ML20059C440

Forwards Description of Resolution of ECCS Evaluation Model Issues & Impact of ECCS Evaluation Model Changes,In Accordance w/10CFR50.46(a)(3)(i) & (Ii),As Clarified in Section 5.1 of WCAP-13541

ML20059C440
Person / Time
Site:	Wolf Creek
Issue date:	10/26/1993
From:	Rhodes W WOLF CREEK NUCLEAR OPERATING CORP.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
ET-93-0121, ET-93-121, NUDOCS 9311010157
Download: ML20059C440 (10)

Text

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'l WQLF CREEK NUCLEAR OPERATING CORPORATION Forrest T. Ahodes Wee Preumnt Engineenng October 26, 1993 ET 93-0121 P

U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Mail Station P1-137 Washington, D.

C.

20555 r

Subject:

Docket No. 50-482:

10 CFR 50.46 Thirty Day Report of ECCS Model Revisions Gentlemen:

This letter describes significant revisions to the Emergency. Core Cooling System (ECCS) Evaluation Models and the estimated effect on the limiting ' ECCS analysis for Wolf Creek Generating Station (WCGS) in accordance with the criteria and reporting requirements of 10 CFR 50.4 6 (a) (3) (1) and (ii), as clarified in Section 5.1 of WCAP-13541, l

" Westinghouse Methodology for Implementation of;10 CFR 50.46 Reporting."

The changes in calculated Peak Clad Temperature ( PCT) ' due to the revisions of Westinghouse ECCS Evaluation Models are reportable per 10 CFR 50.46 guidelines as follows:

1.

For Large Break LOCA (LBLOCA) the net PCT benefit due to Evaluation Model revisions is 25 degrees Fa:"*heit.(*F), for.. a net PCT of 19 61. 2 *F which remains less t.

n 10 CFR 50.46 limit of 2200*F.

2.

For Small Break LOCA (SBLOCA), the net PCT benefit due to Evaluation Model revisions is 13 "F,

for a net PCT of 1548.6 F.

-4 which remains less than the 10 CFR 50.46 limit of 2200 F.

Attachment I describes the resolution of ECCS Evaluation Model issues and the impact of the ECCS Evaluation Model changes.

Attachment II ~

contains the calculated LBLOCA and SBLOCA PCT margin allocations resulting from the permanent changes to Evaluation Models.

Since'the-PCT values determined in the LB and SBLOCA analyses of ~ record,. when combined with all PCT margin allocations, remain well.below the 2200*F regulatory limit, no reanalysis will be performed.

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'I P O Box 411/ Burhngton, KS 66839 / Phone: (316) 364-8831 9311010157 931026 F

An Eqi.al oppodunity Employer M/F/HC/ VET PDR ADOCK 05000482 0

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If you have any questions concerning this matter, please contact me at (316) 364-8831, extension 4002, or Mr. Kevin J. Moles at extension 4565.

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ATTACHMENT I

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SIGNIFICANT CHANGES TO THE WESTINGHOUSE EMERGENCY CORE COOLING SYSTEM EVALUATION MODELS

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t Attachment I to ET 93-0121 Page 2 of 5 Sionificant Chances to the Westinchouse Emeroency Core Coolina System Evaluation Models 1.0 Introduction t

The 1988 revision to 10 CFR 50.46 established reporting requirements for errors or changes in the Emergency Core Cooling System (ECCS)-Evaluation Models. This revision required at least annual reporting of changes and errors in the ECCS Evaluation Models which are not significant.

If the change or error is significant (i.e.,

greater than 50 degrees Fahrenheit

( F) ), the applicant or licensee shall provide-a report to the Nuclear Regulatory Commission within 30 days.

Westinghouse has completed the evaluation.of several potential issues related to the Evaluation Models for ECCS cooling performance. -These issues involve the assumptions made for the physical. models or errors discovered in the program coding.

Each of these issues is discussed in the following sections, which include a description of the issue, the technical evaluation, the resulting change to the Evaluation Model, and the estimated effect of the change on the calculated Peak Cladding Temperature (PCT).

2.0 Evaluation Model Changes 2.1 Safety Injection in the Broken Loop Issue Descrintion Westinghouse recently completed an evaluation of a potential issue concerning the modelirig of Safety Injection (SI) flow into the broken Reactor Coolant System (RCS) loop for Small Break Loss of Coolant Accidents (SBLOCA). Westinghouse previously assumed that SI' l

to the broken RCS loop would result in a lower calculated PCT and, therefore, modeled the ECCS broken loop branch line to. spill _the SI to the containment sump.

The basis for this assumption included consideration for the effect of back pressure on the spilling ECCS line for cold leg breaks, which would see a higher back pressure.for SI connected to the broken RCS loop when compared to spilling against containment back pressure.

Spilling to the higher RCS i

pressure would increase SI to the intact loops, which is a benefit for PCT.

The effect on intact loop SI flow rates as well' as the assumption that. some ' of the SI to the. broken loop would aid' in RCS/ Core recovery resulted in the Westinghouse ECCS model assumption that SI to the broken loop was a benefit.

