ML17261B046

Responds to Issues Discussed During 900323 Telcon Re Inservice Testing Program Status & Relief Request.Current Test Methodology for Seat Leakage Acceptable Based on Application of Direct Measurement Sys

ML17261B046
Person / Time
Site:	Ginna
Issue date:	04/12/1990
From:	Mecredy R ROCHESTER GAS & ELECTRIC CORP.
To:	Andrea Johnson Office of Nuclear Reactor Regulation
References
NUDOCS 9004200442
Download: ML17261B046 (5)

Text

A.f CELEBRATED D1 BUTlON DKMOYiSTRQTION SYSTEM REGULATO INFORMATION DISTRIBUTION STEM (RIDS)

ACCESSION NBR:9004200442 DOG.DATE: 90/04/12 NOTARIZED: NO DOCKET ¹ FACIL:50-244 Robert Emmet Ginna Nuclear Plant, Unit 1, Rochester G 05000244 AUTH. NAME AUTHOR AFFILIATION MECREDY,R.C. Rochester Gas & Electric Corp.

RECIP.NAME RECIPIENT AFFILIATION Document Control Branch (Document Control Desk)

SUBJECT:

Responds to issues discussed during 900323 telcon re inservice testing program status & relief request.

DISTRIBUTION CODE: A047D TITLE: OR Submittal: Inservice COPIES RECEIVED:LTR J ENCL I Inspection/Testing/Relief SIZE:

from ASME Code D

NOTES:License Exp date in accordance with 10CFR2,2.109(9/19/72). 05000244 S

RECIPIENT COPIES RECIPIENT COPIES ID CODE/NAME LTTR ENCL ID CODE/NAME LTTR ENCL PD1-3 LA 1 0, PD1-3 PD 5 5 JOHNSONiA 1 1 INTERNAL: AEOD/DSP/TPAB 1 1 NRR/DET/ECMB 9H 1 1 NRR/DET/EMEB 9H 1 1 NUDOCS-ABSTRACT 1 1 OC 1 0 OGC/HDS2 1 0 G FILE 01 1 1 RES/DSIR/EIB 1 1 EXTERNAL: EG&G BROWNIB 1 1 EG&G RANSOME,C 1 1 LPDR 1 1 NRC PDR 1 1 NSIC 1 1 NCTE LO ALL "RIES" RECIPIENTS:

PLEASE HELP US K) REDUCE KG~i. CGPI'ACZ 'XHE DOCX3KKZ CONTROL DESKS RQCN Pl-37 (EXT. 20079) K) EIZMZNAXR YOUR NME HKH DISTRIBUTION LISTS FOR DOCUMEHIS YOU DGNiT NEED!

TOTAL NUMBER OF COPIES REQUIRED: LTTR 20 ENCL 17

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iuzzzzs yzziz(z AHP izzqjgzi lz~iuzisir i, ZZuEEI IIIIIIIIIII ROCHESTER GAS AND ELECTRIC CORPORATION o 89 EAST AVENUE, -ROCHESTER, N.Y. 14649.0001 April 12, 1990 TELEPHONE AREA CODE 7ld 546-2700 U.S. Nuclear Regulatory Commission Document Control Desk Attn: Allen R. Johnson Project Directorate I-3 Washington, D.C. 20555

Subject:

Inservice Testing Program Status and Relief Request R.E. Ginna Nuclear Power Plant Docket No. 50-244

Dear Mr. Johnson:

This correspondence is in response to the issues discussed. during the conference call between RGGE, the NRC and EG&G on March 23, 1990. The comments affecting the Third Interval Inservice Test Program submittal and the resolution for those comments are addressed below:

1) MECHANICAL EXERCISING VALVES 862A AND 862B (VR-24)

The Containment Spray Pump Quarterly Test. (PT-3Q) currently tests the quarterly opening requirement for the subject valves.

A spring scale is utilized to measure breakaway force for the valve in accordance with the guidance established in OMa-10.

Additionally, manual exercising to the full open position is accomplished while monitoring the applied force. Any evidence of binding or restriction to travel is noted. The test is conservative in nature and meets the requirements of ASME IWV-3522 and OMa-10, paragraph 4.3.2.4(b).

2) TECHNICAL SPECIFICATION TESTING LESS CONSERVATIVE THAN ASME RE UIR1PJHNTS FOR CATEGORY A EVENT V VALVES (GR-5)

Specifically, the test frequency, method and pressure used to quantitatively assess leakage were questioned. The test frequency is correct in the submittal since all valves that can be exercised are tested on a refueling interval basis.

