ML17321A192

Forwards Westinghouse Evaluation of Impact of Reduced Auxiliary Feedwater Flow for Unit 1 & Results of Reanalysis Performed for Unit 2.Feedwater Sys Acceptable for Both Units

ML17321A192
Person / Time
Site:	Cook
Issue date:	10/30/1984
From:	Alexich M INDIANA MICHIGAN POWER CO. (FORMERLY INDIANA & MICHIG
To:	Harold Denton Office of Nuclear Reactor Regulation
References
AEP:NRC:0300H, AEP:NRC:300H, NUDOCS 8411060158
Download: ML17321A192 (26)

Text

REGULATOP INFORMATION DISTRIBUTION TEM (RIDS)

P ACCESSION NBR:8411060158 DOC ~ DATE: 84/10/30 NOTARIZED; NO DOCKET FACIL'!50 "315 Donald C, Cook Nuclear Power Pl anti Unit 1i Indiana 8 05000315 50-316 Donald C, Cook Nuclear Power Pl anti Unit 2~ Indiana 8 05000316 AUTH INANE AUTHOR AFFILIATION ALEXICH,M,P~ Indiana 8 Michigan Electr ic Co, RECIP.NAME RECIPIENT AFFILIATION DENTONiHiRe Office of Nuclear Reactor Regulationi Director SUBJECT; Forwards Westinghouse evaluation of impact of reduced auxi liat y feedwater flow for Uni t 1 L results of reanalysis performed for Unit 2. Feedwater sys acceptable fot both units, DISTRIBUTION CODE: A001D COPIES RECEIVED :LTR ENCL SI E!

TITLE: OR Submittal: General Distribution NOTES: 05000315 OL! 10/25/74 05000316 OL:12/23/72 RECIPIENT COPIES RECIPIENT COPIES ID CODE/NAME LTTR ENCL ID CODE/NAYiE LTTR ENCL NRR ORB 1 BC 01 7 7 INTERNAL: ADM/LFMB 1 0 ELD/HDS3 NRR/DE/MTEB 1 NRR/DL.DIR NRR/DL/ORAB 1 0 N

" ETB NRR/DSI/RAB 1 1 F 0$

RGN3 1 1 EXTERNAL: ACRS 09 6 6 LPDR 03 2 2 NRC PDR 02, 1 1 NSIC 05 1 1 NTIS 1 1 TOTAL NUMBER OF COPIES REQUIRED: LTTR 27 ENCL 24

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IN DIANA 8 N ICHIGAN ELECTRIC CON PAN Y P.O. BOX 16631 COLUMBUS, OHIO 43216 October 30, 1984 AEP:NRC:0300H Donald C. Cook Nuclear Plant Unit Nos. 1 and 2 l Docket Nos. 50-315 and 50-316 @/

License Nos. DPR-58 and DPR-711 REVISED AUXILIARY FEEDMATER FLOM ANALYSES

~~@9 her. Harold R. Denton, Director ~o~oa 1ggyg Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission mc Mashington, D. C. 20555

Dear Hr. Denton:

Indiana 5 Michigan Electric Company (IldECo) letter No. AEP:NRC:0300C, dated November 3, 1980, provided information with regard to the flow design basis of the Donald C. Cook Nuclear Plant auxiliary feedwater system. It has recently been discovered, however, that an error was made in the input to the calculations supporting that earlier letter. The discrepancy has now been corrected, and the results of the revised flow calculations are expected to be transmitted to the NRC in the near future.

It is noted, however, that the revised calculations predict that for a feedwater line break in Unit 1 (for which the feedwater line break was not part of the original licensing basis), approximately 365 gallons per minute (gpm) of auxiliary feedwater flow could be delivered to the intact steam generators prior to operator action; the corresponding auxiliary feedwater flow value for Unit 2 is 375 gpm. Since these flow values are lower than those used in the Final Safety Analysis Report (FSAR) for Unit 2, we have contracted with Mestinghouse Electric Corporation (E) to perform a reanalysis of the feedwater line break accident. g's evaluation of the impact of reduced auxiliary feedwater flow for Unit 1, and the results of the reanalysis performed for Unit 2, are presented in the Attachment to this letter.

