05000263/LER-2007-001

LER-2007-001,

Docket Number
Event date:	01-10-2007
Report date:	03-12-2007
2632007001R00 - NRC Website

Description At approximately 1528 on January 10, 2007, with the unit stable in Mode 1 at approximately 88 percent power, all four turbine control valves [SCV] ramped from approximately 50 percent open to 100 percent open in approximately two seconds with no demand to do so from the pressure control system. The resultant decrease in reactor pressure resulted in a group 1 isolation on low steam line pressure with the MODE switch in RUN, and associated reactor scram. The closure of the Main Steam [SB] Isolation Valves and reactor scram caused reactor pressure to increase and reactor water level to decrease to minus eight inches. The low reactor water level also resulted in a Group II Emergency Safety Feature actuation as designed. All control rods [AA] inserted and all safety systems responded as expected.

During the reactor water level transient, the turbine [JJ] and main feed [SJ] pumps [P] tripped at 48 inches as designed. Maximum water level reached was 78 inches (30 inches below the main steam lines). Control of plant pressure was initially established by manually operating the safety relief valves (SRVs) [RV]. At no point was automatic operation of the SRVs required or challenged. Subsequently, reactor pressure control was established using the high pressure coolant injection system (HPCI) [BJ]. Reactor makeup was established using normal feedwater.

Post scram review disclosed that the turbine control valves did not close on the turbine trip as designed. A post scram walk down revealed that supports [SPT] for the control valve enclosure had failed, allowing the enclosure to transition downward approximately six inches, causing deformation of adjoining support structures. No system breach occurred and there was no release of radioactivity.

Following plant stabilization, the decision was made to proceed to cold shutdown (Mode 4).

As uncertainty existed regarding the status of the main steam system (the control valve enclosure contacted insulation on a section of the main steam line and the control valves failed to close following the turbine trip), the decision was made to cool down using HPCI and the SRVs for pressure control, main feed for makeup, and the residual heat removal system [BO]. Mode 4 was achieved at 0613 on January 11, 2007.

Event Analysis

Pursuant to 10CFR 50.72 paragraphs (b)(2)(iv)(B) for the RPS actuation and paragraph (b)(3)(iv)(A) as an ESF actuation, an eight-hour event notification was made to the USNRC.

Per 10 CFR 50.73 (a)(2)(iv), a Licensee Event report is required for this event.

The event is not classified as a safety system functional failure.

Safety Significance

At approximately 88 percent reactor power, all turbine control valves ramped full open. As reactor pressure dropped, reactor power lowered. The Group 1 isolation occurred at 840 psi as required. In response to the automatic scram at a MSIV position of 90 percent open with the MODE switch in RUN, all control rods inserted as designed. The minimum reactor water level reached during the transient was minus eight inches. No Emergency Core Cooling System (ECCS) actuation was required or occurred. As level lowered below nine inches, the Group 2 isolation occurred as expected. With the feed system in service, feedwater, in conjunction with decay heat, caused reactor water level to rise. At 48 inches the main turbine and the main feedwater pumps tripped as expected. The highest water level reached was 78 inches, approximately 30 inches below the main steam lines. Reactor pressure was controlled initially by manually operating the safety relief valves. No automatic operation of the safety relief valves was required. Water level was subsequently restored to normal levels, feedwater was restarted for makeup, and high pressure coolant injection started and operated in the pressure control mode to remove decay heat.

With the exception of the failure of the turbine control valves to close, all systems and components responded as expected to mitigate the transient and assure core cooling.

Automatic operation of the SRVs and emergency core cooling systems did not occur and were not required. No safety systems or systems credited in the Probabilistic Risk Assessment (PRA) as capable of mitigating an accident were compromised as a result of the structural failure. As a result, there was no significant increase in the risk of a core damage accident or a release of radioactive material to the environment. The transient was well within the analyzed and anticipated frequency of a turbine trip initiator analyzed in the PRA model and bounded by the Pressure Regulator failure analysis in the USAR.

In conclusion, the safety significance in terms of reactor safety and radiological release to the environment as a result of the transient that occurred on January 10, 2007, was not significant.

Cause

The root cause of this event was determined to be latent shortcomings in the design and construction of the turbine control valve enclosure support system during initial construction.

The failure of the support system and associated vertical transition of the control valve enclosure caused the cam that actuates the servo for the control valves to rotate in a direction commanding the control valves to open. This motion overrode signals from the pressure regulating system to close the control valves. Because the cam motion was forced by the transition downward of the control valve enclosure, the control valves remained open despite the turbine trip signal.

Corrective Action The following actions have been completed:

1. The control valve enclosure support system was redesigned assuring that stress levels remained well below code allowable values. The modified design was installed prior to startup.

2. Collateral damage caused by the failure of the control valve supports and the vertical transition of the enclosure was analyzed and repairs completed as required.

3. Inspections and tests of the turbine control system were performed and repairs affected as necessary to assure full system functionality prior to startup.

4. An extent of condition review and plant walkdown was completed which concluded that other similar support configurations were sufficiently robust such that reasonable assurance of their functionality existed.

Failed Component Identification None

Previous Similar Events

No previous similar events were identified.