11-29-2017 | From May 2, 2016 to May 6, 2016, when B Train 122 Control Room Chiller ( CRC) was out-of-service ( OOS) per Technical Specifications (Tech Specs) 3.7.11 Condition A, Unit 2 A Train 23 Containment Fan Coil Unit (FCU) was OOS. According to revision 41 of site procedure C18.1, "Engineered Safeguards Equipment Support Systems," Bus 16 load sequencer and Bus 121 were inoperable when 122 CRC was OOS. Bus 121 supports B Train Diesel Driven Cooling Water Pump and Unit 2 B Train containment cooling (22/24 FCUs). So both trains of containment FCUs were OOS at the same time for approximately 35. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. This would have required entry into LCO 3.0.3 putting Unit 2 in MODE 3 within 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br />, this did not occur. This event is reportable under 10 CFR 50.73(a)(2)(i)(B), Operation or Condition Prohibited by Tech Specs.
The cause was that the Senior Reactor operators failed to utilize Human Performance Tools (Verification/Validation and Procedure Use/Adherence) when assessing the Tech Specs impact to Unit 2 for applying LCO 3.0.6 when 122 CRC was taken OOS.
Corrective actions include independent assessment of shared system LCO's for each unit, revising the LCO database, established a standard for LCO 3.0.6 log entries, and revising the safety function determination program to be more user friendly. |
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Category:Letter
MONTHYEARIR 05000282/20230042024-02-0101 February 2024 Integrated Inspection Report 05000282/2023004 and 05000306/2023004 ML23356A1232024-01-29029 January 2024 Exemption from Select Requirements of 10 CFR Part 73 (Security Notifications, Reports, and Recordkeeping and Suspicious Activity Reporting) ML24024A0722024-01-24024 January 2024 Independent Spent Fuel Storage Installation, Onticello, Supplement to Request for Exemption from Enhanced Weapons, Firearms Background Checks, and Security Event Notifications Implementation ML24017A0182024-01-19019 January 2024 Confirmation of Initial License Examination ML23356A0032024-01-17017 January 2024 Issuance of Amendments Revise Technical Specification 5.6.6, Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR) L-PI-23-034, Response to Request for Additional Information Regarding License Amendment Request to Revise Technical Specification 3.7.8, Cooling Water (Cl) System,2024-01-0202 January 2024 Response to Request for Additional Information Regarding License Amendment Request to Revise Technical Specification 3.7.8, Cooling Water (Cl) System, L-PI-23-035, Response to Request for Additional Information Regarding License Amendment Request to Revise Technical Specifications Section 5.6.6, Reactor Coolant System (RCS) Pressure and Temperature Limits Report2023-12-20020 December 2023 Response to Request for Additional Information Regarding License Amendment Request to Revise Technical Specifications Section 5.6.6, Reactor Coolant System (RCS) Pressure and Temperature Limits Report IR 07200010/20234012023-12-20020 December 2023 Independent Spent Fuel Storage Installation Security Inspection Report 07200010/2023401 ML23349A0572023-12-15015 December 2023 and Independent Spent Fuel Storage Installation, Revision to Correspondence Service List for Northern States Power - Minnesota ML23215A1672023-12-15015 December 2023 Acceptance of Requested Licensing Action Amendment Request to Revise Surveillance Requirement 3.8.1.2 Note 3 IR 05000282/20234012023-12-13013 December 2023 Security Baseline Inspection Report 05000282/2023401 and 05000306/2023401 L-PI-23-033, Supplement to License Amendment Request to Revise Technical Specification 3.8.1, Surveillance Requirement 3.8.1.2, Note 32023-12-0505 December 2023 Supplement to License Amendment Request to Revise Technical Specification 3.8.1, Surveillance Requirement 3.8.1.2, Note 