ML17305A447
| ML17305A447 | |
| Person / Time | |
|---|---|
| Site: | Palo Verde  |
| Issue date: | 12/15/1989 |
| From: | James M. Levine, Shriver T ARIZONA PUBLIC SERVICE CO. (FORMERLY ARIZONA NUCLEAR |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| 192-00603-JML, 192-603-JML, LER-89-003-01, LER-89-3-1, NUDOCS 8912190264 | |
| Download: ML17305A447 (34) | |
Text
ACCELERATED D UTION DEMONSTP TION SYSTEM C
REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)
ACCESSION NBR.8912190264 DOC DATE. 89/12/15 NOTARIZED. NO DOCKET g FACIL:STN-50-529 Palo Verde Nuclear Station, Unit .2, Arizona Publi 05000529 AUTH. NAME AUTHOR AFFILIATION SHRIVER,T.D. Arizona Public Service Co. (formerly Arizona Nuclear Power LEVINE,J.M. Arizona Public Service Co. (formerly Arizona Nuclear Power RECIP.NAME RECIPIENT AFFILIATION
SUBJECT:
LER 89-003-01:on 890216,reactor generator level.
trip due to low steam 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 PD5 LA CHAN,T CODE/NAME 1,1 LTTR ENCL 1 1 ID PD5 PD CODE/NAME DAVIS,M.
LTTR ENCL 1
1 1
1 INTERNAL: ACRS MICHELSON 1 1 ACRS MOELLER. 2 2 ACRS WYLIE 1 1 AEOD/DOA 1 1 AEOD/DS P/TPAB
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1 1 AEOD/ROAB/DSP 2 2 DEDRO 1 . 1 NRR/DET/ECMB 9H 1 1 NRR/DET/EMEB9H3 1 1 NRR/DET/ESGB 8D 1 1 NRR/DLPQ/LHFB11 1 1 NRR/DLPQ/LPEB10 1 1 NRR/DOEA/OEAB11 1 1 NRR/DREP/PRPB11 2 2 NRR/DST/SELB 8D 1 1 NRR/DST/SICB 7E 1 1 NRR/DST/SPLB8D1 1 1 N /DST/NRXB 8E 1 1 NUDOCS-ABSTRACT 1 1 REG FILE 1 1 RES/DSIR/EIB 1 1 ~ 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 NSZC MURPHY,G.'"A 1 1 NUDOCS FULL TXT 1 1 NOTES: 1 1 NOTE TO ALL "RIDS" RECIPIENTS:
PLEASE HELP US TO REDUCE WASTEl CONTACT THE. DOCUMENT CONTROL DESk, ROOM Pl-37 (EXT. 20079) TO ELIMINATEYOUR NAME FROM DISTRIBUTION FULL TEXT CONVERSION REQU1RED TOTAL, NUMBER OF COPIES REQUIRED: LTTR 40 ENCL 40'
ll Arizona Public Service Company PALO VERDE NUCLEAR GENERATING STATION P.O. BOX 52034 ~ PHOENIX. ARIZONA85072-2034 192-00603-JML/TDS/RKR JAMES M. LEVINE December 15, 1989 VICE PRESIDENT NUCLEAR PRODUCTIDM U. S. Nuclear Regulatory Commission NRC Document Control Desk Washington, D.C. 20555
Dear Sirs:
Sub) ect: Palo Verde Nuclear 'Generating Station (PVNGS)
Unit 2 Docket No. STN 50-529 (License NPF-51)
Licensee Event Report'89-003-01 File: 89-020-404 Attached please find Supplement Number 1 to Licensee Event Report (LER) No.
89-003-00 prepared and submitted pursuant to the requirements of 10CFR 50.73.
This supplement is being provided to revise the corrective actions. A Human Performance Evaluation System (HPES) analysis is not being performed as previously stated. Based on the extensive corrective actions described in this report the HPES was determined not to be required. In accordance with 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 truly yours, JML/TDS/RKR/kj Attachment CC'.E. F. Conway E. Van Brunt (all w/a)
J. B. Martin T. J. Polich M. J. Davis A. C. Gehr INPO Records Center p~
g,S'~12190260 PDR e'er g 21, ADOCK O=LI8O5 S pbc
NRC FORM 366 U.S, NUCLEAR REGULATORY COMMISSION APPROVED OMB NO. 31504)104 (669)
EXPIRES: 4/30/92 ESTIMATED BURDEN PEA RESPONSE TO COMPLY WTH THIS INFORMATION COLLECTION REQUEST: 500 HRS. FORWARD LICENSEE EVENT REPORT (LER) COMMENTS REGARDING BURDEN ESTIMATE TO THE RECOADS BRANCH (P630), U.S. NUCLEAR AND REPORTS MANAGEMENT REGULATORY COMMISSION. WASHINGTON, DC 20555, AND TO THE PAPERWORK REDUCTION PROJECT 13)500104), OFFICE OF MANAGEMENTAND BUDGET. WASHINGTON, OC 20503.
DOCKET NUMBER (2) PAGE 3 FACILITY NAME (I)
Palo Verde Unit 2 o 5 o o o 52 9> oFl 5 TITLE (4)
Reactor Trip Due to Low Steam Generator Level EVENT DATE ISI LER NUMBER (5) REPORT DATE (7) OTHER FACILITIES INVOLVED (SI AS SEQUENTIAL DAY YEAR FACILITYNAMES DOCKET NUMBER(s)
MONTH DAY YEAR YEAR NUMBER yT NUMBER MONTH N/A 0 5 0 0 0 0 2 1 6 89 8 9 0 0 3 01 12 1 5 8 9 N/A 0 5 0 0 0 THIS REPORT IS SUBMITTED PURSUANT T O THE RLQVIREMENTS OF 10 CFR ()t /Check one oi mote of the follonenp/ (11 OP E RATING MODE (9) 1 20.402 (b I 20.405(c) 60.73(el(2) (ie) 73.71(b)
POWER 20.406( ~ l(1 )(il SOW (c) (1) 50.73(e)(2)(el 73.71(cl LEYEL OTHER /Specify tn Abtttecr 1 0 0 20.405( ~ ) (I ) (9) 5(L36(c) (21 50.73(el(2) (eit) below end ln TexL HRC Foim 20.405( ~ I (I l(iiil 60.73(e) (2)(i) 60.73(el(2)Niiil(A) 20.405(e) (I l(le) 50.73(el(2)(9) 50.73(e) (2)(eBII(B) Special Report 20.406 (e) Ill(e) 50.73(e ) (2) (iiil 50.73( ~ I(2)(xl 2-SR-89-003 LICENSEE CONTACT FOR THIS LER (12)
NAME TELEPHONE NUMBER AREA CODE Timothy D. Shriver, Compliance Hanager. 6 02 39 3 25 21 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE OESCAIBEO IN THIS REPORT (13)
MANUFAC REPORTABLE,, MANUFAC.
