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LER 89-020-00:on 890527 & 0605,during Local Leak Rate Testing,Valve RHR-V-9 Automatically Isolated.Caused by Procedural Inadequacy & Inadequate Corrective Action Following Second Event.Procedure Modified
	ML17285A606
Person / Time
Site:	Columbia
Issue date:	06/26/1989
From:	Arbuckle J; WASHINGTON PUBLIC POWER SUPPLY SYSTEM
To:	;
Shared Package
ML17285A605	List: ML17285A604; ML17285A606;
References
	LER-89-020, LER-89-20, NUDOCS 8907100289
	Download: ML17285A606 (7)
	v • d • e

NRC Form 3dd U.S. NVCLEAR REGULATORY COMMISSION (94)3 I APPROVED OMB NO. 31504)I04 EXPIRES: 8/31/88 LICENSEE EVENT REPORT (LER)

FACILITY NAME (II DOCKET NUMBER (2) PAGE 3 Washin ton Nuclear Plant - Unit 2 0 5 0 0,03 97 > oFp 7

""'<"Residual Heat Remova Shutdown Coo sng ontasnment so at>on a ve osures (ESF Actuations) Due to Procedure Inadequacies EVE NT D AT E (5I LER NUMBER (6) REPORT DATE (7) OTHER FACILITIES INVOLVED IS)

MONTH OAY YEAR YEAR SEGVENTIAL MONTH OAY YEAR FACILITYNAMES DOCKET NVMBERISI

+Jr, NVMSER vvd NVMSESI 0 5 0 0 0 0 527 89 9 020 0 0 6 268 9 0 5 0 0 0 THIS REPOR'T IS SUBMITTED PURSUANT T 0 THE REOUIREMENTS OF 10 CFR (): /Checfr one oi more of the lollowlnpl )11 OPERATING MODE (8) 73.71(SI 5 20.402(SI 20.405(cl 60.73(el(2)(iv)

POWER 20.405( ~ l(l) (I) 50M(c)() I 60.73(el(2) (vl 73.71(c)

LEVEL p p p 20.405( ~ l(1) (El 50.35(c)12) 60.73(el(2)(vii) OTHER /Specify in AOstrect Oelow end in Text, /Y/IC Foim 20.405(el(1l(ili) 50.73( ~ l(2)(il 60.73(el(2 l(viiil(AI 366A/

20A05( ~ l(1)(iv) 60.73(el(2)(iil 50 73(el(2l(viill(BI 20,405( ~ ) (I l(vl 50.73(el(2)(ill) 50.73( ~ l(2)(cl LICENSEE CONTACT FOR THIS LER (12)

NAME TELEPHONE NUMBER AREA CODE J .D. Arbuckle Com liance En ineer 50 937 7- 211 5 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT (13)

CAUSE SYSTEM COMPONENT MANVFAC TVRER REPORTABLE TO NPRDS c Ilrr SYSTEM COMPONENT MANUFAC.

TVRER REPORTABLE TO NPROS )F ~~~+8 SUPPLEMENTAL REPORT EXPECTED (14) MONTH OAY YEAR EXPECTED SUBMISSION DATE I I SI YES llf yN. COmpleN EXPECTED SUBCI/SS/O/Y DATE/

X NO ABs'TRAcT /L/mlt to /400 soccer, I e., eppioslmetrfy fifteen slnpre specs typewntren lines/ (id)

On the following dates, two related Engineered Safety Feature (ESF) isolations occurred involving the Residual Heat Removal (RHR) System:

(a) On May 27, 1989 an ESF isolation occurred when the Inboard RHR Shutdown Cooling Supply Valve (RHR-V-9) automatically isolated on a high flow signal during the performance of Local Leak Rate Testing (LLRT). During recovery efforts, Plant personnel were also unable to restore RHR Shutdown Cooling within the time frame required by the Plant Technical Specifications. The associated Limiting Condition for Operation (LCO) requires that with no shutdown cooling loop in operation, an alternate method of reactor coolant circulation and decay heat removal be established, and coolant temperatures be monitored at least once every hour.

