05000254/LER-2007-001

LER-2007-001, Quad Cities Nuclear Power Station Unit 1 05000254 1 of 3

Docket Number
Event date:	05-16-2007
Report date:	07-16-2007
Reporting criterion:	10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications
2542007001R00 - NRC Website

FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER (6 PAGE (3) Quad Cities Nuclear Power Station Unit 1 05000254 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 One Main Steam Safety/Relief Valve Outside of Technical Specification Allowed Tolerance Due to Setpoint Drift

A. CONDITION PRIOR TO EVENT

Unit: 1 Event Date: May 16, 2007 Event Time: 1200 hours0.0139 days <br />0.333 hours <br />0.00198 weeks <br />4.566e-4 months <br /> Reactor Mode: 5 Mode Name: Refuel Power Level: 000%

B. DESCRIPTION OF EVENT

On May 16, 2007, Quad Cities Station received as-found test results that showed that two of the four Main Steam Safety Valves (MSSV)[SB] that were removed during the Spring 2007 refuel outage (Q1R19) actuated outside of the +/- 1% band required by Technical Specifications (TS). One valve actuated at -1.4% and one valve actuated at +1.3%. On May 22, 2007, as-found test results were received showing that the set pressure for the safety function of the Main Steam Safety/Relief Valve (SRV) removed during Q1R19 was outside of the +/- 1% band required by TS. The SRV actuated at +2.7%. In all cases, the results were within the +/- 3% ASME Code criteria.

All four of the removed MSSVs and the SRV were replaced during Q2R18 with newly 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 SRV is setpoint drift. No mechanical wear, degradation or foreign material associated with the pilot section of the valve was identified. Based on the results of testing and historical performance, the cause of the out-of-tolerance condition for the MSSVs is also setpoint drift.

D. SAFETY ANALYSIS

The safety significance of this event was minimal. One of the MSSVs was found to have a lift set pressure below (i.e., conservative with respect to) the nameplate value. Both of the MSSVs and the SRV were found to actuate inside the +/-3% Code tolerance. The accident analyses for the fuel cycle during which these valves were installed assumed 3% tolerance for all installed MSSV and SRV valves. This 3% requirement is likewise utilized for the current fuel cycles on both units.

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 FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER (6 PAGE (3) (If more space is required, use additional copies of NRC Form 366A)(17) 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 SRV were replaced during Q2R18 with newly refurbished valves that were certified to be within the +/-1% TS-allowed tolerance.

Quad Cities Nuclear Power Station has submitted a License Amendment request to revise the TS-allowable value for the MSSVs and SRVs to reflect the ASME code allowable. Reference November 7, 2006, letter (RS-06-147), D.M. Benyak (EGC) to U.S. Nuclear Regulatory Commission, "Request for License Amendment to Increase Main Steam Safety Valve Lift Setpoint Tolerance and Standby Liquid Control System Enrichment.

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 setpoint was 2.339% greater than nameplate (LER 265/04-001). Following the Unit 1 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 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.

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.

G. COMPONENT FAILURE DATA

The MSSVs are Model No. 6'-3777-QA-RT Safety Valves manufactured by Dresser Industries/ Consolidated Valve Corporation. The SRV is a Model 7467F Safety/Relief Valve manufactured by Target Rock.