ML17304B393

LER 89-009-00:on 890712,two Reactor Coolant Pumps Load Shed from Power Supply Triggering Reactor Trip Due to Low Coolant Flow.Caused by Failed Fuse in Bus Potential Transformer.Fuse Replaced & Investigation conducted.W/890814 Ltr

ML17304B393
Person / Time
Site:	Palo Verde
Issue date:	08/14/1989
From:	Haynes J, Shriver T ARIZONA PUBLIC SERVICE CO. (FORMERLY ARIZONA NUCLEAR
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
192-00507-JGH-T, 192-507-JGH-T, LER-89-009, LER-89-9, NUDOCS 8908180269
Download: ML17304B393 (22)

Text

ACCELERATED DES+BUTlON SYSTEM DEMONSTRATION REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)

ACCESSION NBR:8908180269 DOC.DATE: 89/08/14 NOTARIZED: NO DOCKET FACIL:STN-50-529 Palo Verde Nuclear Station, Unit 2, Arizona Publi 05000529 AUTH. NAME AUTHOR AFFILIATION SHRIVER,T.D. Arizona Public Service Co. (formerly Arizona Nuclear Power HAYNES,J.G. Arizona Public Service Co. (formerly Arizona Nuclear Power RECIP.NAME RECIPIENT AFFILIATION

SUBJECT:

LER 89-009-00:on 890712,reactor forced flow.

trip due to partial loss of W/8 ltr.

DISTRIBUTION CODE: IE22T COPIES RECEIVED:LTR i ENCL TITLE: 50.73/50.9 Licensee Event Report (LER), Incident Rpt,L SIZE:

etc.

/0 NOTES:Standardized plant. 05000529 l RECIPIENT COPIES RECIPIENT COPIES ID CODE/NAME LTTR ENCL ID CODE/NAME LTTR ENCL PD5 LA 1 1 PD5 PD 1 1 CHAN,T 1 1 DAVIS,M. 1 1 INTERNAL: ACRS MICHELSON 1 1 ACRS MOELLER 2 2 ACRS WYLIE 1 1 AEOD/DOA 1 1 AEOD/DSP/TPAB 1 1 AEOD/ROAB/DSP 2 2 DEDRO 1 1 IRM/DCTS/DAB 1 1 NRR/DEST/CEB 8H 1 1 NRR/DEST/ESB 8D 1 1 NRR/DEST/ICSB 7 1 1 NRR/DEST/MEB 9H 1 1 NRR/DEST/MTB 9H 1 1 NRR/DEST/PSB 8D 1 1 NRR/DEST/RSB 8E 1 1 NRR/DEST/SGB 8D 1 1 NRR/DLPQ/HFB 10 1 1 NRR/DLPQ/PEB 10 1 1 NRR/DOEA/EAB 11 1 1 QR&j DRE~ B 10 2 2 NUDOCS-ABSTRACT 1 1 REG FILE 02 1 1 RES/DSIR/EIB 1 1 ILE 01 1 1 EXTERNAL: EG&G WILLIAMS,S 4 4 FORD BLDG HOY I A 1 1 L ST LOBBY WARD 1 1 LPDR 1 1 NRC PDR 1 1 NSIC MAYS,G 1 1 NSIC MURPHY,G.A 1 1 NOTES: 1 1 FULL TEXT CONVERSION REQUIRED TOTAL NUMBER OF COPIES REQUIRED: LTTR 42 ENCL 42

II Arizona Public Service Company P.O. BOX 53999 ~ PHOENIX, ARIZONA 85072.3999 192-00507-JGH/TDS/RKR August 14, 1989 U. S. Nuclear Regulatory Commission NRC Document Control Desk Washington, D.C. 20555

Subject:

Palo Verde Nuclear Generating Station (PVNGS)

Unit 2 Docket No, STN 50-529 (License No. NPF-51)

Licensee Event Report 89-009-00 File: 89-020-404 Attached please find Licensee Event Report (LER) No. 89-009-00 prepared and submitted pursuant to 10CFR 50.73. In accordance with 10CFR 50.73(d), we are herewith forwarding a copy of the LER to the Regional Administrator of the Region V office.

If you have any questions, please contact T. D. Shriver, Compliance Hanager at (602) 393-2521.

