ML19207B611
| ML19207B611 | |
| Person / Time | |
|---|---|
| Site: | Crane, 05000484  |
| Issue date: | 12/31/1976 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19207B612 | List: |
| References | |
| NUREG-0092, NUREG-92, NUDOCS 7909040320 | |
| Download: ML19207B611 (9) | |
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NUREG-0092 kv$b (Suppl. 3 to NUREG 75/102)
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Dhhkh0 Regulatory Comm s on related to construction of Offico of Nuclear Tyrone Energy Park Unit No.1 Docket No. STN 50-484 Northern States Power Company December 1976 of Wisconsin Northern States Power Company mo a
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NUREG-0092 (Supp. No. 3 to NUREG-75/102)
Cecember 1. 1976 SUPPLEMENT NO. 3 TO THE SAFETY EVALUATION REPORT BY THE OFFICE OF NUCLEAR REACTOR REGULATION U.S. NUCLEAR REGULATORY COM!4ISSION IN THE MATTER O_F NORTHERN STATES POWER COMPANY OF WISCONSIN NORTHERN STATES POWER COMPANY OF MINNES0TA ET AL TYRONE ENERGY PARK U*i!T 1 DOCKET NO. STN 50-484 711136 8
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TABLE OF CONTENTS PAGE
1.0 INTRODUCTION
AND GENERAL DISCUSSION.
1-1 1.1 Intrcouction.
1-1 6.0 ENGINEtRLD SAFETY FEATURES.
6-1 6.3 Erergency Core Cooling System.
6-1 6.3.3 Performance Evaluation.
6-1 21.0 CONELUSIONS.
21-1 APPENDICES PAGE AtPENDIX A - CONTINUATION OF CHRONOLOGY OF RADIOLOGICAL REVIEW OF TYRONE PLANT.
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1.0 INTRODUCTION
AND DISCUSSION 1.1 Introduction The Nuclear Regulatory Comission's (Conmission) Safety Evaluation Report in the tratter of the application by the Northern States Power Company of Wisconsin, Northern States Pc.wer Company of Minnesota, Cooperative Power Association, Dairyland Power Cooperative and Lake Superior District Power Company (applicants) to construct and operate the proposed Tyrone Energy Park Unit I was issued on October 31, 1975. Sup-plement No. I to the Safety Evaluation Report was issued on July 19, 1976. Supplement No. 2 was is<ued on September 14, 1976. We indicated in Supplement No. 2 that there was one outstanding issue, involving our evaluation of the applicants' reanalysis of the errergency core cooling system, which required completion.
The purpose of this supplement is to update our Safety Evaluation Report (and Supple-rent Nos. I and 2) by p-oviding our evaluation of additional information submitted by the applicar;ts since the issuance of Supplement No. 2 concerning a reanalysis of the energency core cooling system.
Each of the sections in this supplement is nunbered the same as the section of the Safety Evaluation Report and Supplement Nos. I and 2 that are being updated, and is supplementary to and not in lieu of the discussion in the Safety Evaluation Report and Supplement Nos. I and 2.
Appendix A is a continuation of the chronoloqy of our principal actions related to the processing of the application.
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6.0 ENGINEERED SAFETY FEATUPES 6.3 QersencyCoreCoolinaSystem 6.3.3 Ferforrance Evaluation In Section 6.3 cf Supplement No. 2 to the Safety Evaluatior Report, we stated that we were evaluating information submitted by the applicants on Septerber 3,1976, con-cerning a reanalysis of the energency core cooling system. We had requested that this analysis be performed by conservatively assuning an upper head coolant temperature equal to the hot leg coolant temperature, since sensitivity studies performed by Westinghouse indicate that the peak clad temperature of the fuel element for a postu-lated loss-of-coolant accident increases with an increase in the upper head coolant temperature.
As a result of our review of the information subcitted by the applicants, we requested additional information on the reana'ysis of the emergency core cooling system. In a letter dated October 15, 1976, the applicants submitted the additional infornation consisting of a loss-of-coolant accident analysis for four postulated large pipe ruptures using the hot leg coolant ter;erature in the upper head region. The Msting-house Topical Report WCAP-SE65 (October 1976), " Westinghouse ECCS - Four-Loop Plant (17 x 17) Sensitivity Studies," which included the appropriate generic break study that used the increased upper head coolant temperature, was referenced by the appli-cants. This generic study indicated that the double-ended cold leg guillotine rupture was still the most limiting break for four-loop plants. These calculations satisfy the break spectrum requirements of Secticn 50.46 of 10 CFR Part 50.
