ML18038B421

Forwards Minutes of Browns Ferry,Unit 3 Restart Panel 950418 Meeting

ML18038B421
Person / Time
Site:	Browns Ferry
Issue date:	04/25/1995
From:	Jerrica Johnson NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To:	Ebneter S NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
Shared Package
ML18038B417	List: ML18038B416 ML18038B418 ML18038B419 ML18038B420 ML18038B421 ML18038B422 ML18038B423
References
NUDOCS 9509190194
Download: ML18038B421 (62)

Text

UNITEO STATES NUCL~MR REGULATORY COMMISSlON REGION II 101 MARIETTASTREET, N.W., SUITE 2900 ATLANTA,GEORGIA 30323-0199 April 25, 1995 MEMORANDUM TO:

FROM:

SUBJECT:

Stewart D. Ebneter, Regional Administrator Jon R. Johnson, Deputy Director Division of Reactor Projects MINUTES OF BROWNS FERRY 3

RESTART PANEL MEETING APRIL 18, 1995 The Browns Ferry Unit 3 Restart Panel met in the Region II offices on April 18, 1995, to review the status of NRC and TVA activities for the restart of this unit.

The next meeting of the NRC panel will be held in the Resident Inspector's offices at the Browne Ferry Nuclear Plant on May~ l995, at 9:00 a.m.,

CDT and the fol,lowup meeting with the licensee will be in the Administrative Building at Browns Ferry from 12:30 p.m., to 2:00 p.m.,

CDT.

Meeting minutes are attached as Enclosure 1.

A Unit 3 Task Checklist is provided as Appendix A and a Unit 3 Issues Checklist is provided as Appendix B.

Appendix C is an executive Summary of the Browns Ferry Multi-Unit PRA.

Attachment:

Browns Ferry Unit 3 Restart Panel Meeting Minutes w/Appendix A, B, and C

cc w/att:

L. A. Reyes, RII E.

W. Merschoff, RII J.

R. Johnson, RII A. F. Gibson, RII J.

P. Stohr, RII M. E. Cline, RII T. A. Peebles, RII C. A. Casto, RII J. J. Blake, RII M. B; Shymlock, RII T. R. Decker, RII M. H. Rankin, RII 9509f90l94 950822 PDR ADOCK 05000296 P

PDR Attachment 4

1 C

cc w/atts:

(Continued)

P. J.Kellogg, RII H. S.

Lesser, /II B. Uryc, RII J.

W. York, RII L. D. Wert, RII K. P. Barr,. RII R.

P.

Zimmerman, NRR S.

A. Varga, NRR G. H. Tracy, EDO F. J.

Hebdon, NRR J.

F. Williams, NRR P.

S. Koltay, NRR

BROWNS FERRY UNIT 3 RESTART PANEL MEETING MINUTES APRIL 18, 1995 Meeting Date:

Meeting Location:

Members Present:

Summary:

April 18, 1995 Region II Office J.

R. Johnson,

Chairman, RII M. S. Lesser, RII C. A. Casto, RII W.

E. Cline, RII K. P. Barr, RII J.

F. Williams, NRR L. D. Wert, SRI The chairman reviewed the minutes of the previous meeting and the status and results of the assigned action items.

The task checklist and selected items from the issues check list were discussed and updated checklists are provided as Appendix A and B.

Appendix C

is.. a copy of the executive summary for the Browns Ferry Multi-Unit PRA.

The panel chairman announced that K. Barr will be replacing W. Cline on the panel.

Previously assigned actions:

NRR (Hebdon)

Schedule ORAT.

2.

3.

(Closed)

Peter Koltay has been identified as the ORAT team leader.

The inspection has been entered in the HIP for the end of October.

J.

Williams will invite P. Koltay to the next panel meeting.

DRP (Lesser)

Arrange for a special (separate) meeting for public comment on the restart of BF3 (September).

(Open)

No action taken on this item yet.

DRP (Wert) Give a status of the number of area turnovers.

(Closed)

A list of the nuwlber of area turnovers was given.

However, it was pointed out that at Browns Ferry, the licensee does not use this information for engineering or technical purposes (Watts Bar does) but uses the information for housekeeping purposes.

The weekly status report will track the number of housekeeping areas turned over.

Attachment

~'

DRP (Lesser)

Schedule Ron Gibbs to perform Module 38703 for replacement components and parts for BF3.

(Open)

Arranged with R. Gibbs and management to perform this inspection and will place this on the HIPS for the week of July 10, 1995.

DRP (Lesser)

Copy of each of our meeting minutes to Peter Koltay to aid him in scheduling the ORAT inspection.

(Closed)

York put P. Koltay on the panel distribution list.

DRP (Johnson)

Arrange for a discussion of two unit operation with the licensee.

(Open)

The licensee will be requested to discuss this at the next meeting.

DRP (Lesser)

Add CATO closeout letter to issues checklist.

(Closed)

We have added this to the Issues Checklist.

NRR (Hebdon)

Discuss with NRR reviewer the possibility for finishing his review of licensee's Appendix R submittal sooner so that inspection can be performed sooner than July.

(Closed)

The NRR reviewer can't finish the review early and Casto will add this team inspection to the inspection schedule for early July.

DRP (Lesser/Wert)

Compare TIs 2512/015 and 2515/074 (employee concerns) to ensure that all applicable points for both are covered for BF3.

(Closed) TI 2515/074 was used to inspect employee concerns for both the Sequoyah and the Browns Ferry 2 star tups and a number of inspections have been completed for the Browns Ferry 3 startup (with no apparent problems revealed).

An inspection of seismic CATDs is to be scheduled by Casto using R. Chou.

The review of TI 2512/015 revealed that it was only applicable to Watts Bar DRSS (Barr/Decker)

Schedule George Kuzo (because of his familiarity with the Watts Bat problems) to inspect or help Dan Jones with PASS or line sampl ing.

(Open) This action is still being formulated, however the scope should include readiness of radiological instrumentation.

DRP (Lesser)

Add four items from NPP identified by the SRI to the list.

Also add Beta tape problea.

(Closed)

The items have been added.

12.

DRS (Casto)

Discuss the restart test program next panel meeting.

(Open)

Casto will assign someone to look into but cannot do this until first two weeks in June since the licensee is behind on procedures.

13.

DRS (Casto/Peebles)

Add inspections/

dates for inspection in EOPs, procedures, maintenance, TSs, etc. to the Master Inspection Plan.

(Open) Still have to compile a list from the Operations Branch.

14.

DRS (Casto)

Check with Lenahan on a

NRR Memo which apparently approved CONAN computer code and may provide information to close IFI 94-12-01.

(Closed) Still have to close IFI, but the inspectors have the necessary information aand IFI is on the Issues Checklist.

Newly Assigned Action Items:

1.

DRP (Lesser/Uryc)

Discuss the status of DOL cases at Browns Ferry.

2.

DRS (J. York) Distribute the executive suamary for the Browns Ferry Multi-Unit PRA.

3.

DRP (York/Turner) Determine problems with identifying inspections on the MIP.

Review and ensure MIP is updated to reflect planned inspections.

These actions shall be completed by the next oversight meeting on May 24, 1995.

The following item was completed on the Task Checklist:

NRC/Licensee Agreement on Restart Issues-The licensee agreed with the Issues Checklist and provided a status report of all items during the April 19, 1995 meeting.

~

~

Dete Printed: Apri 25. 1995 TASK Appendix A

BROWNS FERRY 3

TASK CHECKLIST RESP.

DATE STATUS

'Establ-ish Restart Panel

.Develo

':;Case -S'cific Checkl ist.

'Oevelo

-Restart Action Pl'an RII NRR RI'I, NRR" RII

NRR, 9 22 94' lete Co lete 2'

95

'C'ete Re ional. A'dministrator'. A roves Plan NRR Associate. Director A roves-Plan NRR 2195

• 2 I 95 Co lete C

1'ete

'No'tification" Restart Panel establ-i shed; RON 509:

Licensee:=performs root-cause analysis

and develo s corrective action. lan Lesser Licensee 2/24/95

,Cbmpl ete;;

7/10/91 Compl'ete

NRC,':evaluates:"l,icensee's root'ause

-'determination and corrective action

. lan NRR

.4/I/92

Compl'ete-

'Review licensee generated restart issues Panel 3'/21/95 Compl'ete

.Independent-'NRC identification of

,:.restart:;,items

('onsider,'external

'."sources

:NRC/Licensee,'.agreement',;.on;restart

'ssues".

Obtain ublic comnents.

ress conf Obtain coments from State and Local Officials Obtain coments from applicable Federal a encies Evaluate licensee's readiness self assessment Conduct Operational Readiness Assessment Team Ins ection ORAT Restart issues closed Issue au mented restart covera e

ROI Obtain staff comments on restart Re-review HC 0350 generic restart checklist

'Panel Y

wc 7

'.,Panel:."

Lesser Barr Barr RII Koltay Panel RII RII NRR Panel 3/21/95-4(19/95" 3 21 95 C'ompT'ete C'omjiletc

~

~

S

~

i

Prepare restart recomendation document and basis for restart to Regional Administrator Restart meetin with licensee Restart Panel recommends restart Regional Administrator concurs in restart recommendation SECY a er NRR Associate Director concurs in restart recommendation SECY a er EDO Concurs in restart recommendation SECY a er

'C

'ACRS br,'i'eA'ng-.

Submit Commission a er Commission briefin Commission restart authorization Notif Con ressional Affairs of restart Notif ACRS of restart Notif FEMA of restart Notif Public Affairs of restart Notif State and Locals of restart Monitor restart RII Panel Panel RII NRR NRR c,

'NRR'RR NRR, RII Comm NRR NRR

RII, NRR RII RII RII

,'Not.'Re ui"ied..

