ML20069L749
| ML20069L749 | |
| Person / Time | |
|---|---|
| Site: | McGuire, Mcguire  |
| Issue date: | 06/13/1994 |
| From: | Mcmeekin T DUKE POWER CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NUDOCS 9406200213 | |
| Download: ML20069L749 (6) | |
Text
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s Duke Ptvr Company T C kfe w a n AkGarre Nuilea? Generutnn Department Vice President 1:?ou Hoger< Ferry Road ( AtGolA)
(704)875 4500 Hunterhile, AC2YO M3G (704)M5-4809 hx DUtW POWER June 13, 1994 Document Control Desk U.S.
Nuclear Regulatory Commission Washington, D.C.
20555
Subject:
McGuire Nuclear Station Docket Nos:
50-369 and 370 Supplement to Technical Specification Amendment ECCS - Pump Runout
Dear Sir:
By letter dated November 11, 1993, Duke Power submitted a request for change to Technical Specification 4.5.2.f and 4.5.2.h.
Based on the 4/26/94 NRC request for additional information and based on subsequent technical discussions regarding the subject matter, enclosed are technical responses from our engineering staff as requested.
A copy of this technical information will be provided to the appropriate North Carolina State official.
Should you have any questions, please contact Dwin Caldwell at (704) 875-4328 or John Sawyer (704) 382-6759.
Very truly yours, T.
C. McMeekin xc:
Mr.
S.D.
Ebneter Administrator, Region II U.S.
Nuclear Regulatory Commission 101 Marietta St.,
NW, Suite 2900 l
Atlanta, Ga.
30323 l
Mr. Victor Nerses U.S.
Nuclear Regulatory Commission i
Office of Nuclear Reactor Regulation Washington, D.C.
20555 6200213 940613 o e,s p
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(70 1)3 75-48110 llanterkdle, NC 2h0inKM (704)M75 009 fu DUl(E POWER xc (cont.):
l Mr.
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Maxwell NRC Resident Inspector McGuire Nuclear Station US Regulatory Commission Mr. Dayne Brown Division of Radiation Protection P.
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Box 27687
- Raleigh, N.C.
27611-7687 1
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Snyder Dwin E. Coldwell Jeffery J. Nolln (MNS)
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Bonesole (ONS)
John Sawyer (GO-Safety Analysis) 4 o.
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DUKE POWER RESPONSE REQUEST FOR ADDITIONAI,INFORMATION ON NOVEMilER 11,1993 APPI,1 CATION REl,ATING TO ECCS SUllSYSTEM SURVEll, LANCE REQUIREMENTS Ql.
You state that the information provided by Westinghouse and Dresser / Pacific Pumps indicated that credit could not he taken for an increased pump runout limit due to an excess suction pressure, since casitation is expected to occur on the second stage of the pump for flow rates above the initially proposed runout limits. When and in what documents was Duke Power notilled by Westinghouse and Dresser / Pacific Pumps of the changed pump runout limits? Please provide information on the NPSil limit and the amount of conservatism you will nmv hase on this limit to avoid cavitation.
A.
Duke Power was notified by Westinghouse and Dresser / Pacific in !ctiers DAP-91-074 and DCP-91-074 (Reference 1) dated October 3.1991. Appendix 1 of Reference I discusses cavitation and NPSil requimments for the centrifugal charging pumps (CCPs) and the safety injection pumps (SIPS).
Per Reference 1. Westinghouse and Dresser / Pacific recommend a NPSil of 30 feet in onler to support runout limits of 560 and 675 gpm for the CCPs and SIPS. respectively. Page 6-158 of the MNS FSAR (Refemuce 2) lists the available NPSil values for the SIPS and CCPs for the most limiting conditions. 'The minimum available NPSil for the CCPs is 45.3 f t, and the minimum available NPSil for the SIPS is 48.3 ft luth of which exceed the 30 ft mquirement.
Q2.
Discuss the basis for the change in the residual heat removal (RilR) llow rate. You indicate that the RIIR llow rate will be increased from 3975 gpm to 4025 gpm. Gl 88-17 recommends that HilR llow he reduced for midloop operation to avoid vortexing.
Will this proposed increase in flow rate have im impact on RIIR operation in mid-loop operation?
A.
As mentioned in the Technical Justifications supporting the pmposed changes to the ECCS surveillance requirements. LOCA reanalyses were perfonned to demonstrate the acceptability of the pmposed changes. The weakest MNS/CNS CCP and SIP plant data head curves wem selected for developing the LOCA injected flow predictions. Additional degradation was applied to the weakest head curves in developing the injected flow predictions in onler to build in conscivatism and pump test margin. The strongest CCP and SIP head curves were selected in evaluating runout conditions for the proposed Technical Specification changes. The test dates for the selected CCP and SIP pump head curves are as follows:
Weakest CCP 12N1 Strongest CCP 5/91 Weakest SIP 9/91 Strongest SIP 7/90 The msidual heat mmoval (RilR) pump head curw that supports the proposed TS changes is based uten the weakest vendor data RilR head curve with additional degradation of approximately 12%. This head curve bounds the weakest RilR pump at MNS or CNS.
