (C-1).

All HEPs are rounded.

- HiSh dependence assumed; Swain value of.5 of modified upwards to 1.0.
  - Hign dependence assumed; Swain value of.15 of modified upwards to.002.

a .m-

i j. I t. I i I 1 i END OF FORMAL TRA!NIN5 ASSUMING PRACTKX OF r* EMERGENCIES fN HIGH rN r% 1

%J N'

8 LWITH PRACTICE hs I I '3 OF SIMULATED EMERGENCIES EFFECTIVE COPING WITH EMERGENCIES NO EURTHER PRACTICE 'ow o a s a o o 1 TIME VERTICAL ARROWS REPRESENT PRACTICE SESSIONS Qualitative Effects of Practice and No Practice on Maintenance of Emergency Skills (taken from Swain & Guttman) \\ f J C-6 ~ .sy-- q g

s UNCERTAINTIES: THE MAJOR OBJECTIVE OF THIS RISK ASSESSMENT IS TO DEVELOP A BEST ESTIMATE COMPLEMENTARY CUMULATIVE DISTRIBUTION FUNCTION (CCDF) FOR EARLY AND LATENT FATALITIES FOR THE LIMERICK PLANT. IN ADDITION, THE FOLLOWING CHARACTERIZATION OF UNCERT-AINTIES IS PERFORMED: 1. UNCERTAINTIES FOR SELECTED DOMINANT SEQUENCE PROB-ABILITIES ARE GENERATED, USING A MONTE CARLO SIMULA-TION OF THE SYSTEM MODELS, AND THE INDIVIDUAL COM-PONENT UNCERTAINTY DISTRIBUTION. 2. SUBJECTIVE CHARACTERIZATION OF CCDF UNCERTAINTY, INCLUDING THE UNCERTAINTIES IN: e SEQUENCE EVALUATIONS 9 IN-CORE RADIOACTIVE RELEASE PROCESSES e EX-PLANT CONSEQUENCE CALCULATIONS. T 9 't ~ ~' ~

te i n UNCERTAINTY ANALYSIS e WASH-1400-s. e EPRI (NP-1130) e LIMERICK (ESTIMATE) e

SUMMARY

OF KEY AREAS OF UNCERTAINTY e SEQUENCE PROBABILITY MONTE CARLO SIMULATION 9 e n. . y .. * >. ~ -.. ~ ' ..... ~ ~. ~ - a

~ 10-3 -_ i i e i iiii i 1 6 6 664 6 6

i 4666 6

4 6 68li' d io _ 10-5 x A I 3 U d 10-6 .h km ~ lPWR at N ~ 10., Average curve ~ BWR. 10-8 2 ,,,,,tet i t i i i t ti t t I !'1 ' ' ' 'flif 10-, 0 4 1@ 3 10 2 to 0 101 10 Earty Fatalities. X FIGURE 5-3 Probability Distribution for Early Tatalities per Reactor Year Note: Approximate uncertainties are estimated to be represented by factors of 1/4 and 4 on consequence magnitudes and by factors of 1/5 and 5 on probabilities. - I k y

T t' s k ,A 4 4 4 i : s ii 6 6 i i n ii i I

4. i l i si i i i i l li t i i # i lli 10-3 _

i ~ I 4 10 _ ~ ~ 10-5 ~ x ~ A N g MR

1. 8" ""*

BWR x 4 c 10 ? L E 3 = \\ 73 2 og 10~7 2 \\' 10-8 ~ g. t I t ttttt t t ttttt t t t t i!f f f

t t!ttt t
- t!

0 10 10 10 10 105 1 2 3 4 10 Euty luness. X FIGURE 5-4 Probability Distribution for Early Illness per Reactor Year Note: Approximate uncertainties are estimated to be represented by factors of 1/4 and 4 on consequence magnitudes and by factors of 1/5 and 5 on probabilities. \\' t e l k

Table 3.8.1. g SHIFTS IN ACCIDENT FREQUENCY AND CONSEQUENCES FROM EPRI NP-ll30 Accident Accident Frequency Consecuence WASH-1400 uncertainty factor 5 4 3 Multiplicative shift in median 1/12 1/5 Increase in multiplicative 13 10 uncertainty factor Total multiplicative uncertainty 20 15 including WASH-1400 s.-' _ s - NNkh

i. a m

N ,,. s .......- " ;15 %, "x w p 4 r a.,,ere. y a. a. g 4 / \\ 1M. agy/ 1 'j to e r er, e e t

i,

,re i i. i. i.4 tg i.- to Carty Fataltties (2)

noorontmate vecertatattes are estimated ta te represe$ted by f acters of 1/4 and 4 on coeseuvence =gnatudes and ty f actors of 1/1 and i on arveantitttes.

