ML17304A249

LER 88-016-01:on 880514,reactor Trip Occurred as Control Element Assemblies Inserted Following Attempt to Start Up Reactor.Caused by Nonconservative Operator Performance During Startup.Logkeeping Practices improved.W/880707 Ltr

ML17304A249
Person / Time
Site:	Palo Verde
Issue date:	07/07/1988
From:	Haynes J, Shriver T ARIZONA PUBLIC SERVICE CO. (FORMERLY ARIZONA NUCLEAR
To:	NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM)
References
192-00392-JGH-T, 192-392-JGH-T, LER-88-016, LER-88-16, NUDOCS 8807120562
Download: ML17304A249 (36)

Text

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".'- ACCESSION NBR:8807120562 DOC.DATE: 88/07/07 NOTARIZED: NO DOCKET FACZL:STN-50-528 Palo Verde Vuc'ear Station, Unit 1, Ar'zona Publi 05000528 AUTH. NAME AUTHOR AFF LZATZON SHRZVER,Z.D. Arizona Nuclear Power Project (formerly Arizona Public Serv HAYNES,J.G. Arizona Nuclear Power Project (formerly Arizona Public Serv RECZP.NAME RECIPIENT AFFILIATION

SUBJECT:

LER. 88-016-01:on 880514,reactor trip following earlier than anticipated criticality.

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DISTRIBUTION CODE: ZE22D COPIES RECEIVED:LTR ENCL SIZE:

TITLE: 50.73 Licensee Event Report (LER), Incident Rpt, etc.

S NOTES:Standardized plant. 05000528 RECIPIENT COPIES RECIPIENT COPIES ID CODE/NAME LTTR ENCZ ID CODE/NAME LTTR ENCL PD5 LA 1 1 PD5 PD 1 1 LICITRA,E 1 1 DAVISgM 1 1 INTERNAL: ACRS MICHELSON 1 1 ACRS MOELLER 2 2 AEOD/DOA 1 1 AEOD/DSP/NAS 1 1 AEOD/DSP/ROAB,.

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NRC Form 355

~0 ~4 UA. NUCLEAR REOULATORY COMMISSION (54)3)

APPROVEO OM8 NO. 3)500104 LICENSEE EVENT REPORT ILERI KXPIRES: SISIISS FACILITY NAME III DOCKET NVMSER (2l PA E TITLE Iel Palo Verde Unit 1 o 5 o o o 52 8>orl 8 Reactor Trip Following Earl,ier Than, Anticipated Criticality EVENT DATE IS) LER NUMSKR 15) REPORT DATE I tl OTHER FACILI'TIES INVOLVED IS)

YEAR SEOVENTIAL PRS tl 5 vdlore OAY YEAR FACILITYNAMES DOCKET NVMSER(SI MONTH OAY YEAR NVMSEtt PAS HUMSElr MONTH N/A 0 5 0 0 0 0 5 1 4 8 88 8 0 1 6 0 1 0 7 07 88 N/A 0 5 0 0 0, THIS REPORT IS SUBMITTED PURSUANt T0 THE REOVIREMENTS OF 10 CFR (); IChrch onr or more of thr lollowlnpl (11)

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NAME TELEPHONE NVMSER AREA CODE Timothy 0. Shr iver, Compliance Manager 60 23 9 3- 25 21 COMPLEtE ONE LINE FOR EACH COMPONENT FAILURE OESCRISED IN THIS REPORT (13)

CAUSE SYSTEM COMPONENT MANUFAC MANUFAC. E PORT A 5 LE:5NHXkx'~.gg~~

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ASST R ACT tpecet.

At approximately 0335 HST on Hay 14, 1988, Palo Verde Unit 1 was in Mode 3 (HOT STANDBY) when a reactor trip occurred as the Control Element Assemblies (CEA's) (AA) were being inserted following an attempt to startup the reactor.

The trip occurred when conservative Radial Peaking Factors (RPF) were utilized by the Core Protection Calculator (CPC) (CPU) (JC) as the CEA's were being inserted. There were no other safety system responses (including ESF actuations) and none were necessary. The plant was immediately stabilized in Mode 3.

The CEA's were being inserted after criticality had been achieved earlier than calculated resulting in the CEA's being below the Power Dependent Insertion Limits of LCO 3. 1.3.6. The root cause of the criticality outside established guidelines has been determined to be non-conservative operator performance during the reactor startup. Errors in the information utilized:for calculating the Estimated Critical Condition (ECC) contributed to this event.

The corrective action to prevent recurrence will be to correct the errors in the information utilized for the ECC and improve the administrative controls ~ J for utilizing the ECC. Appropriate disciplinary action will be taken.

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

8807120562 $ 80707 POP. ADGCIE,"05000bPB NIIC rnrm 355 PNI.I

NRC Form 444A 1984 I

~0 ~0 V.S. NUCLEAR REOVLATORY COMMISSION LICENSEE EVENT REPORT (LERI TEXT CONTINUATION APPROVEO OMS NO 8150 0184 EXPIRES: 8/SIISS FACILITY NAME III DOCKET NVMSER ISI LER NVMSER ISI YEAR j~ SEOUENTIAL NUMOER POI A>> REVISION NUMOER Palo Verde Unit 1 o 5 o o o 5 2 8 8 8 01 6 01 02oF 1 8 TEXT IIF mare EFPCe io reeIRRNL Roe mRFo'orMI iYRC ForIII BASSA'oi 1171 This is a supplement to LER 1-88-016-00.

I. DESCRIPTION OF WHAT OCCURRED:

A. Initial Conditions:

On Hay 14, 1988, Palo Verde Unit 1 was in Hode 3 (HOT STANDBY) at normal operating temperature and pressure. A reactor startup was in progress following a trip from 91 percent power which had occurred approximately 38.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> earlier.

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

Event Classification:

Automatic actuation of the Reactor Protection System. Condition prohibited by the plant's Technical Specifications.

