05000366/LER-2009-001, Regarding Safety Relief Valves Allowable Test Range Exceeded Due to Setpoint Drift
| ML091240260 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 05/04/2009 |
| From: | Madison D Southern Nuclear Operating Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| NL-09-0690 LER 09-001-00 | |
| Download: ML091240260 (5) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 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)(ix)(A) 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)(B), Prohibited by Technical Specifications |
| 3662009001R00 - NRC Website | |
text
Dennis R. Madison Southern Nuclear Vice President - Hatch Operating Company. Inc.
Plant Edwin I Hatch 1'1028 Hatch Parkway North Baxley, Georgia 31513 Tel 912.537.5859 Fax 912.3662077 SOUTHERNA COMPANY May 4,2009 Docket No.:
50-366 NL-09-0690 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555-0001 Edwin I. Hatch Nuclear Plant Licensee Event Report Safety Relief Valves Allowable Test Range Exceeded Due to Setpoint Drift Ladies and Gentlemen:
In accordance with the requirements of 10 CFR 50.73(a}(2}(i)(B), Southern Nuclear Operating Company is submitting the enclosed Licensee Event Report (LER) concerning safety relief valves allowable test range exceeded due to Setpoint drift.
This letter contains no NRC commitments. If you have any questions, please advise.
Sincerely,
- J~Y/1~
D. A. Madison Vice President - Hatch DRM/MJK/daj Enclosure: LER 2-2009-001 cc:
Southern Nuclear Operating Company Mr. J. T. Gasser, Executive Vice President Ms. P. M. Marino, Vice President - Engineering RTYPE: CHA02.004 U. S. Nuclear Regulatory Commission Mr. L. A. Reyes, Regional Administrator Mr. A. E. Martin, NRR Project Manager - Hatch Mr. J. A. Hickey, Senior Resident Inspector - Hatch
APPROVED BY OMB: NO. 3150-0104 EXPIRES: 0813112010
,.RC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION HOO7)
, the NRC may not conduct or sponsor, and a person is not required to ruspond to, the Information collection.
- 2. DOCKET NUMBER r,PAGE Edwin I. Hatch Nuclear Plant Unit 2
- 1. FACILITY NAME 05000 366 1 OF 4 14.11TLE Safety Relief Valves Allowable Test Range Exceeded Due to Setpoint Drift
- 5. EVENT DATE
- 6. LER NUMBER
- 7. REPORT DAllE
- 8. OTHER FACILITlES INVOLVED MONTH DAY YEAR YEAR SEOUENTIAL NUMBER REV NO.
MONTH DAY YEAR FACILITY NAME DOCKET NUMBER 05000 FACILITY NAME DOCKET NUMBER 03 12 2009 2009 - 001 -
0 05 04 2009 05000
- 11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR§: (Check all that apply)
- 19. OPERATlNG MODE o 20.2201 (b) o 20.2203(a)(3)(i) o 50.73(a)(2)(i)(C) o 50.73(a)(2)(vii) o 20.2201 (d) o 20.2203(a)(3)(ii) o 50.73(a)(2)(ii)(A) o 50.73(a)(2)(viii)(A) 5 o 20.2203(a)(1) o 20.2203(a)(4) o 50.73(a)(2)(ii)(B) o 50.73(a)(2)(viii)(B) o 20.2203(a)(2)(i) o 50.36(c)(1)(i)(A) o SO.73(a)(2)(iii) o 50.73(a)(2)(ix)(A) o 20.2203(a)(2)(ii) o 50.36(c)(1 )(ii)(A) o 50. 73(a)(2)(iv)(A) o 50.73(a)(2)(x)
- 10. POWER LEVEL o 20.2203(a)(2)(iii) o 50.36(c)(2) o 50.73(a)(2)(v)(A) o 73.71(a)(4) o 20.2203(a)(2)(iv) o 50.46(a)(3)(ii) o 50.73(a)(2)(v)(B) o 73.71 (a)(5) 0.00 o 20,2203(a)(2)(v) o 50. 73(a)(2)(i)(A) o 50,73(a)(2)(v)(C) o OTHER o 20.2203(a)(2)(vi) 181 50.73(a)(2)(i)(B) o SO.73(a)(2)(v)(D)
Specify in AbStract below or in NRC Fonn 366A
- 12. LICENSEE CONTACT FOR THIS LER FACILITY NAME I~ELEPHONE NUMBER (Include Area Code)
Edwin I. Hatch I Kathy Underwood, Perfonnance Improvement Supervisor 912-537-5931
- 13. COMPLETE ONE UNE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT MANU REPORTABLE REPORTABLE MANU*
SYSTEM COMPONENT
CAUSE
CAUSE SYSTEM COMPONENT FACTURER TO EPIX FACTURER TO EPIX RV SB T020 Yes B
- 14. SUPPLEMENTAL REPORT EXPECllED
- 15. EXPECllED SUBMISSION MONTH OAY YEAR DYES (If yes, complete 15. EXPECTED SUBMISSION DATE) 181 NO DATE ABSTRACT (Limit to 1400 spaces, i.e., approximately 15 single-spaced typewritten Jines)
On March 12,2009 at approximately 4:00 pm EDT, Unit 2 was in refuel mode at 0 percent power. On that day, it was detennined that during bench testing at an independent testing facility five Safety Relief Valves (SRVs) experienced setpoint drift that exceeded the allowable plant Technical Specifications (TS) limit above the setpoint value.