Contrary to that, from 0353 hours0.00409 days 0.0981 hours 5.83664e-4 weeks 1.343165e-4 months until 0400 hours0.00463 days 0.111 hours 6.613757e-4 weeks 1.522e-4 months (seven minutes beyond the.3-hour, time limit imposed by the Technical Specifications), shutdown cooling remained inoperable ana an alternate method of reactor coolant circulation and temperature monitoring was not established. The cause of the isolation was unanticipated system interaction during the testing of RHR Shutdown Cooling Supply Valve RHR-V-8 which closed RHR Shutdown Cooling .Supply Valve RHR-V-9. During the test, RHR-V-9 was hydraulically locked shut by pressure inside the valve body and remained closed until the pressure differential was reduced. Valve RHR-V-9 was then opened and shutdown cooling restored at 0400 hours0.00463 days 0.111 hours 6.613757e-4 weeks 1.522e-4 months .

8907100289 890b2b PDR *DOCI(I 05000397 PNU NRC Form 3dd (9 83)

NRC Form 388A U.S. NUCLEAR REGULATORY COMMISSION (9431 LICENSEE EVENT REPORT {LER) TEXT CONTINUATION APPROVEO OMB NO. 3150-0104

~r EXPIRESI 8/3(188 FACILITY NAME (I) OOCKET NUMBER (2) LER NUMBER (8) PAGE (3)

jyo SEQUENTIAI. PQ REVISION Pci& NUMBER '>A'I NUMSER 0 5 0 0 0 TEXT (d moro FPFoo )F roqo)rod, SFF oddr)r'oool iVRC Form 38843) (17)

Abstr act (continued)

(b) On June 5, 1989 whi1 e attempting to compl ete the same LLRT, RHR- Y-9 automatical ly i sol ated when di fferenti al pressure swi tch RHR-DP IS-12B again actuated. However, in this event Residual Heat Removal Shutdown Cooling was not impacted, nor.was there a hydraulic lock condition in the system. Although the isolation could not be immediately reset, Plant Operators were able to de-energize RHR-V-9, open it manually and successfully complete the test.

During troub'leshooting efforts to clear the isolation signal, Plant Operators discovered that RHR-DPIS-12A had been removed from service and RHR-DPIS-12B was reading 5.2 PSID and isolated with the equalizing valve shut.

was,equalized and restored to service and RHR-V-9 was then opened Accordingly,'HR-DPIS-12B by means of the motor operator.

The root cause of the first event is procedural inadequacy in that the RHR-Y-9 actuation had not been anticipated during LLRT pressurization of the RHR Shutdown Cooling Supply line while testing RHR-V-8.'h'e pressurization inadvertently actuated a differential pressure switch (RHR-DPIS-12B) designed to isolate the RHR system on excess flow. When RHk-Y-9 closed on the high pressure line and the pipe was then depressurized, a high differential pressure between the inside of the valve and the RHR system hydraulically locked it shut.

The root causes of the second event are 1) procedural inadequacy and 2) inadequate corrective action'ollowing the first event.

Corrective action consists of 1) modifying the procedure to remove the correct DPIS from service when testing RHR-V-8, 2) performing a Category 1 Root Cause Analysis for both events, 3) performing an assessment of the RHR Shutdown Cooling isolations, and 4) addressing the hydraulic lock issue in a Nonthly Operational Bulletin.

There is no safety significance associated with either event. At the time of both events the reactor vessel head was removed, the fuel pool gates were removed and reactor water level was greater than 22 feet above the reactor vessel flange. These conditions provided a large heat sink for core coolin9.

Plant Conditions a) Power Level - OX b) Plant Node - 5 (Refueling)

Event Descri tion On the following dates, two related Engineered Safety Feature (ESF) isolations occurred involving the Residual Heat Removal (RHR) System:

NRC FORM SSSA

~ U,S, CPOI 1988 S20-S89r00070 (983)

NRC Form 366A U.S. NUCLEAR REOULATORY COMMISSION (9431 LICENSEE EVENT REPORT (LER) TEXT CONTINUATION APPROVED OMS NO. 3160-0106 EXPIR ES: 6/31/SS FACILITY NAME (1) OOCKET NUMBER (2) LER NUMBER (6) PACE (3)

SEQUENTIAL RYAN REVISION NUMBER NUMBER Washin ton Nuclear Plant - 'Unit 2 0 5 0 0 0 3 9 7 8 9 0 2 0 0 0 0 3 OF TEXT /// moro cpoco /1 rcr/vi cd, Ceo odd@'ooo////IC Form 366/('c/ (12)

(a) On tray 27, 1989 an ESF isolation occurred when the Inboard RHR Shutdown Cooling Supply Valve (RHR-V-9) automatically isolated on a high flow signal during the performance of Local Leak Rate Testing (LLRT). During recovery efforts, Plant personnel were also unable to restore Residual Heat Removal (RHR) Shutdown Cooling within the time frame required by the Plant Technical Specifications.

