05000366/LER-2008-004, Safety Relief Valves Allowable Test Range Exceeded Due to Setpoint Drift
| ML082260006 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 08/12/2008 |
| From: | Madison D Southern Nuclear Operating Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| NL-08-1254 LER 08-004-00 | |
| Download: ML082260006 (5) | |
| 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 |
| 3662008004R00 - NRC Website | |
text
D. R. Madison (Dennis)
Vice President - Hatch August 12,2008 Docket No.:
50-366 Southern Nuclear Operating Company, Inc.
Plant Edwin I. Hatch 11028 Hatch Parkway, North Baxley, Georgia 31513 Tel 912.537.5859 Fax 912366.2077 SOUTHERN'\\.
COMPANY Energy to Serve Yt1ur Wor/d SM NL-08-1254 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555-0001 Edwin I. Hatch Nuclear Plant - Unit 2 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 lift setpoint drift in more than one Safety Relief Valve.
This letter contains no NRC commitments. If you have any questions, please advise.
Sincerely,
~:~~
Vice President - Hatch DRM/MJK/daj Enclosure: LER 2-2008-004 cc:
Southern Nuclear Operating Company Mr. J. T. Gasser, Executive Vice President Mr. D. H. Jones, Vice President - Engineering RTYPE: CHA02.004 U. S. Nuclear Regulatory Commission Mr. L. A. Reyes, Regional Administrator Mr. R. E. Martin, NRR Project Manager - Hatch Mr. J. A. Hickey, Senior Resident Inspector - Hatch
NRC FORM 366 (9-2007)
PRINTED ON RECYCLED PAPER NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION (9-2007)
LICENSEE EVENT REPORT (LER)
APPROVED BY OMB: NO. 3150-0104 EXPIRES: 08/31/2010
, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.
- 1. FACILITY NAME Edwin I. Hatch Nuclear Plant Unit 2
- 2. DOCKET NUMBER 05000 366
- 3. PAGE 1 OF 4
- 4. TITLE Safety Relief Valves Allowable Test Range Exceeded Due to Setpoint Drift
- 5. EVENT DATE
- 6. LER NUMBER
- 7. REPORT DATE
- 8. OTHER FACILITIES INVOLVED MONTH DAY YEAR YEAR SEQUENTIAL NUMBER REV NO.
MONTH DAY YEAR FACILITY NAME DOCKET NUMBER 05000 07 01 2008 2008 - 004 -
0 08 12 2008 FACILITY NAME DOCKET NUMBER 05000
- 9. OPERATING MODE 1
- 10. POWER LEVEL 99.8
- 11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR§: (Check all that apply) 20.2201(b) 20.2203(a)(3)(i) 50.73(a)(2)(i)(C) 50.73(a)(2)(vii) 20.2201(d) 20.2203(a)(3)(ii) 50.73(a)(2)(ii)(A) 50.73(a)(2)(viii)(A) 20.2203(a)(1) 20.2203(a)(4) 50.73(a)(2)(ii)(B) 50.73(a)(2)(viii)(B) 20.2203(a)(2)(i) 50.36(c)(1)(i)(A) 50.73(a)(2)(iii) 50.73(a)(2)(ix)(A) 20.2203(a)(2)(ii) 50.36(c)(1)(ii)(A) 50.73(a)(2)(iv)(A) 50.73(a)(2)(x) 20.2203(a)(2)(iii) 50.36(c)(2) 50.73(a)(2)(v)(A) 73.71(a)(4) 20.2203(a)(2)(iv) 50.46(a)(3)(ii) 50.73(a)(2)(v)(B) 73.71(a)(5) 20.2203(a)(2)(v) 50.73(a)(2)(i)(A) 50.73(a)(2)(v)(C)
OTHER 20.2203(a)(2)(vi) 50.73(a)(2)(i)(B) 50.73(a)(2)(v)(D)
Specify in Abstract below or in NRC Form 366A
- 12. LICENSEE CONTACT FOR THIS LER FACILITY NAME Edwin I. Hatch / Kathy Underwood, Performance Analysis Supervisor TELEPHONE NUMBER (Include Area Code) 912-537-5931 CAUSE SYSTEM COMPONENT MANU-FACTURER REPORTABLE TO EPIX
CAUSE
SYSTEM COMPONENT MANU-FACTURER REPORTABLE TO EPIX B
SB RV T020 Yes
- 14. SUPPLEMENTAL REPORT EXPECTED YES (If yes, complete 15. EXPECTED SUBMISSION DATE)
NO
- 15. EXPECTED SUBMISSION DATE MONTH DAY YEAR ABSTRACT (Limit to 1400 spaces, i.e., approximately 15 single-spaced typewritten lines)
On July 1, 2008 at approximately 1:00 pm EDT, Unit 2 was at 2799 CMWTh, which is 99.8 percent of rated thermal power (RTP). On that day, it was determined that during bench testing at an independent testing facility two Safety Relief Valves (SRVs) experienced setpoint drift that exceeded the allowable plant Technical Specifications (TS) limit above the setpoint value. One other SRV exceeded the allowable plant Technical Specifications (TS) limit by being below the setpoint value.
