During the Byron Station, Unit 1, spring 2017, refueling outage, volumetric and surface examinations of the Reactor Vessel Head Penetration (VHP) nozzles identified recordable indications for VHP nozzles 31, 74, 76, and 77 that did not meet the applicable acceptance criteria. The unacceptable indications were identified and repaired prior to returning the reactor head to service. None of the indications were located in the Reactor Coolant System pressure boundary region.
The cause of the P-31 unacceptable indication is attributed to existing welding discontinuities/minor subsurface voids opening to the surface or enlarging due to thermal and/or pressure stresses during plant operation. The cause of the P-74, P-76 and P-77 unacceptable indications is attributed to Primary Water Stress Corrosion Cracking.
The indication in penetration 31 was removed by manual buffing. The indications in P-74, P-76 and P-77 were repaired by manual grinding with no welding required.
This event is reportable in accordance with 10 CFR 50.73(a)(2)(ii)(A) for any event or condition that results in the condition of the nuclear power plant, including its principal safety barriers, being seriously degraded. |
Similar Documents at Byron |
---|
Category:Letter
MONTHYEARIR 05000454/20230042024-02-0202 February 2024 Integrated Inspection Report 05000454/2023004 and 05000455/2023004 ML24022A2722024-01-23023 January 2024 Request for Information for an NRC Post-Approval Site Inspection for License Renewal Inspection Report 05000454/2024011 ML24018A0362024-01-17017 January 2024 Paragon Energy Solutions, Defect with Detroit Diesel/Mtu Fuel Injectors P/N R5229660 Cat Id 0001390618 ML23320A1762023-12-13013 December 2023 Exemption from Select Requirements of 10 CFR Part 73 (EPID L-2023-LLE-0027 (Security Notifications, Reports, and Recordkeeping and Suspicious Activity Reporting)) ML23277A0032023-12-11011 December 2023 Issuance of Amendments Regarding Adoption of TSTF-370 ML23318A5102023-12-0101 December 2023 Relief from the Requirements of the ASME Code IR 05000454/20233012023-11-27027 November 2023 NRC Initial License Examination Report 05000454/2023301 and 05000455/2023301 BYRON 2023-0065, Unit 2 - Notification of Deviation from Electric Power Research Institute (EPRI) Topical Report MRP-227, Revision 1-A, Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guideline2023-11-17017 November 2023 Unit 2 - Notification of Deviation from Electric Power Research Institute (EPRI) Topical Report MRP-227, Revision 1-A, Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guideline BYRON 2023-0063, Day Inservice Inspection Report for Interval 4, Period 3, (B2R24)2023-11-16016 November 2023 Day Inservice Inspection Report for Interval 4, Period 3, (B2R24) IR 05000454/20230032023-11-13013 November 2023 Integrated Inspection Report 05000454/2023003 and 05000455/2023003 and Exercise of Enforcement Discretion RS-23-117, Supplemental Information Letter for Part 73 Exemption Request - Responses to Request for Confirmatory Information2023-11-10010 November 2023 Supplemental Information Letter for Part 73 Exemption Request - Responses to Request for Confirmatory Information ML23317A1192023-11-10010 November 2023 Constellation Energy Generation, LLC - 2023 Annual Report - Guarantees of Payment of Deferred Premiums ML23313A0852023-11-0909 November 2023 Submittal of 2023-301 Byron Initial License Examination Post-Examination Comments RS-23-114, Proposed Alternative for Examination of Pressurizer Circumferential and Longitudinal Shell-to-Head Welds and Nozzle-to-Vessel Welds2023-11-0101 November 2023 Proposed Alternative for Examination of Pressurizer Circumferential and Longitudinal Shell-to-Head Welds and Nozzle-to-Vessel Welds ML23297A2352023-10-26026 October 2023 Information Request for the Cyber