Text

FOIA Request for Listed Pages from Section 19 of General Electric ABWR SSAR
	ML20202H898
Person / Time
Site:	05200001
Issue date:	10/27/1997
From:	Williams O; JRA ASSOCIATES
To:	Racquel Powell; NRC OFFICE OF ADMINISTRATION (ADM)
Shared Package
ML20202H886	List: IA-97-428, Final Response to FOIA Request for Documents.Records in App a Being Made Available in PDR & Encl; ML20202H882; ML20202H898;
References
	FOIA-97-428 OGW-97-175, NUDOCS 9712100385
	Download: ML20202H898 (107)
	v • d • e

{{#Wiki_filter:- FOA5% fEW ._1&Ib auu e; r ww JR B-91 = w 0K* IMP *!h ^ J/R/A ASSOCIATES ' ' ' ~ ~ ~ ' Regulatory information & Support Systenis P.O. Bar 4604 CapitolHeights, MD 20791-4604 (301) 249-9672 October 27,1997 OGW-97-175 Mr. Russell A. Powell, Chief Freedom ofInformation/ Local Public Document Room Branch M/S T6D8 U.S. NUCLEAR REGULATORY COMMISSION Washington, DC 20555

SUBJECT:

FREEDOM OFINFORMATION ACTREQUEST

Dear Mr. Powell:

Pursuant to the Freedom ofInformation Act,5 USC Paragraph 552, and to the Nuclear Regulatory Commission 's regulations, Ihereby request a copy of thefollowingpagesfrom Section 19 of the General Electric ABWR SSAR:

I9D.319 through I9D.3-21
19D.5-48 through 19D.5-80
19D.6-228 through 19D.6-247
I9D. 7-40
191-23 through 191-43
19P-5 through I9P-28 Portions of Chapter 19 were released under accession number 9305200149, which indicated that "the remaining sections will beprovid:d in a nearfuture amendment. At that time, the entire Chapter 19 will be provided in the single column format (similar to SBWR). " Further mention of the complete issuance wasfoundin 9308040020 but those sections have not been locatedin the Public Document Room.

I would appreciateyourprompt response within ten working days of the receipt of this request, as provided by the Code and NRC's policies. I agree to pay suchfees as required under 10CFR Paragraph 9.33 et se_g. withoutfurther authorization, however, please call of thefees exceed $50.00. 9712100385 971203 PDR FOIA WI,L,L, coy.n,,428 PDR IAM97

e i 'Mr. Powell Page 2 Mease callifyou have questions. Thankyouforyour services.

SincerJy, t4 11l1dott.)

e la G. Llams

{ 2sM100 Mov.1 ACWR standardsorelyAner els norers r f lL RCW/HECW. B SURGE TANK di 7 DG ZONE (B) i LOOLING COILS M l (REACTOR BUILDING) l A MAIN CONTROL ROOM g COOLING ColLS (CONTROL BUILDING) A ESSENTIAL ~~ ELECTRICAL EQUIPMENT ROOM (B) COOLING COILS (CONTROL BUILDING) { 4-FE HECW y REFRIGERATOR O (CONTROL BUILDING) II Jk HECW PUMP (CONTROL BUILDING) RCW & RCW FE HECW REFRIGERATOR i i g (CONTROL BUILDING) II Jh HECW PUMP (CONTROL BUILDING) RCW + -RCW ( Figure 19D.3 2 HECW Division B input Dets - Amendment 31 190.3 19

I 23A6100nov.1 ACWR standardserotyAnarysis norors ) RCW/HECW C SURGE TANK IL A D/G ZONE (C) COOLING COILS M (REACTOR BUILDING) A ~ MAIN CONTROL ROOM COOLING COfLS (CONTROL BUILDING) A ESSENTIAL ELECTRICAL EQUIPMENT ROOM (C) COOLING ColLS (CONTROL BUILDING) ) FE HECW REFRIGERATOR p ,g (CONTROL BUILDING) II 4L HECW PUMP (CONTROL BUILDING) RCW + RCW FE HECW q' REFRIGERATOR i i g (CONTROL BUILDING) II JL HECW PUMP (CONTROL BUILDING) RCW & RCW i Figure 19D.3-3 HECW Division C 19D.3 20 input Dats - Amendment 31

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nAs100 nov. s ABWR sondedseraryAnerrsis neron C Table 19D.5 8 Impact of Suppresalon Pool Bypass on Conditional Probabilities for Recovery of Containment Heat Remova! Existing RHR Modified RHR Non-Non Recovery Recovery Accident Core Melt Active inlect'on Probability Probebility l Subclass Arrest la RPV Lower DW (P, ann) (Prawn-Pol IA 1, Illa 1 CM Arrest 0.05 0.06 No Arrest inject 0.01 0.02 No Arrest No in6t 0.1 0.11 181_0 CM Arrest 0.0 0.01 No Arrest inject 0.01 0.02 No Arrest No inject 0.1 0.11 182_0 CM Arrest 0.05 0.06 No Arrest inject 0.1 0.11 No Arrest No inject 0.1 0.11 IB3_0 CM Arrest 0.0 0.01 ( No Arrest inject 0.05 0.06 No Arrest No inject 0.1 0.11 10, IllD CM Arrest 0.2 0.21 No Arrest Inject 0.2 0.21 No Arrest No inject 0.2 0.21 g N ABWR Containment Event Trees - Amendment 33 190.5-47

21A6100 Rev. 3 ACWR standard setery Analysis neporr l Table 19D.5 9 Impact of Modified Values for Recoveiy of Containment Heat Removal on Containment Performance Existing P3sults Modified Results Class lilD sequences with in vessel recovery No Cont Leak 2.16E 10 2.13E 10 RD Open 5.34E 11 5.61E 11 DW Head Fall 5.40E 13 5.67E 13 Class 101 sequences without in vessel recovery or drywell sprays No Cont Leak 2.11E 10 2.08E 10 RD Open 4,43E 11 4.66E 11 DW Head Fall 8.94E 13 9.40E 13 Class ID sequences without in vessel recovery or drywell sprays No Cont Leak 5.07 E 11 5.01E 11 RD Open 1.82E 11 1.88E 11 DW Head Fall 3.67E 13 3.79E 13 ) l 190.5 4 ABWR Contelnment Event Trees - Amendment 33 l

.__._m._. e 23A6100 nov 4 ABWR Stenderd Safety AnsIrsis Report ( 2 h s'r os$!IbTv. t.o a st PAT 3 (Top 8LAti -(TOP 8LAtt ~ L i "14 0 I EQPT AND PER$0NNEL

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i r 1 \\ minim *M WMWR I y o r* i 'v Nek'tL sis $, T5 1 -8 y n 4 E e ~ i m a l Structunt failure occurs at 0.924 MPaG when temperature is 260'C. ( Leakage through drywell head, equipment hatches, personal air locks occur when l temperature exceeds 260'C and pressure exceeds 0.358 MPa. Figure 19D.5-1 ABWR Containment Failure Location and Probabilities Aswa containment event rroes - Amensmont u sso.sas

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23A6100 R2v. 4 ACWR Sr:dard Safety Analpis Repon l .a av mm .o

owso, osu 8

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- QM M%i 1 De u*p

.10 041 Coste eaURg JO W UO temp 1 CO W #tTACT M8ut pt'IR W 9 03 upta A8 l A .10 C0c t4tst! N ji w09 Meet t$ . low m v 0 ?to u*sa ST E CCM #Mumf snAa,t K .08 eenwout 0 FP upsa cont? mit a("' J Rn C4>ag$$ coc eact M60d MV 0 "f u*sa C08t? M ACT 10 0 01 (Cow pc vin, g woug attya ew 0 frg um A C0:49 e Aat n a 4 .19 aos Meeg $$ tow mv 0 771 un a CCe ec n JO )

so M*WE O TPt um a CCe e A5 a

g, J RuOtomt$$ Co c e Aa* N0d M%' 9 44 ue a (*0ptT PMt81f s qg 0 01 OC# #" m 90.aOv is MTa w 9 W w*ea os I O CCJ4T SMURf 'f A .86 COW utT AM te0T M**t55 (cw m v s co unga Je E CCM # Ar,Uent A2 LAME 'O too uswt 0 M u*tA CCe e AO' g J Rs M*at$1 CCM e ACT e0e e c taCa M Y 0 to3 up a A6 g gp go p3, 00i eCe,=v=, MIR (90%' 0 T't 48'ta Coatt shtT ACT p 8

1. - rt M *eES$

LCu r2?Y 0 7't W CC*tt e AOT Il Jo .e ue 0,- ete ~rc J i j Figure 19D.518 Containment Event Evaluation DET for Probability of Early Containment Failure High RV Press and Low Cont Press Sequences l 190.546 ABWR Containment Event Trees - Amendment 34 i L.

I \\ ACWR sunbresafery Anotysis aeron A em sursase esasse 6e sennse everus's apt er leat'Is her eas D&7t IEh labartts Sw rf e E 4 LES ( 4EW449 esekA&.J ANTS 19 %.

== e w sun a==u E it = f

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s past to past 3

en L T 8 4 fig D i 5.0G 4 ese tv temat el0 to f = A g ( 6 el a l a

e WIhat Fy spas, S

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4.0 182_t LDf 8s tLD ,I = 1u:. e...-., j i a ut e..s of..ftG e . f ue, .e..,.,,.., e i figure 19D.519 Containment Event Evaluation DET for Active injection to Lower Drywell ABWR Containment Event Tree:- Amendment 33 190.547

tw100:::v. 3 ACWR Sand:rd $sfaty Analysh Report ~ \\ =, g* n' i ) ro... ) Figure 19D.5 20 Containment Event Evaluation DET for Passive Mitigation 190.548 ABan Containment Event Trer.s - Amendment 33

b nAs100 nov. 3 ACWR ssentedsetoryAnoirsIs neport i = r F,

== = umma so g is e e.am m [. thy .a Mg 4 55 gg. gg g g yyRat @qp MTF esten .*.8 D. tima* g 555 fe.vttTEL teu 4 g a93 .i = te,WEA' 041f 8 3.p mM g s l f 4 3 ifIe _9 ett West 33 e.. .>'r 83 es @ e.. k *M g IO - Os it.At' . Du WWAT e se ta tt. Re wM g e ( fe.wttgrt fW 49 Sy Was9 te @ . D '8 y S e g M M'M11 es tEp - wM g ..,m, AL L eM et re.wfteEL tau .. Waat es 'tp e =. < = = = = ..,m, Figure 19D.5 21 Containment Event Evaluation DET for High Temperature Failure ASWM Containment Event Trees - Amendment 33 1905-69

23A6100 M:v. 3 ACWR stanw surory Anstysis sonn N i sa. I mman.2 e., ime.maars .eng 4.. es.=.se.'t r ue w sat det=to L on <= 4 ss. s e t a.g o j ca v ea t %

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y "' - '.,'_l=,<=, .u m .= s, -m .m g,.. q=; g "' - ' ", - ~

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c <., ,7 45.. .m __,m ..m g e o. ~ m ar 'ren 3 t.e g = -' u .1,.. .m i II . ET t g P.L. .P_ R. _.7 M L.A. W P.'Em m u jn r.l ]7 usie e ert a y ;> "".- m 1 tem +r fr' "' - ya em -m b 1 Figure 19D.5 22 Core Debris Concrete Attack DET I 190.5 70 ABWM Containment Event Tre es - Amendmeet 33

ssAs100C n.s ggfg SaendardSafety Analysis Report = = = , = = = =

== =

<

== r 1 esame. en asume ** 8 g.pv.. tti P.J458tf 5 "M C" W 0 86ft g e . ' O Ik f.