(The Technical Specifications allow shutdown cooling to be secured for 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months during an eight hour period). The Technical Specifications also require that, with no RHR Shutdown Cooling loop in operation, an alternate method of reactor coolant circulation and decay heat removal be established, and coolant .

temperatures be monitored at least once every hour. Due to unanticipated system interactions during the LLRT, RHR-V-9 could not be immediately re-opened to restore shutdown cooling.

In order to perform an LLRT on RHR-V-9, shutdown cooling was secured at 0053 hours6.134259e-4 days 0.0147 hours 8.763227e-5 weeks 2.01665e-5 months with the intent of restoring shutdown cooling no later than 0253 hours0.00293 days 0.0703 hours 4.183201e-4 weeks 9.62665e-5 months and remaining within the bounds of the Limiting Condition for Operation (LCO).

The LLRT on RHR-V-9 was done successfully with enough of the two hour interval left to expect to complete an LLRT on RHR-V-8. The LLRT is performed by closing manually-operated valve RHR-V-113 upstream of RHR-V-8 and RHR-V-9, and pressurizing the piping between RHR-V-113 and either RHR-V-8 or RHR-V-9 depending on which valve is to be tested (RHR-V-9 is between RHR-V-113 and RHR-V-8). In testing RHR-V-8, RHR-V-9 remains open and the piping between RHR-V-8 and RHR-V-9 is then pressurized with a test pump.

When RHR-V-9 received the auto-close signal, the bonnet area of the valve was pressurized to the same pressure as the line (approximately 1000 psig). In preparation for opening RHR-V-9 the operators conducting the LLRT depressurized the test assembly and the pressurized piping. However, the design of RHR-V-9 is such that when it closed the bonnet remained, at the higher pressure.

0234 hours0.00271 days 0.065 hours 3.869048e-4 weeks 8.9037e-5 months the shutdown cooling isolation signal on RHR-V-9 was reset and an At open signal was provided to the motor operator on RHR-V-9. Because the piping had been depressurized and the bonnet remained at higher pressure, the valve could hot open across the resultant differential pressure and the motor operator fuses blew before efforts to open the valve were completed.

At 0300 hours0.00347 days 0.0833 hours 4.960317e-4 weeks 1.1415e-4 months , Fuel Pool Cooling Pump FPC-P-1B was started providing an alternate means of decay heat removal through the fuel pool. ( It was recognized that this did not allow an exit from the LCO action statement because this did not establish sufficient circulation through the reactor core in accordance with the Technical Specification bases). At 0349 hours0.00404 days 0.0969 hours 5.770503e-4 weeks 1.327945e-4 months the fuses for the RHR-V-9 motor had been replaced and the operators recognized the hydraulic lock condition on RHR-V-9. Valve RHR-V-113 was then shut and the line pressurized to about 750 psig. This decreased the differential pressure between the valve body and the system and allowed the valve to be cracked manually off the seat. The valve was then opened by means of the motor operator.

NRC CORM 366A V.S. Crcr IOSS-S2n-SSO.OOOTO IS 63)

NRC Form 366A U.S. NUCLEAR REOULATORY COMMISSION (9.83)

LICENSEE EVENT REPORT (LERI TEXT CONTINUATION OMB NO. 3(50&(04 'PPROVEO EXPIRES: 8/31/88 FACILITY NAME (1) OOCKET NUMBER (2)

LER NUMBER (6) PACE (3I SEQUENTIAL F~'u'EVISION NUMBER .r'N'VMOSR Washington Nuclear Plant - Unit 2 o s o o o 8 9 02 0 00 OF 0 7 TEXT /llmoro F/rood/F ror/rrlrrrd. rrro oddi)r'orro/HRC %%drm 36643) (17)