Security Baseline Inspection, Notification to Perform Inspection RS-23-100, Request for Exemption from Enhanced Weapons, Firearms Background Checks, and Security Event Notifications Implementation2023-10-13013 October 2023 Request for Exemption from Enhanced Weapons, Firearms Background Checks, and Security Event Notifications Implementation RS-23-097, Constellation Energy Generation, LLC, Advisement of Leadership Changes and Submittal of Updated Standard Practice Procedures Plans2023-10-12012 October 2023 Constellation Energy Generation, LLC, Advisement of Leadership Changes and Submittal of Updated Standard Practice Procedures Plans RS-23-108, Proposed Alternative for Examinations of Examination Categories B-B, B-D, and C-A Steam Generator Pressure Retaining Welds and Full Penetration Welded Nozzles2023-10-11011 October 2023 Proposed Alternative for Examinations of Examination Categories B-B, B-D, and C-A Steam Generator Pressure Retaining Welds and Full Penetration Welded Nozzles RS-23-105, Proposed Alternative for Examinations of Examination Category C-B Steam Generator Nozzle-to-Shell Welds and Nozzle Inside Radius Sections2023-10-10010 October 2023 Proposed Alternative for Examinations of Examination Category C-B Steam Generator Nozzle-to-Shell Welds and Nozzle Inside Radius Sections ML23278A0272023-10-0505 October 2023 Operator Licensing Examination Approval - Byron Station, October 2023 RS-23-093, License Amendment to Braidwood Station, Units 1 and 2, and Byron Station, Units 1 and 2, Technical Specifications 3.7.15, Spent Fuel Pool Boron Concentration, 3.7.16, Spent Fuel.2023-09-29029 September 2023 License Amendment to Braidwood Station, Units 1 and 2, and Byron Station, Units 1 and 2, Technical Specifications 3.7.15, Spent Fuel Pool Boron Concentration, 3.7.16, Spent Fuel. ML23251A1622023-09-29029 September 2023 Steam Generator Tube Inspection Reports to Reflect TSTF-577 Reporting Requirements RS-23-094, Relief Request I4R-24, Alternative for Post-Peening Reexamination Frequency for Reactor Pressure Vessel Head Penetration Nozzle Number 752023-09-29029 September 2023 Relief Request I4R-24, Alternative for Post-Peening Reexamination Frequency for Reactor Pressure Vessel Head Penetration Nozzle Number 75 RS-23-091, Relief Request I4R-25, Alternative Requirements for Reactor Pressure Vessel Inservice Inspection Intervals2023-09-26026 September 2023 Relief Request I4R-25, Alternative Requirements for Reactor Pressure Vessel Inservice Inspection Intervals ML23226A0062023-09-19019 September 2023 Review of License Renewal Commitment Number 10 Submittal ML23180A1692023-09-11011 September 2023 Calvert Cliff Units 1 & 2, and R.E. Ginna Plant - Withdrawal of Proposed Alternatives to American Society of Mechanical Engineers (ASME) Requirements (Epids L-2022-LRR-0074, 0076, 0079, 0091, 0092, 0093 and 0094) ML23242A3282023-09-0101 September 2023 Amendment No. 233 Correction IR 05000454/20235012023-08-31031 August 2023 Emergency Preparedness Biennial Exercise Inspection Report 05000454/2023501 and 05000455/2023501 IR 05000454/20230052023-08-30030 August 2023 Updated Inspection Plan for Byron Station (Report 05000454/2023005 and 05000455/2023005) IR 05000454/20230022023-08-0303 August 2023 Integrated Inspection Report 05000454/2023002 and 05000455/2023002 ML23209A7242023-07-31031 July 2023 Request for Information on the NRC Quadrennial Comprehensive Engineering Team Inspection: Inspection Report 05000454/2024010 and 05000455/2024010 ML23192A0362023-07-25025 July 2023 Review of the Spring 2022 Steam Generator Tube Inspection Report ML23122A3022023-07-20020 July 2023 Issuance of Amendments Technical Specifications 2.1.1 and 4.2.1 to Allow a Previously Irradiated Accident Tolerant Fuel Lead Test Assembly to Be Further Irradiated in Unit No. 2 IR 