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se i e Of* 88 tL q 4e sa t ft f Ett f a fLM g se i art ea tt g .n v, 5.3 89 4 are fettJu 9 6t t se are #sft g e5 t#1 { ste f ettJu g ea j e ett f sft gg 49 up WWft 4, e I set totta 1 se gg j ....a ~6 4 y tegensf * '1 ..x. 54 se set f aft ge 5W ps w merM i'1 63 4 seg settJE gg 5 On f 1 a EET Fett gg 55 t/t I e are setta gp g.5 i es EED Fall gg ] a5 t et 54 pa.s.ta ttnsf gg j ....n s us tasats 'f, ey ... ft. ....m t!t ...ft, I i l Figure 199.5 23 Containment Event Evaluation DET for Pedestal Failure i Aswn containment tant nues - Amonsmont ss sso.s n

2SA6100 Mov. 3 ACWR stuhrt sofory Aantnis nunn l .crare tuu esi, arte erm* ** g W 7_5 eE.O...., F g,(his inJha all 648 L J* d %J D be MtM g tu esagte I 8 Bl ime eMa ef ter g 95 14_ t f f f s 1 _he Mf?Y g

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t.e. = m n 33 re tire ru e, rei w - <w u 'a .,e -i,

= =

m ec IsEC, IO8 w - ~m n Figure 19D.S 24 Containment Event Evaluation DET for RHR Recovery Prior to Containment Struct Failure i 190.5 72 ABAR Containment Event Trees - Amendment 33

22A6100 C n s ACWR stentorstsetory Anerysb noport C =- .=. tithi -e 4 .gagan g an A vG sta p .,u.. .t, ne' thft 08f 4 p I et A 60(L 98 I 65 ( \\ l '5" enhL Lies

e. s.tia g

vt g'gts Mg LM List GG)G Figure 19D.5 25 Containment Event Evaluation DET for Suppression Pool Bypass ABWM Containment Event Trees - Amendmsnt 33 190.5 73

23A6100 Mov. 3 ABWR smadossakrykstysmanon R EMB I N: .sur.en.as es.*..<.a.e*' a. 1. r, u. es. mea um LC5 (.[hY.. e5L T tPg JC] C's.5Pma s C _l'eur L5 ,} . to. u.. e.. gg .f>'18 *t g . Me se mi y . UB d> p esm at 4 ** .y e d # s tt s.e totte e er eng11 . f1D LI A8 g .4 49 ft? ..>. tsaa ' .M* '.Mt g es

. 8.M

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2348100 Aey. 3 ACWR ssenterd setery Aner els soport r ( -,.M = .e ese, se ame 6.w 4 e.nr e.me. em.me e F, m e, rse seen sienn wom.n o% = = + f*fh1 ** M'e rk Truf Jc; 10 De Juma s C M E5 It) 4C5 an. sury, e er eats e fe' Ltas g to e sat t ' @ gutw e 984w g, se see We, e e eartl e== e 841Lt te 460 #st6 g 5.54 to st!Dv e e atti e fe' tip g e W Sutt 2 trtw g I" Py gunse o se sumat e= W #6tti k 440 fif t g e.e MCfPt s es e Wres e wee 9 6W em Du tema'

W 8Etg em e tettt Br eth g

eu e @ eM AS .Jn* Ll U g g, 53 e s. E pt* W Ntes 5W , gg g I e f afLi be 4 &S 9 t fL the EE!!w 0I g.95 e e W 8 erst W ft.f Lias_ g 5.4 K Bute e aptw 45 , g ggg d' ?!? 4.5 e 8 4 fLE ap teaD #e?L e= g, 4N W Wet W Wet gg to tense

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.= m f aTLE Br ef".a0 #ett gg 5N es arten e e SWWB as fWat LIM g e== E. turw e WWW gg = W. e ,..m. ...L ..6 g,,,. e ** B Wlu W WWW g e as muset e e meet e== e pall $ gr staf 98ft vt f a t te 5.54 8 ** e W SWEB te Caet Lf M gg 5.4 W Butw te tett gg a W setts 'I E #4fLG hs 4 40 # e ft gg 4.96 d ' ft f e W E8tw W Opt = g9

== ~5, e ... ia ... fL 5.41 a e t. 9, i= e-9,. ....,L Figure 19D.5 27 Containment Event Evaluation DET for Late Containment Status for Sequences with RHR Not Available at Core Damage A8WM Containment Event Trees - Amenctment 33 190505 l

21A6100 Rsv. 3 ABWR standant s:tery Analysis soport ~E l

=

l 7."" ) 1T* l Figure 19D.5-28 Containment Event Evaluation DET for RHR Recovery Prior to Rupture Disk Setpoint Pressure (Class 11) A3WR Containment Event Trees - Amendmeet 33 19D.5 76

2SA0100 nov. 2 Stendard Safety Analysis Report F IL I Y F tytmi.. m I{

e Figure 19D.5-29 Containment Event Evaluation DET for Rupture Disk Opens (Class ll)

ABWR Contsinment Event Trees - Amendment.12 19D.5 77

21M100.M..n 3 .A MC2 (R stent:nrseieryanaa suneoors r s s E s_ -S E_ __ _ 4 i I f 1 g, I g-i 8 ( s t u _._ s 'l I I I 1 I s ts.ar ti.tLie or g l - ) f i i l . - = = .~ ; Figure 19D.5 30 Containment Event Evaluation DET for Core Cooling L 1, m Recovery (Class 11) v1.18D678 -. ABWR Containment Event Trees - Amendment 33 ~

zuisbonr s ABWR .._ h** M W h*

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2 I ...g y ,I - _ 7.

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.t.. s, a.. s. _.s_. L..a _. 5 5 8, s..u a = i 5I I e!j g r -li o <a v r l l l E 3 0r, g I e ', 4s [ l E It f l 6. .j i i o y ep! a It B l 7 i >t \\ ek ~ ~ Il 5 ' - - ~ ~ ~ ~ ' - - - ~ ~ I !!g l >= el 1 g II 8 I m a J.

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f C b ~ e II 1 e o E 5 35 !r I Nf$ Iti 5]E C (- Ha a s19 r[t t 4 ABWR Containment Event Trees - Amendment *a '90'.0'lE (>

-.___.__._.._..,_m_.-_.. __._...-m ......._4_._.~. _... _ _.-~ _.m_ 4 e 1 2348IN Am6 I g6dSgfstYAR3h g I 4 g g 3 ssaass8 8 5 5 8 j I g y 4 k 1 at I _c 1 l}8 1 i i i e i s 1 e i 4 a ~ y !E e 4 I g s-O I e E 4 a [ kk I g N 8I>3 BII" kl l s a I 190.M ABWM Containment Event Trees _mendmenW

~ 9 IN May. I bDW}f ~ ~ ~,., ~..,,,,,,,,,, i l gi Oh! 11

t!!r,e,jj u

d M Oll" 's El p g,- la,l Og-I s$ 0 l yW $ 5 gl <l" i I I I a !k j l g _a s a a r ~ lil U n I a d N e i !!,!<d a -ub k af.. illl Oj ( 96 d ' ' " " < ~ - +.,,,,,,,,,,,, 190.5-227 ~ ^%

a 2S48100 Mov.1 ACWR Standard Safety Analysis Report i llI ' f5 e I Oh! I ~ 5 m i8 E 9 8 a;i Oi! i I m ~' Wd ne .5 n i8!1 8 9 8 q"! j{!E V S 8 e .6 5 5$ avgh j-n n .b I"ii O';;, di ) 190.6-228 Fault Trees - Amendment 31

Mf 00 Mov. t ACWR hated $sfety A3alysis Report l 1 I 81 3 $*l 3!! I, s E yi G 1 _1 k ( n = mW I E e 2 Qd r-b l [ m d E N ! Oli i i e C Fault Trees - Amendment 31 19D.6-229

E h ? as U FAILURE tr Bly MEC1 KCV TO P90 VIDE ECCS CIILI% i v4CVMEC> LDSS F MM LDSS F NM TCV MEC> FAILS TO ESMitA TO C4ML 10 PROVIDE E MRATOR ELEC. EQUIP WELED WATER =0m mmc > =0m. MaC> b k vtttcuac) vacuac) w w CV M sC) g tr e ris e> = 2 TCv P25-FOl6MaC) TE M S pestst ICV P25-r822MEC: TEM SDtSGt FAILS CLOSED tJ41-TE-II34EC) FAILS CLOSED U41-TE-952 OETC) FAILS GETC) (9%B68> FAILS y ,T E 5 vretsMaC> vitii3MaC3 vreeeMac> vrEas a s u e> E. s trot-es res[-or erot-es e.4et-7 p e i q I I rgure 1 g a .~. P Figure 19D.6-23a HVAC Emergency Cooling Water Fault Tree j l

E h D A ECW AfS.C1 FAILS g TO F10.1 { TO pro %IDE Q CHILLED WATER 1 I WHECW1A(9.Q s e .I. ?. 4 FAILURE OF 0W togs of Agey pumpog WYPASSteE AIB.C) RCW TO AC POWER FROM REFROGERATOg PIPE RUPTURED FAILS OPEN PROVIDE COOLN BUS C2(D2.E2) FAftURE FLOW V9YPA98AIB.Q WPRA(8.Q g WRCWA(8.C) EAceC1(01.E1) WPPHEAHISH.CH) ,y IFROM FIG.1sO.014el FROM FIO.190 e-12e) 3 44E48 3, g FIRST PUMP OR SECOND PUMP OR PCV P25 F012A(0.g DFF PRESSUE TEMP SENSOR REFRIGERATOR REFRIGERATOR FAILS FULL TRANSNR P25-H405A(9.Q FAILURE FAILURE OPEN INOFOl P25-DPTNAIS.Q FAILS g, FAILS g P vPuMPA(e.o vPuMmEn vFvrei w e.a vPRooms.o vne,,A..a FROM FIG.Si trROM FIG.31 2.2eE45 1.2eE45 f 2.deE47 .e FIGURE 2 g. O at h Figure 19D.6-23b HVAC Emergency Cooling Water Fauft Tree f

