ML20236F187

Safety Evaluation Supporting Amends 73 & 65 to Licenses NPF-2 & NPF-8,respectively

ML20236F187
Person / Time
Site:	Farley
Issue date:	10/26/1987
From:	Office of Nuclear Reactor Regulation
To:
Shared Package
ML20236F171	List: ML20236F169 ML20236F187
References
TAC-62283, TAC-62284, NUDOCS 8711020035
Download: ML20236F187 (9)

Text

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1. o UNITED STATES f.

g NUCLEAR REGULATORY COMMISSION y

7, ij-WASHINGTON, D. C. 20556

%***o SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 73 TO FACILITY OPERATING LICENSE N0. NPF-2

.AND AMENDMENT NO. 65 TO FACILITY OPERATING LICENSE NO. NPF-8 ALABAMA POWER COMPANY

.[

JOSEPH M.'

FARLEY' NUCLEAR PLANT, UNITS 1 AND 2 Ij

.I JDOCKET N05. 50-348'AND 50 364 1

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1.0 INTRODUCTION

j By letter dated August 25, 1986, superceded June'2, 1987, supplemented-September 16,'and.23, 1987, the Alabama Power Company:(APCo, or the j

licensee),: submitted a request for changes to the Joseph M. Farley. Nuclear Plant, Units 1 and 2, Technical Specifications. This amendment request was noticed on October 7,1986(51FR36082).andJuly. 15, 1987.(52 FR 26582).

.The supplements did not change.the' amendment requested or the determination f

noticed; therefore, the amendment was notirenoticed a third time-The Amendment would revise the Technical Specifications '(TS) to allow an increase in,the allowed steam generator-(SG) tube plugging limit from 5%

to 10% and an increase in the Heat Flux ~ Hot Channel Factor-(F );1imits.

Power (RTP)gewasfrom2.31to'2.32'forgreaterthan'50%; Rate 9 T The F chan 9

and from 4.62 to 4.64 for less than or' equal to 50% RTP. The-licensee had previously'provided a sumbi.ttal,tdated August 25, 1986, which contained proposed TS changes and explanations'of why the effects of the-

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proposed changes on plant transients would not jeopardize safe, operation'of the plant. However, the' staff advised the licenseeLto provide a reanalysis of the emergency core cooling. system (ECCS) analysis for Farley' Units 1 and 2 that supports the large break ~1oss-'of-coolant accident (LOCA)'with a:

l corrected BART code methodology.- By letter dated June:2, 1987', the licensee superceded the August 25, 1956 submittal and addressed the concerns.

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2.0 BACKGROUND

Farley Nuclear Plant currently has a steam generator tube plugging.(SGTP) limit of 5% as shov on TS Figure 2.1-1.

This limit is'. based on the Large

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Break LOCA/ECCS analysis in the FSAR'Section.15.4 which. assumes a 5% SGTP 1

limit. Approximately 2.9% of the steam generator tubes lhave been. plugged l

in Unit.1 and approxima.tely 3.7%-of.the steam generatoritubes have been.

l plugged in Unit 2. This level of SGTPJincludes a11~ r'ow:1 tubes Lin each h

steam generator, which were done as a precautionary measure by the' licensee.

Based on the degradation identified during the last Unit >2. inspection,lthe-expected tube plugging during the October 1987 refueling outage coulu. exceed the current limit of 5%.

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-Therefore,:APCo-hasproposedtherchange?to). increase lthisteam;generatoN N. 4 Etube plugging limit to-10%.to provide. additional marginito;thenlimit (

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4The licensee's originaP submittal,' dated AugustL25,;1986',Jalso pro-1

vided;a revised;ECCSlanalysis for Farley-Unitstliand'2h ; Changes?inL Ag

~ the analysis' assumptions' included. an increase.:inlSGTP from 5%ltoi10%R f

and:an increase in F Efrom 2.31bto 2132 forLRTPlaboy'e 50%-and an d

increase.inF-fromk.62to4'.64L'apenaltyimposed.bytheNRCagainst)

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for RTP?off 50%(or less. The lower;

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O values forlF were the, result of c

.the 1978 ver91on of the Westinghouse ECCS Evaluation Model.

