ML17278A243
| ML17278A243 | |
| Person / Time | |
|---|---|
| Site: | Columbia  |
| Issue date: | 06/12/1985 |
| From: | NRC |
| To: | |
| Shared Package | |
| ML17278A242 | List: |
| References | |
| TAC-57168, NUDOCS 8507010206 | |
| Download: ML17278A243 (3) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, O. C. 20555 SAFETY EVALUATION AMENDMENT NO.
10 TO NPF-21 WPPSS NUCLEAR PROJECT N0.2 DOCKET NO. 50-397 Introduction By [[letter::05000397/LER-1985-014, :on 850214,high Pressure Reactor Scram Initiated Due to Actuation of Turbine Bypass Valve 7 Inches Mercury Low Condenser Vacuum Interlock.Normal Scram Recovery Procedures Followed|letter dated March 14, 1985]], the licensee requested an amendment to the Technical Specifications of the WNP-2 license NPF-21.
Evaluation The proposed amendment would provide relief, for one time only, from the WNP-2 Technical Specifications Surveillance Requirement 4.4.3.2.2, of leak testing three of the eighteen Reactor Coolant System Pressure Isolation Valves.
These valves are designated RCIC-V-66, RCIC-V-13 and RHR-V-23 and are identified in Table 3.4.3.2-1 of the Technical Specifications.
Leak testing of these three valves will require either removal of the containment head or personnel access into the more hazardous areas of the containment.
The licensee proposes to delay the leak testing of these three valves until the first
.scheduled refueling outage.
The valves will be readily accessible at that time because the shield plug and containment head must be removed for refueling.
In the Pacific Northwest, surplus power from hydroelectric veneration results from snow-melt runoff in the spring.
To maximize regional resources, the Bonneville Power Administration has directed that the Supply System is to be on a
12 month scheduled outage cycle that will coincide with this regional surplus power.
The Power Ascension Test Program conducted between licensing (December 20, 1983) and commercial operation (December 13, 1984) required only limited power generation and concomitant minimal fuel burn up during that period.
As a result, refueling is unwarranted at this time but a maintenance outage is scheduled for spring 1985.
The first refueling outage is planned for spring 1986.
Thus, the spring 1985 maintenance outage will not require containment head removal.
Since head removal will not be accomplished, the ability of personnel to perform these valve leak tests is impaired.
Access to these valves under the required test condition (950 + 10 psig) exposes personnel to extreme hazards in the upper elevations o7 the containment and in confined spaces with high pressure test equipment.
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The system design relies.on these v'alves for protection of low pressure pip",'ng.
Extreme pressurization of this low pressure piping can occur upon failure of these valves which is unlikely.
Leakage testing provides an early indication of valve degradition.
The system design is such that any leakage due to degrad-ation that may develop can be readily detected by existing instrumentation because:
'High pressure interface valve leakage pressure monitors (guality Class I) are available with an alarm in the Control Room.
These monitors are under required surveillance by the Technical Specifications.
'Positioii indication on each interface valve is available in the Control Poem.
'Leakage would be diverted to the suppression pool by relief valves provided for over-pressure protection and narrow range suppression pool level indication is available that is sufficiently sensitive to detect significant leakage past both valves in either of these Reactor Coolant System high pressure/low pressure interfaces.
I't should be noted that the operability of these valves is tested at cold shutdown per ASYiE requirements.
To date, no evidence of leakage has been appareiit and the valves have not required maintenance since they were last leak tested.
Had the valves required maintenance, leak testing would have been accomplished at that time as required by the Technical Specifications.
The licensee has determined that these changes have little safety significance and that the proposed amendment will not alter any of the accident analyses.
The staff has reviewed these determinations and the associated changes and conclude that they meet the requirements of the Standard Review Plan, NUREG-0800 and are, therefore, acceptable.
Environmental Consideration This amendment involves a change to the requirement with respect to installation or use of a facility component located within the restricted area as defined iii 10 CFR Part 20.
The staff has determined that the amendment involves no signifi-cant change in the types or significant increase in the amounts of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure.
The Commission has previously issued a proposed finding that this amendment involves no significant hazards consideration and there has been no public comment on such finding.
Accordingly, this amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR Section 51.22(c)(9).
Pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be prepared in connection with the issuance of this amendment.
Conclusioii Me have concluded, based on the considerations discussed above, that:
( 1) there is reasonable assurance that the health arid safety of the public will not be endangered by operation in the proposed manner; aiid (2) such activities will be conducted in compliance with the Commission's regulations and the issuance of this amendment will not be inimical to the coo+on defense and security or to the health and safety of the public.
0ated: Je C 3 meS