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Category:Letter type:LR
MONTHYEARLR-N23-0065, Submittal of 2023 Annual 10 CFR 50.46 Report2023-10-0202 October 2023 Submittal of 2023 Annual 10 CFR 50.46 Report LR-N23-0045, and Peach Bottom Atomic Power Station, Units 2 and 3 - Notice of Proposed Amendment to Decommissioning Trust Agreement2023-09-0808 September 2023 and Peach Bottom Atomic Power Station, Units 2 and 3 - Notice of Proposed Amendment to Decommissioning Trust Agreement LR-N23-0052, Retest Schedule for Drywell to Suppression Chamber Vacuum Breakers Per Technical Specification 4.6.2.12023-07-31031 July 2023 Retest Schedule for Drywell to Suppression Chamber Vacuum Breakers Per Technical Specification 4.6.2.1 LR-N23-0042, Spent Fuel Cask Registration2023-07-12012 July 2023 Spent Fuel Cask Registration LR-N23-0046, Emergency Plan Document Revisions Implemented June 28, 20232023-07-10010 July 2023 Emergency Plan Document Revisions Implemented June 28, 2023 LR-N23-0034, 2022 Annual Radiological Environmental Operating Report (AREOR) - Salem Nuclear Generating Station, Unit Nos. 1 and 2 and Hope Creek Generating Station2023-04-27027 April 2023 2022 Annual Radiological Environmental Operating Report (AREOR) - Salem Nuclear Generating Station, Unit Nos. 1 and 2 and Hope Creek Generating Station LR-N23-0035, 2022 Annual Radioactive Effluent Release Report (ARERR)2023-04-27027 April 2023 2022 Annual Radioactive Effluent Release Report (ARERR) LR-N23-0010, License Amendment Request Revision of Technical Specification (TS) to Delete TS Section 5.5 - Meteorological Tower Location2023-04-21021 April 2023 License Amendment Request Revision of Technical Specification (TS) to Delete TS Section 5.5 - Meteorological Tower Location LR-N23-0009, License Amendment Request (LAR) to Revise the Hope Creek Trip and Standby Auto-start Logic Associated with Safety Related Heating, Ventilation and Air Conditioning (HVAC) Trains2023-04-18018 April 2023 License Amendment Request (LAR) to Revise the Hope Creek Trip and Standby Auto-start Logic Associated with Safety Related Heating, Ventilation and Air Conditioning (HVAC) Trains LR-N23-0024, Submittal of Hope Creek Generating Station Technical Specification Bases Changes2023-03-29029 March 2023 Submittal of Hope Creek Generating Station Technical Specification Bases Changes LR-N23-0006, Report on Status of Decommissioning Funding for Reactors and Independent Spent Fuel Storage Installations2023-03-24024 March 2023 Report on Status of Decommissioning Funding for Reactors and Independent Spent Fuel Storage Installations LR-N23-0019, and Salem Generating Station, Units 1 and 2 - Guarantees of Payment of Deferred Premiums2023-03-21021 March 2023 and Salem Generating Station, Units 1 and 2 - Guarantees of Payment of Deferred Premiums LR-N23-0016, and Salem Generating Station, Units 1 and 2 - Report of Changes, Tests, and Experiments2023-02-28028 February 2023 and Salem Generating Station, Units 1 and 2 - Report of Changes, Tests, and Experiments LR-N23-0018, Technical Specification 6.9.1.5.b - 2022 Annual Report of SRV Challenges2023-02-27027 February 2023 Technical Specification 6.9.1.5.b - 2022 Annual Report of SRV Challenges LR-N23-0012, Annual Property Insurance Status Report2023-02-24024 February 2023 Annual Property Insurance Status Report LR-N23-0014, Stations Submittal of 2022 Annual Report of Fitness for Duty Performance Data Per 10 CFR 26.203(e) and 10 CFR 26.7172023-02-23023 February 2023 Stations Submittal of 2022 Annual Report of Fitness for Duty Performance Data Per 10 CFR 26.203(e) and 10 CFR 26.717 LR-N23-0011, In-Service Inspection Activities - 90 Day Report: Twenty-Fourth Refueling Outage2023-01-19019 January 2023 In-Service Inspection Activities - 