ML20205Q713

Safety Evaluation Re Util 861001 Submittal of Analysis of Large Break LOCA in Response to License Condition 2.C.(12). Analysis Acceptable & License Condition 2.C.(12) Satisfied

ML20205Q713
Person / Time
Site:	Wolf Creek
Issue date:	04/01/1987
From:	NRC
To:
Shared Package
ML20205Q698	List: ML20205Q695 ML20205Q713
References
TAC-57361, NUDOCS 8704060030
Download: ML20205Q713 (4)

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'Nq4 ENCLOSURE .

s> kFETY EVALUATION q l A i s 4_

WOLF CREEK GENEMTING STATION

( i; DOCKET NO. 50-482 4

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5 introduction g 1 By detter dated October 1, 2086, Xansas Gas and Electric Company, the f

licensee for the Wolf Creek linit I nuclear. power plant sut'mitted an analysis of a large break Loss-of-Coolant Accident (LOCA) in compliar.ce with license condition F.C.17, which; requires submittal of the worst large break LOCA analysis for NRC review using an approved ECCS evaluation model, prior to restart following the first refueling outace. (Reference 1). . License condition 2.C.12 was necessitated by the unavailability cf an approved code for the LOCA ' analysis at the time of the Wolf Creek licensing. Pevisions and changes have since been approved for the ECCS large break evaluation model (Reference 2). iThe revisions and corrections refer to: (1) a modeling change in the PREFLODB code which war found to increase the peak cladding temperature by about!20'F and (2) a systematic input error in the PART code which caused low values of hot assen61y bundle ^ power to be used. This error was found to increase peak cladding temperature by about 100'F. The model used for this LOCA analysis satisfies' the 10 CFR 50 Appendix r. reouirements as described in

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10.CrR 50.46.

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Aeolicable Criteria for a iarce Brea'k LOCA  %

The criteria to be satisfied fcr the large LOCA analysis are described in 10 CFR 50.46 and are: .

3. Peak fue? element clad temperature shall not exceed ?,200*F

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2. Cladding chemical interaction with water or steam must not exceed 1.0% of the total amount of Zircaloy fuel cladding l

3. The clad temperature transient must be teninated with core geometry l' menable to cooling during and after the transient.

4 ;The localized maximum cladding"exidation must not exceed 17.0% during or

- after-Quenching

5. After successful initial-operation'of the ECCS core temperature will be riaintained at a low value and decay heat shall be removed for an extended period of time, required by the radioactivity remaining in the core.

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I 8704060030 870401 PDR ADOCK 05000482 l P PDR

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g Description of the Transient .

During full power a laroe break LOCA at eouilibrium condition(assuming)that the reactor is operating at 102% ofde a reactor trip and safety injection signals. The injection of borated water ,

will complement void fomation to shut down the fission process although the

. presence of boron is not accounted for in the LOCA analysis. Similarly no credit is taken for control rod insertion, leaving void fomation as the only credible mechanism to teminate the fission process in the early phase of the transient. Injection of the borated water provides for beat transfer from the core and prevents excessive clad terperature. Vhen the primary coolant depressurizes to 600 psia the accumulators will begin injecting borated water.

The analysis assumes loss of offsite power, hence, reactor coolant pumps are assumed to trip, however, pump coast down is assumed in the analysis. After the depressurization (blowdown) phase of the transient ends. refill of the reactor vessel begins with emergency core cooling water, which was not assumed to be operational up to this time. The refill is completed when the water level reaches the botten of the fuel rods. Reflood then takes place, i.e. the period of the transient when the water covers the core to the extent that the core temperature rise has been terminated. Continued operation of the ECCS pumps supplies water for the long term cooling. The boric acid concentration in the primary is such as to prevent recriticality.

The blottdown, refill and reflood stages of the transient are analyzed with the methodologies described in References 3 through 7. These documents describe the three major phases of a large LOCA transient, the modeling of each phase, the interfacing of the computer codes and the features which ensure compliance with the requirements of Appendix K to 10 CFR 50. The following codes are used to accomplish the analysis:

SATAN-VI: analyzes the thermal hydraulic transient during blowdown (Reference 8).