Appendix B

BROWNS FERRY 3 ISSUES CHECKLIST Date Printed: April 25, 1995 1$5%

DESCRIPT IM TNI ACTION ITENS CTI 2515 NRC LEAD Ik/SEN llCENSEE STATUS NRC ACTINI COUNTS I.D.2.2 I.D.2.3 11.8.3.2 II.8.3.3 II.~.3 4 jy.~@P~gNP~:g>

"', 'IAC:f66 %A'.FOTI'~+ii"-

SPDS Installed TAC N746121 &A F075 Cl 89-06 TAC N73636 F072 SPOS Fully laptaaented TAC N51225 NPA F009 Poet Accident Saoptfng-Correctlve Actions TAC N74613 KA F076 PASS - Procedures TAC N746141 NPA F077; N83122 PASS - Nodlflcationa TAC N44425g NPA F012 paffetjatj~.'::.",'Rmtm':.,a'nd

';RRvfie'ff2:",Coritrit'ppfkp4'-;.iAC,'ft40O8':,',;~SNIPÃ,:~f-

'"?>>>u>>?%Q? g".

'>p'eebtas,'+

? Q+?~

a?e..

r>.

SRI SRI Decker Decker Decker SRI"."' '

ck - "~:.>> ~>':

'-SE"'0i29?/91 93.201'=94'09 94-2f.'~" >'=

SE 2/5/92 IR 95.22 SE 2/5/92 sE 5/2r/8r 94-33 SE 5/27/87; TS aenend 6/21/94 SE 5/27/87 SE 5/23/88'5~j0'>)>>>>'j 1?p~ 4:": ~.

eiaentfelty

coiptete'ield coeptete field ccept ate; testfn 4/96 2/22/95 Design coaptete; 50K fapteeent 12/94 75X i tement 4/95 Cosptegel re+ for

'los'ur'e

'Inspection perforeed

"--/94 revfeMed pro'gim sit laficto Installation veriffed, open pending PNT SINS ready for ctoiure

~ v, V

II.E.4.2.1 4

.Pinktfit>fDIf?.".,".-!1'>>itat[N,

TA0?N44763@NPJL(f 01864>>. '-.

Contafraent Isolation Dependability - Dfverse I so let ion TAC N74615 NPA F078 SRI

?

A?? '?

'v SRI SE 12/22/81 95 10.'<-">'+'x SE 1/6/95 lk 95-16 Cosplatel ready for c t'osur'e TVA to provide coaple't'ion a'tatua 8/95 SlliS shous ready for ctosuri*

fnstallation verified, open pending PNT

DESCRIPTIM NRC LEAD IR/SER LICENSEE STATUS NRC ACTION CQIKNTS STATIIS Contafment Isolation 0ependabl lIty-Contatwent Purge Valves TAC N74616'079 Accident Nonltorlng-Proctdures TAC N74617 NPA F081 Accident Nonltorlng-Noble Gis Nonltor TAC N44905 NPA F020 Iodine/Particulate Nonltor TAC N44976 NPA F021 Containment Nigh Range Nonltor TAC N45047 NPA F022 Contafment Pressure Nonlter TAC N47584 NPA F023 Contafrmnt Mater Level Nonltor TAC N47655 NPA F024 Instruaentat(on for Dettctfon of Inadequate Core Cool fng GL 84-23 TAC N45118 NPA F026 HPCI/RCIC lnftlat lon Ltvtla TAC N45534 NPA F043 ADS Actuation Nodiffcatfon TAC N45682'PA F048 j"84SkI)~j%+M.~4;:".HidNgPd@ksg.

- ~

Qualification of ADS Acclscllatora TAC N48262; NPA F055 SRI Decker Decker Decker Rankin SRI SRI SRI SRI SRI

'RI SRI.

SE 1/6/95 95-16 SER 8/17/90 94-33 SER 12/22/81 SER 12/22/81 SER I/8/82 IR 94-28 95-11 SER 6/16/83 SER 6/16/83 SER 11/1b/86 95-16 SER 9/19/83 90.23 SER 5/29/90 SER 12/3/82 95.16" "'-

SER 7/24/85 Lfcensee tracking wfth individual fnstrunents 4/18/95 4/18/95 Nay 95 ffeld coapfete 7/95 field coopf ate 5/95 field cospfete 2/95 field cocpf ate 1/95 field coaplete 7 12/3/82 stited.

ticeniee'to trick under CRDR as ilternate method;.,

fIeld'c f et eI2/95'"..

field coaplete 4/95 installation verified, open pending PNT open pending review of procedures,

PHT, training Listed in NUREG 1435; not on other lists

~

~

8

I I.K.3.57 DESCRIPT IOI Identlfy Mater Sources Prior to Nanual Actuation of ADS NPA F062 NRC LEAD SRI SRI IR/SER SElt 8/30/82 90'-'37.

LICENSEE STATUS NRC ACTION SIHS shows ready for closure SINS shows reidy for closure CQOENTS closed for ell 3 units

>'TAIItS TEIB>i%ART IISlRlKTI~

TI 2500/020 AIMS BL 83-28 TAC N07931 HPA 0001 SRI SER 1/22/90 IR90.29 IR90-33 95-22 design cooptete; 50X Isptement 12/94 75X I lement 3/95 pending field cooptetion snd testing

's>~ W'~< P$

i.v'+>'s>'"S'<P'".":".g~'Piqg>.'>>o'+~".s'.

Tsj '4gim4" '.jp:iajil>oyae,'Cir'kiri>>>--

'RI,.

's

>s.

IR90'"31" 93:18 93 32c93"43 94-20 95.10

~ s s's-TI 25'lS/089 Tl 2515/095 Tr 251S/099 Tl 2515/109 TI 2515/111 Stress Corrosion Cracking in BMR Piping SN Reclrc Pusp Trip NN Power Oscillation IEB 88.07 TAC N72769 NPA X807 NOV Testing CL 89.10 EDSFI foltoaap Blake SRI Kellogg Casto Shymtock 95-22 SER 4/4/90 TS 179 5/31/94 94-03 field coaptete 6/95 6/2/95 SDX isplement 6/95 75X I lement 9/95 SIHS shows ready for closure TI for GL 84-11; GL M.01 superceded 84-11 and TI cancel led

-',~i"p)fI1ii,,"".";,:lEvst'tiiia>s fn'Env']rona

SRI, s

93-44 TI 2515/118 Service Mater System TAC N73972 MPA L91'7 Kellogg SER 4/23/90 Tl 2515/120 TI R515/121 Tl RS15/1RR Station Blackout TAC N68519:

HPA A022 Installation of Hardened Wtwetl Vent Gi. 89-16 Loss of Fill Oil for Roast Transmitter IEB 90.01 TAC H85363 MPAB122 A4 (.>pe+

., xN!;Fgp,%pal'.~.+>pe,>.'>>,e'>:)re)" $M

-,'-.Tl)2%151'l1IF);""~~ lMit'iF;;:Liwl',;tie't Er'rors

.:.LÃ84a4:~.$ 44;-,.s I+f9j.'jan+-'Ã':>.'.k g >."Q<'I" Lawyer '."

> >s'>>'(

s Shymtock SRI

~

Shymlock IR 93-16 SER 9/16/92 SER 8/16/91 s*.

>> 75X isplement c

lete 2/95 field coeptete 6/95 NRR to issue SER early 1995 TI 2515/128 Plant Hardware Mode to Rx Vessel Mater Level Inst.

Shymlock SER 4/20/94 IR 93.201

~

~

Cl

ISSUE DESCRIPT I I%

ItRC LEAD IR/SER LICENSEE STATUS ItRC ACTION CQIKNTS STATUS NRC NR.LETIRS

.Iii-

e",02 '.'."<'.

IEB 79-12

".IAI>"'II0(017 '%~'4i'"'

Short Period Scram at ONa Slake

'RI 81-18 field carptete 9/95

<<< Refer to I.srge Sore Piptng and Sorts Program Project plan

.>PA<>gyjpj<<<'g0( +gjgygy.'+~)$,,<<>> Pg<'<{(p><(Q<jPYP>

.s)4<"... g ';:SatMfc.:lnalya)sfRr{<A%

Slak'i 4

<< Refer to large Sore Piping and Supports Progrws p

>y'..<I~t>4~~wg< ',=:.AjtIi!3j;::Pnsm,:g:.:.

<<v%~'P>~>$

'Si'i r,".(~:-.

i<a>.,{>~>"".

'RP-23 TVA 6/22/93:,'gtoaid".

for. U2;- L)ce>nsielto silat> t:cloaure",.":a-.

te>'ttir'ot"ta'.-.-

TVA,)0/21/94 certified that IES Nas"previously' coepleted on all 3 tstita

~<:~

ASY+Q@~g~gg ) @48>{~>$8'4@SPpY<><@@+.";My. g<',~~~;,'. ~

g)Kf,'m<".;Qc

',XSP~R~t",,Cental:;+,:-P::::-':,;:

s>M;"~>I Q.:Y'~wcgj~~>j ~Qj:~'i< ~)5<':~g'P%+tTP<$I~"~<'<"lT.

~IEO:,,lS...:-.'.+8~ ",S)reaa,Q'irrrial'NII:Cricking

"<f>5-,".-':'a "',X~"'::in",6'-'-~S!I',Si. I j"'""."2e.