Very truly yours, J. G. Haynes Vice President Nuclear Product,ion JGH/TDS/RKR/kj Attachment cc: W. F. Conway (all w/a)

D. B. Karner E. E. Van Brunt, Jr.

J. B. Hartin T. J. Polich M. J. Davis A. C. Gehr INPO Records Center 8908f80269 8908f4 PDR 8

ADOCK 0 000529 PDC ~Q? 2.

I/I

II NAC For(h 244 UA. NUCLEAR AEOULATOAYCOANCISSION (Ml2)

APPAOVED OMS NO. 2150010l LICENSEE EVENT REPORT (LER) EXPIRES: SI21/SS FACILITYNAME (II DOCKET NUMSER (1l TITLE Ill Pal o Verde Uni t 2 0 5 00 0 529ioF09 Reactor Tri Due to Partial Loss of Forced Flow EVEN'7 DATE ISI LER NUMSER 14) REPOAT DATE (7) OTHER FACILITIES INVOLVED (Sl MONTH DAY YEAR SEQVENTIAL DAY YEAR FACILITYNAMES DOCKET NUMSER(S)

YEAR NVM4ER rRS NUMSEII MONTH N/A 0 5 0 0 0 0 7 1 2 8 9 8 9 0 0 9 00 08 14 8 9 N/A 0 5 0 0 0 THIS REPORT IS SUSMITTED PVASVANT TO THE AEOUIREMENTS OF 10 CFA g IChech one or more of the IOIIOwinpl (11)

OPERATINO MODE ( ~ I 20A02(4) 1 20AOS(c) X 50;724)(2)(>> ,

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NAME TELEPHONE NUMSER AREA CODE Timothy D. Shriver, Compliance Manager 60 23 93 2 52 1 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRI ~ EO IN THIS REPORT (11)

CAUSE SYSTEM COMPONENT MANVFAC. EPORTASLE MANUFAC EPOR'TASL CAUSE COMPONENT TURER TO NPRDS SYSTEM TUREA TO NPRDS

%%%M X E B FU GO 80 X S J V P032 Y SUPPLEMENTAL REPORT EXPECTED (Ill MONTH DAY YEAR EXPECTED SUSMISSION YES IffyN, coinpiete FA'PECPED SVFMISSION DilTEI NO DATE (ISI 12 0'j. 89 ASSTRACT ILimit to IC00 tpecet. I e., epproeimetefy ftfteen tinple.tpece typewritten IinNI (14)

On July 12, 1989 at approximately 2212 HST Palo Verde Unit 2 was operating at approximately 100 percent power when 2 of the 4 reactor coolant pumps were load shed from their power supply (Bus 2E-NAN-S02), resulting in a reactor trip on calculated low DNBR due to low reactor coolant flow. Immediately following the trip, a Safety Injection Actuation Signal (SIAS) and Containment Isolation Actuation Signal (CIAS) Engineered Safety Features occurred, on low Reactor Coolant System (RCS) pressure. Following the event, at approximately 1529 HST on July 13, 1989, a portion of the main feedwater system (HFWS) was overpressurized.

The cause of the load shed was a failed fuse in the bus potential transformer. The cause of the SIAS/CIAS was RCS depressurization due to improper Steam Bypass Control System (SBCS) response and leaking pressurizer spray valves. The cause of the HFWS overpressurization was a failed check valve.

Immediate corrective action taken was to replace the fuse. An independent investigation is being conducted to determine the causes of the incidents which occurred during this event.

This submittal also provides a Special Report in accordance with Technical Specification 3.5.2 ACTION b.

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DESCRIPTION. OF WHAT OCCURRED:

'. Initial. Conditions:

At approximately, 2212 MST on .July 12, 1989, Palo Verde, Unit 2,was in Mode 1 (POWER OPERATION) at approximately 100 percent power.

B. Reportable Event:Description (Including Dates and Approximate Times of Major Occurrences):

Event Classification: Any event or condition that resulted in manual or automatic actuation of any Engineered Safety Feature (ESF)(JE),

including the Reactor Protection System (RPS)(JC)'.

At approximately 2212 MST on July 12; 1989, a reactor (RCT)(AC) trip on low Departure from Nucleate Boiling Ratio (DNBR) occurred due to a partial loss of reactor coolant (AB) flow. The partial loss of,flow occurred when two (2) of the four'(4) Reactor Coolant

.Pumps (RCP)(AB) were load shed from their power supply (Bus 2E-NAN-S02)(BU)(EA). Following the reactor trip, at approximately 2213 HST a Safety Injection Actuation Signal (SIAS)(JE) and a Containment Isolation Actuation Signal (CIAS)(JE) occurred when the Reactor Coolant System (RCS)(AB) pressure decreased to approximately 1823 psia (approximately 14 psi below .the actuation setpoint). All safety .system components actuated as designed. The plant was stabilized at approximately 2322 HST and the event was terminated.