05000454/20234022023-07-18018 July 2023 Baseline Security Inspection Document; 05000454/2023402; 05000455/2023402 ML23198A0372023-07-17017 July 2023 Information Request to Support Upcoming Problem Identification and Resolution (PIR) Inspection at Byron Nuclear Plant ML23178A2422023-06-28028 June 2023 Reassignment of the U.S. Nuclear Regulatory Commission Branch Chief in the Division of Operating Reactor Licensing for Plant Licensing Branch III RS-23-083, Withdrawal - Proposed Alternatives Related to the Steam Generators2023-06-27027 June 2023 Withdrawal - Proposed Alternatives Related to the Steam Generators ML23172A1172023-06-22022 June 2023 Notification of NRC Fire Protection Team Inspection Request for Information RS-23-077, Response to NRC Regulatory Issue Summary 2023-01, Preparation and Scheduling of Operator Licensing Examinations2023-06-16016 June 2023 Response to NRC Regulatory Issue Summary 2023-01, Preparation and Scheduling of Operator Licensing Examinations ML23159A1302023-06-0808 June 2023 Day Inservice Inspection Report for Interval 4, Period 3, (B1R25) ML23159A1542023-06-0808 June 2023 2022 Regulatory Commitment Change Summary Report RS-23-075, Application for Technical Specification Improvement to Extend the Completion Time for Condition B of Technical Specification 3.5.1, Accumulators, Using the Consolidated Line Item Improvement Process2023-06-0707 June 2023 Application for Technical Specification Improvement to Extend the Completion Time for Condition B of Technical Specification 3.5.1, Accumulators, Using the Consolidated Line Item Improvement Process ML23157A2832023-06-0606 June 2023 Notification of NRC Baseline Inspection and Request for Information BYRON 2023-0029, Response to Request for Additional Information Regarding Steam Generator Tube Inspection Reports to Reflect TSTF-577 Reporting Requirements2023-06-0101 June 2023 Response to Request for Additional Information Regarding Steam Generator Tube Inspection Reports to Reflect TSTF-577 Reporting Requirements ML23138A1342023-05-18018 May 2023 Information Meeting with a Question and Answer Session to Discuss NRC 2022 End-Of-Cycle Plant Performance Assessment of Braidwood Station and Byron Station ML23003A7882023-05-11011 May 2023 Report for December 12-16, 2022, Regulatory Audit Regarding Reinsertion of a High Burnup Accident Tolerant Fuel Lead Test Assembly 2024-02-02
[Table view] Category:Licensee Event Report (LER)
MONTHYEAR05000454/LER-2017-0012017-04-25025 April 2017 1 OF 4, LER 17-001-00 for Byron, Unit 1, Regarding Volumetric and Surface Examinations of Reactor Pressure Vessel Head Penetration Nozzles Identify Indications Attributed to Primary Water Stress Corrosion Cracking and Minor Subsurface Void Enlargement from.. 05000455/LER-2016-0012017-02-15015 February 2017 Manual Reactor Trip due to Circuit Breaker Failure that Caused Actuation of Feedwater Hammer Prevention System with Automatic Isolation of Feedwater to Two Steam Generators and Low Steam Generator Levels, LER 16-001-01 for Byron Station, Unit 2 Regarding Manual Reactor Trip Due to Circuit Breaker Failure that Caused Actuation of Feedwater Hammer Prevention System with Automatic Isolation of Feedwater to Two Steam Generators and Low Steam Generator.... 