(b) On June 5, 1989 while attempting to complete the same LLRT, Inboard Shutdown Cooling Supply Valve RHR-V-9 automatically isolated when differential pressure switch RHR-DPIS-128 again actuated. In this particular case, although the isolation could not immediately be reset, RHR Shutdown Cooling was not impacted as during the first event because the Reactor Recirculation (RRC) System was in service providing an alternate method of reactor coolant circulation. In addition, a hydraulic lock condition did not occur because Plant personnel maintained the system in a pressurized state. As a result, Plant Operators were able to de-energize RHR-V-9, open complete the test on the still-pressurized piping.

it manually to 25% and successfully Upon completion of the LLRT, Plant personnel began troubleshooting activities in an attempt to clear the, isolation signal. During alignment of the system in accordance with Plant Procedure (PPM) 7.4.4.3.2.2, "High-Low Pressure Interface Valve Leak Test," a Plant Instrument and Control (I&C) Technician was directed to remove RHR-DPIS-12 from service in accordance with the procedural lineup.

Further investigation found the instrument to be properly equalized and isolated, and labeled RHR-DPIS-12A. However, RHR-DPIS-128 was found to be reading 5.2 PSID and isolated with the equalizing 'valve shut. Accordingly, RHR-DPIS-128 was equalized and restored to service. Valve RHR-V-9 was then opened by means of the motor operator.

Immediate Corrective Action (a) At 0400 hours, Plant operators started RHR Pump 28 (RHR-P-28) and restored shutdown cooling (seven minutes beyond the three-hour time restriction imposed by the Technical Specifications).

(b) Following troubleshooting activities, RHR-V-9 was opened by means of the motor operator.

Further Evaluation and Corrective Action A. Further Evaluation

1. The May 27, 1989 event is reportable under 10CFR50.73(a)(2)(i)(B), "a condition prohibited by the Plant's Technical Specifications," and 10CFR50.73(a)(2)(iv), "an event or condition that resulted in manual or automatic actuation of any Engineered Safety Feature (ESF)."

2. The June 5, 1989 event is reportable under 10CFR50.73(a)(2)(iv), "an event or condition that resulted in manual or automatic actuation of any Engineered Safety Feature (ESF).

3. There were no structures, components or systems that were inoperable at the start of the event that contributed to the event.

NRC FORM 3&6A ~ V.S, 090r 1966 520-569r00020 (903)

NRC Form 355A U.S. NUCLEAR REGULATORY COMMISSION (843)

LICENSEE EVENT REPORT (LER) TEXT CONTINUATION APPROVED OMS NO. 3150M(04 EXPIRES: 8/31/88 FACILITY NAME (11 DOCKET NUMBER (2)

LER NUMBER (5) PAGE (3)

>@a 58OUENT/AL,> ~s RavrsroN NUM88R W.8 NVM88R Washington Nuclear Plant - Unit 2 o s o o o 97 89 0 0 0 0 5 OF 0 7 TEXT /I/moro s/ross is isqrr/rsd, rrss sddidorrsl /VRC %%drm 355A's/ (IT)

4. During the testing of RHR-V-8 dynamics created sufficient pressure spikes in the it is believed that the test pump and pipe pipe that differential pressure switch RHR-DPIS-12B actuated and closed RHR-V-9. The pressure switch is designed to sense high RHR flow rates which could indicate a pipe leak or break and isolate the RHR system, thereby, securing any potential leak.

5. The root causes of these two events are as follows:

(a) May 27, 1989 o Procedural inadequacy - the procedure (PPM 7.4.4.3.2.2) incorrectly required that the DPIS remain in service during the test. Following the event, Plant personnel recognized that the DPIS was within the test pressure boundary and modified the procedure accordingly to remove the DPIS from service during testing on RHR-V-8. "In addition, Plant Management directed that a Category 1 (the highest level) Root Cause Analysis be performed on this event.

(b) June 5, 1989 o Procedural inadequacy - the procedure incorrectly identified RHR-DPIS-128 as RHR-DPIS-N012, and did not specifically identify either RHR-DPIS-12A or RHR-DPIS-12B. Due to the lack of "A" or UBU valve designation, and the absence of another flow element on RHR [RHR-DPIS-12A is on an RHR line and RHR-DPIS-12B is on a-Reactor Recirculation (RRC) System,line], it was assumed that RHR-DPIS-N012 contained the contacts for both logic trains.

