05000354/LER-2004-009, Re as Found Values for Safety Relief Valve Lift Setpoints Exceed Technical Specification Allowable
| ML050270169 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 01/18/2005 |
| From: | Massaro M Public Service Enterprise Group |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| LR-N05-0004 LER 04-009-00 | |
| Download: ML050270169 (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) 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(0) 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 |
| 3542004009R00 - NRC Website | |
text
PSEG Nuclear LLC P.O. Box 236, Hancocks Bridge, New Jersey 08038-0236 0 PSEG Nuclear LLC
'JAN 1 8 2005 LR-N05-0004 U.S. Regulatory Commission Document Control Desk Washington, DC 20555 LER 354104-009-00 HOPE CREEK GENERATING STATION FACILITY OPERATING LICENSE NO. 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 requirements of 1 OCFR50.73(a)(2)(i)(B).
Sincerely, Michael Mansaro Plant Manager - Hope Creek Attachment RFY C
Distribution LER File 3.7 95-2168 REV. 7/99
NRCYORM 366 U.S. NUCLEAR REGULATORY COMMISSIO APPROVED BY OMB: NO. 3150-0104 EXPIRES: 06130/2007 (6-2004)
, the NRC may dir fr ec bnot conduct or sponsor, and a person is not required to respond to, the digits/characters for each block)
- s N, l.*
- 3. PAGE Hope Creek Generating Station 05000 354 1 OF 3
.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 SEQUENTIAL REV FACILITY NAME DOCKET NUMBER MONTH DAY YEAR YEAR NUMBER NO.
MONTH DAY YEAR FACILITY NAME DOCKET NUMBER 11 19 2004 2004 - 009 -
00 1
17 2005
- 9. OPERATING MODE1.THISREPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR§: (Check all that apply) o 20.2201(b) 0 20.2203(a)(3)(i) 0 50.73(a)(2)(i(C) 0 50.73(a)(2)(vii) 5 0 20.2201(d) 0 20.2203(a)(3)(ii) 0 50.73(a)(2)(0)(A) 0 50.73(a)(2)(viii)(A) o 20.2203(a)(1) 0 20.2203(a)(4) 0 50.73(a)(2)(ii)(B) 0 50.73(a)(2)(viii)(B) o 20.2203(a)(2)(i) 0 50.36(c)(1)(i)(A) 0 50.73(a)(2)(iii) 0 50.73(a)(2)(ix)(A)
- 10. POWER LEVEL 0
20.2203(a)(2)(ii) 0 50.36(c)(1)(ii)(A) 0 50.73(a)(2)(iv)(A) 0 50.73(a)(2)(x) 0 20.2203(a)(2)(iii) 0 50.36(c)(2) 0 50.73(a)(2)(v)(A) 0 73.71(a)(4) 0 20.2203(a)(2)(iv) 0 50.46(a)(3)(ii) 0 50.73(a)(2)(v)(B) 0 73.71 (a)(5) 0 0
20.2203(a)(2)(v) 0 50.73(a)(2)(i)(A) 0 50.73(a)(2)(v)(C) 0 OTHER 0
20.2203(a)(2)(vi) 0 50.73(a)(2)(i)(B) 0 50.73(a)(2)(v)(D)
Specify In Abstract below nr in NRC. Fnrm WRRA
- 12. LICENSEE CONTACT FOR THIS LER FACILITY NAME TELEPHONE NUMBER (Infude Area Cofde)
R. Yewdall, Licensing Engineer 1856-339-2469CAUSE SYSTEM COMPONENT MANU-REPORTABLE
CAUSE
SYSTEM COMPONENT MANU-REPORTABLE FACTURER TO EPIX FACTURER TO EPIX B
SI3 RV T020 Y
- 14. SUPPLEMENTAL REPORT EXPECTED
. EXPECTED MONTH DAY YEAR SUBMISSION 0 YES (ff yes, complete 15. EXPECTED SUBMISSION DATE) 0 NO DATE ABSTRACT (Limit to 1400 spaces, i.e., approximately 15 single-spaced typewritten lines)
On November 19, 2004, PSEG determined that the setpoint value for several safety relief valves (SRV) exceeded Technical Specification (TS) allowable tolerance specified in TS 3.4.2.1. This specification requires SRV setpoint limits to be within +1-3% of the specified value. The valves failing to meet limits were Target Rock Model 7567F SRVs. The testing followed Hope Creek fuel cycle 12. In all, a total of five of fourteen SRVs experienced setpoint drift outside of the Technical Specification 3.4.2.1 limit.
The apparent cause for three of the SRV setpoint failures is corrosion bonding/sticking of the pilot disc. The apparent cause for the remaining two valves is under investigation. Immediate corrective action was to replace all five valves with tested and certified spare pilot assemblies. These five valves will be disassembled and inspected to document the cause of the failure. Since the number of SRVs outside of the setpoint tolerance limit (five) 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 1 OCFR50.73(a)(2)(i)(B), as any operation or condition prohibited by the plant Technical Specifications.
