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Category:Letter
MONTHYEARNMP1L3622, Request for Exemption from Certain Requirements of 10 CFR 72.212 and 10 CFR 72.2142025-01-30030 January 2025 Request for Exemption from Certain Requirements of 10 CFR 72.212 and 10 CFR 72.214 NMP1L3618, CFR 50.46 Annual Report2025-01-27027 January 2025 CFR 50.46 Annual Report ML25022A2402025-01-22022 January 2025 Request for Exemption from Certain Requirements of 10 CFR 72.212 and 10 CFR 72.214 for Nine Mile Point Nuclear Station - Holtec HI-STORM FW Ad HI-TRAC Vw IR 05000220/20254032025-01-16016 January 2025 Information Request for the Cyber Security Baseline Inspection, Notification to Perform Inspection 05000220/2025403 and 05000410/2025403 ML24353A1372025-01-15015 January 2025 Proprietary Determination Constellation Energy Generation, LLC 2024 Deferred Premiums 05000410/LER-2024-002-01, Supplement to NMP2 Licensee Event Report 2024-002-00, Automatic Reactor Scram on Turbine Trip Due to Failed Breaker2025-01-10010 January 2025 Supplement to NMP2 Licensee Event Report 2024-002-00, Automatic Reactor Scram on Turbine Trip Due to Failed Breaker ML24358A1832025-01-0707 January 2025 Issuance of Relief Proposed Alternative Request Associated with Ultrasonic Examination of Reactor Pressure Vessel Circumferential Shell Welds ML24344A2742024-12-19019 December 2024 Alternative Request to Use American Society of Mechanical Engineers Boiler and Pressure Vessel Code Case OMN-32 ML24339B7292024-12-18018 December 2024 Amd - Constellation - Adoption of TSTF-591 ML24331A2592024-11-27027 November 2024 Reassignment of the U.S. Nuclear Regulatory Commission Branch Chief in the Division of Operating Reactor Licensing for Plant Licensing Branch III ML24331A2792024-11-26026 November 2024 Supplement to Application to Revise Technical Specifications to Adopt TSTF-591-A, Revise Risk Informed Completion Time (RICT) Program Revision 0 and Revise 10 CFR 50.69 License Condition - Revise LaSalle, Units 1 and 2 Technical Specificati NMP1L3614, Response to Request for Additional Information for License Amendment Request to Adopt TSTF-230, Revision 1, Add New Condition B to LCO 3.6.2.3, RHR Suppression Pool Cooling2024-11-22022 November 2024 Response to Request for Additional Information for License Amendment Request to Adopt TSTF-230, Revision 1, Add New Condition B to LCO 3.6.2.3, RHR Suppression Pool Cooling 05000410/LER-2024-002, Automatic Reactor Scram on Turbine Trip Due to Failed Breaker2024-11-22022 November 2024 Automatic Reactor Scram on Turbine Trip Due to Failed Breaker IR 05000220/20244022024-11-20020 November 2024 Material Control and Accounting Program Inspection Report 05000220/2024402 and 05000410/2024402 (Cover Letter Only) IR 05000410/20240032024-11-0808 November 2024 Integrated Inspection Report 05000220/1014003 and 05000410/2024003 ML24317A1432024-11-0404 November 2024 Constellation Energy Generation, LLC, 2024 Annual Report - Guarantees of Payment of Deferred Premiums ML24268A3382024-10-16016 October 2024 Issuance of Amendment No. 253 Regarding the Modification of TS Surveillance Requirement 4.3.6.a Related to Adoption of TSTF-425, Revision 3 RS-24-093, Response to Request for Additional Information - Alternative Request to Utilize Code Case OMN-32, Alternative Requirements for Range and Accuracy of Pressure, Flow, and Differential Pressure Instruments Used in Pump Tests2024-10-10010 October 2024 Response to Request for Additional Information - Alternative Request to Utilize Code Case OMN-32, Alternative Requirements for Range and Accuracy of Pressure, Flow, and Differential Pressure Instruments Used in Pump Tests NMP2L2890, Submittal of Revision 26 to the USAR and Reference Figures, 10 CFR 50.59 Evaluation Summary Report, TS Bases, TRM Requirements Manual Changes, and 10 CFR 54.37(b) Aging Management Review (Excludes Attachment 6)2024-10-0404 October 2024 Submittal of Revision 26 to the USAR and Reference Figures, 10 CFR 50.59 Evaluation Summary Report, TS Bases, TRM Requirements Manual Changes, and 10 CFR 54.37(b) Aging Management Review (Excludes Attachment 6) ML24275A2442024-10-0303 October 2024 Reassignment of the U.S. Nuclear Regulatory Commission Branch Chief, Division of Operating Reactor Licensing IR 05000220/20243022024-10-0303 October 2024 Initial Operator Licensing Examination Report 05000220/2024302 ML24190A0012024-09-26026 September 2024 Issuance of Amendment Nos. 252 and 197 Regarding the Revision to Technical Specification Design Features Section to Remove Nine Mile Point Unit 3 Project Designation NMP1L3608, Supplemental Information Letter No. 3 - Revision to the Technical Specifications Design Features Sections to Remove the Nine Mile 3 Nuclear Project, LLC, Designation2024-09-20020 September 2024 Supplemental Information Letter No. 3 - Revision to the Technical Specifications Design Features Sections to Remove the Nine Mile 3 Nuclear Project, LLC, Designation RS-24-090, Response to Request for Additional Information - Relief Request Concerning Extension of Permanent Relief from Ultrasonic Examination of Reactor Pressure Vessel Circumferential Shell Welds2024-09-12012 September 2024 Response to Request for Additional Information - Relief Request Concerning Extension of Permanent Relief from Ultrasonic Examination of Reactor Pressure Vessel Circumferential Shell Welds ML24249A1362024-09-0404 September 2024 EN 57304 - Westinghouse Electric Company, LLC, Final Report - No Embedded Files. Notification of the Potential Existence of Defects Pursuant to 10 CFR Part 21 IR 05000220/20240052024-08-29029 August 2024 Updated Inspection