@Pvg<g<gjg~

8 lake',.':,';:

95;22 f)e(d coeplete";<. '""

>A

~ <g~

R83".;55"'IR56" Project-Plan 5,(I

<<~ Refer to GL 85.01

<<~>+. ga8P <>{t'PPv>{y. t> 8$@cg~<;;~g"'.'TQr:re@.", >>t>>ey~jt':

5!set;:,C frgttjt>wtri~ttfttI

.;.tN~ fr'1j,'aa f~".tt'I:Sttj~,

" '{>C'>>

SRI Y

95.22 Project Plan IES 84.02 Failures of GE HFA relays in 1E Safet a

tetm SRI Casto~z'.-".". 88'32'roject Plan

';. ~:Rafer 'to GL 89-10 IEB 86-02 IES 88 03 IES 88 04 Static 0 Rtng DP Suttches Inadequate Latch Engagetsent in ilFA relays by GE TAC 1173554I NPA X503 SR Ptjsp Lose TAC II69890; IS>A X507 Cssto NRR SRI 94-31 SER 8/2/90 SER 4/4/90 field ccmptete TVA 4/11/94 reported coaptation of requirements; inspectlore ccwptete for U3; no robtees found field cceptete 6/95 Project Plan NUREG 1435; not on licensee list; NRR check if issue is closed Project Plan; Tl 2515/087 issued but not required for SF r<<<"~s '~~.:..~m ';.,

<',%$ ><>%~~<++~'4+8{>4X'~~~

""<:>c<diXgh Kel togj'.'<>'Refirto Ti 2515/099 v'

~

~

e

~

0

lEg':,tO)

.Pe<<'.....

DE SCRIPT ISI

W~'i 'of.Fi II'.,4lt: 1ti

-'oeeeoeii,~,Trodi ltteri:,

'.TiC'.N85363". 1NA(i122:;4.,

NRC LEAD IR/SER ShyLatock LICENSEE STATUS NRC ACTIQI SER scheduled for early 1995' CQtKNTS

• Refer to TI 2515/122 STATUS

~:.'<~~a@~~>p~

gz'~"r<!~p :g<gp<'.~o<<<~~ ~~~>

W<>$<3gpSS i@<~~</g. ')gj'jaIt'ag'<<

':q<:"y'- y<~j~<'-"::Caito'r. -"'ER 11/13/92 a Refer to GL 92-08 IES 93.02 Debris Plugging of ECCS Suction Stralners TAC N86537I NPA X302 TAC N89279 NPA 8124

'-Vi'sait::.'liaier~Leyet(ffit;I";";;.

,<<,jAC'.IjI86884!!ejt"-X303@<~$>>

SRI SER 6/28/93 7/19/94 j

SER'4/20/94 IR'93-201 field caoplete 8/95

'-coaptete 3/95 s7odj f1 cit tons new Tl

'refer t'o Tl 2515/128 NRC CKKRIC LETTERS GL 82-33

":GL"'ai"+::";%/".'.j~

"9(j<<j'pwp~~,@2 Inst to fottou course of Accldentl RC 1.97 TAC N51075 NPA A017 Shyaiock Decker

~~',~@~~<<:4'~wgc'g

<<~~qp'r~~~';:Sa~iei"ATMSI':ji'25a0j020'-'~

'.SRI;:".'P4-",

<<MR@:+~>g..

SER 2/8/90 IR90.32 93-201 94.33 O<<<<<'P 94.33; TSjiange 6/21/94 on

~ASS':

j<!~ L<r<A~;

Project Plan; Tt 2515/087 closed IR 90.32 e refer to Tl 2500/020 iC GL 88.01 GL 88 11 GL 88.14 ICSCC ln SMR Aust SS Piping TAC N85296 Radiation Esbrittteeent of Reactor Vessel; RG 1.99 TAC N71469 NPA A023 Instrtaent Air Affact lng SR Syat~

TAC N71633 NPA $ 107 Stake NRR SRI SER 12/3/93 SER 6/29/89 SER 5/9/89 h

V A

4< >.<) <<

field ccoplete licensee states that TS emaendaents

190, 205 162 of 1/8/93 satisfied GL field conplete 9/95 T.;9/jim'i'cbees tied IPE!,fo'ruat I'>mjtsf" Expan'dad:PRA".of.',10

.'mftIIsIH"ops'dLIe:

5

!w)5.4e,~:.4" i<<

Project Plan KUREG 1435; not on licensee list; KRR to revleM for closure listed in KUREG 1435 and Project Plan

~SE<<R due 9/30/94 Saiaalc

'vituatfon;Report'due.to

-NiC)3<i19/<96"~'."-"'"'~-.'~

,"y<:,<<r'j',:IzH<

<~<)

4:::.>!

'< ~

)

"<e,'Refer'.::,to':TN1 'It'ie'1;D.2'.'

~

~

1SQK DESCRtPTtM KRC LEAD IR/SER LlCENSEE STATUS KRC ACTIQI CQtKNTS STATUS GL 89-08 Erosion Corrosion monitoring progras TAC N73459'PA L908 Stake SER 8/21/89 10/9/94 DRS to schedule module 49001 NUREG 1435 end Project Plan

<" Y '~

Y'fOVgaslit?tt Antf.

"tjf.".NTSQ7" %A

-: 110K:.'k4

';8grvfig!Vite,Sttatgag:.'.;f:.;:~

.'.::li>>'t'Rtfatf'iit'..of.':,,'ffarcfstiaP%

tAC!ffT4860":.',QA'!41N9@.7' Casta 2 s..

~<.+Y.9Y9Y<9'y Y< Y,

< 9,.9YPIY

'Kat'tojtg

- S4<?~<a.

SER'/16/9);

<u<>9'.9 Y, <,

V'9

, r., T,,g<<<.o,<.~Y.

i,<~<v.;;;

10</3ii94< "'4'-

e Refer to Tt 2515/109

-'e'R'afar'o Tl 2515/118 e Ra'fer'o Tl 2515/121 Gt. 92-01

'iaaf!e4iii'-':;:ii~:i'1'i'3%"~&~

Reactor Vessel Structurat integrity TAC N83440; NPA 8120

NftR".;.

j<'";-:.:

NRR SER"6/28/94~T SE."9)22/994":

Y< 9X SER 4/19/94 T 9/2/93 5/23/94 7/28/94 identtfy ccanitments; licensee to add to Pro ect Ptan Plant specific revfeMs befng performed on all units NUREG 1435 NR

','I'IS;j"'Y'<'r~D<@

WY

'N< '"'Tl8.':N842<T)"l'%A",l121.",".'",,9:

'SR'~,:,'P:

SER 3/25/N lR93"'l6'-"

11/25/94 Tl coo@lated; Further rey}ea of aods under tES 93'03 GL 92.08 Theraoleg Casto SE 5/11/94 TVA 3/22/95 RHRSII cables use thersetag; and milt be upgraded to configurettons as tested et N prfor to rx vessel hydro;ampecity/cocbustibte analysis by 12/22/95 and abandoned materfet removed by 6/20/96 GL 94.02 Long Tera goin for Theraal Hydraul lc tnatabl titles Peebl as 94-11 TVA 12/22/94 procedures to be revised prior to restart action 1.e closed GL 94-03 tGSCC of Core Shroud TAC N90083 Slake

'94-16 Licensee inspections performed J~ and duly 1994; TVA 9/23/94 concludes U3 cen be operated for at least 1

cycle SE 1/13/95 concluded structural integrity uftl be maintained for at least 1 cycle ufthout need for mod; TVA to reinspect DG

DESCR IPTIM WRESIKYED SAFHY ISIS KRC LEAD IR/SER LICENSEE STATUS KRC ACTION CQIKNTS STATDS

$ R3g~pPi."9+g'@~M~>F"-,,>>r.; >>...

.Nark"..3'.

f,"on0';

Ta'iia':PieBiaol lAC ll0793'1" NPA00<<1.%. '"

Blake 88-19

• Refer, to Long 'fera Torus Inteoifty:Prograe Tl 2515/085(clos>>ed"N

~ 19

?P~<s~>>9@l@pP'~v?4~98 9,"j~":$;>'.

4>.'>",.4 'v>'".S~

4 e Refer to'Tl 2500/020 A-36

,:0u>>a3>l'f Teat Iaq<,"'.,(les'i',,'18

'TJC!

42483'll'%1'lN60?;;-5 i;,'.

Control of Heavy Loads Hear S

t Fuel Pool "ShysIt>ock "

. $4>>p SRI 8841)

.. >>4 94-12

'9i 9> =,

~ $>>4$>x:rÃ4~>.

'. ",?e '>::

~. Refer.*to,EO ProRrasl; Tl 2515/OT6(c)osed N.-)TI -"

Project PLan

'"'PIja'Cracks":'fil.':~if'",,."-,-Slaked'('p g".'w4iv',jligk@N~g4Mc",~)$@gj3?....-

'4Th'.:gyv?~P'.9<<

"'Btitloj"'03ac>>k~'.j'::".g'"",g"",'.'".IShj'iit'ock~"

> r$ ?r @35>

<<.Pg4&$

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..s>><<';

Refer toor GL N-01-

s Refir to TI 2515/120'P~~W,

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'>>44

',:;v?pQ%??gp NRR"'eisslc Evsl Rpt/-

'e jiNIc liEEl~due 3j96'?.'9

'.I? c <</tv

."-~Sm>>g". 4...

