ML17334B294
| ML17334B294 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 01/30/1989 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML17334B293 | List: |
| References | |
| NUDOCS 8902030121 | |
| Download: ML17334B294 (7) | |
Text
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATING TO RE UEST TO CHANGE LICENSING BASIS FOR SI AND RHR CROSSTIES INDIANA MICHIGAN POWER COMPANY DONALD C.
COOK NUCLEAR PLANT UNITS NOS.
1 AND 2 DOCKETS NOS.
50-315 AND 50-316
1, 2, 8) that the NRC staff concur with their interpretation of Technical Specifications (TSs) 3.5.2.e and 3.5.3.d for the Donald C.
Cook Nuclear Plant, Units 1 and 2.
TS 3.5.2 states that two emergency core cooling system (ECCS) subsystems must be operable; it defines an operable ECCS subsystem as including one operable safety injection (SI) pump, one operable residual heat removal (RHR) pump, and associated flow paths.
This TS also allows the operator to remove one ECCS subsystem for up to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> while in Modes 1, 2, or 3 while maintaining an operable flow path for the opposing subsystem.
TS 3.5.3 states that two ECCS subsystems must be operable; it defines an operable ECCS subsystem as including one operable RHR pump and associated flow paths.
This TS allows the operator to remove one ECCS subsystem for up to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> while in Mode 4 while maintaining an operable flow path for the opposing subsystem.
The licensee is proposing to interpret TS 3.5.2 such that the flow path portion of one ECCS subsystem is considered operable when one SI pump or one RHR pump can deliver flow to only two reactor coolant system (RCS) loops while the other pump within the same subsystem is aligned to deliver flow to four loops.
The licensee is proposing to interpret TS 3.5,3 such that the flow path portion of one ECCS subsystem is considered operable when an RHR pump can deliver flow to only two RCS loops.
The licensee also requested that the interpretation be extended to any configuration (such as servicing an RHR heat exchanger) in which flow from a single operating RHR or an SI pump can be delivered to only two RCS loops with the other pump aligned to deliver flow to four loops.
The licensee is requesting to use this interpretation of these TSs so that they would be allowed to close crosstie valves and to operate in this configuration which would permit them to continue uninterrupted operation while performing maintenance and required testing on these systems.
- 2. 0 EVALUATION Closing either RHR crosstie valves (IM0-314, -324) or SI crosstie valves (IMO-270, -275) results in the delivery of emergency core coolant into two instead of four loops thus reducing the flow from that used in the Final Safety Analysis 8902030i2i 890i30 PDR ADOCK 050003i5 P
Report (FSAR) analysis.
Therefore, the licensee evaluated the closure of either low head safety injection (LHSI) or high head safety injection (HHSI) crosstie valves on small and large break Loss-of-Coolant Accident (LOCA) performance, including the impact of crosstie closure on containment long term pressure.
ll ~L" B
k LOCA The D.
C.
Cook SI system consists of two RHR pumps, two centrifugal charging (CCP) pumps and two HHSI pumps.
Each HHSI pump discharge line splits to deliver flow into two of the four cold legs, and a crosstie connects the two pump discharge lines enabling one pump to deliver flow to all four cold legs.
The RHR pump discharge piping is configured the same way as the HHSI pumps.
The design basis large break LOCA analyses assume that flow delivery is available through all four lines from each pump in the SI system.
American Electric Power Service Corporation contracted Westinghouse to evaluate the D,
C.
Cook Unit 1 ECCS performance following a large break LOCA for a scenario in which the HHSI or RHR crosstie line is closed during normal full power operation.
This analysis showed that the only significant impact of the reduced pumped flow was the delay incurred in the vessel refilling while the accumulators are inje~ting.
The result of the slower downcomer fill is an overall penalty of 10 F in calculated peak clad temperature (PCT).
The PCT for the large break LOCA for D.
C.
Cook including )his penalty is 1947 F, so a margin remains to the regulatory limit of 2200 F.
American Electric Power Service Corporation contracted Advanced Nuclear Fuels Corporation to perform the large break LOCA analysis for D.
C.
Cook Unit 2.
The result of this analysis was a maximum PCT of 1988 F which also shows a
margin to the 2200 F regulatory limits.
2.2 Small Break LOCA There are two HHSI pumps in the D.
C.
Cook Unit design.
Each HHSI pump discharge line splits to deliver flow into two of the four cold legs.
A crosstie connects the two pump discharge lines enabling one pump to deliver flow to all four of the cold legs.
The design basis small break LOCA analyses assume that HHSI flow delivery is available through all four lines.
