05000354/LER-2009-002, As Found Values for Safety Relief Valve Lift Setpoints Exceed Technical Specification Allowable
| ML091670248 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 06/03/2009 |
| From: | Jamila Perry Public Service Enterprise Group |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| LR-N09-0123 LER 09-002-00 | |
| Download: ML091670248 (6) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 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 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications |
| 3542009002R00 - NRC Website | |
text
PSEG Nuclear LLC P.O. Box 236, Hancocks Bridge, New Jersey 08038-0236 0 PSEG Nuclear LLC JUN 0 3 2009 LR-N09-0123 1 OCFR50.73 United States Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-001 Hope Creek Generating Station Unit 1 Facility Operating License No NPF-57 Docket No. 50-354
Subject:
Licensee Event Report 2009-002 In accordance with 50.73(a)(2(i)(B), PSEG Nuclear LLC is submitting Licensee Event Report (LER) Number 2009-002.
Should you have any questions concerning this letter, please contact Mr. Timothy R. Devik at (856) 339-3108.
No regulatory commitments are contained in the LER.
Sincerely,
(ýIzr John F. Perry Plant Manager Hope Creek Generating Station Attachment: Licensee Event Report 2009-002 m3D 95-2168 REV. 7/99,
Page 2 LR-N09-0123 Document Control Desk cc:
Mr. S. Collins, Administrator - Region 1 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 B1 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
NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES: 8/31/2010 (9-2007)
, the NRC may not conduct or sponsor, and a person is not required to respond to, the digits/characters for each block) information collection.
j3. PAGE Hope Creek Generating Station 05000354j 1
0FF4
- 4. TITLE As Found Values for Safety Relief Valve Lift Setpoints Exceed Technical Specification Allowable
- 5. EVENT DATE
- 6. LER NUMBER__
- 7. REPORT DATE
- 8. OTHER FACILITIES INVOLVED SEQUENTIALEAEVSEONTNIDLY YEAR FACILITY NAME DOCKET NUMBER MONTH DAY YEAR YEAR NUMBER REV MONTH DAY YEAR N/A FACILITY NAME DOCKET NUMBER 04 18 2009 2009 - 002 - 000 06 03 2009 N/A
- 9. OPERATING MODE
- 11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR§: (Check all that apply)
El 20.2201(b)
[I 20.2203(a)(3)(i)
[I 50.73(a)(2)(i)(C)
El 50.73(a)(2)(vii) 5 0 20.2201(d)
[I 20.2203(a)(3)(ii)
[I 50.73(a)(2)(ii)(A) 17 50.73(a)(2)(viii)(A)
El 20.2203(a)(1)
El 20.2203(a)(4) 0l 50.73(a)(2)(ii)(B)
El 50.73(a)(2)(viii)(B)
El 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 iJ 20.2203(a)(2)(ii)
El 50.36(c)(1)(ii)(A)
El 50.73(a)(2)(iv)(A)
E] 50.73(a)(2)(x)
El 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)
[I 50.46(a)(3)(ii)
[I 50.73(a)(2)(v)(B)
El 73.71(a)(5) 000 [1 20.2203(a)(2)(v)
El 50.73(a)(2)(i)(A)
El 50.73(a)(2)(v)(C)
El OTHER El 20.2203(a)(2)(vi)
[
50.73(a)(2)(i)(B)
[E 50.73(a)(2)(v)(D)
Specify in Abstract below or in The discharge piping analysis contained in NEDC-3251 1 P was re-assessed to ensure that the previous analysis (based on PSA snubbers) was still valid. A review and analysis of the differences between the snubbers determined that the PSA analysis was still bounding the present plant configuration. NEDC-32511 P identifies a maximum increase in the nominal setpoint of "A" SRV to be 3%, without exceeding allowable stresses. The "A" SRV lifted at 5.8% above nominal setpoint. RIS 2005-20 and NRC Inspection Manual Part 9900 Technical Guidance on "Operability Determinations and functionality Assessments for Resolution of Degraded or Nonconforming Conditions Adverse to Quality or Safety" allow the use of the criteria in Appendix F of Section III of the ASME Boiler and Pressure Vessel code for operability determinations. The re-evaluation has determined that the ASME III allowable stresses on the piping system are higher than the stresses that would have been seen had the "A" SRV lifted. There is no present operability concern due to the replacement of this pilot assembly with a fully tested spare.
Based on the above, and because none of the SRVs exceeded the 1250 psig analyzed limit, there was no impact to the health and safety of the public.
The final test results for the SRVs that had setpoint drift outside the tolerance were as follows:
Valve ID As Found TS Setpoint Acceptable Band
% Difference (psig)
(psig)
(psig)
Actual Limit*
F013A 1195 1130 1096-1163 5.80%
3.00%
F013C 1203 1130 1096-1163 6.50%
21.80%
F013F 1163 1108 1075-1141 5.00%
5.50%
F013G 1156 1120 1087-1153 3.20%
8.70%
F013K 1212 1108 1075-1141 9.40%
22.40%
F013L 1170 1120 1087-1153 4.50%
16.30%
The limit is based on the SRV discharge piping mechanical stress limit identified in Table 7-1 of GE analysis (NEDC-3251 1 P) and is known as the "Maximum Allowable Pressure Increase" (MAPI).
A review of this event determined that a Safety System Functional Failure (SSFF) has not occurred as defined in Nuclear Energy Institute (NEI) 99-02.
CAUSE OF OCCURRENCE The specific cause(s) of the failures has not yet been determined. The six SRV pilot assemblies will be disassembled and inspected to determine and document the cause of the failures. The results of the inspection will be evaluated and provided in a supplement to this LER.
PREVIOUS OCCURRENCES
A review of LERs for the three prior years at Hope Creek was performed to determine if a similar event had occurred. There was a similar event during the 2006 Hope Creek refueling outage when three SRVs were found out of the TS required limits of +/- 3%. This event was reported as LER 354/06-003-00. Actions taken at that time were effective in that the following refueling outage did not have any SRV setpoint drift outside the allowable band on the SRVs tested.
CORRECTIVE ACTIONS
- 1. The pilot assembly of each failed SRV was replaced with a pre-tested, certified spare.
- 2.
All 6 SRV pilot valve assemblies that failed will be disassembled and inspected to determine the cause of the setpoint drift. (Corrective Action Program (CAP) number: 70096933 / 0170)
- 3.
All 14 SRV pilot valves will be removed, tested and replaced with pre-tested, certified spare pilot valves during the next refueling outage (RF16). (CAP number: 70096933 / 0180)
- 4.
An equipment apparent cause evaluation is being performed to determine the cause(s) of the failures and determine any additional corrective actions. (CAP number: 70096933 / 0070)
COMMITMENTS
This LER contains no commitments.PRINTED ON RECYCLED PAPER..PRINTED ON RECYCLED PAPER