CAUSE SYSTFM COMPONENT COMPONENT TUBER TO NPRDS
~&A'YSTEM TVRER WRR K4~Ã<
SUPPLEMENTAL REPORT EXPECTED (14) MONTH OAY YEAR EXPECTS'0 SUBMISSION DATE RSI YES /lf yet, Cnmpfete EXPECTED St/SM/SS/DH DATE/ NO AK appr'oximatety 595 (fat onVebNruary 16, 1989, Palo Verde Unit 2 was in Hode 1 (POWER OPERATION) at approximately 100 percent power when a Feedwater Control System (FWCS) malfunction resulted in a reactor trip due to low level in Steam Generator (SG) number 1.
Immediately prior to the trip the Control Room staff observed both SG levels decreasing. Both master controller outputs were cycling full scale. SG number 1 economizer valve was fully shut and SG number 2 economizer. valve was 10 percent open. A Control Room Operator placed SG number 1 economizer valve in manual and inserted approximately 17 percent open demand when the reactor tripped.
Immediately following the trip an Auxiliary Feedwater Actuation Signal was initiated. At approximately 0345 HST a Safety Injection Actuation Signal (SIAS)/Containment Isolation Actuation Signal (CIAS) was generated due to the overcooling of the Reactor Coolant System. SG number 1 level continued to increase and at approximately 0347 HST a Hain Steam Isolation Signal -was received and terminated the cooldown. The cause of the event was a small amount of debris in the restrictor on the vertical relay within the economizer valve pneumatic positioner. Immediate corrective action taken was to replace the pneumatic relays in both FWCS's.
This submittal also provides Special Report 2-SR-89-003 in accordance with Technical Specification 3.5.2 ACTION b.
NRC Form 366 (669)
0 NRC FORM QBA US. NUCLEAR REGULATORY COMMISSION APPROVED OMB NO. 31500(04 (64)9) ~
EXPIRES: 4/30I92 ESTIMATED BURDEN PER RESPONSE TO COMPLY WTH THIS LICENSEE EVENT REPORT ILERI INFORMATION COLLECTION REOUEST: 50.0 HRS. FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE RECORDS TEXT CONTINUATION AND REPORTS MANAGEMENT BRANCH (P 530). U.S. NUCLFAR REGULATORY COMMISSION, WASHINGTON, OC 20555, AND TO THE PAPERWORK REDUCTION PROJECT (3(504))04). OFFICE OF MANAGEMENTAND BUDGET, WASHINGTON. DC 20503.
FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER (6) PAGE (3)
SII'O SEOUENTIAL REVISION NUMSER 'Nl NUMSEVI Pal o Verde Uni t 2 o s o o o 52 98 9 003 01 02 oF 1 5 TEXT (IImove opooe b veevvded, Ivse Rddo'onol HRC Fovvtv 35SI('sl (12)
I. DESCRIPTION OF WHAT OCCURRED:
A. Initial Conditions:
At approximately 0345 MST on February 16, 1989, Palo Verde Unit 2 was in Mode 1 (POWER OPERATION) at approximately 100 percent power when a Feedwater Control System (FWCS)(JB) malfunction resulted in a reactor (RCT)(AC) trip.
B. Reportable Event Description (Including Dates and Approximate Times of Major Occurrences):
Event Classification: Any event or condition that resulted in manual or automatic actuation of any Engineered Safety Feature (ESF)(JE),
including the Reactor Protection System (RPS)(JC).
At approximately 0345 HST on February 16,, 1989, Palo Verde Unit 2 was operating in Mode 1 (POWER OPERATION) at approximately 100 percent power when an FWCS malfunction resulted in a reactor trip on Low Steam Generator (SG)(AB) number 1 level and an Auxiliary Feedwater Actuation Signal (AFAS)(BA) actuation. Subsequent to the trip, number 1 SG was overfed using main feedwater resulting in a Safety Injection Actuation Signal (SIAS)(JE), Containment Isolation Actuation Signal (CIAS)(JE), and Main Steam 'Isolation Signal (HSIS)(JE). A Notification of Unusual Event (NUE) was declared at 0352 MST due to SIAS actuation and was terminated at 0449 MST when the SIAS was reset.
Prior to the event, on February 13, 1989 during Hain Turbine (TRB)(TA) Stop valve (V)(TA) testing, an FWCS transient occurred.
The number 1 economizer valve (FCV)(SJ) started to close in response to the testing but subsequently appeared to stick at approximately 35 percent open. The feedwater pump (FWP)(P)(SJ) speed increased to restore SG level. Manual control of the RB" feedpump was used to gain control of feedwater flow. The SG level fell to a low of approximately 5 percent narrow range (NR). Once SG level was regained (approximately 55 percent NR), the operator returned feedpump speed to automatic control and took manual control of the economizer valve. The valve subsequently responded properly. The secondary operator (utility, licensed) manually stroked the valve approximately 10 to 15 percent open and closed from the 70 percent open position and then restored the valve to its normal position of approximately 85 percent open for 100 percent power operation. With all components returned to automatic, all systems functioned properly. During the transient, oscillations from 50 to 85 percent in the master controller output NRC Fovvn 355A (54)9)
0 0 l
NRC FORM 366A US. NUCLEAR REGULATORY COMMISSION (669) ~
APPROVED 0MB NO. 31500104 EXPIRES: 4/30/92 ESTIMATED BURDEN PER RESPONSE TO COMPLY WTH THIS LICENSEE EVENT REPORT'(LER) INFORMATION COLLECTION REQUESTI 50A) HRS. FORWARD-COMMENTS REGARDING BURDEN ESTIMATE TO THE RECORDS TEXT CONTINUATION AND REPORTS MANAGEMENT BRANCH (P430), U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20555, AND TO THE PAPERWORK REDUCTION PROJECT (3)504)104>, OFFICE OF MANAGEMENTAND BUDGET, WASHINGTON, DC 20503.