However, if the procedure had specifically identified RHR-DPIS-12B as the instrument to be removed from service, this event would not have occurred.

o Inadequate corrective action following the first event - if it had been recognized following the first event that RHR-DPIS-N012 should have been RHR-DPIS-12B, the procedure would have been revised accordingly at that time. It should be noted that Root Cause Analysis efforts for the first event had not been completed by the time this event occurred. Followup efforts associated with this event:discovered the discrepancy with the DPIS designation in the procedure. Plant Management directed that a Category 1 Root Cause Analysis also be performed on this event.

In addition, a- contributing factor comon to both events is system design. The system is designed as "one out of one" logic, which allows no margin for errors or problems during system operation and testing.

NRC FORM 388D i(/ 8.

~ CPOr 1988 520 589r00070 03 83)

NRC FoIm 388A U.S. NUCLEAR REGULATORY COMMISSION (9 83)

LICENSEE EVENT REPORT (LER) TEXT CONTINUATION APPROVED OMS NO. 3150&108 EXPIRES: 8/31/88 FACILITYNAME (1) DOCKET NUMBER (3) LER NUMBER LS) PAGE (3)

SEOUENTIAL 4EVISION YEAR $g) NUMBER 'HN NUM884 Washin ton Nuclear Plant - Unit 2 0 5 0 0 0 9 7 8 9 0 2 0 0 0 6 oF0 TEXT ///moro /Poco l/ /ooolied, use oddio'oool //RC Fon4 38848/ (17)

B. Further Corrective Action

1. The LLRT procedure for testing RHR-V-8 was modified to require removal of RHR-DPIS-12B from service during testing on RHR-V-8.

2. An overall assessment of the RHR Shutdown Cooling isolations which have occurred is currently being performed by the Plant Technical and Nuclear Safety Assurance Groups.

3. Tne Nuclear Safety Assurance Group has addressed the hydraulic lock issue in a Monthly Operational Bulletin. The Monthly Operational Bulletin is required reading for Operations and Maintenance personnel.

4. An evaluation will be performed to consider the feasibility of a design change to incorporate a redundant trip system logic scheme for RHR isolation actuation.

Safety Si nificance There is no safety significance associated with these events. At the time of both events, the reactor vessel head was removed and reactor water level was greater than 22 feet above the reactor vessel flange. These conditions provided a large heat sink for core cooling.

In the first event, the allowed interval for restoration of shutdown cooling was exceeded by seven minutes. Fuel pool cooling was operable providing a natural circulation method of core cooling (although not considered to have been proven to provide adequate assurance of core mixing). When RHR shutdown cooling was restored, temperatures in the RHR system had increased only 11'from 73'o 84'F). This provides adequate assurance that temperatures in the core remained well below 140'F .

I In the second event, the Reactor Recirculation (RRC) System was in service providing an al ternate method of. reactor coolant circulation.

Accordingly, these events posed no threat to the health and safety of either the public or Plant personnel.

Similar Events There have been several LERs associated with the loss of Shutdown Cooling; however, none with the same root causes.

E I IS Information Text Reference EIIS Reference System Component Residual Heat Removal (RHR) System BO Residua'I Heat Removal Shutdown Cooling Supply Valve (RHR-V-9) BO I SV NRC FORM 388A (9.831 i' 'POI )988 530 589/00090

NRC form 388A U.S. NUCLEAR REGULATORY COMMISSION (843)

LICENSEE EVENT REPORT (LER) TEXT CONTINUATION APPROVED OMB NO. 3150M)04 EXPIRES: 8/31/88 FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER (8) PACE (3)

SEOVENTIAL <<3 REVISION YEAR NUMBER 'od NVMSER Washington Nuclear Plant - Unit 2 o s o o o3 97 8 9 0 2 0 000 7 OF 0 7 TEXT /I/moro FIIFEO /8 roqrr/rOd. IIFO Oddro'oool IVRC form 3/ISA 8/ (12)

EIIS Information Text Reference EIIS Reference System Component Valve RHR-V-8 80 ISV RHR-DPI S-128 80 PDI S RHR-DP IS-12A 80 PDIS RHR-DPIS-N012 80 PDI S FPC-P-18 DA P RHR-P-28 80 P Fuel Pool Cooling (FPC) System DA Reactor Recirculation (RRC),System AD NRC FOAM 088A

~ V.S. CPOr )$ 88 S20-S80r00070 (083)

ML17285A606

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