ML20059C857

Provides Results of Evaluation of Ability to Meet Acceptance Criteria for Eccs,In Response to 900804 Notice of Violation. Plant Meets Acceptance Criteria Contained in 10CFR50.46 W/ Valve Logic Design Deficiency in Containment Spray Sys

ML20059C857
Person / Time
Site:	Oyster Creek
Issue date:	08/24/1990
From:	Fitzpatrick E GENERAL PUBLIC UTILITIES CORP.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
C320-90-739, NUDOCS 9009050395
Download: ML20059C857 (3)

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OPU Nuclear NUC MF 100 !nterpace Parkway Parsippany New Jersey 07054 201 263-6500 TELEX 13&482 Wnter's Direct Dial Number:

August 24, 1990 C320-90-739 U. S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555 Centlement i

Subjects Oyster Creek Nuclear Generating Station Docket No. 50-219 License No. DPR-li  ;

Containment Spra? System Design Deficiency i.

l Your letter from the Director, Division of Reactor Projects, Region I, dated April 8, 1990, which forwarded a Notice of Violation (NOV), also requested our evaluation of the ability of Oyster Creek to have met the acceptance criteria for Emergency Core Cooling Systems, as specified in 10 CFR 50.46, with a design deficiency which existed in the Containment Spray System. Our letter dated May 9, 1990, which forwarded the NOV reply, indicated that our evaluation would be submitted within 60 days. This letter provides the results of our evaluation.

The evaluation is based on Containment spray System configuration at the time the design deficiency was discovered.

The containment Spray System is designed to remove energy from primary containment. It is used with the Core Spray System to remove reactor decay heat from containment to the ultimate heat sink following a loss of coolant accident (LOCA). It has two modes of operation. In the containment spray

mode, the drywell end torus are sprayed following a LOCA. System pumps would l have tripped automatically when co.tainment pressure decreased to 2 psig. In the dynamic test (torus cooling) mode water recirculates from the torus through  ;

the containment spray heat exchangers. This permits containment spray loop l operation for test purposes and for controlling torus temperature during normal L station operation.

L i The original safety analysis for containment response following a design basis l LOCA did not address the automatic containment spray pump trip as pressure

} dropa in containment. The analysis assumed Containment Spray System operation I

in the containment spray mode throughout the accident. The dynamic test mode l of operation was only expected to be used for testing and heat removal during normal operation and not for providing long term decay heat removal.

Subsequent analyses show, based on initial conditions assumed, that containment  ;

sprays can depressurize containment to the pump trip setpoint following a LOCA  !

much faster than the original analysis indicated. The emergency operating procedures have also evolved to require dynamic test mode operation if torus cooling is required af ter containment spray pump trip.

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CONTAI~ MENT SPRAY SYSTEM DkSIGN DEFICIENCY PACE 2 A review of Containment Spray System logic in March 1989 revealed that the system would not perform as expected. Following the design basis LOCA, the logic would have prevented system operation in the torus cooling mode, due to the water levcl in the reactor being lower than the low low water level setpoint fe. Containment Opray System initiation. Upon pump trip the Containment Spray System valves could not be realigned to the torus cooling ,

mode. This is consistent with the original design basis. However, this containment spray logic feature is a deficiency which would have prevented the operator from establishing the primary means of decay heat removal required by emergency operating procedures.

There are five acceptance criteria for Emergency Core Cooling Systems contained in 10 CFR 50.46: ,

1. Peak cladding temperature < 2200*F

2. Maximum cladding oxidation < 17%

3. Maximum hydrogen generation < 1%

4. Maintain coolable geometry (< 14 plastic strain in cladding)

5. Long term cooling ,

The first four requirements are satisfied within the first 20 minutes following .

the design basis LOCA when requenching of the fuel rods occurs and are uneffected by the containment Spray System valve logic deficiency discussed ,

above.

In order to meet the fifth requirement, long term cooling, operator action would be required even if the containment spray valve logic deficiency was absent. If the valve logie did not have the deficiency associated with the reactor low low water level condition, the operator would still b6 required to take a manual action to transfer the Containment spray System to the torus cooling mode after the containment spray' pumps tripped on low drywell pressure. With the presence of the valve logic problem, the operator would be required to take a manual action to override the logic or establish another water injection source to the core.

Our analysis shows that the operator han over 2 heure from the time of containment spray pump trip at 2.0 psig drywell pressure until the torus pool temperature reaches the point at whi: there is insufficient net positive suction head available to the core spray pumps for the maximum flow rate assumed in the 10 CFR 50, Appendix K LOCA analyses (4100 gpm). This is approximately 2.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> from the start of the LOCA. Based on engineering judgement, this period of time is adequate for operators and technical support staff to take action to assure that continued long term core cooling will be maintained. Manual action is not prohibited by 10 CFR 50.46 in order to meet the long term cooling requirement.

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a o CONTAINMENT SPRAY SYSTEM  !

• ** DESIGN DEFICIENCY PAGE 3 There are alternatives to restoration of the torus cooling mode of containment spray System operation within the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> time frame. Those have existed since original plant design and operation and involve establishing a reactor pressure vessel (RPV) injection source which is not affected by torus water )

temperature. One method is to align the Core Spray System to take suction from the condensate storage tank. Another method utilizes a tie-in to core spray ,

from the Fire Suppression Water System. Each of two diesel-driven fire pumps can provide approximately 1650 gpm flow to the RPV. As described in the original analysis of the core cooling capability of this method, one fire pump  ;

is adequate to ensure core cooling 20 minutes after the onset of a LOCA. j Although this analysis has not been recently updated, it suggests that the fire 4 pump source should be adequate after 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> when the decay heat removal requirement is significantly reduced. Both of the above core injection methods are already provided in the emergency operating procedures. It should be noted j that the containment spray mode remains available to keep containment pressure I low, i.e. the pumps would have cycled on and off betweer. 3.5 and 2.0 psig, I respectively.

s In conclusion, Oyster Creek meets the acceptance criteria contained in 10 CFR i 50.46 with the valve logic design deficiency in the containment Spray System. .I Manual action is required to satisfy the long term cooling criterion. 1 Suf ficient time is available to take the n.anual action. Currently, operators j are aware of this design deficiency, trained in the nothod to override the  ;

valve control logic and are provided instructions in procedures to override tie l valve logic should it be required. In addition, the containment spray pump l trip setpoint was reduced to 0.6 peig, which allows the pumps to operate I longer.

1 l l Very truly yours, l E. E. Fitzpatrick Vice President & Director Oyster Creek EEF/PFC/crb i'

(C320739C)

I cci Administrator Region I U. S. Nuclear Regulatory Commission l 475 Allendale Road L King of Prussia, PA 19406 l

l NRC Resident Inspector l Oyster Creek Nuclear Generating Station Forked River, NJ 08731 Mr. Alex Dromerick, Jr.

U. S. Nuclear Regulatory Commission Mail Station PI-137 Washington, DC 20555