05000298/LER-2015-004
| Cooper Nuclear Station | |
| Event date: | 05-30-2015 |
|---|---|
| Report date: | 07-29-2015 |
| Reporting criterion: | 10 CFR 50.73(a)(2)(v)(B), Loss of Safety Function - Remove Residual Heat |
| 2982015004R00 - NRC Website | |
Reported lessons learned are incorporated into the licensing process and fed back to industry.
Send comments regarding burden estimate to the FOIA, Privacy and Information Collections Branch (T-5 F53), U.S. Nuclear Regulatory Commission, Washington, DC 20555- 0001, or by internet e-mail to Infocollects.Resource@nrc.gov, and to the Desk Officer, Office of Information and Regulatory Affairs, NEOB-10202, (3150-0104), Office of Management and Budget, Washington, DC 20503. If a means 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.
NRC
FORM 366 U.S. NUCLEAR REGULATORY COMMISSION (See Page 2 for required number of digits/characters for each block)
NO
Cooper Nuclear Station 05000298
PLANT STATUS
Cooper Nuclear Station (CNS) was at 0 psig reactor pressure, in Mode 4, Cold Shutdown, at the time of the event.
BACKGROUND
The safety objective of the Residual Heat Removal (RHR) system [EllS:B0] is to provide core cooling, in conjunction with other Emergency Core Cooling Systems, and to provide containment cooling as required during abnormal operational transients and postulated accidents. The RHR system consists of two heat exchangers [EIIS:HX], four main system pumps [EIIS:P], and associated piping, valves [EllS:V], controls and instrumentation. Irradiated fuel in the shutdown reactor core generates heat during the decay of fission products and increases the temperature of the reactor coolant. The decay heat must be removed to reduce the temperature of the reactor coolant less than or equal to 212 degrees Fahrenheit (F) in preparation for performing Refueling or Cold Shutdown maintenance operations, or the decay heat must be removed for maintaining the reactor in Hot Shutdown condition.
Shutdown Cooling (SDC) is a subsystem of RHR and is placed in operation during a normal reactor shutdown and cooldown. Reactor Coolant is pumped by the RHR main system pumps from recirculation loop "A" through the RHR heat exchangers prior to returning to the reactor vessel through connections to their respective recirculation loop.
Pressure switches [El IS:PS] RR-PS-128A and RR-PS-128B function to protect SDC suction piping from overpressure. Isolation valves [El IS:ISV] RHR-MO-17 and RHR-MO-18 are interlocked to close and remain closed on reactor pressure greater than 72 psig.
EVENT DESCRIPTION
On May 30, 2015, at 03:27, the "B" Loop of RHR was placed in the SDC mode of operations. At 04:15, the plant entered Mode 4. Isolation signals from pressure switches (RR-PS-128A and/or RR-PS-128B) were received at 04:58. SDC suction isolation valves RHR-MO-17, located outside primary containment, and RHR-MO-18, located inside primary containment closed, resulting in a loss of SDC.
Investigation revealed the event was initiated by steam flashing in the SDC line. This flashing created pressure transients, causing RHR-MO-17 and RHR-MO-18 to close. The steam flashing occurred due to temperature in the SDC line being at or near saturation temperature causing localized boiling then void collapse with coolant being drawn from the reactor vessel thru the reactor recirculation system. The pressure perturbations are created as the voids form and then collapse due to the sub-cooled coolant temperature. The pressure switches then actuated due to these pressure perturbations and caused an isolation of the SDC system by closing RHR-MO-17 and RHR-MO-18. SDC was restored at 05:20 on May 30, 2015.
Reported lessons learned are incorporated into the licensing process and fed back to industry.
Send comments regarding burden estimate to the FOIA, Privacy and Information Collections Branch (T-5 F53), U.S. Nuclear Regulatory Commission, Washington, DC 20555- 0001, or by Internet e-mail to Infocollects.Resource@nrc.gov, and to the Desk Officer, Office of Information and Regulatory Affairs, NEOB-10202, (3150-0104), Office of Management and Budget, Washington, DC 20503. If a means 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.
NRC
FORM 366 1 U.S. NUCLEAR REGULATORY COMMISSION (See Page 2 for required number of digits/characters for each block)
NO
Cooper Nuclear Station 05000298 During the loss of SDC, forced circulation of the reactor core was not lost. The Reactor Recirculation '(RR) loop "A" system remained in service. Based on indications from the RR system "A" loop, temperature indications showed in an increase of 18 degrees F, from 186 to 204 degrees F. The vessel drain temperature increased from 190 to 210 degrees F for a heat up of 20 degrees F during the same time period.
BASIS FOR REPORT
This event is reportable under 10 CFR 50.73(a)(2)(v)(B). This was a condition which could have prevented the fulfillment of the safety function of structures or systems that are needed to remove residual heat.
SAFETY SIGNIFICANCE
This is a Safety System Functional Failure. The actual and potential safety consequences of this event were minimal because the isolation of the RHR Shutdown Cooling suction valves occurred due to a momentary isolation signal and did not impact the ability to restore the valves to the open position for restoring SDC to service. The interlock feature performed as designed and caused RHR-MO-17 and RHR- MO-18 to close when high pressure was present.
This event did not challenge the ability to maintain safe shutdown conditions, remove residual heat, control the release of radioactive material, or mitigate the consequences of an accident.
CAUSE
The root cause of the event was determined to be a design vulnerability and subsequent operation of the SDC system that resulted in a trip of the SDC suction valves due to sub-cooling and flashing in the RHR or RR system.
CORRECTIVE ACTIONS
To prevent recurrence of this event, CNS will initiate an Engineering Change request to move the location of the input pressure signals required to meet requirements of Technical Specification 3.3.6.1, Table 3.3.6.1-1, 6(a) from the RR line to the Vessel Steam Dome.
PREVIOUS EVENTS
On October 14, 2012, flashing of the hot reactor coolant to steam occurred, causing a pressure spike that exceeded 72 psig. Consequently, RHR-MO-17 and RHR-MO-18 closed, isolating the RHR SDC loop. This event was reported under LER 2012-004-00, Isolation of Shutdown Cooling Results in Loss of Safety Function, dated December 6, 2012.