Based on the reanalysis performed for Unit 2 with the LOFTRAN computer code, g has concluded that the plant meets the FSAR acceptance criteria assuming the auxiliary feedwater system delivers 375 gpm to the intact steam generators. More specifically, the analysis indicates that the Reactor Coolant System (RCS) and main steam system will remain below 110$ of their respective design pressures, and that the RCS water level will not drop below the top of the core. Ther efore, core integrity would be maintained and the radiological consequences would be only a small fraction of the 10 CFR 100 guidelines.

84.1030

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Mr. Harold R. Denton AEP:NRC:0300H The K evaluation performed for Unit 1 indicates that operator action (i.e.,

action 81) at ten (10) minutes following a postulated feedwater line break would restore heat removal capability. This heat removal capability would be provided via an incr ease in auxiliary feedwater flow to approximately 460 gpm, assuming the availability of control air to the emergency leakoff valves.

Operator actions taken in accordance with actions g1 and g2 of the g evaluation will ensure auxiliary feedwater flow well in excess of 060 gpm (approaching approximately 1200 gpm at the first safety valve setpoint). It is noted that although the feedwater line break is not part of the original Unit 1 licensing basis, the licensing basis does assume operator action at ten (10) minutes following a high energy line break for the purposes of maintaining containment integrity, With this action, g believes that the consequences of a Unit 1 feedwater line break would be bounded by the conclusions provided in the FSAR of plants similar to Cook Plant.

Additionally, it is noted that IMECo has taken the conservative step of temporarily reducing Cook Plant thermal power for both Units while the analysis and evaluation was being performed by W., The enclosed information from gl indicates that the licensing basis for Unit' was not exceeded and, for Unit 2, the public health and safety would not be endangered by a return to full power operation, based upon an assumed auxiliary feedwater flow of 375 gpm to the intact steam generators following a postulated feedwater line break. We expect to return Unit 2 to full power operation upon concurrence of the staff of the Office of Nuclear Reactor Regulation. We believe the attached evaluation Justifies the return of Unit 1 to full power operation.

This document has been prepared following Corporate procedures which incorporate a reasonable set of controls to ensure its accuracy and completeness prior to signature by the undersigned.

Very truly yours, P. Al xichg'U'ice President MPA/dam cc: John E. Dolan W. G. Smith, Jr. - Bridgman R. C. Callen G. Charnoff NRC Resident Inspector at Cook Plant - Bridgman G., Bruchmann

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Mr. Harold R. Denton AEP: NRC:0300H bo: J. G. Feinstein/P. A. Barrett/H. Y. Fouad'/D. A. Medek H. N. Scherer, Jr.'.

J. Markowsky T. 0. Argenta S. H. Steinhar t/J. A. Kobyra R. N. Jurgensen J. B. Shinnock R. F. Kroeger T. P. Beilman - Bridgman J. F. Stietzel - Br idgman,, ,,

D. L. Migginton - NRC DC-N-6015. 1 DC-N-6110

T C SE C 0

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OCT.29 jj. 6 Srs id'usioaer

'84 18:88 ASTI WCWIRE RON AWS- 162 Refereree Ko(s) ..

P. 82 o

Qast<nohous>> Reference No(s).

{Change Control oe'Fg As Applicable) lKSTINNXSE INaZW SAFrrY EVALVATIN CHECX LIST

1) XUCaAR ~rr(S)

2) CRCK LIST APPLICAN.E TO-'t of Red d AFM Flow on thetfcal Feedl)ne Braik Trans]ent at D. G. Cook Unht

3) The ~tten safety evaluation of the reHsed procedure< design change or aodhfkcathon required by 10CFR50.59 has been prepared to the extent required and 3s attached. If a safety evaluation $ s not requIred or fs kncoeplate for any reason, explain on Page 2.