FACILITYNAME (1) DOCKET NUMBER (2) LER NUMBER (6) PAGE (3)
NN SEOOENTIAL N(O REVISION NUM ER NUMSER Palo Verde Unit 2 o s o o o 5 2 9 8 9 0 03 0 1 03o" 1 5 TEXT /I/mort 4/Not 6I rttoirtd. Vtt odd)dont//VRC Form 3664'4/ (17) for FWCS number 1 were observed. These oscillations in the master controller would be expected for the transient induced in the system by the erratic operation of the number 1 economizer valve.
The opportunity to learn from the transient on February 13, 1989 was reduced because the Temporary Data Acquisition System (TDAS)(Ig) data was not reviewed. The Shift Technical Advisor (STA)(utility, non-licensed) did discuss the transient with the Shift Supervisor (utility, licensed), but he inferred from that conversation that there would be no need for the transient data and proceeded to reinitialize the data disk. This reinitialization is done every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. When the data disk was reinitialized, the specific behavior of the number 1 economizer valve in the February 13, 1989 event was lost. Had this information been retained it may have aided in the troubleshooting. It was also noted that the STA was not called immediately when the transient occurred.
After several discussions involving the Plant Hanager (utility, non-licensed), System Engineer (utility, non-licensed), and Operations Standard Advisor (utility non-licensed), a work request was generated to troubleshoot FWCS number 1. Troubleshooting the FWCS was made based on the following considerations:
- 1) There were no apparent problems maintaining SG level.
- 2) There were no apparent problems with the master controller in automatic due to the observed stable FWCS conditions since the February 13, 1989 transient.
- 3) The decision was made to instrument both FWCSs and initiate small perturbations by controller setpoint changes, stop valve testing or initiating rapid steam generator blowdowns during the troubleshooting to observe system response.
Additionally, the decision was made not to perform any high risk Preventive Haintenance Tasks or initiate the scheduled high rate SG blowdown until the system was instrumented.
A work order was generated to perform the troubleshooting (measure the input and output of the economizer valve signal characterizer module, SCH) based on guidance by the System Engineer. The System Engineer would personally direct the troubleshooting work order which would also require Shift Supervisor concurrence. It would also require that the associated parameter (i.e. the economizer valve) be placed in manual when connecting the recorder to the SCH.
Through discussions with the System Engineer and the Instrument and Control (I&C) Supervisor (utility, non-licensed), it was decided to NRC ForRI SeeA (669)
0 NRC FORM 366A ILS. NUCLEAR REGULATORY COMMISSION APPROVED 0MB NO. 3)500)04 (666)
EXPIRES: 4/30/02 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 (P4I30). U.S. NUCI.EAR REGULATORY COMMISSION. WASHINGTON. DC 20555, AND TO THE PAPERWORK REDUCTION PROJECT (3)500104). OFFICE OF MANAGEMENTAND BUDGET, WASHINGTON, DC 20503.
FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER (6)
SEQUENTIAL Mr REVOK>N CC% NUMEE4 NUMSE4 Palo Verde Unit 2 o s o o o 52 9 8 9 00 3 01 0 4 OF 1 5 TEXT /I/movp a>>EP IJ Tuuut)ud, u44 Pddr'0'phu/IVRC Foun 366AB/(17) utilize a standard four-channel Gould recorder for troubleshooting. However, due to mechanical problems with the recorder, it could not be used. After further discussions between I&C personnel and the System Engineer it was determined to use an eight-channel digital recorder.
The digital system recorder was obtained from Heasuring and Test Equipment (HKTE) with prefabricated cables which included triaxial connectors, coaxial wire, coaxial signal wire connector, and individual wires for connection to the system.
At approximately 1830 HST on February 15, 1989, the eight-pen Digital Recorder System was connected to the number 1 FWCS. The digital recorder has an internal circuit wherein the inner shields of the triax connectors are connected together. This is different than the standard four-channel recorder and was not recognized by the System Engineer present or the ILC Technician (utility, non-licensed). This resulted in the SCH for SG number 1 economizer valve being shorted from input to output through the recorder.
With a reasonably stable SG level, the internally cross-connected recorder had very little impact on the master controller output (i.e. the resulting program was very close to the required program). The System Engineer was present for the installation of the recorders as required by the work order.
The recorder was then connected to FWCS number 2. This resulted in cross-connecting the inputs and outputs of both SCHs for SGs number 1 and 2. The effect of tying the two FWCSs together was not observed until the SG 2 economizer valve was returned to automatic control. The Control Room Operator observed SG 2 level increasing abnormally and, at about 70 percent NR, placed the economizer valve in manual. By manually decreasing the controller output .he was able to turn the level rise and return level to its normal range (55 percent NR). The recorder was then removed from FWCS number 2 and the economizer valve controller was returned to automatic.
Proper automatic SG level control of both FWCSs was observed.
There was no abnormal response on number 1 FWCS.
The Shift Supervisor called the Operations Hanager (utility, licensed) and described the event. Based on the fact that the recorder had apparently functioned properly on number 1 FWCS the Operations Hanager directed the Shift Supervisor to consult the Lead System Engineer (utility, non-licensed) and to leave the recorder installed in SG1 FWCS if the Lead System Engineer concurred. The recorder had been .connected to FWCS number 1 for about 9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> prior to the event and the FWCSs appeared to be functioning normally in the automatic mode during this time.
