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Category:Letter
MONTHYEARIR 05000237/20230042024-02-0202 February 2024 Integrated Inspection Report 05000237/2023004 and 05000249/2023004 IR 05000010/20230012024-01-17017 January 2024 NRC Inspection Report 05000010/2023001 (Drss) ML23354A2812024-01-0404 January 2024 Exemption from Select Requirements of 10 CFR Part 73 (EPID L-2023-LLE-0031 (Security Notifications, Reports, and Recordkeeping and Suspicious Activity Reporting)) ML23305A1402023-12-13013 December 2023 Units 1 & 2; Nine Mile Point, Unit 2; Peach Bottom, Units 2 & 3; and Quad Cities, Units 1 and 2 - Issuance of Amendments to Adopt Traveler TSTF-580 RS-23-121, Sixth Ten-Year Interval Inservice Testing Program2023-11-30030 November 2023 Sixth Ten-Year Interval Inservice Testing Program SVPLTR 23-0037, Core Operating Limits Report for Cycle 292023-11-29029 November 2023 Core Operating Limits Report for Cycle 29 RS-23-119, 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 RS-23-107, Relief Request I5R-22, Inservice Inspection Program Relief Request Regarding Examination Coverage for the Fifth Inservice Inspection Interval2023-11-0808 November 2023 Relief Request I5R-22, Inservice Inspection Program Relief Request Regarding Examination Coverage for the Fifth Inservice Inspection Interval RS-23-102, Request for Exemption from Enhanced Weapons, Firearms Background Checks, and Security Event Notifications Implementation2023-10-16016 October 2023 Request for Exemption from Enhanced Weapons, Firearms Background Checks, and Security Event Notifications Implementation ML23285A2342023-10-13013 October 2023 Confirmation of Initial License Examination 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 ML23279A0712023-10-0606 October 2023 Response to NRC Dresden Nuclear Power Station, Units 2 and 3 - Comprehensive Engineering Team Inspection Report 05000237/2023011 and 05000249/2023011 ML23269A0662023-09-28028 September 2023 Proposed Alternative Request RV-23H Associated with the Sixth 10-Year Inservice Testing Interval ML23265A1322023-09-27027 September 2023 Proposed Alternative Request RV-03 Associated with the Sixth 10-Year Inservice Testing Interval ML23205A1362023-09-20020 September 2023 Proposed Alternative Request RV-02D Associated with the Sixth IST Interval IR 05000237/20230112023-09-0606 September 2023 Comprehensive Engineering Team Inspection Report 05000237/2023011 and 05000249/2023011 RS-23-071, Application to Adopt TSTF-564, Safety Limit MCPR2023-08-30030 August 2023 Application to Adopt TSTF-564, Safety Limit MCPR RS-23-080, Constellation Energy Generation, LLC, Application to Revise Technical Specifications to Adopt TSTF-264-A, Revision 0, 3.3.9 and 3.3.10 - Delete Flux Monitors Specific Overlap Requirement SRs2023-08-30030 August 2023 Constellation Energy Generation, LLC, Application to Revise Technical Specifications to Adopt TSTF-264-A, Revision 0, 3.3.9 and 3.3.10 - Delete Flux Monitors Specific Overlap Requirement SRs IR 05000237/20234022023-08-29029 August 2023 Security Baseline Inspection Report 05000237/2023402 and 05000249/2023402 IR 05000237/20230022023-08-0808 August 2023 Integrated Inspection Report 05000237/2023002 and 05000249/2023002 ML23212A9892023-08-0303 August 2023 Subsequent License Renewal Application Public Meeting Summary of Pre-Submittal Meeting ML23174A1502023-08-0303 August 2023 Issuance of Amendment Nos. 282 and 275 Control Rod Scram Times ML23193A0052023-07-28028 July 2023 Subsequent License Renewal Application Public Meeting Summary of Pre-Submittal Meeting ML23192A6292023-07-11011 July 2023 Correction Letter - Issuance of Amendment Nos. 281 and 274 Regarding Transition to GNF3 Fuel (EPID L-2022-LLA-0121) (Non-Proprietary) ML23144A3142023-07-0606 July 2023 Issuance of Amendment Nos. 281 and 274 Transition to GNF3 Fuel (EPID L-2022-LLA-0121) - Nonproprietary ML23187A0462023-07-0606 July 2023 Notification of NRC Baseline Inspection and Request for Information; Inspection Report 05000237/2023004 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 SVPLTR 23-0021, Registration of Use of Casks to Store Spent Fuel2023-06-20020 June 2023 Registration of Use of Casks to Store Spent Fuel RS-23-078, Supplement to License Amendment Request to Revise Technical Specification 3.1.4, Control Rod Scram Times2023-06-16016 June 2023 Supplement to License Amendment Request to Revise Technical Specification 3.1.4, Control Rod Scram Times 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 IR 05000237/20233012023-05-30030 May 2023 NRC Initial License Examination Report 05000237/2023301; 05000249/2023301 RS-23-042, Application to Revise Technical Specifications to Adopt TSTF-580, Provide Exception from Entering Mode 4 with No Operable RHR Shutdown Cooling2023-05-25025 May 2023 Application to Revise Technical Specifications to Adopt TSTF-580, Provide Exception from Entering Mode 4 with No Operable RHR Shutdown Cooling ML23132A0932023-05-17017 May 2023 Information Meeting (Public Open House) with a Question and Answer Session to Discuss NRC 2022 End-Of-Cycle Plant Performance Assessment of Dresden Power Station, Units 2 and 3 RS-23-070, Supplement to License Amendment Request Regarding Transition to GNF3 Fuel2023-05-16016 May 2023 Supplement to License Amendment Request Regarding Transition to GNF3 Fuel SVPLTR 23-0018, Submittal of 2022 Annual Radiological Environmental Operating Report2023-05-15015 May 2023 Submittal of 2022 Annual Radiological Environmental Operating Report IR 05000237/20230012023-05-0404 May 2023 Integrated Inspection Report 05000237/2023001 and 05000249/2023001 SVPLTR 23-0019, ILT 22-1 (2023-301) Svpltr 23-0019 Post-Exam Submittal Letter (1)2023-05-0101 