Prior to the event, at approximately 2001 HST on July 12, 1989, the Control Room received a trouble alarm annunciator (ALM)(ANN), for Bus 2E-NAN-S02. Operati.ons 'personnel (uti.lity, non-licensed) investigated and inspected Bus 2E-NAN-S02. They could not determine the reason for,the alarm on Bus 2E-NAN-S02. At approximately 2020 MST a control room annunc'iator alarm indicated that the unit oscillograph (OSG)'ad operated. Operations personnel (utility, licensed) responded to the annunciator alarm.

The oscillograph indicated that there had been an-:undervoltage condi,tion on Bus 2E-NAN'-S02. Also, the digital fault recorder (XR) printout indicated a disturbance on Bus 2E-NAN-S02. Since no apparent problems were identified with Bus 2E-NAN-S02, personnel'utility, licensed) continued to investigate unit'perations the .problem. From approximately, 2206 HST to approximately 2212 HST the Oscillograph alarmed three more times. Operations personnel were responding to the alarms when at approximatel'y 2212 HST, a load shed actuation occurred on 'Bus 2E-NAN-S02 even though 'Bus 2E-NAN-S02 remained energized. The load shed caused the loads on Bus 2E-NAN-S02 to be deenergized.

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t 2 ir rsuusssE u>> riess>>s h'pC Assr!I srsss'sl o so oo 529 8 9 0 0 9 0 0 03oF 0 9 11T1 The RCPs are powered from non-class 1E 13.8 kv Busses 2E-NAN-S01 (BU)(EA) and 2E-NAN-S02. RCPs 1A and 2A are powered from Bus 2E-NAN-S01 and RCPs 1B and 2B are powered from Bus 2E-NAN-S02. Bus 2E-NAN-S01 was being supplied by a Startup Transformer (XFHR)(EA)

,and Bus 2E-NAN-S02,was .being supplied by the Unit 2 Auxiliary Transformer (XFHR)(EA). Since RCP's 1B and 2B were being. supplied from Bus 2E-NAN-S02, the RCP's were deenergized and a partial loss of flow occurred resulting in a reactor trip on low DNBR.

Approximately one minute after the reactor trip; Reactor Coolant System pressure dropped lower than normal due to improper Steam Bypass Control System (SBCS)(SG) response and leaking pressurizer spray valves (PZR)(AB)(V). This resulted in concurrent safety injection and containment isolation Engineered Safety Features

.(ESF) actuations (JE) when the RCS pressure decreased to approximately 1823 psia, which is 14 psi below the low actuation pressure setpoint of 1837 psia and 1 psi above the minimum allowable trip setpoint value of 1822 psia. Immediately following the safety injection, pressurizer level and pressure stabilized.

Pressure then began to trend toward steady state Hode 3 (HOT STANDBY) conditions (approximately 2250 psia).

At approximately 2223 HST on July 12, 1989, a Notification of Unusual Event (NUE) was declared. The NUE was declared pursuant to EPIP-02, "Emergency Classification" as a result of the SIAS on low pressurizer pressure.

Operations personnel (utility, licensed and non-licensed) investigated the protective relay targets (RLY) on Bus 2E-NAN-S02 and could find no reason for the protective relay actuation. At approximately 2234 HST on July 12, 1989, the load centers powered from Bus 2E-NAN-S02 were reenergized in accordance with an approved procedure. Eight minutes later at approximately 2242 HST, a load shed signal again deenergized the load centers on Bus 2E-NAN-S02.

At approximately 2302 HST, Bus 2E-NAN-S02 was deenergized and taken out of service in order to perform further troubleshooting in accordance with the PVNGS Work Control Program.

At approximately 2322 HST on July 12, 1989 the SIAS/CIAS ESF actuations were secured, plant conditions were stabilized, and the NUE was terminated.

At approximately 0300 HST on July 13, 1989, Protection Relaying and Control (PR&C), the APS group responsible'for investigating the oscillograph operated alarm, reset the alarm in accordance with procedure 42AL-2RKlB, "Panel BOIB Alarm Response".

