ML17304A816

LER 88-014-00:on 881116,reactor Trip Occurred as Result of Low Steam Generator Water Level.Caused by Inadequate Feedwater Flow Due to Leakage of Packing.Procedural Changes Implemented & Evaluation Being performed.W/881214 Ltr

ML17304A816
Person / Time
Site:	Palo Verde
Issue date:	12/14/1988
From:	Haynes J, Shriver T ARIZONA PUBLIC SERVICE CO. (FORMERLY ARIZONA NUCLEAR
To:	NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM)
References
192-00437-JGH-T, 192-437-JGH-T, LER-88-014-01, LER-88-14-1, NUDOCS 8812210144
Download: ML17304A816 (20)

Text

ACCELERATED DISTRIBUTION DEMOYSTRATIQN SYSTEM 1

REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)

ACCESSION NBR:8812210144 DOC.DATE: 88/12/14 NOTARIZED: NO DOCKET N FACIL:STN-50-529 Palo Verde Nuclear Station, Unit 2, Arizona Publi 05000529 AUTH. NAME AUTHOR AFFILIATION SHRIVER,T.D. Arizona Nuclear Power Project, (formerly Arizona Public Serv HAYNES,J.G. Arizona Nuclear Power Project (formerly Arizona Public Serv RECIP.NAME RECIPIENT AFFILIATION

SUBJECT:

LER 88-014-00:on 881116,reactor generator level.

trip due to low steam, W/8 DISTRIBUTION CODE: IE22D COPIES RECEIVED:LTR ENCL SIZE:

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

NOTES:Standardized plant. 05000529 /

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 ARM/DCTS/DAB .1 1 DEDRO 1 1 NRR/DEST/ADS 7E 1 0 NRR/DEST/CEB 8H 1 1 NRR/DEST/ESB 8D 1 1 NRR/DEST/ICSB 1 1 NRR/DEST/MEB 9H 1 1 9H7'RR/DEST/MTB 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/QAB 10 1 1 NRR/DOEA/EAB 11 1 1 NRR/DREP/RAB 10 1 1 NRR/DREP/RPB 10 2 2 NRR/DRISJS 9A 1 1 NUDOCS-ABSTRACT 1 1 BEG Pl-IZ~ 0 1 1 RES/DSIR/EIB 1 1 RES/DSR/PRAB 1 1 RGN5 FILE Ol 1 1 EXTERNAL: EG&G WILLIAMS,S 4 4 FORD BLDG HOY,A 1 ~ 1 H ST LOBBY WARD 1 ~ 1 . LPDR "1 1 NRC PDR 1 1 NSIC HARRIS,J 1 1 NSIC MAYS,G 1 1 S

. NOTES:

NOTE TO ALL "RIDS" RECZPIENIS:

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NRC Form 255 UA. NVCLEAR REOULATOAY COMMISSION (543P Q APPROVEO OMS NO. 21504101 LICENSEE EVENT REPORT HLER) EXPIRES: SISIIES FACILI'TY NAME (I) DOCKET NVMSER (2) PA E Pa 1 o TITLE ICI Verde Uni t 2 0 5 0 0 0 52 9>or-0 8 Reactor Tri Oue to Low Steam Generator Level EVENT DATE (5) LER NUMSER LS) REPORT DATE (7I OTHER FACILITIES INVOLVED (5)

MONTH OAY YEAR SEOUENTIAL ...+c OAY FACILITY NAMES DOCKET NUMSERISI YEAR NUMEEII NA NUMOER MON'TH YEAR N/A 0 5 0 0 0 1116 88 8 8 014 00 1 21 4 8 8 N/A 0 5 0 0 0 THIS REPORT IS SUSMITTEO PURSUANT TO THE REOUIAEMENTS OF 10 CFR $ : IChrch onr ot mote Ol the lollowinII l1'll OPERATINO MODE IS) 20A02(bl 20A05(cl 50.724)(2) (IrI 72.71St)

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NAME TELEPHONE NUMSER AREA CODE Timoth O. Shriver, Compliance Manager 60,23 9 3 2 521

'g COMPLETE ONE LINE FOR EACH COMPONENT FAILURE OESCAISED IN THIS REPORT (12)

CAUSE SYSTEM COMPONENT MANUFAC.