M%?.,g,.4$ +'

Identified not to ha a

'astatt ttels in'SE e.g.

conduit support SE 3/20/92

~ Refer to GL 89-19 NR A-48 IKKRIC SAFHY I GSI 40 GSI 41 Hydrogen Control Neasures and Burn Effects TAC N55955 NPA A019 Safety Concerns Associated Kith Pipe

~reakea in SMR Scraa Syst TAC N43?36 NPA 8065 BM Scrm Discharge Vol~ Syatea TAC N51014 NPA 8058 SRI Blake SRI SER 9/9/86 SER 1/7/86 SER 6/24/83 field coaptete 6/95 Project Plan in MISP NUREG 1435 and Project Plan KuREG 1435)

TI 2515/090 closed in 87-13; Project Plan

>ggq4v".@Ppi@P~'%;9.a>>>>?4)%?<<494'>>>g~r>'.>@4j%;A:S(gg>9,

',:a >r>'j..."

(('g'Ra) fah3ljty;,o>>j,'Aji"'"-q"-'-=".SRI':$5::

5<:%~+4~A%gA!V~>>~rj::Pgi

~ -'$-"'l<.(X'Pc'.

. 4

".- v(>>j,>:4'9>>>>;

?'?'; '

>>?V>>(.;?>>?

" Refer to GL N~ 14 4v> ><<g; e Refer, to Tl 251$/118

'."~$pg~t'>>4 a>>~~j~>@@p~j pj )g f M)jil>>jy9j4<gmr'g '>?PV<>>>>>>'4 7

GSI 75 DESCRIPT IOI Seiee ATMS 4.5.2 and 4.5.3 RPS Teat Alternatives TAC 53966'PA 8093 NRC LEAD IR/SER Shymlock SER 8/17/90 LICENSEE StATUS licensee tracking under GL 83-28 KRC ACTIDN COIENTS GSI 75 Sales ATMS 1.2 Data Capebi lIty TAC l63573

%A 8085 SRI SER 6/12/85 licensee tracking under GL 83-28 IRI.Tl ACTIOI ITEMS

'PA A004 HPA 841 HPA C 10 Appendix J Cont Leak Testing H08717 Fire Protection Final TS N48136 Heavy Leeds Phase I Casto

'PRR NRR NRR SER 10/24/84 SER 10/12/83 SER 6/6/84 A

e Refer'.to GL 88 Renovsl of TS coeptete, NRR revieufng Fire Protection Plan subsittal Project Pion HPA C011 RPS Pouer Supply N08931 NRR SER 6/27/85 6/13/94 Project Plan IFl 84.32.02 IFl 04.41-04 IFI 85 09.02 VIO 85-41-01 IFI 85 51.01

'ontrol)Qelcil kkMF~%k';

Torus Level Inetr~tation Relocation of HPCI Eearg Control Boxes Bolts inadequate on Liaitor actors

~'y4. <8.Pg'<;P~ ~~~4PP<<@4+~

~>(.'

",",'interij'iAccepteb!l lay'f-.;.',",

P79x02 ~~P'~y gyw:I>~sgg+~g~~,;. ~

Cable Tray Sorts Cable Tray S~rt Crlter i~ SeisNc

'SRI SRI SRI Casto Blake Blake Blake IR-'4-"18'9'>>"-'$

5Q 94-27 95-03 field coaplete 5/95

~

~

ISSUE LER 8$ -32 LER 85-33 URl 86-06-02 URl 86-14-03 LER 86-10 LER 86-16 URT Il-02-02 DESCRIPTTM Reevaluation of Design Criteria for FSAR 8.6.2.1 Nonstandard 4" pipe penetrstfons thru sec cont

'.l~orpii,':, SLC,':',6m6~",-

Rx Bldg control Say WAC i

te desi Dverstress of Dryrrall Bema Design Engineers fdentffiad coreectfon of uncNaf if lad pfpfng to contafraent serai lfnes fluid leakage probfm with large bore srxbbers for Torus restraint k?".~g >,k<<'F~@g+:~~~gpgg(~(@~'>>.~g

-;;ffiutrctrf:;Neftgr,'~,;:tiat-';

'.daficfeetiaKC<<~5%5~>':.%.

Lfaftorque gear ratio NRC LEAD Shymlock SRI

~lake Slake

.""",~44<

%K)>><<rrw'>> >

.SRl>'~:.<

? <<".rx":>r,?ca

~lake Casto IR/SER 91-06 88-28

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)>>>>~>>

89-20

> Bk> A.

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-':89.';35."94,:32,":;'"j:-

88-28 88-28

,-a N.33 94.36'-:

.>r r?r> A>)

88-16 LlCErrSEE STATUS r ;: ~ r r NRC ACTlON

~,) >>r?

CQtKNTS STATrrs URT $7-26-03 RBR pusp auction snd noazfe load allowances sfbl exceeded

";t9,;.:Saf i'-'-Cl osreCf '.on!!8'lgi+j,"<<,"

Blake SR f.>>+ <<~9~

,"V,h'>>"'M 88-32 90.08 89'-'35l95;"$ 6:,~)"r

,".89;,35 95, r16jj.';

5>.",'g N@S~~MFPr'. "

+~'.j- )" g>>r":,i~)q~>

ffefd corpfete 2/95 r

r,

='ffef4;c~ieter 1/9$

? >'.

<>'j'p?<<c'Q.

~A'~>>j4'2"%""or". '>F~Y~",r+~a,'l~

IS%K DESCRIPT III NRC LEAD IR/SER LICENSEE STATUS KRC ACTION CQtKNTS STATUS Vio I-04-03 Failure to Correctly transtate deafn requ into draufn}}a SRI 91-02 91-16 field complete I/95 LER $512 ~attery failure concurrent uith LOP/LOCA event auto start Shyatock d9-35 LB N-'l6 Unplanned easel start of ESF due to ra error I<<~~gg:S~m~atfi::"'P-"-g"'~r:"&'"':~: ":e}lcW.-:W')i'.4fHti;;;~< ,';ifNgF"igfdent:Ib'i')iiai':;if -'RI d9-50 '9<<>> "24tscsrd aads needed C LB N32 LB N.P LB N.40 Elect aep requ violated due to deal controls Inadequate deaf gn control process dfscrep fn NVAC duct }5ork Inadequate desi ¹n controls results fn becktp control ayata not aeetf deaf r Shy@tock 90-25 Slake Shystock 91-16 ,pe%<<g+p>@S jcpr'ggjrpy$&r~gp), y. '&'Y.j???'::M'0) SRI-: ";;.;; x?? << 90'".03 95.22 95m 22 &X>>Q! -'55 ?!x, -'Ilk. "%&t}2."4 f'pOi¹afmtc:"xiii'jij'¹,:.,: ~'}lik's'.;.,'5-03 C $";A)~4!rgb>>x,.;~~A'k~~",~ (4}&4@'<<'.t}ae'? Df;"gfIsa}f',y'jjjRt"..'; '-'v'afyeaPjniN.:'.,:()nextxa. .":Mtret'~ "<<'5chf tter:".'8::,"~.- 'SRI = ', k,';".,':xr&58pgA x'>> SRI '5x"}>>! 91.10 5"l'/ 91 10 95.22 field cospleto 4/95 -.>>~i??""~$>:PS14 gg.,z;~x.cW ~>&yp+g>> <<g/5~<<<C~ p.;g! ry5..g <Oei}fiick <<jet;..ln",RN,'.anil;,. ??~gg,"@~V?'?:x:5Ã.x>>MZ~x'AS P'P;:>>:xxx+>>& 'S}tf "" -'P,-.-'. 91 02 95.22 95-22 &,? ??ar.'?"g":r"j'x "!, LER 09-25 Deaf}}n errors fn 250VDC reautta fn manatyzed cond Shylock 90-03 10

N ~ 0

IS%K IcfjRf.'~0M', > DESCRIPT IDN v q Iqp<<jc<<yv<<gp>>Ij<yx'~'<<-

'.

RPS/ARl'.cIIwra)ty< x NRC lEAD

• p>>"

SRI '. IR/SER '; 93-32'95.22 llCENSEE STATUS NRC ACTIQI c. ~NTS vv y'TATUS >>y>>: C<<:(v<< 'yvy>>XNX.

, Oef/i fa'nCfaa',.'fWttie

'"ikw"t" ':",::.:)jest'eticle L'Ikfc~r

.'t'a'N<<t..::;"::.'/N~-~&yN;:<.A,'::,":;.'.:,

Central.': jf.'.,'.Ceit~t O4.'",

Sill'<; c .XS) '95;10 T95>> 10'~); ',":".'",'yc c" c.cc< ~ '. ~ y?zc?c"~."v'<<?'> v < c 'PER,S2l =- C C lER 01-15 , '<<?<<Kc'<<P'.+

.FaljflreyfgffO',,Irvfna,ofy

's.::4'.':;;;"4 ~acf: MPCI dfd not fulfilL safety fcaIctfon froe Low auction pressure during fest start ,,;I<I'jjifock.;>' z>>"';<<,?'s 5 ov C SRI ~<<< ~c',.Cg., 'Z'j:y<%g$<<<k 6,v8".10.;::.9S.$ 0 '" .- g4>4q'.,Vm~ '. 92-11 95-10 xyx '" xcglx' ~'P~'yyg~4p;c,'<<C c? Nod complete, post sod testing ach for 9/95 C .:I $4~<: '~<~~44'.:94 K<Qxxy',:.