Plan for Nine Mile Point Nuclear Station, Units 1 and 2 (Report 05000220/2024005 and 05000410/2024005) IR 05000220/20240102024-08-22022 August 2024 Age-Related Degradation Inspection Report 05000220/2024010 and 05000410/2024010 NMP1L3603, Submittal of Preliminary Decommissioning Cost Estimate and Irradiated Fuel Management Plan2024-08-20020 August 2024 Submittal of Preliminary Decommissioning Cost Estimate and Irradiated Fuel Management Plan ML24222A6772024-08-0909 August 2024 Response to Request for Additional Information for Application to Revise Technical Specifications to Adopt TSTF-591-A, Revise Risk Informed Completion Time (RICT) Program Revision 0 and Revise 10 CFR 50.69 License Condition IR 05000220/20240022024-08-0505 August 2024 Integrated Inspection Report 05000220/2024002 and 05000410/2024002 ML24215A3002024-08-0202 August 2024 Operator Licensing Examination Approval ML24213A1412024-07-31031 July 2024 Requalification Program Inspection NMP1L3601, Supplemental Information Letter No. 2 - Revision to the Technical Specifications Design Features Sections to Remove the Nine Mile 3 Nuclear Project, LLC, Designation2024-07-31031 July 2024 Supplemental Information Letter No. 2 - Revision to the Technical Specifications Design Features Sections to Remove the Nine Mile 3 Nuclear Project, LLC, Designation NMP2L2883, Fourth Inservice Inspection Interval, Second Inservice Inspection Period 2024 Owner’S Activity Report for RFO-19 Inservice Examinations2024-07-24024 July 2024 Fourth Inservice Inspection Interval, Second Inservice Inspection Period 2024 Owner’S Activity Report for RFO-19 Inservice Examinations ML24198A0852024-07-16016 July 2024 Senior Reactor and Reactor Operator Initial License Examinations RS-24-070, Independent Spent Fuel Storage Installation, Nine Mile Point, Units 1 and 2, Quad Cities, Units 1 and 2, R. E. Ginna - Nuclear Radiological Emergency Plan Document Revisions2024-07-12012 July 2024 Independent Spent Fuel Storage Installation, Nine Mile Point, Units 1 and 2, Quad Cities, Units 1 and 2, R. E. Ginna - Nuclear Radiological Emergency Plan Document Revisions RS-24-061, Constellation Energy Generation, LLC, Response to NRC Regulatory Issue Summary 2024-01, Preparation and Scheduling of Operator Licensing Examinations2024-06-14014 June 2024 Constellation Energy Generation, LLC, Response to NRC Regulatory Issue Summary 2024-01, Preparation and Scheduling of Operator Licensing Examinations NMP1L3584, License Amendment Request to Revise Technical Specifications to Adopt TSTF-230, Revision 1, Add New Condition B to LCO 3.6.2.3, RHR Suppression Pool Cooling2024-06-13013 June 2024 License Amendment Request to Revise Technical Specifications to Adopt TSTF-230, Revision 1, Add New Condition B to LCO 3.6.2.3, RHR Suppression Pool Cooling IR 05000220/20244012024-05-30030 May 2024 Security Baseline Inspection Report 05000220/2024401 and 05000410/2024401(Cover Letter Only) ML24079A0762024-05-23023 May 2024 Issuance of Amendments to Adopt TSTF 264 NMP1L3591, Response to Ny State Pollutant Discharge Elimination System (SPDES) Permit Request for Information & Modification Request2024-05-18018 May 2024 Response to Ny State Pollutant Discharge Elimination System (SPDES) Permit Request for Information & Modification Request NMP1L3589, Special Report: Containment High Range Radiation Monitor Instrumentation Channel 12 Inoperable2024-05-16016 May 2024 Special Report: Containment High Range Radiation Monitor Instrumentation Channel 12 Inoperable ML24158A2052024-05-15015 May 2024 Annual Radioactive Environmental Operating Report NMP1L3582, 2023 Annual Radioactive Environmental Operating Report for Nine Mile Point Units 1 and 22024-05-15015 May 2024 2023 Annual Radioactive Environmental Operating Report for Nine Mile Point Units 1 and 2 IR 05000220/20240012024-05-10010 May 2024 Integrated Inspection Report 05000220/2024001 and 05000410/2024001 RS-24-049, Updated Notice of Intent to Pursue Subsequent License Renewal Applications2024-05-0909 May 2024 Updated Notice of Intent to Pursue Subsequent License Renewal Applications RS-24-038, Relief Request Concerning Extension of Permanent Relief from Ultrasonic Examination of Reactor Pressure Vessel Circumferential Shell Welds2024-05-0202 May 2024 Relief Request Concerning Extension of Permanent Relief from Ultrasonic Examination of Reactor Pressure Vessel Circumferential Shell Welds 05000410/LER-2024-001, Automatic Reactor Scram on Turbine Trip Due to Low Condenser Vacuum2024-05-0101 May 2024 Automatic Reactor Scram on Turbine Trip Due to Low Condenser Vacuum NMP1L3581, Independent Spent Fuel Storage Installation (ISFSI) - 2023 Radioactive Effluent Release Report2024-04-30030 April 2024 Independent Spent Fuel Storage Installation (ISFSI) - 2023 Radioactive Effluent Release Report RS-24-041, Alternative Request to Utilize Code Case OMN-32, Alternative Requirements for Range and Accuracy of Pressure, Flow, and Differential Pressure Instruments Used in Pump Tests2024-04-30030 April 2024 Alternative Request to Utilize Code Case OMN-32, Alternative Requirements for Range and Accuracy of Pressure, Flow, and Differential Pressure Instruments Used in Pump Tests 2025-01-07
[Table view] Category:Licensee Event Report (LER)
