ML17305A436
| ML17305A436 | |
| Person / Time | |
|---|---|
| Site: | Palo Verde  |
| Issue date: | 11/30/1989 |
| From: | James M. Levine, Shriver T ARIZONA PUBLIC SERVICE CO. (FORMERLY ARIZONA NUCLEAR |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| 192-00557-JML-T, 192-557-JML-T, LER-89-009, LER-89-9, NUDOCS 8912130271 | |
| Download: ML17305A436 (24) | |
Text
ACCELERATED DIyHK3UTION DEMONSTyWTION SYSTEM REGULATORY INFORMATION DISTRIBUTION 'SYSTEM (RIDS)
ACCESSION NBR:8912130271 DOC.DATE: 89/ll/30 NOTARIZED: NO DOCKET Palo Verde Nuclear Station, Unit 2, Arizona Publi 05000529 g'ACIL:STN-50-529 AUTH. NAME AUTHOR Arizona Public Service Co.. (formerly Arizona Nuclear Power AFFILIATION'HRIVER,T.D.
LEVINE,J.M. : Arizona Public Service'o. (formerly Arizona Nuclear Power RECIP.,NAME RECIPIENT AFFILIATION
SUBJECT:
LER 89-009-01:on 890712,reactor, forced flow.
trip due to partial loss of W/8- = ltr.
DISTRIBUTION CODE: IE22T COPIES RECEIVED:LTR 'ENCL SIZE:
TITLE: 50.73/50.9 Licensee Event, Report (LER), Incident Rpt, etc.
/
NOTES.:Standardized plant. 05000529 RECIPIENT COPIES RECIPIENT COPIES ID CODE/NAME LTTR ENCL ID'ODE/NAME LTTR,ENCL PD5 LA . 1 1 PD5 PD 1 1 CHAN,T 1 1 DAVIS,M. 1 1.
S',
INTERNAL: ACRS,MICHELSON 1 1 ACRS MOELLER 2 2 ACRS WYLIE 1 1 AEOD/DOA 1 1
'AEOD/DS P/TPAB 1 1 AEOD/ROAB/DSP 2 2 DEDRO NRR/DET/EMEB9H3
. 1 1 NRR/DET/ECMB 9H 1,
'1 1 1 NRR/DET/ESGB 8D 1 NRR/DLPQ/LHFB11 1 1 -'NRR/DLPQ/LPEB10 1 1 NRR/DOEA/OEAB11 1 1 NRR/DREP/PRPB1 1 2 2 NRR/DST/SELB 8D 1 1 NRR/DST/SICB 7E 1 1 NRR/DST/SPLB8D1 1 1 /J3ST/SRXB 8E 1 1 NUDOCS-ABSTRACT 1 1 REG F MFILE 02 1 1 RES/DSIR/EIB 1 1 RGN5 01 1 1 EXTERNAL EG&G WILLIAMSi S" 4 4 L ST LOBBY WARD 1 1 LPDR 1 1 NRC PDR 1 1 NSIC MAYS,G 1
'1
- l. NSIC MURPHY,G.A 1 1 NUDOCS FULL TXT 1 NOTES 1 1 D S
D D
NOTE TO ALL "RIDS" RECIPIENTS:
S PLEASE HELP US TO REDUCE WAS'! CONTACI'HE,DOCUMENT CONTROL DESK, ROOM Pl-37 (EXT. 20079) TO ELIMINATEYOUR NAME FROM DISIRIBUTION LISIS FOR DOCUMENTS YOU DON'T NEED)
FULL TEXT CONVERSION REQUIRED TOTAL NUMBER OF COPIES'EQUIRED: LTTR 40 ENCL 40
N ik
.Arizona Public Service Company PALO VERDE NUCLEAR GENERATING'STATION P.O BOX S2034 ~ 'PHOENIX. ARIZONA 85072-2034, 192-'00557-JML/TDS/RKR November 30, 1989 U. S. Nuclear Regulatory Commission NRC Document Control Desk Washington, D.C. 20555
Dear Sirs:
Subject:
,Palo Verde Nuclear Generating Station (PVNGS)
Unit 2 Docket No. STN'0-529 (License NPF-51)
Licensee Event Report 89-009-01 File': 89-020-404
. Attached please find Supplement Number 1 to Licensee Event Report (LER) No.
89-009-00 prepared and submitted pursuant'to the .requirements of 10CFR 50.73.
In accordance with 10CFR 50.73(d), we are. herewith forwarding a copy'f this report to the Regional.'Admini'strator of the Region V Office.
If you have any questions, please contact T. D. Shri.ver, Compliance Manager at (602) 393-2521.