,via';

?X~~pgg~vSZv'@j>>ACkggv,qg..':.)jy') F'..< Sag ffrL:fr<rob)aIN:-;yy,4pr!nff'= I??@U<<>>x':: "TAN@(<<8."x%'Zc". Ix<<<<<<.'cf !':.NIItiii';fj',:.V'Ni..!5j5>."<<':'- " "-fiat'i'farl<4":::~"-'"+~".l"'--'. ':ESF, mtuatfo<l<lc'fry',+lay.: ". Ifa louie"'%~~~"~c'-: ~<~4~:.. s 'y'<<ay>>y>> . Sv)afcjz"."c x54".yy%."A p:C'x<jj"Nv:0( SRl i"'"y 4>>~<>><<'<vl. j& <Zg ( rcc Z v <<+) Sv.y('<A ?Ice<'Ig':.'T a c3:;x>gQ/xc - 94..04".94'%"" 'g.x )?? ec ia<cS y c< cz x $.,< c>>c 94'20":.. "'"-'x Z,.vAC '<ye .,", <<Xv,<<c",g<<xx."x 'c v:x xc'.<< c,- <<f'C.I>>~,',<<'.y:,'v" ".- z v >>z j c y '<y C C Oy.<<x, . ~pMI<<cq>0~/A)<< klNj. '054)' @a"-zgp>> Qmp:~~@~M URI 8-11-01 -'."<x"-cz~>> F~<"-Fz~ 'IccW4~cÃg>< "..'F Nfl'Ur'iLNf7Rmiai.'::.'iiei,'."

favovf""

'":toYktek%Xc.

mji "':.a'atvei~'.:iff'~

i'aecon'da'r'V'Too'nti( remi'...Ntw .'.to'.'Si;ri't'eaae<d.'.tjIriRRC~M!l'.. Neld Dlffererences between the welda asauaed in a rt tl~'jq'y'~V vSRL x>?'A>- . >>v F~x5 (vs b$ SKI? '" v; cc Slake p'Ms 5>g";<PS>+yj>>'4;32'- >'-. ',"* vc<Iic s?<<~ I '.9x5.;,'10." i;:.". " "'c c @..xg Y@S:,) << ~ < c< yA "': v '<<g'CV vfo 92-37-01 '",'<ycg>p '4?gy >xz IFI '8 04 02 IF1 N.OT.L8 Verify Nethod used to install wedge anchors Verfffcatfon of SSD FINIctional R fronts Slake KRR 95-'fS IFI N 07-03 Systcs operational boundary teat fdentfffcatfon

~ ~

IFI 94-11.02 I F I 94-12-01 VIO 94-17.01 lfl 94-1d 02 lFI 94.29.01 IfI %4.29-02 URI %4-29-04 VIO 94 33 01 VIO 93-03-01 Nl N.10.01 NI 5.16.01 DESCRIPT I M

Response

to GL 94-02 Rosmont Trsnsal tter Drift ProbLem Ine trident at Ion Calibration Deficiencies Condl t lan of Contaltaeat Costi Reviau of C~ Concrete c lt Design Nethoda for Anchor Location Tolerance Aapllflcation factors for anchor loads failure to salntaln spacing ln Low voltage cable tr Spring can installation inadequate Second party check by foremn (gate T ) IJork Perforaed on incorrect i KRC LEAD Peebles Shymlock SRI SRI Slake Slake Slake Shyalock Slake Shyalock SRI IR/SER 94-35 LICEKSEE STATUS KRC ACTIOK CQIKKTS STATUS LICE'%SEÃSEKTS TS 339 TS 337 germ Pilot Alr Header Pressure Saltchs TAC Appendix R License Aaewrsh?ent TAC K07902 KRR KRR h~~.$'<@M'<?M<al "<<"'".N~VP<e@<<:?Ã'<<<~Yj~i'-,'j~<'?>>". ~'<<<:~)P$ %i"0+>> '"" 'AO'$$9:> s.8.A"<<$~ V? p?'4 ':. 4 ." g'~Q' p> r><<Q'p(4Q <<Q<'gr5~ pp'<<sMgk p~~ <<"'+pc ~<. ')~gy g$,,:,g?glgP44,:;;0!~L~QNef'Rtol!:;1,+"-'"" KRR rkN~ '~44<@4 ?.Shgddi ':. g;ggg$ yg:>> -;.'. 'y.;-,, ~: SE '3il<3i95 'S 315 Analog Transa it ter/Trlp System TAC N892SO 12

~ ~ t ~

DESCRIPT IM NRC lEAD IR/SER lICENSEE STATUS NRC ACTION COUNTS STATUS N-416 1 N-498-'I 3 end Ravfsad Rgb 3 xP'sa r'ihilTt TAC!N89253 Pressure testing Relief R t Pressure testing Relief R t Stencby coolant s pply P NRll NRR NRR NRR SE 2/24/95 NISC IMKt Procehrea Upgrades TS Changes Kellogg 93-36 SRI >>g OFcc~jgyKW4 P'>>A~gÃp ~I>>>>'"~:4>>SPIES@'~c~p,: Q j!~gg+j~wjg:...j4 4+Ieht'<$0NJ!Rt4.'~l~l,::,>>>'~,~;.'P'e'ebti">': TVA ltr 12/IF/91 'PP pg IV-17 NPCI Controllir I avmenta SRI SSER2 App E NC0560270003 NPP pg III-57 Shroud Naad Bolts (IGSCC) NPP pg II-%4 Online Chea Inst+a Slake Decker SSER2 Sec 4.10 NC0860326292 SSER2 Sec 3.6 NC0860326'l15 NPP pg ll-58 Unresolved CAQa EROS link to NG retinal Revlau licensee CAID closeout letter SRI Barr SSER2 NC0560326143 periodic insp of PERe PROGRANS Ilg'LHKNTHIlg ACCSSAIICE VITN EMIT Cable Aapeci ty 2 PRECEDENT Shymiock 94-35 <<50X iapicment 75X Implement 4/95 complete 5/95 NUREG 1232 V3 S2 revieved prograa and NRC ltr of 6/23/93 revived follaap items Cable Trey supports N 80654 Blake SER 12/I'/91 TVA ltr 3/27/91I <<90X Iapiement c late 2/95 13

~ ~ 0

DESCRIPTION Cooponent and Piece Parts Ruali f1 cation N 83828 Conte iJ>>ent Coatings CRD Insert and IIIthdraual PI I Design Cele Revlem Eg TAC N42482 ISA 8060 flexible Condulta flea NVAC Dmt Rpporta TAC R00300, INI2127 larBe Bore Piping and Supports (IEB 79-02 8

79. 14)

TAC R0017 Nisc Steel fr~ TAC R00297 N80620 Noderate Energy LIJ>> Break Platform Thermal QroJJth TAC R00297, N80620 0-List Seismic Class II/I TAC N80015 NRC lEAD Blake SRI Blake Cesto Shymlock Blake Shymlock Blake 'Olik'e'. Blake Blake Blake Blake NRR SRI: Blake IR/SER SER 12/7/93 94-01 94-09 94-18 94.27 95-03 94-31 94-06 94-27 94-35 SER 13.6 93-02 93-08 SER 10/24/89 7/16/92 IR 93-201 i'Aa, ~~j..w;;sAjp '92'3l'.'93:05"'ER 10/24/89 94.15 94.29 95.03 SER 10/24/89 94-15 94-29 SER 4/20/94 SER 10/24/89 4/20/94 IR 93-201 94-29 SER 12/17/91 lICENSEE STATUS TVA itr 6/12/92 <75X Isptement 50X lsp lament 2/95 coep late 5/95 75X ispl ament 10/94 coepl ate 7/95 >50X isptement 75X ispi ement 11/94 c late 1/95 I 1/19/95 TVA ltr 2/27/91 50X Ispiement 12/94 75X I Iement 5/95 NRC ACTIOI Ifl 94-18-02 opened to track re irs to U3 Inspect es pert of open Item closure COIKNIS Program hss changed; possibly perform module 38703 on rocurement TI 2515/076 closed 92.03 e Refei to Tl 2515/089 'RS kevleued program and found icc table" TVA to submit salti unit PRA 4/95, IPEEE 6/95, Internal fires IPEEE 120da after refueling STAHJS NR 14

~ ~

DESCRIPllDN Splices Thermal Overloads NRC LEAD IR/SER Shymlock 90-22 95-14 Shymlock SER 13.4 LICENSEE STATUS KRC ACTION CQtKNTS SSER2 3.13 found program acceptable 40(I IR satisfactory for alt units STATUS PRSRAIW lRICN DEPART FRCW TK Wll 2 IIPtDKNTATIDNPRECEDENl Cable installation NM682 Conduit Support TAC R00024 N80690 Conflguratton Ngmt/Design ~ass tine N80688 tnstrmont T&tng TAC NBOOM tnstrt>>snt Senatng Linea TAC N80017 Long Tera Tore>> Integrity TAC N80686 Restart lest Program TAC N81791 Small Bore PIplng TAC N80013 R00306 DEPARl FROI LNILT 2 CRllERlA PR Fire Protection; App R. TAC N85254 Louar Dryuat t Ptatfo~ ~nd Nlac Steel TAC N80620 R00303 Shymlock Blake Casto ~lake SRI Blake'RI Cas'to Blake Casto Blake SER 4/8/92 '7/1/94 93.34 94-27 94-35 SER 10/24/89 3/30/92 94-07 SER 11/21/91 94-07 94.20 94. 31 SER 2/4/92 95-03 SER 12/10/92 IR 94.24 SER 2/10/92 94-15 SER 8/30/94 SER 10/24/89 2/4/92 95-03 94-27 95-04 95-07 SER 7/26/88 'IO/24/89 3/19/92 4/20/94 IR94-15, 93.201 94.29 >SON Imptement 75K isptement 10/94 c late 5/95 TVA2/27/91 TVA12/12/91 TVA ttr 2/13/91 TVA ttr 4/29/91 >75K Isplement c late 5/95 TVA Ltr 2/2/94 SON Isptement 6/95 TVA2/27/91 lVA12/12/92 Licensee submittal of 12/20/94, status > SOX Isplement 75K I lement 4/95 TVA 6/12/91 >NC Iapl ament cosptete 1/95 NRR to urtte SER by 7/95 revised bend radius for medtua voltage cabtes Licensee hss cost>tned inst tubing and ssmt t bore I I rograss URI 94-15-01, Spring Can Setttngs SRl - Revleu administrative program; Casto - identify electrical/mechanical tests snd Ins ctor to revieu. Long term design criteria Isptcments 1978 AISC spec PRDCRANB aaeuETETI W ALL TINLEE mtTB 15

