ML19242C006
| ML19242C006 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 07/16/1979 |
| From: | Ziemann D Office of Nuclear Reactor Regulation |
| To: | Counsil W NORTHEAST NUCLEAR ENERGY CO. |
| References | |
| TASK-03-10.C, TASK-RR NUDOCS 7908090648 | |
| Download: ML19242C006 (6) | |
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.,%, V July 16,1979 Docket No. 50-245 M. W. G. Counsil, Vice President Nuclear Engineering and Operations Northeast Nuclear Energy Company Post Office Box 270 Hartford, Connecticut 06101
Dear Mr. Counsil:
RE: TOPIC III-10.C - SURVEILLANCE REQUIREMENTS ON BWR RECIRCULATION PUMP DISCHARGE VALVES Enclosed is a copy of our revised safety assessment of Topic III-10.C, Surveillance Requirements on BWR Recirculation Pump Discharge Valves.
This revision includes ccnsideration of the comments received on the assessment issued by our letter dated August 17, 1978.
Your letter dated November 20, 1978, provided comments on the assessment.
This revision completes our assessment of Topic III-10.C which will be used as input to the integrated review of the Millstone Plant.
If there are any errors in the facts of this revised assessment, please supply corrected infomation within 30 days of the date you receive this letter.
I, no response is received within that time, we will assume that you have no fur'ther comrents or corrections.
Si ncerely, e
Dennis L. Zieman, Chief Operating Reactors Branch !2 Division of Operating Reactors Encl osure:
Revised Assessment for Topic III-10.C cc w/ enclosure:
See next page 7 ri, 'i ET4 Jj'-
t 7 9080 90 [b
Mr. W. G. Counsil July 16,1979 cc William H. Cuddy, Esquire Day, Berry & Howard Counselors at Law One Constitution Plaza Hartford, Connecticut 06103 Anthony Z. Roisman Natural Resources Defense Council 917 15th Street, N. W.
Washington, D. C.
20005 Northeast Nuclear Energy Company ATTN:
Superintendent Millstone Plant P. O. Box 128 Waterford, Connecticut 06385 Mr. James R. Himmelwright Northeast Utilities Service Company P. O. Box 270 Hartf ord, Connecticut 06101 Nuclear Regulatory Commission, Region 1 Office of Inspection and Enforcement ATTN: John T. Shedlosky 631 Park Avenue King of Prussia, Pennsylvania 19406 Waterford Public Library Rape Ferry Road. Route 156 Waterford, Connecticut 06385 K M C Inc.
ATTN: Mr. Richard E. Schaffstall 1747 Pennsylvania Avenue, N. W.
Suite 1050 Washington, D. C.
20006
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SYSTEMATIC EVALUATION PROGRAM Topic III-10.C Surveillance Requirements on BWR Recirculation Pump uischarge Valves Plant: Millstone Nuclear Power Station Unit No. 1 DISCUSSION In July 1976, all BWR facilit es which had completed the Low Press.re Coolant Injuction (LPCI) systen modification, to remove the LPCI loop selection logic, were sent letters recuesting that they apply for a license amendment to incorporate technical specification surveillance recuirements on recirculation pump discharge and bypass valves. The recirculation pump discharge and bypass valves are required, for these plants, to close upon initiation of LPCI. The closure of these valves is necessary to prevent the loss of cooling water by reverse flow through the pumo or its bypass line and o ut the break. The failure of the recirculation pump discharge or bypass valve to close can adversely affect core cooling in a manner similar to the failure of a L?CI valve to cpen.
EVALUATICN This topic apolies t: LPCI modified SWR facilities. Mi'istone Unit No. I ratains LPCI loop 3. election logic and therefore is not subject to the requirements of this topic. Furtherr.. ore, since the unmodified LPCI is susceptible to single failures that can ':!minate all LPCI flow, no credit is given for any LPCI flows in the Millstone Unit No. 1 ECCS analysis. The staff's safety evaluation for Millstone Unit No. I describes the design basis event as the complete severance of the recirculaticm suction linc assuming a failure of the LPCI injection valve (safety evaluation transmi'.ted to licensee October 17,1975). An assuned failure of 'ae LPCI valve pre-vents any LPCI flow from entering the core. An NRC SM e y Evaluation dated December 27, 1974, 'or Millstone Unit No. I discussis the acceptabiTity of the ECCS model used for the above assumptions. The October 17, 1975 evaluation concluded that, with appropriate technical specification changes, Millstone Unit No. I met the performance requirements of 10 CFR 50.46 (Acceptance Criteria and Fmergency Core Cooling Systems for Light Water Nuclear Power Reactors).
The LPCI logic network is designed to direct LPCI flow to the intact recirculation loop in the event of a loss of coolant accident (LOCA). The logic network also was designed to close the suction and discharge valves of the intact loop to prevent LPCI flow from bypassing the core and ficwing out the break. 75e staff review of Topic III-10.C indicates that since the LPCI loop selection logic has not been modified at Millstone Unit No. 1, the primary concern is not applicable as discussed above. However, a different i
2.
requirement does apply to Millstone Ur.it No.1.