E9 A PUMPon TO M 2C RETRfGERATOR D FAILURE / VPUMPA(H.C.DJJ) I ' LOOPS B AND C HAVE TWO PUMPMXS. I ONE LOOP. SET HEREIN AS LOOP 8 M M TOR PUMP HAS BOTH PUMP /HXS RUNMNG FAILURE FRURE LOOP C HAS 6TS SECONO PUMPfHX IN STAND 5Y. WH$RA(9.C.D.EJI WHSPMS.CD.EJi y i n OFTRATOR REFP5GEAATINI STAND 9Y 3 . PUMP PIETNC#8tATORIN FAILS TO FMURE I. FAltuRE START PUMP

MMNTE NANG' e

O WPMHC1A(9.C.D.EJB m RR4. -f,-Ji WRrD1NS.C.DJ. WftMNT4.ffff) 1.94E 44 1_oct43 8.00E 44 sscE m 1 MANUAL gTAND9Y PUMP STAND 9Y PUMP MOPCV DurF PRFSSURE STAfe9Y PUMPIN P21 F025A TRANSMITTE R IMT1Afl0N FAILURE TO LiG CHE CK VA* VE y SIGNALFNLS syAgye P25F001F. MMNTENANG (9.C.EJi F AILS Mg,C,2f3 ( e FMLS TO OFEN* CLDSED Farts a c i A O O O O O j l EMSCONN4...3) WPMSTRT4...-J) WCVH11 f. ff) WPMMNT4.-ffft wpV075Agg.CDJJI VPfD00fAf9.C.DJJ) IFROM FIG. 190 6-15e) 2.27E-03 1.40E44 9.22E 45 2.20E 45 1M-05 g i 9 ar nGuRE 3 g t' j Figure 19D.6-23c HVAC Emergency Cooling Water Fault Tree j I ~ ~J e e

23M100 Aey 1 ACWR sannseedsetery Anstrels naport 1 l -( l i 81-si e af E5 l a li. In (x8 i g as ng .e E E f:a I ] t ( n c 3 h E sy Eg u Ej l5 B a } 8 i <l I I b$. jf ~ $55 N 8 E C Fault Trees - Amendment 31 190.6-233

ll4lllIII \\ i s ay. >ag P5R %$e>2Ta Pj J T L DI s LA Pr 4a e I P A3o Rr e i S G r 4 E T P V v L WA P t l RV a u O a F me t T t. 2 s Df LA y I r 3a S P P \\ r# r n i S 3 / E s Po c. P V v o L e i @A e t AV a c n C 2 e ot s t t cA ?/ e o sF s r i C a P e. I o G r - O r i o L T 3 L W D! 2 t LA c Amr W S r a I a C P i L SI s t e A P t s R 8 Pe I'F Pv v A L t Da Rv w b C 4 .S 6 2 0AN B 0 R t t- ,E 8 6 ,V I E AI D R R D M r c I 9 cs c Tt 2 l 3 1 Dr G La e I Ir Pn r As r P a Rr u s c D l t g Po T Pv v i UL F DA Av n r .ste rn e:-lg:a! $ l I) 1lllll Il lll r

23A6100 nov, t ACWR SenaderdSafety Analysis Report I E G 1 a g lii! (E'l 8. e! [y <.lf g e 1 a -s e s.e ~ f b b i= .i Ef r" A I E W g E, - Ei g' g s! g -1 .e h l us: y ( p:, ~9, ( o s c g* e 5 1 ~ e w E = gia -e 5't N G N6 6 e ~ si d b b fje (.3 d r s u g, b:

n-

'I t e b 5 wg 1 e I E W ( s,' - E. 118! 'l fault Trens - Amendment 31 190.6-235

2SA6100 Mov.1 ABWR sandent serarypatreis nopen I av 7 bb ~ hy b n

g q18 q

= = g a 1.i C E-t <J l!! O$h 7 a .e j 1's = g s ) Rd 88 l e b <Eh h N 3" b N 9 gg E d ~ 5."E N Edd P -'s., u ns Sgha H 173= g *. g s s '!! b!- ( s ~ li C E 1 190,6 236 Fault Trees - Amendment 31

2140100 Mov.1 AGWR sanndardserary Anerysis norors JE = e h de 5 if ('18 e d l I Em I;'1 G i I gi f fji' l '5 C ilig si! Sl-i ~ g; si i I!! Oil I ~ E5 h n h e s E JS f L e a e n h e$ dh N 4 h 5 _e b 4 j$ w E** f bIE K3 e N' i u" b E "" E "I G !! 31i liis g; ig I!! obi is ~ 9 Fault Trees-Amendment 31 190.6-237

22Astoo sev, t ACWR smaantserary Anery,Is neron ) kk 0 8 l$! <! .E5 I li I' g T Ej G j llI 5 p 3 15ECEl f. i E> 0) i 13 c g g. t e!'!C 5 E 'a k Eg e a {EH M h"g g

md N

Ww 4 u a N 15 g i g r' 'E$ I' m e ES ~ 13 1 19D5238 Fault Trees - Amendment St

e 4 2148100 flev, f ewa ~ C q! 8 1- 0 I flE i' il i -In9 Q l-118-r; j ~ i 11! Oli i ~ I: t M E* d E' 4 ~ !;! G i 1 g d 6 Q .E q Id Gj i tje I ~E! C E I'IE g: I! b$ E$ is s Foutt Trees - Amendment 31 yyg_g,ggy

l llll 1lll\\lJ IIll ll jlI h2 ay ~ Na 3 P E R y;;;9>2 3 7 j ~ 1 ~ s w t a :EA s s a vi Aee xuTr er nA tRs n e e a [e t r stv w T sav t lua F m e t m s As Rl y cl sA 3 S F r n c ut r o cE i u sL c aI t e s torP ro s s w t e t B c n oDr A n a t EA va e A sm i s cTr rr e

s nR-A A

u u R r rt a [r t g h A stv r L a F 4 Av 2-6 D 9 1 e rs r 1 uY u G cA s g to 1 0c2 r A L t - i sE aE F R 0 r4 7 rM tA c5 rR cC cs $+ e& eg lg3m23 llll lllI llll llll l ljjj\\\\1l{1, ll Ill,l!ll

2348100 Aev.1 C !!! Oli 5!! <!.! ri p E I, sia D a '"8y Rat e-13 p.. <j8 8 i m a s e E. i, _a i >g. Mg m E d Q' c I j .Y0 gh O ,e y'. .t ( li! G ii <1; _1 { lh a i is <s: p rI a N-O D e i s 30 rl- _a "m';<is E"a! B l{g o m e Isg St E ll1Oli ja <!! (

~ b es a k l1l: e s dj sac o sacrte c: w n BacKte sCIMsg IC'* WIES TD rE 9 D* UKI TD FIG 9 PanaLLtt 70 FIG 9. RELAv VALVE A rgg gg WELAv vaLyt 3 ottAv ram.s rgg gg ran.s - oiv 1 rafts - ery 1 t tstiMt2Ely.14D> stinalet,tansers

setsuacs, r

scease Panatttt scuase PasmottL y envi starts arve envi stetts give- ~ attAvs rait. mELAvs ram. 3 attAv u nc.D> steent rAn.s attav castr> snoemt rarts b g g nsitutet.83ttic> essesuwaser: esitarursr. sacr.i4er> sisresesr.sasr.sasr> .scorurtr. 3.r.isr> sisresesr.s2sr.sasr>

3'IM FIE 18' 680M FE 18)

P&2E-03 WIIDE RIG 4 & 55 eME-93 anDer:G 4 & 53 P e E. S E' 2 rigur e to D l A ? 3 A. a if a Figure 19D.6-24j Reactor Protection System Fault Tree =a a ~ w 9

I 1 sAs100 Mev.1 ABWR StenderdSafety Analysis Report 1 i ~~ h,! M I . = - 4 h E w"! d/ .i I I i s., 158Cl i,! OI,! a s1 we l 9 E ze b h, v B I i ( i j! OI! = i b li 4 3 a E ~ s., es 2 l' Eh l E 5en hs' E Q bE d ~ g$ bb ff $Y l $E $-I ( I 5 ( t= ee t, t e fault Treen - Amendment 31 190.6 243

t norm > < ABWR

- "ha s 1,ry a,,i,a,,,,,q E

I!!,Oli _c e w N~ pg b$ 5 n E 0 = 5" g h}. 19 s a N a ~ 7 ytt yt Ee W E -1 gl Y g [ h"i < be i eg h y ~ e I C /, b s ,x n! Q qi = a 4 5 Cu -l f bk eg u { 5?? NY = I .I.k p h N 5 8 b Y e g s 8 = a E 5 s b i! < d" I eg m i =n 5 {* byg p p big et u rt g 190.& 244 Fault Trees - Amendment 31

e 2$AS100nov.1 ABWR stantant sorety Anerrsis sepan O bWf < .,l' gg-, E i e I d @ OIl ld OIl g EN N Y y 5" NM I Pi O!j ad!Oli t 11 W:. E e 0: 9 !g 3 ( e o E5 i!! Oli S I' ~ "y t i a E E 4 e a e W s au 2 6 ej s! "l 3 = !.,! OE! !;i C i-a ( !!! OIk ( 5 " 'n - ~ a ~ e Fault Trees - Amendment 31 190.6-245

23A6100 Rx 1 ABWR standant setory Aner sis arpon r I 5 6 ag (goc s d >I y g. g3 ~ O' 3" 2 g I., g"!" of E ~ a c i-i l"E g a s!g hje = E' x g ho 3 2 5 ') gf 2 ( !lE ss E l a~ ~ I E e C W o c ~ j 1 _2 ) C 4 w0 De Em A ad c -d hd 9 g ga N a m.: "5 " h k d ~ Ii 3 5 - f"g gj g r 2'9 st cB" lD "e a w G5' 6*o D u ga 5 g-m 3 hG [ 1 25$ u% t !f 2 inf ss E l 190.6-246 fault Trees - Amendment 31

a 2sAs100 nov.1 xwa t (>l ~t I-is y' eg y a e E k" i-! !>guSii J se = g,!i3 11 i E a g <y-e W: = ~ 1 ( c s b E s I u-a e

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d =t + e E' iii '* Du O 2:! CI !Il 3ll k,., ye a = n c 5 t ?!! 31I y-e e e feutt Treen - Amendment 31 ,gg_g_y4 z.,4g

23A8100Mov.3 ABWR sa,ndedsafety Analysis neport ,t Overall CDF sensitivity to HEFs 1.00EM O An Human Actons j. a O Qass AHuman Achons 1.00E-05 Class C Human Actons I 8 1.00E 06 i u 1 1.00E-07 W 0 0 0 l 1 I 1.00E-06 l l @ l @ j } 8 8 8 a NEP Factor 1 ) 4 4 i Factor CDF CDF CDF Class A Class C ALL if29 1.17E 07 1/25 1.17E-07 i 1/20 1.17E 07 1/15 1.17E 07 1/10 1.17E47 1/5 1.18E 07 Base 1.56E 07 5 7.47E 07 10 5.64E-06 15 7.24E-06 20 9.89E 06 25 1.25E 05 29 1.82E 05 Figure 19D.7-4 AB'c1R PRA CDF with all HEPs Multiplied f,] } Simultaneously by a Factor 190.740 Human Enor Prediction - Amendment 33