Figure 3.8.1 Probability Distribution for Early Fatalities Per Year for 100 Reactors (From EPRI NP-ll30) 3-142 ~ a... - t

10 1 l TOTAL MAN-C AUSED RISK l' 2 N g TOTAL NATURAL R S 10 \\ 10'# ^' 10 5 7 0xbc ~~~~ g 10 %*% s O 5 10-7 N w 3 N. m g E ~ N'N h 3 10 / N N LIMERICK N 4EST ESTIMATE \\ s 10'I -sN.,' \\ -10 N 1 10 100 1000 10,000 FATALITIES (x) ....... = Uncertainty Bands Figure 3.8.3 Sumary of Risks Assumed by the Population Surrounding the LGS for Early Fatalities with the Estimated Uncertainty Band. 3-147 4

^ I \\' UNCERTAINTIES e THE CHARCTERIZATIONS OF COMPONENT LEVEL INPUT DATA UNCERTAINTIES (INCLUDING HUMAN ERROR) ARE NOT WELL-DEFINED. THE PROBABILITY DISTRIBUTION FUNCTION IS NOT IN GENERAL KNOWN, ALTHOUGH' LOG NORMAL SHAPES HAVE TYPICALLY BEEN ASSIGNED IN PAST PRAS WHERE UNCERTAINTY ESTIMATES HAVE BEEN . ATTEMPTED. s e THE UNCERTAINTIES IN THE CONSEQUENCE CALCULATIONS ARE EVEN LESS WELL UNDERSTOOD.- THERE DOES NOT APPEAR TO BE A COMPLETE EVALUATION OF THE EXTENT OF POTENTIAL UNCERTAINTIES IN THIS AREA. O PRESENTLY, NO FORMAL MECHANISM FOR PROPAGATING UNCERTAINTIES THROUGH THE ENTIRE PRA (I.E., PROBABILITIES AND CONSEQUENCES) IS IN USE. THERE-FORE, CHARACTERIZATION OF UNCERTAINTIES ON THE CALCULATED CCDFS HAS THUS FAR BEEN SUBJECTIVE. e 9 9

- ~ - Table 3'-2 PARTIAL PRA CHECKLIST FOR COMPLETENESS ANDAfPLICABILITY 8 SUBJECT DISCUSSION METHODOLOGY Elimination of The Reactor Safety Study used a discrete set of Smoothing of release categories and then assi,gned probabil-Release Categories ities to adjacent categories based upon uncer- - tainty in sequence categorization. The use of a larger number of accident release categories provides greater specificity in consequence definition. Based on more precise definition in accident sequence releases, the need for smoothing may be unnecessary. No Repair of In the Re' actor Safety Study very little or no Failed Systems credit is given to the operator for restoring a system to service if it is failed or in mainte-nance. Time Phasing Systems required for operation are considered unavailable if they fail probabilistically. However, in some cases these syst' ems can be restored within a rather short time. Repair of systems is generally not included; however, for specific systems for which explicit operator procedures exist and adverse environment is not considered a problem, repair may be included in the modeling and quantification. Duration of The length of time required for system operation System Operation is set by the definition of successful end state Required used in the analysis, i.e., stable sonditions, hot shutdown, or cold shutdown. The time required to reach these specified conditions directly affects the evaluated conditional input probabilities to the logic model. 5 3 3-17 i S

