JAFP-11-0124, Response to Request for Additional Information for Valve Relief Request VRR-08
| ML112930539 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  |
| Issue date: | 10/20/2011 |
| From: | Joseph Pechacek Entergy Nuclear Northeast, Entergy Nuclear Operations |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| JAFP-11-0124, TAC ME6687 | |
| Download: ML112930539 (5) | |
Text
JAFP-11-0124 October 20, 2011 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555
Subject:
Response to Request for Additional Information for Valve Relief Request VRR-08 (TAC NO. ME6687)
James A. FitzPatrick Nuclear Power Plant Docket Nos. 50-333 License Nos. DPR-59
References:
- 1. Entergy letter, Proposed Relief Request Nos. VRR-07 and VRR-08 for the James A. FitzPatrick Nuclear Power Plant Fourth Interval In-Service Testing Program, JAFP-11-0085, dated July 7, 2011
- 2. NRC email, FitzPatrick - ME6686 & ME6687-Draft RAIs from CPTB Re: Relief Requests VRR-07 & VRR-08, dated September 8, 2011
- 3. Teleconference with NRC, dated September 20, 2011
Dear Sir or Madam:
James A. FitzPatrick Nuclear Power Plant (JAF) submitted two Valve Relief Request (VRR)
[Reference 1] for relief from the testing requirements of ASME OM Code-2001 including 2003 addenda ISTC-3630(a) Leakage Test Frequency (VRR-07) and ISTC-3700 Position Verification Testing (VRR-08).
The NRC provided a draft Request for Additional Information (RAI) [Reference 2] for VRR-08.
By teleconference the RAI questions were clarified on September 20, 2011 [Reference 3].
JAFs responses to the RAI questions as clarified are contained in the attachment to this letter.
There are no commitments contained in this letter.
Entergy Nuclear Northeast Entergy Nuclear Operations, Inc.
James A. FitzPatrick NPP P.O. Box 110 Lycoming, NY 13093 Tel 315-342-3840 Joseph Pechacek Licensing Manager - JAF
JAFP-11-0124 Page 2 of 2 Should you have any questions concerning this letter, or require additional information, please contact Joseph Pechacek, Licensing Manager, at 315-349-6766.
Josep Pechacek Licensing Manager JP/mh
Attachment:
cc:
Responses to Request for Additional Information Regional Administrator, Region I U. S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406-1415 Resident Inspector's Office U.S. Nuclear Regulatory Commission James A. FitzPatrick Nuclear Power Plant P.O. Box 136 Lycoming, NY 13093 Mr. Bhalchandra Vaidya, Project Manager Plant Licensing Branch 1-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Mail Stop 0-8-C2A Washington, DC 20555-0001 Ms. Bridget Frymire New York State Department of Public Service 3 Empire State Plaza, 10 th Floor Albany, NY 12223 Mr. Francis J. Murray Jr., President NYSERDA 17 Columbia Circle Albany, NY 12203-6399 Document Component(s):
001 Transmittal Letter JAFP-11-0124 wI Attachment
Attachment JAFP-11-0124 Responses to Request for Additional Information (2 Pages)
Attachment JAFP-11-0124 Responses to Request for Additional Information Page 1 of 2 The following are the responses from the James A. FitzPatrick Nuclear Power Plant to each NRC request for additional information (RAI) for Valve Relief Request VRR-08 (Relief to implement Performance Based Scheduling of Position Indication Verification Tests):
RAI VRR-08-1 It was stated that the 33 category A solenoid valves are exercised quarterly and their stroke times were measured and compared to ASME OM Code acceptance criteria. Please explain how the stroke time data is tracked and trended for degradation. Also, please describe the exercise method and how valve obturator movement is verified during the exercise test.
Response
During the quarterly exercise testing, each valves stroke time is measured and recorded into the test of record. The test of record contains the applicable ISTC-5152 Code Acceptance Criteria, which allows Operations personnel to immediately identify any discrepancies. If an issue is found, it is documented in the Corrective Action Program and the IST coordinator is notified. If there are no discrepancies with the exercise test values, the values from the test of record are recorded into the IST database (PV-Plus). Within the IST database, the newest test values are compared to the previous six test values by the IST Coordinator. If there is a significant change or an upward trend between the new test and the previous tests, it alerts the IST Coordinator to perform a more detailed analysis of the stroke time data. Also, the IST Database automatically compares the test value with both the applicable ISTC-5152 Code Acceptance Criteria as well as the ISTC-5151(b) Owner Identified limiting criteria. If either of the criteria levels are exceeded, the IST Database displays a warning. This allows for tracking and trending of degrading performance over time.
Thirty of the thirty-three subject valves stroke in less than 2 seconds, and are designated as fast acting. Per ISTC-5152(c), these valves have a maximum limiting stroke time of 2 seconds. Due to how quickly these valves operate, and the inaccuracies that would exist from attempting to manually time the stroke, the stroke time is recorded as less than or equal to 2 seconds or greater than 2 seconds.
During the quarterly exercise testing, the valves are remote manually operated from the control room. Valve obturator position is verified using position indication lights in accordance with ISTC-3530.
Attachment JAFP-11-0124 Responses to Request for Additional Information Page 2 of 2 RAI VRR-08-2 For the valves addressed by this relief request, have evaluations been performed to assess the change in aggregate risk associated with relaxing the interval for position indication verification testing to as long as 60 months? If so, what were the results?
Response
For the valves addressed in VRR-08, a review of the subject valves against the James A.
FitzPatrick Probabilistic Safety Assessment (PSA) model was performed. The results show that only valves 27SOV-141 (Drywell PCV and instrument Air or Normal N2 Cross-Tie Valve) and 27SOV-145 (Drywell Instrument Nitrogen Backup Supply Isolation Valve) are in the model and neither valves failure will impact core damage frequency (CDF) or large early release frequency (LERF) and, therefore, are not risk significant.