We:were:

"d informed by the'licenseefthat'thelK(z) value..had originally lbeen nval.

N uand for a?F ' of12.32L - This' value remained the same'with the,. previous.

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notLrequi.re.a~Ohange?forLK(z). The present model,LBARTeduc v

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no11onger.

4 requires a penalty.(.However 1 Westinghouse had recently. identified

-nonconservative.-assumptionssin-the1BART modeltregarding theleffectsLofJ, e q

control rod thimble filling during reflood;and hot-assemblyfpower.s

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effects../ Reassessment'of the overall4BART model conducted by;...

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Westinghouse.and described in WCAP-9561-P-A.MAddendum 3,; determined; that other conservatism contained in BART compensate for' the-nonconser--

vative thimble filling.and hot assembly assumptions k However, the staffi q.?

-required thatLlicensing actions"be. supported by a'reanalysisLusingtrevis'ed- ' l versionsoftheBARTcode.(WCAP-9561, Addendum.3,-.Revisioh1)1tocomply1 q

with'.10 CFR.50.46.

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Accordingly,;APCo resubmitted the proposed changesito;theiTS in a' 0

submittal'. A new large Break LOCA: analysis was performed byn

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letter dated-June;2, 1987,.which. supersedes)the August'25,(1986.

i Westinghouse for the Farley Nuclear Plant utilizing the'1981".1.

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EvaluationModel--(WCAP-9220-P-Aand'WCAP-9221)with; BASH:(WCAP-10266b Revision 2).

The use'of the BASH methodology has"beenJapproved by<

0 the: NRC' staff in a. letter dated November 13,1986,; from'.Mr.! Charlesi E. Rossi to Mr. E. P. Rahe, Jr. (Westinghouse).. ;In a3teiconiofLAugust:20,-

a 1987, the licensee reaffirmed that:WCAP-9561, Addendum 3,sRevis~ ion 1,_.

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x was used'as required in.the evaluation' 'The NRCDstaff-evaluation'follows..

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q 3.0 EVALUATION Large Break LOCA The analysis toineet7the< requirements of4 Appendix 'K and?10 CFR150.46, 4

for Large Break'1.0CA was performed using theLWestinghouse Evaluation modelwithBART-1A(WCAP-9561-P-A,1984-(Proprietary))andBASHJ (WCAP-10266,_Rev. 2 with Addenda,'1986'(Proprietary)) for/aispectrum.

q of break. coefficients. iSubsequent;to the' completion?of. the.Farley -

o Large Break:LOCA analysis with BASH, Westinghouse: notified Alabama Power Company of enhancements >to theTBASH code and methodology 1 that were'made to improve the reliability and! performance 1of the code H

-in.certain circumstances. The modifications 5to'the? BASH methodology

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which incorporate these enhancement'sL(describedLin4 Addendum:2 to?

WCAP-10266 Revision 2)'were submitted"to the1NRClinlletter..

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.NS-NRC-87-3212,datedMarch26,L1987L'Thisitopicaljreport?hasL L

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. beeni reviewed' by the ~ staff an_d. is approved 1for: application" toJhr,leyL Units :1L

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'and 2. CAPCo"together with Westinghouse:has-evaluated thelimpactiof the

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t BASH code modifications ~ on !the' Farley Largef Break 1LOCA;analysisiwith BASHiin.i 4

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, *.tosthe' June 2 19870 APCo letter and?has, concluded the: Farley; w

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'analysisiremains. con _servative. andLboundi. g.'

