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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-2013-001, Emergency Condenser 11 High Steam Flow Isolation Instrumentation Loss During Plant Startup |
Event date: |
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Report date: |
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Reporting criterion: |
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)(iv)(A), System Actuation
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)(i)(B), Prohibited by Technical Specifications |
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2202013001R00 - 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 Constellation
ý6eDn Energy" 10 NINE MILE POINT NUCLEAR STATION July 12, 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 2013-001, Emergency Condenser 11 High Steam Flow Isolation Instrumentation Loss during Plant Startup In accordance with 10 CFR 50.73(a)(2)(v)(D), please find attached Licensee Event Report 2013-001, Emergency Condenser 11 High Steam Flow Isolation Instrumentation Loss during Plant Startup.
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/JBH
Attachment:
Licensee Event Report 2013-001, Emergency Condenser 11 High Steam Flow Isolation Instrumentation Loss during Plant Startup.
cc:
NRC Project Manager NRC Resident Inspector NRC Regional Administrator
ATTACHMENT LICENSEE EVENT REPORT 2013-001 EMERGENCY CONDENSER 11 HIGH STEAM FLOW ISOLATION INSTRUMENTATION LOSS DURING PLANT STARTUP Nine Mile Point Nuclear Station, LLC July 12, 2013
NRC FORM 366 U.S. NUCLEAR REGULAfCRY COk.,,2]ISSION 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.
l3. PAGE Nine Mile Point Unit 1 05000220 1 OF 5
- 4. TITLE Emergency Condenser 1I High Steam Flow Isolation Instrumentation Loss during Plant Startup
- 5. EVENT DATE
- 6. LER NUMBER _
- 7. REPORT DATE
- 8. OTHER FACILITIES INVOLVED IFACILITY NAME DOCKET NUMBER MONTH DAY YEAR Y
SEQUENTIAL REV MONTH DAY YEAR MOT A ER yER NME NO MOT DAYEAR NA NA NUMBER NO.NAA I I I
I I
FACILITY NAMEf DOCKET NUMBER 05 14 2013 2013 001 00 07 1:2 2013 1 NA NA
- 9. OPERATING MODE 11, THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR§: (Check all that apply)
O 20.2201(b)
[I 20.2203(a)(3)(i) 0l 50.73(a)(2)(i)(C)
El 50.73(a)(2)(vii)
N El 20.2201(d)
El 20.2203(a)(3)(ii)
El 50.73(a)(2)(ii)(A)
E0 50.73(a)(2)(viii)(A)
__ 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)
El 50.36(c)(1)(i)(A)
[E 50.73(a)(2)(iii)
El 50.73(a)(2)(ix)(A)
C] 20.2203(a)(2)(ii)
El 50.36(c)(1)(ii)(A)
El 50.73(a)(2)(iv)(A)
El 50.73(a)(2)(x)
El 20.2203(a)(2)(iii)
El 50.36(c)(2)
El 50.73(a)(2)(v)(A)
El 73.71(a)(4) 003 El 20.2203(a)(2)(iv)
El 50.46(a)(3)(ii)
El 50.73(a)(2)(v)(B)
El 73.71(a)(5)
El 20.2203(a)(2)(v)
El 50.73(a)(2)(i)(A)
El 50.73(a)(2)(v)(C)
El OTHER El 20.2203(a)(2)(vi)
El 50.73(a)(2)(i)(B)
ED 50.73(a)(2)(v)(D)
Specify in Abstract below or in =
I. DESCRIPTION OF EVENT
A. PRE-EVENT PLANT CONDITIONS:
Prior to the event, Nine Mile Point Unit I (NMP1) was conducting a reactor startup from the recently completed refueling outage, with reactor power approximately 3 percent.
B. EVENT:
On May 14, 2013, during reactor startup following the completion of refueling outage N1R22, NMPI experienced both channels of high steam flow instrumentation on the EC System Loop 11 going into a gross fail condition. The channel 12 trip unit associated with transmitter DPT-36-06D went into gross fail and subsequently channel 11 trip unit associated with transmitter DPT-36-06C went into gross fail.
On May 14, 2013 at 1215 EC System Loop 11 was manually isolated in accordance with station operating procedures to comply with Technical Specifications (TS). TS 3.1.3.b was entered requiring EC System Loop 11 to be returned to an operable status within 7-days and the completion of TS surveillance requirement 4.1.3.f.
