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MONTHYEARML0330202682002-10-24024 October 2002 Technical Specification Pages, Amendment 150 Project stage: Other 05000528/LER-2003-001, From Palo Verde, Unit 1 Regarding Pressurizer Safety Valve as Found Lift Pressure Outside of TS Limits2003-04-25025 April 2003 From Palo Verde, Unit 1 Regarding Pressurizer Safety Valve as Found Lift Pressure Outside of TS Limits Project stage: Request ML0312700902003-04-25025 April 2003 Responses to Request for Additional Information to Proposed Amendment to TS 3.2.4, Departure from Nucleate Boiling Ratio, 3.3.1, Reactor Protective System Instrumentation - Operating, 3.3.3, Control Element Assembly Calculators Project stage: Response to RAI ML0312108382003-05-0505 May 2003 & 05/15/2003, Palo Verde, Units 1, 2 & 3, Forthcoming Meeting with Arizona Public Service Co., to Discuss Additional Information Needed for Review of Amendment Request Project stage: Meeting ML0312705492003-05-0707 May 2003 05/14-15/2003, Palo Verde, Units 1, 2, & 3, Meeting Notice with Arizona Public Service, Upgrading Core Protection Calculator System. Change in Time & Discussion of Proprietary Information Project stage: Meeting ML0313505732003-05-14014 May 2003 Meeting Handouts to Discuss the Licensees Application Dated November 7, 2002, to Upgrade the Core Protection Calculator System Project stage: Meeting ML0316813092003-06-17017 June 2003 Request for Withholding Information from Public Disclosure, Core Protection Calculator Upgrade Amendment Request Project stage: Withholding Request Acceptance ML0317502262003-06-24024 June 2003 Summary of Meeting with Arizona Public Service Co. for Palo Verde Units 1, 2 & 3 Project stage: Meeting ML0322407112003-08-12012 August 2003 8/12/03 - Palo Verde Ngs, Units 1,2, & 3 - Notice of Consideration of Issuance of Amds to Facility Operating Licenses, Proposed No Significant Hazards Consideration Determination and Opportunity for a Hearing. (Tacs. MB6726/6727/6728) Project stage: Other ML0325215182003-09-0808 September 2003 Summary of Meetings with Arizona Public Service Co for Palo Verde Nuclear Generating Station Tacs MB6726, MB6727, MB6728 Project stage: Meeting ML0328300272003-09-18018 September 2003 Response to RAI to Proposed Amendments to Technical Specifications 3.2.4, 3.3.1, and 3.3.3 Project stage: Request ML0328703072003-10-0101 October 2003 Special Report 1-SR-2003-002 Re Boron Deposit at Control Element Drive Mechanism Vent Project stage: Other ML0329012052003-10-15015 October 2003 Ngs, Units 1, 2&3- Request for Withholding Information from Public Disclosure for Core Protection Calculator Upgrade Amend Request.(Tac. MB6726/MB6727/MB6728) Project stage: Withholding Request Acceptance ML0330303632003-10-24024 October 2003 Ngs, Units 1, 2 & 3 - Issuance of Amds 150 (3) on the Core Protection Calculator System Upgrade (Tacs. MB6726/MB6727/MB6728) Project stage: Approval ML0332103022003-11-17017 November 2003 Letter, Correction to Amendment No. 150 Project stage: Other 2003-05-14
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LER-2003-001, From Palo Verde, Unit 1 Regarding Pressurizer Safety Valve as Found Lift Pressure Outside of TS Limits |
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10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications |
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r a8LAMS 1 OCFR50.73 Gregg R. Overbeck Mail Station 7602 Palo Verde Nuclear Senior Vice President TEL (623) 393-5148 P.O Box 52034 Generating Station Nuclear FAX (623) 393-6077 Phoenix, AZ 85072-2034 192-01116-GRO/SAB/DJS April 25,2003 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Mail Station P1-37 Washington, DC 20555-0001
Dear Sirs:
Subject:
Palo Verde Nuclear Generating Station (PVNGS)
Unit 1 Docket No. STN 50-528 License No. NPF-41 Licensee Event Report 2003-001-00 Attached please find Licensee Event Report (LER) 50-528/2003-001-00 that has been prepared and submitted pursuant to 10CFR50.73(a)(2)(i)(B). This LER reports the findings and corrective actions taken in response to a single, out of tolerance (OOT) pressurizer safety valve (PSV) which was discovered during post-outage testing The as-found lift pressure for one Unit 1 PSV was outside of the tolerance allowed by Technical Specification Limiting Condition for Operation 3.4.10. The PSVs removed from Unit 1 were as-found lift tested, disassembled, inspected, reassembled and certified at NWS Technologies laboratories.
