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Category:Letter
MONTHYEARL-2024-158, Supplemental Information to License Amendment Request to Adopt Common Emergency Plan with Site-Specific Annexes2024-09-25025 September 2024 Supplemental Information to License Amendment Request to Adopt Common Emergency Plan with Site-Specific Annexes IR 05000266/20240112024-09-18018 September 2024 Biennial Problem Identification and Resolution Inspection Report 05000266/2024011 and 05000301/2024011 L-2024-136, Supplement to License Amendment Request to Adopt Common Emergency Plan with Site-Specific Annexes2024-09-16016 September 2024 Supplement to License Amendment Request to Adopt Common Emergency Plan with Site-Specific Annexes L-2024-145, Notification of Deviation from EPRI MRP-227, Revision 1-A Baffle Bolt Inspection Frequency2024-09-0909 September 2024 Notification of Deviation from EPRI MRP-227, Revision 1-A Baffle Bolt Inspection Frequency ML24207A0202024-08-28028 August 2024 Response to Request for Re-Engagement Regarding the Subsequent License Renewal Environmental Review for Point Beach Nuclear Plant, Units 1 and 2 (Docket Numbers: 50-0266 and 50-0301) IR 05000266/20240052024-08-21021 August 2024 Updated Inspection Plan for Point Beach Nuclear Plant, Units 1 and 2 (Report 05000266/2024005 and 05000301/2024005) IR 05000266/20240022024-08-13013 August 2024 Integrated Inspection Report 05000266/2024002 and 05000301/2024002 L-2024-131, Response to Request for Additional Information Regarding License Amendment Request 300, Modify Containment Average Air Temperature Requirements2024-08-0909 August 2024 Response to Request for Additional Information Regarding License Amendment Request 300, Modify Containment Average Air Temperature Requirements ML24163A0012024-08-0505 August 2024 LTR-24-0119-1-1 Response to Nh Letter Regarding Review of NextEras Emergency Preparedness Amendment Review ML24214A3092024-08-0202 August 2024 Confirmation of Initial License Examination L-2024-113, License Amendment Request 294, Application to Revise Technical Specifications to Adopt TSTF- 577, Revised Frequencies for Steam Generator Tube Inspections2024-07-24024 July 2024 License Amendment Request 294, Application to Revise Technical Specifications to Adopt TSTF- 577, Revised Frequencies for Steam Generator Tube Inspections L-2024-125, Notice of Intent to Provide Supplemental Information to License Amendment Request to Adopt Common Emergency Plan with Site-Specific Annexes2024-07-24024 July 2024 Notice of Intent to Provide Supplemental Information to License Amendment Request to Adopt Common Emergency Plan with Site-Specific Annexes ML24194A1802024-07-24024 July 2024 – Revision to the Reactor Vessel Material Surveillance Capsule Withdrawal Schedule IR 05000266/20244012024-07-23023 July 2024 Public - Point Beach Nuclear Plant Cyber Security Inspection Report 05000266/2024401 and 05000301/2024401 ML24193A2432024-07-12012 July 2024 – Interim Audit Summary Report in Support of Review of License Amendment Requests Regarding Fleet Emergency Plan L-2024-116, Preparation and Scheduling of Operator Licensing Examinations2024-07-11011 July 2024 Preparation and Scheduling of Operator Licensing Examinations L-2024-114, Quality Assurance Topical Report (FPL-1 Revision 31 Annual Submittal2024-07-10010 July 2024 Quality Assurance Topical Report (FPL-1 Revision 31 Annual Submittal IR 05000266/20240102024-07-10010 July 2024 Age-Related Degradation Inspection Report 05000266/2024010 and 05000301/2024010 L-2024-105, License Amendment Request 300, Modify Containment Average Air Temperature Requirements2024-06-26026 June 2024 License Amendment Request 300, Modify Containment Average Air Temperature Requirements L-2024-107, Schedule for Subsequent License Renewal Environmental Review2024-06-25025 June 2024 Schedule for Subsequent License Renewal Environmental Review ML24176A2242024-06-24024 June 2024 Notification of an NRC Biennial Licensed Operator