ML17313A889

LER 99-001-00:on 990310,RT on High Pressurizer Pressure Was Noted.Caused by Loss of Heat Removal.Cr Supervisor Was Removed from Shift Duties for Diagnostics Skills Training. with 990409 Ltr

ML17313A889
Person / Time
Site:	Palo Verde
Issue date:	04/09/1999
From:	Marks D, Overbeck G ARIZONA PUBLIC SERVICE CO. (FORMERLY ARIZONA NUCLEAR
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
192-01042-GRO-D, 192-1042-GRO-D, LER-99-001-02, LER-99-1-2, NUDOCS 9904200382
Download: ML17313A889 (18)

Text

)I 1 J REGULATORY INFORMATION DISTRIBUTION SYSTEM (RZDS)ACCESSION NBR:9904200382 DOC.DATE: 99/04/09 NOTARIZED:

NO DOCKET FACIL:STN-50-528 Palo Verde Nuclear Station, Unit 1, Arizona Publi 05000528 AUTH.NAME-AUTHOR AFFILIATION MARKS,D.G.

Arizona Public Service Co.(formerly Arizona Nuclear Power OVERBECK,G.R.

Arizona Public Service Co.(formerly Arizona Nuclear Power RECZP.NAME RECIPIENT AFFILIATION

SUBJECT:

LER 99-001-00:on 990310,RT on high pressurizer, pressure was noted.Caused by loss of heat removal.CR supervisor was removed from shif t duties for diagnostics skills training.With 990409 ltr.DISTRIBUTION CODE: IE22T COPIES RECEIVED:LTR ENCL SIZE: TITLE: 50.73/50.9 Licensee Event Report (LER), ncident Rpt, etc.NOTES: STANDARDIZED PLANT 05000528 C RECIPIENT ID CODE/NAME LPD4-2 PD INTERNAL: ACRS LE CE ER NRR DIPM IQMB NRR/DSSA/SPLB RGN4 FILE'01 COPIES LTTR ENCL 1 1 1 1 1 1 1 1 1 1 1 1 RECIPIENT ID CODE/NAME FIELDS,M AEOD/SPD/RRAB NRR/DIPM/IOLB NRR/DRIP/REXB RES/DET/EIB COPIES LTTR ENCL 1 1 1 1 1 1 1 1'1 1 EXTERNAL: L ST LOBBY WARD NOAC POORE,W.NRC PDR 1 1 1 1 1 1 LMITCO MARSHALL NOAC QUEENER,DS NUDOCS FULL TXT 1 1 1 1 1 1 NOTE TO ALL"RIDS" RECIPIENTS:

PLEASE HELP US TO REDUCE WASTE.TO HAVE YOUR NAME OR ORGANIZATION REMOVED FROM DISTRIBUTION'LIST OR REDUCE THE NUMBER OF COPIES RECEIVED BY YOU'OR YOUR ORGANIZATION, CONTACT THE DOCUMENT CONTRO DESK (DCD)ON EXTENSION 415-2083 FULL TEXT CONVERSION REQUIRED TOTAL NUMBER OF COPIES'EQUIRED:

LTTR 17 ENCL 3/7

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&every.Paio Verde Nuciear Generating Station Gregg R.Overbeck Vice President Nuclear Production TEL 602/393-5148 FAX 602I3934077 Mail Station 7602 P.O.Box 52034 Phoenix, AZ 85072-2034 192-01042-GRO/DGM/RjH April 9, 1999 U.S.Nuclear Regulatory Commission ATi N: Document Control Desk Mail Station P1-37 Washington, D.C.20555-0001

Subject:

Palo Verde Nuclear Generating Station (PVNGS)Unit 1 Docket No.STN 50-528 License No.NPFP1 Licensee Event Report 99-001-00 Attached please find Licensee Event Report (LER)99-001-00 prepared and submitted pursuant to 10 CFR 50.73.This LER reports an automatic reactor trip on high pressurizer pressure due to a loss of heat removal.In addition to the reactor trip, the unit received an engineered safety feature actuation of the main steam system (MSIS)on high steam, generator water level.There are.no commitments generated as a result of this LER.In accordance with 10CFR50.73(d), a copy of this LER is being forwarded to the Regional Administrator, NRC Region IV.If you have any questions, please contact Daniel G.Marks, Section Leader, Regulatory Affairs, at (623)393-6492.-