The actions taken as a result of the out of tolerance PSV are being performed in accordance with the PVNGS corrective action program. The corrective actions described in this LER are not necessary to maintain compliance with regulations. As such, APS may modify these corrective actions as necessary to improve PSV reliability and performance.
In accordance with 10CFR50.73 (d), a copy of this LER is being forwarded to the NRC Regional Office, NRC Region IV and the Resident Inspector. If you have questions regarding this submittal, please contact Daniel G. Marks, Section Leader, Regulatory Affairs, at (623) 393-6492.
Arizona Public Service Company makes no commitments in this letter.
Sincerely, G RO/SAB/DJS/kg Attachment cc:
E. W. Merschoff N. L. Salgado J. N. Donohew (all with attachment)
APPROVED BY OMB NO. 3150-0104 EXPIRES 7-31-2004
Abstract
On March 5, 2003, lift pressure verification testing was completed on pressurizer safety valves (PSVs) that had been removed during the Unit 1 tenth refueling outage. The testing revealed that the as-found lift pressure for one Unit 1 PSV was outside the Technical Specification limits of +3, -1 percent of design lift pressure.
The out of tolerance as-found PSV condition appears to be the result of a spring whose physical condition relaxed over time and would no longer meet vendor specifications. The impact of the Unit 1 PSV Out of Tolerance (OOT) was evaluated and it was determined the results based on the as-found conditions were bounded by the peak Reactor Coolant System (RCS) pressure results of the current Loss of Condenser Vacuum (LOCV) analysis of record (AOR) (A-PV2-FE-01 60, Rev 2). LOCV is the most limiting event for peak primary pressure that is impacted by the PSV high OOT condition.
Previous similar events have been reported in the last three years.
NRC tFUHI 366 tf-)
(If more space is required, use additional copies of (If more space is required, use additional copies of (If more space is required, use additional copies of (If more space is required, use additional copies of NRC Form 366A) 8
PREVIOUS SIMILAR EVENTS
Similar out-of-tolerance PSV conditions were reported in previously submitted LERs 529/1999-004 and 529/2000-008, where setpoint drift was the cause of the out-of-tolerance conditions. In this instance, the out-of-tolerance condition appears to be the result of spring relaxation. Although previous corrective actions have been effective in reducing the number of out-of-tolerance PSVs, as-found out-of-tolerance conditions periodically occur. APS evaluates industry operating experience for corrective actions that may improve PSV performance and APS may implement additional actions if they are demonstrated to be effective.
