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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-2012-004, Regarding Manual Reactor Scram Due to a Loss of Main Turbine Gland Sealing Steam Resulting in Lowering Condenser Vacuum |
Event date: |
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Report date: |
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Reporting criterion: |
10 CFR 50.73(a)(2)(iv)(A), System Actuation
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)(A), Completion of TS Shutdown
10 CFR 50.73(a)(2)(v), Loss of Safety Function
10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications |
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4102012004R00 - NRC Website |
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text
Michel A. Philippon Plant General Manager P.O. Box 63 Lycoming, New York 13093 315.349.5205 315.349.1321 Fax CENG.
a joint venture of 0h Constellation W
e Energy
Ž'eDF NINE MILE POINT NUCLEAR STATION September 10, 2012 U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 ATTENTION:
SUBJECT:
Document Control Desk Nine Mile Point Nuclear Station Unit No. 2; Docket No. 50-410 Licensee Event Report 2012-004, Manual Reactor Scram due to a Loss of Main Turbine Gland Sealing Steam Resulting in Lowering Condenser Vacuum.
In accordance with 10 CFR 50.73(a)(2)(iv)(A), please find attached Licensee Event Report 2012-004, Manual Reactor Scram due to a Loss of Main Turbine Gland Sealing Steam Resulting in Lowering Condenser Vacuum.
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/DEV
Attachment:
Licensee Event Report 2012-004, Manual Reactor Scram due to a Loss of Main Turbine Gland Sealing Steam Resulting in Lowering Condenser Vacuum cc:
Regional Administrator, NRC Project Manager, NRC Resident Inspector, NRC
/ I,
ATTACHMENT LICENSEE EVENT REPORT 2012-004 MANUAL REACTOR SCRAM DUE TO A LOSS OF MAIN TURBINE GLAND SEALING STEAM RESULTING IN LOWERING CONDENSER VACUUM Nine Mile Point Nuclear Station, LLC September 10, 2012
NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES: 10/31/2013
_;*o-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 2 05000410 1 OF 6
- 4. TITLE Manual Reactor Scram due to a Loss of Main Turbine Gland Sealing Steam Resulting in Lowering Condenser Vacuum
- 5. EVENT DATE
- 6. LER NUMBER _
- 7. REPORT DATE
- 8. OTHER FACILITIES INVOLVED FACILITY NAME DOCKET NUMBER MOTH DA YAR YERSEQUENTIAL REV MONTH DAY YEAR MONTHUMBEDAY.MYEARDAYYEAR None NA NUMBER NO.NoeA I
I l_
I_
I FACILITY NAME DOCKET NUMBER 07 12 2012 2012 004 00 09 10 2012 1 None NA
- 9. OPERATING MODE
- 11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR§: (Check all that apply)
[1 20.2201(b)
El 20.2203(a)(3)(i)
El 50.73(a)(2)(i)(C)
[I 50.73(a)(2)(vii)
E] 20.2201(d)
[I 20.2203(a)(3)(ii)
El 50.73(a)(2)(ii)(A)
El 50.73(a)(2)(viii)(A)
El 20.2203(a)(1)
[I 20.2203(a)(4)
El 50.73(a)(2)(ii)(B)
El 50.73(a)(2)(viii)(B)
- 10. POWER LEVEL E] 20.2203(a)(2)(i)
[I 50.36(c)(1)(i)(A)
[E 50.73(a)(2)(iii)
El 50.73(a)(2)(ix)(A)
[I 20.2203(a)(2)(ii)
[I 50.36(c)(1)(ii)(A)
Z 50.73(a)(2)(iv)(A)
El 50.73(a)(2)(x)
[I 20.2203(a)(2)(iii)
El 50.36(c)(2)
[E 50.73(a)(2)(v)(A)
El 73.71(a)(4) 085 El 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)
[I 50.73(a)(2)(i)(A)
El 50.73(a)(2)(v)(C)
E] 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 =
1. DESCRIPTION OF EVENT
A. PRE-EVENT PLANT CONDITIONS:
Prior to the event, Nine Mile Point Unit 2 (NMP2) was operating at approximately 96 percent of rated thermal power with no inoperable systems affecting this event.