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'9 The new Large Break LOCA'an.alysistassumesianLF 'of;2240.;(Th'ehpresent 9

to 50% RTP was ' equired as ?a.resultLof penalties asses' sed byithe: NRC(against" y

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.the'.1978. version-of theLWestinghouseLECCS Evaluation Mode 1 L 4

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iSince the currentJSma11LBreak;LOCA analysistassume's1AnLF Lof;2.32 andsthei proposed increase 'in F'iis 9

in 2 non-LOCA transient anakysesconservative1ysboundedibyLthe assumptions

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, the proposed changes 6to(increase.thelF j

feoefficientsofLTS.3.2.2Lto2.'32forgreaterthan50%;RTPJand'4.6430r$1 esse S

L cthan 'or' equal,to 50% RTP are consistent with"the design /1_ic'ensing-basis;for)

Farley Nuclear Plant.,

The'fuelsparame'tersusedas-inputifortheLI.0'CATanalysiswerejgeneratedLusing(

M the Revised PAD ThermalLSafety Model, WCAP-C720,' Addendum?2,uwhich weiapproved:

by letter from.C.0. Thomas (NRC)- to E; P. :Rabe,;Jr. '.(Westinghouse)', dated December 19,-

w1 1983. The hydraulic analyses and core' thermal transient;analysesiforithe' Joseph 1 4

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M. Farley Large Break:LOCA analysis;were7 performed?using"102tpercent of L11censeds m.*

NSSS core power, 2652 Mwt'.,

0ther pertinent. assumptions. included:a"-10% SGTPD y

level, minimum and maximum safeguards'ECCS O

Westinghouse fuel design, which is1thh cur l capabilities 017y 17/standardL rent.designefor'both Farley) units,z L

and'an upflow barrel-baffle configuration. ;The upflow barrel-baffle; L

configuration was previously -shownito' represent a;smallfpeakicladJtemperatureL.

M (PCT) penalty; hence, the use of. this configuration :is' conservative'and bounding ?

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on both units. This analysis.also incorporated 'a conservative total reactorc ~

l coolant system flow (1% below TS limit).KPertinent input { parameters areilisted'

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below:

INPUT =PARAMETERSi

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NSSS Power, MWt, 102% of licensed power.

12652 12.49 Peak Linear Power, kw/ft,102% of: design -

Peaking Factor-(At: Design Rating)

Hot Channc1'Enthalpy Rise Factor'

~2.'40 N

1.62c w Accumulator Water Volume d

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_(Cubic Feet per Tank)

'y 21025.0-i Accumulator. Pressure, psi 00.0

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Pumps Operating:(Min ECCS/ Max ECCS) 2/3 l

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Steam Generator Tubes Plugged 10%.(uniform),,

• Minimum' safeguards analysis assumes;2 charging pumps'are operating; J

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and one RHR pump. Maximum safeguardscanalysis assumes threeL charging pumps and two'RHR pumps are; operating.;

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Uniform 10% Steam'GeneratorhTube Plugging assumes _10% SG tubes plugged in each steam generator:and: corresponds to the worst plugging.

level _~inLany steam generator = andfwill ~ bound all: combinations' of :

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x non-uniform plugging as long' as no one; steam ge'nerator; plugging ;

M" level exceeds 10%.'

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Of the three break sizes evaluated, C = 0'.4, Cn=l0.6,~and C = 0.8, the C = 0.4 l

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breakwithminimumECCSsafeguardsprovedtoibsthelimitingf(highestPCT)

The resulting peak clad temperature was '2013*F, which is well' below1the case.

2200 F' allowable limit.

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APCo presented the following conclusions from'their analysis which demonstrate that for breaks up to and. including the double; ended severance of a TreactorL coolant pipe, the ECCS design at'Farley Nuclear Plant.will, meet the acceptance criteria as' presented in 10 CFR 50.46. These'are:'

1.

The calculated peak clad' temperature does not exceed'2200 F based on a

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large break LOCA total peaking factor'of 2.40 and a hot channel enthalpy rise factor of.1'.62.

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The amount of fuel element cladding that reacts chemically with water or l

steam does not exceed 1 percent of the total amount of Zircaloy in thel j

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1 reactor.

