ML20236U779
| ML20236U779 | |
| Person / Time | |
|---|---|
| Site: | Pilgrim |
| Issue date: | 11/30/1987 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20236U776 | List: |
| References | |
| NUDOCS 8712030367 | |
| Download: ML20236U779 (3) | |
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UNITED STATES
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8 NUCLEAR REGULATORY COMMISSION g
j WASHINGTON, D. C,20555 5
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SAFETY EVALUATION BY THE OrFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO.111 TO FACILITY OPERATING LICENSE NO. DPR-35 BOSTON EDISON COMPANY PILGRIM NUCLEAR POWER STATION DOCKET NO. 50-293
1.0 INTRODUCTION
By letter dated July 8, 1987, supplemented by letter dated August 5, 1987, Boston Edison Company (the licensee) submitted Change Request No.
i A7-12 to Facility Operating License No. DPR-35, requesting that Technical Specification (TS) section 4.5. A.3.d, Low Pressure Core In.iection comply with the current Loss of Coolant Accident (LOCA) analysis for the Pilgrim Station.
The proposed change consists of modifying TS surveillance requirement 4.5.A.3.d to specify single LPCI pump performance. The current TS specifies pump performance based on three LPCI pumps.
1 2.0 EVALUATION The proposal changes the TS surveillance requirement 4.5.A.3.d from "Threa l
LCPI pumps sna11 deliver 1A,400 gpm against a system head corresponding to a vessel pressure of 20 psig" to "Each LPCI pump shall pump 4800 gpm at a head across the pump of at least 380 ft."
The licensee states that the current TS requirement is based on l
Pilgrim's original core LOCA analysis in which flow from three LPCI pumps was a limiting case.
The current LOCA analysis, (reviewed in connection with Amendment No. 105 to the Facility Operating License, pertaining to Core Reload 7) does not address the configuration of three l
l pumps infecting into the vessel.
Instead, two and four pump I
configu~ rations are addressed. Therefore, the present wording of the TS I
is not consistent with the current LOCA analysis.
The LOCA analysis update provided in NE00-30767 to support core reload 7 l
for coerating cycle 8 confirmed that the limiting break which is the design basis accident is the complete severance of the recirculation suction line assuming a failure of the LPCI injection valve. The required analysis is in accordance wit'. the requirements of NEDE-24011-P-A-8, General Electric St W ard Application for Reactor Fuel m2988E Bla!!b P
, (GESTAR-11), dated May 1986. NEDO-30767 was reviewed in connectiori with.
Amendnent No. 105 to Facility Operating License No. DPR Pilgrim Nuclear Power Station, dated August 31, 1987, related to Cycle 8-operation. The associated methodology includes an evaluation of other single failures for which two pump and four pump LPCI configurations are used.
In addition, the licensee states that the multiple. pump configuration is not directly tested.
Instead, each pump is tested individually with the
'i results translated by intermediate calculations to verify that multiple
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configuration requirements are satisfied.
The staff review of the proposed change and calculations of LPCI pump performance submitted August 5, 1987, confirmed that the existing technical specification is'not consistent with the current LOCA analysis.
The staff review confirmed that the proposed change does not reduce the LPCI pump capacity requirements. The proposed specification is based;on the licensee's calculations of the pressure drop with three pumps injecting
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into the reactor pressure vessel at 20 psig. To assure conservatism, the licensee used the three pump configuration with the greatest pressure drop for the calculations, and a friction loss of 20% above that for a clean pipe was assumed. The pump head-flow relationship of the pump' design curve was used to determine that the required flow of 14,400 GPM j
could be obtained with an 11% degradat. ion of pump performance.. The j
proposed specification limits the degradation in head-flow performance to 4
9.5% for conservatism.
Therefore, the proposed chenae restates LPCI pump capacity, but does_not change the capacity from that used in Pilgrim's current LOCA analysis.-
Consequently compliance with the LOCA analysis is assured. -The proposed change also assures e riore timely determination of acceptable pump performance without the necessity of intermediate calculations to verify that the multiple configuration performance requirements are met. Additionally, the chance is consistent with current guidance for specifying pump performance provided in NUREG-0123, " Standard Technical' Specification for General Electric Boiling Water Reactors." The staff therefore concludes that the proposed change to TS 4.5.A.3.d is acceptable.
3.0 ENVIRONMENTAL CONSIDERATION
S This amendment involves a change in the Technical Specifications of facility components located within the restricted area as defined in 10 CFR Part 20 and changes a surveillance requirement. -The staff has determined that the amendment involves 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 occupational radiation exposure. The Commission has previously published a proposed finding that the amendnent involves no significant harards consideration and there has been no public coment on such finding. Accordingly, the?
anendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR %51.??(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.
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4.0 CONCLUSION
Ue 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'Connission's regulations, and.the issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public.
5.0 REFERENCES
(1) NUREG-0123, Standard Technical Specifications for General Electric Boiling Water Reactors, Revision.3.
(2) Pilgrim Nuclear Power Station, Updated Final Safety Analysis Report, Chapters 6 and 14 (3) NED0-30767, " Pilgrim Nuclear Power Station Loss of Coolant Accident (LOCA) Analysis Update". September 1984.
(4) 23A 4800 Rev 0 Class I, December 1986, " General Electric Boiling Water Reactor Supplemental Reload Licensing Submittal for Pilgrim Nuclear Power Station Reload 7".
Principal Contributor: Michele 6. Evans, Region I Dated:
November 30, 1987 i
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