Very truly yours,
. M. Levine ice President Nuclear Production JML/TDS/RKR/kj Attachment cc: W. F. Conway (all w/a)
E. E. Van Brunt J. B. Martin T. J.- Polich M. J. Davis A. C. Gehr INPO Records Center S912130271 -S91130 PDR ADOCK 05000529 S PDC
0 0 NRC FORM 366 V.S. NUCLEAR REGULATORY COMMISSION (64)9) APPROVED OMB NO. 31504)104 E XP I R ES: 4/30/92 ESTIMATED BURDEN PER RESPONSE TO COMPLY WTH THIS INFORMATION COLLECTION REOUESTt 50.0 HRS. FORWARD LICENSEE EVENT REPORT (LER) COMMENTS REGARDING BURDEN ESTIMATE TO THE RECORDS "
AND REPORTS MANAGEMENT BRANCH (F630), U.S, NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20555, AND TO THE PAPERWORK REDUCTION PROJECT 131500104). OFFICE OF MANAGEMENTAND BUDGET, WASHINGTON, DC 20603.
FACILITY NAME (II DOCKET NUMBER l2) PAGE 3 Palo Verde Unit 2 o 5 o o o 52 91oF10 TITLE Isl Reactor Trip, Doe to Partial Lo'ss of Forced'Flow EVENT DATE (5) LER NUMBER'(6) REPORT DATE (7) OTHER FACILITIES INVOLVED (6)
MONTH OAY YEAR YEAR 4 SEOVENTIAL REVISION MONTH OAY YEAR FACILITYNAMES DOCKET NUMBER(S)
- N% NUMBER 9):v NUMBER N/A 0 5 0 0 0 0 7 1 2 8 9 8 9 0 09 0 1 11 30 8 9 N/A 0 5 0 0 0 THIS REPORT IS SUBMITTED PURSUANT T 0 THE RLOUIREMENTS OF 10 CF R ('): IChech one or more of the followlnpl (11 OPERATING MODE IS) 20A02(b) 20.405(c) 60,73( ~ ) I 2) I iv) 73.71(III POWER 20.406(el(I) ll) 00.30(c) (1) 60.73 le) l2) iv) 7$ .71(c)
LEYEL 1 0 0 20A05( ~ )(1)IQI 50.30(c)(2) 50.73(el(2)(vill)(AI 50.73(s l(2) (vill l(2)(vill)(B)
OTHER ISpecify in Abrtrect belOw end ln Text, NRC Form 20A06( ~ Ill)I(ill 40.73(s I l2)(l) SSSA/
20A05(el)i)(lv) 40.73(eN2) (6). 50.73( ~
g y%m s
@ Special Report Ak .J.... x~&Y. g4: 20A06( ~ )11)(v) 50.7$ (e) (2((ill) 50.73( ~ ) (2) (x I LICENSEE CONTACT FOR THIS LER (12)
NAME TELEPHONE NUMBER AREA CODE Timothy D. Shriver, Compliance Hanager 6 02 39 3 2 521 COMPLETE ONE LINE'OR EACH COMPONENT'AILURE DESCRIBED, IN THIS REPORT l13) fr ". Rsxnvx.
CAUSE SYSTEM COMPONENT MANUFAC EPORTABLE )~c<f~~
CAUSE SYSTEM COMPONENT MANUFAC EPORTABLE, ~
. ixl TURER TO NPRDS TVRER TO NPRDS
.v..i rX+ PR..
X E B F U G 0 8 0 N A B F 1 2 5 Y X S J' P032Y SUPPLEMENTAL REPORT EXPECTED (14l MONTH
'~vs:~0>c~AC% 5 DAY YEAR EXPECTED SUBMISSION DATE (15)
YES Ilf yer, complete EXPECTED SVBstlSSIOff DATE/ NO ABSTRACT /Limit to /400 rpecn, I e.. epproxlmstely fifteen slnple rptce typewrlnen linn/ (10l On July 12, 1989 at approximately 2212 HST Palo Verde Unit 2 was operating at approximately 100 percent power when 2 of the 4 reactor coolant pumps were shed from their power supply (Bus 2E-NAN-S02), resulting in a reactor 'oad trip on calculated low DNBR due to low reactor coolant flow. Immediately following the trip, a Safety Injection Actuation Signal (SIAS) and Containment Isolation, Actuation Signal (CIAS) Engineered Safety Features occurred on low Reactor Coolant System (RCS) pressure. Following the event, at approximately 1529 HST on July 13, 1989, a portion of the main feedwater system (HFWS) was overpressurized.
The cause of the load shed was a failed fuse in the bus potential transformer. The cause of the SIAS/CIAS was RCS depressurization due to improper Steam Bypass Control System (SBCS) response and leaking pressurizer spray valves. The cause of the HFWS overpressurization was a failed check valve.
Immediate corrective action taken was to replace the fuse.