Sincerely, GRO/DGM/RJH/rlh Attachment cc: E.W.Merschoff P.H.Harrell M.B.Fields J.H.Moorman INPO Records Center 9904200382 990409~PDR ADOCK 05000528 S PDR (all with attachment)

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NRC FORM 366 (5-1998}U.S.N CLEAR REGULATORY COMMISSION LICENSEE EVENT REPORT (LER)(See reverse for required number of digits/characters for each block)APPRO BY OMB NO.3160%104 EXPIRES 06/30/2001 Estimated burden per response to comply with this mandatory information collection request: 50 hrs.Reported lessons learned are incorporated into the licensing process and fed back to industry.Forward comments regarding burden esumate to the Records Management Branch (TW F33), U.S.Nudear Regulatory Commission.

Washington, DC 20555-0001

~and to the PapenNork Reduction Project (3150-0104}, office of Management and BydgeL Washington.

DC 20503.If an information collection does not display a currently valid OMB control number, the NRC may not conduct or sponsor, and a person is not required to respond to, the information cosection FACILITY NAME{1}Palo Verde Nuclear Generating Station Unit1 DOCKETNUMBER{2}

05000528 PAGE{3}1 OF 7 Reactor Trip On High Pressurizer Pressure Due To A Loss Of Heat Removal EVENT DATE (5)LER NUMBER (6)REPORT DATE (7)OTHER FACILITIES INVOLVED (8)MONTH DAY YEAR 03 10 1999 SEQUENTIAL REVISION YEAR NUMBER NUMBER 1999-001 00 MONTH DAY YEAR 04 09 1999 FACILITY NAME N/A FACILITY NAME N/A DOCKET NUMBER 05000 DOCKET NUMBER 05000 OPERATING MODE (9)THIS REPORT IS SUBMITTED 20.2201 b PURSUANT TO THE REQUIR 20.2203 a 2 v EMENTS OF 10 CFR: Check o 50.73 a 2 i ne or more 11 50.73 a 2 vru POWER LEVEL (10)100 20.2203 a 1 20.2203 a 2 i 20.2203 a 2 ii 20.2203 a 2 iii 20.2203 a 2 iv 20.2203 a 3 I 20.2203 a 3 ii 20.2203 a 4 50.36 c 1 50.36 c 2 LICENSEE CONTACT FOR THIS LER (12)50.73 a 2 u 50.73 a 2 iii 50.73 a 2 iv 50.73 a 2 v 50.73 a 2 vii 50.73 a 2 x 73.71 OTHER Specify in Abstract below or ln NRC Form 3BBA NAME anie.Mar s, ection ea er, egu atory A airs TELEPHONE NUMBER (InoNde Area Code)623-393-6492 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT 13 CAUSE SYSTEM COMPONENT MANUFACTURER REPORTABLE To EPIX CAUSE SYSTEM REPORTAB COMPONENT MANUFACTURER LE TO'PIX X SJ COMP F180 BV SB IICNTRL G080 SUPPLEMENTAL REPORT EXPECTED 14 YES (If yes, complete EXPECTED SUBMISSION DATE).X NO EXPECTED SUBMISSION DATE (15)MONTH DAY YEAR ABSTRACT (Limit to 1400 spaces, i.e., approximately 15 single-spaced typewritten lines)(16)On March 10, 1999 at approximately 1325 MST, Palo Verde Unit 1 was in Mode 1 (POWER OPERATION), operating at approximately 100%power, when a reactor trip occurred due to a loss of heat removal capability caused by improper operator intervention following a spurious closure of the main turbine control*valves.An electrical malfunction caused the main turbine control valves to close decreasing the electrical load on the turbine.The steam bypass control system (SBCS)responded to maintain the loss of heat removal caused by the load shed.The control room supervisor directed the secondary operator to place the SBCS to emergency off which initiated a reduction in heat removal and resulted in a reactor coolant pressure increase to the high pressurizer pressure setpoint, subsequently tripping the reactor.In addition to the reactor trip, the unit received an engineered safety feature actuation system actuation of the main steam system (MSIS)on high steam generator-2 level due to the steam generator-2 economizer valve not staying fully closed and allowing level to reach the MSIS setpoint.At approximately 1335 MST, the unit was stabilized in Mode 3 (HOT STANDBY)and the shift manager classified the event as an uncomplicated reactor trip.All safety systems functioned as required.There were no other safety system actuations and none were required.The preliminary cause of the turbine control valves going closed was a failure of the throttle pressure sensing circuit.An equipment root cause of failure is in progress.The cause of the reactor trip was attributed to supervisory methods that did not ensure sufficient information was collected to support an operational decision.No previous similar events have been re orted ursuant to 10CFR50.73.

