ML20065D711

Forwards Description of Revs to ECCS Evaluation Models & Estimated Effect on Limiting ECCS Analysis for Facility & Annual 10CFR50.46 Rept of ECCS Evaluation Model Changes

ML20065D711
Person / Time
Site:	Wolf Creek
Issue date:	03/31/1994
From:	Carns N WOLF CREEK NUCLEAR OPERATING CORP.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
WM-94-0061, WM-94-61, NUDOCS 9404070206
Download: ML20065D711 (17)

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WQLFCREEK NUCLEAR OPERATING CORPORATION -

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Neil S. "Buu" Carns President and . ,,M '

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LUi S.~ Nuclear Regulatory Commission

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Subject:

Docket No. 50-482: 10 CFR 50.46 Annual-Report of

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ECCS Model Revisions

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o s This : letter describes revisions to ' the Emergency Coret Cooling - System (ECCS); <

j LEvaluation Models and the estimated effect'on the. limiting ECCSi_analysisffora i, MN Wolf Creek Generating Station (WCGS) in accordance ~ with1 the e criteria K and? ~ , <

reporting requirements . of 10 CFR . 50j 46 (a) (3) (i)- and J(ii) , :as f clarifie'dlinj ,

Section.5.1'of WCAP-13541, " Westinghouse MethodologyjforfImplementationiof110; CFR 50,46 Reporting." The changes. in calculated: Peak; Cladding Temperatures f *

(PCT) due to the : revisions. of2 Westinghouse EECCS- Evaluation' Models Gare ~'

reportable per 10 CFR 50.46 guidelines ~as'followsr '

1; For'Large Break Loss.of~. Coolant. Accident (LO'A), C the.nethPCTjbenefits

.due to Eva uat l on i Mo ed l ' revisions. 'is 6 Ldegrees. Fahrenheit 1 (*F) ', i .f or J x+  : ,

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'F 6 a net PCT of :1955.2*F which 'remsins' less than' thei10 CFR 50.46s limit 1,1 iv

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. of 2200*F. 1 N.. s

' 2. For' Sma)1. Break LOCA, the -' net PCT benefiti due- to. Evaluation J ModN1) '

N a- revisions is 29'F, for-ainet- PCT of 1532.6*F . which remains /le,ss than:

..*x the 10 CFR:50.46 limit'of 2200*F. .

Attachment I describes the resolutfion of'ECCS EvaluationModel? issues landithe2 ,"

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. s 9 limpact l of .. the ECCS EEvaluation'~ Model changes. Att!achment CIIJ containsj thel >

1 y calculated . Large Break. - LOCA nand M Small1 BreakiLOCA PCT ?margini allocations ? ,i

% !resulting from the permanentichangesLtosthe Evaluatisn?Models. DSince s the jcTl ' / e

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7 values determined;;in,the~Large1 Break-land,lSmall' Break LOCAtahalysisTof? record l7

'which combined with allePCT mardin allocations',' ; remain wellitielow!theS2200*Fi

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7 fregulatory limit,ino reanalysis 1will;beiperformed. <

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WM'94-0061 Page 2 of 2 If you have any questions concerning this matter, please call- me at-(316) 364-8831 extension 4000 or Mr. Kevin J. Moles at extension 4565.

Very truly yours, M-Neil S. Carns NSC/j ra Attachments cc: L. J . Callan (NRC), w/a G. A. Pick (NRC), w/a W. D. Reckley (NRC), w/a L. A. Yandell (NRC), w/a 1

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- Atitachment ' I to WM 94-0061 1- Page 1 of 12 i

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'.Y ATTACHMENT I CHANGES TO THE WESTINGHOUSE EMERGENCY CORE COOLING SYSTEM EVALUATION MODELS d

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. Attachment..I to WM 94-0061 Page.2 of 12 Annuall10 CFR 50.46 Report on Emergency Core Cooling System Evaluation Models Changes n

1.0 INTRODUCTION

N l The Large Break and Small Break . Loss of Coolant ' Accidents '(LOCA) for i Wolf L

. Creek . Generating Station (WCGS) were reanalyzed in 1992 using the ' latest acceptable Westinghouse Evaluation Models to support. the WCGS Power Rerate .

