ML20079D430
| ML20079D430 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 06/26/1991 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20079D429 | List: |
| References | |
| NUDOCS 9107150361 | |
| Download: ML20079D430 (4) | |
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SAFETY EVALUATION-BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NOS. 97 AND 90 TO FACILITY-OPERATING LICENSE _ N0 1 DPR-42 AND DRP-60 NORTHERN STATES POWER COMPANY PRAIRIE ISLAND NUCLEAR GENERATING PLANT, UNIT NOS. 1 AND 2 DOCKET NOS. 50;282 AND 50-306 l
-1.0
_ INTRODUCTION
- By' letter dated February 26, 1991, Northern States Power Company (the licensee) requested an amenenent to the Prairie Island Nuclear Generating Plant Technical Specifications (T5. The proposed TS change would add an action statemeat allowing continued operation of the plant in its existing condition for a period of 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> with one of two steam generator (SG) power operated relief valves
-(PORVs). inoperable. The present action statement allows one hour to restore operability.of the.SG PORVs before action to place the plant in a shutdown condition is required. 'The proposed change also revises the Bases for TS 3.4 to better describe the Bases of the SG PORV specification and to provide guidance in-determining operability of the SG PORVs.
In the-licensee's original submittal, the-_ proposed TS change was justified with respect to aLless restrictive action statement, inadvertently deleted from the TS by. license amendments 91 and 84, approved October 27, 1989. Supplementary
-technical-'information was provided in a submittal dated May 7,1991. As a technical review, this Safety Evaluation (SE) does not address the issue o#
unintentional changes to a facility's-license.
Instead, this SE only addresses the_ significance of any increase in the probabl11ty or consequences of a lpreviously analyzed accident, or any_ decrease in the margin of safety resulting from a change:to the existing TS;
'Each facility at the Prairie Island Nuclear Generating Plant is a two locp-Westinghouse design. The main steam headers direct steam from the two steam generators to the main turbine.
A main steam isolatien valve (MSIV) is located just outside containment. A safety valve header containing five code safety valves and one PORV is connected to each main steam header upstream of the MSIV.
Downstream of.the MSIVs, each main steam header contains a connection for an-
. Two atmospheric steam dump valves are supplied atmospheric steam dump header.
by each header. A single condenser steam dump valve is supplied from the loop B main steam header downstream of the equalizing lir.e which connects the two nain' steam headers.
_The five steam dump valves noted above make up the steam dump system. During power operation, the steam dump system acts as an artificial load to mitigate
-the transient effects of a rapid loss of turbine load.
The steam dump system is also_used to remove stored heat and decay heat from the reactor coolant 9107150361 910626 DR ADOCK 0500 2
2 system. The steam dump system can only function with one or more MSIV's open and the main condenser available.
The steam dump system is classified as a non-s:fe'/ related system.
The two SG PORVs are air operated valves which are normally automatically controlled by a pressure error signcl. The normal set point is below the lowest safety valve lift setting to minimize the lifting cycles on the code safety valver The valves can also be manually controlled from the main control board and the Y shutdown panel. A handwheel mounted on each valo; allows local manual opc "on when power operation is unavailable. A r-. ally operated block valve upstre.n of the SG PORV allows isolation and at power testing of the SG PORV. The SG PORVs are also used to remove stored heat and decay heat from the reactor coolant system when the steam dump system is unavailable. Only the SG PORV pressure boundary is safety related.
The preferred nicans of cooling down the-plant to residual heat removal (RHR) entry conditions is via the steam dump system to the main condenser. The SG PORVs provide an alternate method of plant cooldown when the main condenser or steam dump system is unavailable. The code safety valves provide overpressure protection for-the SGs when the SG PORVs are inoperable or the amount of steam that must be relieved exceeds the capacity of the SG PORVs.
2.0 EVALUATION General Design Criterion 34 of 10 CFR Part 50, Appendix A specifies requirements relating to the system function of remoeing heat from the reactor coolant systen in indirect cycle plants. Guidance based or these requirements is provided in Standard Review Plan 10.3. This guidance includes the provision that the SG PORVs are remotely operable from the contro? room so that cold shutdown can be achieved using only safety-grade components. This guidance, however, was not included in the licensing basis of the Prairie Island Nuclear Generating Plant.