On May 14, 1988, Palo Verde Unit 1 was in Mode 3 (HOT STANDBY) conducting a reactor (AC) (RC) startup. During the reactor startup, the reactor achieved criticality prior to that calculated by the Estimated Critical Condition (ECC). As criticality was achieved below the Power Dependent Insertion Limits of Limiting Condition for Operation 3. 1.3.6, it was decided to insert Control Element Assemblies (CEA)(AA) to calculate a new ECC. As the CEA's were being inserted, a reactor trip occurred at approximately 0335 HST on May 14, 1988. The reactor had been shutdown for approximately 38.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> prior to the trip. The Estimated Critical Rod Position per the ECC was 90" withdrawn on Regulating (Reg) Group 4 with a boron concentration of 1033 ppm for a startup time of 0000 HST.

The startup began at approximately 0100 HST by withdrawing the Shutdown (SD) CEA's banks and the Part Length CEA's (PLCEAs). The operating crew (utility, licensed) completed withdrawal of the SD banks and the PLCEA's at approximately 0159 HST. Withdrawal of the Regulating Groups began at approximately 0304 HST.

The count rate, obtained from the Startup Channels (IG) (XI), was approximately 300 counts per second (cps) when Reg Group 1 was 0 inches withdrawn. The startup was conducted in accordance with 410P-1ZZ03, "Reactor Startup", with the regulating CEA's being withdrawn in 30 inch increments per step 4.3. 12. After each withdrawal increment, a pause was established to allow count rate/power level to stabilize. Additionally, the Shift Technical Advisor (STA) (utility, licensed) was recording count rate after each 30 inch withdrawal. This was started when Reg Group 1 was being withdrawn even though the procedure only requires that power level be recorded and plotted with each 30 inch withdrawal after reaching 60 inches withdrawn on Reg Group 3 and thereafter.

NRC IORM 4444 IS SS>

I l

NRC Form 398A 1983 I

~0 U.S, NUCLEAR REOULATORY COMMISSION LICENSEE EVENT REPORT (LER) TEXT CONTINUATION APPROVEO OMS NO 3150 010a EXPIRES: 8/31/88 FACILITY NAME 111 OOCKET NUMSER IEI LER NUMSER (81 YEAR,.'<<, saauENT>AL,:.rl

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mav>s>0>r em >ruMeem Palo Verde Unit TEXT ///rrroro amoco/a n///rior/ rrso 1

aR/>I'>>e////IC For>sr JSSA'a/11TI o so oo 528 8 8. 01 6 01 03oF 1 8 The Primary Operator (utility, licensed) complied with section 4.3.12 of the procedure and withdrew Reg Groups I and 2 in 30 inch increments. When Reg Group 3 was withdrawn to 30 inches, the Primary Operator (utility, licensed) questioned the STA concerning count rate and was told that it had stabilized (the STA noted that the count rate was approximately 1277 cps). Count rate was noted to have doubled twice since beginning the withdrawal of Reg Group CEA's.

Since criticality was imminent, the Control Room Supervisor (CRS)

(utility, licensed) checked the Power Dependent Insertion Limits (PDILs) of Specification 3. 1.3.6. Technical Specification LCO

3. 1.3.6 specified that in order to enter Mode 2 (STARTUP), the CEAs in Reg Group 3 must be at least 60 inches withdrawn. With the count rate stable at approximately 1277 cps, the Primary Operator pulled Reg Group 3 to 45 inches withdrawn. While the CEA's were being withdrawn to 45 inches, the startup channels (IG) were deenergized in accordance with the procedure at approximately 2000cps. Power level was then monitored on the log power channels (IG) after observing proper overlap on the startup channel and log power channel.

Upon reaching 45 inches withdrawn on Reg Group 3, the startup rate was still not definitely positive and power level had stabilized.

The Primary Operator therefore commenced pulling Reg Group 3 to 60 inches withdrawn. The CEA withdrawal from 45 inches to 60 inches was made in three steps taking approximately one (I) minute to complete.

After the 15 inch withdrawal, the CRS concluded that the reactor was slightly supercritical and, hence, the critical CEA position was between 45 inches and 60 inches. (Note: The measure of criticality is actually based on the indication of a positive startup rate and an increasing power level without CEA motion. Thus, the reactor is actually brought to a supercritical condition.)

The CRS directed the Primary Operator not to allow power to exceed IE-03 percent. The Primary Operator initiated CEA insertions to stabilize power at less than IE-03 percent power. The CRS then conferred with the Shift Supervisor on what action to take. They concurred that it would be inappropriate to be critical while not meeting the PDIL requirements. They decided to insert Reg Group 3 to 0 inches withdrawn 'and investigate the deviations from the ECC. The direction to insert Reg Group 3 to 0 inches was given to the Primary Operator who then complied. It should be noted that Reg Group 3 was 60 inches withdrawn for approximately 2 minutes, 39 seconds.

When the CEA's reached approximately 25 inches withdrawn, an auxiliary trip was generated by Core Protection Calculators (CPC)

(CPU) (JC) Channels B and C on high Radial Peaking Factors. The Reactor Trip Switchgear (SWGR) operated as designed, and CPC channels RA" and RDU tripped as expected.

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I' NRC F4<M 34SA (94(3I

~0 U S NUCLEAR REOULATORY COMMISSION LICENSEE EVENT REPORT ILER) TEXT CONTINUATION APPROVEO OMS NO 3(SOW(04 EXPIRES; SI31/SS FACILITY NAME (II COCKET NUMSER (31 LER NUMSER (SI PACE I3)

YEAR SCQUCNTIAL .i~@ ACVISlON NUMSCN NUMSCA Pal o Verde Uni TEXT IIP NYNP N>>CC l4 NgoSed, NW ~t 1 HRC FONN 3SEASI ((TI o

The plant was immediately stabilized in Node s o o o 01 6 0

3. The event'as 1 0 4 oF 1 8 diagnosed by the Assistant Shift Supervisor (utility, licensed) as an uncomplicated Reactor trip and performance of the appropriate procedure was initiated.

The following information concerns the investigation into the cause of the trip.