TVAER EPOATASLE TO NPRDS 4H~+ CAlln E tlE COMPONENT MANUFAC TVRER EPORTASLE l TO NPADS SP A B I S V B 3 5 0 N RK@% ~4M SUPPLEMENTAL REPORT EXPECTED IICI MONTH DAY YEAR EXPECTED SUSMISSION DATE (15)

YES IIIyrt, Complete EXPECTED SUSMISSIOlY OATSI X NO AssTRAcT ILlmlt to leod tprcrt. I.r., rpptonlmetrty Illtrrn clncl>tprcr typnw/ttrn linrtl (15)

At approximately 0237 HST on November 16, 1988 Palo Verde Unit. 2 was in Mode 1 (POWER OPERATION) at approximately 10 percent power when a reactor trip occurred. Unit 2 was being shut down to identify and repair a reactor coolant system (RCS) leak, which was,within Technical Specification limits for continued operation, when the trip occurred as a result of low steam generator

.water level. The reactor trip was. uncomplicated and stable conditions were achieved at approximately 0247 HST terminating the event. There were no Engineered Safety Feature actuations and none were necessary.

The cause of the low steam generator water level was inadequate feedwater flow due to main feedwater pump speed being too slow for the existing plant conditions. The cause of the RCS leak was excess packing leakage as a result of a broken packing gland follower bolt on an instrument isolation valve.

As corrective action to prevent recurrence, procedural changes have been implemented, an evaluation of the feedwater pump control system is being performed, and programs for implementation of plant modifications will be evaluated and'revised where appropriate.

There have been no previous similar events. ~l 7~

SS12210144 SS1214 PDR ADGCK 05000529 8 PDC NIIC ratm SCE

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HRC Form 844A V.S. HVCLEAR REOVLATORY COMMISSION 19$ >l LICENSEE EVENT REPORT {LER) TEXT CONTINUATION APPROVEO OMS HO SISOWIO4 EXPIRES: 8/81/88 FACILITY NAME (11 DOCKET HVMSER Ql LER HVMSER IS) PACE ISI V EAR ~)OP. SEOVENTIAL REVISION NUM SR NVMSSR Palo Verde Unit TEXT ///more Sooce 14 2

reeoleK Foe ~ ///rmo/H/IC Fo/III JNA'4/117l o 5 o o o 52 98 801 4 00,02 OF 0 8 I. DESCRIPTION OF WHAT OCCURRED:

A. Initial Conditions:

On November 16, 1988, Palo Verde .Unit 2 was in Mode 1 (POWER OPERATION) performing a plant shutdown to investigate and repair a Reactor Coolant System (RCS)(AB) leak. Immediately prior to the reactor (RCT)(AC) trip discussed below,'eactor power was approximately 10 percent.

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

Event Classification: Automatic. actuation of the Reactor Protecti.on System (RPS)(JC).

On November 16, 1988 at approximately 0237 MST Palo Verde Unit 2 was in Mode 1 (POWER OPERATION) at approximately 10 percent power when a reactor trip occurred due to low steam generator (SG)(AB) water level. The low steam generator water level resulted from main feedwater pump (P)(SJ) speed being inadequate to supply feedwater to the steam generators during a power reduction. There were no engineered safety features (ESF)(JE) actuations and none were necessary. The event was properly diagnosed as an Uncomplicated Reactor Trip. At approximately 0247 MST both steam generator levels had been raised to above their trip setpoints and stable conditions were achieved. The event lasted approximately 10 minutes.

Prior to the reactor trip, Palo Verde Unit 2 was being shut down to investigate and repair the cause of 'an unidentified leak. Initial indications of a leak became a'pparent on November 7, 1988 when day-shift operations personnel (utility, licensed and non-licensed) noted an unexplained increase in the reactor cavity sump (WK) level over the previous few days. Initial estimates of the leakage into the sump were approximately 0.5 gallon per hour (0.008 gallon per 'h

.minute).

It should be noted that Technical Specification 3.4.5.2 allows 1.0 gallon per minute (gpm) unidentified Reactor Coolant System (RCS) leakage. Therefore, assuming that the leakage into the reactor cavity sump was unidentified reactor coolant system leakage, continued plant operation was allowed.

Investigation was initiated to identify the cause of the leakage.