Technical Specification 4.3.3.3 (40 month reference) only applies to valves that have not been exercised. since their last test period. The two valves of concern in this case are MOV 878A and 878C. These valves are administratively controlled, closed with power removed, and never required to be opened during any mode of plant operation. Therefore, these valves and the downstream check valves are not exercised (877A,B,F,H).

In the event that the MOVs were opened, they and their associated check valves would be leak tested prior to returning to power operation.

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The differential test pressure is adequate as specified in Technical Specification 4.3.3 and meets Code requirements (IWV-3423). The Franklin Institute TER specifies the application of adjustment calculations when testing is conducted at less than maximum functional differential pressure. This calculation is addressed in our current test procedures and leak data trending is performed satisfying IWV-3426.

The current test methodology for seat leakage is acceptable based on the application of direct measurement systems. The use of rotameters or downstream leak collection devices is presently practiced, even though the Technical Specification allows pressure indicators. Pressure indicators or pressure decay methods are not used to satisfy the ASME requirements.

Attached please find the Franklin Institute Technical Evaluation Report for Primary Coolant System Pressure Isolation Valves dated October 24, 1980, letter from Dennis Crutchfield (NRC) to John Maier (RG&E) dated April 20, 1981, and the sample procedure data sheet for Procedure PT-2.10.4.

3) CZARIFICATION OF IMPACTS DUE TO SYSTEM ISOLATION (GR-3)

GR-3 will be revised to delete CV1713 (Nitrogen Supply to Containment) and CV7226 (Service Air to Containment) and. a Cold Shutdown Justification will be submitted for these valves.

CV1713 is isolated during power operation via locked closed manual valve 1793 thus precluding quarterly testing. CV7226 is used only during refueling outages and is isolated during power operation on both the upstream and 'ownstream sides thus precluding quarterly testing. These valves will be tested during cold shutdown.

CV5393 (Instrument Air Containment Isolation) when isolated, requires among others, that the following valves go to their fail safe position:

Letdown AOV 427 Excess Letdown AOV 310 RCP Seal Return AOV 270B RCP Thermal Barrier AOV 754B Letdown Orifices AOV 200A, 200B, 202 Charging AOV 294, 392A, 392B Pressurizer Spray PCV 431A, 431B This would cause the loss of vital safety and operational functions. Therefore, relief for CV-5393 is as per GR-3.

still applicable

4) PRESSURE DECAY TESTING OF THE DIESEZ GENERATOR STARTING AIR SYSTEM (VR-25)

Pressure decay testing is not feasible since the diesel generator starting air system was not designed to be leak tight.

Numerous potential leak paths such as mechanical unions and instrument connections mask the ability to assess leakage through a single valve in this system. No additional meaningful information would be gathered by implementing such a test.

5) ,RELIEF RE UEST FOR TESTING 392A IN ACCORDANCE WITH IWV-3510 RE UIRrXENTS (VR-16)

RG&E is not requesting relief from all testing requirements, only the requirement to bench test this valve. The bench test requirement stems from the relieving function of the valve even though it is not a standard relief valve. Removal of this valve for bench testing would entail cutting the welded valve out of the charging system piping, disconnecting air supply tubing, de-terminating control wiring and the -subsequent reconnections after bench testing. This would pose excessive burden and is impractical.

6) SAFETY FUNCTION PERFORMED BY EDG FUEL OIL VALVES 5960A AND 5960B (VR-2)

In the-forward direction each check valve ensures that, in the event of DG fuel oil day tank level switch failure, fuel oil would be routed to the DG fuel oil storage tank rather than potentially overfilling the day tank. In the reverse direction each check valve ensures that, when the DG fuel oil transfer pump is in the recirculation mode to the fuel oil storage tank (e.g., when a day tank is at an elevated level), the recirculated fluid does not get. routed into the day tank, potentially overfilling it.

Very truly yours, Robert C. Me redy Division Manager Nuclear Production GJWN098 Attachment XC Mr. Allen R. Johnson (Mail Stop 14D1)

Project Directorate I-3 Washington, D.C. 20555 U.S. Nuclear Regulatory Commission Region I 475 Allendale Road King of Prussia, PA 19406 Ginna Senior Resident Inspector