05000298/LER-2003-008

LER-2003-008, Inadequate Evaluation Leads to Technical Specification Prohibited Operation

Docket Number
Event date:	12-03-2003
Report date:	02-02-2004
Reporting criterion:	10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications
2982003008R00 - NRC Website

PLANT STATUS

Cooper Nuclear Station (CNS) was in Mode 3 (Hot Shutdown) at 000 percent (%) power at the time of the event.

EVENT DESCRIPTION

At 22:02 on November 28, 2003 during normal operation at 100% power, the reactor automatically scrammed on low water level due to a control system malfunction affecting Reactor Feed Pump B (LER 2003-007).

As a result of the scram and the vessel level shrink due to void collapse, both of the Reactor Recirculation pumps tripped, the Reactor Water Cleanup System (RWCU) isolated, and High Pressure Coolant Injection (HPCI) and Reactor Core Isolation Cooling (RCIC) initiated. HPCI and RCIC were manually tripped after reactor water level was restored to the normal level. At 22:10, Operators reset the scram and at approximately 23:08 RWCU Train A was placed back in service (LER 2003-007).

During the recovery from the transient it was noted that the cooldown rate in the bottom head of the reactor had exceeded the 100 degree (°)F per hour Technical Specification limit. This is a known phenomenon that can occur when forced core circulation is lost, while cold water is injected into the bottom of the vessel through the Control Rod Drive system. Exceeding the Technical Specification limit on plant cooldown required entry into the Limiting Condition for Operation (LCO) Condition that governs Reactor Coolant System (RCS) heat-up and cooldown rates.

To exit this LCO it is necessary to determine that the RCS is acceptable for continued operation.

While taking action to reduce reactor vessel stratification Operators raised Reactor Pressure Vessel level to establish natural circulation flow. The establishment of natural circulation flow resulted in the heat-up rate in the bottom head of the vessel exceeding the 100 °F per hour Technical Specification limit. The heat-up rate was restored to within limits and operators commenced depressurization of the reactor vessel to promote natural circulation flow. Heat-up rate in the bottom head drain exceeded the 100 °F per hour Technical Specification limit following the depressurization.

Engineering was engaged to evaluate the cooldown and first heat-up rate transients. A consultant was retained to assist with the evaluation and was provided with the transient information and vessel parameters (e.g., vessel temperatures and pressures). This evaluation reviewed the vessel parameters from just prior to the reactor scram through the first heat-up transient. The initial evaluation results were based on an incorrect assumption about plant conditions and inappropriately concluded that the second heat-up transient was bounded by the vessel design cycle analysis. The engineering evaluation was revised twice in an attempt to resolve operator verbal comments.

Revision 2 was accepted as justification for exiting the LCO, whereupon the LCO was exited and plant Mode was changed from hot shutdown to startup. Further review of the engineering evaluation determined that bottom head heat-up effects resulting from the second heat-up transient had not been properly addressed and the LCO should not have been exited. The unit was returned to hot shutdown until the RCS evaluation was properly completed to permit exiting the LCO and conducting a plant startup. Subsequent analysis determined that the RCS was acceptable for continued operation.

BASIS FOR REPORT

This event is reportable in accordance with 10 CFR 50.73(a)(2)(i)(B) for any operation or condition which was prohibited by the plant's Technical Specifications.

CAUSE

There were two causes to this event, first, human performance errors by several involved individuals and second, a lack of formality in the post-event review and startup approval process.

1. Individual actions by the preparer, independent verifier, supervisor and Shift Technical Engineer were flawed resulting in an inaccurate and incomplete Engineering Evaluation.

2. The post event review and startup approval process lacks the formality and structure necessary to ensure identification and resolution of all startup issues in a rigorous manner.

SAFETY SIGNIFICANCE

The final RCS condition analysis ultimately demonstrated that the stresses induced by the transients remained within the design stress and fatigue limits for the vessel and components. Therefore, there was no actual safety significance for this event. Worst case potential consequences could have been a stress induced fatigue flaw in the nozzle welds, shroud support welds, or bottom head welds leading to an unisolable leak in the Reactor Coolant Pressure Boundary. There were no actual consequences to the vessel due to the failure to complete the analysis prior to entry into Mode 2, and the likelihood of a worse case outcome was very low.

CORRECTIVE ACTIONS

Immediate Actions:

The unit was returned to Mode 3.

The Engineering Evaluation was revised to accurately address stress effects in the bottom head due to heat-up.

Long Term Actions:

Individuals involved in this event have been counseled in the need for formality, rigor, procedure adherance, communications and follow-up.

Revise the Post-Event Review and Startup Approval process to make the process more comprehensive. (Due May 28, 2004) Produce a formal document delineating the standards for preparing, reviewing and approving engineering documents, including handling of formal comments and resolutions. Communicate this to the engineering staff. (Due May 31, 2004) Develop a formal process for exiting LCO action statements based on engineering evaluation. (Due April 30, 2004) Prepare and present training to the engineering staff of events associated with the engineering evaluation of the heat- up and cool-down transients that occurred after the automatic reactor scram on November 28, 2003. (Due June 30, 2004)

PREVIOUS EVENTS

None