3 ML23304A1632023-11-15015 November 2023 Supplemental Information Needed for Acceptance of Requested Licensing Action Amendment Request to Revise SR 3.8.1.2 Note 3 ML23319A3182023-11-15015 November 2023 Request for Exemption from Enhanced Weapons, Firearms Background Checks, and Security Event Notifications Implementation IR 05000282/20230032023-11-0808 November 2023 Integrated Inspection Report 05000282/2023003 and 05000306/2023003 ML23311A3572023-11-0707 November 2023 Core Operating Limits Report (COLR) for Prairie Island Nuclear Generating Plant (PINGP) Unit 2. Cycle 33. Revision 0 ML23285A3062023-10-12012 October 2023 Implementation of the Fleet Standard Emergency Plan for the Monticello Nuclear Generating Plant and the Prairie Island Nuclear Generating Plant ML23270B9022023-09-29029 September 2023 Review of Reactor Vessel Material Surveillance Program Capsule N Technical Report L-PI-23-025, License Amendment Request to Revise Technical Specification 3.8.1, Surveillance Requirement 3.8.1.2, Note 32023-09-28028 September 2023 License Amendment Request to Revise Technical Specification 3.8.1, Surveillance Requirement 3.8.1.2, Note 3 ML23265A2532023-09-26026 September 2023 Review of Reactor Vessel Material Surveillance Program Capsule N Technical Report ML23262B0372023-09-19019 September 2023 Response to NRC Request for Additional Information Regarding the 2023 Monticello and Prairie Island Plant Decommissioning Funding Status Reports ML23256A1682023-09-13013 September 2023 Independent Spent Fuel Storage Installation and Monticello Nuclear Generating Plant - Voluntary Security Clearance Program 2023 Insider Threat Program Self-Inspection L-PI-23-023, Baffle Former Bolts Alternate Aging Management Strategy2023-09-11011 September 2023 Baffle Former Bolts Alternate Aging Management Strategy IR 05000282/20230052023-08-30030 August 2023 Updated Inspected Plan for Prairie Island Nuclear Generating Plant Report 05000282/2023005 and 05000306/2023005 IR 05000282/20230102023-08-17017 August 2023 NRC Inspection Report 05000282/2023010 and 05000306/2023010 ML23222A0122023-08-10010 August 2023 Independent Spent Fuel Storage Installation and Monticello Nuclear Generating Plant - Changes in Foreign Ownership, Control or Influence IR 05000282/20230022023-08-0303 August 2023 Integrated Inspection Report 05000282/2023002 and 05000306/2023002 ML23214A2032023-08-0202 August 2023 Request for Information for an NRC Quadrennial Comprehensive Engineering Team Inspection: Inspection Report 05000282/2024010; 05000306/2024010 ML23206A2342023-07-25025 July 2023 Independent Spent Fuel Storage Installation, and Monticello Nuclear Generating Plant, Changes in Foreign Ownership, Control or Influence ML23202A0032023-07-21021 July 2023 Independent Spent Fuel and Independent Spent Fuel Storage Installation, Monticello Nuclear Generating Plant, Submittal of Quality Assurance Topical Report (NSPM-1) ML23199A0922023-07-18018 July 2023 Notification of NRC Baseline Inspection and Request for Information; Inspection Report 05000306/2023004 ML23195A1732023-07-14014 July 2023 Revision of Standard Practice Procedures Plan L-PI-23-018, License Amendment Request to Revise ISFSI Technical Specification 4.4 to Allow Use of a Code Alternative to ASME Code, NB-5130, Examination of Weld Edge Preparation Surfaces, for the TN-40HT2023-07-14014 July 2023 License Amendment Request to Revise ISFSI Technical Specification 4.4 to Allow Use of a Code Alternative to ASME Code, NB-5130, Examination of Weld Edge Preparation Surfaces, for the TN-40HT ML23181A0192023-06-30030 June 2023 Independent Spent Fuel Storage Installation, Revision to Correspondence Service List for Northern States Power - Minnesota ML23178A2422023-06-28028 June 2023 Reassignment of the U.S. Nuclear Regulatory Commission Branch Chief in the Division of Operating Reactor Licensing for Plant Licensing Branch III L-PI-23-006, License Amendment Request to Revise Technical Specification 3.7.8 Required Actions2023-06-22022 June 2023 License Amendment Request to Revise Technical Specification 3.7.8 Required Actions L-PI-23-016, 2022 10 CFR 50.46 LOCA Annual Report2023-06-14014 June 2023 2022 10 CFR 50.46 LOCA Annual Report IR 05000282/20234202023-06-0101 June 2023 Security Baseline Inspection Report 05000282/2023420 and 05000306/2023420 ML23150A1722023-05-30030 May 2023 Preparation and Scheduling of Operator Licensing Examinations 2024-02-01
