|Prairie Island Nuclear Generating Plant Unit 2|
|Reporting criterion:||10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications|
10 CFR 50.73(a)(2)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident
|3062017003R00 - NRC Website|
|Person / Time|
Northern States Power Company, Minnesota, Xcel Energy
Document Control Desk, Office of Nuclear Reactor Regulation
|Download: ML18011A041 (5)|
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.
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 approximately18 hours
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.
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.
- Surveillance Procedure SP 2099, Unit 2 Main Steam Isolation Valve Logic Test, is not adequately designed to account for outage schedule variation.
- 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.
- 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.