05000354/LER-2006-003, Re as Found Values for Safety Relief Valve Lift Setpoints Exceed Technical Specification Allowable
| ML061720085 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 06/15/2006 |
| From: | Massaro M Public Service Electric & Gas Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| LR-N06-0211 LER 06-003-00 | |
| Download: ML061720085 (4) | |
| 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 |
| 3542006003R00 - NRC Website | |
text
PSEG Nuclear LLC P.O. Box 236, Hancocks Bridge, New Jersey 08038-0236 o PSEG Nuclear LLC LR-N06-0211 JUN 1.5 2006 U. S. Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 LER 354106-003-00 HOPE CREEK GENERATING STATION FACILITY OPERATING LICENSE NPF-57 DOCKET NO. 50-354 This Licensee Event Report entitled, "As Found Values For Safety Relief Valve Lift Setpoints Exceed Technical Specification Allowable" is being submitted pursuant to the requirement of 1 OCFR50.73(a)(2)(i)(B).
Sincerely, Michael J. Massaro Plant Manager - Hope Creek Attachment FDP C
Distribution 95-2168 REV. 7/99
NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSIOt APPROVED BY OMB: NO. 3150-0104 EXPIRES: 06/30/2007 (6>-204)
, the NRC may not conduct or sponsor, and a person Is not required to respond to, the digits/characters for each block)
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- 13. PAGE Hope Creek Generating Station 05000 354 1 OF 3
- 4. TITLE As Found Values for Safety Relief Valve Lift Setpoints Exceed Technical Specification Allowable
- 5. EVENT DATE
- 6. LER NUMBER
- 7. REPORT DATE
- 18.
OTHER FACILITIES INVOLVED SEUNIL I
I I
FACILITY NAMEDOKTNME MONTH DAY YEAR YEAR SEQUENTIAL REV MONTH DAY YEAR IN/A 5FACILITY NAME DOCKET NUMBER 04 21 2006 2006
- - 003 -
00 06 15 2006 N/A
- 9. OPERATING MODE
- 11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR§: (Check all that apply) o 20.2201(b) 0 20.2203(a)(3)(i) 0l 50.73(a)(2)(i)(C) 0 50.73(a)(2)(vii) 5 0 20.2201(d)
[0 20.2203(a)(3)(ii) 0l 50.73(a)(2)(ii)(A) 0l 50.73(a)(2)(viii)(A)
[o 20.2203(a)(1)
El 20.2203(a)(4)
[I 50.73(a)(2)(ii)(B) 0- 50.73(a)(2)(viii)(B) 0o 20.2203(a)(2)(i) 0 50.36(c)(1)(i)(A)
[I 50.73(a)(2)(iii) 0l 50.73(a)(2)(ix)(A)
- 10. POWER LEVEL 0 20.2203(a)(2)(ii) 0l 50.36(c)(1)(ii)(A) 0l 50.73(a)(2)(iv)(A)
[E 50.73(a)(2)(x)
[o 20.2203(a)(2)(iii)
[I 50.36(c)(2) 0l 50.73(a)(2)(v)(A) 0l 73.71(a)(4) 0l 20.2203(a)(2)(iv)
El 50.46(a)(3)(ii)
[I 50.73(a)(2)(v)(B)
[I 73.71(a)(5) 0 0
20.2203(a)(2)(v)
[I 50.73(a)(2)(i)(A)
[I 50.73(a)(2)(v)(C) 0 OTHER o3 20.2203(a)(2)(vi)
ED 50.73(a)(2)(i)(B)
El 50.73(a)(2)(v)(D)
Specify In Abstract below nr In NRC. Fnrm 'RRA
- 12. LICENSEE CONTACT FOR THIS LER FACILITY NAME ITELEPHONE NUMBER (Include Area Code)
F. Possessky, Compliance Engineer J856-339-1160CAUSE SYSTEM COMPONENT MANU-REPORTABLE
CAUSE
SYSTEM COMPONENT MANU-REPORTABLE FACTURER TO EPIX FACTURER TOEPIX B
SBI RV T020 Y
I
- 14. SUPPLEMENTAL REPORT EXPECTED
- 15. EXPECTED MONTH DAY YEAR SUBMISSION El YES (If yes, complete 15. EXPECTED SUBMISSION DATE) 0 NO DATE ABSTRACT (Limit to 1400 spaces, i.e., approximately 15 single-spaced typewritten lines)
On April 21, 2006, PSEG determined that three safety relief valve (SRV) setpoints exceeded Technical Specification (TS) allowable tolerance specified in TS 3.4.2.1. This specification requires SRV setpoint limits to be within +/- 3% of the specified value. The valves failing to meet limits were Target Rock Model 7567F SRVs. The testing followed Hope Creek fuel cycle thirteen. In all, a total of three of fourteen SRVs experienced setpoint drift outside of the Technical Specification 3.4.2.1 limit.
The apparent cause for the three SRV setpoint failures is corrosion bonding/sticking of the pilot disc.
Immediate corrective action was to replace all three valves with tested and certified spare pilot assemblies. These three valves will be disassembled and inspected to document the cause of the failure.