The LOCA reanalyses also incorjorated changes other than just ECCS injected now dilfemnces. 'Ihese other changes include, but are not limited to, an increase in the maximum steam generator tube plugging percentage fmm 10% to 18% an increase in cold Iof3
leg accumulator (CLA) and mfueling water storage tank (RWST) water temperatures and I
various Westinghouse LOCA EM improvements. The increased steam generator tube plugging and the increased CLA/RWST water temperature assumptions were expected to be penalties to the final PCTs. To offset the PCT penalties for the large break analysis, which was closer to the 2200 F 10CFR50.46 PCT acceptance criteria, it was decided to take credit for more RilR injected flow. The proposed Technical Specification changes j
thus reflect this additional RilR injected flow.
The pn posed increase in the RilR surveillance mquirement will have no impact on RIIR operation in mid-loop. The RilR flow rate during mid-kop operation will continue to be limited to 5 3(X)0 ppm, as described in References 3 and 4. Physically, there will be no changes to the RilR system involved with the proposed TS changes. Analytically, cmdit i
will be taken for more RilR injection flow in the LOCA analyses for higher modes of operation. For these modes of operation, the suction sources of the RilR pumps are the RWST (injection phase) and the reactor building sump (sump recirculation phase). In order to take credit for more RilR injection flow, the surveillance mquirement must be increased I
to ensure actual RilR perfomiance remains above analysis assumptions. The latest MNS ND injected flow test data, which is corrected for uncertaintics, indicates that the 4025 gpm proposed TS will be acceptable.
Q3.
You state that the LOCA reanalysis to determine the impact of the proposed TS change met the criteria of 10 CFR 50.46 including the PCT, which was below 2200 F.
Were the changes from the previous analysis such that they are small enough not to be considered to be a significant change (greater than 50 *F)? Specify the "NRC apprmed methodology" for the LOCA analysis. What were the changes in peak cladding temperature for the large and small break LOCAs as a result of the reanalysis?
A.
The LOCA reanalyses were perfomicd by Westinghouse.
The approved LOCA methodologies are given in WCAP-10266 (Reference 5) and WCAP-10054 (Refemnce 6).
The current Westinghouse large break and small break peak clad temperatures, as given in Section 15,6.5 of the MNS FSAR (Reference 2), are 2132 and 1590 F, respectively For simplicity, the above PCTs do not include the effects discussed in Reference 7. The large break and small break PCTs as a result of the LOCA reanalyses are 1945 and 1264 F, respectively. Therefore, the changes in PCTs fmm the previous analyses are large enough to involve significant changes (>50 F). As mentioned in the response to question 2, the LOCA reanalyses incorporated changes other than the ECCS injected flow assumptions, and thus the differences in PCTs cannot be judged solcly by the diffen'nces in the injected flow assumptions. Duke Power has notified the NRC of the significant PCT changes via submittal of Refemuces 8 and 9.
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REl'EltENCES 1.
DAP-9 l-074, DCP-9 l-074. D. L. Fuller (Westinghouse) to R. C. Futrell (Duke), " Emergency Core Cooling System Pump Runout Limit issues," October 3,1991.
2.
Final Safety Analysis Repon, McGuire Nuclear Station, September 1,1993.
3.
McGuire Nuclear Station Procedure OP/li U6200/D1.
4.
McGuire Nuclear Station Procedure OIYEVO 00/nl.
5.
Kabadi, J. N., et al, "'llic 1981 Version of the Westinghouse ECCS Evaluation M(xlel Using the B ASil Code." WCAP-10266-P-A, Rev. 2, March 1987.
6.
N. lxe, et al, " Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code," WCAP-10054-P-A, August 1985.
7.
Letter imm M. S. Tuckman (Duke) to USNRC,"McGuire Nuclear Station Docket Numbem 50-369 and -370, Catawba Nuclear Station, Docket Numbers 50-413 and -414, Repon Pursuant to 10 CFR 50.46, Changes to or Errors in an ECCS Evaluation Model," October 18,1993.
8.
Letter fmm D. L. Rehn (Duke to USNRC, " Catawba Nuclear Station Docket Nos. 50-413 and 50-414, Technical Specifica' ion Amendment, CLA Water Volume and ECCS Subsystem Surveillance Requirements," October 5,1993.
9.
Letter fmm T. C. McMeckin (Dtike) to USNRC, "McGuire Nuclear Station, Units I and 2, Docket Nos. 50-369 and 50-370 Pmposed Technical Specification (TS) Amendment, ECCS Subsystem Surveillance Requirements (TS 4.5.2f and 4.5.2h)," November i1,1993.
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