90 Day Report: Twenty-Fourth Refueling Outage LR-N22-0096, and Salem Generating Station, Units 1 and 2 - Request for Threshold Determination2023-01-0505 January 2023 and Salem Generating Station, Units 1 and 2 - Request for Threshold Determination LR-N22-0094, Emergency Plan Document Revisions Implemented November 21, 20222022-12-14014 December 2022 Emergency Plan Document Revisions Implemented November 21, 2022 LR-N22-0091, Independent Spent Fuel Storage Installation, Report of 10 CFR 72.48 Changes, Tests, and Experiments2022-12-0202 December 2022 Independent Spent Fuel Storage Installation, Report of 10 CFR 72.48 Changes, Tests, and Experiments LR-N22-0075, 2022 Annual 10 CFR 50.46 Report2022-09-30030 September 2022 2022 Annual 10 CFR 50.46 Report LR-N22-0074, Emergency Plan Evacuation Time Estimate2022-09-15015 September 2022 Emergency Plan Evacuation Time Estimate LR-N22-0051, License Amendment Request to Relocate Technical Specification Facility/Unit Staff Qualification Requirements to Quality Assurance Topical Report2022-06-22022 June 2022 License Amendment Request to Relocate Technical Specification Facility/Unit Staff Qualification Requirements to Quality Assurance Topical Report LR-N22-0044, Emergency Plan Document Revisions Implemented November, 20212022-05-19019 May 2022 Emergency Plan Document Revisions Implemented November, 2021 LR-N22-0041, 2021 Annual Radioactive Effluent Release Report (Rerr)2022-04-28028 April 2022 2021 Annual Radioactive Effluent Release Report (Rerr) LR-N22-0040, 2021 Annual Radiological Environmental Operating Report2022-04-28028 April 2022 2021 Annual Radiological Environmental Operating Report LR-N22-0039, Emergency Plan Document Revisions Implemented March 24, 20222022-04-21021 April 2022 Emergency Plan Document Revisions Implemented March 24, 2022 LR-N22-0023, Guarantees of Payment of Deferred Premiums2022-03-21021 March 2022 Guarantees of Payment of Deferred Premiums LR-N22-0017, Submittal of 2021 Annual Report of Fitness for Duty (FFD) Performance Data2022-02-25025 February 2022 Submittal of 2021 Annual Report of Fitness for Duty (FFD) Performance Data LR-N22-0016, Radiological Survey of Site Property to Be Used for Offshore Wind Port Facility2022-02-24024 February 2022 Radiological Survey of Site Property to Be Used for Offshore Wind Port Facility LR-N22-0019, Technical Specification 6.9.1.5.b 2021 Annual Report of SRV Challenges2022-02-24024 February 2022 Technical Specification 6.9.1.5.b 2021 Annual Report of SRV Challenges LR-N22-0011, Response to Request for Additional Information for License Amendment Request to Revise Technical Specification Limits for Ultimate Heat Sink2022-02-0101 February 2022 Response to Request for Additional Information for License Amendment Request to Revise Technical Specification Limits for Ultimate Heat Sink LR-N22-0005, Proposed Relief Request Associated with Reactor Pressure Vessel Water Level Instrumentation Partial Penetration Nozzle Repairs2022-01-0707 January 2022 Proposed Relief Request Associated with Reactor Pressure Vessel Water Level Instrumentation Partial Penetration Nozzle Repairs LR-N22-0007, Request for Exemption from Specific Requirements of 10 CFR Part 26, Fitness for Duty Programs2022-01-0505 January 2022 Request for Exemption from Specific Requirements of 10 CFR Part 26, Fitness for Duty Programs LR-N21-0087, Corrected Hope Creek 10 CFR 50.46 Reports2021-12-0202 December 2021 Corrected Hope Creek 10 CFR 50.46 Reports LR-N21-0078, Hope and Creek Generating Station, Supplement to License Amendment Request to Revise Technical Specifications (TS) to Delete Definitions Found in 10 CFR Part 20 and Delete Figures of the Site and Surrounding Areas from TS2021-11-18018 November 2021 Hope and Creek Generating