01-21-2016 | On November 24, 2015 at 0534 hours0.00618 days <br />0.148 hours <br />8.829365e-4 weeks <br />2.03187e-4 months <br />, the Monticello Nuclear Generating Plant was at 0% power in Mode 3 (Hot Shutdown) for a forced outage. While initially placing Shutdown Cooling ( SDC) in service, the 12 Residual Heat Removal ( RHR) pump tripped approximately 8-10 seconds after start due to the closure of the RHR SDC suction isolation valves. When placing SDC in service, flow rapidly increased after opening the RHR Division 2 Low Pressure Coolant Injection ( LPCI) outboard injection valve causing a localized pressure transient in the reactor recirculation pump suction piping that resulted in an isolation of the SDC suction line. Reactor pressure vessel ( RPV) pressure remained stable at approximately 30 psig.
Prior to attempting to place 'B' SDC in service, the Condensate system and the 'F' Safety Relieve Valve were in service providing decay heat removal. Immediate actions were taken to restore 'B' RHR SDC to operable status, thus an alternative method of decay heat removal was already established by the Condensate system and `F' Safety Relief Valve. |
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Category:Letter
MONTHYEARIR 05000263/20240052024-08-30030 August 2024 Updated Inspection Plan and Follow-Up Letter for Monticello Nuclear Generating Plant, Unit 1 (Report 05000263/2024005) L-MT-24-028, Response to RCI for RR-017 ISI Impracticality2024-08-28028 August 2024 Response to RCI for RR-017 ISI Impracticality 05000263/LER-2024-002, Low Pressure Coolant Injection Inoperable Due to Motor Valve Failure2024-08-27027 August 2024 Low Pressure Coolant Injection Inoperable Due to Motor Valve Failure ML24222A1822024-08-27027 August 2024 – Proposed Alternative Request VR-09 to the Inservice Testing Requirements of the ASME OM Code for Main Steam Safety Relief Valves IR 05000263/20244202024-08-21021 August 2024 Security Baseline Inspection Report 05000263/2024420 - Cover Letter IR 05000263/20240022024-08-14014 August 2024 Integrated Inspection Report 05000263/2024002 ML24218A2282024-08-0505 August 2024 Request for Confirmation of Information for Relief Request RR-017, Inservice Inspection Impracticality During the Fifth Ten-Year Interval ML24208A1502024-07-26026 July 2024 Independent Spent Fuel Storage Installation - Submittal of Quality Assurance Topical Report (NSPM-1) ML24215A2992024-07-23023 July 2024 Minnesota State Historic Preservation Office Comments on Monticello SLR Draft EIS ML24198A2372024-07-18018 July 2024 Information Request to Support Upcoming Biennial Problem Identification and Resolution (Pi&R) Inspection at Monticello Nuclear Generating Plant L-MT-24-022, – Preparation and Scheduling of Operator Licensing Examinations2024-07-0909 July 2024 – Preparation and Scheduling of Operator Licensing Examinations ML24164A2402024-06-10010 June 2024 Minnesota State Historic Preservation Office- Comments on Draft Monticello SLR Draft EIS L-MT-24-019, Submittal of ASME Section XI, Section IWB-3720 Analytical Evaluation in Accordance with 10 CFR 50.55a(b)(2)(xliii)2024-06-10010 June 2024 Submittal of ASME Section XI, Section IWB-3720 Analytical Evaluation in Accordance with 10 CFR 50.55a(b)(2)(xliii) L-MT-24-017, Response to Request for Additional Information Regarding License Amendment Request to Revise Monticello Technical Specification Surveillance Requirement 3.8.6.6 (EPID-L-2023-LLA-01602024-06-0404 