3.

The clad temperature transient-is terminate'd at,a-time when the core geometry is still amenable to cooling.

4.

The cladding oxidation limits of 17% are not exceeded dur.ing or after..

quenching.

5.

The core temperature is reduced and the decay heat is removed for an j

extended period of time, as ' required by the long-lived radioactivity.

y remaining in the core.

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Small Break LOCA As previously stated, APCo requested an increase in SGTP. limit from 5% to 10%.

In justifying the increase to 10% tube p1'ugging, the licensee stated in the~

June 2,1987, letter that -there is evidence that for lowLsteam' generator plugging levels (up to 20%), Small Break LOCA transients' would not be affected by the proposed tube plugging limit. We questioned the licensee about the.

evidence.

In response, the licensee explained in lettersidated August 18, and l

September 16, 1987, their conclusion that for up to 20% tubeLplugging'there would be no affect in the Small Break LOCA analysis. This' explanation is based-l upon an evaluation performed in 1985 for'the Westinghouse designed.Almaraz plant which is similar to the-Farley. design. Both plants are of identical Westinghouse vessel design. Connon features include three coolant loops 157 Lfuel assemblies, standard fuel design (0.374 inch OD),'48' control. rods,144. inch active fuel l

length, and Model 93 reactor coolant pumps. A'few small deviations exist with-respect to operating parameters, but these are insignificant for comparison of response for a small break accident.

For example, reactor power is.2686 MWt for Almaraz; whereas, reactor power for.Farley is 2652;MWt. The licensee i

concluded that evaluations based on the Almarazl plant for' establishing trends L

and sensitivities are equally. applicable for, the Farley units.

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,5Three specific effects of.SGTP were' identified'by the-licensee and evaluated for

-the Almaraz plant as follows:-

1) the impact ofythe ' reduction;of the steam" generator; tube; area' on thel x

~small break ~ transient as'itl relates toitheLabilitf.toltransferJheatL.,

from primary to secondary;and,;thus,! dissipate core; stored energy an'd?

decayt eat, "

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2) the' effect of changes toioperating[ temperatures [(Nimary\\ arid secondary)[

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as aLresult:of.SGTP, and; t

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3) the effect that could betexerted ont the drainingtof the':steain generatory tubesias this has:a direct'effect"on waterfinventoryTin the4vesse1Landi a

. potential for coreLuncovery ;

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.small break-LOCA. These studies,1" Simulation?of Small. Break Type'Beha'vior ofL PUN and SPES using the NORTRUMP ' Code"f and " Limiting; Counter Current? Flow i

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Phenomenon:in Small-lBre'ak LOCA Transients;".were included:in >coceedings'of th'ej, t

Specialists meeting on Small Break LOCA: Analyses in LWR's, Pisa,JItaly. June.

.1985. It was concluded that no effect would be' expected in theLSmall Break' analysis for SGTP levels up' toi20%' for.the three' relevantiphenomena.-identified.

These phenomena-are sumarized as follows:-

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1).only a small portion of the steam. generator. tube!heatitransferiareal..-

H is sufficient'in~a'small break transient to:providefanLeffectiveDheat'

.1 sink to the primary side,

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2) operating temperature differences ~as a.resultiof plugging disappear?

shortly after the break because"the) secondary side pressure reaches; mj" steam generator safety; valve >setpointsjalmost immediately,;and.

3)thecountercurrentflowlimit;(CCFL)charact' eristics!would"be.-such.

s theJinclined! pipe' that the CCFL would still be dominant and limiting'.in;ilegEfor;SGTPE '

d connecting the steam generator: inletj plenu'm.to the. hot levels _up to 20%. -For steam: generator tube plugging levels beyond!20%,

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L a CCFL" calculation in steam generator tube locations would increase.

This would reduce the-dominance of2the CCFL in:thelinclined' pipe,ci.n which case the plugging level would exert an! influence; 7<

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The licensee stated that Farley's current Small Break LOCA analysis is based on j

the WFLASH code.