WFEFLOOD: calculates the refill and reflood phases of the transient (Reference 9).

PART: calculates the fluid and heat transfer conditions in the core during reflood (References 5, 6 and 7).

I COCO: calculates the containment pressure transient during all

' three phases of the LOCA analysis (Reference 10).

LOCTA-TV: computes the thermal transient of the hottest fuel rod 4 - during the three phases (Reference 4).

Initial Conditions and Parameters The initial conditions and the numepical. values of the input parameters have been conservatively detemined from sensitivity studies (References 11,12 and 13). Accordingly the double-ended cold leg guillotine break was considered as the lin.iting large break. The worst break in the spectrum of break sizes has a discharge coefficient (Cn) of 0.4. This was coupled with a 12-second diesel generator start timb delay. The C =0.6 D and 0.8 cases were also analyzed..

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[' 1 Results .

The analyses showed that:

o Peak fuel element clad temperature is 2,100*F which is less than the acceptance criterion of 2,200"F.

o cladding chemical interaction with water or steam is less than 0.3%

compared to the 1.0% acceptance criterion.

o the localized maximum cladding oxidation is 4.547 which is well below the acceptance limit of 17.05.

o the clad temperature transient is terminated at a time when the core geometry is still amenable to cooling as required by the acceptance criteria.

o the core remains coolable after successful initial operation of the ECCS as reouired by the acceptance criteria and at the end of the transient the core temperture is reduced and stabilized and the decay heat can be removed for an extended period of time.

Summary and Cenclusion We have reviewed the large break LOCA analysis submitted by Kansas Gas and Electric Company for the Wolf Creek plant which was required by license condition 2.C.12. The analysis was performed with methodologies and codes which have been previously approved by the NRC and which satisfy the criteria of Appendix K to 10 CFR 50. The input parameters were selected in a conservative manner as defined by previous parametric studies. The results of the analysis are well within the acceptance criteria of 10 CFR 50.46, therefore, the submitted analysis is acceptable and license condition 2.C.12 has been satisfied.

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4 References

1. Letter from G. L. Koester, Kansas Gas and Electric Company to H. R.

Denton, NRR dated October 1, 1986.

2. Letter from C. E. Rossi, NRP to E. P. Rahe Westinghouse " Acceptance for Referencing of Licensing Topical Report WCAP-9561" Addendum 3, Revision 1, dated August 25, 1986

3. Bordelon, F. M. et al. " Westinghouse Emergency Core Cooling System Evaluation Model-Sumary" WCAP-8339, July 1974.

4. Bordelon, F. M. et al., "LO?TA-IV Program: Los-of-Coolant Transient Analysis" WCAP-8301 P and WCAP-8305-NP, June 1984

5. WCAD-9220-P-A Rev. 1, " Westinghouse ECCS Evaluation Motel, 1981 Version" (and WCAP-9721-A non Proprietary) dated February 1982.

6. Young M., et al., "BART-1A: A Computer Code for the Best Estimate Analyzed Reflood Transients" WCAP-9561-P-A,1984

7. Chiou, T.S., et al., "Models for PWR Reflood Calculations Using the BART Code" WCAP-1006?.

8. Rordelon, F. M. , et al., " SATAN-VI Program: Comprehensive Space-Time Dependent Analysis of Loss of Coolant" WCAP-8302-P and WCAP-8306-NP. June 1974

9. Kelly, R.D., et al., " Calculational Model for Core Reflooding After a loss-of-Coolant Accident" (WREFLOOD Code)" WCAP-8170-P and WCAP-8171-NP June 1974

10. Bordelon, F.M., et al., " Containment Pressure Analysis Code (C0C0)"

. WCAP-8327-P and WCAP-8326-NP June 1974

11. " Westinghouse ECCS Evaluation Model Sensitivity Studies," WCAP-8341-P and WCAP-8342-NP July 1974,

12. Salvatori, R., " Westinghouse ECCS - Plan Sensitivity Studies, WCAP-8340-P and WCAP-8356-NP July 1974.

13. Johnson, W. J., et al., " Westinghouse ECCS-Four Loop Plant (17x17)

Sensitivity Studies," WCAP-8565-P-A and WCAP-8566-A-NP July 1975.