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Category:Letter
MONTHYEAR05000270/LER-2024-001, Common Cause Inoperability of Both Trains of Control Room Ventilation System Outside Air Booster Fans Due to Supply Breaker Wiring Deficiency Resulted in a Condition That Have Prevented Fulfillment2024-12-19019 December 2024 Common Cause Inoperability of Both Trains of Control Room Ventilation System Outside Air Booster Fans Due to Supply Breaker Wiring Deficiency Resulted in a Condition That Have Prevented Fulfillment. 05000269/LER-2024-001, Standby Shutdown Facility (Ssf) Pressurizer Level Switch Configuration Caused by Legacy Procedure Deficiency Resulted in Condition Prohibited by Technical Specifications2024-12-19019 December 2024 Standby Shutdown Facility (Ssf) Pressurizer Level Switch Configuration Caused by Legacy Procedure Deficiency Resulted in Condition Prohibited by Technical Specifications 05000269/LER-2022-001-01, Ultrasonic Examination Indication Identifies Degraded Reactor Coolant System Pressure Boundary2024-11-0707 November 2024 Ultrasonic Examination Indication Identifies Degraded Reactor Coolant System Pressure Boundary ML24305A1492024-11-0404 November 2024 Review of the Fall 2023 Steam Generator Tube Inspection Report (O2R31) IR 05000269/20240032024-10-31031 October 2024 Integrated Inspection Report 05000269/2024003 and 05000270/2024003 and 05000287/2024003 (2) ML24255A3322024-10-16016 October 2024 SLRA - Revised SE Letter ML24297A6172024-10-11011 October 2024 PCA Letter to NRC Oconee Hurricane Helene ML24269A0912024-10-0909 October 2024 Request for Withholding Information from Public Disclosure IR 05000269/20243012024-09-23023 September 2024 NRC Operator License Examination Report 05000269/2024301, 05000270/2024301, and 05000287/2024301 ML24145A1782024-08-26026 August 2024 Issuance of Amendment Nos. 430, 432, and 431, to TS 5.5.2, Containment Leakage Rate Testing Program for a one-time Extension of the Type a Leak Rate Test Frequency IR 05000269/20240052024-08-26026 August 2024 Updated Inspection Plan for Oconee Nuclear Station, Units 1, 2 and 3 (Report 05000269/2024005, 05000270/2024005, and 05000287-2024005) ML24220A1092024-08-0808 August 2024 – Operator Licensing Examination Approval 05000269/2024301, 05000270/2024301, and 05000287/2024301 05000287/LER-2024-001, Procedure Deficiency Results in Inadvertent Automatic Feedwater Isolation and Automatic Emergency Feedwater Actuation2024-08-0202 August 2024 Procedure Deficiency Results in Inadvertent Automatic Feedwater Isolation and Automatic Emergency Feedwater Actuation IR 05000269/20240102024-08-0101 August 2024 Focused Engineering Inspection - Age-Related Degradation Report 05000269/2024010 and 05000270/2024010 and 05000287/2024010 IR 05000269/20240022024-07-25025 July 2024 Integrated Inspection Report 05000269/2024002 and 05000270/2024002 and 05000287/2024002 ML24192A1312024-07-15015 July 2024 Licensed Operator Positive Fitness-For-Duty Test ML24183A0972024-07-12012 July 2024 ISFSI; Catawba 1, 2 & ISFSI; McGuire 1, 2 & ISFSI; Oconee 1, 2, 3 & ISFSI; Shearon Harris 1; H. B. Robinson 2 & ISFSI; and Radioactive Package Shipping Under 10 CFR 71 (71-266 & 71-345) – Review of QA Program Changes EPID L-2024-LLQ-0002 ML24183A2352024-06-29029 June 2024 Update 3 to Interim Report Regarding a Potential Defect with Schneider Electric Medium Voltage Vr Type Circuit Breaker Part Number V5D4133Y000 ML24179A1102024-06-27027 June 2024 Submittal of Updated Final Safety Analysis Report Revision 30, Technical Specifications Bases Revisions, Selected Licensee Commitment Revisions, 10 CFR 50.59 Evaluation Summary Report, and 10 CFR 54.37 Update, and Notification ML24141A0482024-05-17017 May 2024 EN 56958_1 Ametek Solidstate Controls, Inc IR 05000269/20240012024-05-0303 May 2024 Integrated Inspection Report 05000269/2024001, 05000270/2024001 and 05000287/2024001 IR 05000269/20244022024-04-24024 April 2024 Security Baseline Inspection Report 05000269/2024402 and 05000270/2024402 and 05000287/2024402 ML24108A0792024-04-16016 April 2024 EN 57079 Paragon Energy Solutions Email Forwarding Part 21 Interim Report Re Potential Defect with Schneider Electric Medium Voltage Vr Type Circuit Breaker Part Number V5D4133Y000 IR 05000269/20244012024-03-28028 March 2024 – Security Baseline Inspection Report 05000269-2024401 and 05000270-2024401 and 05000287-2024401 ML24088A3052024-03-25025 March 2024 Fws to NRC, Agreement with Nlaa Determination for Tricolored Bat