Requalification Program Inspection and Request for Information L-2024-102, Official Service List Update2024-06-19019 June 2024 Official Service List Update ML24149A2862024-06-12012 June 2024 NextEra Fleet - Proposed Alternative Frr 23-01 to Use ASME Code Case N-752-1, Risk-Informed Categorization and Treatment for Repair/Replacement Activities in Class 2 and 3 Systems Section X1, Division 1 (EPID L-2023-LLR-0009) - Letter L-2024-093, Steam Generator Divider Plate Assemblies Bounding Analysis Evaluation for Aging Management Commitment 14 Revision2024-06-10010 June 2024 Steam Generator Divider Plate Assemblies Bounding Analysis Evaluation for Aging Management Commitment 14 Revision IR 05000266/20244202024-06-0505 June 2024 Security Baseline Inspection Report 05000266/2024420 and 05000301/2024420 ML24149A1922024-05-28028 May 2024 Notification of NRC Baseline Inspection and Request for Information ML24141A1382024-05-20020 May 2024 Information Request to Support the NRC Annual Baseline Emergency Action Level and Emergency Plan Changes Inspection.Docx ML24127A0632024-05-0606 May 2024 Supplemental Information to License Amendment Request to Adopt Common Emergency Plan with Site-Specific Annexes IR 05000266/20243012024-04-29029 April 2024 NRC Initial License Examination Report 05000266/2024301; 05000301/2024301 L-2024-067, Annual Monitoring Report2024-04-26026 April 2024 Annual Monitoring Report ML24116A0402024-04-23023 April 2024 Periodic Update of the Updated Final Safety Analysis Report ML24071A0912024-04-22022 April 2024 Issuance of Relief Request I6-RR-03 - Extension of the Unit 2 Steam Generator Primary Nozzle Dissimilar Metal Welds Sixth 10-Year Inservice Inspection Program Interval IR 05000266/20240012024-04-11011 April 2024 Integrated Inspection Report 05000266/2024001 and 05000301/2024001 L-2024-030, Supplement to Relief Request CISl-03-01 for Relief Concerning Containment Unbonded Post-Tensioning System Inservice Inspection Requirements2024-03-27027 March 2024 Supplement to Relief Request CISl-03-01 for Relief Concerning Containment Unbonded Post-Tensioning System Inservice Inspection Requirements L-2024-043, Revised Reactor Vessel Materials Surveillance Capsule Withdrawal Schedules2024-03-25025 March 2024 Revised Reactor Vessel Materials Surveillance Capsule Withdrawal Schedules L-2024-011, and Point Beach, Units 1 and 2 - 10 CFR 50.46 Annual Reporting of Changes to, or Errors in Emergency Core Cooling System Models or Applications2024-03-13013 March 2024 and Point Beach, Units 1 and 2 - 10 CFR 50.46 Annual Reporting of Changes to, or Errors in Emergency Core Cooling System Models or Applications IR 05000266/20230062024-02-28028 February 2024 Annual Assessment Letter for Point Beach Nuclear Plant, Units 1 and 2 (Report 05000266/2023006 and 05000301/2023006) ML24053A3732024-02-22022 February 2024 Operator Licensing Examination Approval Point Beach, March 2024 L-2024-020, Refueling Outage Owners Activity Report (OAR-1) Unit 1 for Inservice Inspections2024-02-22022 February 2024 Refueling Outage Owners Activity Report (OAR-1) Unit 1 for Inservice Inspections ML24036A2652024-02-0505 February 2024 Notice of Inspection and Request for Information for the NRC Age-Related Degradation Inspection: Inspection Report 05000266/2024010 and 05000301/2024010 IR 05000266/20230042024-02-0101 February 2024 Integrated Inspection Report 05000266/2023004 and 05000301/2023004 ML24030A0352024-01-30030 January 2024 Information Request for the Cyber-Security Baseline Inspection, Notification to Perform Inspection L-2024-001, Relief Request CISl-03-01 for Relief Concerning Containment Unbonded Post-Tensioning System Inservice Inspection Requirements2024-01-26026 January 2024 Relief Request CISl-03-01 for Relief Concerning Containment Unbonded Post-Tensioning System Inservice Inspection Requirements L-2024-010, Point Units 3 and 4, Seabrook, Duane Arnold, and Point Beach Units 1 and 2, Nuclear Property Insurance - 10 