NRC Form 366A (64)B)
0 4I
NRC FORM 3SSA US, NUCLEAR REGULATORY COMMISSION (SSB) APPROVED OMB NO.'3(500(04
~
EXPIRES: 4/30/92 ESTIMATED BURDEN PER RESPONSE TO COMPLY WTH THIS LICENSEE EVENT REPORT (LER) INFORMATION COLLECTION, REQUESTI SOA) HRS. FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE RECORDS TEXT CONTINUATION ANO REPORTS MANAGEMENT BRANCH IF@30), U.L NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20555, AND TO THE PAPERWORK REDUCTION PROJECT (31500)04). OFFICE OF MANAGEMENT AND BUDGET, WASHINGTON, DC 20503.
FACILITY NAME (I) DOCKET NUMBER (2) LER NUMBER LS) PAGE (3)
SEOUENTIAI. :D~r: REVISION NUMBER sC.. NUMBER Palo Verde Unit 2 o s o o o 5 2 9 8 9 0 0 3 01 05 1 5 TEXT /I/mt'Pttt it tPqvPtd, IItttddic/Imt/HRC Fomt JSSA'4/ (17)
Prior to the reactor trip on February 16, 1989 at approximately 0345 HST, the Control Room received several alarms (ALH)(IB) including condensate pump (P)(SD) strainer (STR)(SD) Hi DP alarms and FWCS Trouble alarms. The condensate pump strainer Hi DP alarms were concurrent with the perturbation in the feedwater system when the number 1 economizer valve closed. The Secondary Operator (utility, licensed), Shift Supervisor (utility, licensed), and Primary Operator (utility licensed) positioned themselves at Hain Control Boards (CBD) B05 and B06 to evaluate the situation. The Secondary Operator, Shift Supervisor, and Primary Operator observed both SG levels decreasing rapidly with level in SG number 1 decreasing below NR indication. Both master controller outputs were observed to be cycling full scale with one to two second intervals. The feed pumps and number 2 SG economizer control valve followed the oscillations of the number 1 SG FWCS but at a slower rate due to FWCS lead/lag circuits and the physical abilities of the mechanical devices to respond to electronic signals.
Immediately prior to reactor trip, Control Room personnel observed continued decreasing levels in SG number 1, the SG number 1 economizer valve was fully closed, and the SG number 2 economizer valve was 10 percent open. The Secondary Operator observed that the SG number 1 economizer control valve manual/auto controller demand signal was zero and prepared to open the number 1 SG economizer control valve manually in an attempt to restore SG number 1 level. The Secondary Operator took manual control of the SG number 1 economizer valve, and opened the valve to mitigate the underfeed situation that was in progress.
While the Secondary Operator was attempting to manually open the SG number 1 economizer valve, the reactor tripped. This occurred 27 seconds after the initial secondary disturbance was noted. At the time of the trip, the Secondary Operator had manually inserted an approximately 17 percent open demand signal to the number 1 SG economizer control valve. Three seconds after the trip, TDAS indicated that the SG number 1 economizer control valve was 17 percent open. The CRS diagnosed the initial event correctly and the appropriate recovery procedure 42R0-2ZZ05, Loss of Feedwater, was implemented.
At approximately 0345 HST approximately 14 seconds after the reactor trip, an AFAS for SG number 1 was generated due to low low SG number 1 level. The AFAS 1 was a result of SG level "shrink" from the reactor trip and from the excessive main feedwater flow.
The AFAS signal was generated as designed and the Auxiliary Feedwater system performed its intended function.
N R C Fotm 35SA (54)9)
!5 0 1t
NRC FORM ESSA U.S. NUCLEAR REGULATORY COMMISSION (54)9) APPROV ED OMS NO. 31500104 E X P I R ES: 4/30/92 ESTIMATED BURDEN PER RESPONSE TO COMPLY WTH THIS LICENSEE EVENT REPORT (LER) INFORMATION COLLECTION REOUESTI 50.0 HRS. FORWARD COMMENTS REGARDING SVADEN ESTIMATE TO THE RECORDS TEXT CONTINUATION AND REPORTS MANAGEMENT BRANCH <P4)30). V.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20555, AND TO THE PAPERWORK REDUCTION PROJECT (31500104). OFFICE OF MANAGEMENT AND SU DG ET, WASHINGTON, DC 20503.
FACILITY NAME (1) DOCKET NUMBER (21 LER NUMBER (5) PAGE (3)
YEAR, (:4( SEOUENTIAL o(r/ REVISION NUMBER '?..?S NVMSER Palo Verde Unit 2 o s o o o 52 9 8 9 0,.0 3 0 1 0 6 OF 1 5 TEXT ///moro 44444 /4 roorrkod, V44 odd?)1444////IC For??I 3ÃA'4/ (Il)
The Control Room Supervisor (CRS)(utility,, licensed) directed the Control Room Operators to monitor plant safety functions. The Secondary Operator began his post trip safety function monitoring actions as required. The Secondary Operator .verified proper response of the Auxiliary Feedwater System.
During the monitoring of the plant safety functions, the Secondary Operator did not take actions to either restore the economizer control valve controller to auto or to manually close the valve to prevent the cooldown. The Shift Supervisor noted that the level in SG number 1 was increasing but was unaware that the SG number 1 economizer valve was not closed. When SG number 1 level was at approximately 65 percent WR, the Shift Supervisor directed the Secondary Operator to throttle auxiliary feedwater flow to decrease steam generator feedwater flow.
Following the reactor trip, number 1 SG level continued to increase due to the number 1 SG economizer valve being'7 percent open. The number 1 economizer valve 'being in manual defeated the Reactor Trip Override (RTO) automatic controls for the SG number 1 economizer control valve and the valve remained open. Normally fol.lowing a reactor trip,, an RTO of the FWCS occurs to provide initial control of the SG level and limit the Reactor Coolant System (RCS)(AB) cooldown. The RTO logic (which is a non-safety related system) is designed to cl'ose the economizer valves, set the Hain Feedwater Pump Turbines (HFWPTs) to minimum speed, and control downcomer valves to maintain SG level. When SG level increases above the RTO reset level, FWCS control is transferred to single element control for maintaining SG level.