Para A and 8 of this Safety Evaluation Check 11st are to be caapleted only on the bashs of the safety evaluat(on perfonred.

Cmr. LIST - War A (3.1)

(3.2)

Yes Yes No Na

~

~

A change A

to the plant change to procedures as described <n the FSAR2 as described $ n the FSAR2 (3.3)

(3.4)

Yes Yes No No

~

~

A test or.experf~t tet described $ n ihe FQR?

A change to the plant'technical spec1flcathons (Appendix A to ihe Operating Uceme)2

4) CHKCX LIST - PART B (Justtffcat<on for Part B answers ~t be {ncluded on Page 2.)

=

(4.1) Yes No x H11 the probabflf+ of an accident prev)ously evaluated 3n the FSAR be fncreased2

'It 22 Y 2 ~tillthe ti wwNIUNc FSAR he fncreased2 H tt I ~i ly I t it (4.3) Yes  % ~ Nay the possibility of an accident A5ch $ s different than any already evaluated $ n the FSAR be creatad2 (4.4) Yes No X All the probabllfty f a aalfunctlon of equipment $ aportant io safety ~ousb'valuated <n the FSAR be All the consequences of a ealfunctlon of equ<paent 1aportant

)ncreased3'4.5)

Yes Na X to safety preHously evaluated $ n the FSAR he fncreasad?

(4 6) Yes Na X +g the poss~b<14ty Df a salfLecMon of equ$ yaent <sportant to safety different than a~ already evaluated $ n the FSN be created2 (4.7) Yes No X N11 ttle mrgfn of safety as defined $ n the bases to any tedefcal specfficat<on he reduced2 Page 1 of 2

WMI RXN P. 83 OCT.29 '84 18si2 MCGUIRE If the and answtm to any of the above queat$

explain below.

ons are ~~, ~~~~t ~~) ~

If the area+ to any of the abve ~t<on<<n 4) camet b adam~ ]n Mttln safety haslcl Nl tion for 1$ c40$ 0 evalg4tgon> QQ sssfllh+nt sg~tt4ad to gg pggs~nt to le~ 90 W foll'gs~H>ls th 3Mstfffcatfon Mpon the ultan safe~ Ova]oat~on (l) for Liars given $ n Part 5 of the $ aga+ Evaluation C~ List See letter NS-RAT-PTA-84-111

)Raference to document(s) conta<n1ng rattan safety evaluat<o:

for/Description 4con Pave(s): Table(s): F~gMr (s):

mason of Change:

NOT APPLICABLE

~red

'oo&fnatecf hy (Nuclear Rth Safety):

Engineer (s):

bate: l~

Date: l0

~fnat<ng Croup Nenaeer(s): Date: /~ +2 vclear Safety 6roup Nanager: Date:

Paoe 2 nf 0

OCT.29 '$4 18:15 ASTI MCGUIRE RKN P.$ 4 NS-QKS-OPl-N-113 0-BhT-PTLSl-111 OCKBKR 27, 154 3); DoPo XNZNZZS L 8JDk RE: Impact of Reduced Ahf Flew on FeedlirM; Break Transient for D. C. Cook Unit 1.

~can Electric Powe'nfused Mesti~xese of an error in the &uxilia~

feecheater flat calculations for the feedlira break analysis of D. C, Cook Unit

2. Ttds error w&s faed to apply to both mits; however, Unit 1 does not have a feedlirN break analysis as part of its licensing basis and ezplicit csloulatices are not required. The fallming provides & safety evaluation of what action waQd be accessary to Mtigate the consequences of & hypothetical feecGim break at Unit 1i

?he D.C. Cook Unit 1 design basis for operator action fallcadng a high energy 1ire break is 10 mrinlrtes (of. AEP--). Thus credit for the operators to irritiate the appropriate actions at 10 aLnutes is justified.