Following the event, at approximately 1529 HST on July 13, 1989, the secondary plant was being placed in the long path recirculation NAC srs1M )99A

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(LPR) mode in accordance with normal operating procedures. Steam generator levels (SG)(AB) were being maintained using the.

Non-essential Auxiliary Feedwater Pump (P)(BA). Due to back-leakage through a check valve, Auxiliary Feedwater System (AFW)(BA) pressure of approximately 1580 psia was applied to a portion of the 'Hain 'Feedwater System (HFWS)'(SJ). This portion of the Main Feedwater System was rated for at least 1580 psia. When the manual isolation valve (ISV) on the Main Feedwater Pump (HFP)

RBR accord with an approved~rocedure, Hain Feedwater Pumps raAd for 500 psia.

~

bypass line was opened to establish long path recirculation in rpressurized.

the suction piping to the The suction piping is Concurrent with the HFP bypass valve being opened, a low suction RAR and RB" was received in pressure trip for Hain Feedwater Pumps the Control Room. Immediately following the low suction pressure trip, operations personnel (utility,feedwater non-licensed) observed an heater (SN) outlet abnormal decrease in seventh point temperature. To prevent thermal shocking of the feedwater heater, operations personnel 'isolated the Non-essential Auxiliary Feedwater System from the Hain Feedwater System. At approximately 1545 HST, the overpressurization event was terminated.

C. Status of structures, systems, or components that were inoperable at the start of the event that contributed to the event:

There were no structures, systems, or components inoperable at the start of the event which contributed to the event.

D. Cause of each component or system failure, if known:

The cause of the Bus 2E-NAN-S02 load shed malfunction described in Section I.B has been determined to be RCR a failed potential transformer (PT) primary fuse on the phase. The cause of the fuse failure is still under investigation and is expected to be completed by November 1, 1989. The cause will be described in a supplement to this report which is expected to be submitted by December 1, 1989.

The cause of the back-leakage through the MFP bypass check valve (SGN-V431) described .in Section I.B has been determined to be loose fasteners which allowed the check valve disc to drop and not seat on its seating surface. The fastener locking devices required by the Technical Manual were not installed. The cause for the locking device not being installed cannot be determined since our records show no work has been performed on this valve since initial startup of Unit 2. This valve is a "non-safety related" component.

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Oi NAC PS/M SSSA I989 I U.S. NUCLEAR AEOVLATOAYCOMMISSION LICENSEE EVENT REPORT ILER) TEXT CONTINUATION APPAOVEO OM8 NO SI50 OIOS EXPIRES: 8/11/88 PACILITY NAME III OOCKET NUMSER 11I LEA NVMSEA ISI AOE IS)

YEAR ',..y SEOVENS/AL ~ A 1 v IS IO //

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. E Failure mode, mechanism, and effect of each failed component, if known:

The failure of the potential transformer (PT) primary ACA .phase

,fuse resul,ted in a load, shed of,Bus 2E-NAN-S02 .including deenergization of reactor coolant pumps (RCP) 1B and 2B. This resulted in a reactor trip and turbine trip as described in Section I.B. Shortly before the reactor trip, there was indication of a problem with Bus 2E-NAN-S02 by a trouble alarm annunciation for the bus, an oscillograph operated annunciator alarm, oscillograph indication of an undervoltage condition on the bus, and a digital fault recorder indicating a fault on the bus. Operations personnel (utility, licensed and non-licensed) were attempting to identify the problem when the reactor trip occurred. The root cause of failure of the fuse is under investigation and will be included in a supplement to this report.

The failure of check valve SGN-V431 resulted in overpressurizing a portion of the main feedwater system as described in Section I.B.

The cause of the check valve failure is described in Section I.C.

F. For failures of components with multiple functions, list of systems or secondary functions that were also affected:

Not applicable - No component failures had multiple functions which affected other systems or components.

G. For failures that rendered a train of a safety system inoperable, estimated time elapsed from the discovery of the failure until the train was returned to service:

Not applicable - There were no failures that rendered a train of a safety system inoperable.

H. Method of discovery of each component or system failure or procedural error:

The failed potential transformer primary "CA phase fuse was discovered as a result of troubleshooting performed after the event.

The failed check valve SGN-V431 was discovered as a result of troubleshooting performed after the event.

The overpressurization of the MFW,pump suction piping was discovered when troubleshooting the feedwater pump low suction pressure trips as described in Section I.B. The bellows in all six low suction pressure switches were found to be deformed due to the pressure transient.