Program. The results of the reanalyses were submitted to:the NRC as partiof.

. the Cycle 7 and power rerate license amendment requests. The NRC has reviewed'

' and approved these licensing submittals [ Reference 'l and 2] . Since the reanalyses have been reviewed and approved .by the NRC,..these . analyses-effectively replace the previous analysis of record' and have become' the..

licensing basis analyses. Using the calculated Peak Cladding Temperatures-(PCT) from the reanalyses as the reference point for determining margin'toLthe 10 CFR 50.4 6 (b) (1) PCT requirement, .a 30 day report . [Reference. 3)~:was-submitted in October 1993 for the significant PCT changes' associated with the safety' injection mode 1Ang in the broken loop for the Small Break- LOCA' analysis.

Wolf Creek . Nuclear Operating Corporation (WCNOC) has reviewed the annual 10 CFR 50.46 summary report of Emergency Core Cooling System (ECCS) Evaluation Model changes that were implemented by Westinghouse during 1993. The: report- <

includes information concerning changes to and errors discovered in the t

Evaluation Models as well as evaluations performed to address the: identified LOCA-related potential issues. The review concludes that the ~ cumulative :

effect of changes to, or erroro in the Evaluation Models on the limiting -

transient PCT is not significant. Therefore, reporting of the ECCSl Evaluation Model changes can be submitted on an annual basis according'to the reporting; requirements set forth in 10 CFR 50.46 (a) (3) (ii) .

Attachment II . provides an update of PCT margin ' rack-up 'for WCGS. The i PCT L margin rack-up demonstrates that compliance ..with- the requirements. : of 10 CFR 50.46 would be maintained considering the combined effects offthe ECCS Evaluation Model changes with the plant design changes : performed 'under-10 CFR 50.59.

-2.0 EVALUATION MODEL CHANGES

~The following ' sections describe the .. nature of each change or. error and 11ts

. estimated effect on the calculated' PCT for the limiting ECCS analysis.

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.2.1 Vessel And. Steam Generator Calculation Errors.In LUCIFER Epckcround N The LUCIFER code is used to generate the component databases, from raw input \ l

. data, to be used'in the Small Break and Large Break ; LOCA analyses. Errors' 3 were found in the VESCAL subroutine of the LUCIFER code. These errors were inL-the geometric and mass calculations of the vessel.and steam generator portions of the needed data. All LOCA analyses using the LUCIFER code outputs .are ,

affected by these error. corrections. The errors were corrected in.a manner to'z, maintain the' consistency of the' LUCIFER code. ,.

The errors were determined ' to be a Non-discretionary Change .as described in' l

Section 4.1.2 of WCAP-13451, " Westinghouse Methodology. for Implementation 'of.

10 CFR 50.46 Reporting," and were corrected in accordance with- Section. 4.1.3.

of WCAP-13451.

.i Estimated Effect

-For the purposes of tracking PCT, a net PCT effect.of i16 degrees. Fahrenheit

(*F) for a Small Break LOCA and -6*F for .a Large Break LOCA . based on representative plant calculations have been incorporated ~ into - the PCT margin .;

allocations.

2.2 ISHII Drift Flux Errors Backaround -]

An error was discovered both in WCAP410079-P-A,E"NOTRUMP - i. Nodal . Transient-Small Break and General Network Code," and the : relevant- ' coding , in ; NOTRUMP SUBROUTINE ISHIIA that led to an incorrect calculation of' the -drift? flux ? in" NOTRUMP when a laminar film annular flow wes predicted. :The affected equation).

in WCAP-10073-P-A is Equation G-74 wherein a' factor of-)g','the gravitational.