May 2023 ILT 22-1 (2023-301) Svpltr 23-0019 Post-Exam Submittal Letter (1) ML23118A3472023-05-0101 May 2023 County, 1 & 2; Nine Mile Point, 2; and Quad Cities, 1 & 2 - Correction of Amendment No. 193 Adoption of TSTF-306, Revision 2, Add Action to LCO 3.3.6.1 to Give Option to Isolate the Penetration EPID L-2022-LLA-0143 ML23114A2522023-04-28028 April 2023 Request to Use a Provision of a Later Edition of the ASME Boiler & Pressure Vessel Code, Section XI IR 05000237/20234012023-04-26026 April 2023 Cyber Security Inspection Report 05000237/2023401 and 05000249/2023401 ML23081A0382023-04-25025 April 2023 County, 1 & 2; Nine Mile Point, 2; and Quad Cities, 1 & 2 - Issuance of Amendments to Adopt TSTF-306, Rev. 2, Add Action to LCO 3.3.6.1 to Give Option to Isolate the Penetration IR 05000237/20235012023-04-21021 April 2023 Emergency Preparedness Biennial Exercise Inspection Report 05000237/2023501 and 05000249/2023501 ML23110A3202023-04-21021 April 2023 Information Request to Support the NRC Annual Baseline Emergency Action Level and Emergency Plan Changes Inspection RS-23-057, Supplement to License Amendment Request Regarding Transition to GNF3 Fuel2023-04-17017 April 2023 Supplement to License Amendment Request Regarding Transition to GNF3 Fuel SVPLTR 23-0014, Annual Dose Report for 20222023-04-15015 April 2023 Annual Dose Report for 2022 ML23088A2242023-04-0505 April 2023 Regulatory Audit Report Regarding License Amendment Request to Transition to GNF3 Fuel ML23087A0482023-03-28028 March 2023 Operator Licensing Examination Approval - Dresden Nuclear Power Station, April 2023 2024-02-02
[Table view] Category:Licensee Event Report (LER)
MONTHYEARML20135G7442020-05-14014 May 2020 Final ASP Analysis - Dresden 2 (LER 237-90-006) 05000249/LER-2017-0012017-12-27027 December 2017 Unit 3 Standby Liquid Control System Inoperable Due to a Manufacturing Defect Causing a Piping Leak, LER 17-001-01 for Dresden Nuclear Power Station, Unit 3 Regarding Unit 3 Standby Liquid Control System Inoperable Due to a Manufacturing Defect Causing a Piping Leak ML17252B5052017-11-0303 November 2017 LER 77-050/03L-0 for Dresden, Unit 2 Re Local Leak Rate Test of Drywell Personnel Air Lock During Refueling ML17252A3262017-10-21021 October 2017 LER 77-040/03L-0 for Dresden 3 Regarding the Circuit Breaker for Service Water Makeup Valve MO-3-4102 to the Isolation Condenser Found in Off Position ML17252B4922017-08-28028 August 2017 LER 78-028/01X-1 for Dresden, Units 2 and 3 Re NRC Request for Immediate Actions to Be Taken to Mitigate the Potential for a Spurious Closure of a Recirculation Loop Suction Valve with a LOCA Occurring Between the Loop Discharge and Suction 05000237/LER-2016-0032017-05-26026 May 2017 Control Room Emergency Ventilation System Charcoal Filter Bank Failure to meet the Methyl Iodide Penetration Acceptance Criteria, Dresden Nuclear Power Station, Units 2 and 3, Cancellation of Licensee Event Report 237/2016-003-00, Control Room Emergency Ventilation System Charcoal Filter Bank Failure to meet the Methyl Iodide Penetration Acceptance Criteria ML17153A0342017-05-26026 May 2017 Cancellation of Licensee Event Report 237/2016-003-00, Control Room Emergency Ventilation System Charcoal Filter Bank Failure to Meet the Methyl Iodide Penetration Acceptance Criteria 05000237/LER-2016-0042017-01-0909 January 2017 Reactor Building Differential Pressure Less than Technical Specification Requirement, LER 16-004-00 for Dresden Nuclear Power Station, Unit 2, Regarding Reactor Building Differential Pressure Less than Technical Specification Requirement 05000249/LER-2016-0012016-08-25025 August 2016 Alert Declared from Unit 3 HPCI Auxiliary Oil Pump Motor Fire, LER 16-001-00 for Dresden Nuclear Power Station Regarding Alert Declared from Unit 3 HPCI Auxiliary Oil Pump Motor Fire 05000237/LER-2016-0022016-07-15015 July 2016 Unit 2 HPCI Inlet Steam Drain Pot Piping Leak Resulting in HPCI Inoperability, LER 16-002-00 for Dresden Nuclear Power Station, Unit 2 Regarding HPCI Inlet Steam Drain Pot Piping Leak Resulting in HPCI Inoperability 05000237/LER-2016-0012016-04-0808 April 2016 Secondary Containment Differential Pressure Transient, LER 16-001-00 for Dresden Nuclear Power Station, Unit 2 and 3 Regarding Secondary Containment Differential Pressure Transient 05000249/LER-2015-0012016-01-22022 January 2016 Main Steam Line Flow Switches Found Outside Tech Spec Allowed Value, LER 15-001-01, Dresden, Unit 3 Main Steam Line Flow Switched Found Outside Tech Spec Allowed Value ML1015505692009-07-24024 July 2009 Letter LS-AA-1 25-1001, Elevated Tritium Values Identified in 2 Storm Drains Due to Through-Wall Leaks in Underground Piping for Dresden Units 2 and 3. ML0602404072004-05-0505 May 2004 Final Precursor Analysis - Dresden Unit 3 - Unit 3 Scram Due to Loss of Offsite Power and Subsequent Inoperability of the Standby Gas Treatment System for Units 2 and 3 ML0234501722002-12-0202 December 2002 LER 89-029-05, Dresden Unit 2, Elevated HPCI Discharge Piping Temperature Due to Reactor Feedwater System Back Leakage ML17252B5391981-06-10010 June 1981 LER 81-034/03L-0 for Dresden, Unit 2 Re J-3 CRD Was Tested for Overtravel. the Overtravel Alarm Came Up and Rod Position Indication Was Lost ML17252B5411981-06-0808 June 1981 LER 81-030/01T-0 for Dresden, Unit 2 Re Operator Noticed the HPCI Steam Line Was Cold and Filled with Water. the HPCI System Was Declared Inoperable ML17252B5401981-06-0505 June 1981 LER 81-024/03L-0 for Dresden, Unit 2 Re While Conducting Routine Review of Surveillance, It Was Found That the Monthly Surveillance Required by Tech. Spec. 4.4.c.1 for Standby Liquid Control Tank Sampling Was Three Days Overdue