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NRC FORM 366A (6-1998)U.S.NUCLEAR REGULATORY COMMISSION LICENSEE EVENT REPORT (LER)TEXT CONTINUATION FACILITY NAME (1)Palo Verde Nuclear Generating Station Unit 0 TEXT (lf more spaceis required, use additional copies of NRC Form 366A)(17)DOCKET NUMBER(2)05000528 LER NUMBER (6)SEQUENTIAL REVISION NUMBER NUMBER 1999-001-000 PAGE (3)2 OF 7 1.REPORTING REQUIREMENT:

This LER 530/99-001-00 is being submitted to report an event that resulted in the automatic actuation of an engineered safety feature (ESF)(JE), including the reactor protection system (RPS)(JC)as specified in 10CFR50.73(a)(2)(iv).

Specifically, on March 10, 1999, at approximately 1326 MST, Palo Verde Unit 1 was in Mode 1 (POWER OPERATION), operating at approximately 100%power, when a reactor trip occurred following a spurious closure of the main turbine control valves (CV).An electrical malfunction caused the main turbine control valves to close decreasing the electrical load on the turbine.The steam bypass control system (SBCS)responded to the loss of heat removal caused by the load shed.Operations personnel (utility-licensed operator)noted that all eight steam bypass control valves opened and concurrently verifiedno rod bottom lights were lit on the core mimic.The control room supervisor directed.the secondary operator to place the SBCS in emergency off which initiated a reduction in heat removal and resulted in a reactor coolant pressure increase to the high pressurizer pressure setpoint, subsequently tripping the reactor.In addition to the reactor trip, the unit received an engineered safety feature actuation system (ESF)actuation of the main steam system (MSIS)on high level in steam generator-2 due to the steam generator-2 economizer valve not staying fully closed and allowing level to reach the MSIS setpoint.In accordance with 10CFR50.72 (b)(2)(ii), at approximately 1645 MST on March 10, 1999, a one hour non-emergency ENS call (ENS ID 35456)was made for a,RPS/ESF actuation.

2.EVENT DESCRIPTION:

Prior to the automatic trip, at approximately 1325 MST on March 10, 1999, control room personnel observed EHC and other alarms via the RJ computer screen.Subsequently, all eight steam bypass control system (SBCS)permissives activated and turbine bypass control valves quick opened.Control room personnel observed no trip annunciators on the main turbine, no rod bottom lights lit, no CEAs inserted on the CEAC CRT screen, and no reactor power cutback, concurrent with the SBCVs opening and suspected a possible overpower event was occurring and placed the SBCS in emergency off.As a result of the decrease in heat removal, reactor pressure increased to the high pressurizer pressure setpoint 2283 psia and at approximately 1326 MST, the reactor automatically tripped on high pressurizer pressure.All control element assemblies(CEA)(AA)inserted as designed.The reactor trip initiated a main turbine/main generator trip (TA/TB).Secondary system pressure was discharged briefly via atmospheric dump valves (ADVs)and main steam safety valves (MSSVs).The total discharge duration was approximately 12 minutes of which 4 minutes were on the safeties.