This submittal also provides a Special Report in accordance with Technical Specification 3.5.2 ACTION b.
NRC Form 360 (669)
0 0 NRCFORM368A U.S. NUCLEAR REGULATORY COMMISSION (669) APPROVED OMB NO. 31500)04 EXPIRES: 4/30/92 ESTIMATED BURDEN PER RESPONSE TO COMPLY WTH THIS LICENSEE EVENT REPORT ILER) INFORMATION COLLECTION REQUEST) 50.0 HRS. FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE RECORDS TEXT CONTINUATION AND REPORTS MANAGEMENT BRANCH (F430). U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON, OC 20555, AND TO THE PAPERWORK REDUCTION PROJECT (3)504)104). OFFICE OF MANAGEMENTAND BUDGET. WASHINGTON. OC 20503.
FACILITYNAME (1) DOCKET NUMBER (21 LER NUMBER (8) PAGE (3)
?rior) SEQUENTIAL IIEVISION NUMSEII NUMSER Palo Verde Unit 2 TEXT //f moro s/Moo is ror/oirod, oso oddr'iona/ NRC Form 366r4's/ (12) o s o o o 529 8 9 0 0 9 01 0 2o" 1 0 I. DESCRIPTION OF WHAT OCCURRED:
A. Initial Conditions:
approximately 2212 HST on July 12, 1989, Palo Verde Unit 2 was in at approximately 100 percent power.
B.'tReportable'Event Hode 1 (POWER OPERATION)
Description (Including Dates and Approximate Times of Hajor Occurrences):
Event Classification: Any event or condition that resulted in manual or automatic actuation of any Engineered Safety Feature (ESF)(JE),
including the Reactor Protection System (RPS)(JC).
A't approximately 2212 HST on July 12, 1989, a reactor (RCT)(AC) trip on low Departure from Nucleate Boiling Ratio (DNBR) occurred due to a partial loss of reactor coolant (AB) flow. The partial loss of flow occurred when two (2) of the four '(4) Reactor Coolant Pumps (RCP)(AB) were load shed from their power supply (Bus 2E-NAN-S02)(BU)(EA). Following the reactor trip, at approximately 2213 HST a Safety Injection Actuation Signal (SIAS)(JE) and a Containment Isolation Actuation Signal (CIAS)(JE) occurred when the Reactor Coolant System (RCS)(AB) pressure decreased to approximately 1823 psia (approximately 14 psi below the actuation
. setpoint). All safety system components actuated as designed. The plant was stabilized at approximately 2322 HST and the event was terminated.
Prior to the event, at approximately 2001 HST on July 12, 1989, the Control Room received a trouble alarm annunciator (ALH)(ANN) for Bus 2E-NAN-S02. Operations personnel (utility, non-licensed) investigated and inspected Bus 2E-NAN-S02. They could not determine the reason for the alarm on Bus 2E-NAN-S02. At approximately 2020 HST a control room annunciator alarm indicated that the unit oscillograph (OSG) had operated. Operations personnel (utility, licensed) responded to the annunciator alarm.
The oscillograph indicated that there had been an undervoltage condition on Bus 2E-NAN-S02. Also, the digital fault recorder (XR) printout indicated a disturbance on Bus 2E-NAN-S02. Since no apparent problems were identified with Bus 2E-NAN-S02, unit operations personnel (utility, licensed) continued to investigate the problem. From approximately 2206 HST to approximately 2212 HST the Oscillograph alarmed three more times. Operations personnel were responding to the alarms when at approximately 2212 HST, a
,load shed actuation occurred on Bus 2E-NAN-S02 even though Bus 2E-NAN-S02 remained energized. The load shed caused the loads on NRC Form 366A (81)9)
ik NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION (64)9) APPROVED OM 9 NO. 31500104 EXPIRES; 4/30/92 ESTIMATED BURDEN IrER RESPONSE TO COMPLY WTH THIS LICENSEE EVENT REPORT ILER) INFORMATION COLLECTION REQUEST: 500 HRS. FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE RECORDS TEXT'CONTINUATION AND REPORTS MANAGEMENT BRANCH (P.530), U.S. NUCLEAR>>
REGULATORY COMMISSION, WASHINGTON, DC 20555, AND TO THE PAPERWORK REDUCTION PROJECT (31504))04). OFFICE OF MANAGEMENTAND BUDGETt WASHINGTON DC 20503 FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER (6) PAGE (3)
Pied SEQUENTIAL 4g' REVISION NUMBER
~ NUM Pal o Verde Uni t 2 o s o o o 5 2 9 8 9 0 0 901 0 1 0 TEXT N moro totOt it ttqrrritd, Irtt tdditionti Fii(C Form 3%A'tl (17)
Bus 2E-NAN-S02 to be deenergized.