for Oconee Lr 05000287/LER-2023-002, Passive Containment Isolation Device Inoperability Results in Operation or Condition Prohibited by Technical Specifications2024-02-29029 February 2024 Passive Containment Isolation Device Inoperability Results in Operation or Condition Prohibited by Technical Specifications 05000270/LER-2023-001, Inappropriate Procedural Guidance for Planned Online Maintenance Results in Event or Condition That Could Have Prevented Fulfillment of a Safety Function Licensee2024-02-29029 February 2024 Inappropriate Procedural Guidance for Planned Online Maintenance Results in Event or Condition That Could Have Prevented Fulfillment of a Safety Function Licensee IR 05000269/20230062024-02-28028 February 2024 Annual Assessment Letter for Oconee Nuclear Station Units 1, 2 and 3 - (NRC Inspection Report 05000269/2023006, 05000270/2023006, and 05000287/2023006) ML24045A3072024-02-16016 February 2024 Ltr. to Pete Parr Chief Pee Dee Indian Tribe Re Oconee Nuclear Station Units 1, 2, and 3 Section 106 ML24030A0052024-02-16016 February 2024 Ltr. to Brian Harris, Chief, Catawba Indian Nation; Re., Oconee Nuclear Station Units 1, 2, and 3 Section 106 ML24045A2992024-02-16016 February 2024 Ltr. to Eric Pratt Chief the Santee Indian Organization Re Oconee Nuclear Station Units 1, 2, and 3 Section 106 ML24045A3082024-02-16016 February 2024 Ltr. to Ralph Oxendine Chief Sumter Tribe of Cheraw Indians Re Oconee Nuclear Station Units 1, 2, and 3 Section 106 ML24045A3032024-02-16016 February 2024 Ltr. to John Creel Chief Edisto Natchez-Kusso Tribe of Sc Re Oconee Nuclear Station Units 1, 2, and 3 Section 106 ML24045A2952024-02-16016 February 2024 Ltr. to Chuck Hoskin, Jr, Principal Chief Cherokee Nation Re Oconee Nuclear Station Units 1, 2, and 3 Section 106 ML24045A3062024-02-16016 February 2024 Ltr. to Michell Hicks, Principal Chief Eastern Band of Cherokee Re Oconee Nuclear Station Units 1,2, and 3 Section 106 ML24045A3012024-02-16016 February 2024 Ltr. to Harold Hatcher Chief the Waccamaw Indian People Re Oconee Nuclear Station Units 1, 2, and 3 Section 106 ML24045A2972024-02-16016 February 2024 Ltr. to Dexter Sharp Chief Piedmont American Indian Assoc Re Oconee Nuclear Station Units 1, 2, and 3 Section 106 ML24045A3022024-02-16016 February 2024 Ltr. to Joe Bunch United Keetoowah Band of Cherokee Indians in Ok Re Oconee Nuclear Station Units 1, 2, and 3 Section 106 ML24045A2962024-02-16016 February 2024 Ltr. to David Hill Principal Chief Muscogee Creek Nation Re Oconee Nuclear Station Units 1, 2, and 3 Section 106 ML24045A2942024-02-16016 February 2024 Ltr. to Carolyn Chavis Bolton Chief Pee Dee Indian Nation of Upper Sc Re Oconee Nuclear Station Units 1, 2, and 3 Section 106 ML24045A3042024-02-16016 February 2024 Ltr. to Lisa M. Collins Chief the Wassamasaw Tribe of Varnertown Indians Re Oconee Nuclear Station Units 1, 2, and 3 Section 106 ML24045A3052024-02-16016 February 2024 Ltr. to Louis Chavis Chief Beaver Creek Indians Re Oconee Nuclear Station Units 1, 2, and 3 Section 106 ML24011A1482024-02-13013 February 2024 Letter to Steven M. Snider-Oconee Nuclear Sta, Unites 1,2 & 3 Notice of Avail of the Draft Site-Specific Supp. 2, 2nd Renewal to the Generic EIS for Lic. Renew of Nuclear Plants ML24030A5212024-02-13013 February 2024 Letter to Elizabeth Johnson, Director, SHPO; Re Oconee Nuclear Stations Units 1, 2, and 3 Section 106 ML24019A1442024-02-13013 February 2024 Letter to Reid Nelson, Executive Director, Achp; Re Oconee Nuclear Station Units 1, 2, and 3 Section 106 IR 05000269/20230042024-02-13013 February 2024 Integrated Inspection Report 05000269/2023004, 05000270/2023004, and 05000287/2023004; and Inspection Report 07200040/2023001 ML24011A1532024-02-13013 February 2024 Letter to Tracy Watson EPA-Oconee Nuclear Sta, Unites 1, 2 & 3 Notice of Avail of the Draft Site-Specific Supp. 2, 2nd Renewal to the Generic EIS for Lic. Renew of Nuclear Plants ML23304A1422024-02-0101 February 2024 Issuance of Environmental Scoping Summary Report Associated with the U.S. Nuclear Regulatory Commission Staff’S Review of the Oconee Nuclear Station, Units 1, 2, & 3, Subsequent License Renewal Application ML24005A2492024-01-24024 January 2024 Exemption from Select Requirements of 10 CFR Part 73 (Security Notifications, Reports, and Recordkeeping and Suspicious Activity Reporting) ML24017A0582024-01-11011 January 2024 Notification of Licensed Operator Initial Examination 05000269/2024301, 05000270/2024301, and 05000287/2024301 2024-09-23