B. EVENT:
On July 12, 2012 at 02:13, with NMP2 operating at approximately 96 percent power (3,811 MWt), plant operators identified that the indicated offgas system inlet pressure was rising and main condenser vacuum was lowering. In accordance with plant procedures, the operators lowered reactor power to approximately 85 percent by adjusting the reactor recirculation system flow. Offgas system inlet pressure continued to increase, and condenser vacuum continued to degrade, resulting in the operators initiating a manual reactor scram at 02:20. All control rods fully inserted and all systems functioned as expected following the scram. There was no impact on Nine Mile Point Unit 1 from this event.
The immediate cause of the increasing offgas system inlet pressure and flow and, the degrading condenser vacuum was the failure of the turbine gland sealing (TME) system to provide an adequate supply of sealing steam to the main turbine gland seals. This occurred because the steam supply to the inservice 'B' clean steam reboiler isolated due to a low water level condition, and the backup sealing steam supply (from the main steam system) was subsequently unable to provide steam at the pressure needed to effectively perform the gland sealing function.
This event involved the manual actuation of the Reactor Protection System, which resulted in a reactor scram. The NRC notification per 10 CFR 50.72(b)(2)(iv)(B) was completed on July 12, 2012, at 04:49 (Event Number 48097).
C. INOPERABLE STRUCTURES, COMPONENTS, OR SYSTEMS THAT CONTRIBUTED TO THE EVENT:
There were no inoperable structures, systems, or components at the time of the scram that contributed to this event.
NRC FORM 366 (10-2010)
D. DATES AND APPROXIMATE TIMES OF MAJOR OCCURRENCES
July 12, 2012 01:57 02:13 02:20 The steam supply to the inservice 'B' clean steam reboiler isolates due to a low water level condition.
Plant operators reduce reactor power to approximately 85 percent by adjusting reactor recirculation system flow, after identifying that indicated offgas system inlet pressure is rising and main condenser vacuum is lowering.
A manual reactor scram is initiated when offgas system inlet pressure exceeds the procedural limit, and with main condenser vacuum continuing to degrade.
E. OTHER SYSTEMS OR SECONDARY FUNCTIONS AFFECTED
No other systems or functions were affected.
F. METHOD OF DISCOVERY
This event was discovered by the operators when indications of rising offgas system inlet pressure and lowering condenser vacuum were observed in the control room.
G. MAJOR OPERATOR ACTION:
Upon discovery of the rising offgas system inlet pressure and degrading condenser vacuum, the operators lowered reactor power to approximately 85 percent by adjusting the reactor recirculation system flow. When indicated offgas system inlet pressure exceeded 19 psia, the reactor was manually scrammed in accordance with plant procedures.
H. SAFETY SYSTEM RESPONSES:
Following initiation of the manual scram, all control rods fully inserted. No other operational conditions requiring the response of safety systems occurred as a result of this event.
NRC FORM 366 (10-2010)
I1. CAUSE OF EVENT:
The turbine gland sealing (TME) system is designed to provide clean sealing steam from the clean steam reboilers. Backup sealing steam from the main steam system is provided in the event that the normal sealing steam source fails. The sealing steam prevents steam leakage out through the high-pressure turbine shaft and turbine steam control valves, and prevents air in-leakage through the low-pressure turbine shaft. Normal sealing steam operating pressure is approximately 4 psi.
The immediate cause of the increasing offgas system inlet pressure and flow and the degrading condenser vacuum, leading to the manual reactor scram, was the failure of the turbine gland sealing system to provide an adequate supply of sealing steam. The event was initiated by isolation of the steam supply to the in-service 'B' clean steam reboiler due to a low water level condition. The low water condition resulted from excess steam demand on the 'B' clean steam reboiler due to leakage past relief valve 2TME-RV1 35 located on the sealing steam piping downstream of the reboiler. A gag that had been incorrectly installed on 2TME-RV135 during the 2012 refueling outage as part of a permanent design change allowed leakage past the relief valve, resulting in a steam demand that exceeded the capacity of the makeup water supply line to the reboiler.
Following the trip of the 'B' clean steam reboiler, the backup sealing steam supply failed to maintain the turbine seals due to a low setpoint on the controller for initiation of the backup supply. The backup sealing steam supply initiated; however, the supply pressure was too low to maintain proper low-pressure turbine sealing. It was determined that, although the normal sealing steam system was being operated at higher than normal operating pressure (approximately 6 psi) to account for degraded 'A' low-pressure turbine seals, the controller for the backup sealing steam supply was set at 2.7 psi. The cause for this deficiency was inadequate guidance provided in the TME system operating procedure off-normal section. The procedure failed to address the impact of raising TME system pressure to account for the degraded low-pressure turbine seals.