On November 8, 1988, troubleshooting was performed on the reactor cavity sump level indicator (LI)(WK) and it was determined that the level indicator was operating properly. Furthermore, no significant trends or changes were noted in containment (NH)

NRC /ORM 144A 19 SSI

0 NRC for>> 844A U.S. NUCLEAR REGULATORY COMMISSION I844I LICENSEE EVENT REPORT {LER) TEXT CONTINUATION APPROVEO OMS NO. SI50&104 EXPIRES: 8/31/48 fACILII'YNAME III OOCKET NUMSER ICI LER NUMSER IS) ~ AGE IS)

JCGVCNZ/AL ACV/CION NVMCCA ~+ NVMCCA Palo Verde Unit 2 o 5 o o o 5 2 9 8 8 0 1 4 0 0 0 3 QF 0 8 TEXT //l /Atuf cfecu /I /P//cwed, u>> //tu//M/H/IC fPNA JSSA'r/ I IT) atmosphere samples for iodine, noble gas, or particulate

.radioactivity. Therefore, preparations were made to enter the containment to visually search for the cause of the leakage.

The initial containment entry to investigate the cause of the leakage was made on November 9, 1988. Initial attempts were unsuccessful in identifying the source of the leakage so the investigation continued. On November 15, 1988, a steam leak was discovered in the vicinity of the number 1 RCS hot leg. The exact source of the leakage was indeterminate due to the existing radiological conditions and presence of water vapor. Therefore, management prudently decided to shut down Unit 2 to facilitate further inspections and repairs.

At approximately 2000 .HST on November 15, 1988, a power reduction from 100 percent power was commenced. At approximately 15 percent power, the Feedwater Control System (FM)(JB) automatically redirected feedwater flow from the steam generator's economizer region to the downcomer region (i.e., both steam generator economizer regulating valves (FCV)(SJ) shut, both downcomer regulating valves (FCV)(SJ) opened); Concurrent with the redirection of feedwater flow, the operating main feedwater pump speed reduced to approximately 3759 revolutions per minute (RPM),

and the Feedwater Control System control methodology changed such that steam generator level controlled the amount of feedwater flow (vice a combination of measured feedwater and steam flow as well as the steam generator level). As a result of the main feedwater pump speed decreasing, inadequate discharge head was developed to overcome steam generator pressure at the low power conditions.

Steam generator levels decreased until a reactor trip occurred at approximately 0237 HST on November 16, 1988. The reactor trip resulted from a steam generator number,2 low level trip signal.

The event was properly diagnosed as an uncomplicated reactor trip.

Following the trip, the RB" Train Essential Auxiliary Feedwater Pump (BA)(P) and the Non-essential Auxiliary Feedwater Pump were manually started to feed both steam generators't approximately 0247 HST both steam generator levels had been raised to above their trip setpoints and, the event was terminated:

On November 16, 1988, the RCS leak was determined to be valve packing leakage from one of the Plant Protection System (JS)

Channel D Steam Generator differential pressure transmitter (PDT) root isolation valves ( ISV). The valve was appropriately repaired on November 19, 1988.

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4 NRC Form 344A U.S. NUCLEAR REOVLATORY COMMISSION 19S/I LICENSEE EVENT REPORT (LER) TEXT CONTINUATION APPROVEO OMS NO. 3150WIOI EXPIRES'. S/31ISS FACILITY NAME III OOCKET NUMSER (11 LER NUMSER ISI ~ AOE I31 YEAR 5EOUENTIAL AEVISION NUM45II NUM44II Palo Verde Unit Il IeeoeNE 2 o s o o o 5 2 9 8 801 4 00 04 oF 0 8 TEXT IIP more 4Oece Foe edaRNrINI WIC Fonrr SILAS I I Ill During the inspection being conducted to identify the source of the RCS leakage, it was discovered that the water from the leaking valve was flowing through small cracks in a concrete wall for the incore instrumentation (IG) chase. An engineering evaluation of this condition was initiated.

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

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

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

The failed bolt was manufactured from carbon steel and was exposed to boric-acid from the reactor coolant system. The boric acid degraded the bolt until tensile forces resulted in failure.

E. Failure known:

mode, mechanism, and effect of each failed component, if The RCS leak resulted from a broken packing gland follower bolt.

The broken bolt allowed the packing gland follower to cant which reduced the compression on the packing and resulted in excess packing leakage.

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

Not applicable - there were no component failures with multiple functions.

G. For failure 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 - no safety systems were rendered inoperable. The failed bolting resulted in packing leakage significantly below Technical Specification 3.4.5.2 limits.

Hethod of discovery of. each componen'- or system failure or procedural error:

The broken bolting was, discovered during ANPP's investigation into the reason for the reactor cavity sump level increase (see Section I.B). There were no system failures. Procedural inadequacies were discovered as a result of ANPP's Post Trip Review process.