[Table view] Category:Licensee Event Report (LER)
MONTHYEAR05000270/LER-2024-001, Common Cause Inoperability of Both Trains of Control Room Ventilation System Outside Air Booster Fans Due to Supply Breaker Wiring Deficiency Resulted in a Condition That Have Prevented Fulfillment2024-12-19019 December 2024 Common Cause Inoperability of Both Trains of Control Room Ventilation System Outside Air Booster Fans Due to Supply Breaker Wiring Deficiency Resulted in a Condition That Have Prevented Fulfillment. 05000269/LER-2024-001, Standby Shutdown Facility (Ssf) Pressurizer Level Switch Configuration Caused by Legacy Procedure Deficiency Resulted in Condition Prohibited by Technical Specifications2024-12-19019 December 2024 Standby Shutdown Facility (Ssf) Pressurizer Level Switch Configuration Caused by Legacy Procedure Deficiency Resulted in Condition Prohibited by Technical Specifications 05000269/LER-2022-001-01, Ultrasonic Examination Indication Identifies Degraded Reactor Coolant System Pressure Boundary2024-11-0707 November 2024 Ultrasonic Examination Indication Identifies Degraded Reactor Coolant System Pressure Boundary 05000287/LER-2024-001, Procedure Deficiency Results in Inadvertent Automatic Feedwater Isolation and Automatic Emergency Feedwater Actuation2024-08-0202 August 2024 Procedure Deficiency Results in Inadvertent Automatic Feedwater Isolation and Automatic Emergency Feedwater Actuation 05000270/LER-2023-001, Inappropriate Procedural Guidance for Planned Online Maintenance Results in Event or Condition That Could Have Prevented Fulfillment of a Safety Function Licensee2024-02-29029 February 2024 Inappropriate Procedural Guidance for Planned Online Maintenance Results in Event or Condition That Could Have Prevented Fulfillment of a Safety Function Licensee 05000287/LER-2023-002, Passive Containment Isolation Device Inoperability Results in Operation or Condition Prohibited by Technical Specifications2024-02-29029 February 2024 Passive Containment Isolation Device Inoperability Results in Operation or Condition Prohibited by Technical Specifications 05000287/LER-2023-001, Condition Prohibited by Technical Specifications Due to Isolation Valve Exceeding Inservice Testing Leakage Requirements2023-03-0909 March 2023 Condition Prohibited by Technical Specifications Due to Isolation Valve Exceeding Inservice Testing Leakage Requirements 05000269/LER-2022-002, Reactor Coolant Oconee Nuclear Station, Unit 1, System Pressure Boundary Leak on Reactor Coolant Pump Lower Bearing Thermowell2023-01-12012 January 2023 Reactor Coolant Oconee Nuclear Station, Unit 1, System Pressure Boundary Leak on Reactor Coolant Pump Lower Bearing Thermowell 05000269/LER-2022-001, Ultrasonic Examination Indications Identifies Degraded Reactor Coolant System Pressure Boundary2022-12-21021 December 2022 Ultrasonic Examination Indications Identifies Degraded Reactor Coolant System Pressure Boundary 05000287/LER-2022-002, Automatic Actuation of Emergency Feedwater System Due to Malfunctioning Startup Feedwater Control Valve2022-07-0101 July 2022 Automatic Actuation of Emergency Feedwater System Due to Malfunctioning Startup Feedwater Control Valve 05000270/LER-2022-003, Manual Reactor Trip Due to Main Feedwater Control Valve Positioner Malfunction2022-04-21021 April 2022 Manual Reactor Trip Due to Main Feedwater Control Valve Positioner Malfunction 05000270/LER-2022-002, Regarding Automatic Actuation of Emergency Feedwater System Due to Main Feedwater Pump Malfunction2022-04-14014 April 2022 Regarding Automatic Actuation of Emergency Feedwater System Due to Main Feedwater Pump Malfunction 05000270/LER-2022-001, Regarding Automatic Reactor Trip Due to Loss of Power to Reactor Coolant Pumps2022-04-0606 April 2022 Regarding Automatic Reactor Trip Due to Loss of Power to Reactor Coolant Pumps 05000287/LER-2022-001, Response Actions Resulted in Brief Inoperability of Both Onsite and Offsite Emergency AC Power Paths2022-03-28028 March 2022 Response Actions Resulted in Brief Inoperability of Both Onsite and Offsite Emergency AC Power Paths 05000270/LER-2021-005, Automatic Reactor Trip Due to Spurious Trip Signal Concurrent with System Testing2022-02-0808 February 2022 Automatic Reactor Trip Due to Spurious Trip Signal Concurrent with System Testing 05000270/LER-2021-004, More than One Axial Power Shaping Rod Not Aligned within Technical Specification Limits2022-02-0707 February 2022 More than One Axial