ML17252B5271981-06-0303 June 1981 LER 81-023/03L-0 for Dresden, Unit 2 Re While Running HPCI Pump Test, the Operator Observed Torus Wide Range Level Instrument Reading 0 Inches and Narrow Range Level Instrument Reading, About - 2 Inches and Oscillating ML17252B5281981-04-14014 April 1981 LER 81-015/03L-0 for Dresden, Unit 2 Re the Fire Marshall Discovered That the Heat Detectors Were Not Functionally Tested within the 6 Month Interval Required by Tech. Spec. 4.12.A.1 ML17252B5291981-03-23023 March 1981 LER 81-011-03L-0 for Dresden, Unit 2 Re Observed Inadequacies in the Implementation of Administrative or Procedural Controls Which Threaten to Cause Reduction of Degree of Redundancy Provided in Reactor Protection Systems or Engineered Safe ML17252B5301981-02-20020 February 1981 LER 81-080/031-0 for Dresden, Unit 2 Re Abnormal Degradation of Systems Other than Those Specified in Item B.1.e Above Designed to Contain Radioactive Material Resulting from the Fission Process ML17252B5321981-02-17017 February 1981 LER 81-070/03L for Dresden, Unit 2 Re While Conducting Steam Line Area Temperature Switch Calibration Surveillance, Temperature Switch TS 2-261-180 Failed to Trip ML17252B5331981-02-0202 February 1981 LER 81-020/03L-0 for Dresden, Unit 2 Re During Refueling Outage LPCI Test Spray Valve Failed to Shut Electrically, One Set of Thermal Contacts Would Not Reset ML17252B5341981-01-23023 January 1981 LER 80-049/03L-0 for Dresden, Unit 2 Re at Completion of Secondary Containment Leak Rate Test the Reactor Building Ventilation Isolation Valve 2A-5742 Would Not Open ML17252B5351981-01-23023 January 1981 LER 80-050/03L-0 for Dresden, Unit 2 Re the 1501-62D Pressure Switch Actuated at 44.2 Inches of Water Increasing, Tech. Spec. Limit Is Less than or Equal to 41.6 Inches of Water Increasing ML17252B5371980-12-30030 December 1980 LER 80-046-00 for Dresden 2 Regarding a Failure to Receive CRD Volume During a Reactor Scram ML17252B5361980-12-29029 December 1980 LER 80-046/03L-0 for Dresden, Unit 2 Re Conditions Leading to on Operation in a Degraded Mode Permitted by a Limiting Condition for Operation or Plant Shutdown Required by a Limiting Condition for Operation ML17252B5381980-12-0303 December 1980 LER 80-043/03L-0 for Dresden, Unit 2 Re Surveillance Testing of the Condenser Low Vacuum Switches, Pressure Switch 2-503D Was Found to Have a Trip Point of 22.85 Inches Hg ML17252B5421980-11-12012 November 1980 LER 80-042/03L-0 for Dresden, Units 2 and 3 Re That Technical Staff Personnel Revealed That the Ultrasonic Testing of the Scram Discharge Volume Was Not Performed During the 1500-2300 Shift ML17252B5431980-11-12012 November 1980 LER 80-041/03L-0 for Dresden, Unit 2 Re Reactor Building Vent and Refueling Floor Rad Monitor Surveillance (DCP 2007-12) Was Performed ML17252B1841980-10-31031 October 1980 LER 80-40/01T-0 for Dresden Unit 3 Regarding Failure of the HPCI Inboard Steam Supply Valve to Open ML17252B5441980-10-29029 October 1980 LER 80-039/03X-1 for Dresden, Unit 2 Re HPCI Inboard Steam Supply Valve Opened and Had a Dual Light Indication ML17252B5451980-10-23023 October 1980 LER 79-017-01X2 for Dresden, Units 2 and 3 Re Inadequacies Observed in Implementation of Administrative Procedural Controls Which Threaten to Cause Reduction of Degree of Redundancy Provided in Reactor Protection System or Engineered Safety ML17252B5461980-10-0808 October 1980 LER 80-016/01T-1 for Dresden, Unit 2 Re While Shutdown and Conducting 300 Psi Primary System Hydrostatic Leak Inspection, Water Discovered Dripping from LPCI Check Valve Up Stream Drain Line ML17252B5471980-10-0707 October 1980 LER 80-036/03L-0 for Dresden, Unit 2 Re Discovery That the Six Month Functional Test Required by Tech. Spec. 4.12.A.1 of the Cardox System Heat Detectors Was Not Performed within the Allowable Surveillance Interval ML17252B5481980-10-0101 October 1980 LER 80-033/03L-0 for Dresden, Unit 2 Re the Frequency of Water Addition to the Torus Increased. Leak Rate Tests Indicated a Leak in the 2B LPCI Heat Exchanger ML17252B1861980-08-22022 August 1980 LER 80-032/03L-0 for Dresden Unit 3 Regarding During Normal Operations While Performing APRM Rod Block and Scram Functional Tests, APRM 3 Failed to Initiate a Rod Block ML17252B5501980-08-19019 August 1980 LER 80-027/03L-0 for Dresden, Unit 2 Re Dropping Load in Preparation for I.E.B. 80-17 Scram Tests, Low Level of Diesel Generator Coolant Present in the Water Tank Was Noticed by Equipment Operator ML17252B5161980-08-14014 August 1980 LER 80-024/03L-0 for Dresden, Unit 2 Re Drywell Oxygen Concentration Indication Failed Downscale in Violation of Tech. Spec. 3.7.A.6.d ML17252B1881980-08-0404 August 1980 LER 80-030/03L-0 for Dresden Unit 3 Regarding While Reducing Power for Manual Scram Required by I.E. Bulletin 80-17, Routine Surveillance of MSIV Closure Times Showed That MSIVs 3-203-2B and 2D Closed in Less than 3 Seconds ML17252B1911980-06-13013 June 1980 LER 80-024/03L-0 for Dresden Unit 3 Regarding During Normal Operation with Reactor Level Instrument Surveillance in Progress, Level Switch LIS 3-263-58AB Tripped at Less than Tech. Spec. Limit of 84 Inches ML17252B1901980-06-13013 June 1980 LER 80-023/03L-0 for Dresden Unit 3 Regarding During Normal Operation with Reactor Level Instrument Surveillance in Progress, Level Switch LIS 3-263-58A Tripped at Less than Tech. Spec. Limit of 144 Inches ML17252B1921980-06-10010 June 1980 LER 80-025/03L-0 for Dresden Unit 3 Regarding During Normal Operation While Performing Main Steam Line High Flow Isolation Surveillance, Flow Switch DPIS 3-261-2A Was Out of Tech. Spec. Limits, Table 3.2.1 ML17252B5521980-06-0505 June 1980 LER 80-017/03L-0 for Dresden, Unit 