05000454/LER-2016-0012016-05-0303 May 2016 Auxiliary Feedwater Diesel Intake Design Deficiency Related to Turbine Building High Energy Line Break Resulted in an Unanalyzed Condition Due to Insufficient Validation of Vendor Analysis Inputs, LER 16-001-00 for Byron, Unit 1, Regarding Auxiliary Feedwater Diesel Intake Design Deficiency Related to Turbine Building High Energy Line Break Resulted in an Unanalyzed Condition Due to Insufficient Validation of Vendor Analysis Inputs BYRON 2004-0033, Supplemental One to Licensee Event Report (LER) 454-2003-003-00, Licensed Maximum Power Level Exceeded Due to Inaccuracies in Feedwater Ultrasonic Flow Measurements Caused by Signal Noise Contamination2004-03-31031 March 2004 Supplemental One to Licensee Event Report (LER) 454-2003-003-00, Licensed Maximum Power Level Exceeded Due to Inaccuracies in Feedwater Ultrasonic Flow Measurements Caused by Signal Noise Contamination BYRON 2002-0115, LER 02-S001-00 for Byron Station, Units 1 and 2, Unescorted Access Granted Based on Falsified Information Provided by an Individual2002-10-25025 October 2002 LER 02-S001-00 for Byron Station, Units 1 and 2, Unescorted Access Granted Based on Falsified Information Provided by an Individual 2017-04-25
[Table view] |
comments regarding burden estimate to the Information Services Branch (1--2 F43). U.S. Nuclear Regulatory Commission, Washington, DC 20555.0001 or by e-mail to Infccollects.Resource@nrc.gov and to the Desk Officer, Office of Information and Regulatory Affairs, NEOB-10202. (3150-0104), Office of Management and Budget Washington, DC 20503. If a means used to impose an informaton collection does not dsplay a currently valid OMB control number, the NRC may not conduct or sponsor, and a person is not required to respond to the information collection.
3. LER NUMBER
2017 - 00 001
A. Plant Operating Conditions Before the Event:
Event Dates: March 3, 2017, March 5, 2017, and March 6, 2017 Unit: 1 Mode: 6 (defueled) Reactor Power: 0 percent Unit 1 Reactor Coolant System (RCS) [AB]:
Ambient operating temperature and pressure No structures, systems or components were inoperable at the start of this event that contributed to the event.
B. Description of Event:
During the Byron Station, Unit 1, spring 2017, refueling outage (B1R21), NDE personnel performed volumetric and surface examinations of the Reactor Vessel Head Penetration (VHP) nozzles in accordance ASME Code Case N729- 1, as modified by 10 CFR 50.55a(g)(6)(ii)(D). Examination results identified recordable indications for VHP nozzles 31, 74, 76, and 77 that did not meet the applicable acceptance criteria. Each of the affected VHPs is a Core Exit Thermocouple nozzle. None of the unacceptable indications were located in the Reactor Coolant System pressure boundary region.
March 3, 2017:
VHP 31 nozzle had one unacceptable indication identified by surface (PT) examination on an existing Embedded Flaw repaired weld. The unacceptable indication was 7/32 inch compared to the acceptance criteria of 3/16 inch. The indication was located approximately 3 inches from the interface of the nozzle and J-Groove weld.
March 5, 2017:
VHP 76 nozzle had two unacceptable indications in the tube base material identified by volumetric (UT) examination.
The first unacceptable indication was 0.231 inches long and 0.135 inches deep (22 percent through wall) from the nozzle OD surface. The indication was 1.945 inches to 2.18 inches from the end of the nozzle. The second unacceptable indication was 0.316 inches long and 0.153 inches deep (25 percent through wall) from the nozzle OD surface. The indication was located 1.949 inches to 2.265 inches from the end of the nozzle.
March 6, 2017:
VHP 76 nozzle had two additional unacceptable indications in the tube base material identified by volumetric (UT) examination. The third unacceptable indication was 0.354 inches long and 0.179 inches deep (29 percent through wall) from the nozzle OD surface. The indication was located 1.544 inches to 1.898 inches from the end of the nozzle.
The fourth unacceptable indication was 0.355 inches long and 0.130 inches deep (21 percent through wall) from the nozzle OD surface. The indication was located 1.544 inches to 1.898 inches from the end of the nozzle.
VHP 74 nozzle had two unacceptable indications in the tube base material identified by volumetric (UT) examination.