June 2024 Response to Request for Additional Information Regarding License Amendment Request to Revise Monticello Technical Specification Surveillance Requirement 3.8.6.6 (EPID-L-2023-LLA-0160 ML24137A2792024-06-0303 June 2024 Audit Summary for License Amendment Request to Revise Technical Specification 3.8.6, Battery Parameters, Surveillance Requirement 3.8.6.6 IR 05000263/20244012024-05-30030 May 2024 Public - Monticello Nuclear Generating Plant - Cyber Security Inspection Report 05000263/2024401 ML24141A1292024-05-22022 May 2024 Northern States Power Company - Use of Encryption Software for Electronic Transmission of Safeguards Information ML24141A1782024-05-20020 May 2024 Information Request to Support the NRC Annual Baseline Emergency Action Level and Emergency Plan Changes Inspection L-MT-24-015, Response to Request for Additional Information - Alternative Request VR-09 for OMN-172024-05-16016 May 2024 Response to Request for Additional Information - Alternative Request VR-09 for OMN-17 L-MT-24-013, 2023 Annual Radiological Environmental Operating Report2024-05-14014 May 2024 2023 Annual Radiological Environmental Operating Report ML24135A1902024-05-14014 May 2024 Submittal of 2023 Annual Radioactive Effluent Release Report IR 05000263/20240102024-05-13013 May 2024 Age-Related Degrading Inspection Report 05000263/2024010 ML24127A1472024-05-0909 May 2024 Letter to Mille Lacs Band- Section 106 Consultation Regarding the Monticello Nuclear Generating Plant Subsequent License Renew. Application ML24128A0042024-05-0909 May 2024 Letter to Minnesota State Historic Preservation Office- Section 106 Consultation Regarding the Monticello Nuclear Generating Plant Subsequent License Renewal Application L-MT-24-016, 2023 Annual Report of Individual Monitoring for the Monticello Nuclear Generating Plant (MNGP)2024-05-0808 May 2024 2023 Annual Report of Individual Monitoring for the Monticello Nuclear Generating Plant (MNGP) ML24089A2382024-04-29029 April 2024 Summary of Nuclear Property Insurance IR 05000263/20240012024-04-29029 April 2024 Plan - Integrated Inspection Report 05000263/2024001 L-MT-24-012, Reactor Scram, Containment Isolation, and Cooldown Rate Outside of Limits Following Technician Adjustment of Wrong Component2024-04-25025 April 2024 Reactor Scram, Containment Isolation, and Cooldown Rate Outside of Limits Following Technician Adjustment of Wrong Component ML24115A2922024-04-25025 April 2024 Sec106 Tribal, Jensvold, Kevin-Upper Sioux Community ML24115A2912024-04-25025 April 2024 Sec106 Tribal, Jacskon, Sr., Faron-Leech Lake Band of Ojibwe ML24115A2882024-04-25025 April 2024 Sec106 Tribal, Fairbanks, Michael-White Earth Nation.Docx ML24115A3032024-04-25025 April 2024 Sec106 Tribal, Taylor, Louis-Lac Courte Oreilles Band of Lake Superior Chippewa Indians ML24115A3052024-04-25025 April 2024 Sec106 Tribal, Vanzile, Jr., Robert-Sokaogon Chippewa Community ML24115A3012024-04-25025 April 2024 Sec106 Tribal, Seki, Darrell-Red Lake Nation ML24115A2872024-04-25025 April 2024 Sec106 Tribal, Dupuis, Kevin-Fond Du Lac Band of Lake Superior Chippewa ML24115A3002024-04-25025 April 2024 Sec106 Tribal, Rhodd, Timothy-Iowa Tribe of Kansas and Nebraska ML24115A2962024-04-25025 April 2024 Sec106 Tribal, Larsen, Robert-Lower Sioux Indian Community ML24115A2932024-04-25025 April 2024 Sec106 Tribal, Johnson, Grant-Prairie Island Indian Community ML24115A2892024-04-25025 April 2024 Sec106 