2 Re Double Valve Indication Was Observed on the HPCI Inboard Isolation Valve. Upon Trying to Cycle the Valve, the Valve Failed to Close ML17252B1931980-06-0303 June 1980 LER 80-002/03X-1 for Dresden Unit 3 Regarding During Normal Operation While Performing LPCI Monthly Surveillance, MO-3-1501-118 Valve Would Not Close ML17252B5531980-05-27027 May 1980 LER 80-015/03L-0 for Dresden, Unit 2 Re During Steady State Operation, Torus to Drywell Vacuum Breaker 2-1601-33A Was Found Inoperable During Routine Operability Surveillance ML17252B1941980-05-27027 May 1980 LER 80-022/03L-0 for Dresden Unit 3 Regarding During Normal Shutdown Conditions, IRM 12 Downscale Rod Block Was Found to Trip at 4/125 ML17252B1951980-05-0808 May 1980 LER 80-021/01T-0 for Dresden Unit 3 Regarding During Startup Following Refueling Outage, Combinations of 3A Target Rock Valve and 3B, 3C and 3E Electromatic Relief Valves Failed to Open at Rated Reactor Pressure While Performing Operability ML17252B1961980-04-29029 April 1980 LER 80-017/03L-0 for Dresden Unit 3 Regarding During Refueling Outage, While Conducting Operability Surveillance Test, U-3 Diesel Generator Failed to Start 2020-05-14
[Table view] Category:Response to Request for Additional Information (RAI)
MONTHYEARRS-23-119, 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 RS-23-041, Response to Request for Additional Information for Alternative Request RV-23H2023-03-14014 March 2023 Response to Request for Additional Information for Alternative Request RV-23H RS-23-004, Response to Request for Additional Information Regarding Transition to GNF3 Fuel License Amendment Request2023-01-23023 January 2023 Response to Request for Additional Information Regarding Transition to GNF3 Fuel License Amendment Request RS-22-117, Response to Request for Additional Information License Amendment Request Regarding New Fuel Storage Vault and Spent Fuel Storage Pool Criticality Methodologies2022-11-0303 November 2022 Response to Request for Additional Information License Amendment Request Regarding New Fuel Storage Vault and Spent Fuel Storage Pool Criticality Methodologies RS-22-101, Response to Request for Additional Information Related to Relief Request I6R-01 Associated with the Sixth Inservice Inspection Interval2022-08-10010 August 2022 Response to Request for Additional Information Related to Relief Request I6R-01 Associated with the Sixth Inservice Inspection Interval RS-22-027, Constellation, Response to Request for Additional Information Regarding Application to Revise Technical Specifications to Adopt TSTF-541 Revision 2, Add Exceptions to Surveillance Requirements for Valves and Dampers Locked in the Actuated2022-02-23023 February 2022 Constellation, Response to Request for Additional Information Regarding Application to Revise Technical Specifications to Adopt TSTF-541 Revision 2, Add Exceptions to Surveillance Requirements for Valves and Dampers Locked in the Actuated P NMP1L3447, Constellation Energy Generation, LLC - Response to Request for Additional Information - Proposed Relief Request Associated with Reactor Pressure Vessel Water Level Instrumentation Partial Penetration Nozzle Repairs2022-02-0202 February 2022 Constellation Energy Generation, LLC - Response to Request for Additional Information - Proposed Relief Request Associated with Reactor Pressure Vessel Water Level Instrumentation Partial Penetration Nozzle Repairs JAFP-21-0087, Response to Request for Additional Information Regarding Request for Approval of Transfer of Licenses and Conforming Amendments2021-09-16016 September 2021 Response to Request for Additional Information Regarding Request for Approval of Transfer of Licenses and Conforming Amendments RS-21-078, Response to Request for Additional Information for Application to Revise Technical Specification to Adopt TSTF-582, Reactor Pressure Vessel Water Inventory Control (RPV WIC) Enhancements, and TSTF-583-T, TSTF-582 Diesel2021-08-19019 August 2021 Response to Request for Additional Information for Application to Revise Technical Specification to Adopt TSTF-582, Reactor Pressure Vessel Water Inventory Control (RPV WIC) Enhancements, and TSTF-583-T, TSTF-582 Diesel JAFP-21-0044, Response to Request for Additional Information Regarding Request for Approval of Transfer of Licenses and Conforming Amendments2021-06-11011 June 2021 Response to Request for Additional Information Regarding Request for Approval of Transfer of Licenses and Conforming Amendments JAFP-21-0032, Response to Request for Additional Information - Proposed Alternative Concerning ASME Section XI Repair/Replacement Documentation for Replacement of Pressure Retaining Bolting2021-04-20020 April 2021 Response to Request for Additional Information - Proposed Alternative Concerning ASME Section XI Repair/Replacement Documentation for Replacement of Pressure Retaining Bolting RS-21-036, Response to Request for Additional Information Re Dresden License Amendment Request - Proposed Defueled Technical Specifications and Revised License Conditions for Permanently Defueled Condition2021-03-16016 March 2021 Response to Request for Additional Information Re Dresden License Amendment Request - Proposed Defueled Technical Specifications and Revised License Conditions for Permanently Defueled Condition RS-21-029, Response to RAI Re License Amendment Request - Proposed Changes to Byron & Dresden Stations Emergency Plan for Post-Shutdown and Permanently Defueled Condition2021-03-15015 March 2021 Response to RAI Re License Amendment Request - Proposed Changes to Byron & Dresden Stations Emergency Plan for Post-Shutdown and Permanently Defueled Condition RS-20-112, Response to Request for Additional