23A6100 Qett. 3 ABWR standant setoryanalysis nenn Overa;l CDF senshMtyto HEPs 1.00E44 M Human Acons Cass A Human Acons 1.00E 05 f. CVss C Humaq Aacns h 1.00E 06 j 0 0 0 0 3 i-E E 1.00E 07 1.00E 08 l l l l l l l l l l l l l { { HEP Factor ( Fador CDF CDF CDF Class A Class C ALL 1/29 1.51E 07 1.16E 07 1.15E-07 1/25 1.51E-07 1.16E 07 1.15E 07 1/20 1.51E 07 - 1.16E 07 1.15E 07 1/15 1.51E 07 1.16E 07 1.15E 07 1/10 1.53E 07 1.16E 07 1.15E-07 1/5 1.54E-07 1.16E47 1.17E-07 Base 1.56E-07 1.56E47 1.53347 5 1.65E-07 3.86E 07 4.13E 07 10 1.65E-07 3.86E 07 4.13E 07 15 1.65E 07 3.86E 07 4.13E-07 20 1.65E 07 3.86E 07 4.13E 07 25 1.65E 07 3.86E 07 4.13E 07 29 1.65E 07 3.86E47 4.13E 07 Figure 19D.7-3 Results of the ABWH Sensitivity Analysis (in Comparison to the Results of Reference 19D.7-6) 190 7'39 Human Error Prediction - Amendment 33

e 23A6100 Ae11 ABWR stenderdsetery Anstrels noron !C 1 i ? 10 ~ 1 h kh 'g l 1 - i io I F l 3 W l Kle-P !O ,I i E' 5 nft t 8 su E E i g [ t fW f,I 'jO a I )g l I fp Seismic V :: ns Analysis ~ Amendment s's s9122 2

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2SA6100nev,f awa . a C I l e 8# 1 11 io ]? a ( l5 = W f% h h ( Seismic Margins Anotyeie - Amendment 31 19I-25

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2 SAC 100Mov.1 ABWR sonderts*!*ty An*Irsin neron j-li I:n 4o I t ! e i I> 1 E B y l3 $G 5 "4 S f u I ie a h h n l to i 1 lO c.. l! = Seismic Margine Anotyels - Amendment 31 19127

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rs vE E 2 w T 4 S2 v T U U -c s CZ Seismic Ms? gins Analysis-Amendment 37 191'33

23A6 f 00 Mev. 3 ABWR senM setory Anoirsts soport I l r i 5 N j ? ? a s2 o o b b b k b k 8m I h = m 2 V) ,a R g N i 1 j. 11 l i ~ li 3 k i o w c3 l Seismic Margins Analysis - Amendment 33 191-34

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23A6100Mov.2 ABWR senterdsermAner els neron r C i l l l e i B 5 5 6 5 8a l e e l b E E k I 1 e = k d e ( fI k A W N I 8 g v e o h f> 1 n ( Seismic Margins Analnis - Amendment 32 19139

. _ _ _ _ _ _ _ _.. _.. _ _ _ _ _ - _ _ _.. ~ _ _.. - _. - - - _ _ _. _ _ - _.. _. - - -.. _. 23A8f 00 Mey. f ACWR standardsevery An*Ireb nopers l t Y r i, e e, w 5 0 b j la ] 1 5 l g R tw k ) i 5 5 1 L IS 8 e W f E E [ c a i e j 8 s b' ) 191-40 Seismic Margins Analysis - AmenJment 31

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Vacuum Breaker CLASS Min-Max Convolution Event Accident Closes in All Lines i_

SE SVAC CHVLV n SE SVAC 15 BP O.74g 1.12g Pt i Pe M .I v I I Figure 191-25 Suppression Pool Bypass Vie WetweN/DrywoII Vacumn Brooker Lines } I

2348700Aav t ABWR smewseterAnaksisn ers e Case 2 Iow Pressure Core Melt with suppresion pool bypass and actuation of containment rupture disk. Cue 3 High Pressure Core Melt with drywell Head fauure and fire water sprayinidation. Case 4 Suppreulon Pool Decontamination reduction (Not used). Case 5 Large Break LOCA without recovery and with actuation of containment rupture disk. Case 6 High Pressure Core Melt with Drywell Head failure and no Brewater spray initiation. Case 7 Iow Preuure Core Melt with Drywell Head failure and no mitigation. Case 8 High Preuure Core Melt with Early Containment fauure. Cue 9 ATWS event with Dr)well Head failure. ( NCL Normal Containment Leakage to Reactor Building. The offsite exposures for each cue shown in Table 19P 1 were calculated by the CRAC2 code for five representative US regions for the selected individual ABWR sequences as discuued in Subsection 19E.3. Table 19P 2 provides additional detail on the individual contributors to the total core damage frequency. As indicated on Table 19P-2, the core damage frequency is dominated by low pressure transient events (LCLP) (61.4%), followed by high preuure transient events (LCHP) (28.1%) and station blackout sequences (SBRC) (10.3%). Resiew ofTable 19P-1 also indicates that the dominant contributors to the ABWR offsite exposure risk are the relatively low probability (less than 4E 10/yr), high consequence even*s (Cases 6 through 9) which contribute about 82% of the offsite exposure risk. 19P.3 Potential ABWR Modifications Potential modifications to the ABWR design were derived from a suney of various studies indicated in References 19P-1 through 19P-7 and the ABWR design process discussed in Section 19.7. From these, a composite list of modifications was established. This list of potential modifications was resiewed to identify concepts which were already included in the ABWR design or which are not applicable. bI a Veluotion of Potential Aeodifications to the ABWM Design - Amendment 31 199-6

^

2SA6100 Mev.t ABWR sententsareryAnrirsts nerort ) Table 10P 3 summarizes the complete listof modmcatiom and their classification according to the following categorieg 4 (1) Modification is applicable to ABh1 and already incorporated in the ABWR design. No further enluation is needed. (Table 19P.3 prosides a creas reference to the supporting subsection of the SSAR.) (2) Modification is applicable tc, ABWR and not incorporated in ABWR design. O able 19P-4 lists the Category 2 modifications which are evaluated further in this report.) (3) Modiacation is not applicable to the ABWR design due to the basis provided. (4) Modification is applicable to ABWR and is incorponted with the referenced moditication 19P.4 Risk Reduction of Potential Modifications This subsec'lon provides evaluations of the benefits of potential modifications to the AB%% design identified in Table 19P-4. For each modification the basis for the evaluation and the concept is described. Table 19P-5 summarizes the benefit in temis of person-rem averted risk for each of the evaluated modifications. 19P.4.1 Accident Management Accident management is a current topic under generic development within the Industry through the development of Accident Management Guidelines (AMGs) and revisions to Emergency Procedure Guidelines (EPGs). The following modiScations are based on implementation of such generic acthity. 19P.4.1.1 Severe Accident EPGs/AMGs The r/mptom based EPGs, were developed by the BWR Owners Group following the accident at Three Mile Island, Unit 2. CmTently the EPGs are under revision and accident management guidelines (AMGs) are being developed for severe accidents. These should provide a significant improvement which reduces the likelihood of a severe accident. Elements of these guidelines (such as containment pressure and temperature control guidelines) also deal with mitigating the effects of accidents. In the ABWR PRA. Emergency Operating Procedures (EOPs) are based on these guidelines. Additional extensions of %e EPGs and EOPs c<.,uld be made to address arrest of a core melt, emergency planning, radiological release assessment and other areas related to nevere accidents. Since the existing EPGs cover preventive actions and some nnitigative actions, the incremental benefit of this item would be primarily mitigative. It wasjudged that the 19P4 Evolustion of PotentialModi6 cations to the ABWR Design - Amendment 31

nM100 Rn. 4 ABWR standardsetery Anso sis neport r reliability of manual actions usociated with mitigation could be improved by 10%, especially in use of core melt anest procenes. Failure rates for manually initiated mitigative systems were decrened by 10%, to estimate the benefit. The resulting offsite l risk reduction is about 1.5E-.04 penon Sievert over 60 years. I 19P.4.1.2 Computer Aided instrumentation Computer aided artificialintelligence can be added which provides attention to risk issues in man-machine interfaces. Significant computer usisted display and plant status monitodng is already part of the ABWR control room design. Additional artificial intellige nce could be designed which w 1d display procedural options for the operator to evaluate dudng severe accidents. Tne system would be an extension of ERIS to provide human engineered displays of the important variables in the EPGs and AMGs. Operator actions are made significantly more reliable by new features such as Emergency Procedure Guidelines, Safety Plant Par =mner Displays (SPDS), and training on simulaton. If the improvements described in Subsection 19P.4.1.1 are assumed to be implemented, the incremental benefit of additional improvements is expected to be low. The reliability of manually initiated preventive systems was increased by 10% to estimate the benefit.The estimated incremental benefit over severe t accident EPGs (Subsection 19P.4.1.1) is about 3% in core damage frequency (CDF). Because the improvement affects all release cases, the incremental benefit is about 1.0E-04 person-Sievert. 19P.4.1.3 Improved Maintenance Procedures / Manuals For the GE scope of supply this item would provide additional information on the components important to the risk of the plant. As a result ofimproved maintenance manuals and information it would be expected that increased reliability of the important equipment would occur. This item would be a preventative improvement which would address several system or componenu to different degrees. Bued on a 10% improvement in the reliability of the High Pressure Core Flooder (HPCF), Reactor Ccre Isolation Cooling (RCIC), Residual Heat Removal (RHR) and Low Pressure Core Flooder (LPCF) systems, the CDFis reduced by about 9% which has l a corresponding estimated penon-Sievert reduction of about 1.6E-04. 19P.4.2 Decay Heat Removal - Significant improvements in the reliability of ABWR high pressure systems have been made. Among these are RCIC restart (NUREG 0737,II.K.S.13) and isolation reliability improvements (NUREG 0737, II.K.S.15). Additionally, the redundant HPCF is an improvement over early product lines which used the single HPC1'sptem. Evoivation of Potsntial Modifications to the ABWR Design - Amendment S4 19P 7 w