MONTHYEAR05000410/LER-2024-002-01, Supplement to NMP2 Licensee Event Report 2024-002-00, Automatic Reactor Scram on Turbine Trip Due to Failed Breaker2025-01-10010 January 2025 Supplement to NMP2 Licensee Event Report 2024-002-00, Automatic Reactor Scram on Turbine Trip Due to Failed Breaker 05000410/LER-2024-002, Automatic Reactor Scram on Turbine Trip Due to Failed Breaker2024-11-22022 November 2024 Automatic Reactor Scram on Turbine Trip Due to Failed Breaker 05000410/LER-2024-001, Automatic Reactor Scram on Turbine Trip Due to Low Condenser Vacuum2024-05-0101 May 2024 Automatic Reactor Scram on Turbine Trip Due to Low Condenser Vacuum 05000410/LER-2023-001, Supplement to LER 2023-001-00, Automatic Reactor Scram on Low Level Due to Partial Loss of Feedwater2024-01-30030 January 2024 Supplement to LER 2023-001-00, Automatic Reactor Scram on Low Level Due to Partial Loss of Feedwater 05000220/LER-2023-002, Average Power Range Monitors Declared Inoperable Due to Trip of Reactor Recirculation Pump 122023-12-15015 December 2023 Average Power Range Monitors Declared Inoperable Due to Trip of Reactor Recirculation Pump 12 05000220/LER-2023-001-01, Supplement to NMP1 Indication on the N2E Dissimilar Metal Weld Exceeding ASME Acceptance Criteria2023-08-11011 August 2023 Supplement to NMP1 Indication on the N2E Dissimilar Metal Weld Exceeding ASME Acceptance Criteria 05000220/LER-2023-001, Indication on the N2E Dissimilar Metal Weld Exceeding ASME Acceptance Criteria2023-05-12012 May 2023 Indication on the N2E Dissimilar Metal Weld Exceeding ASME Acceptance Criteria 05000410/LER-2022-002-01, Reactor Protection System Actuation While Shutdown2022-12-20020 December 2022 Reactor Protection System Actuation While Shutdown 05000410/LER-2022-002, Reactor Protection System Actuation While Shutdown2022-11-0303 November 2022 Reactor Protection System Actuation While Shutdown 05000410/LER-2022-001, Regarding Automatic Reactor Scram Due to Low Reactor Water Level During Maintenance2022-06-0303 June 2022 Regarding Automatic Reactor Scram Due to Low Reactor Water Level During Maintenance 05000220/LER-2021-002, Isolation of Both Emergency Condensers Due to Loss of UPS 162A2021-11-19019 November 2021 Isolation of Both Emergency Condensers Due to Loss of UPS 162A NMP1L3400, Average Power Range Monitors Declared Inoperable Due to Trip of Reactor Recirculation Pump 132021-05-11011 May 2021 Average Power Range Monitors Declared Inoperable Due to Trip of Reactor Recirculation Pump 13 05000220/LER-2020-001-01, Control Room Air Treatment System Inoperable2020-09-15015 September 2020 Control Room Air Treatment System Inoperable 05000410/LER-2020-002-01, Failure to Meet Technical Specification MSIV Stroke Times2020-08-31031 August 2020 Failure to Meet Technical Specification MSIV Stroke Times 05000220/LER-2020-001, Control Room Air Treatment System Inoperable2020-07-0202 July 2020 Control Room Air Treatment System Inoperable 05000410/LER-2020-002, Failure to Meet Technical Specification MSIV Stroke Times2020-05-0505 May 2020 Failure to Meet Technical Specification MSIV Stroke Times 05000410/LER-2020-001, Manual Scram Due to an Electro Hydraulic Control Fluid Leak on the Turbine Control System2020-05-0404 May 2020 Manual Scram Due to an Electro Hydraulic Control Fluid Leak on the Turbine Control System 05000410/LER-2019-001, High Pressure Core Spray Declared Inoperable2019-12-31031 December 2019 High Pressure Core Spray Declared Inoperable 05000220/LER-2019-004, Average Power Range Monitors Declared Inoperable2019-10-0303 October 2019 Average Power Range Monitors Declared Inoperable 05000220/LER-2019-003-01, Manual Reactor Scram Due to Pressure and-Power Oscillations2019-08-0202 August 2019 Manual Reactor Scram Due to Pressure and-Power Oscillations 05000220/LER-2019-001-01, Automatic Reactor Scram Due to High Reactor Pressure2019-07-26026 July 2019 Automatic Reactor Scram Due to High Reactor Pressure 05000220/LER-2019-003, Manual Reactor Scram Due to Pressure and Power Oscillations2019-06-28028 June 2019 Manual Reactor Scram Due to Pressure and Power Oscillations 05000220/LER-2019-002, Condition Prohibited by Technical Specification Due to Vacuum Breaker Not Locked Closed2019-06-24024 June 2019 Condition Prohibited by Technical Specification Due to Vacuum Breaker Not Locked Closed 05000220/LER-2019-001, Automatic Reactor Scram Due to High Reactor Pressure2019-06-13013 June 2019 Automatic Reactor Scram Due to High Reactor Pressure 05000410/LER-2018-002, Turbine Trip and Scram Due to Unit Differential Relay Trip2018-10-25025 October 2018 Turbine Trip and Scram Due to Unit Differential Relay Trip 05000410/LER-2018-001, For Nine Mile Point Unit 2, Auto Start of Division II Emergency Diesel Generator Due to Loss of Line 62018-07-0909 July 2018 For Nine Mile Point Unit 2, Auto Start of Division II Emergency Diesel Generator Due to Loss of Line 6 ML18018B1122018-01-18018 January 2018 Scram Summary 91-01 Relating to an Event on August 13, 1991 Concerning a Turbine Trip and Automatic Reactor Scram When the Main Transformer Phase B Developed an Internal Fault ML18018B1152018-01-18018 January 2018 Scram Summary 91-01 Relating to a Turbine Trip and Automatic Reactor Scram When the Main Transformer Phase B Developed an Internal Fault on August 13, 1991 05000410/LER-1917-002, Regarding Secondary Containment Inoperable Due to Wind Conditions2017-11-28028 November 2017 Regarding Secondary Containment Inoperable Due to Wind Conditions 05000220/LER-1917-003, Regarding Automatic Reactor Scram Due to Reactor Vessel Low Water Level2017-11-0202 November 2017 Regarding Automatic Reactor Scram Due to Reactor Vessel Low Water Level 05000410/LER-1917-001, Regarding Automatic Reactor Scram Due to High Reactor Pressure2017-10-0404 October 2017 Regarding Automatic Reactor Scram Due to High Reactor Pressure 05000220/LER-1917-002, Regarding Manual Reactor Scram Due to Pressure Oscillations2017-05-18018 May 2017 Regarding Manual Reactor Scram