~ u ... a _ Table 3-2 (Cont 'd) SUBJECT DISCUSSION METH0001.0GY (cont'd) Completeness Finite Number of Sequences Examined: There are an infinite number of possible accident sequences. However, due to hunan and code limitations, only a finite number of these can be examined. Since this set.of sequences represents an approximation to the spectrum of conceivable' accident sequences, sequences are often grouped to conservatively incorporate the events and consequences which may develop. This-and other systematic procedures help to make sure that all significant risk contributors are included in the analysis. l Common Cause Methods to Ensure Common Cause Events are Included: Because common cause failures entail a wide. spectrum of possibilities and enter into all areas of modeling and analysis, common cause failures cannot be isolated as a separate study, but instead must be considered throughout_ all the' modeling and quantification steps. Possible - common cause events are handled both a't the i event tree level, allowing for the failure of one system to cause the failure of another, and at the fault tree level, accounting for similar components to fail in a dependent-manner. Common cause methods used in PRA methodology include: - Functional dependencies anong-systems in event trees, - System fault tree construction extended to include the common harokare dependencies among systems, and - Common human interfaces, test and maintenance among similar components l-G ..... ~ .y v.,m-w r T w e-e -P-

,4 .. _..T~~~ " ~ ~ ~ --A t e Table 3-2 (Cont 'd) ,I t SUBJECT DESCRIPTION ~u 'METH000 LOGY (cont of Binary Nature of Fault trees, as used in PRAs, are inherently. Fault Trees

binary. The trees model only the success or

failure of components or events; no partial successes are usually considered. Because of this, assumptions must be made as to what the y

" success" of a-component refers to. Often,-it is conservatively assumed that any partial failure is a complete failure. Hardware Under the limitations of time.and money, assump-tions are sometimes made regarding the oper-i ability of certain systems. Specifically, non-p safety-rela'ted systems.may be given little credit for their accident mitigating potential. These types of assumptions must be clearly noted. f . DATA Data Available The quantification of event tree / fault tree ' models should be based upon the best available } data base applicable to each specific component i: or system. However, the availability of data appropriate for such quantification is limited on both a generic basis and.a plant-specific basis. I Generally, it fs useful to combine several data i sources; however, variabilities that are usually not in the input data may include: i;. - component size, application, or environment, j -, - manufacturers, and o - component ages. 5 e i Newly constructed plants will not have plante specific' data; therefore, the'use of some generic F data source may be required. m g 6 $h

6 'M

es.

w 6 er ev. 4 e + e e asem e. - - - -,, -- - l,...

Table 3-2 (Cont 'd) .\\ SUBJECT DESCRIPTION' DATA (cont 'd) Plant / Component Data for plants with a long operating history are Age not available. Therefore component failure rate data is in general an average of failure rates over the initial 5 to 10 years of plant oper-ation. This average is anticipated to be repre-sentative even during the wearout phase of plant life. ATWS Frequency Because operating experience is insufficient to adequately characterize the potential for ATWS, there has been a great deal of speculation; Currently the frequency used is that derived from Reference (3). The frequency could be approxi-mately 10 times higher or 10 times lower de-pending upon one's assumptions concerning pre-cursors and rectification. Diesel Failure A single diesel failure rate to start and run for Rates relatively short periods of time are well docu-mented. Failure of two or three diesels is subject to larger uncertainty but can be esti-mated using available data. Failure of four diesels is highly uncertain. PCS Availability A key system used in normal plant shutdown is the Power Conversion System. However, very little data exists to characterize the PCS reliability under a wide variety of the conditions under which it may be required. Constant failure The failure rate is generally assumed to be a con-rate assunption, stant. The time variation of component failure e.g. pipe f ail-rates is not known. Recent EPRI work has shown ures, instrument that higher than normal failure rates may be failures expected during the initial year of plant oper-ation. There is curre..t'.j no charactar !&ation of the end of life performance of major plant compo-nents, i.e. pipes, pumps, and main stream isolation valves. a