The RCPs are powered from non-class lE 13.8 kv Busses 2E-NAN-S01 (BU)(EA) and 2E-NAN-S02. RCPs lA and 2A are powered from Bus 2E-NAN-S01 and RCPs 1B and 2B are powered from Bus 2E-NAN-S02. Bus 2E-NAN-S01 was being supplied by a Startup Transformer (XFMR)(EA) and Bus 2E-NAN-S02 was being supplied by the Unit 2 Auxiliary Transformer (XFMR)(EA). Since RCPs 1B'nd 2B were being supplied from Bus 2E-NAN-.S02, the RCPs were deenergized and a partial loss of flow occurred resulting in a reactor trip on low DNBR.
r Approximately one minute after the reactor trip, Reactor Coolant System, pressure dropped lower than normal due to improper Steam Bypass Control System (SBCS)(SG) response and leaking pressurizer spray valves (P2R)(AB)(V). This resulted in concurrent safety injection and containment isolation Engineered'afety Features (ESF) actuations (JE) when the RCS pressure decreased to approximately 1823 psia, which is 14 psi below the low actuation pressure setpoint of 1837 psia and 1 psi above the minimum allowable trip setpoint value of 1822 psia. Immediately following the safety injection, pressurizer level and pressure stabilized.
Pressure then began to trend toward steady state Mode 3 (HOT
. STANDBY) conditions (approximately 2250 psia).
At approximately 2223 MST on July 12, 1989, a Notification of Unusual Event (NUE) was declared. The NUE was decl'ared pursuant to EPIP-02, "Emergency Classification" as a result of the SIAS on low pressurizer pressure.
Operations personnel (uti,lity, licensed and non-licensed) investigated the protective relay targets (RLY) on Bus 2E-NAN-S02 and could find no reason for the protective relay actuation. At approximately 2234 HST on July 12, 1989, the load centers powered from Bus 2E-NAN-S02 were reenergized in accordance with an approved procedure. Eight minutes later at approximately 2242 HST, a load shed signal again deenergized the load centers on Bus 2E-NAN-S02.
At approximately 2302 HST, Bus 2E-NAN-S02 was deenergized'nd taken out of service in order to perform further troubleshooting in accordance with the PVNGS Work Control Program.
At approximately 2322 HST on July 12, 1989 the SIAS/CIAS ESF actuations were secured, plant conditions were stabilized, and the NUE was terminated.
At approximately 0300 HST on July 13, 1989, Protection Relaying and Control (PR&C), the APS group responsible for investigating the oscillograph operated alarm,, reset the alarm in accordance with procedure 42AL-2RK1B,'Panel BOlB Alarm Response".
NRC ForRI 366A (669)
II NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION (6419) APPROVED 0MB NO. 31504)104 EXPIRES: 4/30/92 ESTIMATED BURDEN PER RESPONSE TO COMPLY WTH THIS LICENSEE EVENT REPORT (LERI INFORMATION COLLECTION REOUEST: 50.0 HRS. FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE RECORDS TEXT CONTINUATION AND REPORTS MANAGEMENT BRANCH (P.530), U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20555, AND TO THE PAPERWORK REDUCTION PROJECT (31504))04). OFFICE OF MANAGEMENTAND BUDGET, WASHINGTON, DC 20503.
FACILITYNAME (1) DOCKET NUMBER (2) LER NUMBER (6) PAGE (3)
'4 SEQUENTIAL .>~ro REVISION v.
NUMBER 3'IJM SR Palo Verde Unit 2 o s o o o 5 2 9 8 9 0 0 9 01 04' 0 TEXT fffmoro 4FRCO fo roquiod, oJP oddl)fooof HRC Form 366A'4) (17)
Following the event, at approximately 1529 HST on July 13, 1989, the secondary plant was being placed in the long path recirculation (LPR) mode in accordance with normal operating procedures. Steam generator levels (SG)(AB) -were being maintained using the Non-essential Auxiliary Feedwater Pump (P)(BA). Due to back-leakage through a check valve, Auxiliary Feedwater System (AFW)(BA) pressure of approximately 1580 psia was applied to a portion of the Hain Feedwater System (HFWS)(SJ). 'This portion of the Hain Feedwater System was rated for at least 1580 psia. When the manual isolation valve (ISV) on the Hain Feedwater Pump (HFP)
RBR bypass line was opened to establish long path recirculation in accordance with an approved procedure, the suction piping to the Hain Feedwater Pumps was overpressurized. The suction piping is rated for 500 psia.