Tribal, Fowler, Thomas-St. Croix Chippewa of Wisconsin ML24115A3022024-04-25025 April 2024 Sec106 Tribal, Stiffarm, Jeffrey-Fort Belknap Indian Community ML24115A3072024-04-25025 April 2024 Sec106 Tribal, Williams, Jr., James-Lac Vieux Desert Band of Lake Superior Chippewa Indians ML24115A2942024-04-25025 April 2024 Sec106 Tribal, Johnson, John-Lac Du Flambeau Band of Lake Superior Chippewa Indians ML24115A3062024-04-25025 April 2024 Sec106 Tribal, Wassana, Reggie-Cheyenne and Arapaho Tribes ML24115A2992024-04-25025 April 2024 Sec106 Tribal, Renville, J. Garret-Sisseton Wahpeton Oyate of the Lake Travers Reservation ML24115A2902024-04-25025 April 2024 Sec106 Tribal, Jackson-Street, Lonna-Spirit Lake Nation ML24115A2952024-04-25025 April 2024 Sec106 Tribal, Kakkak, Gena-Menominee Indian Tribe of Wisconsin ML24115A2972024-04-25025 April 2024 Sec106 Tribal, Miller, Cole-Shakopee Mdewakanton Sioux Community ML24115A2982024-04-25025 April 2024 Sec106 Tribal, Reider, Anthony-Flandreau Santee Sioux Tribe ML24106A1102024-04-24024 April 2024 Mille Lacs Band of Ojibwe -Monticello Sec106 Tribal ML24115A2792024-04-24024 April 2024 Sec106 Tribal, Blaker, Doreen-Keweenaw Bay Indian Community 2024-08-05
[Table view] Category:Licensee Event Report (LER)
MONTHYEAR05000263/LER-2024-002, Low Pressure Coolant Injection Inoperable Due to Motor Valve Failure2024-08-27027 August 2024 Low Pressure Coolant Injection Inoperable Due to Motor Valve Failure L-MT-24-012, Reactor Scram, Containment Isolation, and Cooldown Rate Outside of Limits Following Technician Adjustment of Wrong Component2024-04-25025 April 2024 Reactor Scram, Containment Isolation, and Cooldown Rate Outside of Limits Following Technician Adjustment of Wrong Component L-MT-23-053, Condition Prohibited by Technical Specifications Due to Inoperable Main Steam Line Low Pressure Isolation Switch2023-12-0404 December 2023 Condition Prohibited by Technical Specifications Due to Inoperable Main Steam Line Low Pressure Isolation Switch 05000263/LER-2017-0062018-01-12012 January 2018 Loss of Reactor Protection System Scram Function During Main Steam Isolation Valve and Turbine Stop Valve Channel Functional Tests due to Use of a Test Fixture, LER 17-006-00 for Monticello Regarding Loss of Reactor Protection System Scram Function During Main Steam Isolation Valve and Turbine Stop Valve Channel Functional Tests Due to Use of a Test Fixture 05000263/LER-2017-0052017-09-20020 September 2017 Diesel Generator Emergency Service Water System Automatic Transfer to Alternate Shutdown Panel, LER 17-005-00 for Monticello Nuclear Generating Plant Regarding Diesel Generator Emergency Service Water System Automatic Transfer to Alternate Shutdown Panel 05000263/LER-2015-0042017-08-22022 August 2017 Past Inoperability of Turbine Stop Valve Scram Function Exceeded Technical Specification Requirements, LER 15-004-01 for Monticello Regarding Past Inoperability of Turbine Stop Valve Scram Function Exceeded Technical Specification Requirements 05000263/LER-2017-0042017-08-16016 August 2017 High Pressure Coolant Injection Steam Stop Valve Failed to Open During Test, LER 17-004-00 for Monticello Regarding High Pressure Coolant Injection Steam Stop Valve Failed to Open During Test 05000263/LER-2017-0032017-06-14014 June 2017 Main Steam Isolation Valve Leakage Exceeds Technical Specification Limits, LER 17-003-00 for Monticello Regarding Main Steam Isolation Valve Leakage Exceeds Technical