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NRC FORM 366A (6-1996)FACILITY NAMg (1)U.S.NUCLEAR REGULATORY'COMMISSION LICENSEE EVENT REPORT (LER)TEXT CONTINUATION DOCKET NUMBER(2)LER NUMBER (6)PAGE (3)Palo Verde Nuclear Generating Station Unit1 TEXT (If more space is required, use additional copies of NRC Form 366A)(17)05000528 SEQUENTIAL, REVISION NUMBER NUMBER 1999-001-000 3 OF 7 2.EVENT DESCRIPTION cont': The unit was stabilized at approximately 1335 MST in Mode 3 (Hot Standby)and the Shift Manager classified the event as an uncomplicated reactor trip.At.approximately 1429 MST a main steam isolation signal (MSIS)safety system actuation occurred due to high steam generator-2 levels The level in steam generator-2 increased to the MSIS setpoint of 91 percent narrow range.The level increase in steam generator-2 occurred approximately five minutes after the secondary operator placed both downcomer valves in the automatic position.The cause of the level increase was due to a faulty timer card for the economizer control valve..There were no other safety system actuations and none were required.Required safety systems responded to the event as designed.The timer card was tested and re-installed..

An equipment root cause of failure is underway to determine the problem.3.ASSESSMENT OF THE SAFETY CONSEQUENCES AND IMPLICATIONS OF THIS EVENT: A safety limit evaluation was performed as part of the APS investigation.

The evaluation determined that the plant responded as designed, that no safety limits were exceeded, and that the event was bounded by the current safety analysis.The reactor trip did not result in a transient more severe than those already analyzed in Chapter 15 of the Updated Final Safety Evaluation Report (UFSAR).Scenarios concerning a decrease in heat removal by the secondary system or loss of normal feedwater flow remain bounding for this event.All CEAs inserted as designed.This event did not challenge shutdown margin criteria.Both primary and secondary system pressures remained well below 110 percent of the design pressures.

Four main steam safety valves lifted for approximately 4 minutes during the event.Engineering review of the safety valve performance indicated that all four safety-valves performed as required.Post trip testing of safety valve performance is discussed in the additional information section 8.The specified acceptable fuel design limits (SAFDL)were not exceeded during this event.The pl'ant response was normal for the situation that occurred.The event did'ot result in any challenges to the fission product barriers or result in the release of radioactive materials in excess of quantities allowed by 10CFR20.Therefore, there were no adverse safety consequences or implications as a result of this event.This event di.d not adversely affect the safe operation of the plant or the health and safety of the public.

4i I NRC FORM 366A (6-1998)FACILITY NAME (1)U.S.NUCLEAR REGULATORY COMMISSION LICENSEE EVENT REPORT (LER).TEXT CONTINUATION DOCKET NUMBER(2)LER NUMBER (6)PAGE (3)'alo Verde Nuclear Generating Station Unit 1 05000528 TEXT (If more space is required, use addi Iional copies of NRC Form 366A)(17)SEQUENTIAL REVISION NUMBER NUMBER 1999-001-000 4 OF 7 4.CAUSE OF THE EVENT: 'An independent investigation of this event is being conducted in accordance with the APS corrective action program.The investigation determined, that the reactor tripped due to high pressurizer pressure when secondary heat removal was reduced by operations placing the SBCS to emergency-off.

A preliminary evaluation has determined'he apparent root cause to be personnel error in that supervisory methods did not ensure that sufficient information was collected to support operations decision to place the SBCS in emergency-off.

The apparent cause of the turbine control valves closing was a failure of the throttle pressure sensing circuit.An equipment root cause evaluation is underway for'his issue.The apparent cause of the post trip MSIS was due to a faulty timer card that controls the steam generator economizer valves.During the event the negative ten percent signal to the economizer valve was lost.The loss of the signal was due to loss of power to the card through its back plate connector..

