05000263/LER-2005-003

LER-2005-003,

Docket Number
Event date:	03-08-2005
Report date:	05-09-2005
2632005003R00 - NRC Website

Description On March 8, 2005 the Monticello Nuclear Generating plant was shutdown in a refueling outage, with the reactor head [RPV] removed, and the reactor pressure vessel (RPV) [RPV] flooded up to the flange with the main steam-line plugs installed. The reactor water temperature was 98 degrees F and being cooled by #12 Residual Heat Removal (RHR) [BO] pump [P] aligned in shutdown cooling (SDC) mode.

At 0454, the station was in the process of performing an equipment isolation to support replacement of the Safety Relief Valve (SRV) [RV] solenoids [SOL]. While performing the isolation several unexpected alarms [ALM] annunciated in the Control Room. Operations Shift Supervision decided to remove the isolation and re-evaluate the tagged components. When the isolation was removed the running pump logic sensed a loss of suction flow path to the operating #12 RHR because the position indication for the RHR pump shutdown cooling suction valves [VLV] had been lost due to the equipment isolation.

The indicated loss of suction flow path caused the pump to trip and resulted in the loss of the in-service SDC train.

Operators restored SDC to the unit at 0507 by restarting the #12 RHR pump. The RPV water level remained stable at 651 inches and RPV water temperature remained stable at 98 degrees F.

Based on the loss of shutdown cooling, an NRC Event Notification Report (#41468) was made on March 8, 2005 at 1041.

Event Analysis

In accordance with 10 CFR 50.72 (b)(3)(v)(B), "Event or Condition that could have Prevented Fulfillment of a Safety Function," an eight-hour event notification was made to the USNRC, due to the loss of shutdown cooling which is required to remove residual heat. Per 10 CFR 50.73 (a)(2)(v)(B), a Licensee Event report is required for this event.

The event is classified as a safety system functional failure.

Safety Significance

A review for event significance was performed for the loss of SDC. During the event the RHR and core spray (CS) [BM] injection paths remained available. In addition, Reactor Water Cleanup (RWCU) [CE] was in operation throughout the event. The safety significance of this event with regards to the potential for core damage from the loss of shutdown cooling was determined to be minimal. Several alternative systems and sources of coolant were available and operable, and fully capable of keeping the reactor core covered/cooled.

Operations control room personal identified the loss of SDC and restored it to service in approximately 13 minutes after confirming that the removal of the isolation on the 125 VDC system [EJ] had resulted in the #12 RHR pump trip causing the loss of SDC.

Station procedures provide a means to establish alternate SDC if Operations personnel are unable to restore normal SDC. These procedures assume that the RHR system in SDC mode is either unavailable or incapable of maintaining reactor water temperature, a condition more severe than the actual occurrence. The procedure describes the systems available and how they are to be used to augment SDC in the event of a complete failure of the RHR system. The following alternate systems would be available: RWCU, the CST, Residual Heat Removal Service Water (RHRSW) "A" [BI] cross tie to Low Pressure Coolant Injection (LPCI) [BO], Service water crosstie to condensate [SD]/feedwater [SJ], Firewater [FP] crosstie to LPCI, and SBLC.

Cause

The Root Cause was determined to be failure of Operations and Station Management to effectively oversee implementation of the outage isolation process.

The contributing causes identified several weaknesses in the preparation, review and approval of outage isolations. These weaknesses included failure to require impact statements (e.g.; the effect on safety systems) in outage work packages and isolations, deficiencies in isolation preparations, and inadequate verification of isolations.

Corrective Action Immediate Actions Taken:

1. The SDC system was restored to service.

2. All outage isolations were placed on hold. Each of the outage isolations received an additional review to determine if the isolation had the potential to affect safety related components. If the potential existed, then the work package documentation was reviewed and an impact statement was developed and noted in the isolation placement instructions. The impact statement identified the affected safety related systems or components, and the potential impact on the plant.

3. A stand down was performed with the operations staff to review the event and actions taken.

4. Plant management increased observations during the outage to monitor equipment isolation performance.

Additional Corrective Actions to be taken:

A root cause investigation was completed and corrective actions were identified to revise the work control and outage processes. These actions are being tracked through the corrective action program process.

Failed Component Identification N/A

Previous Similar Events

A review of the station corrective action program database found the following similar event:

November 26, 2001: During Refuel Outage 20, SDC was lost due to an error in a modification installation procedure. The wrong studs for installing a jumper were listed in the procedure causing the "A" LPCI injection valve to close interrupting SDC. The cause was inadequate review of modification installation procedures. Corrective actions were appropriate but limited to the Design Engineering Department's reviews of modification documents.