The CPC trip buffers are not reset until the critical rod height data is taken, as stated in ANPP procedures. The CPC's cannot be reset unless Reg Group 3 is withdrawn sufficiently to reduce the integrated one-pin peak below the auxiliary trip setpoint (at this time, that position, was approximately,27 inches withdrawn). Additionally, 410P-1ZZ03 calls for the CPC reset when Group-3 is 97 inches withdrawn; this accounts for possibly higher peaks at other conditions. This resulted in a loss of actual trip data from the CPC's which would have verified the presence of the auxiliary trip.

Using the CPC Simulator, it was later verified that at less than 30 inches withdrawn on Reg Group 3, an auxiliary trip was correctly generated by the CPC's due to high Radial Peaking Factors. Even though the actual trip buffers for the event were unavailable, the re-creation of the event using the CPC Simulator verified that this was the cause of the reactor trip.

The reactor was subcritical at the time of the trip. No Engineered Safety Features (ESF) actuations were received or required. The Emergency Plan was not initiated and no emergency classification was made.

During ANPP's Post Trip Review evaluation, it was determined that the reactor had gone critical between 50 and 55 inches withdrawn on Group

3. Based upon criticality being achieved below 60 inches withdrawn, Unit 1 operated in a condition prohibited by Technical Specification 3.0.4 in that Mode 2 (STARTUP) was entered without meeting the conditions of LCO 3. 1.3.6.

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

Not applicable - no structures, systems, or components were inoperable at the start of the event which contributed to the event.

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

Not applicable - no component or system failures occurred.

4AC 40AM 344A I9 STi

NRC Form 348A

~o V.S. NUCLEAR REGULATORY COMMISSION 1983 I LICENSEE EVENT REPORT (LER) TEXT CONTINUATION APPROVEO OMS NO 3150&104 EXPIRES I 8/31/88 FACILITY NAME III OOCRET NUMSER IEI LER NVMSER IS)

YEA/I 8~% SEQUENT/AL .err rISVrsrON NvMos/r '-io NUMoso Palo Verde Unit TEXT //F rrroro sooco /4 orRrsorS E.

o>> ~ 1 Failure

/Y/TC Form 3/SSA'4/

mode, mechanism, II3) o and s o o effect of o 52 each 888 failed 01 6 component, 01 if 0 5oF 1 8 known:

Not applicable - no component failures occurred.

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

Not applicable - no component failures occurred.

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

Not applicable - no failures occurred which rendered a train of a safety system inoperable.

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

There were no component or system failures involved. The errors discussed in Section I below were identified during the post trip review process conducted by ANPP.

Cause of event:

The cause of the reactor trip was an Auxiliary Trip generated by the CPC's. The Auxiliary Trip resulted from conservatively high Radial Peaking Factors being generated as Regulating Group 3 CEA's were being inserted below 30 inches. In general, the conservatively high Radial Peaking Factors may result in a reactor trip when Group 3 is less than 95 inches withdrawn and the CPC's are not bypassed.

The cause of the condition prohibited by the plant's Technical Specifications wherein the reactor achieved criticality below the limits of LCO 3. 1.3.6 has been determined to be operator performance which was considered to be less conservative than appropriate for the situation during the reactor startup. It was determined that the control room personnel (utility, licensed) did not act with the desired conservatism in performing the approach to criticality based upon the information available at the time. During the approach to criticality, the control room personnel correctly performed and followed procedures and responded to alarms and permissives to bypass High Log Power trips. However, ANPP Management considers that the degree of conservatism utilized based upon indications of early criticality were not in accordance with management expectations and are considered to be cognitive personnel errors on the part of control room supervision (utility, licensed). As a result of this concern, ANPP performed a Control Room Staff Evaluation. The results of this evaluation are provided in Section V. There were no unusual NRC lOAM 3444 18 83r

~0 NRC SPIIA ESSA 1888 I

~0 ~o U.S. NUCLEAR REQULATORY COMMISSION LICENSEE EVENT REPORT (LER) TEXT CONTINUATION APPROVEO OMS NO EIEOWIC4 EXPIRES: 8/31188 DOCKET NUMSER (11 LER NUMSER (8) ~ ACE IS)

SEQUENTIAL AEVISIQN NUMSEA NUMSSA Palo Verde Unit TEXT IllRAMP soscs is ISPIAnd. Uss 1 o s o o o 5 2 8 8 8 0 1 601 06 OF 1 8 aAthiaW JVitC SPAA CSEA'sl I Ill characteristics of the work location which contributed to this event.

Contributing to the non-conservatism exhibited by the control room personnel, some of the information being utilized by the control room personnel was determined to be incorrect and/or inadequate. The ECC being utilized. by control room personnel contained inaccuracies which resulted from: (1) an inaccuracy in the computer program which calculates transient xenon level and (2) a startup procedure which allowed a 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> deviation from the projected startup time (At the time of the approach to criticality, approximately 3.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> had elapsed from the projected startup time. During this time period Xenon decay caused a positive reactivity change). The boronometer (XI) being utilized for determining boron levels in the Reactor Coolant System (RCS) (AB) may not have provided accurate indication of boron concentration (This issue is being evaluated by engineering and appears to be due to a non-linear response to variations in boron concentration). The information and controls available for use by control room personnel in evaluating the conditions present during the approach to cri.ticality were determined to be inadequate. That is, based upon the fact that the Core Data Book did not contain integrated CEA worth curves for Group 3 below 60 inches, an inverse count ratio plot (1/M plot) was not required by procedure to be started until Group 3 reached 60 inches withdrawn.

J. Safety System Response:

Reactor Protection System Actuation occurred at approximately 0335 MST on May 14, 1988.

There were no other safety system responses (including ESF actuations) arid none were necessary.

K. Failed Component Information:

Not applicable - there were no failed components.

II. ASSESSMENT OF THE SAFETY CONSE(UENCES AND IMPLICATIONS OF THIS EVENT:

There were no safety consequences or implications resulting from this event. As described above, the reactor tripped as designed and all safety responses necessary to place the plant in a stable condition functioned properly.