The card was removed and tested in the rework shop and performed satisfactory.

The card,was re-installed in the field panel and tested several times and functioned'roperly.

An equipment root cause evaluation is underway for this issue.P No unusual characteristics of the work location (e.g., noise, heat, poor lighting)directly contributed to the event.No procedural errors contributed to the event.5.STRUCTURES, SYSTEMS, OR COMPONENTS INFORMATION:

An independent investigation of this event is being conducted in accordance with the APS corrective action program.The investigation has determined that the reactor automatically tripped on high pressurizer pressure following the spurious closing of the main turbine control valves causing a large load reduction.

A prel'iminary root cause of failure analysis (ERCFA)determined that the apparent root cause of the main turbine control valves closing is attributed to an electronic failure in the throttle pressure sensing circuit which controls valve position and regulates steam flow to the main turbine.The failure of the sensing circuit caused all four main turbine control valves to close.The failed pressure sensing circuit is currently in the instrument and control rework facility to determine the specific failure mode.A.new pressure sensing circuit was installed and turbine control valves were returned to service.If the final ERCFA results differ from this determination, a supplement to this report will be submitted to describe the fina3.root cause determination., A transportability review was conducted for Units 2 and 3 to determine if a potential failure of the pressure sensing circuit exists.No abnormal failure modes or repetitive problems were discovered.

~I ig),t NRC FORM 366A (6-1 996)FACILITY NAME (1)U.S.NUCLEAR REGULATORY COMMISSION LICENSEE EVENT REPORT (LER)TEXT CONTINUATION DOCKET NUMBER(2)LER NUMBER (6)PAGE (3)Palo Verde Nuclear Generating Station Unit1 TEXT (if more space is required, use additional copies of NRC Form 366A)(17)05000528 SEQUENTIAL REVISION NUMBER NUMBER 1999-001-000 5 OF 7 5.STRUCTURES, SYSTEMS, OR COMPONENTS INFORMATION cont': An evaluation was also conducted to determine the cause of the MSIS.A preliminary ERCFA determined.that the apparent root cause for the MSIS, post trip, was due to a problem with the timer card on the feedwater control system causing the steam generator-2 economizer valve to remain open approximately 1 percent, increasing the level in steam generator-2 to the high setpoint for MSIS.During the event the negative ten percent signai to the economizer valve was lost.The loss of the signal was due to loss of power to the card through its back plate connector.

The card was re-installed in the field panel and tested several times and fun'ctioned properly.This card is normally inactive during normal operations and is needed during feedwater crossover or during post trip conditions.

The risk of a timer card failure during normal operations is low.After extensive troubleshooting efforts, the problem with the card could not be repeated and was determined to be satisfactory for reinstallation.

The timer card was re-installed in the system for continued use.A root cause evaluation is underway to determine the failure mode.A transportability review will be conducted as part of the ERCFA.The timer card is manufactured by Foxboro and is a Spec 200 part number KY354F2.There are no indications that any structures, systems, or components were inoperable at the start of the event, which contributed, to the event.No other component or system failures were involved.No failures of components with multiple functions were involved.No failures that rendered a train of a safety system inoperable were involved.6.CORRECTIVE ACTIONS TO PREVENT RECURRENCE:

As immediate corrective action, the control room supervisor was removed from shift duties for diagnostics skills training.A human performance evaluation

.(HPES)is underway to determine personnel performance issues surrounding the event.Preliminary results from the HPES indicate that the primary cause was inadequate operator response in that the control room staff did not fully evaluate plant parameters prior to making the decision to place the SBCS in emergency-off.