-constant, was inadvertently omitted from" bothi the ' documentation and 1the The ~ correction of this- error' returned 1NOTRUMP to

-equivalent. coding.

consistency with the ultimate reference for the affected correlation.

This was determined'to be a Non-discretionary Change ass described in_Section:

4.1.2 of WCAP-13451' and was' corrected -in y accordance with '. Section.' 4.1.31of J

~ WCAP.- 13 4 51' . "/'

g m4 Estimated Effect Representative plant analyses .' were used to . estimate a T generic -PCTT effect of-0"F..

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'2.3LNOTRUMP Point Kinetics Error Backaround An ' error' was discovered in 'the coding - used in ' the NOTRUMP - User ' External 2

, , SDBROUTINE .VOLHEAT. The coding did not . correctly perform the ' calculation :

' described -by. Equation 3-12-28 of WCAP-10054-P-A, ' Westinghouse Small ' Break.

ECCS Evaluation Model Using the NOTRUMP Code." This. calculation.is only usedL during the time when the Point Kinetics option'is used-to. determine the core ,

power before reactor trip. Therefore, any analysis that. utilized ' the s more conservative assumption of' constant core power until reactor trip time is.not' affected by this error. The correction of this error returned - NOTRUMP tol J consistency with WCAP-10054-P-A. > ,

This was determined to be a Non-discretionary Change as described in Section 4.1.2 of WCAP-134 51. and was . corrected in accordance with Section 4.1.31 of WCAP-13451.

Estimated Effect

-Representative plant analyses were used to estimate a generic ' PCT effect 'of(

O'F.

,2.4 Core Node Initialization Error

-Backaround An error was discovered in how the properties of , Core Node - components werei

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initialized' for 'non-existent regions 'in the adjoining Fluid.. Node. . ;In particular this led to artificially high core temperatures during' the 1 t ime -

step when ,. the core mixture level crossed a node ~ boundary, Lconservatively, '

- causing slightly more core mixture level depression than appropriate Lduring this_ time step. Correction of this error' allows'for a smoother mixture. level-uncover' transient during node crossings.

. This . was determined to be a Non-discretionary Change. as described':in Section

'4.1. 2 of WCAP-13451 and was corrected in accordance with Section 4.1.3" of t WCAP-13451.

' Estimated Effect 4

The nature of this error led to an estimated. generic PCT effect of 0*F.

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' Attachment I to WM 94-0061 C Page.5 of 12 2.5 NOTRUMP HEAT LINK Pointer Error Backaround An error was discovered.in how NOTRUMP initialized certain HEAT LINK pointeri m,

-variables at the start of a calculation. Correction of this error made': -

NOTRUMP consistent with the original intent of-this section.of coding. ,

This was determined to be a Non-discretionary Change as described in Section ,

4.1.2 of WCAP-13451 and was corrected in accordance with Section 4.le3 Jof WCAP-13451. ,

Estimated Effect Representative plant analyses were used to estimate a generic PCT effect of, 0*F.

2.6 Fuel Rod Model Errors In Small Break LOCA Backoround A number of minor programming errors were corrected in the fuel rod' . heat' up code used in the Small Break LOCA analyses. These corrections'were related' to:

1. Individual rod plenum temperatures

2. Individual rod stack lengths.

3. Clad thinning logic

4. Pellet / clad contact logic j

5. Corrected gamma redistribution

6. Including Zr02 thickness at t-0 initialization'

7. Numerics and convergence criteria of initialization.

These changes were determined'to be Non-discretionary Changes as described-Lin

Section 4.1.2 of WCAP-.13451 and were implemented Jin accordance . with ' Section ,

'4.1.3.of WCAP-13451.

' Estimated Effect The cumulative effect of the error. corrections and convergence criteria:changei

. was found ~ to be .less than 'approximately 14*F. This change'is therefore; judged.

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to have a negligible effect on. PCT and on.a generic basis the estimated effect will be reported. as 0*F.