OESCRIPTISI Neat Code Traceabillty Secondary Conte lrmnt Penetrat lone Melding Progrm Pipe Ilail Tbiming <Gi. dr-01) NRC LEAD Blake Slake Slake ~lake IR/SER SER 5/31/90 SER 4/11/88 SER 5/31/90 SER 8/31/88 LTCENSEE STATUS NRC ACTlON CaraNTS NURUG 1232 V3 S1 sec 2.3 and NRC SER of Nay 31, 1990 revieaed prograa for all 3 mits. Prograa evaluated by April 11, 1988 addressed all 3 mlta Melding concerns adequately addressed per NUGEG 1232 93 51 SER addressed all 3 mits STATUS Sourqis for issues include:

IFS, SIMS,

WISP, NUREG 1435 (Status of Safety Issues at Licensed Power Plants),

BFNPP, NUREG 1232 (SER for Browns Ferry NPP) 16

e

Browns Ferry Multi-UnitPRA ain Rcport Revision 0 I. EXECUTIVE SUiVIMARY The U.S. Nuclear Regulatory Commission (NRC) policy statement on severe accidents in nuclear power plants was published ~n the Federal Register on August 8. 1985. The severe accident policy statement of the NRC concluded that existing plants did not pose an undue risk to the public health and safety. However, the NRC stated that systematic examinations are benefiiciai in identifying plant-specific vulnerabilities to severe accidents that could be fixed with low cost improvements. The NRC's plan for implementing the severe accident policy statement was published on May 25, 1988. The first step in this plan was to request that licensees complete an Individual Plant Examination (IPE). Tlie IPE was intended to be "an integrated systematic approach to an exatnuiation of each nuclear power plant now operating or under'construction for possible significant risk contributors that might be plant specific and might be missed absent a systematic search." On November 23, 19S8, licensees were requested by Generic letter No. S8-20 to perform an IPE/probabilistic risk assessment (PRA) that addressed each plant in order: "(1) to develop an appreciauon of severe accident behavior, (2) to understand the most likely severe accident sequences that could occur at its plant, (3) to gain a more quantitative understanding of the overall probabilities of core damage and fission product releases, and (4) ifnecessary, to reduce the overall probabilities of core damage and fission product releases by modifying, where appropriate, hardware and procedures that would help prevent or miugate severe accidents." A PRA is the usual and preferred method of performing an IPE. A PRA is a structured analysis of postulated events, equipment failures, operator errors, or various combinations of each, which could result in a degraded core and/or a major oQsite release of radioactivity. In response to Generic Letter No. 88-20, TVAcommitted to model BFN Unit 2 and perform a PRA and containment analysis. However, in August 1990, the NRC noted that the three units at BFN share many important safety systems. The NRC expressed a concern with the potential safety implications of shared systems in the various operating modes ofthe BFN units; e.g., all three units operating, Units 1 and 2 operating with Unit 3 shutdown, etc. In response to this concern, TVA committed to perform a multi-unit PRA, which bounds the various combination of units in operation and evaluates the impact of the shared systems on the probability of a degraded core calculated by the BFN PRA. The single unit BFN PRA was submitted for NRC review on September 1, 1992, and approved by the NRC on September 28, 1994: As part of the commitment ofTVAto maintain the BFN PRA current over the life ofthe plant, the PRA that was submitted to the NRC review was revised as a result ofplant modifications and to refine previously modeled plant features. The enclosed report pmvides the BFN Multi-UnitPRA. The results ofthis multi-unit analysis indicated that the most limiting site conQguration is with all three BFN units in operation. The resulting coze damage Gequency for Unit 2, with three units operating, of? SE45 is ETVAW00383)OC.04/08/95

t Cl

Browns Ferry Multi-Unit PRA VIain Report Revision ii approximateiy a tactor of 4 higner than the revised singie unit estimate oi,.6E-06: however. the muiti-unit core damage frequency still represents a very low risk from severe accidents. As shown below. no single initiating event was found to dominate the total frequency oi core damage. .4o plant vulnerabilities were identified for BFN wnen multiple units are in operation. Therefore, no additional enhancements are required to address vulnerabilities.

1.1 BACKGROUND

This report documents the work performed by Tennessee Valley Authority (TVA) and its contractor, PLG, to investigate the influence on the core damage &equency (CDF) at Browns Ferry Nuclear Power Plant associated with the bounding configuration of all three units operating. TVA has previously submitted to the U.S. Nuclear Regulatory Commission a plant-specific probabilistic risk assessment (PRA) for Browns Ferry Unit 2 in September 1992 (Reference 1). That analysis, referred to as Rev. 0, represented the plant conditions at the time of the submittal; namely, Unit 2 operational and Units 1 and 3 defueled. TVA subsequently performed updates, the latest denoted as Rcv. 1A, to the Unit 2 PRA to reduce some of the iniaal modeling conservatisms, to incorporate the effects of design changes at the plant made since the original analysis, and to incorporate selected plant-specific data. In the Rev. 0 and Rev. 1A PRA for Unit 2, plant systems and features shared among units were considered to support Unit 2, as appropriate. TVA committed to the NRC (Reference 2) to perform an expanded PRA that considers the shared plant systems and features, and considers in this study a particular bounding configuration in which all thtcc units are in operation. This rcport presents the results of what is referred to as the Multi-UnitPRA. The methodology used in this study is stimmarizcd in Section 2 and is a straight forward extension ofthe methodology used with previous PRAs on Browns Ferry. The main difference is that this study considers a comprehensive set ofmulti-unit interactions that was not addressed in thc previous PRAs. Potential system and unit interactions are first'identified. Next, a bounding plant configuration is determined. This bounding configuration specifies the initial status of the three units., Initiating events that are specific to multi-unit operation are then jdcntifictL In additio~ system and opcnitor action success criteria specific to multi-unit operation is dctcrminctL The models developed for the Unit 2 Rcv. 1A PRA were used as a starting point in the current analysis. The additions and changes to these models that were ncccssary are documented in Section 3. This report also presents the impact of expanding the PRA models developed for the Rcv. 1A analysis to explicitly consider thc effect of the loss of control bay ventilation. ETVA'W003S.DOC04/08/95 1-2

e

Browns Ferry Multi-UnitPRA blain Report Revision v As part of the Multi-UnitPRA. dependency matrices simiiar to the ones deveioped for the Unit 2 Rev. 0 PRA were updated for Unit" and new ones developed ror Units l and 3. These matrices document the intersystem dependencies that exist between piant systems considered in the PRA and are also provided in Section 3. Section 4 documents the results of an investigation of the multi-unit interactions to veriry no risk significant vulnerabilities were overlooked in selecting the bounding plant configuration. Section 5 describes TVA's participation in review and performing the Multi-l.nitPRA. Section 6 documents the unique strengths of the Browns Feny Nuclear Plant and the assessment of plant vulnerabilities and potential enhancements. Section 7 summarizes the final conclusions of this Multi-UnitPRA. The references of the report are provided in Section 8. and the detailed backup calculations and documentauon are provided in the appendices. 1.2 RESULTS The quantitative findings of the Browns Ferry Multi-UnitPRA are presented in this section. and are compared to the results of the Unit 2 Rev. IA PRA model. The results delineate the principal contributors to risk. The basis for the multi-unit analysis and, therefore, the basis of the comparison of the Multi-UnitPRA results to those of the single unit PRA, is the frequency of core damage. For the Multi-UnitPRA, an initial plant configuration. which is bounding with respect to the availability of systems to avert core datnage, is selected. In this manner, the consideration of the CDF results of the single unit Rev. IAmodel and the Multi-UnitPRA model provides lower and upper bounds, respectively, for the CDFs that would be applicable to the other possible initial plant configurations at Browns Ferry. The same iniuating events were used for both models, plus six additional ones for the multi-unit model quantification. The baseline configuration date for both the Multi-UnitPRA and Unit 2 Rev. 1A is May 31, 1993. The mean value ofthe uncertainty distribution for the total CDF for Browns Ferry Unit 2 under the conditions that aH three units are initiaHy operating at power was found to be 2.8E-05 per reactor-year.~ For the Rev. IAmodel, conesponding to Unit 2 initiaHy at power and Units I and 3 defueled, the mean value ofthe distribution describing the CDF was determined to be 7.6E46 per reactor-year. For both analyses, core danuige is a.mmed for any sequence in which su.tained core uncovery occurs. Per the vulnerability criteria specified for the IPE Rev. 0 report and provided here in Section 6, no vulnerabiTities were identified. The results for CDF were developed in terms of a mean point estimate, as required in 'The unit for the CDF is events per nuclear-powered electric generating unit per calendar year. This deGnition is abbreviated to "per ratctor-year." KTVAQl0038&OC.04/10/95 1-3