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LICENSEE EVENT REPORT (LER) TEXT CONTINUATION APPROVEO OMd HO 3150WIOS EXPIREE: d/31/dd FACILITY NAME 111 OOCKET NUMEEA Ill LER NUMeER lel PACE IS)

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//9///M//V/ICFO//R JISIIA'SI 113) o 5 o o o 5 298 9 009. 00 06 QF 0 9 I. Cause of Event:

The cause of the reactor trip discussed in Section I.B was a failed fuse which caused a load shed signal on Bus 2E-NAN-S02 resulting in a partial loss of flow. The RCR phase potential transformer (PT) is connected to 'Bus 2-NAN-S02. 'The fuse is located between the PT and the bus. Relay 227-S monitors the bus voltage through the PT. The relay load sheds the bus when the voltage is less than or equal to 77 percent of the bus voltage (13.8 kv). When the fuse failed, the relay saw no voltage from the PT and actuated the load shed relays.

2. The cause of the SIAS/CIAS ESF actuation described'.~n Section I.B. was RCS depressurization due to improper Steam Bypass Control System (SBCS) response and excessive leaRage past the pressurizer spray valves.

a. The SBCS response was caused by the method of calibration used for the Proportional Integral (PI) Controller in the SBCS quick open controllers. This allowed the SBCS valves to remain open longer than was required. PVNGS calibrates the PI controller using the dial settings provided by the Combustion-Engineering (CE) setpoint document. The CE setpoint document also provides calibration curves for optimizing the quick open controller response. However, the CE setpoint document did not clearly indicate that these curves were to be used as part of the PI controller calibration.

Therefore, PVNGS calibration procedures only required use of the dial settings and did not require use of the calibration curves to optimize the SBCS response.

b. The cause of the pressurizer spray valve leakage was due to both valves positioners being out of ~libration.

PVNGS was aware that these valves were ldhking prior to this event. However, because these valves are located in containment and inaccessible during power operation, they were scheduled to be recalibrated during the next plant shutdowns The final root cause analysis is not complete for the positioners. The root cause analysis is expected to be completed by November 1, 1989. The results of the root cause will be included in a supplement to this report which is expected to be submitted by December 1, 1989.

3. The cause of the overpressurization of the Hain Feedwater Pump suction piping was due to back-leakage through a check valve as described in Section I.D. This check valve is not safety related and is not currently included in any testing 4RC ~ S/RM SSSR 19 93/

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aMakootl Hale fonaa %CA'tt till o s o o o 5 2 9 9 or maintenance programs. Therefore there were no procedures for maintenance or testing of these check valves at the time of this event. A separate engineering evaluation is ofbeing this conducted on this overpressurization event. A copy evaluation will be provided to the NRC resident inspector at

'PVNGS .and the 'Regional 'Administrator.

J. Safety System Response:

following automatic and manual safety system responses occurred The during this event:

1. Containment Isolation System (automatic)(JH). RBR

2. Low. Pressure Safety Injection 'Trains RAR and (automatic)(BP)

Safety Injection Trains "AR and RBR

3. High Pressure (automatic)(BQ) RAR and RBR Emergency Diesel Generators Trains (automatic)(DG)(EK)

5. Essential Spray Pond System Trains RAR "AR and RBR (automatic)(BS)

6. Essential Chilled Water System Trains and "Bv (automatic)(KH) RAR and RBR

7. Essential Cooling Water System Trains (automatic)(BI) RAR and RBR (automatic)(KA)

8. Condensate Transfer System Trains

9. Containment Spray Trains RAR and RBR,(automatic)(BE)

RA" and "BR

10. Control Room Essential HVAC Trains (automatic)(AHU) RAR and Control Building Essential Ventilation System Trains "Bv (automatic)(AHU)

"AR and RB"

12. Auxiliary Building Essential HVAC System Trains (automatic)(AHU)

K. Failed Component Information:

The failed fuse was manufactured by General. Electric. It is a type EJl, size B, rated at 15.5 kv and 0.5 amps.