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Attachment I to WM 94-0061 Page 6 of 12 hr i-2.7.Large Break LOCA Fuel Rod Model Errors Backcround Minor errors in the rod heat up code used in the t arge - Break LOCA analyses ~

e were. corrected. These errors concerned conditions which exist during' periods of pellet / clad contact and the internal bookkeeping. logic' associated with clad, thinning.

~.i u..ese changes were determined-to be Non-discretionary Changes as described 'in Section. 4.1.2 of WCAP-13451 and were implemented in .'accordance with Section 4.1.3 of WCAP-13451. l Estimated Effect Representative plant calculations have indicated that these' corrections have a-negligible effect on PCT for near Beginning-of-Life (BOL) fuel rod conditions' (i.e. < '2000 MWD /MTU). These effects become prevalent as .burnup iincreases, but are not expected to be of any significance until pellet / clad contactTis predicted for steady-state operating conditions (typically. > 8000 MWD /MTU) .

These' corrections therefore result in a negligible PCT impact.for'Large' Break-LOCA licensing basis PCT which are calculated'with near BOL' conditions. This impact is being . reported generically as 0*F.

2.8 High Temperature Fuel Rod Burst Model Backcround

.A model for calculating the ptediction of ' zircaloy cladding burst behavior L above the previous limit of 1742'F was implemented. .This model was. described-to the NRC in:

Letter ET-NRC-92-3746, N. J. Liparulo' (H) sto- R. C. Jones.-(NRC),

" Extension of NUREG-0630 Fuel Rod Burst Strain and Assembly ~ Blockage z Models to High Fuel Rod Burst Temperatures," September :16,q 1992.

This was determined to be a Non-discretionary Change as. described:'in'Section-4.1'. 2 of WCAP-13451 and was corrected in accordance with Section : 4 ;1.3 ' of-t WCAP-13451.

. Estimated Effect The effect. of the extended burst model has[been determined'. to: be -not applicable-to the analysis of record for.WCGS because rod burst occurs below.

the limit'of 1742 0F. This correction therefore~results in'a' PCT effect of:0 F 0 for the limiting Large Break LOCA analysis.

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?4 Attachment I to WM 94-0061

'Page 7.of.12

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2.9-Hot Assembly Average Rod Burst Tffects

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Backaround a

H,, The rod heat up code used in Small Break LOCA analyses contains a model to-

[' calculate the amount of clad strain that accernpanies rod burst. Historically,- '

the methodology used did not apply this burst .' train model to the hot assembly average rod. This was donc so as to minimize the rod gap and therefore maximize ' the heat transferred to the fluid channel, which in turn ' would :

. maximize the hot rod temperature. However, due ' to mechanisms governing ~ the - '

c zircaloy-water temperature excursion (which is the subject of-the Small: Break LOCA Limiting Time-in-Life penalty for the hot rod) , .modeling . of .- clad . burst :

b strain for the hot assembly average rod can result in-a penalty for the' hot-rod by. increasing the channel enthalpy at the time of PCT. Therefore,-the-

• methodology has been revised such that burst strain will also be modeled on .

the hot assembly average rod. '

'This was determined to be a Non-discretionary Change as described in Section- "

4.1.2 of WCAP-13451 and was corrected in accordance with Section 4.1. 3 - of,

- 'WCAP-13451.

Estimated Effect

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l Representative plant calculations have indicated that. this change introduces an approximate 10 percent increase in the Small Break LOCA Limiting . Time-in-

'sife penalty on the hot rod. However, there is no Small Break;LOCA Limiting ' 2 Tinte-in-Life penalty assessed for WCGS. Therefore, this change wouldLimpose; an effect of 0F 0 to the calculated PCT in the limiting Small Break ' LOCAL analysis.

1 2.10 Revised Burst Strain Limit Model Backaround i A revised burst strain limit model which limits strains is being implemented- '

into.the rod heat up codes used in both the Large Break and Small Break LOCA. r!