Table 3-2 (Cont 'd) SUBJECT DESCRIPTION - DATA (cont 'd) Use of log-normal -Log-normal distributions are generally assumed to describe the to describe component failure probability frequency distribu-distributions. However, sufficient data does j tion of failure not exist to justify this assumption. I rates for certain components Human error Data cited in the Reactor Safety Study (1) and probabilities the Human Reliability Handbook (4) are generally used; however, very little actuaT data exists to support these evaluations. RadionuclIde Decontamination factors have a significant Decontaminations impact on the calculated economic consequences of Factors (DF) radionuclide releases to the environment. There is a significant amount of data on obtainable decontamination factors from small scale tests; however, very little well-documented data exists on large scale decontamination efforts. Th ere-fore, the jus.tific~ation of DFs used in the analysis is necessary. Meteorological One of the key input parameters in calculating Data the distribution of radionuclide in the environ-ment is the meteorological data. Consequence. assessment codes such as CRAC or CRACIT (see Section 7) make use of sampling schemes for weather patterns from a single year of data. The meteorological data input and sampling scheme coupled together can affect the calculated risk. Therefore, it is useful to have a discussion of . the effects anticipated if different years of-meteorological data are used. CONTAINMENT Core Malt Assumptions regarding the timing of events and (General) the physical proceses, e.g.* particle fragmen-tation, core slumping, reactor pressure vessel mel',through, cer: Ote interaction involved in the postulated core melt are,modeled in a simplistic fashion. The implication of these simplifications should be identified. 3-21 .. -....... ~ . a.

Table 3-2 (Cont 'd) 9 6 SUBJECT DESCRIPTION CONTAINMENT (cont 'd) Reactor Pressut: The manner in which the RPV fails during a Vessel (RPV) postulated core melt is' highly uncertain. One Failure method assumes RPV failure from creep rupture - that the RPV ruptures from the stress of the molten core rather than melting through. This model allows the entire bottom head of the vessel to fail at one instant. Other failure modes assume failure from melting, but the manner of melting is also uncertain. If convect' ion currents are assumed, then hot spots would tend to occur on the edges of the vessel and the entire bottom head would fail. If there are not any convection currents, heat transfer would mainly be from conduction and hot spots would occur at the bottom and the entire bottom head would not fail. These last two methods both consider pressure relief during RPV melt by assuming melt through the bottom head pentrations. Molten Core An area of large uncertainty is the manner in Reaction which the molten core will act after it fails the RPV. It is uncertain whether the molten core will: ,e drop onto the floor below in one coherent mass, e fragment and disperse around con-tainment from blowdown of RPV if a large blowdown force occurs, e stay localized below the vessel, or e react with water present causing steam explosion (s). ~ Molten Core In some of the dominant sequences, the oxide layer may be predicted to freeze. The impli-cation of this frozen layer is not well known. It is thought that 'once the. oxide layer freezes the vaporization release stops. Radionuclides i are released during vaporization; the gases (non-l- condensibles) generated at the core / concrete l ~ 3-22 t

Table 3-2 (Co at 'd) SUBJECT DESCRIPTION CONTAINMENT (cont 'd) Molten Core interface bubble up just through the metal layer j (cont 'd) and then through the oxide layer (the oxide layer being on top). The fission products are con-tained in the oxide' layer and are carried away by the gases crossing through. If the oxide layer is frozen, the gases cannot bubble through and escape to containment at the edges of the core and the fission p,roducts do not get transported from the melt. Hydrogen Gener-The amount and timing of hydrogen generation ation during the core melt process may have a signi-ficant effect on the quantity of radionuclides released. Therefore, the models, assumptions, and results should be discussed to indicate the impact on accident sequence release fractions. Steam Explosion The probability of a steam explosion (in-vessel or in-containment) is the subject of wide controversy. The probabilities use'd in the Reactor Safety Study are expected to be an upper bound. More recent PRA 's tend to use lower conditional probabilities for steam explosion based upon Sandia experiments. Hydrogen Explosion It is considered possible that a hydrogen explosion of sufficient magnitude to result' in radionuclide releases comparable to in-vessel steam explosion may occur if the explosive mixture of gases is brought together with an ignition source. RELEASE FRACTION RtALI/LUKKnL IvotL: Radionuclide The CORRAL code uses the Reactor Safety Study Release values for best estimate percent releases for each group of radionuclide. These values are uncertain and recent experimental data indicate the larger numbers are conservative and the lower estimates are too low. Reference (5) offers 3-23 N M sw