Although the SG PORVs can be manually controlled from the main control board and the hot shutdown panel, the SG PORVs are not safety-grade-components and they are not designed to operate remotely on a loss of off-site power. The SG PORVs do provide the capability to cool the plant to RHR entry conditions during a loss of off-sita power by local manual operation of a SG PORV.
The capability of the system is consistent with the licensing basis of the Prairie Island facility.
The proposed revisior. to the Bases of TS 3.4 states that the SG PORVs are required to remove decay heat and cool the reactor down following a steam generator tube rupture event and following a high energy line rupture outside containment (downstream of MSIVs). The proposed revision to the Bases also provides information on the SG PORV block valve, provides a definitien of operability for the SG PORVs, and makes editorial changes. These statements are. consistent with the design of the steam system and the safety analysis of the Updated Safety Analysis Report (USAR). Therefore, the proposed revision to the Bases of TS 3.4 is acceptable.
The proposed change to TS 3.4 would allow one SG PORV to be inoperable for a period of 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> before action to place the plant in the cold shutdown condition is initiated.
Since the block valve :an be closed to isolate a SG PORV, some repairs may be possible with the plant at power. The licensee
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l justified the 48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> time period as the time necessary to safely perform repairs to a SG PORV. Based on past repair history, the licensee states that a major repair to one of the SG PORV1 could be expected to take 34 hours3.935185e-4 days <br />0.00944 hours <br />5.621693e-5 weeks <br />1.2937e-5 months <br />. The licensee also states that, due to the high temperatures and pressures associated with the valves, it is prudent to allow adequate time to repair the valves in a manner which minimizes personal safety risk. The staff considers the proposed completion time acceptable.
The inoperability of one SG PORV will not increase the probability of a previously analyzed accident. Since the SG PORV serves primarily to mitigate-the effects of an accident, inoperability of the valve may increase the consequences of an accident or reduce the margin of safety.
However, the probability of an event occurring during the period of repair which requires operation of the SG PORVs is very small, and the steam dump system and SG code safety valves are normally available to perform the function of the inoperable SG PORV.
In addition, the allowed time for completion of repairs to the SG PORY is reasonable. Therefore, the staff does not consider the increased consequences of a postulated accident or the reduction in the margin of safety to be significant.
Based on the above, the staff finds the proposed change to TS 3.4 acceptable.
The staff has concluded that the_ proposed TS amendment is acceptable based on:
the avEiiooility of alternate components with similar functional capability during the period of repairt the reasonable period of time allowed for the completion of repair; and the low probability, of an event occurring during this period requiring use of the SG PORVs. The proposed changes to the Bases of TS 3.4 are consistent with the design of the steam system and the safety analysis of the USAR and are, therefore, also acceptable.
3.0 STATE CONSULTATION
In accordance with the requirements of 10 CFR 50.91(b) a state consultation was attempted. We were informed, however, that our contact no longer has any official interest in the activities of Minnesota's nuclear power plants.
By neno dated January 4,1991 fron L. B. Marsh to C. Kanmerer, the NRC Office of Governnental and Public Affairs was requested to identify an appropriate Minnesota contact so that state consultation nay continue.
4.0 ENVIRONMENTAL CONSIDERATION
These anendnents change a requirenent with respect to the installation or use of a facility conponent located within the restricted area as defined in 10 CFR Part 20. The staff has deternined that the anendnents involve no significant increase in the anounts, and no significant change in the types, of any effluents that nay be released offsite, and that there is no signifignt increase in individual or cunulative occupational radiation exposure. The Connission has previously issued a proposed finding that these anendnents involve no significant hazards consideration and there has been m public connent on such finding. Accordingly, these anendnents neet the elinibility criteriaforcategoricalexclusionsetforthin10CFRSection51.22(c)(9).
4 i Pursuar;t to'10 CFR 51.22(b),Eno environmental impact statement or; environmental--
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assessment.need bc-prepared-in connection with the issuance of these-amendments..
. 5.0, CONCLUSION:
- The staff.has concluded,l 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 conductedcin compliance with the Commission's regulations, and (3) the issuance of the arendments will not be inimical to the comon defe.ise and security.or to the health and safety of the public.
Principal _ Contributor:
S. Jones
- Dated
- * June 26,-1991 k
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