05000354/LER-2010-003, For Hope Creek Generating Station, RHR Shutdown Cooling Suction Relief Valve Missed Surveillance
| ML103630402 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 12/21/2010 |
| From: | Wagner L Public Service Enterprise Group |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| LR-N10-0436 LER 10-003-00 | |
| Download: ML103630402 (6) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2)(i) 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded 10 CFR 50.73(a)(2)(viii)(A) 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition 10 CFR 50.73(a)(2)(viii)(B) 10 CFR 50.73(a)(2)(iii) 10 CFR 50.73(a)(2)(ix)(A) 10 CFR 50.73(a)(2)(iv)(A), System Actuation 10 CFR 50.73(a)(2)(x) 10 CFR 50.73(a)(2)(v)(A), Loss of Safety Function - Shutdown the Reactor 10 CFR 50.73(a)(2)(v)(B), Loss of Safety Function - Remove Residual Heat 10 CFR 50.73(a)(2)(i)(A), Completion of TS Shutdown 10 CFR 50.73(a)(2)(v), Loss of Safety Function |
| 3542010003R00 - NRC Website | |
text
PSEG Nuclear LLC P. 0. Box 236, Hancocks Bridge, NJ 08038 o PSEG Nuclear LLC DEC 2 1 2010 LR-NI0-0436 10 CFR 50.73 United States Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-001 Hope Creek Generating Station Unit 1 Facility Operating License Number NPF-57 Docket Number 50-354 Subject: Licensee Event Report 2010-003 In accordance with 10 CFR 50.73(a)(2)(i)(B), PSEG Nuclear LLC is submitting Licensee Event Report (LER) Number 2010-003.
Should you have any questions concerning this letter, please contact Mr. Philip J. Duca at (856) 339-1640.
No regulatory commitments are contained in the LER.
Lawrence M. W Plant Manager Hope Creek Generating Station Attachment: Licensee Event Report 2010-003
Page 2 LR-N 10-0436 Document Control Desk cc:
Mr. W. Dean, Administrator-Region I U.S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 Mr. R. Ennis, Project Manager Salem and Hope Creek U.S. Nuclear Regulatory Commission One White Flint North Mail Stop 08 BlA 11555 Rockville Pike Rockville, MD 20852 USNRC Senior Resident Inspector - Hope Creek (X24)
P. Mulligan, Manager IV Bureau of Nuclear Engineering PO Box 415 Trenton, NJ 08625 Hope Creek Commitment Tracking Coordinator (H02)
INPO - LEREvents@lNPO.org
APPROVED BY OMB: NO. 3150-0104 EXPIRES: 10/31/2013 NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION N10-2010)
, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.
- 3. PAGE Hope Creek Generating Station 05000354 1 of 4
- 4. TITLE RHR Shutdown Cooling Suction Relief Valve Missed Surveillance.
- 5. EVENT DATE
- 6. LER NUMBER i
- 7. REPORT DATE
- 8. OTHER FACILITIES INVOLVED MONTH DAY YEAR YEAR SEQUENTIAL IREV MONTH DAY YEAR FACILITY NAME DOCKET NUMBER NUMBER NO.
t DOCKET NUMBER 11 01 10 2010 0 0 3100 12 21 2010
- 9. OPERATING MODE
- 11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR§: (Check all that apply)
[-1 20.2201(b)
El 20.2203(a)(3)(i)
El 50.73(a)(2)(i)(C)
El 50.73(a)(2)(vii) 5 [E
20.2201(d)
[I 20.2203(a)(3)(ii)
El 50.73(a)(2)(ii)(A)
E] 50.73(a)(2)(viii)(A)
El 20.2203(a)(1)
[] 20.2203(a)(4)
El 50.73(a)(2)(ii)(B)
El 50.73(a)(2)(viii)(B)
[] 20.2203(a)(2)(i)
El 50.36(c)(1)(i)(A)
El 50.73(a)(2)(iii)
El 50.73(a)(2)(ix)(A)
- 10. POWER LEVEL El 20.2203(a)(2)(ii)
E] 50.36(c)(1)(ii)(A)
El 50.73(a)(2)(iv)(A)
[E 50.73(a)(2)(x)
E] 20.2203(a)(2)(iii)
El 50.36(c)(2)
El 50.73(a)(2)(v)(A)
El 73.71(a)(4)
El 20.2203(a)(2)(iv)
E] 50.46(a)(3)(ii)
El 50.73(a)(2)(v)(B)
El 73.71(a)(5) 000 El 20.2203(a)(2)(v)
El 50.73(a)(2)(i)(A)
El 50.73(a)(2)(v)(C) 0l OTHER El 20.2203(a)(2)(vi)
Z 50.73(a)(2)(i)(B)
El 50.73(a)(2)(v)(D)
Specify in Abstract below or in DESCRIPTION OF OCCURRENCE (cont'd) test until the Fall 2010 refueling outage. A bounding evaluation was performed in accordance with the procedure for risk assessments of missed or deficient surveillances. The evaluation for the TS surveillance for 1 BCPSV-4425 Shutdown Cooling Suction line safety relief valve function supported a maximum deferral time equal to the normal surveillance interval. The evaluation concluded there was no significant increase in risk as a result of the missed surveillance. The surveillance test was performed during the Fall 2010 refueling outage (H1R16). The test results were unsatisfactory. The setpoint for the valve is 1250 psig.
