ML20206D267
| ML20206D267 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| Issue date: | 04/06/1987 |
| From: | Stewart W VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.) |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| 86-655A, NUDOCS 8704130285 | |
| Download: ML20206D267 (4) | |
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VIRGINIA EtzcrNIC AND Powna COMPANY Ricitwoxn,VINGINIA 202G1 w.L. STEWART Vaca Passinzwr xcci.mi ors =*rion.
April 6, 1987 United States Nuclear Regulatory Commission Serial No.
86-655A Attention: Document Control Desk N0/EJL:jmj Washington, D.C.
20555 Docket Nos.
50-338 50-339 License Nos.
NPF-4 NPF-7 Gentlemen:
VIRGINIA ELECTRIC AND POWER COMPANY NORTH ANNA POWER STATION UNITS 1 AND 2 PROPOSED TECHNICAL SPECIFICATIONS CHANGES SUPPLEMENTAL INFORMATION In our letter dated January 30, 1987 (Serial No.86-655) we proposed changes to the Reactor Trip System Instrumentation sections of the Technical Specifications for North Anna Power Station Units I and 2.
The proposed changes would modify the surveillance frequencies, and allowable out-of-service and test times for selected reactor trip system instrumentation functional units.
Based on comments that we have received from your staff, we are providing a supplement to the information that was contained in the 50.92 Significant Hazards Review. This supplemental information is contained in Attachment 1.
If you or your staff have any questions or need additional information, please contact us.
Very truly yours, l
~
W. L. Stewart Attachment sh 8704130285 870406 PDR ADOCK 05000338
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U. S. Nuclear Regulatory Commission 101 Marietta Street, N.W.
-Suite'2900 Atlanta, GA ~30323 Mr. J. L. Caldwell NRC Senior Resident Inspector North Anna Power Station Mr. Charles Price
-Department of Health 109 Governor Street Richmond, Virginia 23219 s
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Supplemental Information 50.92 Significant Hazards Review 1
The Westinghouse Owners Group submittal of responses to NRC Request No. I for additional information-on WCAP-10271 (letter from Mr. J. J. Sheppard, WOG, to Mr.
C.
O. Thomas, NRC, dated October 4, 1983, letter number OG-106) presented analysis results of the effect of the proposed changes to the technical specifications on Reactor Protection System unavailability, the probability of core melt resulting from an Anticipated Transient Without Trip (ATWT), and the probability of core melt from inadvertent reactor trips. The NRC staff agreed with the conclusions of those analyses as documented in the Safety Evaluation Report that the NRC issued on the topical reports (letter from Mr. C. O.
Thomas, NRC, to Mr. J. J. Sheppard, WOG, dated February 21, 1985).
We have reviewed the requirements of 10CFR50.92 as they relate to the proposed technical specifications changes and have determined that a significant hazards consideration is not involved.
In support of this conclusion, the following analysis is provided.
Criterion 1 - Operation of North Anna Units 1 and 2 in accordance with the proposed license amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated.
Implementation of the proposed changes is expected to result in an acceptable increase in total Reactor Protection System yearly unavailability. This increase, which is primarily due to less frequent surveillance, results in an increase of similar magnitude in the probability of an ATWT and in the probability of core melt resulting from an ATWT. With this slight increase, the probability of ATWT and core melt from ATWT remain within published acceptance criteria.
Implementation of the proposed changes is expected to result in a significant reduction in the probability of core melt from inadvertent reactor trips. This is a result of a reduction in the number of inadvertent reactor trips occurring i
during testing of Reactor Protection System instrumentation. This reduction is primarily attributable to testing in bypass and less frequent surveillance.
The reduction in core melt probability due to inadvertent reactor trip is I
sufficiently large to counter the increase in ATWT core melt probability resulting in an overall reduction in total core melt probability.
The proposed changes do not result in an increase in the severity or consequences of an accident previously evaluated.
Implementation of the proposed changes affects the probability of failure of the Reactor Protection System but does not alter the manner in which protection is afforded nor the manner in which limiting criteria are established.
I Criterion 2 - ihe proposed license amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated.
The proposed changes do not result in a change in the manner in which the Reactor Protection System provides plant protection. No change is being made
4 which alters the functioning of the Reactor Protection System (othei than in a test mode). Rather, the likelihood or probability of. the Reactor ' Protection System functioning properly is affected as described above. Therefore, the proposed changes do not create the possibility of a new or different kind of accident nor involve a reduction in a margin of safety as defined in the Safety-Analysis Report.-
Criterion 3 - The proposed license amendment does not involve a significant reduction in a margin of safety.
The proposed changes do not alter the manner in which safety limits, limiting safety system setpoints or limiting conditions for operation are determined.
The impact of reduced testing other than as addressed above is to allow a longer time interval over which instrument uncertainties (e.g., drif t) may act.
Experience at two Westinghouse plants with extended surveillance intervals has shown the initial uncertainty assumptions to be valid for reduced testing.
Implementation of the proposed changes is expected to result in an overall improvement in safety by:
a.
Fever inadvertent _ reactor trips per unit. This is due to less frequent testing and testing in bypass which minimizes the time spen'. in a partial trip condition.
b.
Higher quality repairs leading to improved equipment reliability due to longer repair times.
Improvements in the effectiveness of the operating staff in monitoring and c.
controlling plant operation. This is due to less frequent distraction of the operator and shift supervisor to attend to instrumentation testing.
As previously stated implementation of the proposed changes results in a slight increase in the probability of ATWT and ATWT core melt. With this increase the probability of core melt from ATWT remains within published acceptance 5
criteria.
Overall core melt probability decreases._ Implementation of the proposed changes does not increase the consequences of a previously analyzed accident nor reduce a margin of safety. Functioning of the Reactor Protection System and the manner in which limiting criteria is established is unaffected.-
Conclusion The foregoing analysis demonstrates tfst the proposed amendment to the technical specifications does not involve a significant increase in the probability or consequences of a previously evaluated accident, does not create the possibility of a new or different kind of accident and does not involve a significant reduction in a margin of safety.
Additionally fewer _ inadvertent reactor trips are expected, equipment reliability is expected to increase and operator effectiveness is expected to improve.
t Based upon the preceding analysis, we co'uclude that the proposed amendment does not involve a significant hazards consideration.
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