nAssoo n:v. 4

l ABWR

.ssuuntsafety Antysis nenn \\ ) 19P.4.2.1 Passive High Pres 6use System This concept would pr adde additional'tingh pr esmur cxpaNhry to remove decay heat. through a diverse isolation ecmdenSes type sym. Sud a sptem would have the advantage of removing not cady decay heat., but corsu&nmerz heat if a similar sptem to that under consideration for the Simplified BWR (SBWR) b employed. The beneSt of this system would be equivalent to an additional diverse RCIC system in addition to an additional containment heat removal sptem. The added sptem was assumed to be 90% reliable, designed to operate independent of offsite power and to be capable ofin-venel core melt arrest. Based on a reduction in the RCIC failure rate, j the benefit is estimated at about 6.9E -04 person Sievert averted. 19P.4.2.2 improved Depressurization This item would provide an improved depressurization system which would allow more reliable access to low pressure sptems. Additional dep essurization capability may be achieved through manually controlled, seismically protected, air powered operators which pennit depressurization to be manually accomplished in the event ofloss of DC control power or control air events. The ABWR high pressure core damage events represent about 28% of the total core damage frequency, but about 46% of the offsite exposure risk. The success of manual initiation was assumed to be improved by 50% and therefore the depressurization failure rate was reduced by a factor of 2. Based on this estimate of benefit offsite person-Sievert is reduced by about 23% and the estimated benefit is about 4.2E-04 person-Sievert. 19P.4.2.3 Suppression Pool Jockey Pump This modification would provide a small makeup pump to provide low pressure decay heat removal from the Reactot Pressure Vessel (RPV) using suppression pool water u a source. The return path to the suppression pool would be through existing piping such as shutdown cooling return lines. The benefit of this modification would be similar to that provided by the firewater injection and spray capability, but it would have the advantage that long term containment inventory concerns would not occur. If the sptem could make low pressure celant makeup systems 10% more reliable, significant reductions in CDF would not be achieved becat se otherlow preem systems are already highly reliable. The estimated benefit is that CDF is reduced 2% ar.d the l averted risk would be 0.2E-04 person-Sievert. 19P-8 Evatustion of Potentiel Mod!Mcations to the ABWR Design - Amendment 54

asAs1co Quv. 4 ABWR staasnt safety Analysis Report C 19P.4.2.4 Sdoty Related Condensate Storage Tank The current ABWR design consists of a standard nonelsmically qualified Condensate Storage Tank (CST). This modification would upgrade the structure of the CST such that it would be available to provide makeup to the reactor following a seismic event. This modificadon only benefits the risks of core damage following seismic events. However, because the suppression pool provides an alternate suction source and the HCLPF for the suppression poolis relatively high (/.ppendix 191), the dominant failure modes are not limited by water availability. Therefore the benefit of this modification is l considered small. A benefit of 10E-04 person Sievert averted was arbitrarily chosen for an upgraded CST. 19P.4.3 Containment Capability l The ABWR containment is designed for about 0.411 MPa internal pressure and includes a containment rupture disk which would relieve excessive pressure ifit develops during a severe accident. By ptoviding the release point from the wetwell airspace, mitigation of releases are achie ed through scrubbing of the fission products in the suppression pool. 19P.4.3.1 Larger Volume Containment This modification would proside a larger volume containment as a means to mitigate the effects of severe accidents. By increasing the size the containment could be able to absorb additional noncondensible gas generation and delay activation of the containment rupture disk or early containment failure. This item would mitigate the consequence of an accident by delaying the time before the severe accident source term is released and allowing more time for radioactive decay and recovery of s) stems. However, if recovery does not occur, eventual release is not prevented and if operation of the containment overpressure rupture disk does not occur, ultimately the containment will fail due to the long term pressurization caused by core concrete interaction and skam generation. If sequences invohing drywell head failure were eliminated (Ca.ses 3,6,7,8 and 9), the l offsite risks would be reduced by about 82% and about 15E-04 person-Sievert would be averted. 19P.4.3.2 Increased Containment Pressure Capacity The design pressure of the ABWR containment is 0.411 MPa (45 psig). The containment rupture disk pressure and ultimate capability are significantly higher. By increasing the ultimate pressure capability of the containment (including seals), the effects of a severe accident could be reduced or eliminated by delaying the time of haluation of Potential Modifications to the ABWR Design - Ament' ment 34 19P 9

nAstooc:v. s ABWR standard setory Aantrsu suport release. If the suength exceeded tbr maximum prenwr obtainable in a se,ere accident, only norm.d contamniera Wksge would cesult This modification would trat get the nent, not cha.uge the core damage frequency and the increased pressure capability may not be suscient to contain the long term pressurization caused by core concrete interaction and steam generation. However,ifit were able to prevent all severe source term release except for normal containment leakage, the person-Sievert risk would be about 2E-04 penon Sievert /60 yem. Therefore, the benefit would be about 16E-04 person-Sievert. 19P.4.3.3 f.nproved Vacuum Breakers The ABWR design contains single vacuum breaker valves in each of eight.dr>well to wetwell vacuum breaker lines. The PRA included failure of vacuum breakers in

Case 2" assuming operation of wetwell spraf.This modification would rr. duce the probability of a stuck open vacuum breaker by making the valves redundantin each line and eliminate the need for operator action.

If Case 2 sequences were eliminated, the benefit of this modification would be about l 4E-07 person Sievert averted. 19P 4.3.4 Improved Bottom Head Penetration Design The ABWR design includes a 50A (2-inch) stainless steel drainline from the bottom of the RPV which is used to prevent thermal stratification in the RPV during operation and to provide cleanup of the bottom head by the CUW system. A carbon steel transition piece connecu the drain line to the RPV. During a severe accident this truisition piece may be susceptible to melting and may provide the earliest path for release of molten core material from the RPV to tl.e containment. The penetrations for the fine motion control rod drives in the ABWR also may proside a pathway for release from the RPV following a severe accident. Failure of the internal blowout suppons on the lower core plate, provided to eliminate the support structure in curtent generation BWRs, and welds of the drives at the bottom of the vessel may allow the CRDs to be panially ejected into the dr)well during the severe accident which would provide a small pathway for release to the containment. The modification is to change the transition piece material to Inconel or Stainless Steel which has a higher melting point. By so doing, additional time would he available for recovery of core cooling systems. This modification also would establish external welds or restraints on the CRDs external to the vessel so that the dris es would not be ejected following failure of the internal welds. The concept would be to make such external welds and supports small enough that the benefit is not lost from eliminating the support beams in current generation BWRs. The benefit of these modifications would be to reduce the probability of in-vessel arrest failure (NO IV). Bued on consideration ] Evaluation of Potentis! Modi 6 cations to the ABWR Design - Amendment 35 19P10

23M100 R:v. 4 ACWR ssentardsevery Anotysis neport of the heatup rate of the bottom head,it has been estimated that making these changes could provide up to two hours additional ume for recovery of systems. It is estimated, based on engineeringjudgement, that this time could result in the in vessel arrest, failure probabilities being reduced by a factor of two. The resulting benefit is about 5/7E-04 person Sievert averted. A potential negative aspect of the modifications is that RPV failure could occur at another unknown location such as the bottom head itself. Although the time of vessel failure would be extended, the falldre mode from these other locations could be poicntially more energetic and lead to unevaluated consequences. 19P.4.4 Containment Heat Removal The ABWR design contains 3 divisions of suppression pool cooling and provisions for a containment rupture disk for decay heat removal. In addition, modifications have been made to use the CUW heat exchangers to the maximum extent possible. Consequently, loss of containment heat removal events contribute only 0.1% ofIS total core damage frequency and offsite exposures. Additional modifications are rsot likely to show l substantial safety benefits. f 19P.4.4.1 Larger Volume Suppression Pool This item would increase the size of the suppression pool so that the heatup rate in the pool is reduced. The increased size would allow more time for recovery of a heat removal system. Since this modification primarily affecu 1 HRC evenu (Table 19P 2), the maximum benefit we,uld be elimination of the LHRC contribution to the Case 9 sequences. These events are mitigated by the containment rupture disk and only contribute about 0.02E-04 person Sieven to the base case risk. The assessed maximum benefit is therefore about 0.02E-04 person-Sievert. 19P.4.5 Containment Atmosphere Mass Removal The ABWR design contains a containment rupture disk which provides containment overpressure protection from the wetwell airspace and utilizes the suppression pool scrubbing feature of the suppression pool to reduce the amount of radioactive material released. One additional modification was considered. 19P.4.5.1 Low Flow Fittered Vent Sorae BWR facilities, esp:cially in Europe, recently have added a filter system extemal to the containment to further reduce the magnitude of radioactive release. The sptems (, typically us: a multi-venturi scrubbing system to circulate the exhaust gas and remove particulate material. In the ABWR, because of the suppression pool scrubbing capability, a significant safety improvement is not expected due to this modification. Evolustion of PotentialModifications to the ABWR Design - Amendment 34 19P 11

23A61007t;v.4 S:ntnf Satty Anhsis Report ABWR The release of radioattive 'tsotopes from thc ABWR follm og severe accidents occurs through the containment rupture disk lor Cass 1, S urWilesequences total about l 8% of the exposure risk.'Ibe remain ~mg sequetces'mvolve derwe.Il head failure or early containment failure which would not be affetted by thh modification. The maximum benefit of the extema) vent system is therefore about L4E-04 person-Sievert assuming l perfect initiation of the filtei i containment vent system. 19P.4.6 Cornbustible Gas Control No additional modifications in the ABWR were identified in this group. 19P.4.7 Conn.inment Spray Systems 19P.4.7.1 Drywell Head Flooding rhis concept would prodde intentional flooding of the upper drywell head such that if high dowell temperaturcs occurred,'he diywell head seal would not fail. Additionally, if the seal were to fail due to overpressurhation of the drywell, some scrubbing of the 2 released fission products would occur.This system would be designed to operate passively or use an AC independent water source. If an -xtem.on of the fire pump to dove 5 spray crosstie were considered for manual ) initiation of upper head flooding, additional reduction in the high temperature containmert failure sequences (Case 8) would result. Additionally, a reduction in the high consequence drywell head failure sequences (Cases 6 and 7) could be achieved. 2 Case 8 sequences were eliminated and Case 6 and 7 source terms were reduced to a level similar to Case 3, the conservative benefit would be 12E-04 person-Sievert. The estiraated benefit of this is about 6.0E-04 person-Sievert assuming a 50% reliability of ini iation. t 19P.4.8 Prevention Concapts i The ABWR design contains an additional division of high p essure makeup capability to impro ve its capability to prevent severe accidents other features such as the fire pump a injection capability and the combustion gas turbine have been included in the design to enhance the plant capability to prevent core damage. The following additional concepts were considered: 19P.4.8.1 Additional Service Water Punips e This item addresses a reduction in the common cause dependencies through such G items as improvec. manufacturer diversity, separation of equipment and support systems 9 such as service water, air supplie:. or bening and venuiation (HVAC). The HPCF, RCIC, and l'CF pumps are diverse in tne ABWR &n,,n since they are either suppised by different rrtnufacturers or have different flow characteristics. Equipment is 5 vnhanon of Potential Modifications to the ABWR Design - Amende.ent 34 19r-12 l