The leaking check valve was manufactured by number Pacific Valves. It is 58809-7-WE(20) an 8 inch one-way flow check valve, figure ASSESSMENT OF THE SAFETY CONSEQUENCES AND IMPLICATIONS OF THIS EVENT:

The Palo Verde Updated Final Safety Analysis Report (UFSAR) accident analysis for loss of reactor coolant system (RCS) flow assumes a total loss of offsite power resulting in a coastdown of a

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1; NRC Porm SSIA 19891. U.S. NUCLEAR REGULATORY COMMISSION LICENSEE EVENT REPORT (LERI TEXT CONTINUATION APPROVEO OMS NO 9150WIOO EXPIRES: 8/SI/88 P A OIL I T Y N AME 111 OOCKET NUMSER I?I LER NUMSER ISI PAGE ISI YEAR r,",g STOVE NT>>AL A1 y IE>>Q N N I/ M e $ A >>>>vMoeo Palo Verde Unit 2 o 5 o,o o 52 9 89 0 09 00 08 OF 0 9 TEXT /// more I/>>PIe>>I e>>>>iire>>E>>roe>>oeloon>>o/ H/IC /rorm JRKAS/117) all four reactor coolant pumps (RCP's). The accident analysis transient is Departure from Nucleate Boiling Ratio (DNBR) limiting. The reduced RCS flow results in an initial rise in RCS average temperature and a reduction in DNBR. Based on this analysis, a reactor trip on low DNBR mitigates this transient and maintains DNBR above the safety limit. For this event, only two RCPs tripped .and coasted down. The Steam Bypass Control System (SBCS) reduced RCS average. temperature following the reactor trip. The accident analysis bounds this event. Based on this, DNBR limits were not exceeded.

Depressurization of the RCS resulted in a SIAS. The primary function of the SIAS for this event type is to maintain RCS inventory and maintain shutdown margin. In this event all control element assemblies (CEA)(AA)(ROD) inserted and RCS average temperature decreased to 551'F.

Adequate shutdown margin was maintained and,pressurizer level remained on scale throughout the event. Therefore adequate RCS inventory was, maintained throughout this event.

The check valve SGN-V431 leakage resulted in overpressurization of a portion of the Main Feedwater System pump suction piping. This portion of the Hain Feedwater System performs no safety function.

All safety systems required to operate performed as designed. The event did not result in any challenges to fission product barriers or result in any releases of radioactive materials. Therefore, there were no safety consequences or implications as a result of this event. This event did not adversely affect the safe operation of the plant or health and safety of the public.

III. CORRECTIVE ACTIONS:

A. Immediate:

The failed fuse was replaced in the potential transformer for Bus 2E-NAN-S02. >>

B. Action to Prevent Recurrence:

The failed fuse's being evaluated by General Electric for the root cause of failure determination. B'ased on the root cause of failure, PVNGS will determine if any additional actions are required to prevent recurrence. Additional actions will be described in a supplement to this report.

2. The SBCS calibration procedure was revised and the SBCS quick open modules were optimized using the nominal setpoint values and the controller performance curve. The SBCS calibration procedure is being reviewed to ensure that all SBCS modules are calibrated using the technique, setpoints and tolerances 'n the Combustion Engineering setpoint document.

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3. The pressurizer spray valves were recalibrated in accordance with an, approved procedure to correct the excessive. leakage condition. The final root cause determination has not been completed. Additional actions will be described in a supplement to this report.

4. The leaking HFP bypass check valve was repaired. Check valves SGN-V431 and SGN-V432 were pressure tested and no leakage was identified. The overpressurization event was evaluated and Hain Feedwater Pump suction piping was walked down. It was determined to be acceptable for continued plant operation.

C. Corrective Actions by Other Units:

Units 1 and 3 will complete the following actions prior to startup from their current outages:

1. Check the optimization of the SBCS quick open modules.

2. Check the calibration of the pressurizer spray valves.

3. Inspect check valves SGN-V431 and SGN-V432.

An independent investigation of this event is also being conducted. Additional actions to prevent recurrence may be developed based upon the results of this independent'valuation. A supplement to this report will be provided to describe additional corrective actions to be taken. The supplement to this report is expected to be submitted by December 1, 1989.

IV. PREVIOUS SIHILAR EVENTS:

There have been:no previous similar occurrences reported pursuant to 10CFR50.73.

'There have been previous reactor trips reported. However, none of,the previous reactor trips were attributable to. the same root cause described in Section I. I. Therefore none of the'previous corrective actions would have been expected. to prevent this event.

V. ADDITIONAL INFORMATION There have been 5 total .accumulated actuation cycles of the Emergency Core Cooling System to date.. This report satisfies the requirements of Technical Specification 3.5.2 ACTION b.

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