The criticality earlier than calculated in the ECC had no adverse safety consequences or implications. As described above, Unit 1 entered Mode 2 with the CEA's below the transient PDIL limit of Specification 3. 1.3.6.

Operation in this condition is permitted for up to two (2) hours pursuant to ACTION Ra" of LCO 3. 1.3.6. .The CEA's were below the PDIL limit for NAC IIIAM ESSA I8 SSI

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NRC PO<lA ESSA 19481

~o V.S. NUCLEAR REGULATORY COMMISSION LICENSEE EVENT REPORT ILER) TEXT CONTINUATION APPROVED OM8 NO 8150-GIOd EXPIRES: 8/Sl/88 DOCKET NUMEER lll LER NUMEER ISI PAGE ISI YEAR rv@ SEGVCHTIAL NVM SA AEVISION HVM TA Palo Verde Unit TEXT //P /Ad/P SAP CP id /Pdvdd//. PAP 1

/ATA/AP/HRC FP/TA 3////AT/ (Ill o s o o o 52 888 0 1 601 07 or-1 8 less than 10 minutes. =-It should be noted that the PDIL 'limits of Specification 3. 1.3.6 are established to ensure that an adequate shutdown margin is maintained and at the same time ensure that the potential effects of a CEA ejection accident are limited to acceptable levels. The function of the shutdown margin requirements is to ensure that the reactor remains subcritical following a design basis accident or anticipated operational occurrence. Shutdown margin requirements vary throughout the core life as a function of fuel depletion and reactor coolant system (RCS) cold leg temperature. The most restrictive condition occurs at the end of core life, with cold leg temperature at no-load operating temperature, and is associated with a postulated steam line break accident, and the resulting uncontrolled RCS cooldown. In the analysis of this accident, the specified shutdown margin is required to control the reactivity transient and ensure that the fuel performance and offsite dose criteria are satisfied. An analysis of the conditions present during the event has determined that the boron concentration was approximately 120 parts per million greater than necessary to meet shutdown margin requirements.

I I I. CORRECTIVE ACTIONS:

A. Immediate:

When control room personnel (utility, licensed) noted that criticality had been achieved earlier than calculated in the ECC, appropriate actions were taken to shutdown the reactor and place in a safe condition by inserting Group 3 to zero inches until the it problems with the ECC could be investigated.

As described above, the reactor trip occurred as the CEA's were being inserted below approximately 25 inches withdrawn. Following the trip, control room personnel (utility, licensed) took the appropriate action to ensure that the plant was in a safe condition.

B. Action to Prevent Recurrence:

Appropriate procedure precautions have been implemented to ensure that control room personnel are aware that reactor trips may occur Regulating Group 3 CEA's are less than 95 inches withdrawn and the if CPC's are not bypassed.

Concerning the cognitive personnel errors described in Section I. I wherein non-conservative operator performance was involved, appropriate disciplinary action and/or counseling will be taken.

Concerning the error in the ECC, the following actions are being taken:

AAC lOAM Sddd I9 8) l

NRC Form 388A I9 83 I

~0 Oo U.S. NUCLEAR REGULATORY COMMISSION LICENSEE EVENT REPORT {LER) TEXT CONTINUATION APPROVEO OMS NO 3/50 GIGA EXPIRES: 8/31/88 FACILITY NAME Ill OOCKET NUMSER 12l LER NUMEER IEI YEAR >PE

>>'pr SEGUENT/AL rr.:9 REV/SIO/r r/UMOE/r OPr NUMOFA Palo Verde Unit TExT ////rro/o N/oco /I oorrrooo, rroo 1 o s o o o 528 88 016-01 08oF 1 8

///I/ro/p//Tc Form JREASl 113)

'dditional controls concerning the time allowance'etween the time the ECC is calculated and the actual approach to criticality will be developed.

'he computer program which calculates transient xenon levels has been modified.

'CS boron samples will be utilized For plant startup in lieu of boronometer readings until the instrumentation is verified to be accurate for all plant conditions.

'nformation and direction for starting inverse count ratio plots earlier in the startup process will be developed.

'n engineeringwillanalysis on the existing ECC calculation methodology be performed. Based upon this analysis, appropriate controls or changes will be delineated.

Concerning the information and methodology for starting up the reactor, the following corrective actions are being taken:

'he integrated CEA worth curves below 60 inches have been included in the Core Data Book.

'he reactor startup procedure will be revised as appropriate to include the information contained in the Core Data Book.

' reactor engineer (utility, non-licensed) will be required to be in the control room (NA) during reactor startups until the appropriate administrative changes are made.

As a result of the Control Room Staff Evaluation, the following corrective actions will be taken:

' review of the Control Room communications during this event will be conducted and guidance on declaring criticality will be promulgated.

'anagement will issue a letter reminding all plant personnel to adopt a conservative approach when conditions or indications are other than expected.

' Human Performance Evaluation System evaluation will be performed by the STA Group.

IV. PREVIOUS SIMILAR EVENTS:

There have been no previous similar events reported pursuant to 10CFR50.73 involving a reactor trip following a criticality earlier than anticipated by the ECC. However, a similar trip occurred as reported in NAC r OAM Seeo

/983/

NRC Forrrr SSEA 1949 I U,S. NUCLEAR REOULATORY COMMISSION LICENSEE EVENT REPORT (LER) TEXT CONTINUATION APPROVEO OMS NO 3150 OIOA EXPIRES: 9/SIISS FACILITY NAME III OOCKET NUMSER IEI LER NUMSER LE) ~ AOE ISI EEOUENTrAL 4UMSEII

'lr AEYrErON Prr'UMFEA Palo'Verde Unit TEXT Iltmrrre NrFce lF mrRrnd, rrm Unit

~ 1 1

iYRC Fnrm SSCASI 1 1tI LER o s o o o 6 28 88 016 88-011-00 when overly conse'rvative Radial Peaking Factors 0 1 09oF 1 8 (RPF) utilized by the CPC resulted in a reactor trip. As discussed in LER 88-011-00, the conservative RPF values are part of the original design of the CPC software. ANPP is currently evaluating the feasibility of modifying the existing software.