Since the number of SRVs outside of the setpoint tolerance limit (three) was greater than the number of SRVs (one) allowed to be inoperable by Technical Specification 3.4.2.1, this condition was determined to be reportable under 10CFR50.73(a)(2)(i)(B), as any operation or condition prohibited by the plant Technical Specifications.
NRC FORM 366 (6-2004)U.S. NUCLEAR REGULATORY COMMISSION (1-2001)
LICENSEE EVENT REPORT (LER)
FACILITY NAME (1)
DOCKET (2)
LER NUMBER (6)
PAGE (3)
YER SEQUENTIAL I REVISION NUMBER NUMBER Hope Creek Generating Station 05000354 2 OF 3 p 2006 003 00 TEXT (If more space Is required, use additional copies of NRC Form 366A) (17)
PLANT AND SYSTEM IDENTIFICATION
General Electric - Boiling Water Reactor (BWR/4)
Main Steam - EIIS Identifier {SB}*
Safety Relief Valves - ElIS Identifier {-/RV}*
- Energy Industry Identification System {EIIS} codes and component function identifier codes appear as {SS/CCC}
IDENTIFICATION OF OCCURRENCE Event Date: April 21, 2006 Discovery Date: April 21, 2006 CONDITIONS PRIOR TO OCCURRENCE Hope Creek was in cold shutdown for the thirteenth refueling outage (RF1 3). No structures, systems, or components were inoperable at the time of discovery that contributed to the event.
DESCRIPTION OF OCCURRENCE On April 21, 2006, Engineering personnel received the results of the Main Steam Safety Relief Valves (SRV){SB/RV}
(Target Rock Model 7567F) setpoint testing required by Technical Specification 4.4.2.2. That report documented the failure of SRV A, C, and K to meet TS 3.4.2.1 limit of +/- 3%.
Action a of TS 3.4.2.1 specifies 'With the safety valve function of two or more of the above listed fourteen safety/relief valves inoperable, be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.'
Valve ID As Found TS Setpoint Acceptable Band
% Difference (psig)
(psig)
(psig)
F013A 1166 1130 1096-1163 3.2%
F013C 1166 1130 1096-1163 3.2%
F013K 1144 1108 1075-1141 3.2%
CAUSE OF OCCURRENCE The apparent cause for the 'A', "C', and "K" SRV setpoint failures is corrosion bonding/sticking of the pilot disc. PSEG Nuclear has continued to experience as-found setpoint failures on SRVs even with the industry recommended coating Ion Beam Assisted Deposition (IBAD) installed on the pilot disc. The initial lift being out of specification, with subsequent lifts within specification and/or the initial failure of the stick test, is an indication of corrosion bonding of the pilot disc.U.S. NUCLEAR REGULATORY COMMISSION (1-2001)
LICENSEE EVENT REPORT (LER)
FACILITY NAME (1)
DOCKET (2)
LER NUMBER (6)
PAGE (3)
IISEQUENTIAL REVISION UjYEAR NUMBER NUMBER Hope Creek Generating Station 05000354 2006 030
_BER 3_F3 112006 003 001
PREVIOUS OCCURRENCES
A review of events for the three prior years at Hope Creek was performed to determine if similar event had occurred.
There was a similar event when 8 SRVs were found during the RF1 I Hope Creek refueling outage and 5 SRVs were found during the RF12 Hope Creek refueling outage out of TS required limits of +/- 3%. These events were reported as LER 354/2004-003-00 and LER 354/2004-009-00. Corrective actions were not successful to prevent recurrence.
SAFETY CONSEQUENCES AND IMPLICATIONS
A bounding analysis was performed and documented in VTD# 322869, NEDC-3251 1 P, "SAFETY REVIEW FOR HOPE CREEK GENERATING STATION SAFETY/RELIEF VALVE TOLERANCE ANALYSIS." This analysis supported the increase in allowable Technical Specification (1S) setpoint drift from + 1 percent to + 3 percent. An individual SRV upper limit setpoint of 1250 psig and 13 SRV's available out of a total of 14 was assumed in the calculation. The calculated peak vessel pressure at the bottom of the reactor vessel was 1331 psig. This provides a margin of 44 psi to the ASME upset limit of 1375 psig. Based on the above, there was no impact to the health and safety of the public because none of the SRVs exceeded the 1250 psig analyzed limit.
In addition, loads on SRV discharge piping were reviewed. The discharge piping analysis establishes an allowable percentage increase for each SRV line such that the allowable stresses will not be exceeded. This condition was previously evaluated and found acceptable for 5.5% above nominal setpoint as identified in LER 354/2004-009-00.
Additional plant systems were evaluated for potential issues regarding the higher pressure from the higher SRV setpoints.
Evaluations concluded that these systems remained operable.
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.
CORRECTIVE ACTION
The pilot assembly for each of the failed SRVs was replaced with a fully tested spare assembly.
All 14 pilot discs were replaced with those of Stellite 21 alternative material, as recommended by the BWROG.
COMMITMENTS
This LER contains no commitments.