Due to Pressure Oscillations 05000220/LER-2017-001, Manual Reactor Scram Due to High Turbine Vibration2017-02-0808 February 2017 Manual Reactor Scram Due to High Turbine Vibration 05000220/LER-2016-002, Regarding Isolation of Both Emergency Condensers Due to Loss of UPS 16282016-09-26026 September 2016 Regarding Isolation of Both Emergency Condensers Due to Loss of UPS 1628 05000220/LER-2016-001, Regarding Secondary Containment Inoperable Due to Simultaneous Opening of Airlock Doors2016-07-12012 July 2016 Regarding Secondary Containment Inoperable Due to Simultaneous Opening of Airlock Doors 05000410/LER-2016-001, Regarding Secondary Containment Inoperable Due to Simultaneous Opening of Airlock Doors2016-06-0606 June 2016 Regarding Secondary Containment Inoperable Due to Simultaneous Opening of Airlock Doors 05000220/LER-2015-004, Regarding Automatic Reactor Scram Due to Main Steam Isolation Valve Closure2015-11-0303 November 2015 Regarding Automatic Reactor Scram Due to Main Steam Isolation Valve Closure 05000220/LER-2015-003, Regarding Secondary Containment Inoperable Due to Simultaneous Opening of Airlock Doors2015-10-0202 October 2015 Regarding Secondary Containment Inoperable Due to Simultaneous Opening of Airlock Doors 05000410/LER-2015-003, Regarding Primary Containment Isolation Function for Some Valves Not Maintained During Surveillance Testing2015-08-21021 August 2015 Regarding Primary Containment Isolation Function for Some Valves Not Maintained During Surveillance Testing 05000220/LER-2015-002, Secondary Containment Inoperable Due to Simultaneous Opening of Airlock Doors2015-04-21021 April 2015 Secondary Containment Inoperable Due to Simultaneous Opening of Airlock Doors 05000410/LER-2015-002, Regarding Manual Reactor Scram Due to Unexpected Reactor Water Level Change2015-04-20020 April 2015 Regarding Manual Reactor Scram Due to Unexpected Reactor Water Level Change 05000220/LER-2015-001, Regarding Secondary Containment Inoperable Due to Simultaneous Opening of Airlock Doors2015-04-10010 April 2015 Regarding Secondary Containment Inoperable Due to Simultaneous Opening of Airlock Doors 05000410/LER-2015-001, For Nine Mile Point, Unit 2, Regarding Secondary Containment Inoperable Due to Sustained High Winds2015-03-12012 March 2015 For Nine Mile Point, Unit 2, Regarding Secondary Containment Inoperable Due to Sustained High Winds 05000220/LER-2014-005, Regarding Secondary Containment Inoperable Due to Simultaneous Opening of Airlock Doors2014-12-12012 December 2014 Regarding Secondary Containment Inoperable Due to Simultaneous Opening of Airlock Doors 05000410/LER-2014-008, Re Secondary Containment Inoperable Due to Reactor Building Exhaust Fan Trip2014-08-0808 August 2014 Re Secondary Containment Inoperable Due to Reactor Building Exhaust Fan Trip 05000220/LER-2014-002, Regarding Unanalyzed Condition Due to Unfused Motor Operated Valve Control Circuit2014-07-0808 July 2014 Regarding Unanalyzed Condition Due to Unfused Motor Operated Valve Control Circuit 05000410/LER-2014-007, Regarding Secondary Containment Inoperable Due to Simultaneous Opening of Airlock Doors2014-06-0202 June 2014 Regarding Secondary Containment Inoperable Due to Simultaneous Opening of Airlock Doors 05000410/LER-2014-006, Regarding Secondary Containment Inoperability Following Auxiliary Boiler Trip2014-05-23023 May 2014 Regarding Secondary Containment Inoperability Following Auxiliary Boiler Trip 05000410/LER-2014-004, Regarding Actuation of the Alternate Rod Insertion System and Subsequent Reactor Scram2014-05-0707 May 2014 Regarding Actuation of the Alternate Rod Insertion System and Subsequent Reactor Scram 05000410/LER-2014-003, Regarding Uninterruptible Power Supply Failure and Subsequent Manual Scram2014-05-0202 May 2014 Regarding Uninterruptible Power Supply Failure and Subsequent Manual Scram 2025-01-10
[Table view] |
LER-2012-006, Technical Specification Required Shutdown Due to Containment Leakage |
Event date: |
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Report date: |
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Reporting criterion: |
10 CFR 50.73(a)(2)(i)(A), Completion of TS Shutdown
10 CFR 50.73(a)(2)(iv)(A), System Actuation
10 CFR 50.73(a)(2)(v), Loss of Safety Function
10 CFR 50.73(a)(2)(i)
10 CFR 50.73(a)(2)(vii), Common Cause Inoperability
10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded
10 CFR 50.73(a)(2)(viii)(A)
10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition
10 CFR 50.73(a)(2)(viii)(B)
10 CFR 50.73(a)(2)(iii)
10 CFR 50.73(a)(2)(ix)(A)
10 CFR 50.73(a)(2)(x)
10 CFR 50.73(a)(2)(v)(A), Loss of Safety Function - Shutdown the Reactor
10 CFR 50.73(a)(2)(v)(B), Loss of Safety Function - Remove Residual Heat
10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications |
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2202012006R00 - NRC Website |
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text
Michel A. Philippon P.O. Box 63 Plant General Manager Lycoming, New York 13093 315.349.5205 315.349.1321 Fax CENG a joint venture of Costellation eF NINE MILE POINT NUCLEAR STATION February 11, 2013 U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 ATTENTION:
Document Control Desk
SUBJECT:
Nine Mile Point Nuclear Station Unit No. 1; Docket No. 50-220 Licensee Event Report 2012-006, Technical Specification Required Shutdown Due to Containment Leakage In accordance with 10 CFR 50.73(a)(2)(i)(A), 10 CFR 50.73(a)(2)(iv)(A) and 10 CFR 50.73(a)(2)(v)(C),
please find attached Licensee Event Report 2012-006, Technical Specification Required Shutdown Due to Containment Leakage.