[Table view] Category:Licensee Event Report (LER)
MONTHYEARML19353A4092019-12-19019 December 2019 Technical Specification 5.6.8 Special Report: Inoperable Containment Isolation Valve Indication Supplemental Report 05000306/LER-2017-0032018-01-11011 January 2018 Both Containment SEa) Pump Control Switches in Pull-out in Mode 4, LER 17-003-00 for Prairie Island Nuclear Generating Plant, Unit 2 Regarding Both Containment Spray Pump Control Switches in Pull-Out in Mode 4 05000306/LER-2017-0022017-12-11011 December 2017 Reactor Coolant System Shutdown Communication Live Vent Through Wall Defect, LER 17-002-00 for Prairie Island, Unit 2, Regarding Reactor Coolant System Shutdown Communication Live Vent Through Wall Defect 05000306/LER-2017-0012017-11-29029 November 2017 23 Containment Fan Coil Unit Operability, LER 17-001-00 for Prairie Island, Unit 2, Regarding 23 Containment Fan Coil Unit Operability 05000282/LER-2016-0062017-02-15015 February 2017 121 Motor Driven Cooling Water Pump Auto Start, LER 16-006-00 for Prairie Island Nuclear Generating Plant, Units 1 and 2 Regarding 121 Motor Driven Cooling Water Pump Auto Start 05000306/LER-2015-0022016-06-22022 June 2016 21 Feedwater Pump Lockout, Unit 2 Reactor Trip Due to Pressure Switch Failure, LER 15-002-01 for Prairie Island, Unit, Regarding 21 Feedwater Pump Lockout, Reactor Trip Due to Pressure Switch Failure 05000282/LER-2016-0042016-06-21021 June 2016 1 OF 4, LER 16-004-00 for Prairie Island, Unit 1, Regarding Missing Fire Barrier Between Fire Area (FA) 59 and 85 / Fire Hazard Analysis Drawings Do Not Match Boundary Description 05000282/LER-2016-0022016-03-25025 March 2016 Listed System Actuation - Motor-Driven Cooling Water Pump Auto-Start, LER 16-002-00 for Prairie Island, Unit 1, Regarding Listed System Actuation - Motor-Driven Cooling Water Pump Auto-Start 05000306/LER-2016-0012016-02-12012 February 2016 Unit 2 Reactor Trip due to a Ground Fault resulting in a Generator Trip, LER 16-001-00 for Prairie Island, Unit 2, Regarding Reactor Trip Due to a Ground Fault Resulting in a Generator Trip L-PI-15-071, Cancellation of License Event Report (LER)50-282/2015-001-00, 14 Fan Coil Unit Leak2015-09-0303 September 2015 Cancellation of License Event Report (LER)50-282/2015-001-00, 14 Fan Coil Unit Leak L-PI-11-023, Reinstatement of Licensee Event Reports Associated with Flooding Scenarios2011-03-31031 March 2011 Reinstatement of Licensee Event Reports Associated with Flooding Scenarios L-PI-07-101, LER 07-03-001 for Prairie Island, Unit 1 Regarding Unanalyzed Condition Due to Breached Fire Barrier2008-01-25025 January 2008 LER 07-03-001 for Prairie Island, Unit 1 Regarding Unanalyzed Condition Due to Breached Fire Barrier L-PI-06-046, Cancellation of Licensee Event Report (LER) 1-05-01, Discovery of Single Failure Vulnerability of Unit 1 Safeguards Buses2006-06-0909 June 2006 Cancellation of Licensee Event Report (LER) 1-05-01, Discovery of Single Failure Vulnerability of Unit 1 Safeguards Buses 2019-12-19
[Table view] |
comments regarding burden estimate to the Information Services Branch (T-2 F43), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by e-mail to used to impose an information collection does not display a currently valid OMB control number, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.
