On November 12, 2017 at 2119, a Control Room board walkdown discovered that both of the Unit 2 Containment Spray Pump control switches had been left in pull-out, when operators transitioned Unit 2 from Mode 5 to Mode 4. With the control switches in pull-out, the pumps would not automatically start as required. Technical Specification (Tech Specs) 3.0.3 was entered as a result of not complying with Technical Specification 3.6.5, Containment Spray and Cooling systems, which required both trains of Containment Spray to be Operable while in Mode 4. This event is reportable under 10 CFR 50.73(a)(2)(i)(B), Condition Prohibited by Technical Specification and 10 CFR 50.73(a)(2)(v)(D), Event or Condition that Could Have Prevented Fulfillment of a Safety Function.
The root cause determined that Surveillance Procedure SP 2099, Unit 2 Main Steam Isolation Valve Logic Test, was not adequately designed to account for outage schedule variation. Contributing causes included that the Unit 2 Startup to Mode 4 procedure does not contain adequate process barriers such that plant configuration meets Technical Specification requirements for Mode 4 entry. Operations personnel failed to uphold standards for panel walkdown requirements.
Corrective actions include revising SP 2099, Unit 2 Main Steam Isolation Valve Logic Test to include steps to reposition Containment Spray Switches to the "as found" configuration and revise Unit 2 start-up procedure to add additional HOLD to have the Shift Manager perform Control Board Walkdown to verify equipment required in Mode 4 is aligned and Operable.
Develop and implement an operations improvement plan specifically targeted to improve Operator standards in the performance of Control Board Walkdowns. |
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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 2017 - 003 - 00
DESCRIPTION OF EVENT
On November 12, 2017, at 2119, a Control Room board walkdown discovered that both of the Unit 2 Containment Spray Pump control switches [CS-46560 (21 Containment Spray Pump) and CS-46561 (22 Containment Spray Pump)] had been left in pull-out, when operators transitioned Unit 2 from Mode 5 to Mode 4. With the control switches in pull-out, the pumps would not automatically start as required. Technical Specifications (Tech Specs) Limiting Conditions for Operation (LCO) 3.0.3 was entered at 2119 as a result of not complying with LCO 3.6.5, Containment Spray' and Cooling Systems, which required both trains of Containment Spray to be Operable while in Mode 4. Unit 2 had entered Mode 4 at 0303 on November 12, 2017. LCO 3.0.3 was exited at 2127 on November 12, 2017, when both Containment Spray Pump control switches were placed in automatic restoring Operability. Preliminary investigation determined that while Unit 2 was in Mode 5, Surveillance SP 2099, Main Steam Isolation Valve Logic Test, had taken the Containment Spray Pump control switches to pull-out but did not re-align the control switches to automatic after the test was completed. This was an 8-hour Non-Emergency report per 10 CFR 50.72(b)(3)(v)(D), Event or Condition that Could Have Prevented Fulfillment of a Safety Function.
EVENT ANALYSIS
The event is being reported under 10 CFR 50.73(a)(2)(i)(B), Condition Prohibited by Technical Specification, and 10 CFR 50.73(a)(2)(v)(D), Event or Condition that Could Have Prevented Fulfillment of a Safety Function. This condition meets the reporting criteria because Prairie Island Nuclear Generating Plant (PINGP) Unit 2 was not in full compliance with LCO 3.6.5, Containment Spray' and Cooling Systems, which required both trains of Containment Spray to be Operable while in Mode 4. Both trains of Containment Spray Pump control switches had been left in pull-out, which defeated the automatic start function for approximately 18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br /> and 24 minutes. Entry into Mode 4 while not meeting LCO 3.6.5 was contrary to LCO 3.0.4.
The primary purpose of the Containment Vessel Internal Spray System is to spray cool water into the containment atmosphere, in the event of a loss-of-coolant accident (LOCA) and thereby ensure that containment pressure does not exceed its design value of 46 psig at 268 degrees Fahrenheit.
SAFETY SIGNIFICANCE
A Westinghouse LOCA analysis concludes that the containment design pressure of 46 psig would not have been operated in Mode 4 with both CS pumps locked out. Further, the nominal spray setpoint of 23 psig would not have been reached. Also, two (2) trains of Containment Fan Coolers, operating with service water as high as 95 degrees Fahrenheit (well above actual conditions), would have been able to remove more than enough heat to offset the decay heat rate at 29 days after shutdown.
The Westinghouse Main Steam Line Break (MSLB) analysis also concluded that in the case of a full double-ended rupture MSLB at the SG discharge nozzle, the 46 psig Containment design pressure and the nominal spray setpoint 23 psig valves would not have been reached, while Unit 2 operated in Mode 4 with both CS pumps locked out.
The Westinghouse analysis that was performed demonstrated this event did not constitute a Safety System Functional Failure (SSFF). (Reference NEI 99-02, Revision 7, Regulatory Assessment Performance Indicator Guideline, Section 2.2, 1 EIIS Component Code - BE 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 - 003 - 00 Mitigating Systems Cornerstone, Safety System Functional Failures, Purpose.) As such, this event will not be reported in the NRC Performance Indicator (PI) for safety system functional failure since an engineering analysis was performed which determined that the long term Containment pressure would have remained below the design pressure, and automatic actuation of the containment spray system would not have been needed to mitigate the consequences of an accident, even though Prairie Island Unit 2 was not in full compliance with Tech Spec 3.6.5 for the Containment Spray and Cooling Systems.
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.
CAUSE(S)
Root Cause
- Surveillance Procedure SP 2099, Unit 2 Main Steam Isolation Valve Logic Test, is not adequately designed to account for outage schedule variation.
Contributing Causes
- Startup procedure 2C1.2-M4, Unit 2 Startup to Mode 4, does not contain adequate process barriers such that plant configuration meets LCO requirements for Mode 4 entry.
- Operations personnel failed to uphold standards such that the panel walkdown requirements are met to ensure that abnormal conditions and deviations from expected values/positions are communicated evaluated and understood by the Control Room Team.
CORRECTIVE ACTION(S)
- SP 2099 will be updated to include steps to reposition Containment Spray Switches to the "as found" configuration.
- Revise procedure 2C1.2-M4 to add additional HOLD in section 5.1.7, to have the Shift Manager perform Control Board Walkdown to verify equipment required in Mode 4 is aligned and Operable.
- Develop and Implement an operations improvement plan specifically targeted to improve Operator standards in the performance of Control Board Walkdowns.
PREVIOUS SIMILAR EVENTS
A review of the Corrective Action Program (CAP) and Licensee Event Reports (LERs) for PINGP revealed no results that involved having both control switches in pull out causing a reportable condition. The time frame for this review was eight years.
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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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