From the reference evaluation of 1985 for the Almaraz plant, they conclude that the effect of SGTP would not be seen in WFLASH analyses

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because WFLASH does not take credit for the CCFL phenomenon.

l Following the incident at Three Mile Island, Unit 2, Westinghouse and the Westinghouse Owners Group developed the NOTRUMP computer codes (WCAP-10079-P-A l

1 and WCAP-10054-P-A (both Proprietary), August 1985) as the new Small Break-LOCA evaluation model, which the NRC staff approved in May 1985, to meet the require-ments of NUREG-0737,Section II.K.3.30.

Small break LOCA analysis performed using NOTRUMP for NUREG-0737 demonstrated that, in general, the NOTRUMP evalua-tion model calculated lower peak cladding temperatures than the WFLASH evalua-tion model. This allowed the WFLASH analyses contained in the Joseph M. Farley Final Safety Analysis Report (FSAR) to remain the licensing basis analysis of record in accordance with NRC Generic Letter 83-35.

APCo concluded that the reference Almaraz plant evaluation is directly appli-

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cable to the Farley plants and that for a 10% SGTP there would be no adverse effect on the WFLASH small break analysis of record.

By projecting effects, if analyzed with NOTRUMP, APCo stated that minimal SGTP effects in PCT would be' expected to be observable at a leve' near 15% to as high as 20% SGTP.

But this would be insignificant compared to the PCT improvement that would be expected by applying the NOTRUMP Evaluation Model. Therefore, APCo concludes that the results of the Farley analysis of record continue to be bounding.

Based on the preceding evaluation, the NRC staff concludes that for the SGTP, as APCo requested (up to 10%), the proposed changes to the TS are acceptable and will not impact or invalidate the current licensing basis for the Small Break LOCA analysis as represented in the Farley FSAR.

In addition, the peak cladding temperature for the Small Break LOCA is 193'F (1820 F vs. 2013'F) lower than that for the Large Break LOCA; thereby, providing the NRC staff with additional assurance that the criteria of 10 CFR 50.46 are satisfied.

Reactor Coolant System Flow The licensee stated that an analysis was performed to determine the effects on the core flow due to the increase of 5% in SGTP, The analysis determined that the increase to 10% SGTP would not decrease reactor coolent system (RCS) flow, below the thermal design flow (TDF) for the Farley Nuclear Plant.

In response to NRC staff questions, the licensee stated that Farley TDF values for Units 1 and 2 are 265,500 gpm for each unit. The licensee also stated that the latest measured total RCS flow values were 283,963"gpm for Unit 1 and 285.767-gpm for Unit 2.

Since tha flow measurement uncertainty for the Farley Units is -

3.5%, these measured flow values correspond to actual flows of at least 274,024 gpm for Unit 1 and 275,765 gpm for Unit 2.

These values are well above the minimum flow requirement of 265,500 gpm in the TS. With'10% of the steam generator tubes plugged, the calculated RCS flow is 94,500 gpm per loop or 283,500 gpm total for Unit 1 and 93,800 gpm per loop or 281,400 gpm total for Unit 2.

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These values are based on best estimate flow caltalations. These flows are acceptable since all are above the minimum flow requirements.

For reference only, the licensee stated that the calculated RCS flow with 0% steam generator i

tube plugging is.289,200 gpm total for' Unit 1 and 287,100 gpm total for Uni t 2.

Non-LOCA Accidents The licensee stated that, since the non-LOCA departure from nucleate boiling (DNB) l transients are based on TDF, which remains applicable, a 10% SGTP limit was determined to have no impact on the non-LOCA DNB transients. The effect of 10%

SGTP upon those non-LOCA accidents which are not DNB related, or for which DNB is-I not the only safety criterion was also evaluated. The only accident of this group which is affected by 10% SGTP is the boron dilution analysis.