Power Shaping Rod Not Aligned within Technical Specification Limits 05000270/LER-2021-003, Conditions Prohibited by Technical Specifications Due to Ssf and Psw Inoperability2022-02-0707 February 2022 Conditions Prohibited by Technical Specifications Due to Ssf and Psw Inoperability 05000270/LER-2021-002, Actuation of the Keowee Hydroelectric Station Due to Loss of AC Power to the Unit 2 Main Feeder Buses2022-01-26026 January 2022 Actuation of the Keowee Hydroelectric Station Due to Loss of AC Power to the Unit 2 Main Feeder Buses 05000270/LER-2021-001, B Motor Driven Emergency Feedwater Pump Past Inoperability Resulted in Condition Prohibited by Technical Specifications2021-10-0404 October 2021 B Motor Driven Emergency Feedwater Pump Past Inoperability Resulted in Condition Prohibited by Technical Specifications 05000287/LER-2020-001, Manual Reactor Trip Due to Reaching Feedwater Heater Level Limit in Operating Procedure2020-06-0909 June 2020 Manual Reactor Trip Due to Reaching Feedwater Heater Level Limit in Operating Procedure RA-19-0376, Core Flooding System Loss of Safety Function2019-09-19019 September 2019 Core Flooding System Loss of Safety Function 05000269/LER-2019-001, Standby Shutdown Facility Reactor Coolant Makeup Pump Oil Suction Tubing Failure2019-08-13013 August 2019 Standby Shutdown Facility Reactor Coolant Makeup Pump Oil Suction Tubing Failure 05000269/LER-2018-002, Loss. of Condenser Vacuum Results in a Main Turbine Trip and a Manual Reactor Trip2018-12-18018 December 2018 Loss. of Condenser Vacuum Results in a Main Turbine Trip and a Manual Reactor Trip 05000287/LER-2018-002-01, Actuation of the Keowee Hydroelectric Station Due to Loss of AC Power to the Unit 3 Main Feeder Buses2018-10-25025 October 2018 Actuation of the Keowee Hydroelectric Station Due to Loss of AC Power to the Unit 3 Main Feeder Buses 05000287/LER-2018-002, Actuation of the Keowee Hydroelectric Station Due to Loss of AC Power to the Unit 3 Main Feeder Buses2018-07-0505 July 2018 Actuation of the Keowee Hydroelectric Station Due to Loss of AC Power to the Unit 3 Main Feeder Buses 05000269/LER-2018-001, Manual Reactor Trip Due to Main Feedwater Flow Control Valve E/P Converter Failures2018-06-0707 June 2018 Manual Reactor Trip Due to Main Feedwater Flow Control Valve E/P Converter Failures 05000287/LER-1917-001, Regarding Reactor Protection System Actuation - Reactor Trip Due to Turbine Trip from Generator Lockout2017-09-20020 September 2017 Regarding Reactor Protection System Actuation - Reactor Trip Due to Turbine Trip from Generator Lockout 05000269/LER-1917-001, Regarding Loss of Both Keowee Hydroelectric Units Due to Human Error2017-08-0909 August 2017 Regarding Loss of Both Keowee Hydroelectric Units Due to Human Error 05000269/LER-2016-003, Regarding Engineered Safeguards Protection System Automatic Actuation Output Logic Bypassed2017-01-24024 January 2017 Regarding Engineered Safeguards Protection System Automatic Actuation Output Logic Bypassed 05000269/LER-2016-002, Regarding Containment High Range Radiation Monitors Inoperable Due to Potential Thermally Induced Current Effects2016-12-16016 December 2016 Regarding Containment High Range Radiation Monitors Inoperable Due to Potential Thermally Induced Current Effects 05000287/LER-2016-001, Regarding Reactor Building Cooling Unit Inoperability Exceeds Technical Specification Completion Time2016-08-26026 August 2016 Regarding Reactor Building Cooling Unit Inoperability Exceeds Technical Specification Completion Time 05000269/LER-2016-001, Regarding RPS Actuation - Unit 1 Reactor Trip Initiated by a Generator Lockout/Turbine Trip2016-05-0505 May 2016 Regarding RPS Actuation - Unit 1 Reactor Trip Initiated by a Generator Lockout/Turbine Trip 05000270/LER-2015-001, Regarding Valid Emergency Feedwater System Actuation Caused by a Main Feedwater System Block Valve Malfunction2015-09-25025 September 2015 Regarding Valid Emergency Feedwater System Actuation Caused by a Main Feedwater System Block Valve Malfunction 05000287/LER-2015-001, Regarding Manual Reactor Trip Due to Unacceptable Main Feedwater Flow Control Valve Oscillations2015-03-31031 March 2015 Regarding Manual Reactor Trip Due to Unacceptable Main Feedwater Flow Control Valve Oscillations 05000269/LER-2014-002, Regarding Deficiency in Loss of Coolant Accident Analysis Adversely Affected Predicted Peak Cladding Temperature2015-01-26026 January 2015 Regarding Deficiency in Loss of Coolant Accident Analysis