NRC FORM 366 (-2004)U.S. NUCLEAR REGULATORY COMMISSION (1-2001)
LICENSEE EVENT REPORT (LER)
FACILITY NAME (1)
DOCKET (2)
LER NUMBER(6)
PAGE63 I
I SEQUENTIAL REVISION 0YER NUMBER NUMBER Hope Creek Generating Station 05000354 2 OF 3 2004 009 00 TEXT (If more space is required, use additonal copies of NRC Form 366A) (17)
PLANT AND SYSTEM IDENTiFICATION
General Electric - Boiling Water Reactor (BWRI4)
Main Steam - EIIS Identifier {SB)*
Safety Relief Valves - EIIS Identifier {-/RV}*
- Energy Industry Identification System {EIIS} codes and component function identifier codes appear as {SS/CCC}
IDENTIFICATION OF OCCURRENCE Event Date: November 19, 2004 Discovery Date: November 19, 2004 CONDITIONS PRIOR TO OCCURRENCE Hope Creek was in cold shutdown for the twelfth refueling outage (RF12). No structures, systems, or components were inoperable at the time of discovery that contributed to the event.
DESCRIPTION OF OCCURRENCE On November 9, 2004, Engineering personnel received the initial 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 B to meet TS 3.4.2.1 limit of +/- 3%. On November 19, 2004 additional test results were received. This testing revealed that following Hope Creek Cycle 12 run, two additional SRVs failed to meet the TS limit.
Upon completion of testing, a total of five SRVs experienced setpoint drift outside of the Technical Specification 3.4.2.1 limit of +/- 3%, (values listed below). 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 Setpolnt Acceptable Band
% Difference (psig)
(psig)
(psig)
F013A 1192 1130 1096 -1163 5.5%
F013B 1171 1130 1096 -1163 3.6%
F013C 1207 1130 1096-1163 6.8%
F013D 1184 1130 1096 -1163 4.8%
F013F 1156 1108 1075-1141 4.3%
CAUSE OF OCCURRENCE The apparent cause for the "B", 'D", and "F" 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 (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 the pilot disc sticking. The other two SRV ("A" and "CT) setpoint failures are still under investigation. The preliminary assessment indicates that the apparent cause may be misalignment of parts causing the high setpoint.U.S. NUCLEAR REGULATORY COMMISSION (1-2001)
LICENSEE EVENT REPORT (LER)
FACILITY NAME (1)
DOCKET (2)
LER NUMBER (6)
PAGE (3)
SEQUENTIAL REVISION YEAR NUMBER NUMBER Hope Creek Generating Station 05000354 3 OF 3 2004 009 00 l
PREVIOUS OCCURRENCES
A review of LERs for the two prior years at Hope Creek and Salem was performed to determine if a similar event had occurred. There was a similar event during the last Hope Creek refueling outage when eight SRVs were found out of the TS required limits of +/- 3%. This event was reported as LER 354/04-003-00. Corrective actions were not successful to prevent recurrence.
SAFETY CONSEQUENCES AND IMPLICATIONS
A bounding analysis was performed and documented in NEDC-3251 I P, "SAFETY REVIEW FOR HOPE CREEK GENERATING STATION SAFETY/RELIEF VALVE TOLERANCE ANALYSIS." This analysis supported the increase in allowable Technical Specification (TS) setpoint drift from + 1 percent to + 3 percent. An individual SRV upper limit setpoint of 1250 psig with 13 SRVs 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 psig to the ASME upset limit of 1375 psig. In addition, loads on SRV discharge piping were reanalyzed. The analysis established an allowable percentage increase for each SRV line such that the allowable stresses would not be exceeded. Four of the five valves met their individual acceptance limits. The "A" SRV valve exceeded its limit of +3% with an as-found setpoint of 5.5%.
Utilizing the calculation basis of 13 of 14 SRVs being available, there were no safety consequences or implications involved as a result of these valves exceeding the allowable tolerance and the analysis is bounding. Therefore, the public health and safety was not affected.
As a follow-up, however, the discharge piping analysis contained in NEDC-3251 I P will be re-assessed to determine the impact of the high setpoint of the "A" SRV. NEDC-32511P identifies a maximum increase in the nominal setpoint of"A" SRV to be 3%, without exceeding allowable stresses. The "A" SRV lifted at 5.5% above nominal setpoint. This requires a re-evaluation of the piping due to the potential for overstressing the line if the SRV A had 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 SRV exceeded the 1250 psig analyzed limit, there was no impact to the health and safety of the public.
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.
The five SRVs will be disassembled and inspected to determine the cause of the high lift points.
Reactor water chemistry will be evaluated for its potential to support the corrosion bonding of the pilot disc seating surfaces.
COMMITMENTS
This LER contains no commitments.