There are no regulatory commitments in this submittal.
Should you have questions regarding the information in this submittal, please contact John J. Dosa, Director Licensing, at (315) 349-5219.
Very truly yours, MAP/GvN
Attachment:
Licensee Event Report 2012-006, Technical Specification Required Shutdown Due to Containment Leakage cc:
NRC Regional Administrator NRC Project Manager NRC Resident Inspector
ATTACHMENT LICENSEE EVENT REPORT 2012-006 TECHNICAL SPECIFICATION REQUIRED SHUTDOWN DUE TO CONTAINMENT LEAKAGE Nine Mile Point Nuclear Station, LLC February 11, 2013
NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES: 10/31/2013 (10-2010)
, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.
- 3. PAGE Nine Mile Point Unit 1 05000220 1 OF 6
- 4. TITLE Technical Specification Required Shutdown Due to Containment Leakage
- 5. EVENT DATE
- 6. LER NUMBER
- 7. REPORT DATE
- 8. OTHER FACILITIES INVOLVED FACILITY NAME DOCKET NUMBER MONTH DAY YEAR YEAR SEQUENTIAL REV MONTH DAY YEAR NA NA NUMBER NO.NAA FACILITY NAME DOCKET NUMBER 12 13 2012 2012 006 00 2
11 2013 1NA NA
- 9. OPERATING MODE
- 11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR§: (Check all that apply)
El 20.2201(b)
El 20.2203(a)(3)(i)
[E 50.73(a)(2)(i)(C)
El 50.73(a)(2)(vii)
N 0l 20.2201(d)
El 20.2203(a)(3)(ii)
El 50.73(a)(2)(ii)(A)
[I 50.73(a)(2)(viii)(A)
El 20.2203(a)(1)
[E 20.2203(a)(4)
El 50.73(a)(2)(ii)(B)
El 50.73(a)(2)(viii)(B)
- 10. POWER LEVEL I] 20.2203(a)(2)(i)
El 50.36(c)(1)(i)(A)
El 50.73(a)(2)(iii)
El 50.73(a)(2)(ix)(A)
El 20.2203(a)(2)(ii)
El 50.36(c)(1)(ii)(A)
Z 50.73(a)(2)(iv)(A)
[I 50.73(a)(2)(x)
[I 20.2203(a)(2)(iii)
El 50.36(c)(2)
El 50.73(a)(2)(v)(A)
El 73.71(a)(4) 100 [1 20.2203(a)(2)(iv)
[I 50.46(a)(3)(ii)
El 50.73(a)(2)(v)(B)
El 73.71(a)(5)
El 20.2203(a)(2)(v)
[D 50.73(a)(2)(i)(A)
[
50.73(a)(2)(v)(C)
El OTHER El 20.2203(a)(2)(vi)
[E 50.73(a)(2)(i)(B)
El 50.73(a)(2)(v)(D)
Specify in Abstract below or in C. INOPERABLE STRUCTURES, COMPONENTS, OR SYSTEMS THAT CONTRIBUTED TO THE EVENT:
The degraded sealing surfaces of drywell purge isolation valves IV-201-31 and IV-201-32 contributed to this event. There were no other inoperable structures, systems, or components that contributed to the event.
D. DATES AND APPROXIMATE TIMES OF MAJOR OCCURRENCES
December 10, 2012 21:00:00 Condition Report CR-2012-011157 was initiated after an adverse trend was identified in nitrogen makeup to primary containment.
December 12, 2012 17:56:33 Significant leakage was noted between the drywell purge isolation valves IV-201-31 (outboard) and IV-201-32 (inboard) based upon displaced oxygen detected at the downstream vent.
December 13, 2012 15:08:27 LLRT was initiated for drywell purge isolation valves IV-201-31 (outboard) and IV-201-32 (inboard). The test was subsequently declared a failure.
16:30:10 The primary containment was declared inoperable due to a primary containment leakage rate that was in excess of the TS 3.3.3.a limit.
16:45:00 A normal orderly plant shutdown was commenced.
19:12:46 A manual reactor scram was initiated from 18 percent power in order to reduce the reactor coolant temperature to a value less than 215 degrees F within 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> of declaring the containment inoperable.