The transmitter equalizing valves for both transmitters were opened and subsequently closed to equalize the sensing lines static head. When this was done the respective trip unit readouts and voltage output readings returned to normal. Afterward, the voltage readings were monitored for a period of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> to confirm that the instrument differential pressure remained normal and allowing steam condensation to fill the sensing lines. The EC System Loop 11 high steam flow instruments were declared operable followed by restoration of EC System Loop 1 to a standby condition and declaring it operable.
This event did not affect Nine Mile Point Unit 2.
An event notification was made in accordance with 10 CFR 50.72(b)(3)(v)(D) for the loss of EC Loop II isolation capability on high steam flow on May 14, 2013 at 1720 (EN 49029).
C. INOPERABLE STRUCTURES, COMPONENTS, OR SYSTEMS THAT CONTRIBUTED TO THE EVENT:
Prior to the event, channel 12 high steam flow instrumentation on the EC System Loop 11 was inoperable.
D. DATES AND APPROXIMATE TIMES OF MAJOR OCCURRENCES
May 14, 2013 0115 The reactor mode switch is placed in startup 0840 The gross fail set point is reached for EC System Loop 11 channel 12 steam flow differential pressure transmitter DPT-36-06D. TS Table 3.6.2.c note (f) action is entered for one channel less than the required minimum for one trip system, requiring the channel to be placed in a trip condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
1145 The gross fail setpoint is reached for EC System Loop 11 channel 11 steam flow differential Dressure transmitter DPT-36-06C. TS Table 3.6.2.c note (f) action is entered for one channel less NRC FORM 366 (10-2010)
than the required minimum for both trip systems. The action requires tripping one channel within an hour and tripping the remaining channel within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, or alternatively to take the action of TS 3.6.2.a (3), which requires declaring the affected EC System Loop 1 I inoperable and taking the action of TS 3.1.3.b.
1215 EC System Loop 11 is isolated. TS 3.1.3.b is entered.
1500 Both transmitters were equalized, the readings returned to normal, and gross fail alarms reset.
Monitoring for 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> commenced.
May 15, 2013 0600 EC System Loop 11 high steam flow instrumentation is declared operable. TS Table 3.6.2.c note (f) actions are exited.
0612 EC System Loop 11 is restored to a standby lineup. The system is declared operable and TS 3.1.3.b is exited.
E. OTHER SYSTEMS OR SECONDARY FUNCTIONS AFFECTED
No other systems or functions were affected.
F. METHOD OF DISCOVERY
On May 14, 2013 at 1145 the control room received Reactor Protection System Analog Trip System (RPS ATS) trouble light for channel 11.
G. MAJOR OPERATOR ACTION:
Operators initiated TS actions resulting in isolating the EC System Loop 11. Following troubleshooting and
- execution of a confidence run, EC System Loop 11 high steam flow instrumentation system was declared operable, followed by returning EC System Loop 11 to a standby configuration.
H. SAFETY SYSTEM RESPONSES:
No operational conditions requiring the response of safety systems occurred as a result of this event.
II. CAUSE OF EVENT
The cause of the failure of channel 11 and 12 high steam flow instrumentation for EC System Loop II is leakage from the shared high pressure sensing line for differential pressure transmitters DPT-36-06C and DPT-36-06D. The apparent cause of the leakage in the transmitter high pressure sensing line is leak by on instrument blowdown valves VLV-36-374 and VLV-89-72.
This event has been entered into the Nine Mile Point Nuclear Station corrective action program as condition report number CR-2013-004347.
NRC FORM 366 (10-2010)
III. ANALYSIS OF THE EVENT
The event is reportable in accordance with 10 CFR 50.73(a)(2)(v)(D) as a condition that could have prevented the fulfillment of the safety function of structures or systems that are needed to mitigate the consequences of an accident.
There were no actual nuclear safety consequences associaied with this event. The event was caused by leakage in the shared high pressure sensing line for differential pressure transmitters, resulting in both channels of high steam flow instrumentation for the EC System Loop 11 being declared inoperable. Although EC System Loop 11 was manually isolated, the conditions satisfying the automatic initiation of the EC System Loop 11 were not present, and no automatic isolation due to high steam flow was required.
The safety function of the EC High Steam Flow instruments (DPT-36-06C and DPT-36-06D) is to isolate EC System Loop 11 in the event of an EC System Loop 11 steam leak. Each of these transmitters monitors steam flow to the EC System Loop 11 and each provides an isolation signal on high steam flow indicative of a line break. Normally, with no steam flow, as is the case with the EC in a standby configuration, the differential pressure seen by the transmitters is approximately zero. The sensing lines are a static column of water from the transmitters to where the instrument lines connect to the steam flow elbow. During normal operation, any minor leakage from the sensing lines is made up from condensation from the steam line to both the low and high pressure sensing lines.