Table 3-2 (Cont 'd) SUBJECT DESCRIPTION RELEASE FRACTION xtAtt/LuxxAL ICUEL: (cont'a) Radionuclide some additional insight into this area where data Release is sparse and attempts at modeling are under dev el opment. Containment - Containment integrity may be assumed to be main-Integrity at tained with the temperatures > 250 F and at ~ High Temperatures internal pressures above the design limits over and Pressure extended periods of time. These conditions are ~ beyond the design limits of containment, there-fore current best estimates may be optimistic. EX-PLANT EFFECTS Radionuclide The model used to define the plume of radio-transport active material as it traverses large distances and dispersion (>20 miles).has not been verified experimentally. ~ Evacuation model The calculation of early fatalities is very sensitive to the assumption on evacuation of the population. Many aspects of~ evacuation are untested. Shielding There exists a significant variation in the type effectiveness of structure available in the environs of a power plant for sheltering of the population and the shielding offered by these structures. ~ Dose-Mortality The applicability of given dose mortality response response curve curve is strongly dependent upon the health of a person and the degree of medical attention he receives once exposed. The calculation of early fatalities is very sensitive to the assumption made in selecting a response curve. I 9 e 9 6 3-24 'a-v' --,m

c ~ Table 3-2 ' (Cont 'd) n SUBJECT i DESCRIPTION EX-PLANT EFFECTS (cont'd) Threshold effect There has been 'a long-standing controversy on the on latent cancers existence of-a dose threshold; that is, a thres-- hold below which latent cancer risk to a person f is zero. Duration of radio-The release of radionuclides calculated by nuclide relcase CORRAL to escape with each containment failure mode and accident sequence is assumed to occur r over a 30 minute period. This permits modeling the. release as a puff. If the release is of longer duration it' is possible,.because of wind . direction cha'qes, that concentrations of radionuclides v ill be lower than if a " puff" release is assamed. 4' E y l I i f }. L y e 3-25 ,v- - -~

c b 4 'i UNCERTAINTY IN SEQUENCE FREQUENCY .1 - r 9 IDENTIFY THE DOMINANT SEQUENCE IN.EACH RADIONUCLIDE l RELEASE CLASS 8 DETERMINE THE UNCERTAINTIES t e-INITIATOR: OPERATING EXPERIENCE DATA COMPONENT INPUTS: WASH-1400 i e MONTE CARLO SIMULATION OF DOMINANT SEQUENCES 9-CALCULATE THE UNCERTAINTY IN THE FREQUENCY OF THE i SEQUENCE. 1-1 e f i f l i I l }

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t t' t t t t te j* so' se2 W ses -,a % r .x PROBABILITY DISTRIBUTI0fl FOR EARLY FATALITIES PER REACTOR YEAR , NOTE: APPROXIMATE UNCERTA NTIES ARE ESTIMATED Td BE REPRESENTED SY i FACTORS OF ld AND ON CONSEQUENCE MAGNITUDES AND BY. FACTORS OF.L/> AND ON PROBASILITIES. ~~ r---

1 -- ~ -6 TRAwiltNT W/ Ctattatll[0 1055 or vt NT vtWT 151183fE0 CL A15 OT t 0(CAT HCAT OP(N (OgfROL 3*t[UP LIOutNCE 5f0UChCC PC57ULAft0 ALM 0 VAL INIIIALLY PAINIAIN(3 WAf(A OC51C::ATCA pac 6 4 1LITY CORC MILi 8 i g k fu MOOC 1 M00C 2 r100C 3 TV Og PCOC 3 ' 2.2 s 10"O Class !!! MODE 2 1.1 a 10*I Class II t00( 2.3 2.2 s 10'8 Class !!!

CCC 1 1.1 a 10*I Class !!

Figure 3.4.11. " Bridge" Event Tree Providing the Link Between Identified Accident Sequences which Result in Containment Overpressure and the Containment Event Sequences Following Core Melt. t ACCICINT salact CONTAIM(NT d INit! ATOR$ + 5(Qu(NCE + TA([ + (VENT (vtNT TACES (IFREQU18t0) TAtt figures 2.4.1 Figure 3.4.11 Figure 3.4.14 t.hrougn 3.4.10 Figure 3.4.15 Figure 3.4.12. Flow Chart of the Event Trees Used to Define Accident Sequences. 4 i I l l 6 n

LIMERICK SITE COMPARISON ~ l (Relative to WASH 1400) l e Realistic Site l; i l e Higher Population i l' e Different Weather l e Same Evacuation Model i .s.,.. lt

LIMERICK DESIGN COMPARISON i (Relative to WASH 1400 BWRL e MK ll Reinforced Concrete Steel-lined Containment e Larger Standby Gas Treatment Sy' stem e Containment Overpressure Relief ~ e More and improved Safety / Relief Valves e improved Piping ~ e ' improved Shutdown System i ~ 'l

Spray Pond for Emergency Cooling Water e improved Emergency Pump Capability

~ 'i e Four Dedicated Emergency Diesel Generators e More Reliable Offsite Power' ~ ~ ..o... l.