This model, which is identical to that previously.- approved forf use - for :

Appendix E K analyses of Upper Plenum Injection plants ' with c WCOBRA/TRACi as -

described in WCAP-10924-P-A, Revision 1, Volume 1, Addendum 4; " Westinghouse:

Large; Break LOCA Best Estimate Methodology: ,Volbme -l i . Model

Description:

and-:

' Validation,-Addendum 4t- Model ' Revisions , .1991.

This has been determined to be a . Non-discretionary Change < as . discussed in l

Section . 4.1.2 ' of - WCAP-134 51 and . is being implemented in i accordance 'with -

.Section 4.1.3 of WCAP-13451.

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e, Attachment I to hH 94-0061 Page 8 of 12 Estimated Effect The estimated effect on the Large Break LOCA PCT ranges from negligible to a moderate unquantified benefit, which will be inherent in calculations once ..

this model is implemented. In the Small Break LOCA, implementation of this change would not affect the calculated PCT because fuel rod burst was calculated not to occur at any time in life fuel conditions.

3.0 RESOLUTION OF POTENTIAL ISSUES Westinghouse has completed the evaluation of several LOCA-related potential issues. Each of these issues is discussed in the following sections, which include a brief description of the issue, the technical evaluation, and the estimated effect of the change on the calculated PCT.

3.1 Charging / Safety Injection System Issues Backcround Westinghouse has recently completed its evaluation of a potential safety issue regarding four specific issues related to the design and use of the miniflow line for the charging / safety injection pumps. Two of these issues involved Small Break LOCA PCT penalties for certain plants. One issue involves the j; operation of the centrifugal charging pump (CCP) miniflow line during accident conditions. A CCP runout condition may occur if the CCP injection lines were balanced with the CCP miniflow path closed and credit was taken for operator action to isolate the miniflow line during the accident. Also, the existence of this condition may impact the ECCS flows assumed in plant specific Small Break LOCA analyses. The other issue involves miniflow orifices that are used~

for the charging / safety injection pumps. Westinghouse has supplied two different orifice types: 60 or 70 gpm orifice at a differential head of 6000 feet.

Additional confirmation testing indicates that the orifice plates will allow a higher than design flow rate through the orifice at the design differential head. As a result, a discrepancy may exist between the installed miniflow line capacity and the ECCS analysis assumptions. The discrepancy would occur if the ECCS analysis assumed that the miniflow line resistance was based on the orifice allowing design flow at the design head as opposed to the higher as tested flow and head.

Consequently, the miniflow path may permit more flow than previously determined which may reduce safety injection flow-during injection, j

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Page 9 of 12' Technical Evaluation The concern.of this issue is that flow diversion to'the mini-flow:line duringi a Small Break LOCA may _ result in degradation of the available; centrifugal charging flow for Reactor Coolant System (RCS) injection _ and therefore i the 4

_ total pumped safety. injection flow to the RCS may -be Lless than that assumed.

for the LOCA analyses. However, this concern is'not' applicable to WCGS based on.the current design and use of the charging / safety ' injection :miniflow :line-during a . transient - event . Under normal operation, the mini-flow iso? ation -

valve is open to provide recirculation flow to the CCP. Following receipt _of a-safety injection signal, the mini-flow isolation' valve 'will' remain . open . if '

there -is not - suf ficient flow passing through the- flow switch in . order to - a maintain adequate minimum flow to the CCP. This recirculation. flowi is necessary at high RCS pressure in order to prevent damage.to.the CCP under low flow' conditions. The mini-flow isolation valve also receives an' auto-closure signal to automatically isolate the recirculation line when sufficienti flow has been sensed by the-flow switch.

-It should be noted that the charging / safety injection flowrates' used in i the Small Break 'LOCA analyses were conservatively;' generated based' on ~the configurations in which the CCP mini-flow isolation valves remainLcontinuously- '

open during the entire L transient. This was done 'for added.~ conservStism in simulating the ECCS performance during accident conditions and --also ~for - a possible design change that would convert the automatic safety. grade miniflow.

control system to a continuously open miniflow valve.