However, when SI is modeled to enter into the broken loop, a significant PCT penalty is calculated by the NOTRUMP small break Evaluation Model.

(approximately 150 F for a typical Westinghouse 3-loop design).

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Attachment I to ET 93-0121 Page 3 of 5 Technical Evaluation The PCT penalty occurs as a result of competition between the steam venting out the break and the SI to the broken loop, which also exits through the break.

The competition between the steam and the SI renults in higher RCS pressures for the identical core steaming rates.

Since the ECCS uses centrifugal pumps, higher RCS pressure l

results in lower delivered SI flow rates to the' intact RCS loops, j

leading to the calculated PCT penalty.

This penalty is somewhat aggravated by the use of the Moody two-phase break flow model, which is a thermal equilibrium model being used to model a clearly non-equilibrium process. However, the penalty is large enough such that I

a change to a non-equilibrium break flow model would not be expected to offset the break flow RCS pressure interaction seen when SI. is assumed to enter into the broken loop.

When a newer conservative model, based on prototypic test which 1

models the configuration of the SI piping to the RCS cold leg in a l

Westinghouse designed Pressurized Water Reactor (PWR), is used, a net PCT benefit is calculated.

Improved condensation of the loop steam in the intact loops results in lower RCS pressure and larger SI flow rates.

The increase in SI flow rates, due to lower RCS pressure, leads to the lower calculated PCT.

Thus, the negative effects of SI into the broken loop can be offset by an improved SI condensation model in the intact RCS loops.

The improved condensation model is based on data obtained from the COSI test facility.

The COSI test facility is a 1/100 scale representation of the cold Icg and SI injection ports in a Westinghouse designed PWR.

More detailed information on the COSI tests is provided in Reference A.

The COSI tests demonstrated that the current NOTRUMP condensation model under-predicted condensation in the intact loops during SI and thus is a conservative model. Use of the improved condensation model has demonstrated that the current NOTRUMP Small Break.LOCA analyses without the improved condensation model and no SI into the broken loop is more conservative (higher calculated PCT) than a case which includes SI into the broken loop and the improved condensation model.

Additionally, the effects of SI in the broken loop have been determined to not change Reactor Coolant Pump trip symptoms developed in response to NRC Generic Letters83-10C and 85-12 or SI i

termination criteria found in the Westinghouse Owners Group (NOG)

.j Emergency Response Guidelines.

i Chanae to the Evaluation Model The COSI tests demonstrated that the current NOTRUMP condensation model under-predicted condensation in the intact loops during S! and thus is a conservative model.

Furthermore, recent evaluations have shown that the current NOTRUMD SBLOCA analyses without the improved condensation model and no SI into the broken loop is more

' ' Attachment I to'ET 93-0121 Page 4 of 5 conservative (higher calculated PCT) than a case which includes SI into the broken loop and the improved condensation model.

Since.

current NOTRUMP based SBLOCA analyses have a conservative calculated

PCT, Wolf Creek Nuclear Operating Corporation concurs with'.

Westinghouse's position that incorporation of these changes into the current NOTRUMP based SBLOCA evaluation models is not necessary at this time.

However, the WOG is reviewing this issue and possible development of a generic program for resolution.

Estimated Effect For the purposes of tracking

PCT, a net PCT change of 0F

( +150 F/~150*F) was assessed for this issue since the negative effects of SI flow into the broken loop can be offset by an improved SI condensation model in the intact RCS loops.

2.2 NOTRUMP Drift Flux Flow Regime Map Errors I

Issue Descriotion Errors were discovered in both WCAP-10079-P-A [ Reference B) and related coding in NOTRUMP SUBROUTINE DFCORRS where the improved TRAC-P1 vertical flow regime map is evaluated.

In Evaluation Model-applications, this model is only used during counter-current flow conditions in vertical flow links.

The affected equation in WCAP-10079-P-A is Equation G-65 which previously ' allowed for unbounded 1

values of the parameter C= contrary to the intent of the original -

source of this equation.

This allowed a discontinuity to exist ~1n the flow regime map under some circumstances.

This was corrected by placing an upper limit of 1.3926 on the parameter C=

as reasoned from the discussion in the original source.

As stated,. this correction returned NOTRUMP to consistency with the original source for the affected equation.

- t Further investigation of the DFCORRS uncovered an additional ~ closely-related logic error which led to discontinuities under certain other circumstances.

This error was also corrected and returned the coding to consistency with WCAP-10079-P-A.

This was determined to be a Mon-discretionary Change as described in

.Section 4.1.2 of WCAP-1345..

[ Reference C) and was corrected in accordance with Section 4.1.3 of WCAP-13451.

Estimated Effect Representative plant calculations by Westinghouse indicated PCT ef f ects ranging from ~13*F to -55*F.

For the purposes of tracking PCT,. the minimum benefit of

-13*F has been assigned to these changes.