Concurrent with the HFP bypass valve being opened, a low suction pressure trip for Hain Feedwater Pumps "AH and RBR was received in the Control Room. Immediately following the low suction pressure trip, operations personnel (utility, non-licensed) observed an abnormal decrease in seventh point feedwater heater (SN) outlet temperature. To prevent thermal shocking of the feedwater heater, operations personnel isolated the Non-essential Auxiliary Feedwater System from the Hain Feedwater System. At approximately 1545 HST, the overpressurization event was, terminated.
C. Status of structures, systems, or components that were inoperable at the start of the event that contributed to the event:.
There were no structures, systems, or components inoperable at the start of the event which contributed to the event.
D. Cause of each component or system failure, if known:
The cause of the Bus 2E-NAN-S02 load shed malfunction described in Section I.B has been determined to be a failed potential transformer (PT) primary fuse .on the RC" phase. An original equipment manufacturer (General Electric) analysis concluded that the fuse failure was probably initiated by a previous current surge on the system. The surge damaged the fuse which effectively reduced its rated continuous current capacity. This caused the fuse to melt slowly over time at currents below its minimum interrupt rating. The current surge .may have been caused by an over voltage condition, which would have caused the PTs to.saturate and draw momentary heavy currents.
The cause of the back.-leakage through the HFP bypass check valve (SGN-V431) described in Section I.B has been determined to be loose fasteners which allowed the check valve disc to drop and not seat NRC F olin 366A (669)
II NRC FORM 366A (689) 0 US. NUCLEAR REGULATORY COMMISSION APPROVED OMB NO.31500104 E XPI RES L4/30/92'STIMATED BURDEN PER RESPONSE TO COMPLY WTH THIS LICENSEE, EVENT REPORT tLER) INFORMATION COLLECTION REQUEST: 500 HRS. FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE RECORDS TEXT CONTINUATION AND REPORTS MANAGEMENT BRANCH (F430); U.S. NUCLEAR REGUL'ATORY COMMISSION, WASHINGTON. DC 20555. AND TO THE PAPERWORK REDUCTION PROJECT (31504)104),.OFFICE OF MANAGEMENTAND BUDGETWASHINGTON,DC 20503.
FACILITY NAME O) DOCKET NUMBER (2) LER NUMBER (6) PAGE YEAR SEOVENTIAL REVISION NUMBER NVMSSR (3)'alo Verde Uni,t 2'EXT
/IImoro rproo /4 rrh/o/Nd. ore odChr/orho/ NRC Form 3/hSA 3/ (I2 I o 5 o o o 5 2 9 8 9 00901,05 OF 1 0 on its seating surface. The .fastener locking devices required by
=
the Technical Manual'ere not installed., The cause .for the .locking device not being installed cannot be. determined since our records show no work has been performed on this valve since initial startup of Unit 2. This valve is a Rnon-safety related" component.
Failure known:
mode, mechanism, -and:
'r effect of each, failed component, if The failure of .the potential transformer (PT) primary RCR phase fuse resulted in a load shed of Bus 2E-NAN-S02 including deenergization of reactor coolant pumps (RCP) 1B and 2B. This-resul.ted in a reactor trip and turbine -trip as described in Section I.B. Shortly before the .reactor trip, there was indication of a problem wi:th Bus 2E-NAN-S02 by a trouble, alarm annunciation for the bus, an oscillograph operated annunciator alarm, oscillograph indication of an undervoltage condition on the bus, and a digital-fault recorder indicating a fault. on the bus. Operations personnel (utility, licensed and non-licensed) were attempting to identify,-
the problem when the. reactor trip occurred. The cau'se of failure of the fuse is described 'in Section I.D.
The failure of check valve SGN-V431 resulted in overpressurizing a portion of the main feedwater system as'escribed in Section I.B.
The cause of the check valve failure is described. in,Section I.C.
'F. 'or failures of components with multiple, functions, list of systems or'econdary functions that were also affected:
Not applicable - No component .failures had multiple functions which-affected other systems or components.
G. For failures that rendered a train of a safety system inoperabl'e, estimated time elapsed from the discovery of the fai,lure unti-1'he train was, returned to service:
Not appl'icable - There were no failures that rendered a train of a safety system inoperable.
Method of discovery of each component or system failure or procedural error:
The failed potential transformer primary RCR phase fuse, was discovered as a result of troubleshooting performed after the event.
The failed check'valve SGN-V431 was discovered as a result of troubleshooting performed after the event.
NRC Form 366A (689)
I~ 0 NRC FORM 366A U.S. NUCLEAR RFGULATORY COMMISSION (669) APPROVED OMB NO. 31500104 EXPIRES: 4/30/92 ESTIMATED BURDEN PER RESPONSE TO COMPLY WTH THIS LICENSEE EVENT REPORT (LER) INFORMATION COLLECTION REQUEST: 50.0 HRS. FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE RECORDS TEXT CONTINUATION AND REPORTS MANAGEMENT BRANCH (P 530). U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20555. AND TO THE PAPERWORK REDUCTION PROJECT (31500104). OFFICE OF MANAGEMENTAND BUDGET, WASHINGTON, DC 20503.