Limsrick Preliminary Risk Assessment Design Features Comparison i-1 I N I I 10

AL MAN-CAUSED RISK 100 1

TOTAL NATURAL RISK 1000 N 1 ~ 10,000 FREQUENCY 1 (events / year) 100,000 1 t 1 MILLION WASH 1400 BWR WITH UPDATED METHODS AND DATA AT LIMERICK SITE' I i 'DOeS NOT INCLUDE i 10 MILLION h LIMERICK DESIGN i N renguReS 100 MILLION LIMERICK AS DESIGNEDN \\ 1 { 1 BILLION 1 10 100 1000 10,000 FATALITIES ) onn. I

LIMERICK DATA COMPARISON (Relative to WASH 1400) c o Larger Data Base e initiating Frequencies .l,

Equipment Reliability o Maintenance Times j

I l i -l

I LIMERICK METHODOLOGY COMPARISON ~; i (Relative to WASH 1400) i e improved Computer Models e More Comprehensive Treatment of Transients e Updated Decontamination Factors e Updated Treatment of Hydrogen / Steam Physics

Ii e Use of Emergency Operator Guidelines l

l i I

t l l ~

SUMMARY

OF DIFFERENCES i (Relative to WASH 1400)

o i'

j o Site Effect - Up ) o Methods and Data Effect - Mixed i i ~ e Limerick Design Effect - Down I

ll e Net Effect - Down eser.so i

i i i i

RESULTS e Limerick Site Higher Population e Limerick Plant Better Than WASH 1400 Plant l e Limerick Plant-Site Together Lower Risk Than WASH 1400 BWR ,t a e Limerick Generating Station Presents Minimal Risk to Public 9547 9 1 !

'l Uncertainty Bands on Limerick Best Estimate CCDF -1 N I N TOTAL MAN-CAUSED RISK A/ TOTAL NATURAL 'i 5 -4 N r. f10'I 24 10-6 -m f '*ss O \\, $10'I \\ 5 's 10-8 \\ f \\ s LIMERICK s 4EST ESTIMATE N 10'I \\ ~,'g m, -10

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ENCLOSURE 4 COMMENTS OF MR. ALAN NOGEE The following are comments received by NRC from Mr. Alan Nogee during a meeting recess period on February 11, 1982. Mr. Nogee noted that 32 contentions related to the PRA were recently posed in the Limerick proceeding and that many of these contentions pertained to the basis of comparison between the PRA and the WASH-1400 study. He was of the opinion that selective use was made of updated methodology and data in preparing the PRA, leading to biased results and a questionable basis for comparison. Mr. Nogee also took issue with the use of 1970 population figures for both Limerick and WASH-1400. He stated that this represented data compiled 5 years prior to completion of Peach Bottom (WASH-1400) but 15 years prior to the projected completion of Limerick. Mr. Nogee also questioned why a 25 mile radius evacuation zone was assumed in the PRA and WASH-1400 study while the Limerick Emergency Plan assumes a 10 mile radius. Mr. Nogee also took issue with the fact that GE participated in the prepa-ration of the PRA. He felt that this amounted to GE evaluating the risks of its own design. He noted that the Indian Point and Zion PRAs were prepared by independent groups. Also, according to Mr. Nogee, the minimum x time required to perform a PRA should be 18 months, and he noted that PECO completed its PRA in only 10 months. Mr. Nogee also stated that Keystone Alliance and Limerick Ecology Action were interested in assuring that the NRC staff performed a thorough, independent review of the PRA and offered to work closely with the staff in its review. i bESIGNATEDORIGINAh* bGI'tified By _ W _- i i}}

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ML20052A093

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SUMMARY

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