American Electric Power Service Corporation contracted Westinghouse to evaluate the D.
C.
Cook Units 1 and 2
ECCS performance following a small break LOCA for a scenario in which the HHSI crosstie line is closed during normal full. power operation.
Their small break LOCA analysis was performed for a reference plant applying the limiting four-inch equivalent diameter cold leg break for the D.
C.
Cook Units 1 and 2 licensing basis WFLASH analysis.
The four inch break was chosen since it is also the limiting break size for the reference four loop plant when analyzed with the NOTRUMP evaluation model.
Their analysis did note that a reduction in SI can, in some instances, cause a shift in limiting break size towards the smaller breaks, but Westinghouse determined that the four inch break we'll remain limiting.
The result of this analysjs was a calculated PCT of 1427 F for the four inch break in Unit 1 and a 1482 F
PCT in Unit 2.
The NRC staff was concerned that breaks smaller than four inches which do not rely upon
accumulator injection for recovery could possibly be more limiting than the four inch break.
As a result of these NRC staff concerns, Westinghouse performed analyses (Refs.
3 and 4) of break sizes less than four inches.
This evaluation showed that smaller breaks that do not result in depressurization of the RCS to the accumulator setpoint will not result in the most limiting PCT.
The four inch diameter cold leg break will remain the limiting break size with the HHSI crosstie valves closed.
Previous evaluations performed by Westinghouse have determined that the effects of the HHSI crosstie closure will cause the PCT for the three inch and four inch breaks to increase.
However, the PCT increase for the four inch case was determined to be greater than the increase for the three inch case.
Therefore, it was concluded that the four inch diameter cold leg break will remain the limiting small break LOCA event for D.
C.
Cook Units 1 and 2 with the SI flow reduction resulting from closure of th~ HHSI crosstie line and that a margin exists to the regulatory limit of 2200 F.
2.3 Containment Lon Term Pressure The licensee's containment analysis (Refs.
8 and 9) is intended to address two issues:
- 1) operation with the crosstie valves closed, and 2) operation at reduced RCS temperature and pressure.
Operation of Unit 1 at reduced temperature and conditions is proposed by the licensee for its next fuel cycle (Cycle ll).
The NRC staff finds that reduced temperature and pressure operation is not a concern from the standpoint of peak containment pressure because operation at reduced RCS pressure and temperature results in lower mass and energy releases and, in turn, a lower peak containment pressure following a LOCA.
With regard to operation with the crosstie valves closed, there will be less SI flow available for steam condensation, which will affect the mass and energy releases following a LOCA.
To evaluate the impact of this on the containment
- pressure, the licensee performed a calculation of mass and energy releases followed by an analysis of peak containment pressure.
In the mass and energy release calculations, the licensee used the method described in Westinghouse Topical Report WCAP-10325 with one exception.
The NRC staff previously found WCAP-10325, entitled "Westinghouse LOCA Mass and Energy Release Model for Containment Design - March 1979 Version," to be acceptable.
The safety evaluation report for WCAP-10325 was transmitted to Westinghouse by a letter dated February 17, 1987.
The exception noted by the licensee has to do with the steam/water mixing model used in the broken loop.
In their analysis, the.licensee assumed a complete thermal equi librium mixing condition for the steam and emergency core cooling injection water during the reflood phase.
This condition was described in WCAP-10325, However, at that time, Westinghouse took credit for steam/water mixing only in the intact loop and not in the broken loop.
In the analysis for D; C.
Cook, Westinghouse justified the applicability of the above model by the 1/3 scale test data documented in the report EPRI 294-2, entitled "Mixing of Emergency Core Cooling Water with Steam:
1/3 Scale Test and Summary."
The test data were reviewed and discussed in WCAP-10325, which indicated that very effective mixing, rapid steam condensation, and complete thermal equilibrium were obtained at a very short distance downstream of the injection nozzle.
The licensee also provided additional justification for using the mixing model in the broken loop in this analysis.
The limiting break for the containment peak pressure analysis is identified as the double ended reactor coolant pump suction line break.
As this break location, and because of the nearby location of fhe emergency core cooling injection nozzle, the released steam via reverse flow through the reactor coolant pump encounters emergency core cooling injection water as it passes through the broken loop cold leg.
Complete mixing occurs, and a portion of it is condensed.
It is this portion of steam which is condensed that is taken credit for in the analysis.
Based on the test results presented in WCAP-10325 and the justification regarding the conditions for the limiting break location, steam/water mixing in the broken loop is considered a realistic characterization.