Information Related to License Amendment Request to Adopt TSTF-5682020-09-0303 September 2020 Response to Request for Additional Information Related to License Amendment Request to Adopt TSTF-568 RS-20-038, Response to Request for Additional Information for the License Amendment Request to Change Technical Specifications to Increase Allowable MSIV Leakage Rates and Revise Secondary Containment Surveillance2020-05-0606 May 2020 Response to Request for Additional Information for the License Amendment Request to Change Technical Specifications to Increase Allowable MSIV Leakage Rates and Revise Secondary Containment Surveillance RS-19-112, Response to Request for Additional Information Regarding License Amendment Request to Revise Technical Specifications Following Adoption of TSTF-542, Reactor Pressure Vessel Water Inventory Control2019-12-0505 December 2019 Response to Request for Additional Information Regarding License Amendment Request to Revise Technical Specifications Following Adoption of TSTF-542, Reactor Pressure Vessel Water Inventory Control RS-19-101, Seismic Probabilistic Risk Assessment Report, Response to NRC Request for Information Pursuant to 10 CFR 50.54(f) Regarding Recommendation 2.1 of the Near-Term Task Force Review of Insights From...2019-10-30030 October 2019 Seismic Probabilistic Risk Assessment Report, Response to NRC Request for Information Pursuant to 10 CFR 50.54(f) Regarding Recommendation 2.1 of the Near-Term Task Force Review of Insights From... JAFP-19-0057, Response to Request for Additional Information - Proposed Alternative to Utilize Code Cases N-878 and N-8802019-06-0404 June 2019 Response to Request for Additional Information - Proposed Alternative to Utilize Code Cases N-878 and N-880 NMP1L3279, Response to Request for Additional Information - Proposed Alternatives to Utilize Code Cases N-878 and N-880 for Plants2019-05-0101 May 2019 Response to Request for Additional Information - Proposed Alternatives to Utilize Code Cases N-878 and N-880 for Plants JAFP-19-0006, Response to Request for Additional Information - Proposed Alternative to Utilize Code Cases N-878 and N-8802019-01-0808 January 2019 Response to Request for Additional Information - Proposed Alternative to Utilize Code Cases N-878 and N-880 RS-18-143, Supplemental Information Supporting License Amendment Requests for Approval of Changes to Emergency Plan Staffing Requirements2018-11-29029 November 2018 Supplemental Information Supporting License Amendment Requests for Approval of Changes to Emergency Plan Staffing Requirements RS-18-097, Response to Request for Additional Information and Supplemental Information-License Amendment.2018-07-27027 July 2018 Response to Request for Additional Information and Supplemental Information-License Amendment. RS-18-061, Response to Request for Additional Information Regarding Decommissioning Funding Plans for Independent Spent Fuel Storage Installations (Isfsis)2018-05-0202 May 2018 Response to Request for Additional Information Regarding Decommissioning Funding Plans for Independent Spent Fuel Storage Installations (Isfsis) RS-18-022, Response to NRC Request for Additional Information for Request for License Amendment to Revise Technical Specification 5.5.12, Primary Containment Leakage Rate Testing Program.2018-02-14014 February 2018 Response to NRC Request for Additional Information for Request for License Amendment to Revise Technical Specification 5.5.12, Primary Containment Leakage Rate Testing Program. RS-17-088, Response to March 12, 2012, Request for Information Enclosure 2, Recommendation 2.1, Flooding, Required Response 3, Flooding Integrated Assessment Submittal2017-09-0808 September 2017 Response to March 12, 2012, Request for Information Enclosure 2, Recommendation 2.1, Flooding, Required Response 3, Flooding Integrated Assessment Submittal RS-17-086, Supplement to Application to Revise Technical Specifications to Adopt TSTF-542, Reactor Pressure Vessel Water Inventory Control.2017-07-13013 July 2017 Supplement to Application to Revise Technical Specifications to Adopt TSTF-542, Reactor Pressure Vessel Water Inventory Control. RS-17-044, Response to Request for Additional Information Regarding Proposed Alternative for the Use of Encoded Phased Array Ultrasonic Examination Techniques in Lieu of Radiography2017-03-13013 March 2017 Response to Request for Additional Information Regarding Proposed Alternative for the Use of Encoded Phased Array Ultrasonic Examination Techniques in Lieu of Radiography RS-16-244, Additional Information Regarding Request for License Amendment to Address Secondary Containment Access Openings2016-12-12012 December 2016 Additional Information Regarding Request for License Amendment to Address Secondary Containment Access Openings RS-16-124, Spent Fuel Pool Evaluation Supplemental Report, Response to NRC Request for Information Pursuant to 10 CFR 50.54(f) Regarding Recommendation 2.1 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident2016-08-31031 August 2016 Spent Fuel Pool Evaluation Supplemental Report, Response to NRC Request for Information Pursuant to 10 CFR 50.54(f) Regarding Recommendation 2.1 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident RS-16-166, Supplemental Information Related to Inservice Inspection Interval Relief Request I5R-082016-08-22022 August 2016 Supplemental Information Related to Inservice Inspection Interval Relief Request I5R-08 RS-16-159, County, and Quad Cities - Response to Request for Additional Information Regarding License Amendment to Address Secondary Containment Access Openings2016-07-28028 July 2016 County, and Quad Cities - Response to Request for Additional Information