Estimated Effect The PCT effect on the Small Break LOCA Evaluation'Model for.this11ssue varied depending on the affected plant ECCS configuration and capability.. An;'

assessment of this issue concluded that the concerns-regarding the design'and-use of the charging / safety injection miniflow line are not applicable:to WCGS-and'would not involve Small Break LOCA PCT penalties'at-this time--based:on; the' current ECCS configuration and capability.

3.2 Double-Disk Gate Valve Pressure Equalization Backorcund d

Westinghouse completed the. evaluation of a' potential. issue:concerning the_use-of double-disk gate . valves in the ECCS as hot leg;isolationLvalves. -Use'of'

-these double-disk gate valves may involve an. inner: disc pressure equalizationi Lline'that could set up a11eak path into'the hot leg during' cold leg' injection' following a LOCA. This condition could lead to inadequate cold leg 11njectioni

.resulting in an increase in PCT.

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Attachment I'to WM'94-0061 Page 10 of 12-The design characteristic of a double-disk gate Lvalve provides -isolation 1by' the downstream disk sealing against the valve seat. The mechanical seating force and the hydraulic force from the upstream pressure (safety l injection 1 pump) act to provide force to the valve seal surfaces. LThe double-disk gate; valve design results in'a volume of fluid which is' enclosed between the discs i when the valve is closed. As the fluid volume heats up, pressure greater than-system pressure may develop and may cause the disks.to bind against~the' seats to the extent that the valves cannot be cpened. .To avoid this, many. double-disk gate valves have been modified to include a pressure equalization line orL a small hole in one of the disks to relieve the pressure between'the disks.

Based on generic leakage calculations it was determined that the double-disk' j gate valves modified to eliminate concerns for thermal. binding'could leak as much as 30 gallons per minute per valve. This leakage into the RCS.hotKlegs will increase steam binding during reflood and result in an increase.in the: <

calculated PCT.

Estimated Effect This: issue would only affect plants that have modified the configuration of-

the double-disc gate valves that could be susceptible to pressure locking. A review of the current ECCS configuration indicates that the configuration of -

the double disc gate valves installed in the ECCS'does not involve afpressure, equalization device, e.g., a bypass line installed or a relief hole. drilled in-one of the discs. Therefore, the concern is.not. applicable to WCGS at:this time and a Large Break LOCA PCT penalty does not need to be assessed.

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3.3 Large Break LOCA Rod Internal Pressure Issues ~

Issue Descrintion Westinghouse recently completed an evaluation of a-potential. issue concerning

'the impact of increased beginning of life rod internal pressure. uncertainties' "

on LOCA analyses. ' Historically, beginning of ' life 1 fuel' -pressureL;and temperature uncertainties, were' based upon end of life considerations. -These ,

rod internal pressure uncertainties were found to ;be- Lpotentially nonconservative. During the evaluation of this issue, a.second. issue related-

, .to the ' applicability of the generic Integral Fuel Burnable Absorber (IFBA) fuelianalyses to the updated ECCS Evaluation Models was also identified and

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combined with this issue since the underlying mechanisms were the same.

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Attachment!' I to WM 94-0061 Page 11 of 12 Technical Evaluation

..This issue concerns the potential that a lower rod internal'. pressure for

, reload fuel than that assumed in the analyses may result in a Large Break'LOCA PCT penalty. The. initial fuel rod internal pressure is'important to the.LOCA' analyses since it affects- several Appendix' K requirements, l'ncluding -

accounting for .the effects of fuel ' rod swelling 'and rupture ~ and 'any flow.

blockage during reflood. The Large Break LOCA analysis is sensitive to this' parameter, with higher pressures generally being more limiting if the. nominal- l design fuel rod pressurization is greater than 275 pounds per square inch-

. gauge: (psig).