Therefore, the NRC staff finds this analysis assumption to be acceptable.
The licensee's calculation of mass and energy release values included both cases of maximum and minimum SI flowrate.
For the case of minimum SI flow, the RHR crosstie valve is assumed to be closed, and the pump discharge head is assumed to be degraded by lOX.
The assumption of a closed RHR crosstie valve is more limiting than that of a closed SI crosstie valve because it results in a greater reduction of SI flow.
For the case of maximum SI, the crosstie valves are assumed to be open, with no degradation in pump discharge head.
The mass and energy releases for the case of minimum SI.were determined to be bounding.
The licensee stated that the assumptions used in its calculation of mass and energy releases are consistent with the guidelines of Standard Review Plan (SRP)
Section 6.2. 1.3.
The NRC staff finds the above approach for determining mass and energy release rates to be consistent with staff guidance and, therefore, acceptable.
The licensee's containment peak pressure analysis used the Westinghouse computer code LOTIC and the calculated mass and energy release values as part of the input.
The LOTIC code is described in Westinghouse Topical Report WCAP-8354-P-A, entitled "Long Term Ice Condenser Containment Code," which was found acceptable by the NRC staff in a safety evaluation report transmitted to Westinghouse by letter dated January 28, 1976.
The worst single failure assumed in the containment analysis is the failure of one diesel train, which provides minimum post-accident engineering safety features, i.e.,
one of two containment spray
- pumps, one of two containment spray heat exchangers, and one of two containment air return fans.
The assumptions used in the licensee s containment analysis are consistent with the guidelines of SRP Section 6.2.1. l.b.
Based on these assumptions, the calculated peak containment pressure was 11.89 psig which is less than the design pressure of 12.0 psig.
Therefore, the NRC staff finds the peak containment analysis to be acceptable.
Based on the above, the NRC staff concludes that operation of the Donald C.
Cook Nuclear Plant, Units 1 and 2, with the RHR or SI line crosstie valves closed is acceptable since post-LOCA containment conditions are within acceptable limits and in accordance with the requirements of General Design Criteria 16, 38 and 50 for containment performance.
Operation at reduced RCS temperatures and pressures is not a concern from the standpoint of peak containment pressure because of the resulting lower mass and energy releases, and therefore, no further consideration in this regard is necessary.
3.0 NRC STAFF POSITION The NRC staff has reviewed the licensee's analysis of closing the crosstie valve in either RHR of HHSI pump delivery systems.
This closure results is a slight reduction in LOCA mitigation performance and an increase in calculated PCT and containment pressure, but there still exists an acceptable PCT margin to the regulatory limits of 2200 F in Appendix K to 10 CFR Part 50 and 12. 0 psig containment design pressure.
Therefore, the NRC staff approves the licensee's request to allow closure of either RHR or the HHSI crosstie valves but not both crossties in D.
CD Cook Units 1 and 2.
Furthermore, the containment long term pressure analysis is acceptable for supporting operation of the D.
C.
Cook Nuclear Plant at reduced temperature and pressure (RTP) conditions.
This approval does not permit operation at RTP conditions.
Additional analyses supporting the RTP program were submitted by the licensee on October 14, 1988, and are currently under staff review, and will be the subject of a future licensing action
4.0 REFERENCES
1.
- Letter, M.
P. Alexich (IMEC) to H.
R.
Denton (NRC), AEP:NRC: 1024, March 23, 1987.
2.
- Letter, M.
P. Alexich (IMEC) to T.
E. Murley (NRC),
MEP: NRC: 1024A, May 13, 1987.
3.
- Letter, M.
P. Alexich (IMEC) to T.
E. Murley (NRC),
AEP:NRC: 1024C, October 13, 1987.
4.
- Letter, M.
P. Alexich (IMEC) to T.
E. Murley (NRC), AEP:NRC: 1024E, February 29, 1988 5.
"Westinghouse LOCA Mass and Energy Release Model for Containment Design-March 1979 Version," WCAP-10325-P-A, May 1983.
6.
EPRI 294-2, Mixing of Emergency Core Cooling Water with Steam:
1/3 Scale Test and
- Summary, (WCAP-8423), Final Report June 1975.
7.
"Long Term Ice Condenser Containment Code - LOTIC Codes,"
WCAP-8354-P-A, April 1976.
8.
- Letter, M.
P. Alexich (IMPC) to T.
E. Murley (NRC), AEP:NRC: 1024D, August 22, 1988.
9.
- Letter, M.
P. Alexich (IMPC) to T.
E. Murley (NRC),
AEP:NRC: 1024F, January 12, 1989.