ML19294B320
| ML19294B320 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley |
| Issue date: | 01/09/1980 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19294B318 | List: |
| References | |
| TAC-8132, NUDOCS 8002280178 | |
| Download: ML19294B320 (3) | |
Text
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UNITED STATES o
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NUCLEAR REGULATORY COMMISSION g
WASHINGTON, D. C. 20555 4
's, SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION ZIRCONIUM - WATER CORRECTION TO THE ECCS ANALYSIS FOR FACILITY OPERATING LICENSE NO. DPR-66 DUQUESNE LIGHT COMPANY OHIO EDISON COMPANY PENNSYLVANIA POWER COMPANY BEAVER VALLEY POWER STATION, UNIT NO.1 DOCKET NO. 50-334 Introduction By letter dated May 7, 19791 Duquesne Light Company (the licensee) submitted a revised ECCS analysis for the Beaver Valley Power Station Unit 1.
This analysis was presented in response to the Order for Modification of License which the NRC issued on April 21, 1978.2 The orden:was related to the requirer.2nts of 10 CFR 50.46 (a)(1) that ECCS performance be calculated in accordance with an acceptable calculational model which conforms to the provisions in Appendix K,10 CFR.
Reference 2 contains as an enclosure the safety evaluation which describes the ECCS evaluation model error and which justifies issuance of the order.
The order also requested the licensee to submit, as soon as possible, a reevaluation of the ECCS performance calculated in accordance with the corrected and a pproved model.
The ECCS analysis evaluated here uses the modified Westinghouse evaluation model3 4 which was recently reviewed and approved by the staffs.
It includes the correction fer the Zr-H 0 reaction error.
This present submittal fulfills 2
the requirement of the order.
- Duquesne Light Company (DLC) letter (C. N. Dunn) to NRC (Schwencer) dated May 7,1979.
2NRC Letter (Schwencer) to DLC (C. N. Dunn) dated April 21, 1978.
3 CAP-9220, Westinghouse ECCS Evaluation Model, February 1978 Version, W
WCAP-9220-P-A (proprietary), WCAP-9221-A (Non-proprietary), February,
'78.
4Westinghouse letter (Eicheldinger) to NRC (Stolz) dated April 7,1978.
5NRC letter (Stolz) to Westinghouse (Anderson) dated August 29, 1978.
8002280
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. Evaluation The licensee analyzed the ECCS performance for the large break LOCA using the modified Westinghouse evaluation model.
Previous ECCS analysis submitted for Beaver Valley 1 identified the C =0.4 break as limiting.
D The present analysis was performed for a double ended guillotine cold let break (DECLG) with discharge coefficient of C =0.4.
D The sensitivity analysis provided as an attachment to Reference 1 shows that the C =0.4 break results in a PCT more than 100 F higher than tne D
0.8 and 0.6 breaks. The sensitivity study addresses two important aspects which determine the likelihood of a shift in the limiting break.
First of all, it shows that the PCT is much higher for the C =0.4 D
DECLG than for the C =0.6 and 0.8 DECLG breaks and a major increase in the 0
PCT for the C =0.6 or CD=0.8 DECLG break with no corresponding increase D
in the PCT for the C =0.4 would be required to shift the break. However, D
since the Zr-H O reaction is enhanced at higher temperatures, the increased 2
heat output due to the Zr-H 0 correction would be greater for the C =0.4 2
D break than for the CD=0.6 and C =0.8 breaks.
This would argue against D
any change in the limiting break discharge coefficient.
In addition, the sensitivity study shows that none of the breaks result in a shift of the PCT from an unburst node to a burst node. A shift of PCT from an unburst node to a burst node could be an indication of the possibility of a shift in limiting break discharge coefficient.
The temperature margin between unburst node and burst node in the corrected model for all breaks is more than 200 F.
This indicates the unlikelihood of a shift of PCT from an unburst node to a burst node. This eliminates the possibility of a shift in limiting break discharge coefficient due to a shift in break node type.
We judge that the sensitivity study backed by previous experience provides sufficient justification for accepting the CD=0.4 break as limiting for Beaver Valley Unit 1.
Input parameters assumed in the analysis are listed below:
Core Power: 102% of 2652 Mwt (rated power)
Peak Linear Power: 102% of 12.07 kw/ft Peaking Factor: 2.32 Accumulator Water Volume: 1025 ft3 per tank Steam Generator Tube Plugging: 1.0% uniform The steam generator plugging of 1.0% assumed in the analyses is conservative.
tio tubes have been plugged to date.
The analysis shows that the C =0.4 break results in a PCT of 2124 F, a D
maximum local Zr-H O reaction of 6.2 % and a total Zr-H2O reaction of 2
less than 0.3 %.
The limits of 10 CFR 50.46 require a PCT of.less than 2200"F, a local Zr-H O reaction of less than 17% and a total Zr-H20 reaction of less 2
than 1%. All the values presented in the analysis satisfy these limits and are acceptable.
The licensee did not include a small break analysis since neither steam generator plugging nor correction of the Zr-H O error affect significantly 2
tL results of the analysis.
Environmental Consideration We have determined that this action does not authorize a change in effluent types or total amounts nor an increase in power level and will not result in any significant environmental impact. Having made this determination, we have further concluded that this action is insignificant from the standpoint of environmental impact, and pursuant to 10 CFR Sl.5(d)(4) that an environmental impact statement or negative declaration and environ-mental impact appraisal need not be prepared in connection with the issuan:e of this action.
Concl usions Based on the review of the submitted documents, the staff concludes that tre results of the ECCS reanalysis, performed with the February 1978 version of the Westinghouse ECCS evaluation model, corrected for Zr-4 0 reaction 2
error and incit Jing the assumption of the 1.0% uniform steam generator tube plugging, yield the values of LOCA parameters which are conservative relative to the 10 CFR 50.46 criteria.
The staff considers the submitted ECCS reanalysis acceptable for operation of the plant with up to 1.0% steam generator tubes plugged.
Further, we have concluded, based on the considerations discussed above, that:
(1) because the termination of the NRC's April 21,1978 Order for Modification of License relating the the Zirconium-water error action does not involve a significant increase in the probability or consequences of accidents previously considered and does not involve a significant decr=ase in a safety margin, the action does not involve a significant hazards consideration, (2) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (3) such activities will be conducted in compliance with the Commission's regulations and the temination of the April 21, 1978 Order will not be ininical to the common defense and security or to the health and safety of the public.
Cate: January 9,1980