Adversely Affected Predicted Peak Cladding Temperature 05000269/LER-2014-001, Regarding Unanalyzed Condition Associated with the 13.8kV Emergency Power Cables Located in the Underground Trench2014-05-27027 May 2014 Regarding Unanalyzed Condition Associated with the 13.8kV Emergency Power Cables Located in the Underground Trench 05000269/LER-2013-004, Regarding High Cycle Fatigue Resulted in Reactor Coolant Leak and Unit Shutdown2014-01-10010 January 2014 Regarding High Cycle Fatigue Resulted in Reactor Coolant Leak and Unit Shutdown 05000287/LER-2013-001, Re Manual Reactor Trip Due to Main Feedwater Flow Oscillations2013-12-19019 December 2013 Re Manual Reactor Trip Due to Main Feedwater Flow Oscillations 05000269/LER-2013-003, Regarding Keowee Hydroelectric Station Unit 2 - Emergency Power Lockout2013-11-0404 November 2013 Regarding Keowee Hydroelectric Station Unit 2 - Emergency Power Lockout 05000269/LER-2013-002, LPI and RBS Trains Inoperable When 1LP-21 Was Closed Due to Human Error2013-08-26026 August 2013 LPI and RBS Trains Inoperable When 1LP-21 Was Closed Due to Human Error 05000269/LER-2011-004-01, Regarding Inability to Detect RCS Leak Rate Using the Particulate Radiation Monitor2013-05-22022 May 2013 Regarding Inability to Detect RCS Leak Rate Using the Particulate Radiation Monitor 05000269/LER-2013-001, Regarding Inadequate HVAC Load Analysis and Design Impacts on Emergency Power Equipment2013-04-0808 April 2013 Regarding Inadequate HVAC Load Analysis and Design Impacts on Emergency Power Equipment 05000287/LER-2012-001, For Oconee, Unit 3, Regarding Three Main Steam Relief Valves (MSRV) Lift Pressure Exceeds + 1% Tolerance2012-06-0707 June 2012 For Oconee, Unit 3, Regarding Three Main Steam Relief Valves (MSRV) Lift Pressure Exceeds + 1% Tolerance 05000269/LER-2012-001, Regarding Unanalyzed Conditions Exist for Standby Shutdown Facility Mitigated Events2012-06-0404 June 2012 Regarding Unanalyzed Conditions Exist for Standby Shutdown Facility Mitigated Events 05000269/LER-2011-006-01, Regarding Pressurizer Heater Capacity Non-Compliant with Technical Specification 3.4.92012-05-29029 May 2012 Regarding Pressurizer Heater Capacity Non-Compliant with Technical Specification 3.4.9 05000269/LER-2011-007-01, Regarding Inoperable Containment Isolation Valve2012-05-14014 May 2012 Regarding Inoperable Containment Isolation Valve 05000269/LER-2011-002-01, Regarding Tech Spec Required Shutdown for an Inoperable Containment Isolation Valve2012-05-14014 May 2012 Regarding Tech Spec Required Shutdown for an Inoperable Containment Isolation Valve 05000269/LER-2011-005-01, For Oconee Nuclear Station, Unit 1, Reactor Protection System Overpower Flux/Flow/Imbalance Channels Inoperable2012-02-23023 February 2012 For Oconee Nuclear Station, Unit 1, Reactor Protection System Overpower Flux/Flow/Imbalance Channels Inoperable ML11364A0462011-12-23023 December 2011 Special Report Per Technical Specification 5.6.6, Problem Investigation Process Nos.: 0-11-13855, 0-11-14092 05000269/LER-2011-003-01, 1 for Oconee, Units 1, 2, and 3 Regarding Inoperability of the Standby Shutdown Facility Diesel Generator2011-12-19019 December 2011 1 for Oconee, Units 1, 2, and 3 Regarding Inoperability of the Standby Shutdown Facility Diesel Generator 2024-08-02
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ENERGY.
Scott L Batson Vice President Oconee Nuclear Station Duke Energy ONOIVP 1 7800 Rochester Hwy Seneca, SC 29672 o: 864.873.3274 f 864.873.4208 Scott.Batson@duke-energy.com 10 CFR 50.73 ONS-2014-001 January 10, 2014 Attn: Document Control Desk U. S. Nuclear Regulatory Commission 11555 Rockville Pike Rockville, MD 20852-2746 Subject: Duke Energy Carolinas LLC (Duke Energy)
Oconee Nuclear Station (ONS) Unit 1 Docket No. 50-269 Licensee Event Report 269/2013-04, Revision 0 Problem Investigation Program No.: 0-13-13168 Enclosed is Licensee Event Report 269/2013-04, Revision 0 for Oconee Nuclear Station, Unit 1. This report is being submitted in accordance with 10 CFR 50.73 (a)(2)(i)(A),
completion of a nuclear plant shutdown required by Technical Specifications, 10 CFR 50.73 (a)(2)(i)(B), operation or condition prohibited by Technical Specifications, and 10 CFR 50.73 (a)(2)(ii)(A), degradation of a principal safety barrier.
There are no regulatory commitments contained in this report. Any questions regarding the content of this report should be directed to Sandra N. Severance, Regulatory Affairs, at (864) 873-3466.