FACILITY NAME (I) DOCKET NUMBER (2) PAGE (3)
LE R NUMBE R (61 SEQUENTIAL (u> REVISION NN NUMSEII ~ >23 NUMBER Palo Verde Unit 2 0 5 0 0 0 5 2 9 8 9 0 0 9 0 1 06 '"
TEXT iilmors sosse is rsr)eked, Irse sddr'osrMI A'RC Farm 366A'4) ()2)
The overpressurization of the HFW pump suction piping was discovered when troubleshooting the feedwater pump low suction pressure trips as described in Section I.B. The bellows in all six low suction pressure switches were found to be deformed due to the pressure transient.
'Cause of Event::
The cause of the reactor trip discussed in Section I.B was a =
failed fuse (Component Failure, SALP Cause Code E) which caused a load shed signal on Bus 2E-NAN-S02 resulting in a partial loss of flow. The "CR phase potential transformer (PT) is connected to Bus 2-NAN-S02. The fuse is located between the PT and the bus. Relay 227-S monitors the bus voltage through the PT. The relay load sheds the bus when the voltage is less than or equal to 77 percent of the bus voltage,(13.8 kv). When the fuse failed, the relay saw no-voltage from the PT and actuated the load shed relays.
- 2. The cause of the SIAS/CIAS ESF actuation described in Section I.B. was RCS depressurization due to improper Steam Bypass Control System (SBCS) response and excessive leakage past the pressurizer spray. valves.
- a. The SBCS response was caused by the method of calibration used for the Proportional Integral (PI) Controller in the SBCS quick open controllers. This allowed the SBCS valves to remain open longer than was required. PVNGS calibrates the PI controller using the dial settings provided by the Combustion Engineering (CE) setpoint document. The CE setpoint document. also provides calibration curves for optimizing the quick open controller response. However, the CE setpoint document did not cl,early indicate that these curves were to be used as part of the PI controller calibration.
Therefore, PVNGS calibration procedures only required use of the dial settings and did not require use of the calibration curves to optimize the SBCS response (Defective Procedure, SALP Cause Code D) .
- b. The cause of 'the pressurizer spray valves not closing was improper post maintenance adjustments due to inadequately defined work steps and work scope. This resulted in misalignment of the positioner (Defective Procedures, SALP Cause Code D).
NRC Form 366A (64)9)
Ik NRC FORM 366A US. NUCLEAR REGULATORY COMMISSION (669) APPROVEO 0MB NO. 31500)04 EXPIRES: 4/30/92 ESTIMATED BURDEN PER. RESPONSE TO COMPLY WTH THIS L'ICENSEE'EVENT REPORT ILER) INFORMATION COLLECTION REQUEST: 500 HRS. FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE RECORDS TEXT CONTINUATION ANO REPORTS MANAGEMENT BRANCH (F430), U.S. NUCLEAR REGULATORY COMMISSION, IVASHINGTON;OC 20555, AND TO THE PAPERWORK REDUCTION'PROJECT "(31504)104). OFFICE OF MANAGEMENTAND BUDGET, WASHINGTON, OC 20503. "
FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER (6) PAGE (3)
Wi; SEQUENTIAI REVISION YEAR $
IPR 'UMBER NUMBER Palo Verde Unit 0 5' 0 0 5 2 9 8 9 0 0 9 0 1 07 o" 1. 0 TEXT ///mont Seot itntOIA(td, ott tddit/ont/iVRC Fonm 366A'4/(17)
It was 'also determined that, the positioner for, one of the valves (2JRCEPV100F)'ad failed'. As a result of APS's investigation, it has .been determined that the cause of failure for the posi.tioner is accelerated deterioration of the positioner's pressure retaining o-rings due to long term exposure to a significant high temperatu're .and h'igh radiation environment. PVNGS was aware that these valves were leaking prior to this event. However, because these valves are located in containment.and inaccessible during power operation, 'they were scheduled to be reworked during, the next plant shutdown.
3.. The cause of the overpressurization of the Hain Feedwater-Pump suction piping was due to back-leakage through a check
,valve (Component Failure, SALP Cause. Code E)'s described in Secti'on I.D. This check valve is not safety related and is not currently incl.uded in .any testing or maintenance programs-.. Therefore there were no procedures for maintenance or testing of these check valves at the time of this event.
APS is developing a program to periodically inspect selected check valves. A separate engineering evaluation has been, conducted on this overpressurization event. . A copy of this evaluation.'was,provided to ',the NRC resident inspector at PVNGS and the Regional Administrator (Reference APS letter
.number 102-01448-JNB/TDS/RKR, dated October 3, 1989).
