05000272/LER-2008-002, Missed Containment Spray Valve Surveillance Per Technical Specification 4.0.5
| ML090420490 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 02/06/2009 |
| From: | Braun R Public Service Enterprise Group |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| LR-N09-0030 LER 08-002-00 | |
| Download: ML090420490 (5) | |
| 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)(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 |
| 2722008002R00 - NRC Website | |
text
PSEG Nuclear LLC PO. Box 236,, Hancocks Bridge, NJ 08038-0236 0 PSEG Nuclear L.L. C.
FEB 0 6 2009 1 OCFR50.73 LR-N09-0030 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington DC 20555-001 LER 272/08-002 Salem Nuclear Generating Station Unit 1 Facility Operating License No. DPR-70 NRC Docket No. 50-272
SUBJECT:
Missed Containment Spray Valve Surveillance per Technical Specification 4.0.5.
This Licensee Event Report, "Missed Containment Spray Valve Surveillance per Technical Specification 4.0.5." is being submitted pursuant to the requirements of the Code of Federal Regulations 10CFR50.73(a)(2)(i)(B).
The attached LER contains no commitments. Should you have any questions or comments regarding this submittal, please contact Mr. Howard Berrick at 856-339-1862.
Sincerely, Robert Braun Site Vice President Salem Generating Station Attachments (1)
Document Control Desk Page 2 FEB:,,
2009 LR-N09-0030 cc Mr. S. Collins, Administrator - Region I U. S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 Mr. R. Ennis, Licensing Project Manager - Salem U. S. Nuclear Regulatory Commission Mail Stop 08B1 Washington, DC 20555-0001 USNRC Senior Resident Inspector - Salem (X24)
Mr. K. Tosch, Manager IV Bureau of Nuclear Engineering PO Box 415 Trenton, NJ 08625
NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES: 08/31/2010 (9-2007)
, the NRC may not conduct or sponsor, and a person is not required to respond to, the
/3. PAGE Salem Generating Station - Unit 1 05000272 1 of 3
- 4. TITLE Missed Containment Spray Valve Surveillance per Technical Specification 4.0.5.
- 5. EVENT DATE
- 6. LER NUMBER
- 7. REPORT DATE
- 8. OTHER FACILITIES'INVOLVED FACILITY NAME DOCKET NUMBER MONTH DAY YEAR YEAR SEQUENTIAL REV MONTH DAY YEAR NUMBER NO.
DOCKET NUMBER 12 09 2008 2008 002 00 02 07 2009
- 9. OPERATING MODE
- 11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR§: (Check all that apply)
El 20.2201(b)
El 20.2203(a)(3)(i)
[E 50.73(a)(2)(i)(C)
El' 50:73(a)(2)(vii) 1E 20.2201(d)
El 20.2203(a)(3)(ii)
El 50.73(a)(2)(ii)(A)
El 50.73(a)(2)(viii)(A)
[I 20.2203(a)(1)
El 20.2203(a)(4)
El 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)
[I 50.73(a)(2)(iii)
El 50.73(a)(2)(ix)(A)
- 10. POWER LEVEL El 20.2203(a)(2)(ii)
El 50.36(c)(1)(ii)(A)
El 50.73(a)(2)(iv)(A)
El 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)
El 50.46(a)(3)(ii)
El 50.73(a)(2)(v)(B)
El'73.71(a)(5) 99.8%
El 20.2203(a)(2)(v)
[E 50.73(a)(2)(i)(A)
El 50.73(a)(2)(v)(C)
El OTHER El 20.2203(a)(2)(vi)
[D 50.73(a)(2)(i)(B)
El 50.73(a)(2)(v)(D)
Specify in Abstract below nr in NR(h. Fnrm RRRA
- 12. LICENSEE CONTACT FOR THIS LER FACILITY NAME TELEPHONE NUMBER (include Area Code)
Howard Berrick, Senior Licensing Engineer 856-339-1862CAUSE SYSTEM COMPONENT MANU-REPORTABLE
CAUSE
SYSTEM COMPONENT MANU-REPORTABLE FACTURER TO EPIX FACTURER TO EPIX A
BE N
- 14. SUPPLEMENTAL REPORT EXPECTED 15; EXPECTED MONTH DAY YEAR SUBMISSION [E YES (If yes, complete 15. EXPECTED SUBMISSION DATE)
[
NO DATE ABSTRACT (Limit to 1400 spaces, i.e., approximately 15 single-spaced typewritten lines).