19:13:09 The High Pressure Coolant Injection (HPCI) system automatically initiated on low Reactor Pressure Vessel (RPV) water level as expected due to RPV level shrink following the scram.
19:13:12 RPV level was restored above the HPCI system low level actuation setpoint and the HPCI system initiation signal was reset.
23:33:00 The reactor coolant temperature was below 215 degrees F.
E. OTHER SYSTEMS OR SECONDARY FUNCTIONS AFFECTED
Other reactor and secondary systems functioned as expected.
NRC FORM 366 (10-2010)
4
F. METHOD OF DISCOVERY
This event was discovered by the operators when the LLRT on drywell purge isolation valves IV-201 -
31 (outboard) and IV-201-32 (inboard) was declared a failure.
G. MAJOR OPERATOR ACTION:
Operators initiated the manual reactor scram per plant procedure. The HPCI system initiation signal was reset after the RPV level was restored above 53 inches. Pressure control was established on the Turbine Bypass Valves, which is the preferred system.
H. SAFETY SYSTEM RESPONSES:
Following initiation of the reactor scram, all control rods fully inserted. The HPCI system automatically initiated on low RPV level as expected due to RPV level shrink following the scram. No other operational conditions requiring the response of safety systems occurred as a result of this event.
II. CAUSE OF EVENT
The cause for this event was excessive leakage through drywell purge isolation valves IV-201-3 land IV-201-
- 32. The general corrosion layer that developed on the purge line carbon steel piping over time became disturbed causing portions of the corrosion to break loose and collect inside the containment vent and purge piping. Containment vent and purge events carried loose corrosion from within the piping through the containment vent and purge isolation valves causing increased leakage due to uneven valve seating. This event was entered into the Nine Mile Point Nuclear Station corrective action program as condition report number CR-2012-011247.
III. ANALYSIS OF THE EVENT
This event is reportable in accordance with 10 CFR 50.73(a)(2)(i)(A) as a plant shutdown required by Technical Specifications, with 10 CFR 50.73(a)(2)(iv)(A) as an actuation of the high pressure coolant injection system, and with 10 CFR 50.73(a)(2)(v)(C) as a condition that could have prevented the fulfillment of a safety function of a system needed to control the release of radioactive material.
There were no actual nuclear safety consequences associated with this event. All control rods fully inserted following the reactor scram. The HPCI system automatically initiated on low RPV level as expected, due to RPV level shrink following the scram. There were no other automatic initiations of safety systems, and immediate actions performed by the operators were adequate and appropriate in placing and maintaining the reactor in a safe shutdown condition. The reactor scram was without complications and was not risk significant.
The closest related transient described in the NMP1 Updated Final Safety Analysis Report (UFSAR) is the Turbine Trip with Partial Bypass (Low Power) event described in UFSAR Section XV-B.3.14. The maximum reactor pressure and peak neutron flux reached during the December 13, 2012 event were both less than the calculated values presented in the UFSAR analysis for a low power turbine trip with partial bypass flow. In addition, this transient event does not challenge the Minimum Critical Power Ratio safety limit and, therefore, is not evaluated on a reload cycle basis.
NRC FORM 366 (10-2010)
I
Based on the above discussion, it is concluded that the safety significance of this event is low and the event did not pose a threat to the health and safety of the public or plant personnel.
This event affects two NRC Regulatory Oversight Process (ROP) performance indicators (PIs). The PI for Unplanned Power Changes per 7000 Critical Hours is projected to rise from 0.85 to 1.73, compared to the Green-to-White threshold value of 6. Also, the PI for Safety System Functional Failures rises from 3 to 4, compared to the Green-to-White threshold value of 6. Both PIs will remain green.
IV. CORRECTIVE ACTIONS
A. ACTION TAKEN TO RETURN AFFECTED SYSTEMS TO PRE-EVENT NORMAL STATUS:
- 1. Disassembled, inspected and cleaned containment vent and purge isolation valves IV-201-09, IV-201-10, IV-201-31 and IV-201-32 and associated piping. All retrievable loose corrosion was removed.
- 2. Performed local leak rate testing on valves IV-201-07, IV-201-08, IV-201-09, IV-201-10, IV-201-16, IV-201-17, IV-201-31 and IV-201-32.
- 3.
Pressurized containment to approximately 1.5 psig and trended containment pressure to verify there was no significant leakage.
B. ACTION TAKEN OR PLANNED TO PREVENT RECURRENCE:
- 1. During the next refueling outage, remove corrosion from the internal surfaces of the four horizontally-oriented containment vent and purge isolation valves IV-201-09, IV-201-10, IV-201-31 and IV-201-32 and associated piping.
- 2. During the next refueling outage, coat the internal piping and associated attachments to prevent corrosion for the four horizontally-oriented containment vent and purge isolation valves IV-201-09, IV-201-10, IV-201-31 and IV-201-32.
- 3.
During the next refueling outage, inspect the internal piping and associated attachments of the four vertically-oriented containment vent and purge isolation valves IV-201-07, IV-201-8, IV-201-16 and IV-201-17 and, based on the inspection findings, develop a cleaning/coating strategy (if required).
V. ADDITIONAL INFORMATION
A. FAILED COMPONENTS:
There were no other failed components that contributed to this event.
B. PREVIOUS LERs ON SIMILAR EVENTS:
There were no previous LERs on similar events.