Station, Supplement to License Amendment Request to Revise Technical Specifications (TS) to Delete Definitions Found in 10 CFR Part 20 and Delete Figures of the Site and Surrounding Areas from TS LR-N21-0081, Submittal of Updated Final Safety Analysis Report, Revision 25, Summary of Revised Regulatory Commitments2021-11-15015 November 2021 Submittal of Updated Final Safety Analysis Report, Revision 25, Summary of Revised Regulatory Commitments LR-N21-0056, License Amendment Request to Amend the Technical Specifications to Revise Surveillance Requirements for Electric Power Monitor Channels for the Reactor Protection System (RPS) and Power Range Neutron Monitoring System2021-11-0303 November 2021 License Amendment Request to Amend the Technical Specifications to Revise Surveillance Requirements for Electric Power Monitor Channels for the Reactor Protection System (RPS) and Power Range Neutron Monitoring System LR-N21-0079, Submittal of Hope Creek Generating Station Technical Specification Bases Changes2021-11-0303 November 2021 Submittal of Hope Creek Generating Station Technical Specification Bases Changes LR-N21-0071, 2021 Annual 10 CFR 50.46 Report2021-09-30030 September 2021 2021 Annual 10 CFR 50.46 Report LR-N21-0065, License Amendment Request - Revision of Salem and Hope Creek Generating Station Technical Specification (TS) to Delete Definitions Found in 10 CFR Part 20 and Delete Figures of the Site and Surrounding Areas from TS2021-09-29029 September 2021 License Amendment Request - Revision of Salem and Hope Creek Generating Station Technical Specification (TS) to Delete Definitions Found in 10 CFR Part 20 and Delete Figures of the Site and Surrounding Areas from TS LR-N21-0069, Emergency Plan Document Revisions Implemented August 10, 20212021-09-0909 September 2021 Emergency Plan Document Revisions Implemented August 10, 2021 LR-N21-0059, In-Service Inspection Activities - 90 Day Twenty-Third Refueling Outage2021-08-13013 August 2021 In-Service Inspection Activities - 90 Day Twenty-Third Refueling Outage LR-N21-0054, Correction to Prior Spent Fuel Cask Registration Letter2021-07-15015 July 2021 Correction to Prior Spent Fuel Cask Registration Letter LR-N21-0046, Submittal of Salem Generating Station Updated Final Safety Analysis Report, Revision 32, Summary of Revised Regulatory Commitments for Salem, 10 CFR 71.106 Review Results and 10 CFR 54.37(b) Review Results for Salem2021-06-17017 June 2021 Submittal of Salem Generating Station Updated Final Safety Analysis Report, Revision 32, Summary of Revised Regulatory Commitments for Salem, 10 CFR 71.106 Review Results and 10 CFR 54.37(b) Review Results for Salem LR-N21-0040, Response to Request for Additional Information SNSB-RAI 1 License Amendment Request to Revise Low Pressure Safety Limit to Address General Electric Part 21 Safety Communication2021-05-27027 May 2021 Response to Request for Additional Information SNSB-RAI 1 License Amendment Request to Revise Low Pressure Safety Limit to Address General Electric Part 21 Safety Communication LR-N21-0042, Core Operating Limits Report, Reload 23, Cycle 24, Revision 212021-05-24024 May 2021 Core Operating Limits Report, Reload 23, Cycle 24, Revision 21 LR-N21-0025, License Amendment Request to Revise Technical Specification Limits for Ultimate Heat Sink2021-05-0707 May 2021 License Amendment Request to Revise Technical Specification Limits for Ultimate Heat Sink LR-N21-0039, Deviation from EPRI Document 33002012244 Inspection Requirements2021-04-30030 April 2021 Deviation from EPRI Document 33002012244 Inspection Requirements LR-N21-0018, Response to Requests for Additional Information SNSB-RAI 2 and SNSB-RAI 3 License Amendment Request to Revise Low Pressure Safety Limit to Address General Electric Part 21 Safety Communication2021-04-29029 April 2021 Response to Requests for Additional Information SNSB-RAI 2 and SNSB-RAI 3 License Amendment Request to Revise Low Pressure Safety Limit to Address General Electric Part 21 Safety Communication 2023-09-08
