ML20113D386

Responds to RAI Re GL 95-07, Pressure Locking & Thermal Binding of Safety-Related Power-Operated Gate Valves

ML20113D386
Person / Time
Site:	Cooper
Issue date:	06/27/1996
From:	Mueller J NEBRASKA PUBLIC POWER DISTRICT
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20113D390	List: ML20113D386 ML20113D450 ML20113D490
References
GL-95-07, GL-95-7, NLS960122, TAC-M93453, NUDOCS 9607030140
Download: ML20113D386 (5)

Text

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COOPER NUCLE AR STATON 7 P.O. BOX 98, BROWNVLLE. NE BRASKA 68321 Nebraska Public Power District TAM"" .

NLS960122 June 27,1996 U. S. Nuclear Regulatory Commission ATfN: Document Control Desk Washington, DC 20555-0001

Subject:

Response to Request for Additional Information - Generic Letter 95-07 Cooper Nuclear Station, NRC Docket 50-298, DPR-46

References:

1. Letter from D. L. Wigginton to G. R. Ilom, dated May 28,1996," Request for Additional Information - Generic Letter 95-07, ' Pressure Locking and Thermal Binding of Safety-Related Power-Operated Gate Valves,' Cooper Nuclear Station (TAC No. M93453)"

2. Letter from G. R. Horn to NRC Document Control Desk, dated February 13,1996,"I 80 - Day Response to Generic Letter 95-07" Gentlemen:

Attached is Nebraska Public Power District's (the District's) response to your request for additional information (Reference 1) conceming the District's 180 - day response to Generic Letter 95-07 (Reference 2).

Should you have any additional questions conceming this matter, please contact my office.

Sincerely, d.IIL Mueller Site Manager

/cct Attachment 9607030140 960627 PDR ADOCK 05000298 P PDR qts.W # "

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!. U. S. Nuclear Regulatory Commission June 27,1996 Page 2 of 2 cc: Senior Project Manager USNRC - NRR Projects Directorate IV-1 l

Senior Resident Inspector USNRC - Cooper Nuclear Station Regional Administrator j USNRC - Region IV NPG Distribution l

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Attachment 1 to NLS960122 Page 1 of 2 Fallowing are the NRC staff's requests for additional information and the Dictrict's response:

1. The licensee 's submittal states that steady-state heat transfer calcidations were performedfor HPCI-Af0V-MO58, HPCISuppression PoolSuction Valve, and RCIC-A10V-A1041, RCICSuppression PoolSuction Valve, which conservatively demonstrated that the heat dissipates rapidly in the suction lines and no increase in valve temperature above ambient will occur. Please provide these heat transfer calcidationsfor our review.

In addition, please provide the basisfor the assumption of121 degrees F as the marimum design basis accident temperaturefor the suppressionpool.

Revision 1 of NPPD calculation NEDC 96-001," Determination of Dead Leg lieat Transfer in Piping Associated w/ IIPCI-MOV-MO58 and RCIC-MOV-MO41"is attached as Appendix A.

The District's 180 - day response to GL 95-07 indicated that 121 degrees F was the assumed m:.ximum design basis accident temperature for the suppression pool. Revision 0 of NPPD calculation NEDC 96-001 incorrectly referenced 121 degrees F as the maximum design basis accident suppression pool temperature for the IIPCI/RCIC response scenario. The correct maximum temperature is 116 degrees F. However, the actual suppression pool temperature  ;

assumed in calculation NEDC 96-001 to determine the temperature profile in the HPCI and  !

RCIC suction lines is 140 degrees F. 140 degrees F is more conservative and is based on the l temperature above which NPSII for the llPCI and RCIC systems can no longer be assured.

NEDC 96-001 has been revised to clarify that 116 degrees F is the maximum suppression pool temperature for the HPCl/RCIC response scenario. The calculation remains unchanged with 140  ;

degrees F assumed as the actual temperature to determine the temperature profile in the HPCI l

and RCIC suction lines.

2. HPCISteam Isolation Valve, HPCI-M014, ifaflexible-wedge, split wedge, or double-disk gate valve, may be potentially susceptible to thermally-inducedpressure locking ifit exists in a configuration where condensate could become trapped in the valve bonnet.

ifIso, this valve, ifaflexible-wedge or solid wedge gate valve, may be potentially susceptible to thermal binding ifit is openedfor HPCItesting, closed while hot, and subsequently allowed to cool. This valu: was not identified as susceptible to nressure locking or thermal binding in the licensee 's ":bmittal. Please provide the wails ofthe pressure locking / thermal binding reviewfor this valve.

HPCI-MOV-M014 is a flexible-wedge gate valve. HPCI-MOV-Mol4 is located in a configuration where condensate cannot be trapped in the valve bonnet. A steam line drip leg drain pot is located directly upstream of HPCI-MOV-M014 to prevent the accumulation of steam condensate from occurring in the valve bonnet. The drip leg contains an alarm to alert operations personnel should the drip leg fail to drain properly. Therefore, this valve is not considered to be susceptible to pressure locking. Additionally, this valve is not considered susceptible to thermal binding. While in the closed, standby position following IIPCI system surveillance operation, valve temperature is maintained by exposure to reactor steam at nominal

Attachment I to NLS960122 Page 2 of 2 operating pressure and temperature on the upstream side of the insulated valve. Therefore, the cooling effect necessary to cause thermal binding does not exist.

3. Through review ofoperational experiencefeedback, the staffis aware ofinstances l where licensees have completed des l;;n orprocedural modifications to preclude pressure locking or thermal binding which may have had an adverse impact on plant safety due to incomplete or incorrect evaluation ofthe potential effects ofthese modifications. Please describe evaluations and trainingforplantpersonnel that have been conductedfor each design or procedurid modification completed to addresspotentialpressure locking or l thermal binding concerns.

In 1995, physical modifications were performed on four valves that were determined to be potentially susceptible to pressure locking (CS-MOV-MOl2A and B, IIPCI-MOV-MOl9, and RCIC-MOV-MO21). In each case the physical modification involved the placement of a one quarter inch hole in the downstream (reactor) side of the flexible wedge disk. The safety evaluation for this modification included consideration of maintaining the tight shutoff capability for each valve given that one seating surface would be rendered nonfunctional by the one quarter inch hole. Periodic Appendix J testing of CS-MOV-M012A and B assures that the remaining functional seat is performing satisfactorily. IIPCI-MOV-MOl9 and RCIC-MOV-MO21 were tested subsequent to the modification as part of the ASME Section XI 10-year hydrostatic test.

Additionally, the remaining functional seat for each valve modified is assisted in the closed direction by reactor pressure. Training on the modification was provided to licensed operators.

In 1993, a procedure modification was performed to address pressure locking concerns with RHR-MOV-MO25A and B. The procedure modification altered the lineup of existing valves to vent the bonnet volume to the downstream (reactor) side of the flexible wedge disk. The potential for increased packing leakage was recognized due to the venting of the bonnet volume to the reactor side of the valve.

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l. LIST OF NRC COMMITMENTS l ATTACHMENT 3 l l

Correspondence No: NLS960122 The following table identifies those actions committed to by the District in this document. Any other actions discussed in the submittal represent intended or planned actions by the District. They are described to the NRC for the NRC's information and are not regulatory commitments. Please notify the Licensing Manager i

at Cooper Nuclear Station of any questions regarding this document or any associated regulatory commitments.

COMMITTED DATE COMMITMENT OR OUTAGE None l

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l PROCEDURE NUMBER 0.42 l REVISION NUMBER 1.1 l PAGE 9 OF 11 l