As a result of the economizer valve in manual, the RTO logic was defeated and excessive feedwater occurred. The resulting overfeeding of the SG caused a cooldown of the primary system. At approximately 0345 HST a SIAS/CIAS, was gener ated due to overcooling of the primary system.
The SIAS/CIAS setpoints are selected to assure adequate makeup of RCS coolant in the event of a loss of RCS inventory. The volumetric decrease in the RCS liquid, due to the cool'ing, results in a pressure reduction similar in nature to a loss of inventory.
The'High Pressure Safety Injection (HPSI)(Bg) pumps responded as required. The design response for SIAS at high pressure is for the HPSI pumps to inject .as they did in this event. The HPSI pumps started due to the SIAS and responded properly to inject water into the RCS to restore RCS pressure and pressurizer level. The Safety Injection (SI) system, as well as charging pumps, repressurized the RCS. The SIAS actuat'ion was verified to have occurred per design during the resetting of the SIAS actuation. The CIAS-also was NRC Form 3SSA (54)9)
4l IIRC FORM366A U.S. NUCLEAR REGULATORY COMMISSION APPROVED OMB NO. 31500104 (669)
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FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER (6) PAGE (3) yEAR QR@ SEOVENTIAL g~'> REVISION NVMSSR NVMSSR Palo Verde Unit 2 o s o o o 5 2 9 89 0 03 01 07 OF 1 5 TEXT (IImsm FAPCS JI TPSSPsd, IIJp Pddicensl iVRC %%dmI 36643) ((2) verified to have provided containment isolation per design.
With the number 1 SG economizer valve still in manual, overriding the RTO trip logic, the SG continued to 0347 HST, an HSIS was received at 91 percent NR in SG number 1.
fill: At approximately The HSIS isolated main feedwater which terminated the RCS cooldown.
At approximately 0352 HST on February 16, 1989, the Shift Supervisor declared a Notification of Unusual Event (NUE). The NUE was declared pursuant to EPIP-02 (Emergency Classification) as a result of the initiation of a SIAS on low pressurizer pressure. At approximately 0400 HST on February 16, 1989 the appropriate state and local agencies were notified via the Notification and Alert Network (NAN). The Nuclear Regulatory Commission (NRC) Operations Center was notified at approximately 0444 HST on February 16, 1989. The actions of EPIP-02 were performed in a timely manner.
Stable conditions were achieved and the NUE was terminated at 0449 HST on February 16, 1989.. The NUE was reported in Special Report 2-SR-89-002.
The STA was notified and responded to the control room. The CRS did not direct the STA to perform Appendix BB per 42EP-2ZZ01, Emergency Operation. This was contrary to an approved procedure.
Consequently, Appendix BB was not performed. Normally the Duty STA will initiate Appendix BB following a reactor trip without CRS direction.
The control room staff stabilized the plant and directed efforts towards resetting ESFAS actuations per the appropriate procedures.
Following plant stabilization, the STA completed an Event Notification Worksheet which was reviewed by the Shift Supervisor .
A notification to the NRC via the Emergency Notification System (ENS) phone was made by the STA. Initial observations by the Control Room staff indicated that there was a problem with the actuation of the RAR Essential Chiller following the SIAS. The initial notification via the ENS stated that all ESF equipment operated correctly. Based on the information available at the time of the call, the information relayed to the NRC by the STA was not accurate (i.e. that the chiller did not appear to automatically start as designed). A follow-up call was made by Compl.iance (utility, non-licensed) to the NRC via the 'ENS phone stating that "the RAH Essential Chiller did not start automatically. However, a control room operator manually started the RAR Essential Chiller successfully." Subsequent evaluation has determined that the RAR Essential Chiller did function as designed in that the automatic start was concurrent with the operator's attempt to manually start the RAH Essential Chiller.
NRC F oIRI 366A (669)
NRC FORM 266A U.S. NUCLEAR REGULATORY COMMISSION (64)9) APPROVED 0M B NO. 31500104 EXPIRES: 4/30/92 ESTIMATED BURDEN PER RESPONSE TO COMPLY WTH THIS LICENSEE EVENT REPORT ILER) INFORMATION COLLECTION REOUEST: 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 (3)504)104). OFFICE OF MANAGEMENTAND BUDGET. WASHINGTON, DC 20503.
FACILITY NAME (1) DOCKET NUMBER (2) LER NVNIBER (6) PAGE (3)
Iso sEGUENTIAL NVMSER ~ REvlsloN gj2 NUMBER Palo Verde Unit TEXT /// moro Ersso /s 2
rrrr((rd, osr rdd4/oor/ HRC %%dnn 36643/ (I2) o s o o o 5 2 9 89 00 3 0 1 0 8 >>15 During the course of the event, the RCP Vibration readings required to be taken at 0400 were missed. A review was made of the on-line Bently-Nevada vibration monitoring data during the time frame that the readings were missed and no abnormalities were noted in RCP vibration. The control room staff's work load at the time of the missed readings was the major cause of the deficiency.
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 that contributed to the event.
D. Cause of each component or system failure, if known:
Not applicable - there were no component or system failures.
E. Failure known:
mode, mechanism, and effect of each failed component, if Not applicable - no failed components were involved.
F. For failures of components with multiple functions, list of systems or secondary functions that were also affected:
Not applicable - no fai-led components were involved.
For failures that rendered a train of a safety system inoperable, estimated time elapsed from the discovery of the failure until the train was returned to service:
Not applicable - no safety systems were rendered inoperable.
H. Method of discovery of each component or system failure or procedural error:
The Primary Operator was unsuccessful in shutting down the Essential Chiller on the first attempt because of procedural inadequacy in Emergency Procedure 42EP-21ZOI. The procedure did not provide sufficient guidance to ensure that all required actions were taken prior to securing the essential chillers. In this case, the diesel generators must be shutdown prior to securing the essential chiller.