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Browns Ferry Multi-UnitPRA Main Report Revision 0 'AUREG-i335 (Reference 3). as we11 as the previousiy cited uncertainty distribution. The presentation of the total CDF in terms of the uncertainty distribution is shown in Figure 1-1 tor the Multi-UnitPRA and Lnit" Rev. 1A PRA. Note that the Monte Carlo process used to determine the uncertainty distributions yields a slightly different value for the mean than the point estimate mean reported elsewhere in this report. This deviation between point estimate and Monte Carlo means is normal and results from small numerical uncertainties associated with the Monte Carlo sampling process. Descriptive parameters of the uncertainty distributions are as follows: PRA 5th Percenti)e 50th Percentile Mean 95th Percentile Multi-Unit Unit 2 Rev. IA Unit 2 Rev. 0 4.5E-06 1.6E-06 5.6E-06 1.5E-05 4.5E-06 2.8E-05 7.6E-06 4.8E-05 8.2E-05 2.3E-05 1.1E-04 In the quantification of the Level 1 event sequence models, the principal contributors to the CDF were identified from several vantage points. The results and contributors arc summarized in this section for the multi-unit model and compared to the Rev. 1A model. The Multi-UnitFRA was iniually based on Unit 2 Rev. 1 PRA model. In the process of developing the Multi-UnitPRA, refinements to the Unit 2 Rev. 1 model were provided, and the Unit 2 PRA was updated to Rev. 1A by TVA. 1.2.1 IMPORTANT CORE DAMAGESEQUENCE GROUPS The importance of initiating events was examined by determining the contributions of core damage sequences grouped by type of initiating event. The ranked results are shown in Table 1-1 and Figure 1-2 for major initiating event categories. As can been seen, thc mean CDF corresponding to the multi-unit configuration, while still small, is about a factor of 4 greater than the corresponding mean CDF ofthe single unit configuration. The reason for the increase is the change in success criteria for shared systems for initiating events that could impact two or three reactor units concurrently. SpeciGcally, the impact of the change in success criteria for such shared features as diesel. generators, emergency equipment cooling water system (EECW), and residual heat removal service water system (RHIKW) is evident for initiator categories such as loss of ofBitc power that could impact all three units concurrently. For initiators (such as those that comprise thc category "transients with reactor not isolated" ) that involve csscntiaHy a single unit to respond, the impact of shared features is much more modest. ETVA'%00383)OC.04/08/95

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Browns Ferry iviulti-UnitPRA Main Report Revision 0 A detailed listing oi the contribudon of each initiating event to the CDF is given in Appendix C. and is summarized below for Unit 2 in the Multi-UnitPRA and compared to the Unit 2 Rev. 1A PRA: Scenarios initiated by a loss of offsite power contribute 39% oi the CDF in the Multi-UnitPRA as compared to 20% for the Unit 2 Rev. IAPRA. Scenarios initiated by internal floods contribute 22% to the CDF for the Multi-Unit PRA as compared to 15% for the Unit 2 Rev. IAPRA. No internal flooding scenarios lead directly to core damage but require additional hardware failures. Flooding initiators were postulated in the Unit 2 reactor building, in the Unit I or 3 reactor building, in the turbine building,'nd at the intake pumping station. One flooding sequence, initiated by a flood in the turbine building, has a mean frequency greater than 1.0E46 {1.2E46) in the Multi-UnitPRA. No individual sequence in the Unit 2 PRA was greater than 1.0E-06 in frequency. Support system failure initiators (specifically, loss of plant air; loss of raw cooling water; loss of unit prefetred power; loss of either instrumentation and control bus 2A or 2B; or instrument tap failures) contribute 21% to the total CDF for the Multi-Unit 'RA as compared to 2% for the Unit 2 Rev. IAPRA. Transients with the reactor not isolated contribute 8% to the CDF for the Multi-Unit PRA as compared to 28% for the Unit 2 Rev. IAPRA. Turbine aip and loss of feedwater are two specific examples of initiators in this group. Transients with the reactor isolated as a result of the initiating event (initiator) contribute 7% to the CDF in the Multi-UnitPRA as compared to 25% for the Unit 2 Rev. IAPRA. Closure of the main steam isolation valves (MSIV) and turbine trip without bypass are two specific examples of initiators in this group. Large and medium loss of coolant accidents (LOCA) and interfacing systems LOCAs (ie., when the boundary between a high and a low pressure system fails and the lower pressure system overpressurizes) make up only a small part (2%) of the total CDF for the Multi-UnitPRA as compared to 7% for the Unit 2 Rev. IAPRA. The absolute change in contribution to CDF actually decreased slightly (5.0E-08) due to modeling refinements incorporated into the Multi-UnitPRA but not into the Umt 2 Rev. 1A PRA. ~ Scenatios initiated by the inadvertent opening of one or mote relief valves contribute only.a smaH part (1%) to the CDF for the Multi-UnitPRA as compact to 3'%or the Unit 2 Rev. 1A PRA. 'Ihree distinct initiators are considered: opening of one safety relief valve (SRV), opening of two SRVs, and opening of three or more SRVs. A review of the top 25 sequences leatHng to cote damage provides some insigitt as to the varying nature of core datmtge sccnazios, far the Multi-UnitPRA. Twenty~ af these QVAVl0038J)OC.04/12/95 1-5

Browne Ferry Multi-UnitPRA Main Report Revision 0 seouences were initiated by "multiple unit" iniriators (plant disturbances that have the potenual to impact more than one operating unit). Specirically, these initiators that appear in the top 2S seouences are Internal Flood in the Turbine Building (eight scenarios), Loss oi Offsite Power (eight scenarios), and Loss or Raw Cooling Water (five scenarios). Of the four "single unit" scenarios in the top twenty-five, three were initiated by vessel isoladon events (Closure of AllMSIVs. Loss of Condenser Vacuum. and Turbine Trip without Bypass). The remaining "single unit" scenario was initiated by a Loss of Feedwater. The top two seouences are of a similar nature. Both are initiated by a "multiple unit" ininator (Internal Flood in the Turbine Building and Loss of Raw Cooling Water) followed by hardware failure of all four RHR pumps. Me total frequency for these two sequences is 2.28E-06 (or about 8% of the total CDF). Hardware failure of the four RHR pumps is common to ten of the top 2S scenarios. The increased importance of RHR failures in the Multi-Unitstudy is primarily due to the reduced availability of the interunit RHR crossties for multiple unit events. Table 1-2 summarizes the functional failure group contributions to core dating frequency. Failure of heat removal is characteristic of three additional sequences of the top twenty-five. In two sequences,'all four RHR heat exchangers fail, and in the reinaining sequence, the RHR pumps fail due to the loss of pump cooling (specificaQy, loss of the fan coolers). The third sequence overall is initiated by a loss of offsite power foQowed by hardware failur of aQ diesel generators. This is the most limiting station blackout sequence and represents about 2% of the total CDF. Two other sequences in.the top twenty-Gve are related: sequence 7 is a loss of offsite power foQowed by failure of the Unit 1/Unit 2 f'uel oil transfer pumps; and, sequence 22 is a loss of ofBite power foQowed by hardware failure of the four Unit l/Unit 2 diesel generators only. Transient initiators followed by loss of two vital DC power supplies characterize six of the top twenty-five scenarios. Transients initiators followed by ittadequate EECW Qow characterize three of the top twenty-Five scenatios. 122 ANALYSISOF INDIVIDUALSEQUENCES No single core damage sequence was found to dominate the total frequency af care damage. A large, number of sequaices make up the total CDF. Table 1-3 provides information on the distribution of core damage sequences across the &eqttency range far the Multi-UnitPRA as compared to the Unit 2 Rev. 1A PEA. The noted decrease in the number of seqtiences in the highest fluency category is due to the added complexity of the Multi-UnitPRA model that results in additional split &action branching; e.g., more sequences but at lower values. 'ee Appetidix C for further details. QVAW003833OC.04/12J99

Browns Ferry Multi-UnitPRA Main Report R.evision 0 1.2.3 IMPORTAi>T OPERATOR ACTIONS The importance of a specific operator action was determined by summing the frequencies of the sequences invoiving failure of that action. and comparing that sum to the total CDF. The importance is the ratio of that sum to the total CDF. This analysis provides a relative importance of the operator action. as it only determines the CDF impact of sequences that include the operator action. but does not distinguish whether the sequence failure is due to the operator action or the component failures. Table 1-4 summarizes the important operator action failures ranked in order of their impact on the total CDF for the Multi-UnitPRA and the Unit 2 Rev. 1A PRA. The operator actions to recover offsite electric power are not included in Table 1-4 because they are a complex function of the time available and the specific equipment failures involved. The offsite power recovery actions split fraction importance is shown in Table C-13. 12.4 IMPORTANTPLANT HARDWARE CHAIMCTERISTICS An importance analysis of plant system failure modes to the total CDF was also performed. Only hardware failures involving the system itself are considered in Table 1-5, which provides a ranking in order of their impact on the total CDF for the Multi-UnitPRA. The Unit 2 Rev. 1A PRA impacts are also shown in the table for comparison. The system importance measure is the &action of the CDF involving partial or complete failure of the indicated system. These importance measures are not strictly additive because multiple system failures may occur in the same sequence. The importance rankings account for failures within the systems that lead to a plant trip, or failures that limitthe capability of the plant to mitigate the cause of a plant trip. Consequential failures resulting from dependencies on other plant systems (e.g., the loss of drywell control air due to failure of reactor building closed cooling water) are not included in this importance ranking. Care must be taken when comparing the results of the multi-unit PRA to the Unit 2'PRA as gauged by the PRA importance since this quantity is merely a relative meastire. For example, RPS system failures appear in 7% of the core dannage scenarios in the Multi-UnitPRA; the corresponding importance measure for the Unit 2 PRA is 20%. The relative'nature ofthe measure is apparent when 38% ofthe multi-unit CDF 2.8E-05 (or 1.96E-06 is compared to 20% ofthe Unit 2 CDF equal to 7.6E-06 (1~~. RPS is "more important" in absolute CDF impact in the Multi-UnitPRA, than in the Unit 2 PRA, a fact not communicated solely by the importance measures. What is apparent in Table 1-5 is that systems that are shared among the units to a significant degree (such as the diesel generators, REGKW, and EECW) are relatively more important in the Multi-UnitPRA, as compared to the Unit 2 PRA. hTVAN0038330C.04/10/95 1-7