23A6100 Rsv. 4 ABWR statant s:1:ty Aa:tysis neport separated in the AE*VR design in accordance with Regulatory Guide 1.75. Thus, no further improvement is expected with regard to separation. A reduction in common cause deper.dencies from support systems such as senice water systems, could conceivably reduce the plant risk through an improvement in system reliability. The concept for this item would be to provide an additional cooling water system capable of supporting each of the four divisional systems identified above. The current design provides support to these sptems from one of three dhisions. Thus, the effect of this change would be to include a diverse and additional support sys*em. In addition, diversityin instrumentation which controls these systems could be included so that redundant indicauon and trip channels would rely on diverse instnur - son. A 10% increase in the reliability of the four systems was assumed which is t.he same improvement that mpy be derived from improved maintenance (Subsection 19P.4.1.3). l This results in an estimated benefit of about 1.6E 44 person-Sievert. 19P.4.9 AC Power Supplies The current ABWR electrical desi n is improved through applicatia of a gas-turbine C (- ginerator to augment the offsite electrical g.id. The following concepts were considered for additional onsite power supplies. 19P.4.9.1 Steam Driven Turbine Generator A steam driven turbine generator could be installed which uses reactor steam and exhausts to the suppression pool. The system would be conceptuallysimilar to the RCIC system with the generator connected to the offsite power grid. The benefit of this item would be similar to the addition of another gas turbine generator, but would be somewhat less due to the relative unreliability of the steam turbine compared with a diesel generator and its unavailability after the RPV is depressurized. Ifit were sized large enough,it could have the adantage of providing power to additional equipment. If the system has a '0% availability for all events, the benefit is similar to an 80% reduction it, the diese,enerator common mode failure rate. Evaluation of the FRA l indicatet that the resulting beneSt is about 5.2E-04 person-Sievert. 19P.4.9.2 Alternate Pump Power Source The ABWR provides separate diesel driven power supplies to the HPCF and LPCF pumps. Offsite power supplies the feedwater purnps. This modification would proside a small dedicated power sourca such as a dedicated diesel or gas turbine for the feedwater, or condensate pumps so that they do not rely on offsite power. Evntustion of Potential Moo.fications to the ABWR Cesign - A*v.r.dment 34 19F 13

23A6:00Rev 4 ABWR stuhrd Scisty Anlysis Report The benent would beless dependence onlow pressure 9 stems during loss of offsite power events and station blackota evenu.lf the feedwater mirm were made to be 90% _',3 available duringloss of offsitt power events and stadon bbawls, the benefit would be similar to adding an additional RCIC svstem (Subsection 19PA2.1). The resulting l benefit would be about 6.9E-04 person-Sievert. 19P.4.10 DC Power Supplies The ABWR contains 4 DC divisions with sufficient capacity to sustain 8 hours of station ,) blackout (with some load shedding). This represents an improvement over current operating plant design:. 19P.4.10.1 Dedicatso DC Power Supply This item addresses the use of a diverse DC power system such as an additional battery or fuel cell for the purpose of previding motive power to certa'n components. Conceptually a fuel cell or sepa ate battery could be used to power a DC motor / pump combination and provide high pressure RPV injection and containment cooling. With pwper startin;; controls such a system could be sized to provide several days capability. Providing a separate DC powered high pressure injection capability has a benefit of ) iurther reducing the station blackout and loss of offsite power event risks which represent about 75% of the total CDF, but only a small fraction oDhe offsite risk. If the effective unavailability of the RCIC is reduced by a facter of 10 cae to the availability of a diverse system, one benefit we.ild be similar to adding a power supply for feedwater ] (Subsection 19P.4.9.2) and the benefit would be about ti.9E-04 person-Sievert. 19P.4.11 ATWS Capability The current ABWR design provides improvements in containment heat removal and l detection of ATWS events to limit the impact of this class of events. The PRA indicates that ATWS e>ents contribute about 0.1% of the core damage frequency (Table 19P-2) and about l'.% of the offsite risk (Case 9). 19P.4.11.'l ATWS Sized Vent This modi 5 cation would be available to remove reactor heat from A'IWS events in addition to severe s.cridents and Class 11 events. It would be similar to the containment rupture disk (which is currently sized to pass reactor power consistent with that generated during RCIC iniection), but it would be of the larger size iequired to pass the additional steam associated with LPCF imiection.The system would need to be manually initiated. The benefit of this venting conceptis to p. event core damage and to reduce the source term available for release following ATWS events. The evaluation shows that an ATWS j sized vent manually initiated with a 100% reliability would have a maximum benefit of Evnivation of Potential MJdifications to the ABWR Design - Amendment 34 19P.14

2sAs100n n 4

=

AGWR ssenderdsevery Aner sis soport r l reducing the offsite dose by about 3.0E 04 person-Sievert by reassigning the consequences from case 9 to case 1. I 19P.4.12 Seismic Capability The current ABWR is designed for a Safe Shutdown Earthquake of 0.Sg acceleration. The seismic margins analysis (Appendix 191) addresses the margins associated with the seismic design and concludes that there is a 95% confidence that existing equipment has less than a 5% probability of failure at twice the SSE level. This capability is considered adequate for the ABWR design and no additional changes are considered. l 19P.4.13 System Simplification i This item is intended to address system simplification by the elimination of unnece sary l interlocks, automatic initiation of manual actions or redundancy as a means to redt'ce overall plant risk. Elimination of seismic and pipe whip restraints is included in the concept. 1-While there are several examples of redundant systems, valves and features en the ASWR design which could conceinbly be simplified, there are several areas in which _[ the ABWR design already has been improved and simplified, especially in the area of A cc-ntrols and logic. System interactions during accidents were included in this category. One area was identi6ed in which simple modification of an existing system could provide some benefit. 19P.4.13.1 Reactor Building Sprays This concept would use the fitewater sprays in the reactor building to mitigate releases of fission products into 'he reactor building following an accident. The concept would require additional valving and nozzles, separate from the fire protection fusible links, to spray in areas vulnerable to release, such as near the containment overpressure relief line routing. The benefit of this modification could 1.4 to reduce the impact of events which do not involve the operation of the containment rupture disk. Such events release fission products from the containment into the reactor building. Releases from normal containment leakage and cases 3,6,7,4 and case 9 sequences could potentially be reduced. lf10% cf these releases from these cases were arbitrarily mitigated by this l method, the e e it would be about 1.7E-04 person-Sievert. 19P.4.14 Core Retention Devices Core retention features are incorporated into the ABWR Design. As discussed in Subsection 19E.2.2(FS),if a severe accident has resulted in a loss of RPV integrity, accident management guidance specifies that drywell sprays be initiated which will cause the suppression pool to overflow into the lower drywell after a few hours and Evaluatior, of Potentist Modi 5 cations to the ABWR Design - Amendment 34 19P 15

21A6100 R:v. 4 ABWR sundwsetery Artysis aeron T quench the debras bed After the molten core has been quent bed, no further ablation of concrete is expected and the decay heat can be removed by normal containment c'ooling methods such as suppie Tion pool cooling. If sprays can not be initiated, the Lower Drywell Flooder System described in Subsection 9 5.12 tools a debris bed by flooding over the molten core in the lower drywell with wun from the suppression pool. TMs system is similar to the

Post Accident Flooding' concept included in Referenc:: 19P-4. One f.dditional concept from Reference 19P-4 is included.

19P.4.14.1 Flooded Rubble Bed This concept consists of a bed of refactory pebbles which fill che lower drywell casity and are flooded with water. The bed impedes the flow of molten corium and increases the available heat transfer area which enhances debris coolability.The use of thoria (Th0 ) 2 pellets in a multiple layer geometry has been shown to stop melt penetration; thus, preventing core <oncrete interaction. Drawbacks to using thorium dioxide include cost, toxici:y, and the radiological impact of radon gas release into the lower drywell via the radioactive decay of thorium. Other refractories such as alumiaa slow :orium l penetution but may fail to stop core-concrete contact. Other refractories may be susceptible to :hemical attack by the corium and ' ay melt at lower temperatures. Pebbles composed of refractories other than th also may be susceptible to floating ) l because they have lower density than the corium. A major drawback common to all flooded rubble bed core retention systems is the need for further exparimental testing in order to validate the concept in BWR applications. I The benefit of this modification lies in the potential elimination of core-concrete interaction and a corresponding decrea:e in non-condensable gas generation. 9EC to Appendix 19E indicates a 90% certainty that debris on a concrete floor covered with wa*.er will be coolable in the current ABWR design. Only sequences in which no liquid injection to the drywell occurs will result in core-concrete interaction. A conservative estimate of the benefit of this concept over the existing desigt vould be elimination of sequences with core-concrete interaction except those with containment cooling failure. A resiew of Subsection 19E.2 indicates that this would effect about 1% of Cases 1,6 and 7. This corresponds to about 0.10E-04 person-Sievert averted. 19P.5 Costimpacts of Potential Modifications As discussed in Subsection 19P.I.S.1, rough order of magnitude costs were assigned to each modification based on the costs of systems detennined by GE. These costs represent the incremental costs that would be incurred in a new plant rather than ecsts that would apply on a backfit basis. Credit for the onsite costs averted by the I modification are discussed in Subsection 19P.1.3.2. For each modification which reduces the core damage frequency an estimate of the impact was raade and then i Evaluation of Potential Modificatl>ns tu the ABWR Design - Amendment 34 19P 16

i 4 1 nAssoo Msv. 4 ACWR s ndantseroryAner sisner:rt r applied to the potential averted offsite cost. This subsection summarize 3 the cost basis for each of the modification ev luated in Subsection 19P.4. This basis is generally the i cost estimate less the credit for onsite averted costs. Table 19P-6 summarizes the results. The cosu were biased on the low side, but all known or reasonably expected costs were i-accounted for in order that a reasonable assenment of the minimum cost would be ) obtained. Actual plant costs are expected to be higher than indicated in this evaluation. f All costs are refer-nced to 1991 U.S. dollars based on changes in the Consumer Price Index. I 19P.5.1 Accident Mana9ement i l 19P.5.1.1 Severe Accident EPGs/AMGs The cost of extending the EPGs would be largely a one-time cost which should be j prorated over several plants if accomplished by the BWROG. Current industry acchity 1. j. addressing this as part of Accident Management Guidelines (AMG). If plant specific, j; synptom based, severe accident emergency procedttres were to be prepared based on j AMGs, the cost would be at least $600,000 for plant specific modifications to EOPs.

