05000265/LER-2006-001
| Docket Numbersequential Revmonth Day Year Year Month Day Year N/Anumber No. N/A | |
| Event date: | 04-10-2006 |
|---|---|
| Report date: | 07-27-2007 |
| Reporting criterion: | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications |
| 2652006001R01 - NRC Website | |
FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER (6 PAGE (3) Quad Cities Nuclear Power Station Unit 2 05000265 NUMBER NUMBER (If more space is required, use additional copies of NRC Form 366A)(17)
PLANT AND SYSTEM IDENTIFICATION
General Electric - Boiling Water Reactor, 2957 Megawatts Thermal Rated Core Power Energy Industry Identification System (EIIS) codes are identified in the text as [XX].
EVENT IDENTIFICATION
Two Main Steam Safety Valves and Two Main Steam Safety/Relief Valves Outside of the Technical Specification Allowed Tolerance A.C CONDITION PRIOR TO EVENT Unit: 2 Event Date: April 10, 2006E Event Time: 1400 hours0.0162 days <br />0.389 hours <br />0.00231 weeks <br />5.327e-4 months <br /> Reactor Mode: 5 Mode Name: Refueling Power Level: 000%
B. DESCRIPTION OF EVENT
On April 10, 2006, at approximately 1400 hours0.0162 days <br />0.389 hours <br />0.00231 weeks <br />5.327e-4 months <br />, Quad Cities Nuclear Power Station determined that two of the four Main Steam Safety Valves (MSSVs) [V] [SB] removed from Unit 2 during the Spring 2006 refuel outage (Q2R18) had been found during as found testing to have lift set pressures 1.9% and 1.6% below nameplate. These values are outside of the +/-1% Technical Specification (TS)-allowed tolerance.
Both of the MSSVs had lift set pressures inside the +/-3% ASME Code tolerance.
Also, it was determined that the Main Steam Safety/Relief Valve (SRV) [RV] removed from Unit 2 during a planned outage in April of 2005 (Q2P03) had been found during as-found testing to have a lift set pressure 5.4% above nameplate, which is outside of both the +/-1% TS allowed tolerance and the +/-3% ASME Code tolerance.
Finally, it was determined that the SRV installed on Unit 2 during Q2P03 and removed during Q2R18 had been found during as-found testing to have a lift set pressure 3.7% above nameplate, which is outside of both the +/-1% TS-allowed tolerance and the +/-3% ASME Code tolerance.
All four of the removed MSSVs and the SRV were replaced during Q2R18 with refurbished valves that were certified to be within the +/-1% TS-allowed tolerance.
C. CAUSE OF EVENT
Based on the results of testing and valve disassembly and inspection, the cause of the out-of-tolerance condition for the MSSVs and the SRV removed during Q2R18 and for the SRV removed during Q2P03 is setpoint drift. No mechanical wear, degradation or foreign material was identified.
FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER (6 PAGE (3) Quad Cities Nuclear Power Station Unit 2 05000265 NUMBER NUMBER (If more space is required, use additional copies of NRC Form 366A)(17)
D. SAFETY ANALYSIS
The safety significance of this event was minimal. Both of the MSSVs were found to have a lift set pressure below (i.e., conservative with respect to) the nameplate value and inside the +/-3% Code tolerance. The analysis completed for the April 2004 Unit 2 SRV out-of-tolerance event (LER 265/04-004) bounds the test results described above. That analysis showed that the acceptance criteria for the Anticipated Transient Without Scram, ASME overpressure, and Appendix R analyses were met. Therefore, the valves were capable of performing the safety function.
This condition is being reported in accordance with 10 CFR 50.73(a)(2)(i)(B), which requires reporting of any operation or condition that was prohibited by the plant's TS.
E. CORRECTIVE ACTIONS
All four of the removed MSSVs and the removed SRV were replaced during Q2R18 with refurbished valves that were certified to be within the +/-1% TS-allowed tolerance.
Quad Cities Nuclear Power Station is pursuing a revision to the TS-allowable value for the MSSVs and SRVs to reflect the ASME code allowable value.
Quad Cities Nuclear Power Station continues to monitor and contribute to industry efforts to improve SRV performance through improved refurbishment process.
Additionally, benchmarking efforts are in progress to determine what changes can be made to station processes to improve SRV performance.
Some improved SRV performance is anticipated in response to the greatly diminished Main Steam Line vibration levels as a result of installation in 2006 of Acoustic Side Branches on the MSSV risers.
F. PREVIOUS OCCURRENCES
There have been previous instances of MSSVs and SRVs being outside of the TS allowed value (+/-1%). Following the Unit 1 refuel outage in October of 2000 (Q1R16), the SRV setpoint was 2.203% lower than nameplate, one MSSV setpoint was 2.0643% greater than nameplate, and one MSSV setpoint was 1.20% greater than nameplate. Following the Unit 2 refuel outage in February of 2002 (Q2R16), the SRV setpoint was 2.026% greater than nameplate, one MSSV setpoint was 2.8% less than nameplate, one MSSV setpoint was 1.8% less than nameplate, and one MSSV setpoint was 1.5% less than nameplate. Following the Unit 1 refuel outage in November of 2002 (Q1R17), the SRV setpoint was 2.203% greater than nameplate and one MSSV setpoint was 1.2% lower than nameplate. Following the Unit 2 refuel outage in March 2004 (Q2R17), the SRV setpoint was 6.8% greater than nameplate and one MSSV refuel outage in April 2005 (Q1R18), one MSSV was 1.7% lower than nameplate, one MSSV was 2.3% lower than nameplate, and one MSSV was 2.0% lower than nameplate.
Following the Unit 2 refuel outage in Spring 2006 (Q2R18), one MSSV setpoint was found 1.9% below nameplate, one MSSV was found 1.6% below nameplate, an SRV removed FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER (6 PAGE (3) Quad Cities Nuclear Power Station Unit 2 05000265 NUMBER NUMBER (If more space is required, use additional copies of NRC Form 366A)(17) during a mid-cycle outage was found to be 5.4% above nameplate, and the SRV removed during Q2R18 was found to be 3.7% above nameplate. Following the Unit 1 Spring 2007 refuel outage (Q1R19), one MSSV setpoint was 1.4% lower than nameplate, one MSSV setpoint was 1.3% above nameplate, and the SRV setpoint was 2.7% above nameplate.
For every case except the Q2R17 and Q2R18 SRVs, the setpoint was within the ASME code allowable of +/-3%, and therefore there was no effect on functionality. For the Q2R17 and Q2R18 SRVs, specific assessments were performed to show that the safety valve function was met.
Based on the history described above, Quad Cities Nuclear Power Station has submitted a revision to the TS-allowable value for the MSSVs and SRVs to reflect the ASME code allowable.
COMPONENT FAILURE DATA G.
The MSSVs are Model 6'-3777-QA-RT Safety Valves manufactured by Dresser Industries/ Consolidated Valve Corporation. The SRVs are Model 7467F Safety/ Relief Valves manufactured by Target Rock.