NRC FORM 366 (10-2010)
C. THE ENERGY INDUSTRY IDENTIFICATION SYSTEM (EIIS) COMPONENT FUNCTION IDENTIFIER AND SYSTEM NAME OF EACH COMPONENT OR SYSTEM REFERRED TO IN THIS LER:
COMPONENT Reactor Protection System High Pressure Coolant Injection System Reactor Pressure Vessel Containment Isolation Valve Nitrogen Injection System Turbine Bypass Valve D. SPECIAL COMMENTS:
None IEEE 803 FUNCTION IDENTIFIER N/A N/A RPV ISV N/A N/A IEEE 805 SYSTEM IDENTIFICATION JC BJ AD NA NA JI NRC FORM 366 (10-2010)
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05000220/LER-2012-001, Regarding Automatic Reactor Scram Due to Electronic Pressure Regulator Failure | Regarding Automatic Reactor Scram Due to Electronic Pressure Regulator Failure | 10 CFR 50.73(a)(2)(iv)(A), System Actuation 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition 10 CFR 50.73(a)(2)(v)(A), Loss of Safety Function - Shutdown the Reactor 10 CFR 50.73(a)(2)(v)(B), Loss of Safety Function - Remove Residual Heat 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2)(i) 10 CFR 50.73(a)(2)(x) 10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded 10 CFR 50.73(a)(2)(iii) 10 CFR 50.73(a)(2)(i)(A), Completion of TS Shutdown 10 CFR 50.73(a)(2)(ix)(A), Prevented Safety Function in Multiple System 10 CFR 50.73(a)(2) 10 CFR 50.73(a)(2)(viii)(A) 10 CFR 50.73(a)(2)(viii)(B) | 05000410/LER-2012-001, Forced Shutdown Due to an Increase in Drywell Leakage in Excess of Technical Specifications Limit | Forced Shutdown Due to an Increase in Drywell Leakage in Excess of Technical Specifications Limit | 10 CFR 50.73(a)(2)(iv)(A), System Actuation 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition 10 CFR 50.73(a)(2)(v)(A), Loss of Safety Function - Shutdown the Reactor 10 CFR 50.73(a)(2)(v)(B), Loss of Safety Function - Remove Residual Heat 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2)(i) 10 CFR 50.73(a)(2)(x) 10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded 10 CFR 50.73(a)(2)(iii) 10 CFR 50.73(a)(2)(i)(A), Completion of TS Shutdown 10 CFR 50.73(a)(2)(ix)(A), Prevented Safety Function in Multiple System 10 CFR 50.73(a)(2)(viii)(A) 10 CFR 50.73(a)(2)(viii)(B) | 05000220/LER-2012-002, Regarding Automatic Reactor Scram Due to Automatic Generator Protective Trip | Regarding Automatic Reactor Scram Due to Automatic Generator Protective Trip | 10 CFR 50.73(a)(2)(iv)(A), System Actuation 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition 10 CFR 50.73(a)(2)(v)(A), Loss of Safety Function - Shutdown the Reactor 10 CFR 50.73(a)(2)(v)(B), Loss of Safety Function - Remove Residual Heat 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2)(i) 10 CFR 50.73(a)(2)(x) 10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded 10 CFR 50.73(a)(2)(iii) 10 CFR 50.73(a)(2)(i)(A), Completion of TS Shutdown 10 CFR 50.73(a)(2)(ix)(A), Prevented Safety Function in Multiple System 10 CFR 50.73(a)(2)(viii)(A) 10 CFR 50.73(a)(2)(viii)(B) | 05000410/LER-2012-002, Regarding Loss of Isolation Function on RHR Shutdown Cooling Suction Line Due to Breaker Trip | Regarding Loss of Isolation Function on RHR Shutdown Cooling Suction Line Due to Breaker Trip | 10 CFR 50.73(a)(2)(iv)(A), System Actuation 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition 10 CFR 50.73(a)(2)(v)(A), Loss of Safety Function - Shutdown the Reactor 10 CFR 50.73(a)(2)(v)(B), Loss of Safety Function - Remove Residual Heat 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(i) 10 CFR 50.73(a)(2)(x) 10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded 10 CFR 50.73(a)(2)(iii) 10 CFR 50.73(a)(2)(i)(A), Completion of TS Shutdown 10 CFR 50.73(a)(2)(ix)(A), Prevented Safety Function in Multiple System 10 CFR 50.73(a)(2)(viii)(A) 10 CFR 50.73(a)(2)(viii)(B) | 05000220/LER-2012-003, Regarding Loss of Isolation Function on Shutdown Cooling System Suction Line Due to an Operating Procedure Deficiency | Regarding Loss of Isolation Function on Shutdown Cooling System Suction Line Due to an Operating Procedure Deficiency | 10 CFR 50.73(a)(2)(iv)(A), System Actuation 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition 10 CFR 50.73(a)(2)(v)(A), Loss of Safety Function - Shutdown the Reactor 10 CFR 50.73(a)(2)(v)(B), Loss of Safety Function - Remove Residual Heat 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2)(i) 10 CFR 50.73(a)(2)(x) 10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded 10 CFR 50.73(a)(2)(iii) 10 CFR 50.73(a)(2)(i)(A), Completion of TS Shutdown 10 CFR 50.73(a)(2)(ix)(A), Prevented Safety Function in Multiple System 10 CFR 50.73(a)(2)(viii)(A) 10 CFR 50.73(a)(2)(viii)(B) | 05000410/LER-2012-003, Regarding Suppression Pool Level Below Technical Specification Limit During Mode Change | Regarding Suppression Pool Level Below Technical Specification Limit During Mode Change | 10 CFR 50.73(a)(2)(iv)(A), System Actuation 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition 10 CFR 50.73(a)(2)(v)(A), Loss of Safety Function - Shutdown the Reactor 10 CFR 50.73(a)(2)(v)(B), Loss of Safety Function - Remove Residual Heat 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2)(i) 10 CFR 50.73(a)(2)(x) 