05000244/LER-2007-001

LER-2007-001, Loss of Electrical Generation Results in Plant Trip

Mdocket Nu Bersequential Revmonth Day Year Year Month Day Yearnumber No. 05000
Event date:	01-27-2007
Report date:	3-23-2007
2442007001R00 - NRC Website

I. PRE-EVENT PLANT CONDITIONS:

On January 27, 2007 the R.E. Ginna Nuclear Power Plant (Ginna) was in Mode 1 at approximately 100% steady state reactor power.

A. EVENT:

On January 27, 2007, at approximately 2040 EST, Ginna experienced a loss of load transient when gross electrical generation output decreased from approximately 607 MWe to less than 100 MWe over an approximate ten second time period. The loss of electrical generation is attributed to a failure in the Main Turbine Electro Hydraulic Control (EHC) system that caused the four high pressure turbine control valves to rapidly close simultaneously. The loss of electrical generation resulted in a heat-up of the reactor coolant system (RCS) and an automatic reactor trip due to an actuation of the Reactor Protection System (RPS) on Over-Temperature Differential Temperature.

The Control Room operators performed the appropriate actions of Emergency Operating Procedure E-0 (Reactor Trip or Safety Injection). The operators then transitioned to Emergency Operating Procedure ES-0.1 (Reactor Trip Response) when it was verified that both reactor trip breakers were open, all control and shutdown rods were inserted, and safety injection was not actuated or required.

TO THE EVENT:

None

C. DATES AND APPROXIMATE TIMES OF MAJOR OCCURENCES:

• January 27, 2007, 2040 EST: automatic reactor trip due to an actuation of the Reactor Protection System (RPS) on Over-Temperature Differential Temperature.

D. OTHER SYSTEMS OR SECONDARY FUNCTIONS AFFECTED:

None

E. METHOD OF DISCOVERY:

The reactor trip was immediately apparent due to plant response, alarms, and indications in the Control Room.

F. SAFETY SYSTEM RESPONSES:

Both motor driven auxiliary feedwater (AFW) pumps and the turbine driven AFW pump started due to a signal from the anticipated-transient-without-scram (ATWS) mitigation actuation circuitry (AMSAC), and functioned properly. As a result of the RCS heat-up transient, the RCS experienced a pressurization event that caused both Power-Operated Relief Valves (PORVs) to open twice. The peak main steam pressure obtained during the heat-up of the RCS also caused one of the eight main steam safety valves to open momentarily. Additionally, an automatic trip of the A Main Feedwater Pump as the result of a low seal water differential pressure occurred approximately seven minutes after the reactor trip.

III.� CAUSE OF EVENT:

The immediate cause of the reactor trip was a Reactor Protection System (RPS) actuation on Over-Temperature Differential Temperature. The RPS actuation was the result of a loss of electrical generation event.

The underlying cause of the loss of electrical generation event was attributed to a failure in the Main Turbine Electro Hydraulic Control (EHC) System. An EHC System electrical card which provides a common output signal to all four of the high pressure turbine control valve servo cards had an edge connector that was spread and temporarily lost contact with the receiver cabinet side of the circuit. The result was a common output signal that went to zero volts, causing the control valves to close.

The design of the EHC System electrical card was a known potential failure mechanism as evidenced by warning placards inside the control panel. Although the placards provided some degree of information to the workers, the station failed to Incorporate adequate guidance into the maintenance procedures for inspecting the card edge connections and seating the cards. This card is a single point vulnerability with regards to a turbine/reactor trip.

IV. ASSESSMENT OF THE SAFETY CONSEQUENCES OF THE EVENT:

This event is reportable In accordance with 10 CFR 50.73, Licensee Event Report System, item (a)(2) (iv)(A), which requires a report of, "Any event or condition that resulted in a manual or automatic actuation of any of the systems listed in paragraph (a)(2)(iv)(B) of this section". The automatic reactor trip is an actuation of the Reactor Protection System, and AFW pump starts are actuations of a PWR auxiliary feedwater system.

An assessment was performed considering both the safety consequences and implications of this event with the following results and conclusions:

There were no operational or safety consequences or implications attributed to the reactor trip because:

• The two reactor trip breakers opened as required.

• All control and shutdown rods inserted as designed.

• The loss of load transient is similar to the loss of load transients described in the Updated Final Safety Analysis Report (UFSAR). The loss of electrical load transients are performed to evaluate RCS over-heating events and their impact on the following design criteria:

I)�minimum DNBR, ii) maximum RCS pressure and, iii) maximum Main Steam and SG pressure The UFSAR transients were examined and compared to the plant response for the actual event. The plant behavior was found to be bounded by the events detailed in the accident analysis. The UFSAR transients were found to be bounding due to a combination of less limiting actual plant conditions and proper operation of plant equipment In responding to the plant trip.

• The momentary opening of the main steam safety valve is bounded by the UFSAR accident analysis event, which assumes operation of the main steam safety valves.

• The Control Room operators were in the process of manually stopping the A Main Feedwater Pump at the time it automatically tripped and the event did not adversely affect the ability of the operators to respond to the plant transient.

• The Conditional Core Damage Probability (CCDP) of the event was calculated as 8.53E-07.

Based on the above and the review of post trip data and past plant transients, it can be concluded that the plant operated as designed, that the public's health and safety was assured at all times.

V.�CORRECTIVE ACTIONS:

A. ACTION TAKEN TO RETURN AFFECTED SYSTEMS TO PRE-EVENT NORMAL STATUS:

• The card edge connector was repaired, and a full Inspection of the card was performed.

Testing was completed to verify proper operation of the card and the EHC System.

B. ACTION TAKEN OR PLANNED TO PREVENT RECURRENCE:

• Maintenance procedures will be revised to Include criteria and operating experience for card edge connector inspection.

• Instrument and Control Technician initial and continuing training programs will be evaluated for specific content based on this event.

• A single point vulnerability classification and development of mitigating strategies will be completed for all single point vulnerabilities in the EH system.

• The affected EHC System card has since been replaced during a forced outage.

VI. ADDITIONAL INFORMATION:

A.�FAILED COMPONENTS:

No other structures, systems, or components failed as result of this event.

B. PREVIOUS LERs ON SIMILAR EVENTS:

A similar Ginna LER event historical search was conducted which resulted In no similar events.

C. THE ENERGY INDUSTRY IDENTIFICATION SYSTEM (EIIS) COMPONENT FUNCTION

THIS LER:

COMPONENT IEEE 803� IEEE 805 FUNCTION NUMBER�SYSTEM IDENTIFICATION Pump� P� SJ Pump P BA Valve, Relief� RV� SB Valve, Relief RV AB Connector� CON� JJ