Specification Limits 05000263/LER-2017-0022017-06-13013 June 2017 Main Steam Isolation Valve Closure Time Outside of Technical Specification Requirements, LER 17-002-00 for Monticello Nuclear Generating Plant Regarding Main Steam Isolation Valve Closure Time Outside of Technical Specification Requirements 05000263/LER-2017-0012017-06-13013 June 2017 Reactor Scram and Group II Isolation Due to 11 Reactor Feed Pump (RFP) Removal from Service with 12 RFP Isolated, LER 17-001-00 for Monticello Regarding Reactor Scram and Group II Isolation Due to 11 Reactor Feed Pump (RFP) Removal from Service with 12 RFP Isolated 05000263/LER-2016-0012017-05-25025 May 2017 High Pressure Coolant Injection System Cracked Pipe Nipple Caused Oil Leak, LER 16-001-02 for Monticello Regarding High Pressure Coolant Injection System Cracked Pipe Nipple Caused Oil Leak 05000263/LER-2016-0032017-05-25025 May 2017 HPCI Declared Inoperable Due to Excessive Water Level in Turbine, LER 16-003-01 for Monticello Regarding HPCI Declared Inoperable Due to Excessive Water Level in Turbine 05000263/LER-2016-0022016-09-30030 September 2016 Inadequate Appendix R Fire Barrier Impacts Safe Shutdown Capability, LER 16-002-00 for Monticello Regarding Inadequate Appendix R Fire Barrier Impacts Safe Shutdown Capability 05000263/LER-2014-0022016-07-13013 July 2016 Torus to Drywell Vacuum Breaker Did Not Indicate Closed During Testing, LER 14-002-01 for Monticello Regarding Torus to Drywell Vacuum Breaker Did Not Indicate Closed During Testing 05000263/LER-2014-0032016-07-13013 July 2016 Torus to Drywell Vacuum Breaker Dual Indication During Testing, LER 14-003-01 for Monticello Nuclear Generating Plant RE: Torus to Drywell Vacuum Breaker Dual Indication During Testing 05000263/LER-2015-0072016-01-21021 January 2016 Loss of Residual Heat Removal Capability, LER 15-007-00 for Monticello Regarding Loss of Residual Heat Removal Capability 05000263/LER-2015-0062016-01-21021 January 2016 - Reactor Scram due to Group 1 Isolation from Foreign Material in the Main Steam Flow Instrument Line, LER 15-006-00 for Monticello Regarding Reactor Scram due to Group 1 Isolation from Foreign Material in the Main Steam Flow Instrument Line ML1015505712009-09-12012 September 2009 Event Notification for Monticello on State Offsite Notification Due to Not Meeting Permit Requirements L-MT-05-035, LER 50-004-00 for Monticello Nuclear Generating Plant Regarding Voluntary LER for Control Rod Drive Insert Line Leakage2005-05-12012 May 2005 LER 50-004-00 for Monticello Nuclear Generating Plant Regarding Voluntary LER for Control Rod Drive Insert Line Leakage ML0216100952002-05-15015 May 2002 LERs 02-001-01 & 02-002-01 for Monticello Nuclear Generating Plant Re Mechanical Pressure Regulator Failure Causes Reactor Scram & Application of Instrument Deviation Acceptance Criteria Allowed As-Found Settings to Be Outside Tech Spec Val 2024-08-27
[Table view] |
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EVENT DESCRIPTION
On November 24, 2015 at 0534 hours0.00618 days <br />0.148 hours <br />8.829365e-4 weeks <br />2.03187e-4 months <br />, the Monticello Nuclear Generating Plant was at 0% power in Mode 3 (Hot Shutdown) for a forced outage. While initially placing Shutdown Cooling (SDC) [KE] in service, the 12 Residual Heat Removal (RHR) [BO] pump [P] tripped approximately 8-10 seconds after start due to the closure of the RHR SDC suction isolation valves [ISV].