Prairie Island Nuclear Generating Plant 05000-306 NUMBER NO.
Unit 2 2017 - 001 - 00
DESCRIPTION OF EVENT
During an extent of condition review for an issue with correct application of Technical Specifications (Tech Specs) LCO 3.0.6 that occurred August of 2017, a similar condition was discovered. While 122 Control Room Chiller (CRC) was out-of- service (OOS) due to chiller oil temperature outside operability limit per Tech Specs 3.7.11 Condition A, from May 2, 2016 to May 6, 2016, 23 Containment Fan Coil Unit (FCU1) was OOS due to a problem with the discharge damper. According to guidance in C18.1, "Engineered Safeguards Equipment Support Systems", Bus 16 load sequencer and Bus 121 are inoperable when 122 CRC is OOS (safeguards room cooling is provided by CRC). Supported systems including 21 Safeguards Screenhouse Roof Exhaust Fan (powered from Bus 121), supported B Train Diesel Driven Cooling Water Pump, and B Train containment cooling (22/24 FCUs) were also OOS.
Having both trains of containment FCUs OOS at the same time for approximately 35.6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> would require entry into LCO 3.0.3 for Unit 2, actions required to be in MODE 3 within 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br />; MODE 4 within 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br />; and MODE 5 within 37 hours4.282407e-4 days <br />0.0103 hours <br />6.117725e-5 weeks <br />1.40785e-5 months <br />.
This did not occur. The Senior Reactor Operator failed to correctly assess the Technical Specifications (Tech Specs) impact to Unit 2 when applying Tech Specs 3.0.6. This event is reportable under 10 CFR 50.73(a)(2)(i)(B), Operation or Condition Prohibited by Tech Specs.
EVENT ANALYSIS
The event is reportable under 10 CFR 50.73(a)(2)(i)(B). The licensee shall report any operation or condition which was prohibited by the plant's Tech Specs. This condition meets the reporting criteria because both trains of containment FCUs were OOS at the same time for approximately 35.6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, this required entry into LCO 3.0.3, putting Unit 2 in MODE 3 within 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br />. Tech Specs 3.0.3 for Limiting Condition for Operation was not entered and the required actions were not initiated within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> to place the unit, as applicable, in MODE 3 in 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br />; MODE 4 within 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br />; and MODE 5 within 37 hours4.282407e-4 days <br />0.0103 hours <br />6.117725e-5 weeks <br />1.40785e-5 months <br />.
The Containment Air Cooling System consists of four fan coil units, a duct distribution system, and the associated instrumentation and controls. During normal operation the fans may be run at high or low speed and during post-accident conditions the fans run at low speed. The Containment Air Cooling System is designed to recirculate and cool the containment atmosphere in the event of a loss-of-coolant or main steam line break accident and thereby ensure that the containment pressure cannot exceed its design value of 46 psig at 268 degrees F (100% relative humidity).
Two of the four containment cooling units and one containment spray pump provide sufficient heat removal capability to maintain the post-accident containment pressure and temperature below the design value, assuming that the core residual heat is released to the containment as steam. Analysis has shown that the operation of one containment spray pump during the injection phase and the heat removal capability equivalent to a single fan coil unit at maximum fouling conditions is sufficient to maintain containment pressure less than design. While B Train FCU's 22 and 24 were OOS and A Train FCU 23 was OOS, A Train FCU 21 was operable. Because analysis has shown that single FCU with Containment Spray is sufficient to maintain containment pressure and 21 FCU was operable the event is not reportable under 10 CFR 50.73(a)(2)(v)(D) as an event or 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 (safety system functional failure). Even though 23, 22 and 24 FCU's were inoperable per Tech Specs, 21 FCU was operable and the ability to mitigate postulated accidents was not lost and the system was not in an unanalyzed condition as described 10 CFR 50.73(a)(2)(ii).