The root cause of the SRV setpoint drift above the setpoint value is corrosion-induced bonding between the pilot disc and seating surface. The cause of the SRV setpoint drift below the setpoint value is initially identified to be a reduction in spring force.
Immediate corrective actions for this event included replacement of the SRVs with refurbished pilot valves which have been certified to actuate within 11.5 psig of the setpoint.
Each of the four pilot discs from the valves removed for testing will be replaced with a pilot disc made from Stellite 21 material. Evaluation of additional actions to further improve SRV performance will be tracked under the plants corrective action program.
(If more space is 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 (EIIS Code XX).
DESCRIPTION OF EVENT
On July 1, 2008 at approximately 1:00 pm EDT, Unit 2 was at 2799 CMWTh, which is 99.8 percent of rated thermal power (RTP). On that day, it was determined that during bench testing, at an independent testing facility, two Safety Relief Valves (SRVs); (EIIS Code SB) experienced setpoint drift that exceeded the allowable plant Technical Specifications (TS) limit above the setpoint value. One other SRV exceeded the allowable plant Technical Specifications (TS) limit below the setpoint value. The following is a tabulation of the test results for the four SRVs removed from the plant tested:
MPL Number Pilot Serial Number As-Found Lift Pressure Percent Drift 2B21-F013A 1187 1109 96.4 2B21-F013D 305 1266 110.1 2B21-F013E 1188 1175 102.2 2B21-F013L 316 1190 103.5 These valves were removed from service during a planned maintenance outage and replaced with like kind valves that were serviced and tested in accordance with plant procedures.
CAUSE OF EVENT
The root cause of the SRV setpoint drift above the setpoint value is corrosion-induced bonding between the pilot disc and seating surface.
For the SRV setpoint drift below the allowed range, internal inspection identified no issues with valve seating surfaces or pilot disc that would correlate to the valve set-point decrease.
During testing it was noted that the spring force was reduced. Further investigation will be performed to determine the cause of this set-point failure; however the cause of failure is initially identified as the spring force reduction at this point in the investigation. This investigation is scheduled for completion by October 2, 2008. If a different cause or a change in corrective action(s) results from the ongoing investigation, then a revision to this LER will be submitted. (9-2007)
LICENSEE EVENT REPORT (LER)
CONTINUATION SHEET U.S. NUCLEAR REGULATORY COMMISSION
- 1. FACILITY NAME
- 2. DOCKET
- 6. LER NUMBER
- 3. PAGE YEAR SEQUENTIAL NUMBER REVISION NUMBER Edwin I. Hatch Nuclear Plant Unit 2 05000366 2008 004 0
3 OF 4
PRINTED ON RECYCLED PAPER 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 EIIS 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 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 safety relief valves must accommodate the most severe pressurization transient which, for the purposes of demonstrating compliance with the ASME Code limit of 1375 psig peak vessel pressure, has been defined as a closure of all MSIVs with a failure of the direct reactor protection system trip from the MSIV 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 four tested SRVs along with the conservative assumptions that one SRV is out of service and the other six valves lift at the maximum technical specification allowed limit of 1184.5 psig demonstrated that the resultant peak pressure was within the ASME Code limit.
Furthermore, the plant Technical Specifications overpressure safety limit of 1325 psig 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 for the MSIV Closure AOO with the MSIV position switches providing the reactor protection system trip, the resultant dome pressure was within the plant Technical Specifications Safety Limit.
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 (i.e., electrical signal) to actuate the SRVs. During the run cycle the redundant electrical system was available. The system was procured to Class 1E 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 four pilot valves have been replaced with refurbished pilot valves which have been certified to actuate within 11.5 psig of the setpoint. (9-2007)
LICENSEE EVENT REPORT (LER)
CONTINUATION SHEET U.S. NUCLEAR REGULATORY COMMISSION
- 1. FACILITY NAME
- 2. DOCKET
- 6. LER NUMBER
- 3. PAGE YEAR SEQUENTIAL NUMBER REVISION NUMBER Edwin I. Hatch Nuclear Plant Unit 2 05000366 2008 004 0
4 OF 4
PRINTED ON RECYCLED PAPER Each of the four 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.
SNC will continue to participate in industry working groups in evaluating potential solutions to this industry issue. Any additional actions to further improve SRV performance will be tracked under the plants corrective action program.
As stated in LER 2-2007-006, each of the eleven 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.
ADDITIONAL INFORMATION
Other Systems Affected: None
Failed Components Information
Master Parts List Number: 2B21-F013 EIIS System Code: SB Manufacturer: Target Rock Reportable to EPIX: Yes Model Number: 7567F Root Cause Code: B Type: Relief Valve EIIS Component Code: RV Manufacturer Code: T020 Commitment Information: This report does not create any new permanent licensing
commitments
Previous Similar Events
LER 1-2008-002; identified multiple SRV setpoint drift for three of the eleven SRVs.
Corrective actions for that LER, replacement of disc with stellite 21 disc, were implemented for the Unit 1 SRVs but not yet implemented for the Unit 2 SRVs and thus could not have prevented the current event.
LER 2-2007-006; identified multiple SRV setpoint drift for five of the eleven SRVs.
Corrective actions for this LER, replacement of disc with stellite 21 disc, were not yet implemented for the Unit 2 SRVs 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 setpoint drift was performed.