J. Safety System Response:
The following automatic and manual safety system responses occurred during this event:
- l. Containment Isolation System (automatic)(JH).
- 2. 'Low Pressure Safety Injection Trains "AR and "BH (automatic)(BP)
Emergency Diesel Generators Trains RAR and "B" (automatic)(DG)(EK)
- 5. Essential Spray Pond System Trains "AR and RB" (automatic)(BS)
- 7. Essential Cooling Water System Trains RAR and RBR (automatic)(BI)
- 9. Containment Spray Trains "AR and "B" (automatic)(BE)
NRC Form 366A (669)
II NRC FORM 356*
(669) 4 U.S. NUCLEAR REGULATORY COIAMISSION APPROVED OMB NO. 31500104 EXPIRES: 4/30/32 ESTIMATED BURDEN PER RESPONSE TO COMPLY WTH THIS LICENSEE EVENT REPORT ILER) INFORLIATION COLLECTION REOUESTI 50.0 HRS. FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE RECORDS TEXT CONTINUATION AND REPORTS MANAGEMENT BRANCH (F430). U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20555. AND TO THE PAPERWORK REDUCTION PROJECT (31504)104), OFFICE OF MANAGEMENTAND BUDGET, WASHINGTON, DC 20503.
FACILITY NAME (11 DOCKET NUMBER (2) LER NUMBER ISI PAGE IS)
I REVISION YEAR If4% 5EQVS NVMSER NVMSSR Palo Verde Unit 2 o s o o o 52 989 0.09 0 1 0 8 OF 1 0 TEXT ///mort totot /4 rtrrrrrrtd, ort tddr)/oot/ HRC Form 3554'4/ I'l)
Control Building Essential Ventilation- System Trains RAH and RBR (automatic)(AHU)
K. Failed Component Information:
The failed fuse was manufactured by General El'ectric. It is'a,type EJI, size B, rated at 15.5 kv and 0.5 amps.
The leaking check valve was manufactured by Pacific Valves. It is an 8 inch one-way flow check valve, figure number 58809-7-WE(20) .
The valve positioner was manufactured by Fisher Controls. It is a type 3590 Electro-Pneumatic Valve Positioner, serial number 6558897.
I I. ASSESSMENT OF THE SAFETY CONSEQUENCES AND IMPLICATIONS OF THIS EVENT:
The Palo, Verde Updated Final Safety Analysis Report (UFSAR) accident analysis for loss of reactor coolant system (RCS) flow assumes a total loss of offsite power resulting in a coastdown of all four reactor coolant pumps (RCPs). The accident analysi's transient is Departure from Nucleate Boiling Ratio (DNBR) limiting. The reduced RCS flow results in an initial-rise in RCS average temperature and a reduction in DNBR.
Based on this analysis, a reactor .trip on low DNBR mitigates this transient and maintains DNBR above the safety limit. For this event, only two RCPs tripped and coasted down. The Steam Bypass Control System (SBCS) reduced RCS average temperature following the reactor trip. The accident analysis bounds this event. Based on this, DNBR limits were not exceeded.
Depressurization of the RCS resulted in a SIAS. The primary function of the SIAS for this event type is to maintain RCS inventory and maintain shutdown margin. In this event all control element assemblies (CEA)(AA)(ROD) inserted and RCS average temperature decreased to 551'F.
Adequate shutdown margin was maintained and pressurizer level remained on scale throughout the event. Therefore adequate RCS inventory was maintained throughout this event.
The. check valve SGN-V431 leakage resulted in overpressurization of a portion of the Main Feedwater System pump suction piping. This portion of the Main Feedwater System performs no safety function.
All safety systems required to operate performed as designed. The event did not result in any challenges to fission product barriers or result in any releases of radioactive materials. Therefore, there were no safety consequences or implications as a result of this event. This event did not adversely affect the safe operation of the plant or health and safety of the public.
NRC Form 366A (64)S)
IS NRC FORM
-0 U.S. NUCLEAR REGULATORY COMMISSION 35SA'SSBI APPROVED 0MB ND, 31500(04 E XP I R E S'/30/92 ESTIMATED BURDEN PER RESPONSE TO COMP(.Y'WTH THIS LICENSEE EVENT REPORT ILER) INFORMATION COI.LECTION REOUESTI 60.0 HRS. FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE RECORDS TEXT CONTINUATION AND REPORTS MANAGEMENT BRANCH (PS301, U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON, OC 20555, AND TO THE PAPERWORK REDUCTION PROJECT (3(5001041, OFFICE OF MANAGEMENTAND BUDGET. WASHINGTON, DC 20503, FACILITY,NAME (1I DOCKET NUMBER (2I LER NUMBER (SI PAGE (3)
YEAR SEQUENTIAL I"RPI REVISION NUMSER NUM ER
.Palo Verde Uni.t 2 0' 0 0 0 '5 2 9 8 9 0 0 9 0 1 OF TEXT llfmort tpttt lt rtto/rN/ ost tddrllont/ HRC Form 355r('4/ (171
'III. CORRECTIVE ACTIONS:
A. ,Immediate:
The failed fuse was replaced in the potential transformer for Bus 2E-NAN'-S02.