The technical evaluation of this issue concluded that both the rod internal-pressure uncertainty and the current IFBA designs with 200 psig initial' fill!

pressure fuel will typically result in a. maximum il5'F PCT ' variation.

Consequently, rod internal pressure manufacturing uncertainties and ;200. psig initial fill pressure'IFBA fuel do not have significant effects on the Large Break LOCA analyses. Also, based on these results, it was concluded that only nominal rod internal pressureL(with.an upper bound bias).should-be'used in the

.LOCA analyses for fuel designs with an initial cold fill pressure 2 200 psig.

This is consistent with past LOCA analyses.

q Estimated Effect Based on the current fuel rod design that utilizes an ' initial' backfill ,

pressure of 275.psig and the core configuration'that uses no IFBA. fuels,Jtherei is no plant specific PCT change associated with this ~ issue 'for WCGS at -this ,

. time. However, any.new reloads which would utilize. low (<200 ' psig) .' initial ~

fill pressure fuel would.be specifically analyzed or evaluated.

3.4 Small Break LOCA Limiting Time In Life -- _

Zircaloy/ Water Oxidation Temperature Excursion lasue Descrintion 4

Westinghouse recently completed an evaluation.of a potent 1El. issue:with regard-to burst / blockage modeling ' in the Westinghouse . Small ; Break LOCA Evaluationi  ?

Model. This potential issue involved a number of synergistici ef fects , . . 'all- 1 related to the manner in which the Small Break LOCA'model: accounts-for the-swelling and burst of fuel rods, modeling of the rod - burst strain, and' resulting effects on clad temperature and oxidation from the metal / water; reaction models and channel blockage, g

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-Technical Evaluation The limiting Small Break LOCA is typically assumed ' to : occur at 'beginning .of-

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life fuel rod conditions where temperatures are the highest. The highest = clad <.

-temperature would be expected to occur at a' time'when the fuel is' most active; N However, .during the life of the fuel, the rod internal - pressure ._ increases '- ,

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greatly while the temperature decreases only slightly. 1It:is.. postulated that-L the fuel rod could burst during'a transient as a result of the increased rod-  !~

h internal pressure. Fuel rod burst during the course of a Smail Break LOCA 1 analysis was found to potentially result -in a significant- temperature- i excursion above the clad temperature transient for-a non-burst case. Since.

i the nethodology for Small Break LOCA analyses.had been to perform the analyses- ,

at a near beginning of life' condition,- where rod ; internal pressure-- is relatively Iow,.most analyses did not result-in the occurrence.of rod burst, .m and therefore may.not have reflected the most limiting time in' life PCT. Fo r - ~

WCGS, this concern has been incorporated into the current analysis of.recordi <

for the Small Break LOCA. Because rod burst is most~ sensitive to rod' internal--

l pressure, a plant specific calculation was performed under the1 highest-. fuel rod pressures which occur at the fuel end of life conditions - (i'.e. 6 0,000 : O MWD /MTU) to determine whether or not fuel rod burst would occur. "The results of this calculation indicated that fuel rod burst would not occur:at the very!

high rod internal pressures. Because fuel rod burst was not . calculated to occur when the fuel would undergo the most severe rod; internal' pressure : ";

excursion, fuel rod burst would not be calculated to occur at.any'other time-in life fuel conditions.

Estimated Effect Because fuel rod burst would not be calculated to occur at any time in life-

. fuel conditions, the beginning of life fuel conditions.~are limiting for. WCGS;

~

and therefore no Time-in-Life penalty on the hot rod--needs to be assessed'at -

this time for the limiting Small Break LOCA.

4.0 REFERENCES

1. Wolf Creek Generating Station - Amendment No . ~ 61 to ' Facility Operating. a License No. NPF-42, dated March 30, 1993.

2. Wolf Creek Generating Station Amendment No. 69 to Facility ' Operating - ,

License No. NPF-42, dated November 10, 1993. ,

3. 1,etter ET 93-0121 dated ' October.. 26, .1993, ' from F. T. . Rhodes , ' WCNOC, t;o-USNRC.