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2.3 Structural Metal Heat Modeling Issue Descrir> tion i

A discrepancy ras discovered during review of the. finite element heat conduction model used in the WREFLOOD-INTERIM code of the LBLOCA Evaluation Model to calculate heat-transfer from structural ~

metal in the vessel during the reflood phase.

It was noted that the material properties available in the code corresponded to those of stainless steel. While this is correct for the internal structures, it is inappropriate for the vessel wall which consists of - carbon steel with a thin stainless internal clad.

This was defined as'a non-discretionary change per Section 4.1.2 of WCAP-13451 [ Reference-C], since there was thought to be potential for increased PCT with a more sophisticated composite model.

The model was revised by replacing it with a more flexible one that allows detailed specification of structures.

-1 Estimated Effect The estimated effect of this correction is a 25*F PCT benefit.

3.0 References

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A.

Westinghouse Letter ET-NRC-93-3971,

" Notification of a Significant Change to the Westinghouse Small Break LOCA ECCS Evaluation Model, Pursuant to 10 CFR 50.46 (a) (3) (ii) : Safety Injection (SI) in the Broken Loop," dated September 21, 1993.

B.

WCAP-10079-P-A, 'NOTRUMP-A Nodal Transient Small Break And General Network Code," August, 1985.

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C. WCAP-13451, " Westinghouse Methodology for Implementation of 10 CFR 50.46 Reporting," October, 1992.

D.

Safety Evaluation attached to NRC letter, " Wolf Creek Generating Station - Amendment No. 61 to Facility Operating. License No. NPF-42 (TAC No. MB4946)," dated March 30, 1993.

E.

WCAP-12909-P,

" Westinghouse ECCS Evaluation Model: Revised Large Break LOCA Power Distribution Methodology."

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-Attachment II to ET 93-0121

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ATTACHMENT II ECCS liVALUATION MODEL PCT MARGIN ASSESSMENTS

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

.A.

ANALYSIS OF RECORD 1 1

Evaluation Model 1981 Evaluation Model with BASH Peaking Factor:

FQT-2.50, Fog-1.65 SG Tube Plugging:

10%

Power Level / Fuel:

3565MW /17x17.V5H w/IFM, non-IFBA t

Limiting transients Cp=0.4, Min. Safeguards, Reduced.Tavg 0

Peak Cladding Temperature (PCT):

1916 F 0

B.

PRIOR PERMANENT ECCS MODEL ASSESSMENTS DPCT =

0F C.

10 CFR 50.59 EVALUATION 0

1. RCS Loose Parts DPCT = +20.2 F D. CURRENT 10 CFR 50.46 CHANGES-1993 0

1.

Structural Metal Heat Modeling DPCT = -25 F E TEMPOPJGY USE OF PCT MARGIN 0F2 0

1.

Power Shape Assumption DPCT =

F. OTHER MARGIN ALLOCATIONS

1. Transition Core (STD/V5H)

DPCT

+50 F3 0

0F4 0

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Cold Leg Streaming Temperature DPCT -

Gradient NET PCT Result 1961.20F Notes:

1. Based on the reanalysis that was performed to support Wolf Creek rerating program.

The results of the reanalysis were submitted' to the NRC as part of the Cycle 7 Technical Specification change package.

The NRC has reviewed and approved the proposed changes

[ Reference D)

2. The process described in Reference E is used to assure that cycle-specific power distribution will not lead to results more limiting-j 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 basis for reloads utilizing both V5H (with IFMa) and STD fuel until a full core'of V5H is achieved.

0 25 F was assessed.

For the purposes of tracking

4. A PCT benefit of <

0 a benefit of 0 F has been assigned to this change.

PCT,

i Attachment II to ET 93-0121' Page 3 of 3

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

A, ANALYSIS OF RECORD 1 Evaluation Model:

1985 Evaluation Model with NOTRUMP Peaking Factor:

FQT=2.50, FDH=1.65 I

SG 1bbe Plugging:

10%

Power Level / Fuel:

3565MW /17x17 V5H w/IFM, non-IFBA e

Limiting transient:

3-inch Break 0

Peak C1 adding Temperature (PCT):

1510 F.

7 0

B.

PRIOR PERMANENT ECCS MODEL ASSESSMENTS DPCT =

0F C.

10 CFR 50.59 EVALUATION 1.

RCS Loose Parts DPCT = +44.60F D.

CURRENT 10 CFR 50.46 CHANGES-1993 0

1.

Effect of SI in Broken Loop DPCT = +150 F 0

2.

Effect of Improved Condensation Model DPCT = -150 F 0

3. Drift Flux Flow Regime Errors DPCT =

-13 F 0

0F E.

TEMPORARY USE OF PCT MARGIN DPCT =

l F. OTHER MARGIN ALLOCATIONS

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Cold Leg Streaming Temperature DPCT =

+7F Gradient NET PCT Result 1548.60F Notes:

1.

Based on the reanalysis that was performed to support Wolf Creek rerating program.

The results of the reanalysis were submitted to the NRC as part of the Cycle 7 Technical Specification change package.

The NRC has reviewed and approved the proposed changes (Reference D).

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