NRC Form 366A (64)9)
II 0 NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION APPROVEO 0MB NO. 31500104 (64)9)
EXPIRES: 4/30/92 ESTIMATED BURDEN PER RESPONSE TO COMPLY WTH THIS LICENSEE EVENT REPORT (LER) INFORMATION COLLECTION REOUESTt 50.0 HRS. FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE RECORDS TEXT CONTINUATION AND REPORTS MANAGEMENT BRANCH IP.530). U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON. OC 20555, AND TO THE PAPERWORK REDUCTION PROJECT (31504>104>. OFFICE OF MANAGEMENTAND BUDGET, WASHINGTON, DC 20503.
FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER (6I PAGE (3)
YEAR 4/4 SEOVENTIAL ?r,ij t>EVISION NVM66>I 4m NVM 64 Palo Verde Unit 2 o s o o o 5 2 9 8 9 0 0 3 01 09 OF 1 5 TEXT ii/ moto tpooo it roOuirod, oto oddiobno////IC Form 3654'4/ (17)
I. Cause of Event:
Reactor Trip The cause of the reactor trip and AFAS was a malfunction of the number 1 steam generator economizer valve. Debris in the pneumatic positioner of the number 1 SG economizer valve initiated the erratic behavior of the FWCS. This was also the cause of the transient encountered on February 13, 1989. A small amount of debris, approximately 10 mils in diameter (by microscopic examination), was present in the restrictor on the vertical relay within the Fisher Pneumatic Positioner Model 3570 on the SG 81 economizer valve. This would have prevented the valve from operating properly. The manufacturer concurs .in this evaluation.
The number 1 economizer valve exhibited erratic pneumatic relay operation during testing conducted after the event. 18C personnel observed that the valve initially opened nominally, but as the open demand incr eased the valve slowly drifted closed by itself (i .e.
the valve closed fully with a 75 percent open demand).
The debris was evaluated in an attempt to determine its source.
The debris appearance was not that of "desiccant" but looked to be representative of a metal particle. The overall instrument air system is currently under monitoring activities associated with generic letter 88-14 and has been demonstrated to meet ANSI standards for instrument air system. Further analysis of the debris can not be performed as the particle was lost due, presumably, to normal air movements during the examination process.
Additionally, the inlet filter on the pneumatic regulator for the number 1 SG economizer valve was examined and no indications of contaminants were identified. This regulator is on a branch line immediately before the number 1 SG economizer valve positioner.
Other investigations of the FWCSs (calibration checks, physical inspection, and testing) found no deficiencies. Testing was also performed on the FWCSs Reactor Trip Override (RTO)(JB) circuit which verified proper operation.
The event was compounded by improper installation of the recorder leads due to a personnel error on the part of the II)C Technician (utility, non-licensed) and the System Engineer (utility, non-licensed). The leads cross-connected the input to the output of the SCH as described in Section I.B. The improper installation of the recorder equipment compounded the February 16, 1989 transient by rendering the number 1 FWCS master controller ouput inoperable during a portion of the event. During subsequent NR C Form 366A (669)
0 II NRC FORM 366A U S." NUCLEAR REGULATORY COMMISSION APPROVEO 0MB NO. 31504)104 (64)9)
E XP IR E 5: 4/30/92 ESTIMATED BURDEN PER RESPONSE TO COMPLY WTH THIS LICENSEE EVENT REPORT ILER) INFORMATION COLLECTION REQUEST'00 HRS. FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE RECORDS TEXT CONTINUATION ANO RFPORTS MANAGEMENT BRANCH (F430). V.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20555, AND TO THE PAPERWORK REDUCTION PROJECT 131504)104). OFFICE OF MANAGEMENTAND BUDGET,WASHINGTON. DC 20503.
FACILITY NAME (11 DOCKET NUMBER (2) LER NUMBER (6) PAGE (3)
SR+ SEQUENTIAL 366 REVISION O?~6 NUMSER @>6 NUMSER Palo Verde Unit 2 o s o o o 5 2 9 8 9 0 0 3 01 10oF TEXT /ifmoro g>>or is rtqrrdtd. Irto Ed'/orm/IYRC Form 36SE'4/ (17) testing with the digital recorder connected across the economizer valve SCH (i.e., input to output shorted) and with representative input values, it was found that the master control output would "clamp" to zero. The "clamping" would remain in place until an interruption of either the input or output of the master control occurred.
The number 1 economizer valve slowly closing caused both FWCS master controller outputs to oscillate in an attempt to control level. During this period, the number 1 FWCS master controller output reached an output value low enough to cause the recorder to "clamp" the output to zero. Since the number 1 FWCS master controller was "clamped", the number 1 economizer valve received a zero signal and number 1 HFWP was controlled by the number 2 FWCS master controller. An auctioneer circuit for the HFWP selects the highest signal for either the number 1 or 2 FWCS master controller (except during RTO).
SG number 2 feedwater flow increased (due to increased feed pressure) as number 1 economizer valve .closed. FWCS number 2 responded by reducing both feedwater pump speeds and repositioning number 2 economizer valve. The number 2 FWCS master controller responded properly to this event by controlling pump speed and number 2 economizer valve position to maintain SG number 2 level.
CIAS/CIAS and HSIS The cause of the SIAS/CIAS and HSIS was an overfeeding to the number 1 SG due to the number 1 economizer valve being left in manual at 17 percent open and the excessive feedwater header pressure due the "8" HFWP being in manual at a high speed setting (5450 rpm vice approximately 3800 rpm for this condition). The R8" HFWPT controller was transferred to manual just prior to the reactor trip. Hembers of both operating crews, the offgoing and the oncoming, were interviewed in an attempt to determine when was placed in manual. None of the individuals interviewed
'8'FWP could recall placing the pump speed controller in manual.
PVNGS determined that the pump went to manual 8 seconds prior to the reactor trip. As indicated by the TDAS plots, the pump speed was tracking feedwater control system demand until 8 seconds prior to the trip.