Regarding License Amendment to Address Secondary Containment Access Openings RA-16-049, Response to Request for Additional Information Regarding Requests to Withhold Emergency Preparedness Documents from Public Disclosure2016-05-26026 May 2016 Response to Request for Additional Information Regarding Requests to Withhold Emergency Preparedness Documents from Public Disclosure RS-16-105, Supplemental Information in Support of Request for License Amendment Regarding Transition to Areva Fuel2016-04-26026 April 2016 Supplemental Information in Support of Request for License Amendment Regarding Transition to Areva Fuel RS-16-077, Response to Request for Additional Information Regarding Request for License Amendment to Reduce the Reactor Steam Dome Pressure Specified in the Technical Specification 2.1.1, Reactor Core Sls.2016-04-14014 April 2016 Response to Request for Additional Information Regarding Request for License Amendment to Reduce the Reactor Steam Dome Pressure Specified in the Technical Specification 2.1.1, Reactor Core Sls. RS-16-018, Response to Request for Additional Information Regarding Request for License Amendment Regarding Transition to Areva Fuel2016-01-28028 January 2016 Response to Request for Additional Information Regarding Request for License Amendment Regarding Transition to Areva Fuel RS-15-258, County, and Quad Cities - Response to Request for Additional Information Regarding License Amendment Request Adding Technical Specification for Inservice Leak and Hydrostatic Testing Operation2015-09-29029 September 2015 County, and Quad Cities - Response to Request for Additional Information Regarding License Amendment Request Adding Technical Specification for Inservice Leak and Hydrostatic Testing Operation RS-15-185, County - Supplemental Information Regarding Seismic Hazard Risk Evaluation and Seismic Hazard Prioritization Results - Response to NRC Request for Information Pursuant to 10 CFR 50.54(f) Regarding Recommendation 2.1 of the Near.2015-07-22022 July 2015 County - Supplemental Information Regarding Seismic Hazard Risk Evaluation and Seismic Hazard Prioritization Results - Response to NRC Request for Information Pursuant to 10 CFR 50.54(f) Regarding Recommendation 2.1 of the Near. ML17252B1311978-09-0101 September 1978 Letter Regarding Supplemental Response to IE Inspection Report 05000249/1978012 - Dresden Unit 3 IR 05000249/19780121978-09-0101 September 1978 Letter Regarding Supplemental Response to Ie Inspection Report 05000249/1978012 - Dresden Unit 3 ML17252B1381976-11-0808 November 1976 Letter Regarding Response to the Item Noted Under Enforcement Action in Inspection Report 05000249/1976020 - Dresden Unit 3 ML17252B1451976-06-28028 June 1976 Letter Regarding 06/14/1976 Letter Concerning Actions Taken in Response to Request for Commitments in Areas of Concern - Dresden Unit 3 and IE Inspection Report 05000249/1976007 ML17252B1571975-10-16016 October 1975 Letter Referring to 09/19/1975 Letter Informing of Steps Taken to Correct Items of Noncompliance Brought to Attention in 08/29/1975 Letter - Dresden Units 2 and 3 ML17252B0491974-01-0707 January 1974 Answers to AEC Questions on Quality Assurance for Operations - Dresden Unit 3 ML17238A7511973-08-31031 August 1973 Quad-Cities Stations Units 1 and 2 - 10 CFR 50, Appendix J ML17238A7351972-04-26026 April 1972 Additional Information to Final Report of Safety Valve Operation Following a Feedwater Transient - Dresden Unit 3 ML17238A7491972-02-22022 February 1972 Hydrogen Control Equipment - Dresden Units 2 and 3 ML17238A7151971-09-13013 September 1971 Performance of the Dresden Units 2 and 3 Emergency Core Cooling Systems ML17252A4281966-05-26026 May 1966 Letter Concerning Amendments Nos. 2 and 3 to the Unit 3 Plant Design and Analysis Report ML17252A4311966-05-17017 May 1966 Response to the 5/12/1966 Letter Requesting Additional Information and Amendment No. 2 to the Unit 3 Plant Design and Analysis Report 2023-03-14
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Commonwealth Edison Company ONE Fl.RST NATIONAL PLAZA* CHICAGO, ILLINOIS Address Reply to, POST OFFICE BOX 767 *CHICAGO, ILLINOIS 60690 April 26, Mr. Donald* J. Skovholt Assistant Director for Reactor Operations Division of Reactor Licensing U.S.* Atomic Energy Commission
.Washington, D.C. 20545 5 o- 2 4 I
Subject:
Additional Information to Final Report of Safety Valve Operation following a Feedwater Transient - Dresden Unit 3 (DPR-25)
Dear Mr. Skovholt:
The attached additional information is in response to your letter of March*29, 1972 concerning the December 8, 1971 incident at Dresden Unit 3.
In addition to one signed original, 39 copies of this, informatic:m,,are ali;;<:> 1:submitted. Should you desire any additional information, please let us know.
~ry ::lyl_Ci.
~lLee, Jr.-
Assistant to the 0 1 Preside!liii-:!!~.....,
(9 ).>"'~">ll.i4.M~
cc: Boyce H. Grier, Region III 233.2
ENCLOSURE Question 1:
What is the approximate number and nature of grounds that developed in the i2s volt de systems? What annunciations we~e lost or were erroneous as a result of the grounds? Is the information provided by the annunciator panel also provided elsewhere in the control room?
Do you have procedures for required action following loss of annunciator panels?
RESPONSE
There were five grounds experienced during the December 8, 1971 incident. These were:
1 - When the "3A" electromatic valve was damaged by the steam jet from the "3F" safety valve the damage to the control wiring resulted in a ground on the 125VDC operating power supply and on the annunciator power supply. This damaged circuitry caused the fuses in the power supply to the 903-3 annunciator panels to blow.