V. ADDITIONAL INFORHATION A. The following information was developed as a result of a Control Room Staff evaluation conducted by ANPP:

SHIFT SUPERVISOR (Utility, Licensed)

The Shift Supervisor (SS) was in the "horseshoe" area. It was his intention to maintain a broad perspective on overall plant response and therefore was not directly involved with the specifics of the criticality. When he was consulted about the PDILs and the critical rod position by the CRS, he concurred with the CRS's recommendation that the Group 3 CEA's be reinserted to 0 inches. ANPP believes the Shift Supervisor should have been more involved in this evolution.

CONTROL ROOM SUPERVISOR/ASSISTANT SHIFT SUPERVISOR (Utility, Licensed)

The CRS was directing the Reactor Startup activities. The CRS was using the correct procedure for the evolution. The Startup was proceeding in a controlled and "unhurried" manner. The CRS had discussed the potential for an "early" criticality due to Xenon decay with his Reactor Operators. The Primary Operator indicated he understood the discussion.

When Group 3 was at 30 inches, it was apparent that, based on the count rate information, the reactor would go critical "...very close to 60 inches...". Due to the apparent large difference between the suspected early criticality of approximately 60 inches on Group 3 and the ECC of 90 inches" on Group 4, the CRS should have taken a more conservative approach and reevaluated the ECC prior to continuing the startup. When Reg Group 3 was at 60 inches, the CRS recognized that the reactor had gone critical during the last rod withdrawal. He then directed the Primary Operator to maintain reactor power less than 1E-03 percent of rated thermal power while he consulted with the SS. At this time, the CRS was primarily concerned with Technical Specification limits on CEA position (PDILs).

It was the understanding of the CRS that the Reactor Operator actually pulling CEAs is the one who actually "calls" criticality.

The CRS, upon recognizing that the reactor was critical, asked the Primary Operator, "What are the indications of criticality?". This was done in order to prompt the operator to "call" criticality. In 4rrc ilrAM ldl<

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~e NRC form 344A IQ 83)

~0 U.S. NUCLEAR REOULATORY COMMISSION LICENSEE EVENT REPORT ILER) TEXT CONTINUATION APPROVEO OM8 NO 319)M)84 EXPIRES: 8/31/88 PACILITY NAME III OOCKET NUMEER )1)

LER NUMEER )4)

YEAR g>' NTIAL g+p) hSVrSION A, NVM 4 fr ~ .m NVMOoh Palo Verde Unit 1 o s o o o 52'8 8 8 0 1 '6 0 1 1 0 OF 1 8 TEXT /// moro /oooo /4 ooO/4'rE ooo oR/Orion/Ar/IC Forrh 3/)/)A'4/ I)7)...

this case the CRS should have been more direct with his communications to the Primary Operator with regard to what information he wanted with respect to the condition of the reactor, i.e., by asking "Is the reactor critical?". It should also be recognized that. there are no formal guidelines regarding who on the Control Room staff should or must "declare criticality." ANPP Hanagement believes that the CRS should have directed the evolution be stopped when it became apparent that the criticality could be achieved earlier than anticipated.

Following the Reactor Trip, the CRS directed the Operators to maintain their safety functions and the plant was stabilized in Hode 3.

NO III - PRIHARY OPERATOR (Utility, Licensed)

The Primary Operator was pulling the CEA's under the direction of the CRS. Ke observed the power level increase above the point where the Log Power Channel could be bypassed and the CPC channels become "active". Based on the interview with the Primary Operator, he believed the reactor to be critical at approximately 60 inches withdrawn on Reg Group 3. Actions were taken by the Primary Operator to insert the CEA's in order to maintain the reactor at less than lE-03 percent power at the direction of the CRS. Before the reactor was stabilized and the critical point data could be taken, it was decided to reinsert Group 3. Therefore, criticality was not formally stated nor entered in the Control Room logs. Criticality should have been entered in the Control Room logs as a late entry.

The indications present with Group 3 at 30 inches indicated that subsequent withdrawals would be very near, The Primary Operator should have shown more concern with these if not at, criticality.

indications, and at least questioned, the CRS. A more conservative action would have been to recalculate the ECC prior to continuing the Startup. The Primary Operator should have recognized indications of criticality prior to being "prompted" by the CRS.

ANPP believes the Primary Operator should have stopped the evolution when it became apparent that criticality would be achieved earlier than anticipated.

NO III - SECONDARY OPERATOR (Utility, Licensed)

The Secondary Operator was performing the Hain Turbine Warmup in preparation for secondary plant startup.

NO III - CONTROL ROOM (Utility, Licensed)

Was not directly involved in startup.

NNC IONM )44A I4 4 3 >

NRC Po<m 3SSA I 9831 V.S. NUCLEAR REQVLATORY COMMISSION LICENSEE EVENT REPORT (LER) TEXT CONTINUATION APPROI/EO OM8 NO 3150 OIOO EXPIRES; 8/31/88 OOCKET NUMEER LTI LER NUMSER (Sl ~ AOE 13I 1"K SEOUENT/AL ar'8 OEVR/ON NUM E/I NS NVMOOO Palo Verde Unit TEXT //I mo/o N>>co 4 POP//od. o>> ~ 1 A//TC fo/m 3RLI 9/

SHIFT TECHNICAL ADVISOR 1 ill o s o o o (Utility, Lice'nsed) 5 2 8 88 01 6 01 11pF1 8 The Shift Technical Advisor (STA) was observing the progress of the startup and recorded count-rates periodically during withdrawal of the Shutdown groups and Regulating groups. He indicated that the count rates had doubled twice during the course of the rod withdrawal. The STA should have been more aggressive in providing this information to the Control Room staff. This would have provided additional indication to the Control Room on their nearness to criticality. Since the CEA worth curves are not available in the Core Data Book, it was not possible to perform a I/H plot.