Should a feedlire break occur at Unit 1 the reactor protection system would respond as fciLlcws:

?be reacta'c+id trip on a steaa generatcc'ow water level ooirNident with stea/feed water f1'ismatch a stae generator low-lac water level.

Hen the stean generator 1~lcm level is reached in aa stein general',

the aotor driven auxiliar7 feed pmps (NP) start. ?he worst single failure for a feecRim break is a failure of the NP which feeds two intact stean generators. FollarLng reaotcl'rip and subsequent turbine+ trip, the lcM lcm'ater level in a second stein generate would be reached and the turbim driven auxQiary feed paap (?DQVP) is starMd. Ls the steaa generators depressurize, the steea supyly to the ThLMP is reduced and eveatua11y la4. Thus caoe ice stean supply to the TMPMP is loot, cnly a ainiaal Ilant af auxQiar7 fee4rater fXar wculd be ave,able fa. heat pcRcv &1 a Operator &ation should be perfcmed to restore beat remcrval capability, i.e.

450 gpa to the intact stean generators, kraQ&ble operator actions are as fciLlms:

1. He oan Relate kFbl flee to the faulted SG and defeat the Q.car retention logio for the intact steaa generahx fed by the sane MDP. This results in orM, NP delivering 4R gpn to the intact stean gener&ter. This action is SiaQar tO actiOn required by other ylantS Sindlar tO D.C. COOk Unit 1 that here a feeQine br'eak analysis as part af their'esign basis.

2. He can restore stean supply to the TMRfP. This oan be eccccnplished ty closing the aein steaQ.inc isolatioe valve ce the faulted stean the NSIVs for the lires that connect to the TDQVP.

general'r

OCT.29 '84 18:i8 ASTI NVCWIRE RQN P. 85

</y Mith the asempt;ion of action 1 above et 10 minutes, Westinghouse believes the feedlfm break coreequenoes would be bounded by the eonolusions (RCS pressure less than 130S of the design pressure, core integrity maintained, and dose releases are a saall fractXea of the 10 CFR Part 100 guidalitss) provided in the FShR of plots s~iar to D. C. Cook stree the D. C. Cook hht syeten design assures iN flat to the intaot steaa generators price to the 10 alake operator aotion time. Thus provided operator action results in an adequate beat sink, Mestfn~e sees no safety issue regarding the reduced hN flex fcc' hypothetical feecQ.fre break.

Di S QXE Pl>et Trinsfent hnelysis Operating Plant AmQysis K. P. M855E, Manager P. A. , Ravager Pleat Transient kaQysis Operating Plmt inalysis

OCT.29 '84 i8:2Z ASTI ~IFK RXN P. 86 MS-RAT-P7h-Sl-112 FRN: NS PTh MIN 2%~5 DATF.l 0 tober 29 1884 KIBJECT: D, C. COOK UNIT 2 (AMP) FHUX INK BKAK REhRhLYSIS D. P.

4 Daafnicis P. Suda MHC ~

701- 25 CC: A. L. Sterdis Lofts NC ~

NC 0-09k P. A J. L. Little J. Parvtn NNC ~k Mesti~ouse Site Malaya D. S. Love MHC 0-09h D. C. Richardson HNC 0-15 Please find attached the report docmenting the feedline br eak reanalysis assuming ~5 gxn of auxiliary feeckater flex to the 3 intact stean generators fa D. C. Cook Unit 2. As sham in the attached, the FSAR acceptance criteria of no substantial overpressurization of the reaehcr coolant systan and the core readns covered are met, Sbould ycv have any questions the mdersigned may be contacted.

This expedited effcrt was charged to sbop order AKKP-485.