Sincerely, Scott L. Batson Vice President Oconee Nuclear Site Enclosure www.duke-energy.com
NRC Document Control Desk January 10, 2014 Page 2 cc: Mr. Victor McCree Administrator, Region II U.S. Nuclear Regulatory Commission Marquis One Tower 245 Peachtree Center Ave., NE, Suite 1200 Atlanta, GA 30303-1257 Mr. Richard Guzman, Senior Project Manager (by electronic mail only)
Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission 11555 Rockville Pike Mail Stop O-8C2 Rockville, MD 20852 Mr. Eddy Crowe NRC Senior Resident Inspector Oconee Nuclear Station www.duke-energy.com
Abstract
On November 8, 2013, the Oconee Unit 1 Control Room received an alarm associated with the containment atmosphere particulate radiation monitor. Reactor Coolant System (RCS) leakage of
<0.1 gpm was identified. On November 11, 2013, upon verification of un-isolable reactor coolant system pressure boundary leakage on the 1B2 High Pressure Injection (HPI) Injection line, Oconee Unit 1 was shut down as required by Technical Specifications. The shutdown was orderly and without complication. The cause evaluation determined that mechanical, high-cycle fatigue resulted in a through wall crack in the stainless steel butt weld between the HPI nozzle safe end and HPI piping. Inadequate procedural guidance existed for the conduct of Augmented Examinations and appropriate disposition of Ultrasonic Testing (UT) examination results where conditions limited the weld volume that could be examined.
This event is reportable under 10 CFR 50.73(a)(2)(i)(A), as completion of a shutdown required by Technical Specifications, 10 CFR 50.73(a)(2)(i)(B), operation or condition prohibited by Technical Specifications, and 10 CFR 50.73(a)(2)(ii)(A), degradation of a principal safety barrier. High pressure injection capability was maintained, and containment integrity was not impacted.
NRC FORM 366 (10-2010)
EVALUATION:
BACKGROUND At the time elevated Reactor Coolant System (RCS) [EIIS:AB] leakage was identified, Oconee Nuclear Station (ONS) Unit 1 was operating in Mode 1 at approximately 100 percent power. No significant structures, systems or components were out of service at the time of this event that contributed to this event. Unit 1 [EIIS:NH] experienced reactor coolant pressure boundary leakage in the High Pressure Injection (HPI) System [EIIS:BG] that required a plant shutdown. Technical Specification (TS) 3.4.13, RCS Operational Leakage, Required Actions B.1 and B.2 require that the unit be in Mode 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in Mode 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> of identification of pressure boundary leakage.
During normal operation, the HPI System controls the RCS inventory, provides the seal water for the Reactor Coolant Pumps (RCP) [EIIS:P], and recirculates RCS letdown for water quality maintenance and reactor coolant boric acid concentration control. The discharge of the HPI pumps connects to a nozzle on each of the four reactor inlet pipes downstream of the reactor coolant pumps. The reactor coolant which is letdown is normally returned to the RCS through two of these nozzles (1A1 and 1A2).
During emergency operation, the HPI System supplies borated water from the Borated Water Storage Tank (BWST) to the RCS and the RCP seals. Three parallel HPI pumps have the capability to take suction from the BWST and discharge through two redundant flow headers into the RCS, utilizing four injection lines (two per header). The stainless steel HPI injection lines terminate at injection nozzle [EIIS:NZL] assemblies located on each of the reactor inlet pipes downstream of the RCPs. Each nozzle assembly consists of a carbon steel nozzle (stainless steel clad on the inside),
to which a stainless steel safe end is welded. The HPI piping is welded to the other end of the safe end. Inside the safe end is a stainless steel thermal sleeve, which extends into the main RCS flow path.
EVENT DESCRIPTION
On November 8, 2013, at approximately 1837 hours0.0213 days <br />0.51 hours <br />0.00304 weeks <br />6.989785e-4 months <br />, while Oconee Unit 1 was operating at approximately 100% Full Power (FP), the Control Room received an alarm associated with the containment atmosphere particulate radiation monitor (i.e., RIA-47 [EIIS:IL]) used for RCS leakage detection. At 2324 hours0.0269 days <br />0.646 hours <br />0.00384 weeks <br />8.84282e-4 months <br />, an RCS Leakage Calculation was performed and unidentified leakage of 0.020 gpm was noted, consistent with leakage calculation values obtained within the prior week.
Reactor building particulate sample results indicated no detected activity, and the radiation monitor counts stabilized.
On November 9, 2013, at 0323 hours0.00374 days <br />0.0897 hours <br />5.340608e-4 weeks <br />1.229015e-4 months <br />, another RIA-47 alarm was received. No immediate signs of RCS leakage were observed. At 0614 hours0.00711 days <br />0.171 hours <br />0.00102 weeks <br />2.33627e-4 months <br />, a second RCS Leakage Calculation was performed, and unidentified leakage was found to be 0.088 gpm. This value exceeded the baseline mean by three standard deviations, indicating the leakage results were valid and action was warranted. Based on the leakage calculation results, increased activity on the radiation monitor, and an observable increase in reactor building normal sump rates, a reactor building entry was conducted to identify the leakage
source. Although the general location of the leak was identified, the source could not be determined.
On November 10, 2013, at 1630 hours0.0189 days <br />0.453 hours <br />0.0027 weeks <br />6.20215e-4 months <br />, a second reactor building entry was conducted and video images and robotically obtained leakage samples were evaluated. Absent conclusive evidence of the leak source, the Shift Manager initiated a power reduction to allow for a direct visual inspection.