CFR 50.54(w)(3)2024-01-25025 January 2024 Point Units 3 and 4, Seabrook, Duane Arnold, and Point Beach Units 1 and 2, Nuclear Property Insurance - 10 CFR 50.54(w)(3) ML24005A3242024-01-24024 January 2024 Exemption from Select Requirements of 10 CFR Part 73 (EPID L-2023-LLE-0040 (Security Notifications, Reports, and Recordkeeping and Suspicious Activity Reporting)) ML23352A2752024-01-23023 January 2024 Issuance of Amendment Nos. 274 and 276 Regarding Revision to Technical Specification 5.5.17, Pre-Stressed Concrete Containment Tendon Surveillance Program L-2023-173, Quality Assurance Topical Report (FPL-1) Revision 30 Update2023-12-15015 December 2023 Quality Assurance Topical Report (FPL-1) Revision 30 Update L-2023-174, Subsequent License Renewal Application - Third Annual Update2023-12-13013 December 2023 Subsequent License Renewal Application - Third Annual Update L-2023-176, Supplement to Exemption Request Regarding Enhanced Weapons, Firearms Background Checks, and Security Event Notifications Final Rule2023-11-29029 November 2023 Supplement to Exemption Request Regarding Enhanced Weapons, Firearms Background Checks, and Security Event Notifications Final Rule L-2023-155, Supplement to Response to Request for Additional Information, Revised NextEra Common Emergency Plan, and Revised Site-Specific Emergency Plan Annexes Regarding License Amendment Request for Common Emergency Plan Consistent with NUREG-06542023-11-28028 November 2023 Supplement to Response to Request for Additional Information, Revised NextEra Common Emergency Plan, and Revised Site-Specific Emergency Plan Annexes Regarding License Amendment Request for Common Emergency Plan Consistent with NUREG-0654, 2024-09-09
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ber 25, 2011
SUBJECT:
ERRATA TO POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2; COMPONENT DESIGN BASES INSPECTION (CDBI) REPORT 05000266/2011009; 05000301/2011009
Dear Mr. Meyer:
On October 18, 2011, the U.S. Nuclear Regulatory Commission (NRC) issued Component Design Bases Inspection (CDBI) Report 05000266/2011009; 05000301/2011009 (ML11291A094). In the inspection report, the summary of findings incorrectly annotated a cross-cutting aspect for Non-Cited Violation (NCV)05000266/2011009-03; 05000301/2011009-03. As indicated in the body of the inspection report, no cross-cutting aspect was identified for this issue. In addition, the alphanumeric identifier for the cross-cutting aspect for finding (FIN)05000266/2011009-04; 05000301/2011009-04 was incorrect. As indicated in the body of the inspection report, the correct alphanumeric is H.4.c. Please replace pages 1 and 2 of Inspection Report 05000266/2011009; 05000301/2011009 with the enclosed corrected pages.
We apologize for any inconvenience to you and your staff.
Sincerely,
/RA/
Ann Marie Stone, Chief Engineering Branch 2 Division of Reactor Safety Docket Nos. 50-266; 50-301 License No. DPR-24; DPR-27 Enclosure: Errata to Inspection Report 05000266/2011009; 05000301/2011009 cc w/encl: Distribution via ListServ
SUMMARY OF FINDINGS IR 05000266/2011009, 05000301/2011009; 8/01/2011 - 9/02/2011; Point Beach Nuclear Plant, Units 1 and 2; Component Design Bases Inspection (CDBI).
The inspection was a 3-week onsite baseline inspection that focused on the design of components. The inspection was conducted by regional engineering inspectors and two consultants. Four Green finding were identified by the inspectors. Three of the findings were considered Non-Cited Violations (NCVs) of NRC regulations. The significance of most findings is indicated by their color (Green, White, Yellow, Red) using Inspection Manual Chapter (IMC)
0609, Significance Determination Process (SDP). Findings for which the SDP does not apply may be (Green) or be assigned a severity level after NRC management review. The NRCs program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, Reactor Oversight Process, Revision 4, dated December 2006.