The trip unit gross fail function is a design feature to monitor for an off-normal electronic instrument loop component output. For both of these trip units, the signal at the trip unit was reading low and the gross fail could not be reset. A low output signal is non-conservative in that the high steam flow isolation may not occur when required by design. A low output reading is indicative of a lowered high pressure instrument line water column.
The amount of seat leakage has been evaluated for consideration of impact on containment requirements. The total leakage remains well within the allowable value to meet IOCFR50 Appendix J requirements. The leakage, if internal to the drywell, is monitored as part of the floor drain leak rate criteria, as required by Technical Specifications.
Currently the instrument loops are operable with any minor leakage being maintained by condensing steam as designed. The instruments are calibrated and will perform their isolation function.
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 the NRC Reactor Oversight Process (ROP) Performance Indicators (PIs) for NMP I Safety System Functional Failures (SSFF). The SSFF PI will increase from 4 to 5 and remains green. The green to white threshold for this PI is greater than 6. No other NRC performance indicators were impacted by this event.
NRC FORM 366 (10-2010)
IV. CORRECTIVE ACTIONS
A. ACTION TAKEN TO RETURN AFFECTED SYSTEMS TO PRE-EVENT NORMAL STATUS:
- 1.
Transmitters DPT-36-06C and DPT-36-06D were equalized via the equalizing valve and the trip units verified to be reading normally.
- 2.
EC System Loop 11 was returned to a standby lineup.
B. ACTION TAKEN OR PLANNED TO PREVENT RECURRENCE:
- 1.
A drywell walkdown of the high pressure sensing line for differential pressure transmitters DPT-36-06C and DPT-36-06D was performed on July 9, 2013, during the recent planned outage confirming that there are no leaks from the sensing line into the drywell.
- 2.
An interim methodology has been developed and is being implemented to monitor the operability of the transmitters DPT-36-06C and DPT-36-06D instrument loops for times when the instrument is required by TS and there is no reactor steam for instrument line makeup.
- 3.
Replace instrument blowdown valves VLV-36-374 and VLV-89-72 during the next refueling outage at NMP1.
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.
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 IEEE 803 FUNCTION IDENTIFIER N/A IEEE 805 SYSTEM IDENTIFICATION JC EC System Differential Pressure Transmitter D. SPECIAL COMMENTS:
None PDT BL NRC FORM 366 (10-2010)
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05000220/LER-2013-001, Emergency Condenser 11 High Steam Flow Isolation Instrumentation Loss During Plant Startup | Emergency Condenser 11 High Steam Flow Isolation Instrumentation Loss During Plant Startup | 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-2013-001, Regarding Reactor Core Isolation Cooling System Isolation Due to a Temperature Switch Unit Failure | Regarding Reactor Core Isolation Cooling System Isolation Due to a Temperature Switch Unit 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)(viii)(A) 10 CFR 50.73(a)(2)(viii)(B) | 05000220/LER-2013-002, Regarding Unanalyzed Condition Caused by Unfused Control Room DC Ammeters | Regarding Unanalyzed Condition Caused by Unfused Control Room DC Ammeters | 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-2013-002, For Nine Mile Point Unit 2, Regarding Failure of High Pressure Core Spray System Pressure Pump Due to Motor Winding Failure | For Nine Mile Point Unit 2, Regarding Failure of High Pressure Core Spray System Pressure Pump Due to Motor Winding 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)(viii)(A) 10 CFR 50.73(a)(2)(viii)(B) | 05000410/LER-2013-003, Regarding Unanalyzed Condition Caused by Unfused Control Room DC Ammeters | Regarding Unanalyzed Condition Caused by Unfused Control Room DC Ammeters | 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-2013-004, Regarding Manual Reactor Protection System Actuation Due to Loss of Reactor Recirculation Flow | Regarding Manual Reactor Protection System Actuation Due to Loss of Reactor Recirculation Flow | 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-2013-005, Regarding Secondary Containment Inoperabilities Due to Differential Pressure Not Meeting Technical Specification Surveillance Requirement 3.6.4.1.1 | Regarding Secondary Containment Inoperabilities Due to Differential Pressure Not Meeting Technical Specification Surveillance Requirement 3.6.4.1.1 | 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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