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NRC Porro 3OSA (843 I U 8 NUCLEAR REOVLATORY COMMISSION LICENSEE EVENT REPORT {LER) TEXT CONTINUATION APPROVEO OMS NO 3ISOWIOO EXPIRES: 8/31/88 PACILITY NAME III OOCIIET NVMSER Il) LER NUMSER IS) PACE I3)

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akNliorW H/IC Porrrr 38EA8/ I IT) o s o o o 529 88 0 1 4 0 0 05or 0 8 I. Cause of Event:

There were concurrent contributory causes which resulted in the reactor trip. The first cause is that the automatic Feedwater Control System response was not adequate in maintaining a sufficient feedwater pump speed following the period that feedwater flow to the steam generators shifted from the economizer region to the downcomer region. In part, this was due to main feedwater pump speed adjustments which were made by operations personnel during normal power operations (The adjustments procedurally were required to minimize economizer valve oscillations). The adjustments, coupled with a programmed main feedwater pump speed limitation when feedwater flow shifts to the downcomer region, resulted in pump speed being too slow for the existing plant conditions. It should be noted that a site modificati.on was installed in July 1988 which lowered the main feedwater pump minimum speed. The site modification was originally implemented to resolve Feedwater Control System (FWCS) performance problems in Unit l. It had been necessary to take manual control of the FWCS, vice leaving it in automatic, during low power operations in order to prevent overfeeding the steam generators. The site modification was also prepared and implemented in Units 2 and 3 and,the resulting FWCS was more versatile; however, the system potentially required operator action during certain plant conditions. Inadequate reviews were conducted for identifying and delineating the necessary operator actions. This was especially important for Unit 2 since the steam pressure is approximately 20 psi higher in Unit 2 than Unit I during low power operations.

Another contributory cause is that procedural controls utilized at low power operations (i.e., below twenty percent power) did not contain explicit .guidance for 'ensuring that the proper adjustments were made to the automatic main feedwater pump speed control to compensate for the adjustments made at normal power operations. As a result of implementing the site modification to reduce minimum feedwater pump speed, procedure revisions were not initiated which would have provided additional'uidance for ensuring adequate feedwater supply.

Another concern, which may have had an impact on this event, involves operator performance. Control Room operating personnel (utility, licensed) on-shift during the power reduction did not take the appropriate compensatory measures which would have maintained main feedwater pump speed at an adequate level for feeding the steam generators. This would have required adding significant positive bias to the feed pump speed controller which was not addressed in the procedure. If additional information regarding this concern is identified which would significantly alter the perception of this event, a supplement to this report will be submitted.

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U.S. NUCLEAR REOULATOR Y COMMISSION LICENSEE EVENT REPORT ILER) TEXT CONTINUATION APPAOVEO OM8 NO. 3)SOLI Of EXPIRES: 8/3) /88 FACILITY NAME I'l OOCI)ET NUMSER )3)

LE/I NUMSEI) (4) ~ AOE )3)

@j Sf QVE NT/AL Pa) /IEV/SION v.,'VMSf/I Mv. NI/Mf4 A Palo Verde Unit 2 o so o o 52 9 88 0 1 4 0 0 06or- 0 8 TEXT //P /lMlP f//PCP /4 nPPV/)P/L V>> /RIRNPM//V/IC //Pnll 34SAY/ I)7)

There were no unusual characteristics of the work location which contributed to this event. Except as noted above, procedural controls have been determined to be adequate.

The concrete walls surrounding the incore instrumentation chase consist of mass concrete placed to form the chase, access shaft, ventilation shaft, and reactor cavity. The walls are variable thickness and are load bearing in that 'they transfer loads from the primary shield above to the containment basemat. The walls are under constant compression with relatively low stress levels.

There are no flexual or tensile stress loads. The cracks identified in the incore chase concrete wall are vertical cracks probably induced by mass volume changes which resulted from temperature changes. This type of crack formation is not unusual.

J. Safety System Response:

The following manual and automatic safety system responses occurred:

Plant Protection System automatic initiation of reactor trip.

Essential Auxiliary Feedwater Pump "B" was manually started by Control Room personnel.

K. Failed Component Information:

The broken packing gland follower bolt was supplied as part of the valve manufactured by Borg Warner. The model number of the valve is 77540. The failed bolt is manufactured from A540 Grade B23 carbon steel material.