The first unacceptable indication was 0.394 inches long and 0. 133 inches deep (21 percent through wall) from the nozzle OD surface (through wall dimension is 21 percent through wall). The indication was located 1.822 inches to 2.216 inches from the end of the nozzle. The second unacceptable indication was 0.394 inches long and 0.128 inches deep (20 percent through wall) from the nozzle OD surface. The indication was located 1.507 inches to 1.901 inches from the end of the nozzle.
comments regarding burden estimate to the Information Services Branch (T-2 F43). U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by e-mail to NEOB-10202. (3150-0104). Office of Management and Budget, Washington. DC 20503. If a means used to impose an information collection does not display a currently valid OMB control number. the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.
3. LER NUMBER
2017 - 00 001 VHP 77 nozzle had one unacceptable indication in the tube base material identified by volumetric (UT) examination.
The unacceptable indication was 0.315 inches in length and 0.156 inches deep (25 percent through wall) from the nozzle OD surface. This indication was located 1.823" to 2.138" from the end of the nozzle.
This condition is reportable to the NRC in accordance with 10 CFR 50.73 (a)(2)(ii)(A), as a condition that resulted in a principle safety barrier being seriously degraded.
This LER is being submitted in follow-up to three emergency notification system ENS calls made in accordance with 10 CFR 50.72(b)(3)(ii)(A):
ENS 52591 dated March 3, 2017 at 2052 EST ENS 52592 dated March 5, 2017 at 1605 EST ENS 52592, supplemented March 6, 2017 at 1750 EST.
C. Cause of Event:
The cause of the P-31 recordable indication is attributed to existing welding discontinuities/minor subsurface voids opening to the surface or enlarging due to thermal and/or pressure stresses during plant operation.
The cause of the P-74, P-76 and P-77 recordable indications is attributed to Primary Water Stress Corrosion Cracking (PWSCC).
D. Safety Consequences:
This event is not considered an event or condition that could have prevented fulfillment of a safety function. The indications were identified in a timely manner and repaired prior to through-wall leakage occurring. The indications were identified as part of a required periodic inspection. Potentially, if the indications remained undetected, any one could have propagated over time through the alloy 600 weld material to form a leak path through the reactor coolant boundary.
E. Corrective Actions:
Immediate Actions Completed P-31 - This indication was reduced to an acceptable dimension by manual buffing in accordance with ASME Section Xl.
P-74, P-76 and P-77 - These indications were repaired by grinding with no welding required in Accordance with ASME Section Xl.
Corrective Actions - Longer Term as a mitigating strategy to address the potential for further PWSCC degradation. During B1R21, 100 percent of the VHPs were peened in the inside diameter and the outside diameter.
F. Previous Occurrences:
Byron Station, Unit 2. Licensee Event Report (LER) 455-2007-001-00, "Reactor Pressure Vessel Head Control Rod Drive Mechanism Penetration Nozzle Weld Indication Due to an Initial Construction Weld Defect Allowing the Initiation of Primary Water Stress Corrosion Cracking," (June 8, 2007).
comments regarding burden estimate to the Information Services Branch (T-2 F43), U.S. Nuclear Regulatory Commission. Washington. DC 20555-0001, or by e-mail to Infocollects.Resource@nrc gov, and to the Desk Officer, Office of Information and Regulatory Affairs, used to impose an information collection does not display a currently valid OMB control number, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection
3. LER NUMBER
2017 - 00 001 Byron Station, Unit 1. Licensee Event Report 2011-002-00, "Unit 1 Reactor Pressure Vessel Head Penetration Nozzle Weld Flaws Attributed to Primary Water Stress Corrosion Cracking," (May 18, 2011).
Byron Station, Unit 2. Licensee Event Report 2014-004-00, "Byron Station Unit 2 Reactor Pressure Vessel Head Control Rod Drive Mechanism Penetration Nozzle Weld Indication attributed to Primary Water Stress Corrosion Cracking," (December 5, 2014).
A review of these LERs concluded that these events are similar; however, the causes and corrective actions taken would not have been expected to prevent this event from occurring.
G. Component Failure Data:
Manufacturer Nomenclature Model Mfg. Part Number Westinghouse Reactor Vessel Integrated Head 1718E72 N/A Package Termination