05000366/LER-2009-003
| Docket Number Sequential Revmonth Day Year Year Month Day Year 05000Number No. | |
| Event date: | 06-20-2009 |
|---|---|
| Report date: | 08-10-2009 |
| Reporting criterion: | 10 CFR 50.73(a)(2)(iv)(A), System Actuation |
| 3662009003R00 - NRC Website | |
PLANT AND SYSTEM IDENTIFICATION
General Electric - Boiling Water Reactor Energy Industry Identification System codes appear in the text as (El IS Code XX).
DESCRIPTION OF EVENT
On June 20, 2009 at 14:17 EDT, Unit 2 was in mode 1 with an approximate reactor power of 2804 CMWTh. At this time the control room received a `GEN INLET TEMP HIGH' alarm followed within minutes by a 'GEN PROTECTION CIRCUIT ENERGIZED' alarm. This protective circuitry initiated a main generator (El IS Code TB) runback due to a high Stator Water Cooling (SWC, El IS Code TJ) outlet temperature. The electro-hydraulic control, Mark VI system (EHC, EllS Code TG) cores processor began reducing turbine load by closing the turbine control valves (TCV, EllS Code TA) and subsequently opening the turbine bypass valves (TBV, EllS Code TA). Operations personnel began reducing reactor load in response to the main generator runback by reducing recirculation flow (El IS Code AD). The reactor power was reduced to approximately 66 percent reactor thermal power and could not be reduced any lower without the insertion of control rods (El IS Code AA). Due to turbine load reducing faster than the manual actions taken to reduce reactor load, reactor pressure began to increase. Reactor pressure increased to a high of 1074 psig and the Reactor protection System (RPS) (EllS Code IG), initiated a reactor scram due to high reactor pressure.
Reactor water level decreased resulting in a primary containment valve Group 2 (EllS Code JM) valve isolation per design. Safety Relief Valves (EllS Code SB) did not actuate nor were they required to based on the maximum reactor pressure reached.
CAUSE OF EVENT
The direct cause of this event was improper set-up of a valve controller.
During the investigation of the 'GEN INLET TEMP HIGH' alarm, it was found that the generator SWC heat exchangers were on full bypass when they should have been providing flow through the heat exchanger given the current system operating temperature. When the valve was manually stroked by maintenance, it was noticed that the piping temperatures down stream of the valve immediately dropped. From review of the trend data taken by rounds it was determined that the valve was not maintaining the SWC temperature by design. Therefore, the cause of this event was failure of valve 2N43F100 (SWC Temperature Controller) to properly control the SWC temperature.
The SWC temperature instrument loop consists of (2N43F100) a Fisher model 667-Y AOV three way valve with a model 3582 pneumatic valve positioner. A Fisher 4160B temperature controller feeds the instrument air signal to the above positioner and controls the SWC temperature by its established set points. The function of this instrument loop is to control the inlet SWC temperature within a given operating range. The valve controls this temperature by opening a bypass around the SWC heat exchangers when SWC inlet temperatures begin to drop or vice-a-versa when the temperature increases. This function � PRINTED ON RECYCLED PAPERNRC FORM 366A (9-2007) U.S. NUCLEAR REGULATORY COMMISSIONNRC FORM 366A LICENSEE EVENT REPORT (LER)(9-2007)
CONTINUATION SHEET
2009 - 003 - 0 was lost when maintenance was performed on the instrument loop during the 2R20 refuel outage. The positioner and the controller were not set so that the valve would function per design. Both the positioner for valve 2N43F100 and its temperature controller 2N43R310, proportional band were set up as "Direct' acting. In this configuration, the valve responded opposite of its design. Therefore; due to seasonal increases in river temperature, the temperature control set points were reached and the SWC flow was automatically placed on full bypass. Removal of system heat exchangers from the system caused a rapid heat build up in the stator cooling water since the generator was online. When the system went on bypass, the generator stator bar temperatures increased to a point where the generator runback set point was reached.
REPORTABILITY ANALYSIS AND SAFETY ASSESSMENT
This event is reportable per 10 CFR 50.73(a)(2)(iv)(A) because unplanned actuations of a safety feature system listed in 10 CFR 50.73 occurred. In this instance, a reactor protection system (RPS) actuation resulting in a reactor scram.
An increase in reactor vessel pressure during reactor operation compresses the steam voids and results in a positive reactivity insertion. This causes the neutron flux and thermal power transferred to the reactor coolant to increase which could challenge the integrity of the fuel cladding and the reactor coolant pressure boundary. Therefore, the reactor is shut down automatically on high reactor vessel steam dome pressure to limit the neutron flux and thermal power increase. The automatic reactor shutdown on high pressure, along with the SRVs, limits the peak reactor vessel pressure to less than the American Society of Mechanical EngineersSection III Code limits.
In this event, protective circuitry initiated a main generator runback due to a high SWC outlet temperature. EHC, Mark VI, system began reducing turbine load by closing the turbine control valves (TCV) and subsequently opening the turbine bypass valves. Operations personnel began reducing reactor load in response to the main generator runback by reducing recirculation flow. The reactor power was reduced to approximately 66 percent reactor thermal power and could not be reduced any lower without the insertion of control rods. Due to turbine load reducing faster than the manual actions taken to reduce reactor load, reactor pressure began to increase. Pressure increased to a high of 1074 psig and the RPS initiated a reactor scram due to high reactor pressure. Reactor water level decreased resulting in primary containment Group 2 valve isolation per design. No SRVs opened nor were any required to open to limit or reduce reactor vessel pressure. No Emergency Core Cooling Systems actuated nor were any required to actuate to recover or maintain water level during or following this event. All automatic functions operated per design in response to the pressure increase and the automatic reactor shutdown.
Based on this analysis, it is concluded that this event had no adverse impact on nuclear safety.
� PRINTED ON RECYCLED PAPER there was a potential for a similar configuration to exist on another valve. It was determined that no other vulnerabilities of this nature exist.
Plant personnel confirmed that the Unit 1 SWC instrument loop is set up so it will function as designed.
ADDITIONAL INFORMATION
Other Systems Affected: None Failed Components Information:
Master Parts List Number: 2N43F100 ElIS System Code: TJ Manufacturer: Fisher Controls (F130) Reportable to EPIX: Yes Model Number: 3582 Root Cause Code: B Type: Valve Positioner ElIS Component Code: TCV Failed Components Information:
Master Parts List Number: 2N43R310 ElIS System Code: TJ Manufacturer: Fisher Controls (F130) Reportable to EPIX: Yes Model Number: 4160B Root Cause Code: B Type: Temperature Controller ElIS Component Code: 23 Commitment Information:
This report does not create any new permanent licensing commitments.
Previous Similar Events:
There are no similar events within the past two years in which a valve control system was set-up improperly that resulted in a reactor scram.
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