M. R Adler Hant Tmnsient kalysis Approved:

tl. P. Osborne, Hanager Plmt Transient kalysis

OCT.P3 '84 i8:25 &ST OJSE WCWIRE RON P. 87 OCT.29 '84 18:28 WST t&CWIRE RXN P. 83 The PBkR feedlira break analysis fcr D. C. Cook Unit 2 was perfumed assming

%0 gpn of auxiliary fee%ater (hFM) flat is delivered to the intact steer.

generators. The follcwing demonstrates that the plant meets the FBAR acceptance criteria (described on ~ge 14.2.8-2 of the FSAR) assuming the LN aystan delivers 375 gpm to the intact steai generators.

-This analysis was perfcrmed using the same methodology and assuagions as stated in the FShR with the foDcwing exceptions, The LOFTS computer code was used, Zhfs code simulates the g.ant thermal kinetics, reacta'oolant system (RCS) including natural circulation, pressurizer, steaa generators, and feechrater systen, and computes pertitant variables including the pressurizer pressure, pressurizer ~ater level, ard reactor coolant temper at<'re.

Reaotcr trip and auxiliary fee%ater initiation are assuaed to occur on l~lcac water level in the faulted stean gener ator.

The lnost restrictive singe failure ih the AN systaa is assLzoed. This is the loss of the motor driven aw&iary feeckater pmp that rem:.My wculd have supp'.iei.'. fee&ster to taro intact stean 'eterators in one a.inute faDcwing reactor trip. The other tIotcc driven auxiliary fee4tater puap and the turbid driven auxiliary fee&ater pmp deliver a total of 375 gxn to the three intact:tean generators witt.i>> one migrate follming the resets trip.. Ln additiorel 280 is assessed befcc e the feed liras are purged and the relatively coldseconds (1K F) auxiliary fee%ater enters the intact stean generators.

Figures 1-7 Qluatrate the key plant paraneters calculated follcwing a fmdline rupture. The calculated sequence of events are listed in Tah1e 1i Results presented in Figures 4 and 6 sheet that pressures in the RCS and main stean systen ramdn belcw 11(5 of the respective desig r.ra sure.. Fressuri~u pressure irerease util reactor trip ce i~lcm stean generator level.

Prmsure then decreases, due to the loss of heat input, LatD steailine isolation occurs on lcm'teaQine pressure in the faulted loop. Coolant expansion occurs due to reduced beat transfer capabGity in the steau 6enerators; the [~surizer ~ety valvule open to maintain primary coolant systan pressure at an acceptahle valw.

OCT.29 '84 i8:32 (ASTI ~RE NXN P. 89 Re reaata'ore ranains covered with water throughout the transient, as water relief due to themal ~asion is limited by the beat ranoval capability of 4e avxiliary frater ayatan.

Kx a~steec'uxiliary fee&ster flm rate is capable of reaovi~ all af the decay heat "3200 seconds after reactor tzf.p. After this ~, core decay heat 4wreases bales the auxiliary frater hes5 ranoval capacity and reactor ooolaat tan~ra~~es and pressure decrease, Results of the analysis axe that fcr the postulated feedlim rupture, the assigned auxQiary feeAater aystca~ capacity (375 gpn) ia adequate to r above

~'.ecay heat and to prevent the water 3 arel in the RCS from dropping to the top of the core, REFQKHCE 1: Bunatt, T.M.T,, et al., 4LOFQUN Code Deso<ption,"

VQP-7g07-k, LprQ, 19%.

TABLE 1: TQK SEVEN(X OF EVEh'IS EVERT TDK Feedline R~ture Occurs 10.0 Reactor trip on L~Lcv 15.4 stean generator water level Rod motion and power lost 17 4 to the reactor coolant pmps Auxiliary feehrater is de'ivered 75.4 to 3 intact stean generators Lm stem'ra pressure 149.4 setyoint reached Steanl ine isolation occurs 157.4 Pressurizer safety valve 612.0 setpoint reached Stean generator safety va've 700.0 setpoint reached ir intact stean generators reached Core decay heat decreases to 3200 aux'l iary feeOcater heat removal capacity