Several different items were investigated to see if there was any other mechanism that could have caused the HFWP to shift to manual. The only mechanism identified would be a momentary loss of power on the shift during the fast bus transfer. It is a loss of power did not occur based on the fact that the same concluded'hat NRC Form 366A (Er69)
0 NRC FORM 566A U.S. NUCLEAR REGULATORY COMMISSION APPROVED OMB NO. 31500104 (669)
EXPIRES: 4/30/92 FSTIMATED'BURDEN PER RESPONSE TO COMPLY WTH THIS LICENSEE EVENT REPORT INFORMATION COLLECTION REQUEST: 500 HRS. FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE RECORDS ILER)'EXT CONTINUATION AND REPORTS MANAGEMENT BRANCH IP430), U.S. NUCLEAR RFGULATORY COMMISSION, WASHINGTON, DC 20555. AND TO THE PAPERWORK REDUCTION PROJECT (31504)(04), OFFICE OF MANAGEMENTAND BUDGET. WASHINGTON. DC 20503.
FACILITY NAME ()) DOCKET NUMBER (2) LER NUMBER (6) PAGE (3)
(OP SEQI/ENTIAL Pig AEVISlON i?4 NUMBER NUMFFR Pal o Verde Uni t 2 o s o o o 5 2 9 89 00 3 01 11 >> 1 5 TEXT ///ovvo g>>co/4 /POvt)od. v>> odde'ono/P//IC Form 36643/ (IT) power supply is shared by both feed pumps and the 'A'FWP stayed in automatic. In addition, the manual/auto controller will not revert to manual unless power is lost for greater than 1 second.
PVNGS has concluded that the manual pushbutton was inadvertently depressed at the time indicated.
In response to the initial transient, both feedwater pumps followed the demand signal from FWCS number 2. During RTO, the control for the HFWP is shifted to its applicable FWCS master controller versus going through the auctioneering circuit. After the trip, the speed decreased to 3670 rpm as designed due to the RTO logic.
'A'FWP However, the 'B'FWP remained at high speed as a result of being in manual.
Both FWCSs entered the RTO mode. FWCS number 1 master controller was still "clamped" low at this time. FWCS number 2 master controller functioned as required; however, the 'B'FWP was in manual and remained at high speed as designed. The number 2 downcomer valve immediately responded to the RTO by opening due to the average RCS temperature being above the setpoint and subsequently responded properly thereafter. The economizer valve went closed as designed.
FWCS number 1 master controller output was "clamped" at the onset of the transient but became Runclamped" as the FWCS number 1 came out of RTO when SG number 1 level increased to 52 percent prior to the HSIS. At that time, resetting RTO caused the master controller output to control the number 1 downcomer valve as, designed.
Transient information was provided to I&C Engineering for an independent assessment. Additionally, the vendor design engineer (contractor, non-licensed) for the FWCS and a vendor system design engineer (contractor, non-licensed) were involved in the analysis at PVNGS. The vendor of the FWCS electronic modules was also consulted.
The economizer valve left approximately 17 percent open and the controller in manual added to the severity of the cooldown 'B'FWP transient and subsequently caused an HSIS actuation to occur which further complicated recovery operations. This was a cognitive personnel error in that the control room personnel did not recognize actual plant conditions and was contrary to procedural guidance.
There were no unusual characteristics of the work location (i.e.
Control Room) except the addition of various alarms which annunciated and the rapid pace of events which occurred. However, simulator training and emergency procedures are adequate to provide NR C Form 366A (689)
!5 4l
!IRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION APPROVED OMB NO, 31500104 (669) ~
EXPIRES; 4/30/92 ESTIMATED BURDEN PER'RESPONSE TO COMPLY WTH THIS LICENSEE EVENT REPORT (L'ER) INFORMATION COLLECTION REQUEST: 500 HRS. FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE TEXT CONTINUATION RECORDS'ND REPORTS MANAGEMENT BRANCH (P430). U.S. NUCLEAR REGULATORY COMMISSION; WASHINGTON, DC 20555, AND TO THE PAPERWORK REDUCTION PROJECT (315001041. OFFICE OF MANAGEMENTAND BUDGET. WASHINGTON. DC 20503.
FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER (6) PAGE (3)
(NN SEQUENTIAL
REVISION MPZ NUMBER A< NUMSER Palo Verde Unit 2 o so o o 52 9 89 , 00 3 0 1 12 oF1 5 TEXT /I/ more Spree ir rEII/'rrd. err rdd/r/orM/ HRC FINm 3664 4/ (12) operators with experience to compensate for these conditions.
During recovery operations an auxiliary feed flow indicator indicated less than the expected flow rate. Further testing determined that the auxiliary feedwater- flow instrumentation functioned properly through the performance of the loop calibration and the performance of a pump run test. The auxiliary feed flow indicator (AFB-FI-41B) was calibrated, minor problems were noted in the as found calibration.
The auxiliary feedwater flow loop converts a delta-p signal into a flow signal through the use of a Square-root extractor. With the existing plant configuration, the Square-root extractor does not allow the instrument to function below approximately 10 percent of the range of the instrument (0-2000 gpm) or below 200 gpm. It is concluded that the reason for the apparent lack of flow indication for 2JAFB-FT41B was that the flow was below the threshold of the Square-root extractor.
J. Safety System Response:
The following safety systems actuated automatically as a result of the event.
- 1) Emergency Diesel Generators (DG) (EK) Train RAR and RB",
- 2) Essential Spray Pond Systems (BS) Train RA" and RB",
- 3) Essential Chillers (KH) Train RAR and "B",
- 4) Essential Cooling Water (BI) Train RAR and RB",
- 6) Essential Auxiliary Feedwater (BA) Trains "A" and RB",
- 7) HPSI Trains RAR and 'B",
- 8) Low Pressure Safety Injection (BP). Trains RAR and RB",
- 9) Containment Spray (BE) Trains RAR,and RB", and
- 10) HSIS K. Failed Component Information:
Not applicable - no failed components were involved.
