05000354/LER-2013-007, Regarding As-Found Values for Safety Relief Valve Lift Set Points Exceed Technical Specification Allowable Limit
| ML14035A069 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 01/16/2014 |
| From: | Carr E Public Service Enterprise Group |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| LR-N14-005 LER 13-007-00 | |
| Download: ML14035A069 (6) | |
| 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 |
| 3542013007R00 - NRC Website | |
text
PSEG Nuclear LLC P.O. Box 236, Hancocks Bridge, New Jersey 08038-0236 0 PSEG
\\ T tclear LLC JAN 1 6 2014 1 0CFR50.73 LR-N14-005 United States Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-001 Hope Creek Generating Station Unit 1 Renewed Facility Operating License No. NPF-57 Docket No. 50-354
Subject:
Licensee Event Report 2013-007-00 In accordance with the requirements of 10 CFR 50.73(a)(2)(i)(B), PSEG Nuclear LLC is submitting the enclosed Licensee Event Report (LER) Number 2013-007-00, "As-found Values for Safety Relief Valve Lift Setpoints Exceed Technical Specification Allowable Limit."
i, :
If you have any questions or require additional information, please contact Mr. Philip Duca at (856) 339-1640.
There are no regulatory commitments contained in this letter.
Sincerely, Eric S. Carr Plant Manager Hope Creek Generating Station Attachment: Licensee Event Report 2013-007-00
LR-N14-005 1 OCFR50.73 Page 2 of 2 cc:
W. Dean, Regional Administrator-Region I, NRC
- - eJ. Hughey, Project Manager - US NRC NRC Senior Resident Inspector - Hope Creek (X24)
P. Mulligan, Manager, NJBNE LER uploaded to ICES P. Bonnett - Hope Creek Commitment Tracking Coordinator (H02)
L. Marabella - Corporate Commitment Tracking Coordinator (N21)
NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES: 10/31/2013 110-2010)
, the NRC maý 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 4
- 4. TITLE As-Found Values for Safety Relief Valve Lift Set Points Exceed Technical Specification Allowable Limit
- 5. EVENT DATE
- 6. LER NUMBER__
- 7. REPORT DATE
- 8. OTHER FACILITIES INVOLVED MONTH DAY YEAR YEAR SEQUENTIAL REV MONTH FACILITY NAME DOCKET NUMBER NUMBER NO.
N/A 11 22 2013 2013 - 007 -
00 01 16 2014 N/A DOCKET NUMBER
- 9. OPERATING MODE
- 11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR§: (Check all that apply)
[l 20.2201(b)
El 20.2203(a)(3)(i)
El 50.73(a)(2)(i)(C)
El 50.73(a)(2)(vii)
El 20.2201(d)
[I 20.2203(a)(3)(ii)
El 50.73(a)(2)(ii)(A)
El 50.73(a)(2)(viii)(A)
El 20.2203(a)(1)
El 20.2203(a)(4)
[I 50.73(a)(2)(ii)(B)
El 50.73(a)(2)(viii)(B)
El 20.2203(a)(2)(i)
[I 50.36(c)(1)(i)(A)
El 50.73(a)(2)(iii)
[I 50.73(a)(2)(ix)(A)
- 10. POWER LEVEL [E
20.2203(a)(2)(ii)
[I 50.36(c)(1)(ii)(A)
El 50.73(a)(2)(iv)(A)
[I 50.73(a)(2)(x)
El 20.2203(a)(2)(iii)
E] 50.36(c)(2)
El 50.73(a)(2)(v)(A)
F-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) 100 [1 20.2203(a)(2)(v)
[I 50.73(a)(2)(i)(A)
El 50.73(a)(2)(v)(C)
El OTHER El 20.2203(a)(2)(vi)
Z 50.73(a)(2)(i)(B)
El 50.73(a)(2)(v)(D)
Specify in Abstract below or in The extent of condition for this event is to expand the scope of the SRV Group 1 valve testing, per ASME OM Code Section 1-1320 for Class 1 Pressure Relief Valves. However, since all 14 SRV pilot stage assemblies were removed and replaced with tested spares during the refueling outage (H1 R18), the extent of condition scope was satisfied.
CAUSE OF EVENT
The cause of the setpoint drift is attributed to corrosion bonding, which is consistent with prior determinations at HCGS and within the BWR industry. Corrosion bonding occurs when 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 pilot stage assemblies failing high during initial lift tests. Subsequent lifts following the initial 'as-found' lift are typically within setpoint tolerances. The five SRVs which exceeded the lift settings during the 'as-found' testing were within +/-3% tolerance for the second lift test, which confirms that corrosion bonding caused the high lift set drift.
The combination of materials used for the pilot disc and the pilot seat has been a known industry issue since the design of the Target Rock 2-stage SRV was initially installed. 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 and 2 with limited success. Option three has not been considered due to mixed industry results/performance.
Following a previous outage (H1 R1 5), 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.
SAFETY CONSEQUENCES AND IMPLICATIONS
There are no safety consequences associated with the 'as-found' setpoint drifts experienced during H1 R1 8. The safety valve function of the SRVs operates to prevent the reactor coolant, system from being pressurized above the Safety Limit of 1375 psig, which is 110% of the design pressure of 1250 psig, in accordance with the ASME Code.
A total of 13 operable SRVs are required to limit reactor pressure to within ASME III allowable values for the worst case transient. The five SRVs that were found above the allowable +3% were below the design pressure and Safety Limit; therefore, the five SRVs were capable of fulfilling their design function and reactor vessel overpressure protection was not compromised. With regard to Emergency Core Cooling System (ECCS) performance, no SRVs are required to open, and therefore, setpoint drift is not a concern. For a small break LOCA, the Automatic
Depressurization System (ADS) will cycle open SRVs. Two of the five SRVs (A' and 'D') were ADS valves; however, the setpoint drift had no impact on the ADS (electronically controlled) or manual function of the valves.
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."
PREVIOUS EVENTS A review of events for the past four years at Hope Creek was performed to determine if a similar event had occurred. Similar events occurred during the 2009 (H1 R1 5), 2010 (H1 R1 6), and 2012 (H1 R1 7) Hope Creek refueling outages when multiple SRVs were found out of the TS required limits of +/- 3%. These events were reported as LER 354/2009-002-01, LER 354/2010-002-01, and LER 354/2012-004-01.
A previously completed SRV Setpoint Drift root cause evaluation documented that the pilot stage assemblies in the Target Rock 2-Stage SRV design have an industry wide chronic history of corrosion bonding leading to setpoint drifting.
CORRECTIVE ACTIONS
- 1. All 14 SRV pilot stage assemblies were removed and replaced with pre-tested, certified spare pilot valves (HIR1 8).
- 2. Replace the currently installed Target Rock two-stage SRVs with a design that eliminates setpoint drift events exceeding +/-3% and improve SRV reliability.
COMMITMENTS
This LER contains no commitments.
FORM 366A (10-2010)