ML20087P111
| ML20087P111 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 03/06/1984 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20087P109 | List: |
| References | |
| NUDOCS 8404060060 | |
| Download: ML20087P111 (4) | |
Text
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 66 TO FACILITY LICENSE NO. DPR-71 AND AMENDMENT NO. 92 TO FACILITY LICENSE NO. DPR-62 CAROLINA POWER & LIGHT COMPANY BRUNSWICK STEAM ELECTRIC PLANT, UNITS 1 AND 2 DOCKET NOS. 50-325 AND 50-324 INTRODUCTION By letter dated September 7, 1982 Carolina Power & Light Company (CP&L) requested eight miscellaneous revisions to the Technical Specifications (TS) for the Brunst:ick Unit 1 and 2 operating licenses.
The staff has completed the safety evaluation of six of the requested changes. These amendments would modify certain TS requirements, to provide a clarification of certain requirements, provide editorial and/or administrative corrections of certain requirements, and provide changes to reflect consistency with the actual plant design. The revisions are described as follows:
revise the monthly channel check in the surveillance requirements, for Reactor Vessel Water Level to "Not Applicable" for instruments B21-LT-N0170-3 and B21-LSH-N017D-3, revise the valve group number from 7 to 8 for the reactor vessel head spray isolation valves, and revise the valve group number from 8 to 2 for the RHR discharge isolation valves to radwaste and the RHR proc-ess sampling valves, revise the minimum number of flame, heat, and smoke instruments required to be operable in their defined fire zones and add additional fire zones that have been established, revise the surveillance requirement for demonstrating Safety / Relief Valve (S/RV) operability, and revise the snubber list to reflect typographical corrections, snubber additions, and snubber deletions.
This amendment request also requested changes to the Technical Specifications regarding surveillance requirements for primary containment integrity and regarding by-passing the actuation of isolation functions associated with a loss of vacuum in the main turbine condenser. This portion of the amendment request will be addressed in a separate action.
EVALUATION Remote Shutdown Monitoring Instrumentation (Units 1 and 2)
The Technical Specifications define CHANNEL CHECK as a qualitative assessment of channel behavior during operation by observation. This determination shall include, where possible, comparison of the channel indication and/or status with other indications and/or status derived from 8404060060 840306 PDR ADOCK 05000324 P
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2-independent instrument channels measuring the same parameter. This requires that there be indication of the measured parameter. Table 3.3.5.2-1 (Remote Shutdown Monitoring Instrumentation) and Table 4.3.5.2-1 (Remote Shutdown Monitoring Instrumentation Surveillance Requirements) list reactor vessel i
water level transmitter B21-LT-N017D-3 and its associated level switch B21-LSH-N017D-3 and water level transmitters B21-LT-3331 and B21-LT-N026A and their associated level indicators B21-LI-3331 and B21-HI-R604AX.
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The licensee has stated.that the actual plant design is such that the
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reactor vessel water level transmitted B21-LT-N017D-3 and level switch
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B21-LSH-N017D3 input to bistable logic and do not provide input to an
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indicator.
The bistable logic functions to cutoff reactor core isolation i
cooling (RCIC) on high reactor vessel water level.
Since it is not possible to perform a interchannel comparison of display indications the 1
request is to change the testing requirement to "Not Applicable" in Table 4.3.5.2-1.
The requirements for-a quarterly channel calibration would remain unchanged.
In addition, Tables 3.3.5.2-1 and 4.3.5.2-1 have been modified by i
segregating the reactor vessel level instrumentation into groups to reflect the channels that have indication and those that do not. This clarifies the Technical Specifications to reflect the plant design.:
i We find the request to modify Tables 3.3.5.2-1 and 4.3.5.2-1 to be acceptable based on the above discussion and since it is consistent with the surveillance requirements established by the Standard Technical Specifications for BWR plants.