When placing SDC in service, flow was rapidly increased after opening the RHR Division 2 Low Pressure Coolant Injection (LPCI) outboard injection valve [INV], causing a localized pressure transient in the reactor recirculation pump suction piping that resulted in an isolation of the SDC suction line. The RHR High Reactor Pressure annunciator [PA] was received and immediately cleared as the pressure switch [PS] upstream of 12 Recirculation Pump [AD] Suction valve in the 'B' Recirculation Loop actuated causing a Group 2 containment isolation signal. However, this was not expected as Reactor Pressure Vessel (RPV) [RPV] steam dome pressure remained stable at approximately 30 psig.
At 0535 hours0.00619 days <br />0.149 hours <br />8.845899e-4 weeks <br />2.035675e-4 months <br />, immediate actions were taken to restore 'B' RHR SDC to operable status.
At 0545 hours0.00631 days <br />0.151 hours <br />9.011243e-4 weeks <br />2.073725e-4 months <br />, an alternative method of decay heat removal was established by utilizing the Condensate [SD] system and 'F' Safety Relief valve [RV].
At 1354 hours0.0157 days <br />0.376 hours <br />0.00224 weeks <br />5.15197e-4 months <br />, the 12 RHR pump and 12 RHR Service Water pump were successfully placed in service on SDC and the plant reached Mode 4 (Cold Shutdown) at 1428 hours0.0165 days <br />0.397 hours <br />0.00236 weeks <br />5.43354e-4 months <br />.
EVENT ANALYSIS
The event was determined to be reportable in accordance with 10 CFR 50.73(a)(2)(v)(B) as an Event or Condition that Could have Prevented the Fulfillment of the Safety Function of Structures or Systems that are Needed to Remove Residual Heat. This event is considered a Safety System Functional Failure per NEI 99-02, Revision 7.
SAFETY SIGNIFICANCE
Prior to attempting to place 'B' SDC in service, the Condensate system and the 'F' Safety Relieve Valve were in service providing decay heat removal. These systems remained in service and, as demonstrated by steadily lowering RPV pressure and temperature, provided adequate decay heat removal until SDC was placed in service. Additionally, the Reactor Water Cleanup System [CE] was available for decay heat reject if needed. After the closure of the SDC suction valves and subsequent trip of the 12 RHR pump, immediate actions were taken to restore SDC to operable status. Since the reactor remained adequately cooled, there were no actual consequences as a result of the initial failed attempt to place SDC in service. There was no impact to the health and safety of the public.
CAUSE
Both reactor high pressure SDC isolation pressure switches are located on the 'B' Recirculation Suction Piping. When initially placing SDC in service the LPCI outboard injection valve was opened and flow into the 'B' Recirculation system increased to approximately 4000 gpm in several seconds. This rapid flow increase caused a localized pressure transient in the 'B' Recirculation pump piping that resulted in the isolation of the SDC suction valves. Closure of the SDC suction valves subsequently caused a trip of the 12 RHR pump due to loss of pump suction. Written documentation in the operations manual did not adequately address the sensitivity of the pressure switches while placing 'B' SDC in service.
CORRECTIVE ACTION
Since the Condensate system and the 'F' Safety Relieve Valve were already in service providing decay heat removal, an alternate method of decay removal did not need to be established. Immediate actions were taken to restore 'B' SDC to operable status.
The Operations Manual used to place 'B' SDC in service was re-performed in its entirety to verify proper valve alignment, ensure the piping was full of water, and verify acceptable temperatures existed prior to attempting to place the system in service. This included venting the RHR suction and discharge lines prior to placing 'B' SDC in service. Existing procedural guidance allowed the associated LPCI injection valve to be slowly throttled open to achieve required RHR pump flow without introducing a pressure transient that would challenge the reactor high pressure SDC isolation setpoint.
The 12 RHR pump was successfully started and placed in SDC mode to cool down the plant to MODE 4. The Operations Manual has been updated to provide additional guidance for placing SDC in service including the pressure switch sensitivity to injection flow rate changes when changing the position of the LPCI outboard injection valve.
PREVIOUS SIMILAR EVENTS
There were no Licensee Event Reports with similar causes of loss SDC within the 3 last years.
ADDITIONAL INFORMATION
The Institute of Electrical and Electronics Engineer codes for equipment are denoted by [XX].