The analyses were performed with a nodified version of the Westinghause evaluation rodel used for the previous analyses reported in Section C.3.3 of Supplement No. 1 to the Safety Evaluation Report. This modified rodel was also found to be acceptable as docurented in the Comission's letter to Westinghouse datad May 13,1976.
The analysis to determine the containment backpressure for the reanalysis of the SNUFPS emergency core cooling systen was perforred in the same way as the previous analysis discussed in Section 6.2.1 of the Safety Evaluation Report. This analysis was perforned with a reference containment using assumptions which meet the require-ments defined in Section 6.2.1 of the Safety Evaluation Report. The resultant contain-ment backpressuro from this analysis was then used in the reanalysis of the emergency core cooling systan.
The applicants also performed a containment backpressure analysis us'ng best estinate parameters for 6 typical SNUPPS containment. This analysis showed that higher contain-ment backpressures would occur than for the reference containment. Higher containnent 6-1
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backpressures will result in iower peak clad t6..;;ratures in the analysis of emergency core coolin7 systems. Therefore, we conclude that the applicants have shown that the reference containment analysis 15 conservative for the reanalysis of the emergency core cooling system for the SNUPPS plants.
We reaffim cur conclusion, stated in Sectiot 6.2.1 of the Safety Evaluation Report, that the plant-dependent inforn,'.tiore used for the emergency core cooling aystem con-tainment pressure analysis for the SNUPPS plants is conservative, and that the cal-culated containment pressure is in accordance with Appendix K to 10 CFR Part 50.
The new analyses identified the worst break as a double-ended cold leg break with a discharge coefficient (Moody mul iplier) of 1,0.
The peak clad temperature of the fuel element was calculated to be 2148 degrees Fahrenheit, which is below the accept-able limit of 2200 degrees Fahrenheit as specified in Section 50.46 of 10 CTR Part 50.
In addition, the calculated maximum local metal-water reaction of 6.7 percent and a total core-wide metal-water reaction of less than 0.3 percent are well below the allowable limits of 17 percent and one percent, respectively. These aralyses were performed using a total peaking factor of 2.32, 'J2 percer.t of the rated nuclear steam supply system power level of 3411 megawatts theral and 1C2 percent of the peak linear power density of 12.6 kilowatts per foot. The analyses alsr conservatively assuned the loss of of fsite power and the loss of or,e low head safety injection pump.
On the basis of this evaluation and our previous evaluation described in Section 6.3.3 of Supplement No. I to the Safety Evaluation Report, we conclude that the emergency core cooling system perfomance for the SN'JPPS plants conforms to the acceptance criteria in Section 50.46 of 10 CFR Part 50.
Therefore, we reaffirm our conclusion tnat the design of the Tyrone emergency core cooling system complies with the Final Acceptance Criteria. On this basis, the emergency core cooling systen also complies with General Design Criterion 35 of Appendix A to 10 CFR Part 50.
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21.0 CONCLUSION
S Our evaluation of the reanalysis of the energency core cooling system has confirmed that the system still complies with the Final Acceptance Criteria. Hence, in Supple-ment Nos. I and 2 and in this supplement, we have discussed each of the outstanding issues identified in Section 1.8 of the Ofety Evaluation Report and have indicated a favorabic resolution of each matter. Therefore, we reaffirm our conclusions ds set forth in Section 21.0 of the Safety Evaluation report.
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APPENDIX A CONTINUATION OF CHRONOLOGY OF RADIOLOGICAL REVIEW OF TYRD_5[ PLANT September 8,1976 Meeting with SNUPPS to discuss fire protection criteria and steam tunnel design criteria.
September 8, 1976 Letter from applicants incorporating SNUPPS letter of September 3, 1976.
September 14, 1976 Issuance of Supplement No. 2 to the Safety Evaluation Peport.
September 23, 1976 Letter to applicants requesting additional information concerning ECCS analysis.
September 28, 1976 Letter from SNUPPS providing submittal date for addi-tional information concerning ECCS analysis.
September 30, 1976 Leti'r to applicants requesting information concerning fi rt protection evaluation.
October 15, 1976 Letter from SNUPPS attaching a docure:.t entitled, " Major Reactor Coolant System Pipe Ruptures (Loss of Coolant Accident)," in response to request of September 23, 1976.
October 18, 1976 Letter from applicants incorporating SNUPPS letter of October 15, 1976.
October Z 1976 Letter from applicants advising that information on fire protection will be submitted by April 1,1977, in response to request of September 30, 1976.
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