II. ASSESSHENT OF THE SAFETY CONSEQUENCES AND IHPLICATIONS OF THIS EVENT:
The economizer valve was inadvertently left in manual and open approximately 17 percent which contributed to a 47 degree cooldown of the primary system. This exceeded the cooldown limit of Technical Specification 3.4.8. 1, and an engineering evaluation was performed in accordance with the ACTION requirement. The evaluation determined that there were no adverse effects on the structural integrity, of the RCS and the RCS remains acceptable for continued operation. A review of the FSAR shows that during the cooldown event of February 16, 1989, the NRC Form 366A (669)
NAG FORM 366A US. NUCLEAR REGULATORY COMMISSION (64)9) APPROVED OMB NO. 31500104 EXPIRES: 4/30/92 ESTIMATED 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 (P430). 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 (1) DOCKET NUMBER (2l LER NUMBER (6) PAGE (3)
SEQUENWAL REVISION NUMSER NUMSER Palo Verde Unit 2 o s o o o 52 9 8 9 0 0 3 01 1 3 1 5 TEXT (I/moro Eooco (I nqvied. o44 oddroooo) iVRC Fomr 3654'4) (17) plant was bounded by the analysis conducted for the Hain Steam Line Break event.
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 the health or safety of the public.
I I I. CORRECTIVE ACTIONS:
A. Immediate:
The pneumatic relays were replaced in both Feedwater Control Systems.
B. Action to Prevent Recurrence:
Controls have been adopted requiring the labeling of test equipment leads, e.g. labeling the leads as (+) and (-) rather than relying on the assumption of polarity based on work practices which vary among disciplines.
Guidance has been provided to the Shift Supervisors emphasizing the importance of contacting the STA for any transient.
The STA section has developed guidance for STAs with respect to TDAS reinitialization so that transient data is retained for analysis. This will ensure that TDAS data is preserved when required.
The existing work control troubleshooting procedures for work on components critical for power operation have been revised to identify critical components requiring higher management review before troubleshooting activities can begin. In addition the procedure has been revised to require the use of the Equipment Instruction Hanuals (EIH) prior to using unfamiliar electronic test equipment or using testing equipment in unfamiliar applications to insure proper application, use, and to identify any peculiarities associated with the equipment operation.
The digital recorders and isolation amplifiers have been hard wired together eliminating the accessible connections on the recorder.
The only accessible input connections are on the isolation amplifier.
An investigation, determined that there is a problem with the NRC Form 366A (689)
0 NRC FORM 266A U.S. NUCLEAR REGULATORY COMMISSION APPROVED 0M B NO. 31500104 (64)0)
EXPIRES: 4/30/92 ESTIMATFO BURDEN PER RESPONSE TO COMPLY WTH THIS LICENSEE EVENT REPORT (LER), INFORMATION COLLECTION REOUEST( SOA) HRS. FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE RECORDS TEXT CONTINUATION AND RFPORTS MANAGEMENT BRANCH (F430), U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON. OC 20555, AND TO THE PAPERWORK REDUCTION PRO)ECT (31500104). OFFICE OF MANAGFMENT AND BUDGET. WASHINGTON, DC 20503.
FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER (6) PAGE (3)
SEQUENTIAL ?AW REVISION NVMSSII NVM FII Palo Verde Unit 2 o s o o o 52 989 0 03 01 14 OF 1 5 TEXT /llmare <<woe /4 rer)vtred, vee eddro'oael/VRC Form 3SI/A3/ (17) availability of Equipment Instruction Hanuals for test equipment.
A task force has been formed to review this problem. The task force is expected to complete its review and provide recommended correcti,ve actions by January 2, 1990. These recommendations are expected to be implemented by Harch 30, 1990.
Control Room personnel in Units 1, 2, and 3 have reviewed a summary of this event with emphasis on team skills and implications of manual operations following reactor trip events.
An Operation Plant Guideline has been issued regarding when it is acceptable, desirable, and necessary to transfer system controllers to manual. This guideline is applicable to all major control systems.
The crew involved was removed from shift to perform the following:
Perform a self critique Participate in the Post Trip Review Report (PTRR) investigation Assist in the determination of corrective actions including required crew upgrades and/or procedural enhancements The SS involved has discussed management expectations regarding this event with the Plant Director and Operations Hanager.
The crew participated in several simulator scenarios with special emphasis on the following:
Communications Team Work Plant Awareness Procedural Compliance An evaluation determined the Auxiliary feedwater flow could be raised to 250 gpm. Procedures have been revised to include this flow rate.
The STA involved was counseled on the requirement for performing Appendix BB of 42EP-2ZZOl. All STAs were briefed on the requirements to perform Appendix BB of 42EP-2ZZ01.
An Instruction Change Request to 42EP-2ZZ01 has been initiated to provide additional guidance for securing the essential chillers.
NRC FomI 366A (SBS)
0 II NRC FORM'366A US. NUCLEAR REGULATORY COMMISSION APPROVED 0MB NO, 31500104 (64)9)
E XP I R E S. 4/30/92 ESTIMATED BURDEN PER RESPONSE TO COMPLY WTH THIS LICENSEE EVENT REPORT (LER) INFORMATION COLLECTION REQUEST: 500 HRS. FORWARD, COMMENTS REGARDING BURDEN ESTIMATE TO THE RECORDS TEXT CONTINUATION AND,REPORTS MANAGEMENT BRANCH IP4)30), U.S. NUCLEAR
.REGULATORY COMMISSION. WASHINGTON. DC 20555. AND TO THE PAPERWORK REDUCTION PROJECT (31504))04). OFFICE OF MANAGEMENTAND BUDGET, WASHINGTON, DC 20503.
FACILITY NAME (1I DOCKET NUMBER 12) LER NUMBER (6) PAGE (3)
SEQUENTIAL REVISION NUMBER NUMBER Palo Verde Unit 2 o s o o o 52 989 0 0 3 0 1 1 5 OF 1 5 TEXT /I/ mort tttct /t toII/)td. Btt tdd4'oot//VRC Form 356A'4/ (12)
All the STAs have been instructed to verify and understand all related plant information to ensure complete and accurate.
information is provided to the NRC.
IV. PREVIOUS SIMILAR EVENTS:.
There have been no previous similar occurrences reported pursuant to IOCFR50.73.
There have be'en previous reactor trips reported as a result of low .SG level. 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. ADDITIIONAL INFORMATION:
There have been 4 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.
NR C FomI 366A (689)
0