II. ASSESSMENT OF THE SAFETY CONSEQUENCES AND IMPLICATIONS OF THIS EVENT:

There were no safety consequences or implications resulting from this event. There was no impact on public health and safety. The uncomplicated reactor trip occurred per design as a result of the low steam generator water level. Water'evel remained above the point which would have required an automatic Auxiliary Feedwater Actuation (JE)(BA). Adequate heat removal capabilities existed throughout the event. This event could not have occurred at higher power levels because the Feedwater Control System swapover from the economizer to the downcomer is controlled by Nuclear Instrumentation at 15 percent power.

There were no safety consequences or implications resulting from the bolting failure on the differential pressure transmitter isolation valve. Leakage through the valve's. packing as a result of the failed bolt remained below Technical Specification limits for continuous operation. There are two packing gland follower bolts on the affected valve. The other bolt remained functional throughout the event.

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2 o s o o o 529 8 8 0 1 4 00 07 OF 0 8 TEXT ////INv4 44444 444 aAM/on4////IC fomI 355l'4/ 1171 There are no safety consequences or implications resulting from the cracks in the incore instrument chase concrete wall. The type of crack formation is not unusual. Since the cracks are vertical, no load transfer path is interrupted and no other design function is compromised. The structural design basis of the containment internal structure is unchanged by the cracks.

I I I. CORRECTIVE ACTIONS:

A. Immediate:

The failed bolt as well as the other remaining bolt on the instrument root valve have been replaced.

B. Action to Prevent Recurrence:

As action to prevent recurrence, additional instructions have been included in operating procedures to ensure that operations personnel take the appropriate measures for maintaining adequate main feedwater pump speed during low power operations.

An engineering evaluation of Unit 2's automatic feedwater control system operation will be performed. As an initial result of the Engineering Evaluation, an enhancement to the preventive maintenance task for the feedpump governor is being implemented.

A Human Performance Evaluation is being performed to address factors which contributed to the operations personnel performance concern. Additionally, this event will be reviewed by the appropriate operations department personnel from Units I, 2, and 3 during normally scheduled periodic training. If additional corrective action is identifie'd as a result of the Human Performance Evaluation which significantly alter impact the perception of this event, a supplement to this report will be issued.

As discussed in Section I. I, the appropriate procedure changes did not get implemented as a result of the site modification. As a corrective action, the current site modification administrative controls will be evaluated and improved where appropriate. The system engineer program and the system engineer/Plant Standards and Control interface will be evaluated to determine can be implemented to ensure that necessary procedure changes are if improvements incorporated following plant modifications. A representative sample of cur'rent site modifications will be reviewed to determine if additional procedure revisions are necessary.

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t 2 o s o o o 5 2 9 8 8 0 1 4 '00, 08 or- 0 8 oee edd/dorM/A/RC form 3//SA'e/1171 Concerning the bolt failures discussed in Section I.B, the problem with boric acid causing premature failures had been previously identified. An engineering evaluation had been performed and as corrective action, a new bolting material was specified (ASTH A564 TP 630). The bolting is being changed out on an "as-needed" basis.

Concerning the cracks in the incore instrument chase concrete wall, a procedure for sealing the cracks is being developed. The cracks will be sealed during Unit 2's next refueling outage. It should be noted that this is a long-term solution for corrosion protection.

A design or structural repair is not required.

IV. PREVIOUS SIHILAR EVENTS:

There have been no previous similar .events reported pursuant to IOCFR50.73. It should be noted that other reactor trips have been reported which resulted from feedwater flow problems; however, none involve the sequence of events or root cause described in this LER.

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0 0 P Arizona Nuclear Power Project P.O. BOX 52034 ~ PNOENIX. ARIZONA85072-2034

,192-00437-JGH/TDS/DAJ December 14,, 1988 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 88-014-00 File: 88-020-404 Attached please find Licensee Event Report (LER) No. 88-014-00 prepared and submitted pursuant to 10CFR 50.73. I'n accordance with 10CFR 50.73(d)., we are herewith forwardi'ng a copy of the LER to the Regional Administrator of the Region V office.

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

Very tru y yours, I/g J'. G. Haynes Vice President Nuclear Production JGH/TDS/DAJ/kj Attachment CC: D. B. Karner (all w/a)

E. E. Van Brunt, Jr.

J. B. Hartin T. J. Polich H. J. Davis A. C. Gehr INPO Records Center

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