The initially identified cause of the SRV setpoint drift above the setpoint value is corrosion-induced bonding between the pilot disc and seating surface.
Immediate corrective actions for this event included replacement of the SRVs with refurbished pilot valves with discs made from stellite 21 that have been certified to actuate within 11.5 psi of the setpoint. Each of the pilot discs from the valves removed for testing have been replaced with a pilot disc made from Stellite 21 material. Evaluation of additional actions to further improve SRV perfonnance will be tracked under the plant's corrective action program.
PRINTED ON RECYCLED PAPER NRC FORM 366 (!HOO7)
(If more space ;s required, use additional copies of NRC Form 366A)
PLANT AND SYSTEM IDENTIFICATION
General Electric - Boiling Water Reactor Energy Industry Identification System codes appear in the text as (EllS Code XX).
DESCRIPTION OF EVENT
On March 12,2009 at approximately 4:00 pm EDT, Unit 2 was in refuel mode at 0 percent power. On that day, it was determined that during bench testing at an independent testing facility five Safety Relief Valves (SRVs) (EllS Code SB) experienced setpoint drift that exceeded the allowable plant Technical Specifications (TS) limit above the setpoint value.
All eleven SRV's were tested. Of those eleven, five failed the as found testing. The following is a tabulation of the test results for the five SRVs that failed the as-found test:
MPLNumber Pilot Serial Number As-Found Lift Pressure Percent Drift 2B21-F013A 302 1193 103.7 2B21-F013B 315 1209 105.1 2B21-F013D 314 1200 104.3 2B21-F013H 307 1204 104.7 2B21-F013M 1005 1187 103.2 These valves were removed from service during a planned refuel outage and replaced with like kind valves that were serviced and tested in accordance with plant procedures.
CAUSE OF EVENT
The initially identified cause of the SRV setpoint drift above the setpoint value is corrosion induced bonding between the pilot disc and seating surface.
REPORTABILITY ANALYSIS AND SAFETY ASSESSMENT This event is reportable per 50.73(a)(2)(i)(B) because an event occurred which is prohibited by Technical Specifications (TS). Specifically, multiple test failures of the SRVs is defined as reportable in NUREG-1022, Revision 2, dated October 2000, in section 3.2.2, example 3, titled "Multiple Test Failures."
The SRVs, which are located on the four main steam lines within the drywell between the reactor vessel and the inboard main steam isolation valves (MSIV EllS Code SB), are required during Modes 1,2, and 3 to limit the peak pressure in the nuclear system such that it will not exceed the applicable ASME Boiler and Pressure Vessel Code limits for the reactor PRINTED ON RECYCLED PAPER
NRC FORM 368A (9-2007)
LICENSEE EVENT REPORT (LER) u.s. NUCLEAR REGULATORY COMMISSION CONTINUATION SHEET
- 1. FACILITY NAME
- 2. DOCKET
- 6. LER NUMBER
- 3. PAGE Edwin 1. Hatch Nuclear Plant Unit 2 05000366 YEAR I SEQUENTIAL IREVISION NUMBER NUMBER 3
OF 4
2009 001 0
coolant pressure boundary. Per TS Surveillance Requirement 3.4.3.1, the valves are tested in accordance with the In-service Testing Program to verify the safety function lift setpoints are within the specified limits.