P. 18

'84 18:35 ASTI t&CWIRE FKXN OGT.P3 S,use 00Ã0 g

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<@iNij N85$ g TOE Fl.N VERSUS TINE FIGURE 1: HAIR FEEDLINE RUPTURE ACCIDENT CORE HEAT

OCT.29 '84 i8:38 (ASTI WCWIRE NXN ii e 4Ã04 g

40

~ ~ ~ ~

0 gggQJ m gpss FIGURE 2: MAIN FEEDLINE RUPTURE ACCIDENT REACTOR COOLANT FLSt VERSUS TINE

~ 29 '84 i8 4Z '4ESTI NVCWIRE RCXN P. 0 f85~

QSOe 750 L00 I

ae 0 44~i a NsSig NASL ORO

'IGURE 3: NIN fEEOt.IHE RUPTURE ACCIDENT PRESSURIZER ltATER VOL!9K VERSUS TIKE

OCT.29 '84 i8:45 ASTI l&CWIRE RCKN P. 13 NN,O Q. O

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FIGURE 4: HAIN FEEDLINE RUPTtif ACCIDENT PRESSURIZER PRESSURE VERSUS TIME

sat P ~

m.a 85 l0 HOT I~ ~N 8 St$ e N COLD UJ I gg}

RLN me 0% N sat, I~ NLm Tm ML Vl

%% N TCOLO MN 4 ec~lcg 5 MNg ~ jQ+

FIGURE S; MAIN FEES.IHE RUPTURE ACCIDENT FAULTED AND INTACT LOOPS RCS TEMPERATURES VERSUS TIHE

'84 18:52 ASTI t&CWIRE KXN P,15

~

- OCT.29 tNOi0 INTACT YEAH GENE RATORS i%i.N y

FAlKTED STEN GENERATOR

~ i 4~ NN5gfg me g~

aalu

~ ~ ~ ~ a FIGURE 6: HAIN FEED.INE RUPTURE ACCIQEHT STEAM GENERATOR PRESSURE VERSUS TI%

l&CWIfK

~.29 '84 i8:55 ASTI ROC'NTACT STPN GfHERATOR$

FAULTEQ STQg 'GEgEy LD FIGURE 2: MAIN FEEDLINE RUPTURE ACCIlKNT STEN GENERATOR NSS VERSVS 75%

Distribution CDocket fileiwo/encl.

ORB81 Rdg. wo/encl.

CParrish w/encl.

DWigginton w/encl..

October 25, 1984 DOCKET NO(S). 50-315 and 50-316 Mr John Dolan, Vice President Indiana and Michigan Electric Company c/o American Electric Power Serwlce CorporOihon 1 Mverside Plaza

SUBJECT:

Columbus, Ohio 43215 Co COOK NU(XZAR PLANT,

'ONALD UNITS 1 AND 2 The following documents concerning our review of the subject facility are transmitted for your information.

IC Cj Notice of Receipt of Application, dated D Draft/Final Environmental Statment, dated D Notice of Availability of Draft/Final Environmental Statement, dated CI Safety Evaluation Report, or Supplement No. , dated Cj Notice of Hearing on Application for Construction Permit, dated C3 Notice of Consideration of Issuance of Facility Operating License, dated C3 Monthly Notice; Applications and Amendments to Operating Licenses Involving no Significant Hazards Considerations, dated C3 Application and Safety Analysis Report, Volume CI Amendment No. to Application/SAR dated D Construction Permit No. CPPR- , Amendment No. dated O Facility Operating License No. , Amendment No. , dated.

CI Order Extending Construction Completion Date; dated EK Other (Specify/ Monthl Notice aoverin eriod -throu h October 24 1984 iration date for hearin re uests and comments Novdmber 26, 1984.

~ \

Division of Licensing Office of Nuclear Reactor Regulation

Enclosures:

As stated cc w/enclosure oui>cm> ORB/I 1:DL suRNA'ML~ CParrish/ s DATE+ 10/ /jI/84 NRC FORM 318 {1/84)'NRCM 0240

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