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| 05000529/LER-2003-001, Regarding Reactor Trip with Loss of Forced Circulation Due to Failed Pressurizer Main Spray Valve | Regarding Reactor Trip with Loss of Forced Circulation Due to Failed Pressurizer Main Spray Valve | 10 CFR 50.73(a)(2)(v), Loss of Safety Function | | 05000530/LER-2003-001, Regarding Main Steam Safety Valve as Found Lift Pressures Outside of TS Limits | Regarding Main Steam Safety Valve as Found Lift Pressures Outside of TS Limits | 10 CFR 50.73(a)(2)(v), Loss of Safety Function | | 05000528/LER-2003-001, From Palo Verde, Unit 1 Regarding Pressurizer Safety Valve as Found Lift Pressure Outside of TS Limits | From Palo Verde, Unit 1 Regarding Pressurizer Safety Valve as Found Lift Pressure Outside of TS Limits | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | | 05000529/LER-2003-002, Regarding ESF Actuation - Unit 2 Emergency Diesel Generator Actuation | Regarding ESF Actuation - Unit 2 Emergency Diesel Generator Actuation | | | 05000528/LER-2003-002, Manual Reactor Trip Due to Degraded Main Condenser Tube Plug | Manual Reactor Trip Due to Degraded Main Condenser Tube Plug | 10 CFR 50.73(a)(2)(iv)(A), System Actuation 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition 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)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 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)(x) 10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded 10 CFR 50.73(a)(2)(iii) 10 CFR 50.73(a)(2)(i)(A), Completion of TS Shutdown 10 CFR 50.73(a)(2)(ix)(A), Prevented Safety Function in Multiple System 10 CFR 50.73(a)(2)(viii)(A) 10 CFR 50.73(a)(2)(viii)(B) | | 05000530/LER-2003-002, Re RCS Pressure Boundary Leakage Caused by Degraded Inconel Alloy 600 Components | Re RCS Pressure Boundary Leakage Caused by Degraded Inconel Alloy 600 Components | | | 05000529/LER-2003-003, Corrected Copy of Letter for LER 03-003-00 for Palo Verde, Unit 2 | Corrected Copy of Letter for LER 03-003-00 for Palo Verde, Unit 2 | | | 05000528/LER-2003-003, Regarding Technical Specification Violation for Failure to Meet Shutdown Cooling Trains Operable Action Statements | Regarding Technical Specification Violation for Failure to Meet Shutdown Cooling Trains Operable Action Statements | 10 CFR 50.73(a)(2)(iv)(A), System Actuation 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition 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)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(x) 10 CFR 50.73(a)(2)(i)(A), Completion of TS Shutdown 10 CFR 50.73(a)(2)(ix)(A), Prevented Safety Function in Multiple System | | 05000530/LER-2003-003, Regarding Technical Specification Violation for Failure to Calibrate Nuclear Instrumentation | Regarding Technical Specification Violation for Failure to Calibrate Nuclear Instrumentation | 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2)(iii) 10 CFR 50.73(a)(2)(i)(A), Completion of TS Shutdown 10 CFR 50.73(a)(2) 10 CFR 50.73(a)(2)(viii)(B) | | 05000529/LER-2003-004, Mode 3 Entry with an Auxiliary Feed Water Pump Inoperable - Technical Specification Violation | Mode 3 Entry with an Auxiliary Feed Water Pump Inoperable - Technical Specification Violation | 10 CFR 50.73(a)(2) | | 05000528/LER-2003-004-04, Regarding Cracks in Contact Block of Main Control Room Handswitches Result in Inoperable Equipment | Regarding Cracks in Contact Block of Main Control Room Handswitches Result in Inoperable Equipment | 10 CFR 50.73(a)(2)(iv)(A), System Actuation 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition 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)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 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)(x) 10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded 10 CFR 50.73(a)(2)(iii) 10 CFR 50.73(a)(2)(i)(A), Completion of TS Shutdown 10 CFR 50.73(a)(2)(ix)(A), Prevented Safety Function in Multiple System 10 CFR 50.73(a)(2)(viii)(A) 10 CFR 50.73(a)(2)(viii)(B) | | 05000530/LER-2003-004, Reactor Trip with Loss of Forced Circulation Due to an Electrical Grid Disturbance | Reactor Trip with Loss of Forced Circulation Due to an Electrical Grid Disturbance | 10 CFR 50.73(a)(2)(iv)(A), System Actuation 10 CFR 50.73(a)(2)(v), Loss of Safety Function |
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