05000263/LER-2015-007

LER-2015-007, Loss of Residual Heat Removal Capability

Monticello Nuclear Generating Plant
Event date:	11-24-2015
Report date:	01-21-2016
Reporting criterion:	10 CFR 50.73(a)(2)(v)(B), Loss of Safety Function - Remove Residual Heat
LER closed by
IR 05000263/2016004 (13 February 2017)
2632015007R00 - NRC Website

LER 15-007-00 for Monticello Regarding Loss of Residual Heat Removal Capability

ML16022A223
Person / Time
Site:	Monticello
Issue date:	01/21/2016
From:	Gardner P A Northern States Power Co, Xcel Energy
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
L-MT-16-006 LER 15-007-00
Download: ML16022A223 (4)

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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].