An input to the boron dilution analysis for Modes 1 and 2 is the RCS active volume,.i.e., the total RCS volume minus the volumes of the pressurizer, the pressurizer surge line, l

the dead volume of the reactor vessel head, and plugged steam generator tubes.

Reduction of the RCS active volume is directly proportional to the reduction in operator response time for the boron dilution event described in the' Farley FSAR.

APCo estimated that 10% tube plugging will reduce the Farley active volume by approximately 4%. However, the licensee stated that, from the boron dilution.

analysis done for Farley, it can be shown that the RCS active volume'can be minutes) y more than 4%; thus, the required operator action time (of at least 15 reduced b would still be adequate. Therefore, the 10% SGTP limit for the Farley Nuclear Plant will not change the' conclusions of the safety analysis.

In response to a question on the effect of tube plugging on pump coastdown, the licensee in a letter dated October 13, 1986, stated that an analysis was performed which determined that a 10% SGTP limit would not decrease RCS flow below the TDF for the Farley Nucle'ar Plant.

The initial RCS flow used in the pump coastdown analysis is based on TDF which must be maintained to comply with the TS. Therefore, the modeled pump coastdown used in the current non-LOCA analysis will not become more severe. The pump coastdown for, Farley Nuclear Plant is modeled using the PH0ENIX computer code.

A description of the model.is found in WCAP-7993, " Calculation of Flow Coastdown After Loss of Reactor Coolant Pump (PH0ENIXCode)

Technical Specification Changes The Technical Specification changes for Farley Units 1 & 2 are as follows:

1.

In Figure 2.1-1, Reactor Core Safety Limit, applicability was changr to account for the increase in the SGTP limit from 5% to 10%. Thir was discussed in the evaluation above and is acceptable.

2.

In TS 3/4.2.2 Heat Flux Hot Channel Factor - F (z) the value for F was changed 9

n from 2.31 to 2.32 for greater than 50% RTP and from 4.62 to 4.64 for less than original to 50% RTP.

These changes are acceptable for the reasons explained herein.

3.

In TS 3/4.2.1, Bases, Axial Flux Difference, an editorial change was made to account for the new F value of 2.32 instead of 2.31. 'This is acceptable.

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~4.0 ENVIRONMENTAL' CONSIDERATION These amendments' change'a requirement with respect.to_ installation or...

use of a facility component l located within the restricted areas, as: defined '

in'10 CFR Part 20', and. change the surveillance requirements. The staff; has determined:that these amendments; involve.no significant increase in:the.

. amounts, and no. significant change.in0thbitypes, of: any effluentsuthat.may, s

be released off site ~and that there is no significant increase.in individual or. cumulative occupational radiation exposure. The Comission :has'.previously

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issued,a proposed finding that!tnese amendments: involve;no.significant hazards-si consideration, and there: has/been:no public coment on such: finding.. Accord-ingly,.these amendments meet thte eligibility criteria 1for categorical exclu-sionsetforth.in10CFR,51.22(c)(9).-Pursuantto10CFR51.22(b).no.-'

environmental. impact' statement or. environmental assessment need be? prepared-i, in connection with the issuance'of these amendments.

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5.0 CONCLUSION

The Comission made a proposedfdetermination that this' amendment. involves-no significant hazards consideration which w'as' published-in-the; Federali

- Register (51 FR 36082) on' 0ctober 7,; 1986,J and,(52 FR.26582) onl July. 15,;1987,-

'q and consulted with the -State of' Alabama'.. No public"coments or requests for hearing were received and the-State:of, Alabama'did-not.have any comments.

The staff has concluded, based on the considerationsEdiscussed above,-

that: (1) there is reasonable assurance that the health and safety of the pu'olic will not be endangered by operation in the proposed' manner, Land:

(2) such activities will.be conducted in compliance.with the Comis'sion's regulations-and the issuance of these amendments will'not be inimical"to-the-comon defense' and' security lor to the health and-safety; of the public.;

Principal Contributors:

H. Balukjian E. Reeves Dated: October 26, 1987 j

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