[Table view] Category:Licensee Event Report (LER)
MONTHYEAR05000354/LER-2017-0012017-07-0707 July 2017 1 OF 3, LER 17-001-00 for Hope Creek, Unit 1 Regarding Secondary Containment Door Not Latched in Closed Position 05000354/LER-2016-0062017-05-0404 May 2017 LER 16-006-01 for Hope Creek, Unit 1, Regarding Mode Change Without B Channel Level Instrumentation Operable 05000354/LER-2016-0052017-03-13013 March 2017 Reactor Protection System Actuation While the Reactor Was Shutdown, LER 16-005-01 for Hope Creek, Unit 1, Regarding Reactor Protection System Actuation While the Reactor Was Shutdown 05000354/LER-2016-0032017-03-0808 March 2017 As-Found Values for Safety Relief Valve Lift Set Points Exceed Technical Specification Allowable Limit, LER 16-003-001 for Hope Creek Generating Station Unit 1 Regarding As-Found Values for Safety Relief Valve Lift Set Points Exceed Technical Specification Allowable Limit 05000354/LER-2016-0042016-12-20020 December 2016 1 OF 3, LER 16-004-00 for Hope Creek Regarding Operations With a Potential to Drain the Reactor Vessel (OPDRV) Without Secondary Containment 05000354/LER-2015-0052016-01-0505 January 2016 Reactor Scram Due to Invalid RRCS Actuation, LER 15-005-01 for Hope Creek, Unit 1, Regarding Reactor Scram Due to Invalid RRCS Actuation LR-N12-0114, Retraction of Licensee Event Report 2011-0012012-04-13013 April 2012 Retraction of Licensee Event Report 2011-001 LR-N05-0143, Special Report 05-001-01 Regarding the Cause of Failure and Channel Restoration2005-03-31031 March 2005 Special Report 05-001-01 Regarding the Cause of Failure and Channel Restoration LR-N04-0526, Special Report 354/04-012-002004-11-18018 November 2004 Special Report 354/04-012-00 LR-N04-0489, LER 04-08-00 for Hope Creek Generating Station Re Potential for Uncontrolled Radiological Release - Reactor Water Clean-up Isolation2004-10-28028 October 2004 LER 04-08-00 for Hope Creek Generating Station Re Potential for Uncontrolled Radiological Release - Reactor Water Clean-up Isolation 2017-07-07
[Table view] |
Text
PSEG Nuclear LLC P.O. Box 236, Hancocks Bridge, New Jersey 08038-0236 SPSEG APR 13 2012 Nuclear LLC LR-N12-0114 10 CFR 50.73 NUREG 1022 United States Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-001 Hope Creek Generating Station Unit 1 Facility Operating License Number NPF-57 Docket Number 50-354
Subject:
Retraction of Hope Creek Licensee Event Report 2011-001 In accordance with NUREG 1022 Sections 2.8 and 5.1.2, PSEG Nuclear LLC is retracting (formally withdrawing)
Licensee Event Report (LER) Number 2011-001.LER 2011-001 was transmitted to the NRC via letter LR-N1 1-0294 dated September 22, 2011.The LER reported, under 10CFR50.73 (a) (2) (v) (D), an event or condition that could have prevented the fulfillment of the safety function of an SSC that is needed to mitigate the consequences of an accident.Subsequent to submittal of LER 2011-001, Hope Creek Engineering performed analyses (discussed in the Attachment) using more accurate input conditions.
These analyses have determined that the HPCI system (the system that was the subject of the LER) was able to perform its safety function.
Therefore PSEG Nuclear is retracting LER 2011-001.LER H201 1-001 reported that a supplement to the LER was anticipated.
This retraction eliminates the need for that supplement.
Should you have any questions concerning this letter, please contact Mr. Philip J. Duca at (856)339-1640.No regulatory commitments are contained in this correspondence.