Browns Ferry Multi-Unit PRA Main Report Revision u L3 SENSITIVITY ANAI.YSIS: EXTENDED DC POPOVER AND ALTERS'ATE INSECTION CAPABILITY An analysis was performea to determine the risk reauction potential or the following: Using the diesel-driven fire protection system pump to iniect water into the reactor vessel upon loss of AC power. Providing an alternative source of power to the SRVs solenoid valves to permit depressurization of the reactor following loss of AC power and depletion of batteries. These improvements are evaluated in conjunction with the hardened wetwell vent because ox the interaction each improvement has on the other. Although separately each has benefit, taken together they provide an open loop cooling mode for the vessel with a flow path from the diesel driven fire pump into the vessel, through the SRVs into the suppression pool, and out of the hardened wetwell vent. During the preparation of the Unit 2 PRA Rev. 0 (issued September 1992), TVA recognized the potential of using the diesel~ven fire pump for vessel injection or debris bed cooling and subsequently prepared a system notebook for the high pressure fre protection system. However, the results have not yet been incorporated into the PRA model. The pump is capable of removing decay heat only after about 4 hours, therefore successful initial vessel level control (such as provided by HPCI or RCIC) is required. The SRVs are capable of extended operation in that a nitrogen gas supply can be aligned. DC control power to the solenoid valves is still required. The valves required to open for the hardened wetwell vent path also have a backup nitrogen gas supply and require.DC power. These valves are located outside containment and can be locally operated via handwheels (pxior to any postulated core damage). The analysis was performed by running a number of sensitivity cases using the Multi-Unitand Unit 2 PRA models. For each model, two cases werc evaluated The difference between the two cases is that one assumes thc availability of a supplemental DC power supply for the SRV solenoid valves, where the other case requires that offsitc power be restored within 6 hours in order to provide a DC source for the SRVs. Both models used the same probabilistic values for the avaihbility ofthe diesel driven fixe pump and the manual actions required to align the pump low path for vessel injection and remote manual operation of the haxdened wctwcll vent. For thc Multi-UnitPRA, the supplemental DC power case pxoduccs a CDF of2.6E~ from the loss ofoffsitc power mitiator, while the ofhitc power recovery required case produces a CDF of2.7E46. The baseline Multi-UnitPEA CDF, duc to the loss ofoffsitc power initiator, is 1.IE<5. The Rcv. 1A Unit 2 PRA showed similar results with the supplemental DC power case CDF due to thc loss ofoffsitc power of 5.7E7, while the offsitc power required case produced a CDF of 5.9E-07. Thc baseline Unit 2 CDF, duc to the loss of offsitc power initiator, is 1.5EM. ETVAW00385)OC04/l0/95 1-8

Drowns Ferry,~lulti-L'nit PRA ~lain ikcport Revision 6 These results indicatea that the use of the diesel-driven fire pump in an open loop mode of core cooling re'.1ects a reduction in the computea core damage rreauency due to the loss of AC power..'viost oi the gain in risk reduction is achieved through use of the diesel-driven fire pump ana the nardened wetwell vent. which are already in place.. roviding an alternate source of power to the SRVs is not.warranted. This is especially so once consideration is given to the fact that the 4-hour battery depletion time is based on a conservative calculation and that relatively low current is required to maintain a solenoid open to allow an SRV to function. Based on this. TVA has no plans to provide an alternate source of power to the automatic depressurization system solenoid valves. Use of the diesel-driven fire pump as an alternate low pressure iniection source is already discussed in the Emergency Operating Instructions. ETVAtN0038.DOC.04/04/95 1-9

Table 1-1. Initiating Event Group Contributions to Core Damage Frequency Multi-lJnit PRA I/nit 2 PRA Initiating Event Category Loss of Offsite Power Internal Floods Support System Failures Transients with Reactor Not Isolated Transients with Reactor Isolated Loss of Coolant Accidents Stuck-Open Relief Valves Interfacing Systems LOCAs Total Mean CDF (per Reactor-Year) 1.1E-05 6.1E-06 5.8E-06 2.3E-06 2.0E-06 4.6E-07 1.9E-07 4.6E-08 2.8E-OS Pcrcentagc of Total 39% 22% 21% 8% 7% 2% 1% <<1% ]00% Mean CDF (per Reactor-Year) I 5E-06 1 IF-06 1.7E-07

2. I E-06 1.9E-06 5.11:-07 1.9E-07 4.6E-08 7.6E-06 1 cfcculagc of Total g0n~

15'! u 2"0 28",n 250/ 70/ 30/ < 10/ 100%

Table 1-2. Functional Failure Group Contributions to Core Damage Frcqucncy Multi-UnitPRA Unit 2 PRA Accident Sequence Group Loss of RHR Degraded EECW Transient followed by Loss of Vital DC Power (250V Boards 2 and 3) Anticipated Transient without Scram Station Blackout Transient with Vessel at High Pressure Blackout of Unit i/Unit 2 ~Not calculated. Mean CDF (per Reactor-Year) l.lE-05 3.8E-06 3.3E-06 1.7E-06 1.7E'-06 S. IE-07 5.0E-07 Percentage of Total 39 14 12 Mean CDF (per Reactor-Year) 2.5E-06 1.6E-06 5.2E-07 Percentage. of Total 33 21

Urowns &erry 11ulti-bnit l'RA Main Report zc,evtsloll v Table 1-3. Breakdown of Core Damage Sequences in Each Frequency Range Multi-UnitPRA Frequency Range (Events per Yearj 10E-06 to 10E-05 10E-07 to 10E-06 10E-08 to 10E-07 10E-09 to 10E-08 10E-10 to 10E-09 10E-11 to 10E-10 10E-12 to 10E-11 10E-13 to 10E-12 'Aumber of Sequences 299 ",817 9,240 1,030 13 Percentage of CDF 17 31 Number of Sequences 1,071 2.199 Percentage of CDF 14 35 38 13 'The number of sequences in this range may be reduced by truncanon. No initiator was considered with a cutoff less than 1.0E-10. ~ Not determined. >TVAW0038XlOC.04/04/95 1-12

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Browns Ferry Multi-UnitPRA Main Report Revision 0 Table 1-5. Browns Ferry Unit 2 Important Systems System Residual Heat Removal Service Water Svstem PRA Importance"'luiti Uait ~ Unit Z PRA o6i I 009 Diesel Generators Residual Heat Removal System 250V DC Battery Boards Emergency Equipment Cooling Water System High Pressure Coolant Injection System Reactor Core Isolation Cooling System Reactor Protection System Shared Actuation Instrumentation Main Steam System Including Turbine Trip Standby Liquid Control System Control Rod Drive Hydraulic System Condensate and Feedwater System 0.40 0.38 0.21 0.12 0.09 0.08 0.07 0.04 0.04 0.02 0.01 0.15 0.2 0.51 0.07 0.06 0.2 0.07 0.08 0.04 0.04 .01 ~Fraction of CDF associated with sequences in which the failures occur in the indicated system. ~ Less than 0.01. hTYAW0038.DOC.04/08/95 1-14

8 b0p S 50th BROWNS FERRY 45E 06 UNIT2 REV. 1A PRA TOTAlCORE DAMAGE FREQUENCY 50th 1.5E-05 SROWNS FERRY MULTI-UNITP RA TOTALCORE DAMAGE FREQUENCY MEAN 7.6E-OS MEAN 2.8E-05 5th I.SECS 95Ih 95th 4.5E-06 2'3E 05 62E 05 1E-7 1E-6 1E-5 1E-4 ) E-3 ) E-2 FREQUENCY EVENTS PER YEAR Figure 1-1. Total CDF for Brogans Ferry Multi-Unitand Unit 2 Rev. IA I'RAs

8 W b0p2 MULTI-UNITPRA TOTALCOP = 2.8E-05 UNIT 2 PRA TOTALCDF = 7.6E-06 TRANSIENTS WITH REACTOR LSOLATED1k INTERFAGING SYSTEMS LOCABcc1 TRANSIENTS WITH REACTOR NOT ISOMTED8% SUPPORT 8YSTEtl FAN.URES 2115 STUCK.OPEN RELIEF VALVES1% LOSS OF OFFSITE POWER 3$% LOSS OF COOLANl'CCIDENTS 2% LOSS OF OFFSITE POWE 2'RANSIENTS WITH REACTOR ISOLATEO 25% 1RANSIENlS WITH REACTOR tlOT ISOLATEO 2511 ItITERNALFLOODS 15% LOSS OF COO! Alit ACCIDENTS 7% S1UCK OI'ENlittILI'AIVLS 3 7s SIIPPORT SYSTCM FAllIIALS2. INTERFACIIIGSYSILLIS I OCns ~ i.. Figure 1-2. Browns Ferry CDF by Initiating Event Category}}