2 - Grounds were found on the LPCI manually operated maintenance isolation valves 26A and 26B indication circuits inside the primary containment. These grounds were caused by moisture and cleared when the indicating switches were dried out.
3 - Grounds were found in the annunciator power supply for "3C" and "3D" electromatic va.lves. These grounds were caused by moisture and cleared when switches were dried out.
Erroneous annunciation developed cin electromatic valves "3A", "3C",
and "3D" indicating that they were open when, in fact, they were closed. The damage to the "3A" valve referenced above also resulted
in the valve indicating open; valve position indication for "3C" and "3D" valves showed th.em to be closed. When the fuse blew, all annunciation for panel 903-3 was lost with the exception of I
window 124 which signalled the loss of power. These annunciations i
were only lost for the short periods of time necessary to install replacement fuses. The majority of the information conveyed by these annunciators can be obtained from other indications in the control room. The remaining information can be obtained locally, either through observation of local instruments or through ,
observation of the individual systems.
Procedures for loss of an annunciator panel address themselves to prompt restoration of the panel rather than to an elaborate and time consuming verification of the individual pieces of information.
Enough information is available in the control room to monitor system performance until the~ annunciator panel is restored to operation.
The existing procedures will be revised, however, to ensure that these indications are monitored during this period when annunciators are unavailable.
Question 2:
.What other functions could be lost by similar grounds in the de systems (125 volt and 250 volt)? What is the relative susceptibility of these systems to similar types of grounds?
RESPONSE
There are no 250 VDC services brought into the drywell and, therefore, the 250 VDC system is not susceptible to the drywell environment.
In addition to the Automatic Blowdown System (ABS) , the 125 VDC system is used as a power supply for:
1 - Various (LPCI, Core Spray, etc) check valve position indication circuits.
2 - Various (LPCI, Core Spray, Standby Liquid) manually operated maintenance shutoff valves position indication switches.
3 - The DC solenoid operated pilot ~alve for the MSIV's.
4 - Drywell floor drain sump level alarms.
5 - Drywell equipment drain sump level alarms.
6 - Drywell equipment drain sump temperature alarms.
The position indication switches on the valves and the DC solenoid operated pilot valves on the MSIV's are designed to withstand the environment during pla.nt operation. The position* switches and solenoids are the totally *enclosed type and are designed to operate in a 1000/o relative humidity environmrmt. Physical damage can be done to the equipment which could result in a circuit ground, short or open. These conditions are protected against by properly
4 protecting these circuits with fuses and circuit breakers.
The DC solenoid operated pilot valves on the MSIV's are normally energized _and a failure would be in the direction to de-ene+gize the solenoid and to close the MSIV.
The drywell sump alarm contacts are also totally enclosed construction and are designed to operate in 100% relative humidity.
Question 3:
What is the effect on ECCS systems in the event of grounds in the de supply in the drywell?
RESPONSE
The General Electric Company document NED0-10139 dated June, 1970 titled "Compliance of Protection Systems to Industry Criteria:
General Electric BWR Nuclear Steam Supply System" among other things discusses the effect of shorts, opens and grounds on the DC supply systems to the ECCS systems.
Section 3 of this document provides information for each ECCS system.
The paragraphs titled Equipment Qualification, Channel Integrity, Channel Independence, Control and Protection Interaction and Environmental Considerations for e_ach ECCS system discusses the effect of DC supply faul~s on each ECCS system.
The paragraph concerning control and protection interaction states that annunciator circuits using contacts of sensor relays or logic
relays c~nnot impair operability of the system control because of the electrical separation between controls of the two subsystems.
The grounds oh the DC supply to annunciator panel 903-3 experienced in the December 8, 1971 incident on Unit 3 at Dresden substantiates this conclusion since all ECCS systems except the ABS performed normally during this incident.
The 1
' 3A" ~lectromatic valve solenoid operation was damaged by* the steam jet from the "F" safety valve and, therefore, the ABS system was not completely operable. The remaining four electromatic valves were tested and found to be operable after this incident occurred. Only four electromatic valves are required for adequate ABS operation.
Question 4:
Are changes planned to further protect these de systems from similar kinds of failure?
RESPONSE
Subsequent to the Dresden December 8, 1971 incident the '1 3F" safety valve discharge "rams horn" was repositioned to direct the opening away from the "3A" electromatic relief valve solenoid operator.
The other safety valve discharge paths were inspected to verify that they were not directed towards an electromatic valve operator.
This should preclude any damage to an electromatic valve from the discharge of a saf~ty valve.
The other equipment in the drywell which gave evidence of grounds on the DC supply system was dried and cleaned.
The ABS ~ystem is the only system affected by grounds on its DC supply inside the drywell. A discussion of the operation of the ABS system with various faults on the DC supply appears in NEDd-10139 paragraph 3.5.2.2 and concludes that any sirtjl.e fa.ilure of this nature could not disable more than one valve. This conclusion was also substantiated during the December 8, 1~71 incident at Dresden. The "3A" electromatic valve had been physically damaged and at least a partial ground existed on its associated wiring yet the remaining four valves were found to operate properly subsequent to the incident.
It is Edison's conclusion that the ECCS systems are not adversely affected by grounds on the DC supplies to components in the drywell because of proper circuit protection and separation practices -~ *.
incorporated in the Dresden installation.
While the separation criteria used in the Dresden installation do not separate the DC supply to the valve solenoids to the same extent as the system analyzed in NED0-10139 (i.e. the cables share penetrations and may share cable parts since all are in division I) the analysis is still valid where the faults applied to the DC solenoids are concerned.
Because the ECCS system operation was not de.graded by the appearance of grounds on the DC system in the drywell as experienced d~ring the December 8, 1971 incident, there are no further modifications planned for the DC sy8tem other than those outlined above which were complete prior to restoring the unit to service following the incident.
QUESTION 5:
As a result of the incident, have any changes been made or been considered in operating procedures, operator training or operator retraining?