The impact of the "as found" setpoints for these safety relief valves was analyzed using the most severe pressurization transient which, for the purposes of demonstrating compliance with the ASME Code limit of 1375 psi peak vessel pressure, has been defined as a closure of all MS IVs with a failure of the direct reactor protection system trip from the MS IV position switches. The reactor ultimately shutdowns from a high neutron flux trip. Analysis of this event using the as-found bench test results for the tested SRV's concluded that there would have been at least a 50 psi margin to the 1375 psi ASME Boiler and Pressure Vessel Code overpressure limit. Even though this transient is not an anticipated operational occurrence (AOO), the analysis demonstrates that even under the extreme conditions assumed adequate margin to the ASME Code limit of 1375 psi still exists.
The plant Technical Specifications overpressure safety limit of 1325 psi dome pressure must be met during normal operations and for anticipated operational occurrences (AOOs). The analysis of the as-found test results also showed that there is approximately a 35 psi margin to the 1325 psi Tech Spec Safety Limit during the limiting MSIVF event in Hatch-2 Cycle 20.
In addition, a non-credited electrical actuation system was installed in 1993 to ensure proper actuation of the SRVs. This system provides a redundant, independent method (Le., electrical signal) to actuate the SRVs. During the run cycle the redundant electrical system was available. The system was procured to Class IE environmental and seismic standards, and is deemed highly reliable.
Based on this analysis, it is concluded that this event had no adverse impact on nuclear safety.
CORRECTIVE ACTIONS
All pilot valves have been replaced with refurbished pilot valves which have been certified to actuate within 11.5 psi of the setpoint and have disc made from stellite 21 material.
Each of the pilot discs from the valves removed for testing will be replaced with a pilot disc made from stellite 21 material. Implementation will be tracked under the corrective action program.
Any additional actions to further improve SRV performance will be tracked under the plant's corrective action program.
ADDITIONAL INFORMAnON Other Systems Affected: None NRC FORM ae6A (&-2007)
PRINTED ON RECYCLED PAPER
NRC FORM 386A (9-2007)
LICENSEE EVENT REPORT (LER)
CONTINUATION SHEET U.S. NUCLEAR REGULATORY COMMISSION Edwin 1. Hatch Nuclear Plant Unit 2
- 1. FACIUTY NAME 05000366
- 2. DOCKET YEAR I SEQUENTIAL IREVISION NUMBER NUMBER
- 6. LER NUMBER 4
OF 4
- 3. PAGE 2009 001 0
Failed Components Information
Master Parts List Number: 2B21-F013 Manufacturer: Target Rock Model Number: 7567F Type: Relief Valve Manufacturer Code: Ta20 EllS System Code: SB Reportable to EPIX: Yes Root Cause Code: B EllS Component Code: RV Commitment Information: This report does not create any new permanent licensing
commitments
Previous Similar Events
LER 2-2008-004; identified multiple SRV setpoint drift for three of the four tested SRV's.
Corrective actions for that LER, replacement of discs were implemented but discs made of stellite 21 for the Unit 2 SRV's were not available for all of the replaced discs and thus could not have prevented the current event.
LER 1-2008-002; identified multiple SRV setpoint drift for three of the eleven SRV's.
Corrective actions for that LER, replacement of discs with stellite 21 discs, were not yet implemented for the Unit 1 SRV's and thus could not have prevented the current event.
LER 2-2007-006; identified multiple SRV setpoint drift for five of the eleven SRV's.
Corrective actions for this LER, replacement of discs with stellite 21 discs, were not yet implemented for the Unit 2 SRV's and thus could not have prevented the current event.
LER 1-2006-003; which identified an error in reporting multiple SRV setpoint drift, also described results from the previous three outages where multiple SRV setpoint drift had occurred. Corrective actions for this LER focused on ensuring the proper reporting of SRV se!point drift was performed.
NRC FOFlM 361SA (9-2007)
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