ML17264A426
| ML17264A426 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 03/27/1996 |
| From: | Mecredy R ROCHESTER GAS & ELECTRIC CORP. |
| To: | Andrea Johnson NRC (Affiliation Not Assigned), NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NUDOCS 9604020442 | |
| Download: ML17264A426 (6) | |
Text
CATEGORY 1
SUBJECT:
Submits addi justification/basis for revised relief requests VR-8 6 VR-9,as discussed w/IST program reviewer K Dempsey.
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TITLE: OR Submittal: Inservice/Testing/Relief from ASME Code GL-89-04 NOTES:License Exp date in accordance with 10CFR2,2.109(9/19/72).
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'OC.DATE: 96/03/27 NOTARIZED:
NO DOCKET I 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 JOHN(ON,A.R.
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P ND ROCHESTER GASANDElECTRIC CORPORATION ~ 89 EASTAVENUE, ROCHESTER, N. Y ldbf9-000I AREACODE7I6 54'6-27M ROBERT C. MECREDY Vice President Nvcteor Operotions March 27, 1996 U.S. Nuclear Regulatory Commission Document Control Desk Attn:
Allen R. Johnson Project Directorate I-1 Washington, D.C.
20555
Subject:
Submittal of Additional Justification/Basis to Support Revision of Relief Requests VR-8 and VR-9 R.
E. Ginna Nuclear Power Plant Docket No. 50-244 Ref. (a):
RG&E Correspondence, Revision of Relief Requests VR-8, VR-9, and PR-9 of Inservice Testing (IST)
Program for Pumps and Valves 1990-1999 Third 10-Year
- Interval, Revision 2, dated January 12, 1996.
Dear Mr. Johnson:
Rochester Gas Electric (RG&E) hereby submits additional justification/basis for revised relief requests VR-8 and VR-9 as discussed with IST program reviewer Ken Dempsey.
During a
telephone discussion with Mr. Dempsey, conducted on March 7, 1996, supplemental information was requested to provide additional basis to facilitate the NRC review of the subject relief requests.
These relief requests were submitted to alter IST of the safety injection (SI) accumulator discharge check valves
- 842A, 842B, 867A and 867B from disassembly, full-stroke exercise and inspection conducted at a frequency of once per six years to full-flow exercise testing conducted at a once per three refueling outage frequency.
Additional justification/basis was also requested for the revision of relief request PR-9 for IST of the charging pumps.
This information will be provided in a subsequent correspondence.
The additional justification/basis for NRC consideration for approval of full-flowtesting of these check valves at a once per three refueling outage frequency is as follows:
These full-flowcheck valve tests are considered significant infrequently performed evolutions (SIPEs) in accordance with conservative plant management directives due to the potential for adverse impact to plant operation as a result of the off-normal plant configuration required for conducting the tests.
Analyses performed in preparation for these tests addressed the potential for nitrogen injection into the Reactor Coolant System (RCS) while residual heat removal (RHR) is in operation and for thermal shock of the SI accumulator nozzles.
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Although, due to controls established for. the performance of these tests, the probability for nitrogen injection into the RCS is extremely low and the thermal impact to the SI accumulator nozzles is not significant, these challenges to systems and components important to safety are not activities that RG&E believes are necessary and prudent to perform more that reasonably needed.
As stated in the basis for relief for the existing NRC approved relief requests VR-8 and VR-9, which employs check valve disassembly frequencies of once per six years, the mechanical condition of these valves has been found to be excellent when disassembled.
The successful full-flowtesting conducted during the 1995 refueling outage validated this excellent mechanical condition.
Since these valves only experience flow during testing and do not experience flow
'during any normal system-operational configuration, check valve degradation and wear is minimal.
3.
Due to the system design of Ginna Station, these check valves do not function as part of the RHR System nor are they employed during RHR System operation as they are in other nuclear power plant system designs.
Licensees who test their SI check valves at a frequency less than once every three refueling outages typically employ system designs where these check valves are utilized for RCS cooling using the RHR System which subjects these check valves to significantly larger amounts of degradation and wear.
Based on the preceding additional justification, RGGE continues to
'request relief from ASME Section XI requirements in order to full-flow inservice test these check valves at a frequency of once per three refueling outages.
In support of the impending outage
- schedule, responses to relief requests VR-8 and VR-9 are requested by April 30," 1996.
Very truly yours,
~~Q~c ~
Robert C. Mecredy xco Mr. Allen R. Johnson (Mail Stop 14B2)
Project Directorate I-1 Washington, D.C.
20555 U.S. Nuclear Regulatory Commission Region I 475 Allendale Road King of Prussia, PA 19406 US NRC Ginna Senior Resident Inspector-
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