RESPONSE
The existing procedures have been augmented by an Operating Order which amplifies the existing procedures and specifically addresses the tripping of the Reactor Feed Pumps .on high water level. This Operating Order is posted on the Control Panel in the Control Room.
In addition, a complete revision of the Operating Procedures is in the final review stages. This revision will incorporate the provisions of this Operating Ordt~r.
Operator training in proper response to transients such as occurred in this incident have been reviewed by all Operating Shifts. In addition, operator retraining at the BWR Simulator has been implemented for all licensed personnel. At this time, 9 SRO's and 15 RO's have completed this retraining.
- a-Question 6:
What are your present instructions regarding manual vs autom,atic feedwater control?
RESPONSE
Chapter 32 of the Operating and Procedures states, as a criteria, the following:
A controller may be placed in the manual mode from the automatic mode whenever, in the judgement of the operator, continued automatic operation is unsafe, or when it may cause any unnecessary transients
- In addition, the Operating Order mentioned in the reply to Question 5 states:
1 - Run the Auto Level Set Point down* to +10 to +15".
2 - Leave the Feedwater Control System in the Automatic Mode.
3 - Close the low flow feedwater valve.
4 - If the water level continues to increase past the high level alarm point at +44", trip all Reactor Feed Pumps immediately.
Question 7:
With reference to your evaluation of modifications to the feedwater control sys tern in pa.r agraph E .1 7 of your f ina 1 report , provide the following information:
- a. What is the current status of your evaluation?
- b. Provide the proposed details of the reactor feedwater pump high reactor water level trip and the bases for selecting the trip water level.
- c. 'Does elimination of the run out circuitry remove overload protection from the feedwater pump.and motor drive?
RESPONSE
- a. Status (Numbers Reference Report)
(17.a) The design of the high reactor water level reactor feedwater pump trip is complete and installation in accordance with this design has been approved by the Dresden Station Review Board. Installation will be completed during the first convenient unit outage.
(17.b) An additional pipe anchor for the start-up feedwater regulator valve piping has been designed and approved for installation. Installation will be completed during the first convenient unit outage.
(17.c) The alternate type valve positioner will be tested as part of a feedwater control system improvement program. The purpose of th.is program is increasing the stable response rate of the feedwater control system.
The program consists of four parts. The first part of the program consists of tests of the existing Rystem with the intent of isolat frig th1:?. cause of the tendency to instability at high cont~ol system response rates. The last three part.:.:; yf the progrc.!M involve adding damping to the existing valve positioner, changing to a different type positioner, r:i.'1d C.":.ddin:~.: lag-lc2d compensation to the
control loop. Each part of the program involves extensive testing of the installed control system with the unit operating. The four parts will be carried out sequentially until the required system perfr>rmance is accomplished. The de?tailed test procedure has been prepared and is currently being reviewed. It is presently scheduled to begin testing by May 1, 1972.
(17.d) The improved amplifier will be installed at the first convenient outage.
(17.e) Feedwater regulator valve position indication will be installed in the control room during the first convenient outage after materials are available.
(17.f) Although the feedwater system is not directly a safety system, a.n evaluation of elimination of "Run-out" is being performed on the basis that it may provide safe-guard protection which compliments the required ECC systems. The evaluation currently in progress includes predicting the performance of the feedwater system with and without the run-out circuit in the event of a small sized loss of coolant.
- b. The reactor feed pump trip initiation signal is taken from the turbine trip contacts set at reactor water level 48 inches.
Through appro?riate relaying, the pumps will be tripped by a
~ . ....
e DRE St> EN ST"TtON ,
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u M ,.,. .3 RSAC:TOl't Flt&bWAT'Soff Pl) MP tl ' i'°' REU\c:Toet WAT'IE!ir 1..tn S. lo. T flt.\ P P~S\.111-llN"lf\i' t>S6t~N I l ~t.OSfZ .S o/IJ ~ L6111f1-
~ 48 * /Nellt!.S
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two out of two logic circuit °(an electrical schematic is (l'; VI ;l'.) ; 1\
attached}. The level set point was selected the same as the
. .. ~"'j)\
~o ~ ~ra g ;\q,\ ~' .
turbine trit.p .. and' high level isolation, becaus~ isolation.
1
.. ..... ; .** "~-:,~:-i 4'!'~~*11t:t\.a.u . f . t a'nd *. t u.rb **ine.¥1.'f~~J::l1.e:L~~I}. 1a e maJor 1 oss o primary sys em
.. cmt~:ii ~1.iiw~~1fa.4.1ro1 water inveneory and the need for feedwater. This level setpoint was also selected to terminate a high level transient before water ~eached the HPCI, isolation condenser or main steam nozzles.*
- c. No. if run-out is ,eliffiinated, the pump and motor protection will remain without change.
Question 8:
.'.What are your plans for the study of the phenomena which caused the premature lifting of the safety.valves?
RESPONSE
The* preliminary program inc.ludes three parts: Monitoring the valves in place on Dresden Unit 3, mathematical modeling/analysis of the valves and primary system, and full sc;:a);e,:*.,.)...aboratory testing *
.' ;:; j, >;:~~:.~';;'*;*:~,~r ,\ *;! ";, .
- Negotiations are currently in progr~~*s- with v'a;r.:!,.qus corfsul tarits
. . ~* ..*):' "-' )~~. :/~ . ' ; :** '>.. .
to provide .the last two phases of. the program~. ;t:~\ is planned to
. . i \ ,, .;'\:... ~ *<;~* *. *\
monitor pressure transducers instal'led *for reactor.* dynamics testing Jr',.* : * *' I' I.~ , ~ ' *
.,1
. . . . ..,1\ . .
and temperature during flooding o( the .*ma'.i'n. s.~eaJlf lines at the. next
. '. ~
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scheduled outage on Dresden Unit 3.
2 3 31-A
RECEIVED 1972 APR 2 PM 4 ;6 U.S.A TOt!IC ENERGY COMM*
" . RtGUL "'i"Y *
'MAlk&.. RECOROSSECTIO~
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