The evaluation also concluded that operations procedures, command and control, communications, shift staffing and stress levels were all satisfactory,and did not contribute to the event.If the final HPES evaluation results differ from those previously stated, a supplement to this report will be submitted to describe the final HPES determination.

ig)IP NRC FORM 366A (6-1998)FACILITY NAME (1)U.S.NUCLEAR REGULATORY COMMISSION LICENSEE EVENT REPORT (LER)TEXT CONTINUATION DOCKET NUMBER(2)LER NUMBER (6)PAGE (3)Palo Verde Nuclear Generating Station Unit 1 TEXT (if more space is required, use addi!ional copies of NRC Form 366A)(17)05000528 SEQUENTIAL REVISION NUMBER NUMBER 1999-001-000 6 OF 7 6.CORRECTIVE ACTIONS TO PREVENT RECURRENCE cont'd: An ongoing evaluation is being conducted in accordance wi;th the APS Corrective Action Program to address the root cause of failure for the main turbine control valve and steam generator economizer control valve failures.As immediate corrective action, the main turbine control valve throttle pressure sensing circuit was replaced.An ERCFA is underway to determine the failure mode and will be completed by July 30, 1999.'uring the event the negative ten percent signal to the economizer valve was lost.The loss of the signal was due to loss of power to the card through its back plate'connector.

The card was re-installed in the field panel and'ested several times and functioned properly.The timer card for the steam generator economizer control valve was tested satisfactorily, re-installed and returned to service.After extensive troubleshooting efforts, the problem with the card could not be repeated and was determined to be satisfactory for reinstallation.

The timer card was re-installed in the system for continued use.A root cause evaluation is underway to determine the failure mode and will be completed by July 30, 1999.A transportability review will be conducted as part of the ERCFA.To enhance operator awareness of the anomalies associated with this event, the lessons learned will be placed in the Licensed Operator Continued Training (LOCT).As part of LOCT, addi:tional training will be given on anomalies associated with isolation of the steam generator economizer valves by August 15, 1999.Operations management wi.'ll evaluate the need for additional diagnostic skills training for licensed operators following completion of the HPES evaluation by June 30, 1999.7.PREVIOUS SIMILAR EVENTS: Although previous similar events have been reported pursuant to 10CFR50~73 in the past three years for automati'c actuation o'f an ESF, including RPS, the causes discussed in the previous events have not been similar to this event.Therefore, the corrective actions taken to address previous similar events, would not have prevented this event.

Ol ig>I NRC FORM 366A (6-1996)U.S.NUCLEAR REGULATORY COMMISSION LICENSEE EVENT'REPORT (LER)TEXT CONTINUATION FACILITY NAME (1)~DOCKET NUMBER(2)LER NUMBER (6)PAGE (3)Palo Verde Nuclear Generating Station Unit 1 05000528 TEXT (If more space is required, use additional copies of NRC Form 366A)(17)SEQUENTIAL REVISION NUMBER NUMBER 1999-001-000 7 oF 7 8.ADDITIONAL INFORMATION Unit Restart Based on the contingency action plan and on reviews by the Plant Review Board, the Management Response Team, and the Incident Investigation Team, unit restart was authorized by the Operations Director in accordance with approved procedures.

At approximately 2228 on March 12, 1999,'nit 1 entered Mode l,and at approximately 0121 MST on March 13, 1999 Unit 1 was synchronized on the grid.MSSV Performance Subsequent to the reactor trip, secondary system pressure rose above 1250 psig resulting in the operation of four MSSVs.Based on available instrument readings, four MSSVs lifted within their expected lift pressure settings and limited secondary peak pressure to approximately 1282 psig during the event.One MSSV (SGE-PSV-561), with a design lift of 1250 psig, did not lift during the event.Engineering performed an evaluation of SGE-PSV-561 performance and determined that the actual steam line pressure that this valve experienced during the event may not have been sufficient to cause it to lift.Post trip testing was conducted for this MSSV to confirm the lift setting was within Technical Specification acceptance limits.The as found lift setting was determined to be approximately 3.8 percent above the design lift pressure which is outside the Technical Specification limit.The MSSV was declared inoperable at approximately 1434 MST, on March 13, 1999 and the lift settings were adjusted and returned to service within the LCO completion time.Although SGE-PSV-561 post trip test results were outside of Technical Specification limits, there were no safety consequences's a result of this condition in that the MSSVs performed as expected during the event and prevented over-pressurization of the secondary system.

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