05000254/LER-2007-001, Regarding Two Main Steam Safety Valves and One Main Steam Safety/Relief Valve Outside of Technical Specification Allowed Tolerance Due to Setpoint Drift
| ML072190603 | |
| Person / Time | |
|---|---|
| Site: | Quad Cities  |
| Issue date: | 07/16/2007 |
| From: | Tulon T Exelon Generation Co, Exelon Nuclear |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| SVP-07-045 LER 07-001-00 | |
| Download: ML072190603 (4) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications |
| 2542007001R00 - NRC Website | |
text
Exel n@
Exelon Generation Company, LLC www.exeloncorp.com Nuclear Quad Cities Nuclear Power Station 22710 206h Avenue North Cordova, IL 61242-9740 July 16, 2007 SVP-07-045 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555 Quad Cities Nuclear Power Station, Unit 1 Renewed Facility Operating License No. DPR-29 NRC Docket No. 50-254
Subject:
Licensee Event Report 254/07-001, "Two Main Steam Safety Valves and One Main Steam Safety/Relief Valve Outside of Technical Specification Allowed Tolerance Due to Set Point Drift" Enclosed is Licensee Event Report (LER) 254/07-001, "Two Main Steam Safety Valves and One Main Steam Safety/Relief Valve Outside of Technical Specification Allowed Tolerance Due to Set Point Drift," for Quad Cities Nuclear Power Station, Unit 1.
This report is submitted in accordance with the requirements of the Code of Federal Regulations, Title 10, Part 50.73(a)(2)(i)(B), which requires the reporting of any operation or condition that was prohibited by the plant's Technical Specifications.
There are no regulatory commitments contained in this letter.
Should you have any questions concerning this report, please contact Mr. W. J. Beck at (309) 227-2800.
Respectfully, Ti thy J. Tulon Site Vice President Quad Cities Nuclear Power Station cc:
Regional Administrator - NRC Region Il NRC Senior Resident Inspector - Quad Cities Nuclear Power Station
Abstract
On May 16,
- 2007, Quad Cities Station received as-found test results that showed that two of the four tested Main Steam Safety Valves actuated outside of the +/-
1% set pressure band required by Technical Specifications.
On May 22, 2007, as found test results were received showing that the Main Steam Safety/Relief Valve set pressure was outside of the +/-
1% band required by Technical Specifications.
In all
- cases, the results were within the +/-
3% ASME Code criteria.
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.
The safety significance of this event was minimal.
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.U.S. NUCLEAR REGULATORY COMMISSION (7-2001)
LICENSEE EVENT REPORT (LER)
TEXT CONTINUATION FACILITY NAME (1)
DOCKET NUMBER (2)
LER NUMBER (6 PAGE (3)
YEAR SEQUENTIAL REVISION Quad Cities Nuclear Power Station Unit 1 05000254 NUMBER NUMBER 2007 001 00 2 of 3 (If more space is required, use additional copies of NRC Form 366A)(1 7)
PLANT AND SYSTEM IDENTIFICATION
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 (QlRl9) 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 QlRl9 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 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 Q2Rl8 with newly refurbished valves that were certified to be within the +/-I% 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 (+/-l%).
Following the Unit 1 refuel outage in October of 2000 (QIRl6),
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 (QlRl7),
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 (QIRl8),
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 Q2RI8 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.