ANPP Management believes that the STA should -have been more involved in monitoring the startup activities and providing direct communication that 'the reactor was nearing criticality. He should have recommended to the SS that the evolution be stopped when it became apparent that criticality could be achieved earlier than anticipated.

B. Following the event, it was determined that the information provided in the 4-hour call made via the Emergency Notification System (ENS) was not accurate. During the ENS notification, the reactor trip occurred as the CEA's were being inserted in order it was discussed that to calculate a new ECC, and the CEA's were being inserted since the reactor was approaching criticality prior to the ECC. However, was not discussed that the reactor had achieved earlier criticality it and the CEA's were also being inserted due to concerns about meeting PDIL limitations.

ANPP believes that the criticality and PDIL concerns should have been discussed in the initial report.

Investigation into this'spect of the event is continuing and will address whether additional reporting requirements were applicable.

Based upon the results of the investigation, corrective actions will be implemented as appropriate. However, as an immediate corrective action additional administrative controls will be implemented to provide more explicit directions for NRC notifications.

The results of this investigation are provided in Section VI.D.

C. Exact discussions of the event were impacted by information available in the various logs. ANPP will evaluate this aspect and determine changes are required to enhance the current log keeping techniques.

if The results of this evaluation are discussed in Section VI.A and VI.C.

D. As previously discussed, additional evaluations/investigations are being conducted as a result of this event in both the reporting/notification aspects and in the area of Human Performance NOC //IOM 34OO 19 8) i

~0 NRC Form SCCA yO I9 SSI V,S. NUCLEAR REOULATORY COMMISSION LICENSEE EVENT REPORT ILERI TEXT CONTINUATION APPROVEO OMS NO SISO 0194 EXPIRESI 8/SI/SS FACILITY NAME III OOCKET NUMSER 111 LER NVMSER ISI PACE IS)

YEAR >YR> SCovCNT>AL s'sp >>Cy>s>QN NVMC4/l:9 A NVMFC>>

Palo Verde Unit 1 o s o o o 5 2 8 8 8 01'6 01 12oF 1 8 TExT //T mort sosco /4 oco/'r>N/. Fss /rooo>s/H/Ic Fsy>T> sr/I/4'4/ I ITI Evaluation System. Based upon the results of these evaluations a supplement to this report will be issued.

The results of the Human Performance Evaluation System review are contained in Section VI.A.

VI. SUPPLEMENTAL INFORMATION The information in this section is provided as a result of ANPP's on-going investigation into the circumstances surrounding the event described in this LER.

A. An evaluation was conducted to,address those errors identified during the approach to criticality on May 14; 1988. The evaluation was performed using the INPO Human Performance Evaluation System (HPES) which was developed specifically for addressing HUMAN PERFORMANCE problems at Nuclear Power Plants. During the HPES evaluation, the problems identified during the Post Trip Review (PTR) process as discussed in Section I. I were. analyzed. The HPES is intended to identify the "causal factors" affecting the root causes discussed in Section I. I. Also during the HPES evaluation, additional problems were identified and analyzed.

The following provides a summary of the HPES concerns evaluated and their respective corrective actions.

1. Criticality was not declared by the Control Room staff when it occurred.

Corrective Actions:

a. Job performance standards and requirements for the Control Room staff delineating responsibility for the declaration of reactor status will be developed and implemented.

b. Procedures, instructions, and programs will be revised to incorporate the specific requirements or responsibility for declaration of criticality.

c. Simulator training (initial and requalification) and on-the-job training will address whose responsibility it is for the declaration of reactor status based upon the policies and procedures developed.

2. The Control Room Supervisor (CRS) did not terminate the reactor startup even though count rate nearly doubled twice when Reg Group 3 CEA's were withdrawn from 0 to 30 inches and it became apparent that criticality would be achieved near the Power Dependent Insertion Limits (PDIL's).

N>IC >ON>s S444 19 SS>

V NRC Form SSSA I98SI

~ ~o V.S. NUCLEAR REOVLATORY COMMISSION LICENSEE EVENT REPORT (LER) TEXT CONTINUATION APPROVEO OMS NO SI50&ICO EXPIRES: 8/8 1 /88 FACILITY NAME III OOCKET NVMSER IXI LER NVMSER ISI ~ AOE ISI YEAR SFK ssQUENT/AL 'i Asyrslorr

".'/op NUMPTA ~

rrv>>eeo Palo Verde Unit TEXT IN more a>>ce /p /eee/reN, /roe 1

pdc/oor>>I HIIC Fe/m JSEAS/ II Tl o s o o o 52 888' 01 6 01 13oF1 8 Corrective Actions:

a. A policy which requires more formal communications between members of the Cohtrol Room staff will be developed and implemented. The policy will be incorporated into the Conduct of Shift Operations procedure and subsequently incorporated into simulator training.

b. The administrative procedures governing the STA role in Control Room operations will be reviewed and revised as necessary to ensure that the STA is more effectively utilized on shift.

c. Simulator training involving scenarios where there is a large error between the Estimated Critical Condition (ECC) and the actual critical condition will be provided.

3. The reactor was taken critical below the PDIL's.

(Note: A contributory factor in this concern is that the SS on duty was a relief crew SS filling in for the normal SS).

Corrective Actions:

a. Guidance will be established on the standardization and conduct of operations between crews to insure consistency between the on-shift and replacement crew members.

b. Simulator training involving scenarios where there is a large error between the ECC and the actual critical condition will be provided.

c. The operations crew supervision involved will be re-instructed regarding their responsibilities for unit operations including that they should be directly involved in critical evolutions by providing guidance and ensuring all aspects of the task are understood prior to task performance.

4. The ECC used for the startup was calculated for 0000 HST (approximately 3 hours and 25 minutes prior to the time criticality was achieved). An ECC for 0200 was calculated; however, it was not finalized nor was it utilized to establish a new boron concentration. The 0200 calculation was only used to predict the expected change due to xenon.