The as-found setpoint was 1346.3 psig which is outside the + 3% acceptable range [Reference -ASME OM Code-2001, Mandatory Appendix I (Inservice Testing of Pressure Relief Devices in Light-Water Reactor Nuclear Power Plants), Section 1-1300 Guiding Principles, Subsection 1-1320 Test Frequencies, Class1 Pressure Relief Valves, Paragraph (c) Requirements for Testing Additional Valves, Subsection (1)].
Therefore, this failed late surveillance is reportable per 10 CFR 50.73(a)(2)(i)(B).
CAUSE OF OCCURRENCE A review of available plant documents, including all applicable procedures and the original program data base, and discussions with the IST Engineer identified the apparent cause of the error in grouping as a technical rigor application deficiency that occurred at the inception of the Hope Creek relief valve program development in 1985.
The relief valve was replaced with a spare and the removed valve was disassembled. No replacement parts were required. The valve was reassembled, bench calibrated, and returned to stock as a spare.
The most likely cause of the failure of the valve to actuate within the acceptable band is corrosion bonding or bridging.
PREVIOUS OCCURRENCES
A review of LERs for the three prior years at Hope Creek was performed to determine if a similar event had occurred. Two similar events were identified:(1) During the 2009 Hope Creek refueling outage when six Main Steam Safety Relief Valves (SRVs) were found out of the TS required limits of +/- 3%.
This event was reported as LER 354/2009-002-00 and its supplement 354/2009-002-01(2) During the 2010 Hope Creek refueling outage when six Main Steam SRVs were found out of the TS required limits of +/- 3%. This event was reported as LER 354/2010-002.
SAFETY CONSEQUENCES AND IMPLICATIONS
1 BCPSV-4425 is located on the common RHR shutdown cooling suction line. The valve is located between inboard and outboard isolation valves H1BC-1BCV-071 and H1-1 BCV-164 respectively. The valve is a safety relief valve with a minimum design capacity of 0.1 gpm (ref. Specification 10855-M-141(Q), Appendix C). The valve is designed to relieve system pressure in the event of a primary containment isolation signal that would cause the referenced valves to close following the system
reaching maximum design values. Per the design piping specification the maximum pressure and temperature this piping is rated for is 1375 psig and 575 OF. During the as found testing the relief valve lifted at 1346.3 psig, lower than the maximum rating of the piping. The valve moving off of its seat would relieve enough pressure to meet the requirement of 0.1 gpm, since the full open capacity is 100 GPM at 10% accumulation. Therefore a 10% accumulation for this application is very conservative and the valve would have performed its design function upon initial opening.
A review of this event determined that a Safety System Functional Failure (SSFF) did not occur as defined in Nuclear Energy Institute (NEI) 99-02, "Regulatory Assessment Performance Indicator Guideline".
CORRECTIVE ACTIONS
(1) Upon discovery of the improper grouping of 1BCPSV-4425, an immediate, detailed, item by item review of all relief devices in the IST program was performed individually by the IST Program Manager and the Relief Valve Program manager. This review was focused specifically on the ASME classification of the valves and whether or not any other components were not appropriately grouped per their classification. This review confirmed that 1 BCPSV-4425 is the only IST relief device, other than the Main Steam Safety Relief Valves, that are Class 1 components. All other IST relief devices are Class 2 and 3 and assigned to groupings with the appropriate testing frequency based on Classification and function. (Complete)
(2) The operating experience of this event was rolled to the Programs Engineering Group by the Manager of that group as a re-enforcement for the need for thorough technical rigor application in the production and review of engineering products. (Complete)
(3) The procedure for testing of Hope Creek ASME class 1, 2, 3 safety/relief valves was revised to establish a new valve group for 1 BCPSV-4425. (Complete)
(4) The Maintenance Plan for 1 BCPSV-4425 was revised to 18 months not to exceed 24 months.
(Complete)
(5) 1 BCPSV-4425 was replaced with a spare (that was satisfactorily tested), rebuilt, bench calibrated and returned to stock. (Complete)