ML20081A908
| ML20081A908 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 03/08/1995 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20081A901 | List: |
| References | |
| NUDOCS 9503150341 | |
| Download: ML20081A908 (2) | |
Text
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UNITED STATES f
j NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 30666 4 001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NOS. 115 AND 108 TO FACILITY OPERATING LICENSE NOS. DPR-42 AND DPR-60
)
NO.RTHERN STATES POWER COMPANY PRAIRIE ISLAND NUCLEAR GENERATING PLANT. UNIT NOS. 1 AND 2 DOCKET NOS. 50-282 AND 50-306
1.0 INTRODUCTION
(
By letter dated January 13, 1995, the Northern States Power Company (NSP or the licensee) submitted a proposed revision to the Technical Specifications (TS) appended to Facility Operating License Nos. DPR-42 and DPR-60 for the Prairie Island Nuclear Generating Plant, Unit Nos. I and 2.
The proposed revision would revise -TS 4.4.D.1 to extend the interval for residual' heat removal (RHR) system leakage testing from the current "once every 12 months" 2
to "once every refueling outage "
t Following a design-basis, loss-of-coolant accident, the RHR system becomes an extension of the containment once a changeover from the safety injection phase to the recirculation phase occurs. To minimize post-accident leakage from the' i
RHR system, TS 4.4.D.3 imposes a limit of 2 gallons per hour from either RHR i
train, at a system pressure of 350 psig. This limit ensures that the incremental offsite exposure from this source will be insignificant when compared to the exposure resulting from direct containment leakage following the design-basis accident.
TS 4.4.D.1 requires that those portions of the RHR system located outboard of the containment isolation valves-(which are open during post-' accident operation) be hydrostatically tested at regular intervals to ensure that this limit is not exceeded.
2.0 EVALUATION Under the currently specified 12-month test interval, the RHR leakage test t
must be performed during power operation, during which time the RHR system is not in operation.
The test is conducted by pressurizing the RHR system to the i
350 psig test pressure using coolant supplied via a letdown line from the chemical 'and volume control system (CVCS).
Leakage is determined by visual observation. Because the CVCS system is connected to only one RHR train, testing of the other train requires that a cross-connect valve between the two trains be temporarily opened. This valve is normally maintained closed to provide train separation and redundancy.
If the leakage test is conducted during a refueling outage, the RHR system is i
operating in the shutdown cooling mode and, during the initial stages of 9503150341 950300 I
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i cooldown, is pressurized to above the 350 psig test pressure. LNo change in valve configuration is necessary to test either RHR train. - Accordingly,.
a performance of the. leakage test during an outage is_less complex and requires fewer manual. actions than when performed during power operation.
The proposed reduction in the testing frequency increases. the potential for leakage to go undetected for a longer period of time-(approximately 6 months).
l However, previous. testing on a 12-month interval has not disclosed:significant
. leakage and extending the interval.to' refueling outages greatly reduces the complexity of the test.. It is also likely that during the. routine quarterly functional testing and inspection of the RHR system (atia much lower pressure than 350 psig), any significant leakage would be identified. Additionally, 1
leakage testing of the RHR system on a refueling interval is consistent with~
j the Westinghouse Standard Technical Specifications (NUREG-1431,.Rev.0) and.
with the interval for Type B and C containment penetration testing specified in Appendix J of 10 CFR Part 50.
3.0 CONCLUSION
Based on the above evaluation, we find the revision to TS 4.4.D.1 extending-the RHR system. leakage test interval from "once every 12 months" to "once i
every refueling outage" for Prairie Island Units 1 and 2 to be acceptable.
4.0 STATE CONSULTATION
In accordance with the Commission's regulations, the Minnesota State' official was notified of the proposed issuance of the amendments. The State official l
had no comments.
i
5.0 ENVIRONMENTAL CONSIDERATION
The amendments change surveillance requirements. The NRC staff has determined that the amendments involve no significant increase in the amounts,'and no significant change in the types, of any effluents that may be released-offsite, and that there is no significant increase in individual or cumulative i
occupational radiation exposure. The Commission has previously issued a proposed finding that the amendments involve no significant hazards consideration and there has been no public comment on such finding 1
(60 FR 6308). Accordingly, the amendrhents meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).
Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need i
be prepared in connection with the issuance of the amendments.
6.0 CONCLUSION
The Commission has concluded, based on the considerations ~ discussed above, that:
(1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public.
Principal Contributor:
H. Abelson Date: March 8, 1995
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