~ AC ///AM SOOA 19 SSr

~

N/IC form 348A (983 I

~0 U.S. NUCLEAA IIEOULATOAYCOMMISSION LICENSEE EVENT REPORT (LER) TEXT CONTINUATION APP/IOVEO OMS NO 3/50 OI04 EXPISES: 8/31/88 FACILITY NAME III OOCKET NUMSE/I IEI LE/I NUMSEA (8) 54QUENT/AL 4 4 V IS lO N Illr/M44rr rrr/Moe A Pal o Verde Uni t 1 o s o o o 5 2 8 8 8 0 1 6 0 1' 4 QF 1 8 TEXT lllmoro afoco /I 44//ror/. r/fo or//'/ro//Y/IC forrrr 38EAS/ I I TI Corrective Action:

The ECC and Reactor Startup procedures will be modified to include specific guidance on how close the projected time of criticality used for the ECC must be with respect to the actual time of criticality.

5. The Reactor Operator did not recognize the nearness of the reactor to criticality when Reg Group 3 was withdrawn at 30 inches and 45 inches.

Corrective Actions:

a. An evaluation of the RO involved in the startup to determine if he possesses sufficient practical knowledge skills in applying reactor theory to instrument indications being performed. Training will be provided as needed.

b. Simulator training will include non-ideal or off-normal startup scenarios (e.g., shortly after a reactor trip or when errors in boron or xenon concentration are present).

Simulator training will be conducted to ensure that operators can apply theory to plant operations by applying 1/H plots and other methods allowed by the procedure for determining critical CEA position or Boron concentration.

6. Rod Worth data for Reg Groups 1 and 2, and Group 3 below 60 inches were not included in the Core Data Book.

Corrective Actions:

a. The review process for changes related to core reloads will be upgraded.

b. The procedural controls for the Core Data Book will be reviewed and revised as appropriate to ensure that the data provided adequately meets the needs of the users.

c. The training and/or qualification requirements for the Engineering Evaluations Department Reactor Engineering staff will be upgraded to include an integrated knowledge of the effect of core reloads.

7. New core reload calculations included high radial peaking factors associated with Reg Group 3 CEA's. (Note: This concern is also related to the event described in LER 88-011-00 involving a reactor trip caused by conservative peaking factors.)

roC IOAM 3444

/9 83r

~0 0 NRC fons 388A

/9831 U.S< NUCLEAR REGULATORY COMMISSION LICENSEE EVENT REPORT ILER) TEXT CONTINUATION APPROVED OMB NO 3180-010O EXPIRES; 8/31/88 FACILITY NAME 111 DOCKET NUMBER (31 LER NUMBER ld)

YEAR o" Sdavonr/AL 8$ NUMOER

~o:

~

sdv/Sio/I

.> NUM ER Palo Verde Unit TEXT //f /so/o oooco /I /od///sNL voo ~ 1 A//IC Fons 3/IEA3/1131 Corrective Actions:

o s o o o 52 88 8 . 16 01 15oF 1 8

a. A more effective and productive interface will be established with Combustion Engineering (CE) concerning specific operating practices at Palo Verde.

b. The review process for changes related to core reloads will be reviewed and upgraded.

c. The training and/or qualification requirements for the Fuels Hanagement staff responsible for core reloads will be reviewed and upgraded as necessary.

d. Transient Data Acquisition System data will be transmitted to the Safety Analysis group.

8. The "Xenon" computer program used to calculate the reactivity due to xenon had large uncertainties due to incorrect coefficients.

Corrective Action:

The administrative control requirements for the xenon program (e.g. determining and verifying the correct xenon distribution coefficients) will be evaluated and upgraded as necessary. The administrative controls will be sufficient to ensure that the xenon program is accurate for: I) Determining xenon worth during transients, 2) Predicting criticality with xenon present, and 3) guantifying the effects of xenon on shutdown margin.

9. The Compliance representative (utility, non-licensed) did not notify the Compliance Hanager (utility, non-licensed) prior to making the 4-hour ENS notification.

Corrective Action:

A Department Instruction prescribing the requirements for ENS notifications be implemented and provided to Unit Operations Supervision and Compliance Engineers. As an interim measure, the contents of a 1986*letter requiring Hanagement notification of ENS calls was updated and disseminated to the Compliance Engineers..

10. The Shift Technical Advisor (STA) (utility, licensed) did not update his log concerning the events surrounding the approach to criticality until two days after the event.

nsc oosM Soos

/9 83i

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NRC POIIR 348A

~o V.S, NUCLEAR RECULATORY COMMISSION 18431 LICENSEE EVENT REPORT (LER) TEXT CONTINUATION APPROVEO OMB NO 3150 010E EXPIRES; 8/31/88 fACILITYNAME (I I OOCKET NVMSER (21 LER NUMBER ISI SEOUENr/AL REVISION NUMSEA NUMPEA Palo Verde Uriit 1 o s o o o 52 888 0 1 6 01 16oi 1 8 TEXT //f /RIPP EPPES /E /PEMPPd. VPP ~ //PP'/M////IC A/I/I388A3 I I 1 71, Corrective Actions:

a. The STA work schedule is being evaluated to determine present schedule is the most effective use of the STA if the resource;

b. The need for accuracy and completeness in the areas of shift turnover and log taking will be reaffirmed.

c. The STA involved in this event has been counseled to assure that he has a proper understanding of the requirements for completion of all on shift tasks, particularly logkeeping.

'l. The Unit Log and Control Room Log did not accurately reflect entry into operational modes or entry into Technical Specification ACTION statements. For example, criticality was not recorded in the Unit or Control Room Logs and entry into ACTION "aR of LCO 3. 1.3.6 was not recorded in the Unit Log. This information should have been entered into the logs as "late entries".'orrective Actions:

a. Conduct of Shift Operations will be revised to require the logging of significant actions occurring during an abnormal event as late entries time they happened.

if those actions were not logged at the

b. Conduct a

of Shift Operations will be evaluated to determine revision is necessary regarding the need for an on-shift if supervisory review of the Control Room Log prior to the shift turnover to the oncoming crew to assure completeness of the logs.

c. The Unit Operations Management will periodically review Control Room and Unit Logs to assure that the logs meet the standards established in the Conduct of Shift Operations.

d. The logkeeping requirements during Simulator requalification training will be evaluated.