A. NRC-Identified and Self-Revealed Findings Cornerstone: Initiating Events
- Green. The inspectors identified a finding of very low safety significance involving the licensees failure to meet the requirements of the American Institute of Steel Construction (AISC) Specification. Specifically, the licensees design basis calculation failed to ensure the turbine building structural steel floor beams met the AISC specification. This finding was entered into the licensees corrective action program.
No violation of NRC requirements was identified.
The performance deficiency was determined to be more than minor because the finding was associated with the Initiating Events Cornerstone attribute of design control and adversely affected the cornerstone objective to limit the likelihood of those events that upset the plant stability and challenge critical safety functions during shutdown as well as power operations. The finding screened as of very low safety significance (Green)
because the transient initiator would not contribute to both the likelihood of a reactor trip and the likelihood that mitigation equipment or functions will not be available. This finding had a cross-cutting aspect in human performance, work practice because the licensee did not ensure effective supervisory and management oversight of work activities, including contractors, such that nuclear safety was supported. Specifically, the licensee failed to have adequate oversight of design calculation and documentation for establishing structural adequacy of the turbine building structural steel beams at EL. 44-0. H.4(c) (Section 4OA5.1.b.(2))
Cornerstone: Mitigating Systems
- Green. The inspectors identified a finding of very low safety significance (Green) and associated Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, involving the licensees failure to correctly translate design basis assumptions into procedures or instructions. Specifically, the licensee failed to monitor average outside air temperature which was one of the design input criteria for the temperature heat-up calculation associated with rooms which housed safety-related equipment. This finding was entered into the licensees corrective action program.
1 Enclosure
The performance deficiency was associated with Mitigating System Cornerstone and determined to be more than minor because, if left uncorrected, it could lead to a more significant safety concern. The finding screened as very low safety significance (Green)
because the finding was not a design or qualification deficiency, did not represent a loss of system safety function, and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. The finding had a cross-cutting aspect in the area of human performance, resources because the licensee did not ensure adequate training and qualification of personnel. Specifically, the licensee failed to adequately train licensed operators to ensure adequate knowledge with respect to the interface between functionality of a non-safety system component and the impact of a failure on the operability of safety-related equipment. H.2(b). (Section 1R21.3.b.(1))
- Green. The inspectors identified a finding of very low safety significance (Green) and associated Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the failure to ensure a minimum AFW flow of 275 gpm as specified in the accident analysis for the Loss of Normal Feedwater event. This finding was entered into the licensees corrective action program.
The performance deficiency was associated with the Mitigating Systems Cornerstone attribute of design control and was determined to be more than minor because, if left uncorrected, it would have the potential to lead to a more significant safety concern.
Specifically, an AFW flow rate of less than 275 gpm as specified in the procedures did not ensure the pressurizer would not become water solid and cause an over-pressure condition within the Reactor Coolant System during the Loss of Normal Feedwater. The finding screened as of very low safety significance (Green) because the finding was not a design or qualification deficiency, did not represent a loss of system safety function, and did not screen as potentially risk-significant due to a seismic, flooding, or severe weather initiating event. This finding had a cross-cutting aspect in the area of human performance, resources because the licensee did not maintain design documentation in a complete and accurate manner. Specifically, the licensee failed to maintain Emergency Procedures consistent with the design basis analysis for LONF. H.2(c).
(Section 1R21.6.b.(1))
Cornerstone: Barrier Integrity
The performance deficiency was determined to be more than minor because it was associated with the Barrier Integrity Cornerstone attribute of design control and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers (fuel cladding, reactor coolant system, and containment) protect the public from radionuclide releases caused by accidents or events. This finding is of very low safety significance (Green) because there was no actual barrier degradation.
The inspectors did not identify a cross-cutting aspect associated with this finding because this was a legacy design issue; and therefore, was not reflective of current
SUBJECT:
ERRATA TO POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2; COMPONENT DESIGN BASES INSPECTION (CDBI) REPORT 05000266/2011009; 05000301/2011009
Dear M