A normal downpower to 20% was initiated on November 10, 2013, at 2141 hours0.0248 days <br />0.595 hours <br />0.00354 weeks <br />8.146505e-4 months <br />. On November 11, 2013, at 0520 hours0.00602 days <br />0.144 hours <br />8.597884e-4 weeks <br />1.9786e-4 months <br />, the approximately 0.1 gpm leakage was visually determined to be un-isolable RCS pressure boundary leakage from the 1B2 HPI Injection Line. TS 3.4.13, RCS Operational Leakage, Conditions B.1 and B.2 were entered. At 0848 hours0.00981 days <br />0.236 hours <br />0.0014 weeks <br />3.22664e-4 months <br />, an Emergency Notification System call to the NRC was made reporting the degraded condition and the TS required shutdown. On November 12, 2013, at 0244, Unit 1 entered Mode 5 and exited the TS Mode of Applicability.
Upon confirmation that the unidentified leakage being investigated was RCS pressure boundary leakage, the pressure boundary leakage had existed for a time period greater than the Technical Specification allowed COMPLETION TIME.
CAUSAL FACTORS A Failure Investigation Team was created to develop a repair plan, and a Root Cause team was created to investigate the causal factors for this event. Although the Root Cause Evaluation is still being finalized, two Root Causes for this event have been identified. If the final root cause conclusions result in substantive changes to the results presented below, Duke Energy will submit a supplement to this LER.
RC-1: The un-isolable reactor coolant system leak on the 1 B2 HPI line was caused by mechanical, high-cycle fatigue which resulted in a through wall crack in the butt weld between the safe end and HPI piping.
The crack initiated at two separate locations at the weld root due to an unspecified high vibration event, likely associated with the 2008 1 B2 reactor coolant pump seal failure.
These cracks then merged into a single crack and continued to propagate through wall in the intervening years. Crack propagation is attributed to HPI Full Flow testing in subsequent refueling outages (i.e., for test purposes, flow is directed through a single header, rather than splitting into multiple headers) until normal reactor coolant system operating conditions grew the crack thru-wall. For each unit, HPI line vibration is greatest on the B2 line and is bounded by the 1 B2 HPI line. The 1 B2 HPI line is dedicated for emergency injection only, and it is generally stagnant during normal plant operation.
RC-2: Inadequate procedural guidance existed for the conduct of Augmented Examinations and appropriate disposition of Ultrasonic Testing (UT) examination results where conditions limited the weld volume that could be examined.
A UT limitation is defined as any obstruction or condition that limits the extent of angle beam scanning or limits the extent of required coverage using straight beam scanning. When adequate weld volumes could not be examined on the 1 B2 HPI Nozzle safe end-to-pipe butt weld, no procedural guidance provided weld volume acceptance criteria or directed these limitations to be entered into the corrective action program for evaluation. During the root
cause investigation, metallurgical analysis documented that the crack propagated over several operating cycles. Historical Non-Destructive Examination (NDE) data revealed that the crack was visible in existing radiographs. Had the failed weld volume been adequately interrogated, the crack would have been identified before propagating through wall.
CORRECTIVE ACTIONS
Numerous actions were taken to investigate, determine extent of condition, and repair the HPI line.
Those most relevant are included below.
Immediate:
- 1. Inspected 1 B2 injection line pipe and pipe support configuration to verify conformance with design. The piping and supports were installed per design requirements and tolerances. There were no indications of damage or pipe rubbing due to vibration.
- 2. Performed Radiography Testing (RT) of 1 B2 High Pressure Injection (HPI) nozzle and associated original Babcock & Wilcox 2-ply thermal sleeve assembly. RT images were compared to previous, historical RTs of the thermal sleeve and no visible cracks were found on the thermal sleeve. In addition, the position of the thermal sleeve and visible gaps within the expansion area remained unchanged.
- 3. Removed 1 B2 injection line safe end-to-pipe butt weld and adjacent piping for metallurgical analysis. Metallurgical analysis identified high cycle mechanical fatigue (vibration) as the most likely failure mechanism.
- 4. Performed surface conditioning to achieve maximum coverage of weld volume to be inspected and Non-Destructive Examination (NDE) Evaluations on 1A1, 1A2, and 1B1 HPI nozzle safe end-to-piping stainless welds, including penetrant testing (PT), conventional UT, and phased array encoded UT. UT, PT, and phased array encoded UT results were acceptable for the 1A1, 1A2, and 11B1 HPI nozzle-to-safe end welds.
- 5. As part of an extent of condition review, performed surface conditioning to achieve maximum coverage of weld volume to be inspected and NDE Evaluations on 2A1, 2A2, 2B1, and 2B2 HPI nozzle safe end-to-piping stainless welds, including PT, conventional UT, and phased array encoded UT during the scheduled refueling outage. UT, PT, and phased array encoded UT results were acceptable for each of the Unit 2 HPI nozzle-to-safe end welds.
- 6. Performed dimensional checks of Unit 1 and Unit 2 HPI nozzle safe end and stainless steel pipe to determine stress and fatigue sensitivity to any observed discrepancies from design conditions. Of these eight lines, 1 B2 safe end-to-pipe dimensions had the largest safe end variation and highest offset angle. The worst case geometry was analyzed and found to meet Code requirements.
- 7. Replaced 1 B2 HPI safe end-to-pipe butt weld and adjacent piping. Replacement allowed for correction of the offset angle that was identified with the previous weld.