[

19P.S.1.2 Computer Alded instrumentation , \\ Additional software and development costs associated with modifying existing Safety l Plant Display Systems are estimated to cost at least $600,000 for a new plant. This estimate is based on assumed additions ofisolation devices to transrtit data to the computer and in-plant wiring. Because this modification reduces the frequency of core dange events, a present worth of $400 onsite cc,sts are averted and the cost basis is - $599,6v0. 19P.5.1.3 Improved Maintenance Procedures / Manuals f The cost of at least $300,000 would be required to identify components which should i receive enhanced maintenance attention and to prepare the additional detailed [ procedures or recommended information beyond that currently planned. Credit for E reduction in onsite costs reduces the cost basis to $299,000. 19P.5.2 Decay Heat Removal i= l 19P.S.2.1 Passive High Pressure System [ The cost of an additional high pressure system for core cooling would be extensive since it would not only require additional sptem hardware which would cost at least i $1,200,000, but it would also require additional building costs for space available for the j system. Assuming the sptem could be located in the reactor building withoutincreasing its height, building costs are estimated to be another $550,000. The credit for averted j onsite costs is about $6,000 which brings the cost basis to $1,744,000. ~ i Evolustion of Potentist Modifications to the ABWR Design-- Amendment.14 19'*-17 i i i 4

23A6100 Rev. 2 ACWR steaant setery Aurys's nepors ) 19P.5.2.2 Improved Depressurization The cost of the additionallogic changes, pneumatic supplies, piping and quali6 cation was estimated for the GESSAR design (Reference 19P-1). A similar cost would be expected for the ABWR design. The cost is estimated to be at least $600,000 for an improved system for depressurization. This estimate assumes no building space increase for the added equipment. The credit for averted onsite costs was evaluated to be $1,400 which makes the cost basis $598,600. 19P.5.2.3 Suppression Pool Jockey Pump The cost of an additional small pump and associated pipirg is estimated at more than $60,000 including installation of the equipment. It is assumed that increases in nower supply capacity ad building space are not required. Controls and associated wiring could cost an additional $60,000 for a total cost of at least $120,000. A credit of $200 for I averted onsite costs makes the cost basis $119,800. 19P.S.2.4 Safety Related Condensate Storage Tank l Es$ mating the cost of upg ading the CST structure to withstand seismic events requires a detailed structural analysis and resultant material. It isjudged that the final cost increase would be in excess of $1,000,000. No credit for onsite cost averted was assumed ) for this modification. 19P.S.3 Containment Capability 19P.5.3.1 Larger Volume Cc.itainment Doubling the containment volume requires an increase in the concrete and rebar. If 2 2 l structural costs of the containment can be made for $12,900/m ($1,200/ft ), doubling the containment volume without increasing its height, the cost would be at least $8,000,000. This estimate does not include reanalysis and other documentation costs. Since this modification is mitigative, no credit for onsite averted costs was assumed. 19P.S.3.2 increased Containment Pressure Capacity The cost of a stronger containment design would be similar in magnitude to increasing its size (Subsection 19P.5.3.1). If the costs are primarily due to denser rebar required during installation and additional analysis, an estimate of at least $12,000,000 could be required. Since this modiScation is mitigative, no credit for onsite averted costs was assumed. 19P.5.3.3 Improved Vacuum Breakers The cost of redundant vacuum breakers including installation and hardware is estimated at more than $10,000 per line. Instrumentation associated with this modification is not included. For the eight lines the cost of this modification is more 19P.18 Evaluation of Potential Modifications to the ABWR Design - Amendment 32

22Astoo nov, s ACWR sanndantsareryAnalysis aerort than $100,000. Since this modiScation is mitigative, no credit for onsite averted costs was assumed. 19P.5.3.4 Improved Bottom Penetration Design The cost increue of using a stainless or inconel transition piece as opposed to carbon steel would be expected to be small in comparison to the engineering and documentation change costs assc ciated with the change. Costs, associated with external welds and suppon for the CRDs isjudged to be at least $1000 per drive. In addition, j about $500,000 of analysis would be required to develop the changes. This would i dominate the cost of this modiScation when applied to all 205 drives. Such changes are estimated to be atleast $750,000. Since this modification is mitigative, no credit for averted onsite costs applies, i 19P.S.4 Containment Heat Removal 19P.5.4.1 Larger Volume Suppression Pool This concept would result in similar costs as item Subsection 19P.5.3.1 for prosiding a ( - larger containment. An estimate of $8,000,000 is assigned to this item. 19P.S.5 Containment Atmosphere Mass Removal 5 19P.5.5.1 Low Flow Fittered Vent The cost of added equipment associated with the FILTRA system (excluding a test program) was estimated to be abo'.it $5,000,000 in Reference 19P-4. Although a de, tailed estimate was not prepared for the ABWR, an estimate of $3,000,000 has been assumed for the purpose of this evaluation. Since this modification is mitigative, no credit for averted onsite costs applies. 19P.5.6 Combustible Gas Control No additional modifications to the ABWR were identified in this group. 4 19P.5.7 Containment Spray Systems 19P.5.7.1 Drywell Head Flooding An additional line to flood the drywell head using existing Stewater piping would be a relatively inexpensive addition to the current symm. Instmmentation and controls to permit manual control from the control room would be needed. It !s estimated that the i total modification costwould be at least $100,000 for the engineering, piping, valves and cabling. haluation of Potential Modithetions to tr.2 ABWR Cesign - Amendment 31 19P.19

4 23AS100 R:v. 4 ABWR sandard setery Aantysis nep:rt [ \\ Because this modification is mitigative, no credit for aserted onsite costs has been applied. 19P.S.8 Prevention Concepts 19P.5.8.1 Additional Service Water Pump The use of diverse instrumentation would not presumably have a significant equipment cost, but there would be an increased cost of maintenance and spare parts due to less interchangeability and less standardization of procede es. These costs, however, are probably lovr in comparison with the extra support systems for air supply and service water. Equipment, power supplies and structural changes to include these new sptems are estimated to cost at least $6,000,00n. A small credit for averted onsite costs makes the cost basis for this item $5,999,000, based on the benefits discussed in Subsections 19P.4.1.3 and 19P.5.1.3. 19P.5.9 AC Power Supplies 19P.5.9.1 Steam Driven Turbine Generator The cost of the system should be similar to that for the RCIC sutem, but additional cost ) wt uld be needed for structural changes to the reactor building plus the generator and its controls. This item is expected to cost at least $6,000,000. With credit for averted onsite costs, the cost basis for this item becomes $5,994,300. 19P.S.9.2 Alternate Pump Power Source A typical feedwater pump for an ABWR sized plant could require a 4000 kWe sized generator, at $300 per kWe, a separate diesel generator and the supporting auxiliaries could cost at least $1,200,000. This cost would include wiring and installation of the alternate generator, but does not assume additional structural costs. With credit for averted onsite costs, the cost basis for this item becomes $1,194,000. 19P.S.10 DC Power Suppliec 19P.S.10.1 Dedicated DC Power Supply Fuel cells are largely a c'evelopmental technology, at least in the large size range required for this application in addition the process involves some risk of Sre.To address these concerns a cost of at least $6,000,000 would be expected. A separate battery would be less expensive than fuel cells, but would involve additional space requirements -hich could make this modiScation more expensive than adding a diesel generator as discussed in Subsection 19P.5.9.2. A battery bank capable of supplying 400 Kwe would be about 50 times larger in capacity than the emergency Evslustion of Potentist Modifications to the ABWR Design - Amendment 34 \\ 19P 20

22A6100 kv. 5 ACWR Standard Safety Analysis Report 'r'~ i L 2 2 batteries. This number of batteries would require at least 464 m (5,000 ft ) og,p,ce, 2 assuming extensive stacking and without concern for seismic response. At $5,382/m 2 ($500/ft ) construction cost, the additional space required would amount to $2,500,000 for this modification. Additional costs would be requ: red for DC pumps, cabling and instrumentation and controllers. A total cost would be at least $3,000,000. 19P.5.11 ATWS Capability 19P.5.11.1 A1WS Sized Vent Larger piping and additional training would be required to extend the existing rupture disk feature to be available during an ATWS event. Additionalinstrumentation and. cabling would be required to make the vent operable from the control room. It is estimated that the incremental cost would be at least $300,000. 19P.S.12 Seismic Capability l No modifications were considered for this group. 19P.5.13 System Simplification 19P.S.13.1 Reactor Building Sprays The cost of this modification isjudged to be similar to the concept of drywell head s flooding (Subsecdon 19P.5.5.1) ifit only involves piping and valves which are tied into the firewater system. An estimate of $100,000 has been assigned to this item. Onsite cleanup costs also could be affected by this modification. If the cleanup costs were eliminated an averted cost would conservatively be about $5,000. 19P.S.14 Core Retention Devices 19P.S.14.1 Flooded Rubble Bed Reference 19P-4 estimated that the refractory material needed for this modification would cost approximately $2,203/kg ($1,000/lb). If the lower drywell were filled with about 0.458 m (1.5 ft) of this material, which would remain well below the senice 3 3 platform, at least 35.4 m (1250 ft ) of material would be required. Ifit weighs 3 3 2422 kg/m (19 lb/ft ), the material cost alone would amount to $18,750,000. 19P.6 Evaluation of Potential Modifications A ranking of the modifications by $/ person-Sievert avened is shown in Table 19P-7 based on the results and estimates provided in Subsections 19P.4 and 19P.5. [ The lowest cost / person Sieven averted modification is more than 1600 times the target n ( l criteria of $100,000 per person Sievert averted. Clearly none of the modifications is justifiable on the basis of costs for person-Sievert averted. This can be attributed to the Evatustion of Potentist Modifications to the ABWR Design - Amenoment 35 19P 21

21A6100 Rev. 5 ACWR standard sarety Anstysis neport ) low probability of core damage in the ABWR with the modifications to reduce risk already installed. 19P,7 Summary of Conclusions Potentially attractive rrodifications were identified from previous enluations of potential prevention and mitigation concepts applicable during severe accidents and discussion with the NRC staff. Potential modifications were reviewed to select those which are applicable to the ABWR design and which have not already been implemented in the design. Of these modifications, twenty one were selected for l additional review. The low level of risk in the ABWR is demonstrated by the total 60 year offsite exposure risk of 26.9E-04 person-Sievert. At this level only modifications which cost less than $269 l can bejuvified. Based on this low level no modifications arejustified for the ABWR. Based on the PRA results, none of the modifications pro.ided a substantial improvement in plant safety. 19P,8 References 19P1 Evaluation ofProposed biodifications to the GESSAR L1 Design, NEDE S0640, Class ) III, June 1981. 19P 2 Supplement to the FinalEnvironmentalStatement -Iimsick Generating Station, Units 1 and 2, NUREG-0974 Supplement, August 16,1989 19P-S Issuance ofSupplement to the FinalEnvinmmental Statement-Couanche Peak Steam Electric Station, Units 1 and 2, NUREG 0775 Supplement, December 15,1989 19P-4 Survey of the State of the Art in biitigation Systems, NUREG/CR-3908, R&D Associates, December 1985 19P-5 Assessment of Severe Acadent Preventwr and biitigation Fea*ures, NUREG/CR-4920, Brookh.ven National Laboratory, July 1988. 19P-6 Design und Feasibility ofAcadent Afitigation Systemsfor Light WaterReacon, NUREG/CR-4025, R&D Associates, August 1985 199 7 Seve;- mdent Risks: An AssessmentforFive US Nuclear Power Plants, NUREG 1150,,; iary 1991. 19P-8 Technical Guidancefor Siting Critena bevelopment, NUREG/CR 2239, Sandia National Laboratories, December 1982. i 19P 22 Evaluation of Poten:ial Modifications to the ABWR Design - Amendment 35 3

e 23A8 f 00 Mev. 6 ADWR sandsid safety Analysbneport Table 19P 1 Offsite Accident Cases Person-Sievert Frequency Exposure'(per Contribution Case ~ (per yr) event) (per 60 yr) (%) l Case 1 2.1E-00 -138 1.7E-4 6.3 l Case 2 7.8E 1