The cause of the event is attributed to inadequate management guidance to ensure that changes in scope during engineering change package development are adequately reviewed and assessed.
The original scope of the design change regarding relief valve 2TME-RV135 was to perform a review to determine if a setpoint change would be necessary to support operation at extended power uprate conditions; however, the design change scope was subsequently altered to install the gag instead. The revised design change package did not contain an adequate level of detail in the installation instructions for gagging the relief valve, and did not contain adequate testing provisions to verify that the gag was installed properly. These deficiencies resulted in the loss of gland sealing steam to the turbine generator and the subsequent manual reactor scram. This event was entered into the Nine Mile Point Nuclear Station corrective action program as condition report number CR-2012-006615.
NRC FORM 366 (10-2010)
III. ANALYSIS OF THE EVENT
This event involved a valid actuation of the Reactor Protection System which resulted in a reactor scram, and is reportable in accordance with 10 CFR 50.73(a)(2)(iv)(A).
There were no actual nuclear safety consequences associated with this event. All control rods fully inserted following initiation of the manual reactor 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 loss of condenser vacuum is modeled in the probabilistic risk assessment (PRA) and contributes approximately 2 percent to the base core damage frequency. The manual reactor scram was without complications and was not risk significant. Based on this 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.
The NRC performance indicator for Unplanned Scrams per 7,000 Critical Hours is projected to rise to approximately 1.96 and remains green. No other NRC performance indicators were impacted by this event.
IV. CORRECTIVE ACTIONS
A. ACTION TAKEN TO RETURN AFFECTED SYSTEMS TO PRE-EVENT NORMAL STATUS:
The gag installed in relief valve 2TME-RV1 35 was adjusted to secure the valve disk from opening, and a design change was implemented to increase the initiation setpoint for the backup sealing steam supply to address operation at the higher turbine gland sealing system pressure. Analyses have determined that sufficient margin exists within the turbine gland sealing system to continue operation at the higher operating pressure until repairs can be made to the 'A' low-pressure turbine seals. The turbine gland sealing system was returned to service, and the plant was subsequently returned to power operation (Mode 1) on July 15, 2012.
B. ACTION TAKEN OR PLANNED TO PREVENT RECURRENCE:
- 1. Revise the engineering change package management procedure to address required reviews when scope changes are identified. The revised procedure will also require that implementing departments review the installation and testing sections of design changes to ensure an adequate level of detail.
- 2. Revise the TME system operating procedure to require the following if the system is to be operated at pressures above the normal operating pressure: (1) prepare and issue an engineering change package (ECP) to reflect revisions to backup sealing steam supply initiation pressure; and (2) develop and implement a plan to monitor system margin and capacity to assure no adverse consequences.
- 3. A work order has been prepared to inspect/refurbish the 'A' low-pressure turbine seals during the 2014 refueling outage.
NRC FORM 366 (10-2010)
NJRC FOIRM 366A U.S. NUCLEAR REGULATORY COMMISSION
'10-2010)
LICENSEE EVENT REPORT (LER)
CONTINUATION SHEET
- 1. FACILITY NAME
- 2. DOCKET
- 6. LER NUMBER
- 3. PAGE YEAR SEQUENTIAL REVISION YA NUMBER NUMBER Nine Mile Point Unit 2 05000410 6
OF 6
112012 004 00 NARRATIVE
V. ADDITIONAL INFORMATION
A. FAILED COMPONENTS:
There were no failed components that contributed to this event.
B. PREVIOUS LERs ON SIMILAR EVENTS:
LER 2006-001, Automatic Reactor Scram due to a Loss of Main Turbine Gland Sealing Steam, submitted on May 5, 2006, describes an event in which NMP2 automatically scrammed from 86 percent power. The scram was caused by a main turbine trip due to low condenser vacuum that resulted from a failure in the main turbine gland sealing system. A high level condition caused isolation of the steam outlet valve from the inservice 'A' clean steam reboiler, and the backup sealing steam supply (from the main steam system) failed to function due to a disconnected mechanical linkage for the backup system pressure indicating controller. The actions taken following this event would not have prevented the July 12, 2012 event from occurring.
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 Condenser Main Turbine Main Turbine Gland Sealing System Clean Steam Reboiler Relief Valve IEEE 803 FUNCTION IDENTIFIER N/A COND TRB N/A RBLR RV IEEE 805 SYSTEM IDENTIFICATION JC SG TA TC TC TC D. SPECIAL COMMENTS:
None 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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