10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded 10 CFR 50.73(a)(2)(iii) 10 CFR 50.73(a)(2)(i)(A), Completion of TS Shutdown 10 CFR 50.73(a)(2)(ix)(A), Prevented Safety Function in Multiple System 10 CFR 50.73(a)(2)(viii)(A) 10 CFR 50.73(a)(2)(viii)(B) | 05000220/LER-2012-004, Regarding Automatic Reactor Scram Due to a Generator Load Reject | Regarding Automatic Reactor Scram Due to a Generator Load Reject | 10 CFR 50.73(a)(2)(iv)(A), System Actuation 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition 10 CFR 50.73(a)(2)(v)(A), Loss of Safety Function - Shutdown the Reactor 10 CFR 50.73(a)(2)(v)(B), Loss of Safety Function - Remove Residual Heat 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2)(i) 10 CFR 50.73(a)(2)(x) 10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded 10 CFR 50.73(a)(2)(iii) 10 CFR 50.73(a)(2)(i)(A), Completion of TS Shutdown 10 CFR 50.73(a)(2)(ix)(A), Prevented Safety Function in Multiple System 10 CFR 50.73(a)(2)(viii)(A) 10 CFR 50.73(a)(2)(viii)(B) | 05000410/LER-2012-004, Regarding Manual Reactor Scram Due to a Loss of Main Turbine Gland Sealing Steam Resulting in Lowering Condenser Vacuum | Regarding Manual Reactor Scram Due to a Loss of Main Turbine Gland Sealing Steam Resulting in Lowering Condenser Vacuum | 10 CFR 50.73(a)(2)(iv)(A), System Actuation 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition 10 CFR 50.73(a)(2)(v)(A), Loss of Safety Function - Shutdown the Reactor 10 CFR 50.73(a)(2)(v)(B), Loss of Safety Function - Remove Residual Heat 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2)(i) 10 CFR 50.73(a)(2)(x) 10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded 10 CFR 50.73(a)(2)(iii) 10 CFR 50.73(a)(2)(i)(A), Completion of TS Shutdown 10 CFR 50.73(a)(2)(ix)(A), Prevented Safety Function in Multiple System 10 CFR 50.73(a)(2)(viii)(A) 10 CFR 50.73(a)(2)(viii)(B) | 05000220/LER-2012-005, Regarding Feedwater Level Control Failure, HPCI Initiation and Reactor Scram | Regarding Feedwater Level Control Failure, HPCI Initiation and Reactor Scram | 10 CFR 50.73(a)(2)(iv)(A), System Actuation 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition 10 CFR 50.73(a)(2)(v)(A), Loss of Safety Function - Shutdown the Reactor 10 CFR 50.73(a)(2)(v)(B), Loss of Safety Function - Remove Residual Heat 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2)(i) 10 CFR 50.73(a)(2)(x) 10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded 10 CFR 50.73(a)(2)(iii) 10 CFR 50.73(a)(2)(i)(A), Completion of TS Shutdown 10 CFR 50.73(a)(2)(ix)(A), Prevented Safety Function in Multiple System 10 CFR 50.73(a)(2)(viii)(A) 10 CFR 50.73(a)(2)(viii)(B) | 05000410/LER-2012-005, Regarding Automatic Diesel Actuation Due to the Loss of a 115 Kv Offsite Power Source | Regarding Automatic Diesel Actuation Due to the Loss of a 115 Kv Offsite Power Source | 10 CFR 50.73(a)(2)(iv)(A), System Actuation 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition 10 CFR 50.73(a)(2)(v)(A), Loss of Safety Function - Shutdown the Reactor 10 CFR 50.73(a)(2)(v)(B), Loss of Safety Function - Remove Residual Heat 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2)(i) 10 CFR 50.73(a)(2)(x) 10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded 10 CFR 50.73(a)(2)(iii) 10 CFR 50.73(a)(2)(i)(A), Completion of TS Shutdown 10 CFR 50.73(a)(2)(ix)(A), Prevented Safety Function in Multiple System 10 CFR 50.73(a)(2)(viii)(A) 10 CFR 50.73(a)(2)(viii)(B) | 05000220/LER-2012-006, Technical Specification Required Shutdown Due to Containment Leakage | Technical Specification Required Shutdown Due to Containment Leakage | 10 CFR 50.73(a)(2)(iv)(A), System Actuation 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition 10 CFR 50.73(a)(2)(v)(A), Loss of Safety Function - Shutdown the Reactor 10 CFR 50.73(a)(2)(v)(B), Loss of Safety Function - Remove Residual Heat 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2)(i) 10 CFR 50.73(a)(2)(x) 10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded 10 CFR 50.73(a)(2)(iii) 10 CFR 50.73(a)(2)(i)(A), Completion of TS Shutdown 10 CFR 50.73(a)(2)(ix)(A), Prevented Safety Function in Multiple System 10 CFR 50.73(a)(2)(viii)(A) 10 CFR 50.73(a)(2)(viii)(B) | 05000220/LER-2012-007, Regarding High Pressure Coolant Injection System Logic Actuation Following an Automatic Turbine Trip Signal Due to High Reactor Water Level | Regarding High Pressure Coolant Injection System Logic Actuation Following an Automatic Turbine Trip Signal Due to High Reactor Water Level | 10 CFR 50.73(a)(2)(iv)(A), System Actuation 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition 10 CFR 50.73(a)(2)(v)(A), Loss of Safety Function - Shutdown the Reactor 10 CFR 50.73(a)(2)(v)(B), Loss of Safety Function - Remove Residual Heat 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2)(i) 10 CFR 50.73(a)(2)(x) 10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded 10 CFR 50.73(a)(2)(iii) 10 CFR 50.73(a)(2)(i)(A), Completion of TS Shutdown 10 CFR 50.73(a)(2)(ix)(A), Prevented Safety Function in Multiple System 10 CFR 50.73(a)(2)(viii)(A) 10 CFR 50.73(a)(2)(viii)(B) |
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