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- Attachment II to WM 94-0061 Page 1 of 3 1

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ATTACHMENT II- y l

ECCS EVALUATION MODEL PCT MARGIN ASSESSMENTS A

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Attachment II to WM 94-0061 ,

e Page 2 of 3

• • • Large Break LOCA PCT Margin Rack-Up Summary ***

P A. ANALYSIS OF-RECORD 1 .

Evaluation Model: 1981 Evaluation Mo' del with-BASH ,

Peaking Factor: FQT-2.50, FDH-1.65 SG Tube Plugging: 10 percent Power Level / Fuel: 3565MWt /17x17 V5H w/IFM,'non-IFBA ,

Limiting transient Cp=0.4, Min. Safeguards, Reduced'Tavg Peak Cladding Temperature (PCT) : 1916 0F.

B. PRIOR PERMANENT ECCS MODEL ASSESSMENTS DPCT =.-25 0 F C. 10 CFR 50.59 EVALUATION

1. RCS Loose Parts DPCT = +20.~20F <

D. 1993 10 CFR 50.46 MODEL ASSESSMENTS (Permanent Assessment of PCT Margin)

1. LUCIFER Error Corrections DPCT = -60 F E., TEMPORARY USE OF PCT MARGIN DPCT = 0 0F F. OTHER MARGIN ALLOCATIONS

1. Transition Core (STD/V5H) DPCT ='+50 0 F3 2.-Cold Leg Streaming Temperature DPCT = 0F4' 0

Gradient NET PCT Result 1955.20F 6

Notes:

1. Based on the reanalyses that was performed to support _the WCGS Power Rerate ,

Program. The results of the reanalyses have been reviewed and approved by:

the NRC, i

2. The Power Shape Sensitivity Methodology ' (PSSM) - $s used to ' assure : that -

. cycle-specific power distribution 'will not lead ta results more . limiting' than those of the analysis of record. Therefore, there is no. PCT effect-assessed for this issue.

3. Transition core penalty applies on a cycle-specific I basis for nreloads-utilizing both V5H (with IFMs)' and STD fuel until'a full core of' V5H - is achieved.

4. A PCT benefit of < 25 F0was-assessed. For the purposes of tracki'ng PCT,ca benefit of 0 0F has been assigned to this change.

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kf.. , .' i ,7 4 Attachment II'to WM 94-0061

. Page 3 of 3

• • • Small Break LOCA PCT Margin Rack-Up Summary ***

A. ANALYSIS OF RECORD 1 Evaluation Model: 1985 EM with NOTRUMP Peaking Factor: FQT=2.50, FDH-1.65 SG Tube Plugging: 10 percent Power Level / Fuel: 3565MWt /17x17 V5H w/IFM Limiting transient: 3-inch Break Peak Cladding Temperature ' (PCT) : 15100 F B. PRIOR PERMANENT ECCS MODEL ASSESSMENTS DPCT = 0F0 h

C. 10 CFR 50.59 EVALUATION

?? 1. RCS Loose Parts DPCT = +44.6 0F.

D. 1993 10 CPR 50.46 MODEL ASSESSMENTS (Permanent Assessment of PCT Margin) .

1. Effect of SI in Broken Loop DPCT = +1500F

2. Effect of Improved Condensation Model ~DPCT = -150 0F.

3. Drift' Flux Flow Regime Errors DPCT'= -130 F

4. LUCIFER Error Corrections DPCT = -160 F E. TEMPORARY USE OF_ PCT MARGIN- DPCT =' 0F0 F. OTHER MARGIN ALLOCATIONS

2. Cold Leg Streaming Temperature 'DPCT.=- +70 F Gradient NET PCT' Result '1532.60 F j<

. Notes:

1. Based on the~ reanalyses that was performed to support the WCGS Power Rerate:

Program. LThe results of the' reanalyses have been reviewed and approved by? ' '

'~

the NRC. ,

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