- 8. Collected vibration data on the 1 B2 line during HPI Full Flow testing. Also collected vibration data on all four Unit 2 HPI lines during HPI Full Flow test during the refueling outage. Based on this review and historical data, the vibration levels observed for the HPI Lines during the Full Flow test are largest for the B2 lines on all three units, with the 1 B2 line having the highest reported vibration levels of all 12 HPI Lines. The current vibration levels on the 1 B2 line are within the design acceptance criteria of the piping.
- 9. Reviewed the population of Oconee Unit 3 welds that are inspected by the same NDE UT procedure as the 1 B2 HPI safe end-to-pipe weld and determined they were adequately examined in 2010 to detect the presence of cracking. This includes each Unit 3 HPI nozzle safe end-to-pipe weld. The only UT inspections with coverage limitations (i.e., valve-to-valve butt welds) were further dispositioned by RTs at that time. The Unit 3 HPI nozzle thermal sleeve RTs from the spring of 2012, taken to assess thermal sleeve tightness and position, were also reviewed. Where visible in the RTs, the safe end-to-pipe weld showed no crack-like indications.
Based on the reviews performed and the current negligible Unit 3 RCS unidentified leakage, there are no concerns with the operation of Unit 3.
Planned:
- 1. Modify the HPI system to increase 1 B2 Emergency Injection line's resistance to piping vibration.
- 2. Revise NDE procedures to provide prescriptive guidance for maximizing examination coverage when performing augmented examinations, including entry into the corrective action program for evaluation by functional owner when limitations or indications of degradation are detected.
- 3. Revise the Section XI Functional Area Manual to require augmented examination owners to document evaluation of augmented NDE results, including evaluation of exam limitations, and to take appropriate actions commensurate with risk associated with the NDE results.
SAFETY ANALYSIS
While at 100% power on November 8, 2013, RCS leakage of <0.1 gpm was detected on ONS Unit 1. The leak was later found to be from a circumferential crack located at the safe end-to-pipe butt weld (1-RC-201-105) located between the 182 HPI injection nozzle and valve 1HP-152. The total circumferential extent of the aggregate crack was about 1.2 inches on the pipe inside diameter and about 0.1 inch (-1/8") on the outside surface.
The 1 B2 HPI line is dedicated for emergency injection only, and it is generally stagnant during normal plant operation. The leak in this line remained small, and an orderly shutdown was performed. The leak was much less than what is considered in the Probabilistic Risk Analysis (PRA). However, an un-isolable leak in the reactor coolant system (RCS) pressure boundary
constitutes degradation of a principal safety barrier and is reportable to the NRC. The leak was entirely within the reactor building containment and no radioactive releases were made.
Based on the metallurgical evidence and comparison of this approximately 0.1 gpm leak to the 1997 ONS Operating Experience associated with a similarly located and larger HPI safe end-to-pipe weld leak, the leak before break capacity of this material was demonstrated. There is reasonable assurance that the line would not have catastrophically broken, even during a design basis event, based on a comparison of the materials (same), loading (similar) and flaw extent (smaller) for these two leaks. Additionally, had the leak location failed catastrophically, the 2 1/2 inch pipe break (approximately 0.025 square feet), would have constituted a small break Loss Of Coolant Accident. Breaks at this location are bounded by analyses in the Oconee UFSAR which concludes that this break can be handled without core damage.
ADDITIONAL INFORMATION
A review of the ONS corrective action program data was conducted to include all Root Causes for the last five years with similar event and cause codes as well as appropriate key word searches for this event. Selected apparent cause evaluations were reviewed as well. Two or more of the same events that involved the same equipment, same administrative controls or the same personnel actions were not discovered during the review of ONS corrective action program data; therefore, a similar/recurring event does not exist.
Energy Industry Identification System (EIIS) codes are identified in the text as [XX]. This event is considered INPO Consolidated Events System (ICES) Reportable. There were no releases of radioactive materials, radiation exposures or personnel injuries associated with this event.
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05000287/LER-2013-001, Re Manual Reactor Trip Due to Main Feedwater Flow Oscillations | Re Manual Reactor Trip Due to Main Feedwater Flow Oscillations | 10 CFR 50.73(a)(2)(iv)(A), System Actuation | 05000269/LER-2013-001, Regarding Inadequate HVAC Load Analysis and Design Impacts on Emergency Power Equipment | Regarding Inadequate HVAC Load Analysis and Design Impacts on Emergency Power Equipment | 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)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000269/LER-2013-002, LPI and RBS Trains Inoperable When 1LP-21 Was Closed Due to Human Error | LPI and RBS Trains Inoperable When 1LP-21 Was Closed Due to Human Error | 10 CFR 50.73(a)(2)(v), Loss of Safety Function | 05000269/LER-2013-003, Regarding Keowee Hydroelectric Station Unit 2 - Emergency Power Lockout | Regarding Keowee Hydroelectric Station Unit 2 - Emergency Power Lockout | 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) | 05000269/LER-2013-004, Regarding High Cycle Fatigue Resulted in Reactor Coolant Leak and Unit Shutdown | Regarding High Cycle Fatigue Resulted in Reactor Coolant Leak and Unit Shutdown | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded 10 CFR 50.73(a)(2)(i)(A), Completion of TS Shutdown |
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