The staff has required that all SWR-3's perform a modification tr ensure that the recirculation line suction valves remain open when LPCT is initiated on a LCCA signal.
Motor-operneo valves are placed or SWR recirculation suction and discharge lines. Fellowing a loss-of-coolan; accident (LCCA), if either of these valves or the unbroken recirculati3n line closes and if the low pressure coolant injection system (LPCI) su) plies ECCS water to that loop, then the LPCI *aater will flow through the j tt pump nozzles into the lower plenum where it will centribute to core raflooding.
If neither of the valves closes, the LPCI water could flow backwards through the unbroken loops' recircJlation pump, around the downComer, and out the brcak, thereby not contributing to core reflooding. To provice redundancy, BWR-3 ECC2 designs incorporated automatic closure of both the suction and discharge valves (on the unbroken loop only) jpon receipt of a LCCA signal.
However, assumed single failure ;f the loop selection logic system can result in selectico of the wrong loop as the broken icon. This would cause the following two events:
1.
All LPCI flow from both LPCI systems would be directed to the broken loop and would be lost out the break.
This effect has beer. :ansidered in SWR-3 ECCS-LCCA analyses; as a result, no credit is assumed for LPCI flow.
2.
Both the suctior /alve and the discharge valve on the broken recircu-lation line would close.
If the break location were between those two valves, the break would be isolated from the reactor vessel. Although this could L:. advantageous under certain conditions, under other conditions it could irtroduce undesirable effects which have not been adequately considered in previously performed ECCS-LCCA calculations. That is, for a certain range of break sizes, it :s possible that core uncovery could occur with vessel pressure above the LPCI pump shutoff head.
If break isolation were to occur at that time, LPCI flow could be delayed and/or reduces, resulting in a later core reflooding and a higher PCT.
With respect to Item 2 above, compensating effects exist that partially or wholly compensate for the above undesirable e"fects. The Feedwater Coolant Injection (FWCI) and the Automatic Pressure Relief (APR) would complete depressurization to the point where LPCI could function. Although such LPCI operation would be delayed, credit can be asse ad for the full complement of ECCS equipment since the required single failure has already been assumed (loop selection logic failure, selection of the wrong loop).
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... Preliminery calculations indicate + hat the above described compensating effects would SesultinPCTis for the w rs.,ize is latable break that are below 22C0 F.
However, a fully approve ' - M1 meeting all requirements of Appendix K to 10 CFR 50.46 does not exist which it capable of calculating a postulated break that becomes isolated. Also, the preliminary calculations were not performed for all sizes of SWR-3's.
Consequently, it is not possible to categorically state that 10 CFR 50.46 requirements are met for all isola-table breaks for all BWR-3's Therefore, General Electric Company reccmmended, and we require, that the autcmatic closure feature on the suction valve be disabled.
This makes break isolation a non-credible event which does not recuire analysis: Two independent failures are necessary, i.e., closure of 9e discharge valva in the broken loop (requiring 1000 selection logic failur,), and closure of the suction valve in the same loop (for example, by operato error).
No cre.iit has been assured for closure of the suction va.ve in any safety analyses other than ECCS-LOCA analyses.
For ECCS-LOCA analyses, closure of the suction valve provided a backup function for closure of the discharge valve on the unbroken loop. Witn the recommended mcdification (suction valve closure disconnected), singla failce to close of the discharga valve on the unbroken loop will now cause failure cf the LPCI system. However, this LPCI failure has already been taken into account by the ECCS-LOCA analyses on all 3WR-3 plants. No credit is assumed for L?CI operation on SWR-3 plants, since singl( failure of the loop selection logic can cause complete failure of LPCI. Stated another way, the recocrended change merely creates another potential path
.to a failure that is already accounted for in the ECCS-LCCA analyses, that is failure of LPCI; however, the recommended change precludes possibility of an event which has not been acounted for in the analyses, i.e., break isolation.
By letter dated April 24, 1978, Northeast Fuclear Energy rompany informed the NRC that on April 21,'1978, the breakers for the recirculation system suction valves were racked out with the valves in the opan position for the Millstone Unit No. I facility.
On May 24, 1979, we were infor ed (telecon: N., DiBenedetto, NUSCO, Dente and McGuinness) that the LPCI loop selectic.
gic was being changed, during the current outage, to negate the LPCI signal that closes the recirculat.en line suction valves. This modi #ication will permit the closure of the suction valves.by remote manual action uut will remove any automatic closure associated with LPCI system operation.
As stated above, elimination of the automatic closure feature on BWR-3 plants is a desiraole change since it aliminates the potential for an event which involves unreviewed safety concerns. The change does not create any new unreviewed safety concerns. We, therefore, find accootable Millstone Unit No. I operatica with disables suction valve automatic c 1sure following a LOCA signal.
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On the basis of ar review, we conclude that Topic III-10.C is acceptably resolved for Millstone Unit.10. I and no further action is required.
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