ML20043F168

LER 90-001-00:on 900512,reactor Tripped Due to lo-lo Water Level in Steam Generator 2A.Caused by Procedural Inadequacy. Procedure FNP-2-STP-151 Revised to Provide Addl Initial Conditions to Prevent Subj event.W/900608 Ltr

ML20043F168
Person / Time
Site:	Farley
Issue date:	06/08/1990
From:	Hairston W, Dennis Morey ALABAMA POWER CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
LER-90-001-01, LER-90-1-1, NUDOCS 9006140214
Download: ML20043F168 (4)

Text

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.' Allb:m2 Power Company 40 inverness Center Parkway

-. Post Othee Box 1295 Dirmingham. Alabama 35201 Telephone 205 86B-5501 i, W. G. Ha:raton, til

?t' Senior Vice President Nuclear Operation. Alabama Power June 8. 1990 E 10CFR$0.73 I Docket No. 50-364 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Vashington, D.C. 20555 o

Gentlemen Joseph M. Farley Nuclear Plant - Unit 2 Licensee Event Report No. LER 90-001-00 Joseph M. Farley Nuclear Plant, Unit 2 Licensee Event Report No. LER 90-001-00 is being submitted in accordance with 10CFR50.73.

If you have any questions, please advise, n

Respectfully submitted, g). ph W. G. Hairston, III VGil,lII/JARimgd 16.19 b

Enclosure cc Mt. S. D. Ebneter Mr. G. F. Maxwell 9004140214 9004,00 ' ./h

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$UtvilSION 449449ren. e.",otege $K9tCilO $\/0ht!SSION Detll kO l l l Auf m AC, es , M om a... ..n w, r , .,, ece evo.~,r ., >,w von At 0615 on 5-12-90, during a unit startup, the reactor tripped due to low-lov vater level in the 2A steam generator. The low-lov vater level occurred when the steam supply valves for the on-service steam generator feedvater pump (SGFP) closed. Testing being performed in accordance with FNP-2-STP-151.4 (Main Turbine Protective Device Test) resulted in lov electro-hydraulic (Ell) fluid pressure which caused the SGFP valves to close.

This event was .aused by procedural inadequacy. FNP-2-STP-151.4 did not provide adequate guidance concerning the initial conditions required to perform the procedure. Testing performed subsequent to the reactor trip snoved that it is not appropriate to perform this procedure while feedvater is being provided by a SGFP.

FNP-2-STP-151.4 has been revised to provide additional initial conditions to prevent performing the procedure when an SGFP is in service.

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Plant and System Identification Vestinghouse - Pressurized Vater Reactor Energy Industry Identification System codes are identified in the text as [XX).  :

i Summary of Event i At 0615 on 5-12-90, during a unit startup, the reactor tripped due to lov-lov vater level in the 2A steam generator [AD). The low-lov vater level occurred when the steam supply valves for the on-service steam generator feedvater pump  :

(SGFP) [SJ) closed. Testing being performed in accordance with FNP-2-STP-151.4 (Main Turbine Protective Device Test) resulted in lov electro-hydraulic (EH) l fluid [TG] pressure which caused the SGFP valves to close.

Description of Event At approximately 0615 on 5-12-90, a limit switch on main steam isolation valve (MSIV) 3370A vas placed in the closed position in accordance with an established surveillance procedure (FNP-2-STP-151.4). By plant design this caused the turbine governor, intercept and reheat stop valves to close due to draining of EH fluid through the dump valves for each turbine valve. In this condition (all turbine valves closed), the turbine appears to be tripped. However, the turbine was still '

latched and the turbine control system was calling for the valves to be open. The 5 operators did not recognize this condition nor did they realize this placed a large demand on the EH fluid system causing EH pressure to decrease rapidly. This EH fluid pressure drop was not noticed by the operators. However, the EH pressure.

drop caused the SGFP steam supply valves to close resulting in a loss of feedvater flow and the decrease in steam generator levels.

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In response to the loss of feedvater, the operators started all auxiliary feedvater penps and decreased reactor power. However, these act!ons failed to restore steam generator level prior to the 2A steam generator level decreasing to the low-lev i level setpoint. A reactor trip occurred at 0615 on 5-12-90. Reactor power had l been decreased to approximately 3 percent when the trip occurred.

( Following the trip, the operators implemented FNP-2-EEP-0 (Reactor Trip or Safety Injection) and FNP-2-ESP-0.1 (Reactor Trip Response), ensuring that the unit was l safely in Mode 3 (Hot Standby). The unit was maintained in a stable condition.

' ause of Event This event was caused by procedural inadequacy. FNP-2-STP-151.4 did not provide adequate gufh nce concerning the initial conditions required to perform the >

procedure. Testing performed subsequent to the reactor trip shoved that it is not appropriate to perform this procedure while feedvater is being provided by an SGPP.

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__.~ -C-m 1 Corrective Action PNP-2-STP-151.4 has been revised to provide additional initial conditions to  !

! prevent performing the procedure when an SGFP is in service. I Reportability Analysis and Safety Assessment i

This event is reportable because of the actuation of the reactor prctection system. After the trip, the following safety systems cperated as designed:

- main feedvater was isolated with flow control valves and bypass valves closed, source range nuclear instrumentation automatically energized, and pressurizer heaters and spray valves operated automatically as required to maintain system pressure.

There was no effect on the health and safety of the public.

Additional Information No components failed during this event.

This event would not have been more severe if it had occurred under different operating conditions.

A reactor trip occurred during performance of FNP-2-STP-151.4 on May 13, 1986.

Review of the trip report in 1990 for this previous event indicates it was similar to the current event. However, personnel in 1986 performed an l inadequate root cause analysis. In May of 1986, the trip vas believed to have been caused by loss of EH fluid pressure due to a governor valve EH fluid dump valve sticking open when the turbine was being latched. This improper cause analysis is believed to have been made due to the surveillance procedure improperly instructing the operator to latch the turbine after testing the MSIV 3370A limit switch. This guidance was improper and misleading since the l

r turbine does not receive a trip signal by testing the MSIV 3370A limit switch.

As a result of procedure reviev unrelated to the May 1986 trip, this instruction was removed from the procedure.

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Corrective actions taken in 1986 included adding precautions in procedures for l monitoring EH pressure when latching the turbine and tripping the turbine if l lov EH fluid pressure threatens SGFP operation. These corrective actions did not prevent recurrence, since the turbine was not being relatched and therefore EH fluid pressure vas not monitored. The actual cause of both reactor trips was the performance of FNP-2-STP-151.4 which by design causes a large drop in EH fluid pressure. Determination of root cause for events has

-since been stressed and improved, as evidenced by current proper determination of the May 12, 1990 trip.

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