05000354/LER-2010-002, Regarding as Found Values for Safety Relief Valve Lift Setpoints Exceed Technical Specification Allowable
| ML103630404 | |
| 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-N10-0443 LER 10-002-00 | |
| Download: ML103630404 (7) | |
| 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 |
| 3542010002R00 - NRC Website | |
text
PSEG Nuclear LLC P.O. Box 236, Hancocks Bridge, New Jersey 08038-0236 DEC 2 1 2010 LR-NI 0-0443 0 PSEG Nuclear LLC 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 2010-002 In accordance with 10 CFR 50.73(a)(2)(i)(B), PSEG Nuclear LLC is submitting Licensee Event report (LER) Number 2010-002.
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.
_nerJ Plant Manager Hope Creek Generating Station Attachment: Licensee Event Report 2010-002 95-2168 REV. 7/99
Page 2 LR-N 10-0443 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 P0 Box 415 Trenton, NJ 08625 Hope Creek Commitment Tracking Coordinator (H02)
INPO - LEREvents@lNPO.org
NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES: 10/31/2013 (10-2010)
, 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 5
- 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 MONTH DAY YEAR SEQUENTIAL REV M
FACILITY NAME DOCKET NUMBER YEAR NUMBER NO.
MONTH DAY YEAR N/A 10 25 2010 2010 - 002 -
00 12 21 2010 N/A DOCKET NUMBER
- 9. OPERATING MODE
- 11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR§: (Check all that apply)
[E 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)
[1 20.2203(a)(3)(ii)
El 50.73(a)(2)(ii)(A)
El 50.73(a)(2)(viii)(A)
[] 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)
[1 50.36(c)(1)(i)(A)
El 50.73(a)(2)(iii)
El 50.73(a)(2)(ix)(A)
- 10. POWER LEVEL E] 20.2203(a)(2)(ii)
E] 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)
[3 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)
[I 50.73(a)(2)(v)(C)
[3 OTHER El 20.2203(a)(2)(vi) 0 50.73(a)(2)(i)(B)
[E 50.73(a)(2)(v)(D)
Specify in Abstract below or in The analysis performed by GE (NEDC-3251 1 P) remains valid for the existing plant configuration and the maximum allowable percent increase (MAPI) above the SRV nominal setpoints can still be applied. For the H1RF16 testing the as-found setpoints of SRV-C, G, K, L and P remained below the value which is the lesser of 1250 psig or MAPI limit.
Thus the as-found setpoints remained within the analyzed limits (NEDC-3251 1 P). SRV-A drifted above the MAPI value of 3.0%. The Technical Evaluation concluded that if the setpoint drift of SRV A reached +5.8%, that the stresses imposed by the increased lift setpoint would have been below the ASME Section III, Appendix F, value for failure. For the Cycle 16 drift of +4.2%, this value is bounded by the +5.8% drift previously evaluated in the H1RF15 Technical Evaluation.
Therefore, the increase in the six SRV setpoints would not have impacted the vessel overpressure protection or the torus and torus attached piping.
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 1177 1130 1096-1163 4.20%
3.00%
F013C 1186 1130 1096-1163 5.00%
21.80%
F013G 1199 1120 1087-1153 7.10%
8.70%
F013K 1172 1108 1075-1141 5.80%
22.40%
F013L 1192 1120 1087-1153 6.40%
16.30%
F013P 1157 1120 1087-1153 3.30%
27.4%
- 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) did not occur as defined in Nuclear Energy Institute (NEI) 99-02, "Regulatory Assessment Performance Indicator Guideline".
CAUSE OF OCCURRENCE An oxide forms between the mating surfaces of the Pilot Disc (solid Stellite 21) and the seat in the Pilot Body (Stellite 6 overlay). This bridging oxide fractures when the pilot disc lifts. The load required to fracture this bridging oxide increases the lift point and can lead to pilots failing high during lift tests.
The apparent cause of the setpoint drift is corrosion bonding, which is consistent with industry experience. The materials combination for the pilot disc and the pilot seat have been a known industry issue since the design of the Target Rock 2 stage SRV. The oxygen content of the steam, in the pilot disc area, aggravates the natural corrosive reaction in the pilot disc seating area. Numerous industry attempts to resolve the oxide formation have failed to improve performance. A summary of the BWROG recommendations to improve SRV reliability with regard to setpoint drift was documented in NRC Regulatory Issue Summary 2000-12 dated August 7, 2000:
"Resolution of Generic Safety Issue B-55, Improved Reliability of Target Rock Safety Relief Valves". The three modification options recommended were: (1) the installation of ion beam implanted platinum (IBAD Process) pilot valve discs, (2) the installation of Stellite 21 pilot valve discs, and (3) the installation of additional pressure actuation switches. Hope Creek has implemented options 1 & 2 with limited success. Option three has not been
considered due to mixed industry results/performance.
Following H1RF15, Southwest Research was contracted to metallurgically evaluate the Pilot Body and Disc from SRV-K (setpoint failure at +9.4%) using both stereomicroscopy and scanning electron microscopy (SEM) to determine if evidence of bonding between the mating surfaces of the disc and body was present. The SEM examinations of the seating area on the Pilot Disc showed clear evidence of brittle oxide fracture along the seating line. These sharp fracture lines are typically produced as a brittle oxide grown between two surfaces fractures as the surfaces are separated, leaving islands of the oxide on each surface. Spectra taken from various regions along the seat confirmed that portions of the oxide were being removed from the Pilot Disc seat, i.e., left behind on the seat face, as the disc lifted off the seat. These results confirm that an oxide had formed between the mating surfaces of the Pilot Disc and the seat in the Pilot Body and that this bridging oxide fractured when the disc lifted. The load required to fracture this bridging oxide increases the lift point and can lead to pilots failing high during lift tests.
Based on these previous examinations and the fact that the second lift of five of the six SRVs was within the +/-3%
tolerance, corrosion bonding is the apparent cause for five of the six SRVs.
SRV-G is the only SRV, where repeated lifts did not produce a satisfactory setpoint. For the other five pilots the second lift was within the +/-3% tolerance. SRV-G was as-found tested with the first three lifts above the +3%
setpoint tolerance (first lift = +7.1%; second lift = +4.7%; third lift = +3.1%; fourth lift = +1.8%; fifth lift = +0.4%; sixth lift = +0.5%). With corrosion bonding, industry experience has shown that the first lift breaks the bond & all successive lifts are within setpoint. Five of the six setpoint failures during HiRF16 had the classical performance related to a corrosion bonding condition. SRV-G, however, had three successive lifts out of tolerance.
The most probable cause of the setpoint drift of the sixth SRV is a parts misalignment condition. SRV-G will be disassembled and inspected to determine the cause.
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 2009 Hope Creek refueling outage when six 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.
CORRECTIVE ACTIONS
- 1. During H1RF16 all 14 SRV pilot valves were removed and replaced with pre-tested, certified spare pilot valves.
- 2.
All 14 SRV pilot valves will be removed, tested and replaced with pre-tested, certified spare pilot valves during the next refueling outage (H1RF17).
- 3.
All six pilot valves that failed to meet the + 3% TS setpoint tolerances will be disassembled and inspected to validate that the cause is corrosion bonding.
- 4. SRV-G will be disassembled and inspected to determine the cause of the successive out of tolerance as-found lifts.
- 5.
A proposal to convert to 3-stage safety-relief valves is being considered through the plant modification process.
COMMITMENTS
This LER contains no commitments.
FORM 366A (10-2010)