LER-2004-007, Regarding Exceeding the Maximum Power Level Specified in Operating License Condition 2.C(1) |
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| 5282004007R00 - NRC Website |
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V LAFS A sthsidiar ofqfPinnarle ll.vt Capial Corporation Palo Verde Operating Licenses Palo Verde Nuclear Generating Station David M. Smith Plant Manager Nuclear Production Tel.
623-393-6116 Mail Station 7602 Fax.
623-393-6077 RO. Box 52034 e-mail: DSMITH10@apsc.com Phoenix, AZ 85072-2034 192-01151 -DMS/SAB/DJS August 13, 2004 ATTN: Document Control Desk U. S. Nuclear Regulatory Commission Washington, DC 20555-0001
Dear Sirs:
Subject:
Palo Verde Nuclear Generating Station (PVNGS)
Units 1, 2, and 3 Docket No. STN 50-528, 50-529 and 50-530 License No. NPF-41, NPF-51 and NPF-74 Licensee Event Report 1-2004-007-00 Attached please find Licensee Event Report (LER) 50-528/2004-007-00 that has been prepared and submitted pursuant to the Palo Verde Operating Licenses (OLs). This LER reports findings related to exceeding the Maximum Power Level for Units 1, 2 and 3 by approximately 1 percent.
In accordance with 1 OCFR50.4, a copy of this LER is being forwarded to the NRC Region IV Office and the Senior 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. The corrective actions described in this LER are not necessary to maintain compliance with regulations.
Sincerely, DMS/SAB/DJS/kg Attachment cc:
B. S. Mallett M. B. Fields N. L. Salgado NRC Region IV Regional Administrator NRC NRR Project Manager NRC Senior Resident Inspector for PVNGS I---
APPROVED BY OMB NO. 3150-0104 EXPIRES 7-31-2004
Abstract
On July 14, 2004, during a review of historical operating data, Engineering concluded that the maximum Palo Verde specific calorimetric error was approximately 38.76 MW in Units 1 and 3 and 39.90 MW in Unit 2 or approximately 1 per cent (1 %). The error resulted in core power levels above the Operating License limits of 3876 MW thermal for Units 1 and 3 and 3990 MW thermal in Unit 2 while the ultrasonic flow measurement (UFM) instrument was in service. The non-conservative feedwater flow input to the secondary calorimetric calculation had been in place in Units 1 and 3 since 1999 and in Unit 2 since 2000. This review and its results were based on a letter received from the vendor (Caldon) stating there maybe an error in the flow measurement.
The cause of the event was an error in the velocity profile factor used by the UFM.
The corrective action for this event was the removal of the UFMs from service.
In the past three years, no previous similar events were reported.
NRC FORM 366 (7.2001)
(f 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 (if more space Is required, use additional copies of NRC Form 366A)
New software from the vendor, Caldon, that includes newly developed velocity profile factors (the old profile factor was non-conservative) and the associated uncertainties analysis, is being reviewed and a plan that includes more intense validation and monitoring is being developed for placing the UFMs back into service.
No date for placing the UFMs back into service is available at this time.
8. PREVIOUS SIMILAR EVENTS
In the past three years, no previous similar events were reported.
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| 05000529/LER-2004-001, Regarding Steam Generator Tube Leak | Regarding Steam Generator Tube Leak | 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)(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 | | 05000530/LER-2004-001, Re RCS Pressure Boundary Leakage Caused by Degraded Alloy 600 Component | Re RCS Pressure Boundary Leakage Caused by Degraded Alloy 600 Component | | | 05000528/LER-2004-001, Regarding Reactor Shutdown Due to Reactor Coolant System Pressure Boundary Leakage | Regarding Reactor Shutdown Due to Reactor Coolant System Pressure Boundary Leakage | | | 05000530/LER-2004-002, Regarding an Automatic Reactor Trip on Low DNBR Following a Main Turbine Control System Malfunction | Regarding an Automatic Reactor Trip on Low DNBR Following a Main Turbine Control System Malfunction | | | 05000530/LER-2004-002-01, 1 for Palo Verde, Unit 3 Regarding Main Turbine Control System Malfunction Results in Automatic Reactor Trip on Low DNBR | 1 for Palo Verde, Unit 3 Regarding Main Turbine Control System Malfunction Results in Automatic Reactor Trip on Low DNBR | 10 CFR 50.73(a)(2)(x)
10 CFR 50.73(a)(2)(ix)(A), Prevented Safety Function in Multiple System
10 CFR 50.73(a)(2) | | 05000529/LER-2004-002, On 07/14/2004 Reactor Tripped on Low Departure from Nucleate Boiling Ratio (DNBR) | On 07/14/2004 Reactor Tripped on Low Departure from Nucleate Boiling Ratio (DNBR) | 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)(vi)(A)
10 CFR 50.73(a)(2)(x)
10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded
10 CFR 50.73(a)(2)(i)(A), Completion of TS Shutdown
10 CFR 50.72(b)(3)(ii), Degraded or Unanalyzed Condition
10 CFR 50.73(a)(2)(viii)(A)
10 CFR 50.73(a)(2)(viii)(B) | | 05000528/LER-2004-002, Technical Specification Violation - Exceeded 20% RTP with LCO Not Met | Technical Specification Violation - Exceeded 20% RTP with LCO Not Met | | | 05000529/LER-2004-003, Regarding Actuation of Plant Emergency Diesel Generators | Regarding Actuation of Plant Emergency Diesel Generators | 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)
10 CFR 50.73(a)(2)(viii)(A)
10 CFR 50.73(a)(2)(viii)(B) | | 05000528/LER-2004-003, Unit I, Regarding Manual Reactor Trip - Slipped Control Element Assembly | Unit I, Regarding Manual Reactor Trip - Slipped Control Element Assembly | 10 CFR 50.73(a)(2)(iv)(A), System Actuation | | 05000528/LER-2004-004, Re Missed Surveillance Requirement for Temperature Detector Calibration | Re Missed Surveillance Requirement for Temperature Detector Calibration | | | 05000528/LER-2004-005, Re Missed Surveillance Tests on Shutdown Cooling Valve RCS Pressure Interlocks | Re Missed Surveillance Tests on Shutdown Cooling Valve RCS Pressure Interlocks | 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)(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)
10 CFR 50.73(a)(2)(viii)(A) | | 05000528/LER-2004-006, Regarding a Loss of Offsite Power (LOOP) and the Subsequent Reactor Trip | Regarding a Loss of Offsite Power (LOOP) and the Subsequent Reactor Trip | | | 05000528/LER-2004-007, Regarding Exceeding the Maximum Power Level Specified in Operating License Condition 2.C(1) | Regarding Exceeding the Maximum Power Level Specified in Operating License Condition 2.C(1) | | | 05000528/LER-2004-008, Regarding Improper Contact Configuration on Containment Isolation Valve | Regarding Improper Contact Configuration on Containment Isolation Valve | | | 05000528/LER-2004-009, Supplemental Report to Licensee Event Report 2004-009-00 | Supplemental Report to Licensee Event Report 2004-009-00 | | | 05000528/LER-2004-011, Re Missed Surveillance Requirements for Containment Lvs Test/Drain Valves | Re Missed Surveillance Requirements for Containment Lvs Test/Drain Valves | 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)(ix)(A), Prevented Safety Function in Multiple System
10 CFR 50.73(a)(2)
10 CFR 50.73(a)(2)(viii)(A)
10 CFR 50.73(a)(2)(viii)(B) |
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