IEEE Component Code — FCU comments regarding burden estimate to the Information Services Branch (T-2 F43), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by e-mail to used to impose an information collection does not display a currently valid OMB control number, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.
3. LER NUMBER 1. FACILITY NAME Prairie Island Nuclear Generating Plant 05000-306 - 001 2017 - 00
SAFETY SIGNIFICANCE
Safety function was not lost, because with both B Train FCU's and one A Train FCU inoperable per Tech Specs, the ability to mitigate postulated accidents was not lost and the system was not in an unanalyzed condition. Analysis has shown that the operation of one containment spray pump during the injection phase and the heat removal capability equivalent to a single fan coil unit at maximum fouling conditions is sufficient to maintain containment pressure less than design.
There were no radiological, environmental, or industrial impacts associated with this event, and PINGP did not adversely affect the health and safety of the public. This event report does not identify any safety system functional failures.
CAUSE
Cause evaluation determined that the Senior Reactor operators failed to utilize Human Performance Tools (Verification/Validation and Procedure Use/Adherence) when assessing the Technical Specification impact to Unit 2 for applying LCO 3.0.6 when 122 CRC was taken 00S.
CORRECTIVE ACTION(s) 1. Revise operations work instructions SWI 0-200.3, TECHNICAL SPECIFICATION ENTRY & EXIT to require independent assessment of shared system LCO's for each unit. This action is complete.
2. Revise the LCO database to Limit the use of "Unit 0" to ISFSI Technical Specifications. This action is complete.
3. Establish the standard that LCO 3.0.6 log entries, carried over a shift, are in Narrative Logs using the Open Item option.
This action is complete 4. Revise site work instruction 5AWI 3.15.8, SAFETY FUNCTION DETERMINATION PROGRAM to be more user friendly.
Including graphical explanations. This action is expected to be completed by the end of the year.
PREVIOUS SIMILAR EVENTS
A review of the Corrective Action Program (CAP) and Licensee Event Reports (LERs) for PINGP revealed one similar event over the last three years.
On September 11, 2015, it was identified that 122 Control Room Chiller was removed from service and control valve CV-31837 (121/122 Control Room Chiller Outlet) and CV-31838 (121/122 Control Room Chiller Inlet) were closed. This isolated Train B Safeguards Chilled Water and rendered Bus 16 Unit Cooler non-functional, which will result in unacceptable temperatures in the associated bus room during a postulated High Energy Line Break (HELB). Bus 16 would not have performed its safety function and was inoperable for greater than the time allowed by Tech Specs. Tech Specs 3.8.9 for Distribution Systems-Operating was not entered and the required actions were not taken to restore to an operable status within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> or to enter MODE 3 in 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and MODE 5 in 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. This is a reportable event under 10 CFR 50.73(a)(2)(i)(b), Operation or Condition Prohibited by Tech Specs.
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05000306/LER-2017-001 | 23 Containment Fan Coil Unit Operability LER 17-001-00 for Prairie Island, Unit 2, Regarding 23 Containment Fan Coil Unit Operability | 10 CFR 50.73(a)(2)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2)(i)(b) 10 CFR 50.73(a)(2)(ii) | 05000306/LER-2017-002 | Reactor Coolant System Shutdown Communication Live Vent Through Wall Defect LER 17-002-00 for Prairie Island, Unit 2, Regarding Reactor Coolant System Shutdown Communication Live Vent Through Wall Defect | 10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded | 05000306/LER-2017-003 | Both Containment SEa) Pump Control Switches in Pull-out in Mode 4 LER 17-003-00 for Prairie Island Nuclear Generating Plant, Unit 2 Regarding Both Containment Spray Pump Control Switches in Pull-Out in Mode 4 | 10 CFR 50.73(a)(2)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications |
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