Sincerely, David P. Lewis Plant Manager Hope Creek Generating Station
Attachment:
LER 2011-001 Retraction
".A C Page 2 LR-N12-0114 Document Control Desk cc: Mr. W. Dean, Administrator
-Region I U.S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 Mr. John Hughey, Project Manager Salem and Hope Creek U.S. Nuclear Regulatory Commission One White Flint North Mail Stop 08 BlA 11555 Rockville Pike Rockville, MD 20852 USNRC Senior Resident Inspector
-Hope Creek (X24)P. Mulligan, Manager IV Bureau of Nuclear Engineering PO Box 415 Trenton, NJ 08625 Hope Creek Commitment Tracking Coordinator (H02)INPO -LEREvents@lNPO.org LR-N12-0114 Document Control Desk Attachment 1 LER 2011-001 Retraction Page 1 of 2 Background In August of 2011, Hope Creek Engineering identified a condition in which the high pressure coolant injection (HPCI) system potentially could be prevented from fulfilling its safety function.The HPCI room ventilation delta- temperature trip of 70 degrees F, which isolates HPCI in the event of a steam leak, had the potential to isolate HPCI prematurely in extreme winter conditions.
This would impact the ability of HPCI to fulfill its design function during the accident scenario listed in UFSAR Table 6.3-6 where one of the assumed single failures listed is the loss of an emergency diesel generator (EDG) coincident with a loss of coolant accident (LOCA) and a loss of offsite power (LOP). This would result in the loss of the HPCI Room Coolers.During a design basis event (DBE) the HPCI Room is supplied by the Filtration Recirculation Ventilation System (FRVS). The design for the HPCI System is to trip when room temperature exceeds 160°F or when the ventilation air temperature difference across the HPCI room exceeds 70'F. During August of 2011 a technical evaluation concluded that HPCI room ventilation outlet to inlet temperature might exceed 70°F if normal operating Safety Auxiliary Cooling System (SACS) loop temperature was below 45 0 F. SACS is the cooling water supply for the FRVS. At that time, a realistic time history of FRVS supply temperature to HPCI room associated with the lowest SACS temperature was not available and as such a constant conservative low FRVS supply temperature to the HPCI Room was used. Based on this evaluation it was concluded that HPCI potentially could be unable to fulfill its safety function during extreme cold environmental temperatures.
Therefore LER 2011-001 was submitted.
In order to preclude this potential, a design change package (DCP) was developed and implemented.
This DCP lowered the set-point temperature of the HPCI Room Coolers so that if the accident scenario discussed above did occur the starting ambient temperature of the HPCI room would be lower (closer to the FRVS inlet temperature to the room) and the ventilation air temperature difference would not exceed 70 0 F. This resulted in, using a conservative 35 0 F FRVS supply temperature as input, the calculated differential room temperature not reaching the 70°F trip set-point.
Implementation of this DCP eliminated the potential loss of the HPCI safety function during the loss of an emergency diesel generator (EDG) coincident with a loss of coolant accident (LOCA) and a loss of offsite power (LOP).Discussion Subsequent to submittal of LER 2011-001, a winter time Reactor Building Ventilation System analysis was performed and this analysis provided realistic low values of temperature time history of FRVS supply temperature to the HPCI Room. The analysis was performed using a GOTHIC model of the Reactor Building under winter conditions.
The analysis was performed in two steps. First simultaneous transient analysis of SACS loops temperatures and FRVS supply temperatures was performed.
The output of this analysis was used as input to the HPCI room ventilation analysis.
The HPCI room ventilation analysis then calculated the maximum temperature difference across the HPCI Room when the HPCI system is operating and the room coolers are not running.
LR-N12-0114 Document Control Desk Attachment 1 LER 2011-001 Retraction Page 2 of 2 The GOTHIC model analyzed eight cases. The results of all the cases were used to determine the sensitivity of various parameters on the FRVS supply temperature.
The case that provided the most realistic low values of SACS loop time temperature time histories and FRVS supply temperature to the HPCI Room was used as the input to the HPCI room ventilation analysis.
A design basis that established the realistic low values of SACS and FRVS supply temperatures was created. The results of this analysis showed that the HPCI room ventilation outlet to inlet temperature would not exceed 65 0 F. This analysis was performed using the room cooler set-points that existed before implementation of the DCP.Therefore, HPCI was always able to fulfill its safety function during extreme cold environmental temperatures during the loss of an emergency diesel generator (EDG) coincident with a loss of coolant accident (LOCA) and a loss of offsite power (LOP).