S3.28 3.9E-7 0.01 l

Came 3 0 3714 0 0.00 l Case 4 0 2064 0 0.00 Case 5 7.5E-12 933.80 4.2E-7 0.02 [ Case 6 3.1E-12 24,160 4.5E-6 0.17 l Case 7 3.SE-10 27,260 C.4E-4 23.8 l Case 8 4.1E 10 32,020 7.9E-4 29.3 l Case 9 1.7E-10 33,120 3.4E-4 12.6 l NCL 1.3E-07 96 7.5E-4 27.6 Total 1.6E-07 26.9E-4 100.0 For case descriptions see Table 19E.34; frequencies are based on Table 19P-2. t Average of regionalvalues used;see Subsection 19E.3. L a Eve'ueCon cf Pctentist Modmcations to the ABWR Design - Amendment 35 19P 23 t

21A6100 Rev.1 ACWR senassed scroly Aner sis neport r Table 19P-2 Core Damage Frequency Contributors' Event Sequence init. Evem 1A 1B1 182 1B3 1D 11 lilD IV Total Contrib Screm 1.1E-08 4.3E 10 9.5E-13 1.1E 08 7.3 Turb Trip 6.8E-09 2.7E-10 3.7E-11 7.1E-09 4.5 Isolation 1.8E-08 7.1E 10 1.1E 11 1.9E-08 11.9 LOOP 2 4.1 E -09 1.5E 11 4.2E 13 4.1E-09 2.6 LOOP 8 2.4E-09 9.6E 12 1.4E 12 2.4E-09 1.5 LOOP 8+ 5.8E-10 1.1E-09 6.0E 11 1.7E-09 1.1 SB02 6.6E 12 6.7E 08 6.7E-08 .9 SBOS 2.6E-08 2.6E-08 1E.7 SBO8+ 1.5E 08 8.9E 10 1.65-08 10.3 IORV 1.1E-09 2.0E 10 9.5E 13 1.3E 09 0.8 SBLOCA 2.5E-10 2.5E-10 0.2 ATWS 1.5E 10 1.5E-10 0.1 ) TOTAL 4.4E-08 2.6E 08 1.5E 08 8.9E 10 7.0E-08 1.1E-10 2.5E-10 1.5E-10 1.57E-07 100 ~Offsite Release Group" LCHP SBRC LCu* LHRC LBLC ATWS Total Case Case 1 3.4E-09 7.9E-10 1.6E-08 5.1E-11 2.0E-08 Case 2 7.8E 11 7.8E 11 Case 3 1.3E-12 1.3E 12 Case 4 0 Case 5 6.3E-12 6.3E-12 Case 6 1.2E 10 1.2E-10 Case 7 1.1E-10 2.6E 10 3.70E-10 Case 8 2.1E 10 2.1E 10 Case 9 1.1E-12 1.5E 10 1.5E-10 NCL (N) 4.0E-08 1.5E 08 8.0E-08 2.0E-10 1.4E-07 Total 4.4E-08 1.6E-08 9.6E 08 1.1E.12 2.5E 10 1.5E 10 1.57E-07 Contrib % 28.1 10.3 61.4 0.122 0.2 0.1 100 Derived from PRA event trees (1904 and 19D5); some numerical differences exist due to simplification. " For description see Subsection 19E.2.2 19P-24 Evaluation of Potential Modincations to the ABWR Design.\\mendment 31

23A6100 Rsv.1 ABWR standard sarery Anssysis neport Table 19P 3 Modifications Considered Basis (SSAR Reference Modification Category Reference) (19P.X) 1. Accident Management a. Severe Accident EPGs/AMGs 2 1 b. Compuier Aided instrumentation 2 1 c. Improved Maintenance procedures / Manuals 2 1 d. Preventive Maintenance Features 4 1c 1 e. Improved Accident Mgt instrumentation 4 1b 1 f. Remote Shutdown Station 1 (7.4.2.2) 1 g. Security System 1 (13.6.3) 1 h. Sh ulator Training for Severe Accidents 4 1b 1 2. Reactor Decav Heat Removat a. Passive High Pressure System 2 1 b. Improved Depressuritation 2 1,2,3,5 ( c. Suppression Poo'l Jockey Pump 2 1 d. Improved High Pressure Systems 1 (6.3.2.2.1) 1 I c. Additional Active High Pressure System 1 (6.3.2.1.1) 1,5 f. Improved Low Pressure System (Firepump) 1 (5.4.7.1.1.10) 1,2,3 g. Dedicated Suppression Pool Cooling 1 (3.2.2) 1,2 h. Safety Related Condensate Storage Tank 2 1 1. Extended Station Blackout injection 4 10e 1 l. Improved Recirculation Mode 3 PWR 3 3. Conto'.nment Capability a. Larger Volume Containment 2 1,4 b. Increased Conteinment Pressure Capacity 2 1 c. Improved Vacuum Breakers 2 1 d. Increased Temperature Marga for Seals 1 (19F.3.2.2) 1,4 e. Improved Leak Detection 1 (7.2.2.2) 3 f. Suppression Pool Scrubbing 1 (19E.2.13) 5 g. Improved Bottam Penetration Design 2 4 Evstustion of Potentist Modifications to the ABWR Ossign - Amendment 31 1SP.25

23A8700A;v.1 AGWR studerds:rei Azer sis nepors r l Table 19P 3 Modifications Considered (Continued) Basis (SSAR Reference ModWication Category Reference) (19P X) 4. Containment Heat Removal a. Larger Volume Suppression Pool 2 1 b. CUW Decay Heat Removal 1 (5.4.8) 2 c. High Flow Suppression Pool Cooling 1 (6.2.2) 1,4 d. Passive Overpressure relief 1 (6.2.5.2.6) 1,2,5 5. Containment Atmosphere Mass Removal a. High Flow Un'iltered Vent 3 Mark Ill 1,4 b. High Flow Filtered Vent 3 Mark 111 1,4 c. Low Flow (filtered) Vent 2 1,2,3,4 d. Low Flow Vent (unfiltered) 1 (6.2.5.2.6) 1,2,3,4 6. Combustible

ass Control a.

Post Accident inerting System 3 Inerted 1,4 b, Hydrogen Control by Venting 3 Inerted 1,4 ) c. Preinerting 1 Inerted 1,4 l d. Ignition Systems 3 Inerted 1,3,4,6 e. Fire Suppression System inerting 3 Inerted 1,4 7. Containment Spray Systems a. Drywell Head Flooding 2 2 b. Containment Spray Augmentation 1 (5.4.7.1.1.10) 1,2,3,6 8. Prevention Concepts a. Additional Service Water Pump 2 3 b. Improved Operating Response 1 (15.0) 1 c. Diverse Injection System 4 2a 1 d. Operating Experience Feedback 1 1 e. Imptoved MStV/SRV Design 1 (5.4.5,5.4.13) 1 I ) 1 19P-26 Evstantion of Potentis! Modifications to the ABWR Design - Amendmunt 31 7 L

e ', a 23A6100 Rzv.1 AGWR standardsarery Anotysis neport 1 Table 19P 3 Modifications Considered (Continued) Basis (SSAR Reference l Modification Category Reference) (19P X) 9. AC Power Supplies a. Steam Driven Turbine Ge, Jrator 2 1 b. Alternate Pump Power Source 2 1 c. Deleted d. Additional Diesel Generator 1 (P.3.1) 1,3 e. Increased Electrical Divisions 1 (8.3.1) 1 f. Irnproved Uninterruptable Power Supplies 1 (8.3.1) 1 g. AC Bus Cross ties 1 (8.3.1) 1 h. Gas Turbine 1 (9.5.11) 1 1. Dedicated RHR (bunkered) Power Supply 4 2g 1

10. DC Power Supplies a.

Dedicated DC Power Supply 2 1 ( b. Additional Eatteries/ Divisions 4 10e 1 c. Fuet Cells 4 10e 1 d. DC Cross-ties 1 (8.3.2) 1 e. Extended Station Blackout Provisions 1 (19E.2.1.2.2) 1,3

11. ATWS Capability a.

ATWS Sized Vent 2 2,4,6 b. Improved ATWS Capability 1 (19.7.2(2),(4)) 1,5,6

12. Seismic Capability a.

increased Seismic Margins 1 (190 8 b. Integral Basemat 3 Mark lli 1

13. System Simplification a.

Reactor ltuilding Sprays 2 2 b. System Simpliilcation 1 (Table 1.3-2) 1 c. Reduction in Reactor Bldg Flooding (19.7.3(4)) 5

14. Core Retention Devices a.

Flooded Rubble Bed 2 1,2,4,6 b. Reactor Cavity Flooder 1 (9.5.12) 3 c. Basaltic Cements 1 (19.7.3(4)) 1,4 Evaluation of Potential Mod &cstions to the ABWR Design - Amendment 31 19P 27

23A6100 Mov.1 ACWR stezdard surety A:stysh ;;sp:rt Table 19P 3 Modifications Considered (Continued) Basis 8 (SSAR Reference Modification Category Reference) (19P X) 4. Containinent Heat Removal a. Larger Volume Suppression Pool 2 1 b. CUW Decay Heat Removal 1 (5.4.8) 2 c. High Flow Suppression Poci Cooling 1 (6.2.2) 1,4 d. Passive Overpressure relief 1 (6.2.5.2.6) 1,2,5 5. Containment Atmosphere Mass Ron. oval a. High Flow Unfiltered Vent 3 Mark 111 1,4 b. High Flow Filtered Vent 3 Mark ll1 1,4 c. Low Flow (filtered) Vent 2 1,2.3.4 d. Low Flow Vent (unfiltered) 1 (6.2.5.2.6) 1,2,3,4 6. Combustible Gas Control a. Post AccidentIrierting System 3 Inerted 1,4 ) b. H,edrogen Control by Venting 3 Inerted 1,4 c. Preiserting 1 Inerted 1,4 d. Igniiltn Systems 3 Inerted 1,3,4,6 e. Fire Suporession System inerting 3 Inerted 1,4 7. Containment Spray Systems a. Drywell Head Fiooding 2 2 b. Containment Spiay Augmentation 1 (5.4.7.1.1.10) 1,2,3,6 8. Prevention Cont.epts s. Additional Sers ice Water Pump 2 3 h. Improved Operating Response 1 (15.0) 1 c. Divme injection System 4 2a 1 d. Operating Experience Feedback 1 1 e. Imptoved MStV/SRV Design 1 (5.4.5,5.4.13) 1 l 19P-20 Evstustion of Petentis! Modificataons to the ABWR Design - Amendment 31 l . _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _}}

ML20202H898

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