Primary Containment Isolation Valves (Unit 2 only)
J Table 3.6.3.1oftheUnit2TSindicatestheresidualheatremovalL(RHR)-
discharge isolation valves to radwaste (EII-F040 and EII-F049) 'and the RHR process sampling valves (EII-F079A, B and EII-F080A, B) are group 8-i valves.
In fact, these valves are group 2. valves. The revision from 8 to-2 for the valve groups for those valves,-in accordance-with the proposed-TS page in Attachment 2 to CP&L's September 7,1982 letter is acceptable
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since the revision provides consistency between TS information and the actual approved plant design.
Primary Containment Isolation Valves (Unit 1 only)
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The TS revision for Unit 2 primary containment isolation valves is also applicable to the same' valves-for Unit 1 and-is acceptable for Table:
3.6.3.1 of the Unit 1 TS. An additional revision to Table 3.6.3.1 of the Unit 1 TS is required to correct.a typographical error that occurred'during.
the transition from custom.TS to BWR Standard TS.
Specifically, the-revision from "7".to "8" has been reviewed by the staff for the valve'.
groups for the reactor vessel head spray isolation valves (EII-F022 and-
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. EII-F023) in accordance with the proposed TS page in Attachment 3 to CP&L's September 7, 1983 letter. The change is an administrative change that corrects the TS and is therefore acceptable.
Fire Detection Instrumentation (Units 1 and 2)
TS Table 3.3.5.7-1 lists the fire zones and the minimum number of operable flame, heat and smoke instruments for each zone.
The number of installed fire detection instruments has been modified and the fire zones have been redefined by the licensee.
Revisions to the applicable TS sections are necessary to reflect these changes.
Accordingly, the proposed revisions to TS Table 3.3.5.7-1 for Units 1 and 2 in Attachment 4 to CP&L's September 7, 1982 letter have been reviewed by the staff and are acceptable since the revision provides consistency between the TS requirements and the actual plant design.
Safety / Relief Valves (S/RVs) (Units 1 and 2)
The three-stage S/RVs, which are equipped with bellows, have been replaced with two-stage S/RVs which are not equipped with bellows. TS 4.4.2 requires the demonstration of S/RV operability through a bellows integrity check once every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. This check is no longer applicable to Brunswick Units 1 and 2 and cannot be performed. Therefore, on this basis, we find acceptable CP&L's proposed change to TS 4.4.2 which would delete the bellows integrity check and add the provision that S/RV operability be demonstrated in accordance with the requirements of TS 4.0.5.
The proposed page revisions for Units 1 and 2 TS in Attachment 5 to CP&L's September 7, 1982 letter have been reviewed by the staff and are acceptable.
Safety-Related Hydraulic Snubbers (Units 1 and 2)
TS Table 3.7.5-1 provides information on safety-related hydraulic snubbers. Subsequent to publication of the current Table 3.7.-5-1,.the licensee has performed plant modifications resulting in the rerouting or removal of some system lines.
This in turn has caused the relocation or removal of slected safety-related hydraulic snubbers.
The licensee has also discovered typographical errors in some snubber numbers accessibility information in the current Table 3.7.5-1.
The staff has reviewed the proposed changes to Table 3.7.5-1 of the TS for Units 1 and 2 in Attachment 6 of CP&L's September 7,1982 letter and finds them acceptable because they are administrative corrections and update the TS to be consistent with the actual plant design.
ENVIRONMENTAL CONSIDERATIONS We have determined that the amendments dc 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-l
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- determination, we have further concluded that the amendments involve an action which is insignificant from the standpoint of environmental impact and pursuant to 10 CFR 51.5(d)(4), that an environmental impact statement, or negative declaration and environmental impact appraisal need not be prepared in connection with the issuance of these amendments.
CONCLUSIONS We have 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, and (2) such activities will be conducted in compliance with the Commission's regula-tions and 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 Contributors:
L. L. Wheeler, M. Wigdor and D. Hoffman Dated:
March 6, 1984
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