12. The Event Notification Worksheet (for making ENS notifications) did not include the fact that the reactor was critical. This resulted in incomplete information being communicated to the NRC.

NRC /OAM SEER IS 83/

NRC Poem ESSA

/983 I U.S. NUCLEAR REGULATORY COMMISSION LICENSEE EVENT REPORT (LER) TEXT CONTINUATION APPROVEO OMS NO SI SOWIOE EXPIRES; 8/3 I /88 OOCKET NUMSER (SI LER NUMSER Lel VS*A jgo SEOI/ENTIAL NI/MOEA $OX0> REVISION

~

NUM EA Palo Verde Unit 1 o s 0 0 o 5 2 8 8 8 01'6 01 1 7 OF 1 8 TEXT lll/I>>/o u>>co /d /odowAE vdo /IT/o'ono//VRC Porm JSSA 9/ l ill Corrective Action:

A program or procedure will be developed to ensure that accurate and adequate information is obtained by the Compliance representative for NRC notification.

B. Further investigation has been conducted to determine the length of time it took to withdraw Reg Group 3 to 60 inches withdrawn. Based upon ANPP's investigation, the following Reg Group 3 to 60 inches withdrawal information is provided. The information provided is ANPP's best approximation of the withdrawal sequence and is based upon operators statements and times for starting/stopping CfA motion provided by the Plant Computer (ID)(CPU).

Reg Group Time Reg Group 3 Position hrs. min sec ~3Acti v i t ~Arox

• 03 23 37 Withdrawal St'art 0 03 24 32 Withdrawal Stop 30 03 26 55 Withdrawal Start 30 03 27 21 Withdrawal Stop 45 03 30 03 Withdrawal Start 45 03 30 25 Withdrawal Stop 52 03 30 26 Withdrawal Start 52 03 30 44 Withdrawal Stop 58 03 30 48 Withdrawal Start 58 03 30 50 Withdrawal Stop 60

• CfA positions are based upon'perators statements for stops at 30, 45, and 60 inches.

C. ANPP's investigation into this event included an evaluation of the logs that have been maintained during reactor startups. This investigation included an evaluation of the administrative controls concerning logkeeping, an evaluation of the scope and adequacy of the logs that were maintained during this event, and an evaluation of the logs that were maintained during previous reactor startups.

An evaluation of Control Room logs including the logs maintained during this event identified that the logkeeping practices were inconsistent, that is, the details included in the logs and the actions recorded appeared to be dependent upon the individual making the entry instead of following pre-established guidelines. A subsequent evaluation of the guidance provided in this area determined that additional controls were NAC I RAM Sddd Ie elI

NAC Form SSSA I9031

~ t LICENSEE EVENT REPORT ILER) TEXT CONTINUATION

~o U.S. NUCLEAR AEGULA'TORY COMMISSION APPAOVEO OM8 NO 3150M)04 EXPIRES: 8/3)/88 FACILITY NAME III OOCIIET NUMSER IT)

LER NUMSER (8) ~ AGE IS)

YEAR SEOVENTIAL <EV IS lG N NVM EA NVMSER Palo Verde Unit 1 o s o o o 52 88 8 0 1 6 0 1 1 8 oF1 8 TEXT IIImaP u>>cu 0 mewvvL uPP aA4tiaW H)IC Am4 JGEA'41 I )7) necessary to establish consistency in the information recorded in the logs. Based upon these findings, ANPP considers that the administrative controls providing guidance for proper logkeeping practices could be enhanced. Therefore, 40AC-9ZZ02 "Conduct of Shift Operations" will be revised to provide more prescriptive guidance for the information required to be entered. into the logs.

D. ANPP Licensing Department conducted an independent evaluation of the reportability aspects of this event. The evaluation was specifically conducted to determine if a one-hour notification was required pursuant to 10CFR50.72. As a result of this evaluation, it was determined that the four-hour notification conducted following the event was appropriate and that no one-hour notification was required.

E. A formal ANPP incident investigation program is being developed. This will include improvements to the Post Trip Review (PTR) process.

F. ANPP's continuing investigation into this event included an evaluation to determine if additional operator actions were required as a result of being critical below PDIL specifications. A clarification of Technical Specification Surveillance Requirement 4. 1. 1.2. l.b was necessary since it cross references Specification 3. 1.3.6 (PDIL requirements). The concern being that if the Control Element Assemblies (CEA's) are not within the Transient Insertion Limits of Technical Specification 3. 1.3.6, is immediate boration required in accordance with Technical Specification

3. 1. 1.2 ACTION AaA or is there a two-hour period to restore CEA's in accordance with Specification 3. 1.3.6 ACTION Ua".

ANPP has determined that operators have two hours to restore the CEA's to within the PDIL limits. Immediate boration pursuant to Technical Specification 3. 1. 1.2'ACTION AaA is not required unless the two hour ACTION of Specification 3.1.3.6 cannot be met.

~ AC 40AM )444 I9 8)i

Arizona Nuclear Power Project P.O. BOX 52034 ~ -

PHOENIX, ARIZONA85072-2034 192-00392-JGH/TDS/DAJ July 7, 1988 U. S. Nuclear Regulatory Commission NRC Document Control Desk Washington, D.C. 20555

Subject:

Palo Verde Nuclear Generating Station (PVNGS)

Unit 1 Docket No. STN 50-528 (License NPF-41)

Licensee Event Report 1-88-016-01 File: 88-020-404 Attached please find Supplement No. 1 to Licensee Event Report (LER) No.

88-016-00 prepared and submitted pursuant to the requirements of 10CFR 50.73(d). We are herewith forwarding a copy of this report to the Regional Administrator of the Region V Office.

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

Very trul yours, J. G. Haynes Vice President Nuclear Production JGH/TDS/DAJ/kj Attachment cc: E. E. Van Brunt, Jr. (all w/a)

J. B. Martin T. J. Polich E. A. Licitra A. C. Gehr INPO Records Center