On December 9, 2008, with Salem Unit 1 in Mode 1, it was identified that Containment Spray pressure relief (vacuum breaker) valve 1CS12 could not be located to perform a required post removal as-found surveillance test in accordance with the requirements of the Technical Specifications (TS) and the ASME OMa-1 988, Part 1, Requirements for Inservice Performance Testing of Nuclear Power Plant Pressure Relief Devices. The inability to perform the test because of the loss of the 1CS12 resulted in a conservative determination that the valve would not have passed the TS surveillance pressure test.
The valve misplacement is attributed to failure to follow work order instructions to properly retain the valve for testing. The valve testing. scope was expanded to the second redundant valve on the tank. The test of the redundant valve concluded that the valve would have performed its function. All pressure relief valves on the containment spray additive tank were replaced with new valves..
This event is reportable in accordance with 10CFR50.73(a)(2)(i)(B), Operation or Condition Prohibited by Technical Specifications.
PLANT AND SYSTEM IDENTIFICATION
Westinghouse - Pressurized Water Reactor Containment Spray/Safety Valves {BE/RV}*
- Energy Industry Identification System (EIIS) codes and component function identifier codes appear as {SS/CC}
CONDITIONS PRIOR TO OCCURRENCE The plant was in Mode 1. No structures, systems, or components were inoperable at the time of discovery that contributed to the event.
DESCRIPTION OF OCCURRENCE On September 24, 2008, it was identified that the valve grouping for the pressure relief devices (vacuum breakers) on the Containment Spray Additive Tank {BE/RV}, i.e., valves 1CS12 and 1CS13, had not been tested as required by the American Society of Mechanical Engineers (ASME) Code and they had exceeded their required Technical Specification (TS) code frequency. The surveillance for the 1CS12 valve was scheduled and the valve was swapped out with a new replacement vacuum breaker valve.
This work was completed on October 9, 2008. In accordance with plant procedures and ASME Code OMa-1 988, Part 1, components that are replaced shall be as-found tested within three (3) months after being removed.
On December 9, 2008, it was discovered that the old 1CS12 vacuum breaker valve, removed during the fall outage for as-found testing, could not be located. After discussion with the Maintenance department, it appears that this component was inadvertently disposed of prior to the as-found testing scheduled date.
Since the surveillance on the replaced. valve could not be performed,. the TS 4.0.5 surveillance is considered a test failure.
CAUSE OF OCCURRENCE The inability to perform the required TS surveillance test was caused by a failure to follow the written instructions provided in the work order to retain these valves for an as-found test. Since the. surveillance on the removed vacuum breaker valve could not be performed, the TS 4.0.5 surveillance is considered a test failure.
NRC FORM 366 (9-2007)
PRINTED ON RECYCLED PAPER NRC FORM 366 (9-2007)
PRINTED ON RECYCLED PAPER
(If more space is-required, use additional copies of NRC Form 366A)
PRIOR SIMILAR OCCURRENCES A review of Salem LERs since 2005 identified no occurrences where a valve requiring TS surveillance testing had been misplaced prior to the test. This, review also identified no occurrences where the failure to follow instructions had resulted in a reportable event.
SAFETY CONSEQUENCES AND IMPLICATIONS
The Containment Spray Additive Tank (SAT) pressure relief valves are not specifically Credited in the UFSAR Chapter 15 safety analyses for their pressure (vacuum) relief benefits. The 100% redunda"nt SAT pressure relief valves (i.e., CS12 and CS13) prevent a vacuum from forming in the SAT by maintaining atmospheric pressure on the top of the fluid in the tank during post-accident operations.
Although the 1 CS12 valve was not available for as-found testing, the redundant 1CS13 vacuum breaker valve was removed for expanded scope testing. Testing determined that the' removed 1 CS13 valve did not meet the ASME Code requirements, but the valve would have provided vacuum relief capability for the SAT. The safety function of the system would not have been affected.
A review of this event determined that a Safety System Functional Failure (SSFF) as defined in NEI 99-02, Regulatory Assessment Performance Indicator Guidelines, did not occur. This event did not prevent the ability of a system to fulfill its safety function to either shutdown the reactor, remove residual heat, control the release of radioactive material, or mitigate the consequences of an accident.
CORRECTIVE ACTIONS
- 1. All Unit 1 Containment Spray pressure relief devices were replaced with new vacuum breaker valves
- 2. Expanded scope test of the remaining valve in this group, i.e., 1CS13, which was removed and replaced with a new valve, was completed. The removed pressure relief valve would have provided vacuum relief capability for the SAT.
- 3. All Unit 2 Containment Spray pressure relief devices were checked and confirmed to be within periodicity of their testing requirements.
- 4. The maintenance plans for these valves will be revised to include as-found testing of removed vacuum breaker valves.
COMMITMENTS
The corrective actions cited in this LER are voluntary enhancements and do not constitute. commitments.