B. Action to Prevent Recurrence:
Based on the probable cause of the potential transformer fuse failure, APS engineering has identified other potential transformer fuses that may 'have been subjected to an over-voltage condition. These fuses will be replaced with new fuses at the next outage of sufficient duration in Units 1 and 2. No fuses have been identified 'in Unit 3 which have been subjected to an over-voltage condition.
- 2. The Units 2 and 3 SBCS quick open modules have been optimized using the. nominal setpoint values and the controller performance curve in accordance, with .an approved work order.
Optimizati'on of the SBCS quick open modules in Unit 1 will,be completed prior to startup from the current outage. The SBCS-calibration procedure has been reviewed to ensure that all SBCS modules were calibrated using the technique, setpoints and tolerances in the Combustion Engineering setpoint document. The SBCS calibration procedure is being revised to incorporate the controller performance curve. The procedure revision is expected to be issued by January 1990.
- 3. The failed Unit 2 positioner was replaced. The positioner.on the other spray valve in Unit 2 has been replaced since the last refueling outage and does not require rework until the next refueling outage. The Unit 2 pressurizer spray valves
.were readjusted in accordance with .an approved procedure to correct the excessive leakage condition. One of the Unit 3 positioners was replaced in'ugust 1988. The other Unit 3 positioner was replaced as a result of this event. This positioner was disassembled to evaluate the components. The evaluation did not .identify any accelerated degradation.
Based on this evaluation no further action is required for the positioner installed in August 1988 until the next refueling outage. The Unit 3 pressurizer spray valve adjustment has also been checked. The Unit 1 positioners will be replaced or reworked and the adjustment of the pressurizer spray valves will be checked prior to startup from the current outage.
A preventive maintenance task is also being developed to require replacement of the components in the positioner that NRC Form 3SEA (BS91
NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION APPROVED OMB NO. 31500104 (6691 EXPIRES: 4/30/92 ESTIMATED BURDEN PER RESPONSE TO COMPLY WTH THIS LICENSEE EVENT REPORT ILER) INFORMATION COLLECTION REQUEST 500 HRS. FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE RECORDS TEXT CONTINUATION AND REPORTS MANAGEMENT BRANCH (P$ 30I. U.S. NUCLEAR REGULATORY COMMISSION. WASHINGTON. DC 20555. AND TO THE PAPERWORK REDUCTION PROJECT (31500104). OFFICE OF MANAGEMENTAND BUDGET, WASHINGTON, DC 20503.
FACILITY NAME (II DOCKET NUMBER (21 LER NUMBER (61 PAGE (31
':.+ SEOI/ENTIAL REVISION NUMSER NIJMSER Palo Verde Unit 2 o s o o o 52 9 8 9 00 9 01 1 0 OF 1 0 TEXT //f moro 4/roro /4 roSIh/rod, SM odChr/ohM/ NRC Form 366AB/ (17) are subject to accelerated degradation due to radiation and temperature each refueling outage . This preventive maintenance task is expected to be developed by January 1990. A maintenance procedure is also being developed to provide detailed instructions for adjustment of the valves. '
The procedure is expected to be issued by January 1990.
- 4. The leaking HFP bypass check valve (SGN-V431) was repaired.
HFP bypass check valves SGN-V431 and SGN-V432 were pressure tested and no leakage was identified. The overpressurization event was evaluated and Hain Feedwater Pump suction piping was walked down. It was determined to be acceptable for continued plant operation. Check valves SGN-V431 and SGN-V432 have been inspected in Units 1 and 3. APS is developing a program to periodically inspect selected check-valves. This program is expected to start during the next refueling outage for Units 1, 2 and 3.
1V. PREVIOUS SIHILAR EVENTS:
There have been no previous similar occurrences reported pursuant to 10CFR50.73.
There have been previous reactor trips reported. However, none of the previous reactor trips were attributable to the same root cause described in Section I. I. Therefore none of the previous corrective actions would have been expected to prevent this event.
V .. ADDITIONAL INFORMATION There have been 5 total accumulated actuation cycles of the Emergency Core Cooling System to date. This report satisfies the requirements of Technical Specification 3.5.2 ACTION b.
NRC Form 366A (689)
/i '0 J