ML20054H498
| ML20054H498 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 06/11/1982 |
| From: | Lombardo J Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20054H497 | List: |
| References | |
| NUDOCS 8206240135 | |
| Download: ML20054H498 (3) | |
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UNITED STATES 4
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NUCLEAR REGULATORY COMMISSION g.
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULA FOR OYSTER CREEK NUCLEAR GENERATING STATION SUPPORTING AMENDMENT NO. 61 TO PROVISIONAL OPERATING LICE GPU NUCLEAR CORPORATION AND JERSEY CENTRAL POWER & LIGHT COMPANY DOCKET NO. 50-219
1.0 INTRODUCTION
B~y letter dated May 28, 1982 GPU Nuclear Corporation and Je.rsey Central Power & Light Company (the licensees) requested an amendment to Provisional Operating License No. DPR-16 for the Oyster Creek Nuclear Generating Station.
The amendment would authorize a one time only change to Section 1.12, Definition of Refueling Outage, to extend the time between successive tests of surveillances, prior to cycle 10 only, from not exceeding 20 months to not exceeding 30 months.
The last refueling / maintenance outage (cycle 9) was completed July 20, 1980. From July 20, 1980 to the present approximately ten months of unscheduled maintenance outages have occurred.
It is the expressed purpose of this request to extend the surveillance interval to be in keeping with the projected cycle 10 refueling / maintenance outage scheduled for January 15, 1983.
2.0 DISCUSSION AND EVALUATION l
The present Appendix A Technical Specification Section 1.12, Definition of Refueling Outage, requires that after the first refueling outage, the time between sucessive tests or surveillance shall not exceed 20 months.
Nuclear Corporation has requested that for one time only, prior to theGPU cycle 10 refueling outage, this requirement be extended from 20 months to 1
30 months.
inspection of four torus-to-drywell vacuum breakersThe surveillance five safety valves, (3) inspection of the core spray, s(pa)rger,lacement of 2 rep visual inspection of the interior of the torus.
and (4)
For Item 1, GPU Nuclear Corporction has stated that the vacuum breakers are operability tested once cach month which would tend to increase pressu,re to the suppression chamber pe l
Technical Specification Section 4.5.I.5.a.
l will give assurance that the vacuum breakers are able to perform theirThis m l
safety function during extension of the surveillance interval.
8206240135 820611 PDR ADOCK 05000219 P
. With respect to Item 2, GPU fluclear Corporation has stated that the valves will perform as required to relieve pressure to prevent the reactor pressure from exceeding the safety limit of 1375 psig.
Sixteen main steam safety valves from another nuclear power plant which are of the same type as those used at Oyster Creek, were tested at the Wyle Laboratories. These tests indicated that 4 valves were within proper setpoint tolerance and that the remaining 12 valves had "as found" set points of 20 - 50 psi below trie required set points.
This experience along with other similar valves indicated that when valves were out of setpoint tolerance the drift was also downward.
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In regard to Item 3, GPU Nuclear Corporation states there is a high probability that additional cracking if any during the requested surveillance extension time will not be of significance to affect the intended function of the core spray system. Their conclusion is based on (1) extensive additional cracking is not>likely in the next cycle of operation; (2) existing, and new cracks, should there be any are not likely '.o propagate during normal operation; (3) the only significant operationa' loads that can be postulated are those associated with initiatior of the core spray system, and significant crack propagation is not expected; (4) for reasonable assumptions on the existing crack size ad propagation, analyses demonstrates that the design nominal flow and distribution of the core spray system can be maintained within accepta' ole limits and that a margin remains for new cracks should there be any; (5) the clamp assemblies presently in place at all significant crack locations provide for additional structural capability of the sparger and minimize the potential for geometric changes that are significant to the hydraulic performance to occur; and (6) flow out of the cracks is unlikely to significantly affect the overa11' spray
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distribution. A repair proposal and Safety Evaluation was presented to the NRC on April 1, 1980. Based on a review of the licensee's submittal, Amendment No. 47 to Provisional Operating License No. DPR-16 for the Oyster Creek Nuclear Generating Station was issued on May 15, 1980.
For Item 4, GPU Nuclear corporation stated that based on past inspection history of not finding any significant problems, the requested suryeillance extension would not create a condition adverse to the health and safety of the public. Their conclusion is supported by the following discussion.
In the early seventies, inspections resulted in their finding loose baffles which were removed at that time. The remaining baffles, although not loose, were removed in 1977. Moreover, a general corrosion on the shell was noted during these inspections and corrective action was taken to remove this corrosion and to build up the areas with welding. The corrosion problem was not considered significant.
Based on our review we conclude thet the proposed change to the Technical Specifications would not adversely impact the safety of the plant. We, therefore, find that the request to extend the surveillance interval from 20 months to 30 months on a one time basis prior to the Cycle 10 refueling outage is acceptable. This change request does not affect Section 1.24, Surveillance Requirements, in any way.
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3.0 ENVIRONMENTAL CONSIDERATION
We have determined that the amendment 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 the amendment involves an action which is insignificant from the standpoint of envircnmental impact and, pursuant to 10 CFR 551.5(d)(4), that an environmental impact statement or negative declaration and environmental impact appraisal need not be prepared in connection with the issuance of this amendment.
4.0 CONCLUSION
We have concluded, based ~on the considerations discussed above, that:
(1) because the amendment does not involve a significant increase in the probability or consequences of accidents previously considered and does not involve a significant decrease in a safety margin, the
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amendment 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 issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public.
5.0 ACKNOWLEDGMENTS
This evaluation has been prepared by J. Lombardo.
Date:
June 11, 1982
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