PNO-V-87-030A, on 870415,Region V Augmented Insp Team Commenced Insp Activities Re 870410 Loss of Suction to RHR While at Half Loop.Team Expects to Stay Onsite Through 870421.Summary of Sequence of Event Encl

PNO-V-87-030A:on 870415,Region V Augmented Insp Team Commenced Insp Activities Re 870410 Loss of Suction to RHR While at Half Loop.Team Expects to Stay Onsite Through 870421.Summary of Sequence of Event Encl

ML20212R637
Person / Time
Site:	Diablo Canyon
Issue date:	04/20/1987
From:	Mendonca M, Narbut P NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V)
To:
References
PNO-V-87-030A, PNO-V-87-30A, NUDOCS 8704270289
Download: ML20212R637 (8)

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PRELI9lINARY NOTIFICATION OF EVENT OR UNUSUAL OCCURRENCE--PNO-V-87-30-A Date 04/20/87 l This preliminary notification constitutes EARLY notice of events of POSSIBLE safety or public interest significance..The information presented is as initially received without

'vsrification or evaluation and is basically all that is known by Region V staff on this s

. < date.

FACILITY: PACIFIC GAS & ELECTRIC COMPANY Emergency Classification Notification of Unusual Event DIABLO CANYON UNIT 2 DOCKET NO. 50-323 Alert SAN LUIS OBISP0 COUNTY, CA Site Area Emergency General Emergency XX Not Applicable

SUBJECT:

LOSS OF SUCTION TO RESIDUAL HEAT REMOVAL SYSTEM WHILE AT HALF-LOOP Th2 Region V AIT team members arrived at the site on the afternoon of Tuesday, April 14, and commenced their inspection activities at 8:00 a.m. on April 15.

Interviews'with members of the operating crew were conducted through the evening (until approximately 9:30 p.m.) on April 15, 1987.

The AIT team expects to remain onsite through approximately Tuesday, April 21, 1987.

A sumary sequence of the events is attached.

Tha above information was obtained from the AIT team leader, and is current as of 10:00 a.m.

April 20, 1987.

CONTACT: M. Mendonca, RV P. Narbut FTS 463-3720 805-595-2354 b

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CA (Reactor Licensees) f t,> REGION V: FORM 211 PDR Resident Inspectorj//6 (Revised 3/14/83) tjg ofe6W'^{ Co*

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j' - t Diablo' Canyon, Unit.2 4 04/10/87 RHR EVENT SUl9tARY Plant Status: The Unit was in the^ seventh day f the first refueling outage following a shutdown on 04/03/87'at 2352 hrs. The plant was:in Mode 5:with the reactor coolant system temperature.being maintained at.approximately 87*F.

Preparations-were in progress to install-the steam. generator nozzle dams. The

reactor vessel level was being maintained at approximately half . loop to'-

support the installation of.the nozzle dams.-

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-Opsrational - Residual Heat Removil pump 2-1 was in service providing flow through both RHRheatexchangers(trains;arecross-tied)..

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Configuration

- Reactor vessel level was being maintained by balancing letdown flow to t the VCT with the charging flow back to the primary system (constant VCT

' level). Letdown was from the RHR pump ~ discharge via.HCV-133 and charging

was by gravity flow from the VCT via the normal charging path (through a

{ non-operating centrifugal charging pump).

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! - Reactor Coolant System b'oron concentration was approximately 1997 ppm.

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The containment equipment hatch and personne1~ air lock were open. The ,

i emergency personnelLhatch was closed. Various jobs were in progress inside of containment and'a continuous purge was in progress with containment ventilation exhaust fan E-3 discharging through RCV-11 and 12 to the plant vent.

, Plant Equipment -

Residual Heat Removal pump 2-1 was in service and the 2-2 RHR pump was Configuration: available for service. All instrumentation associated with the RHR system was in service.

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Both Safety Injection pumps'were " cleared"* and unavailable for service.

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Centrifugal Charging Pump 2-2 was~ 0PERABLE and available for immediate

service. CCP 2-1 and the positive displacement charging pump were 1

administrative 1y tagged out but were available for service. '

The Refueling Water Storage Tank was available as a borated' water source

with level at approximately 97%.

l All four accumulators have been cleared

• and drained.

- Boric Acid Storage Tank 2-2 was at 80% level with a boron concentration

] of 22050 ppeb.' BAST 2-1 was empty. Boric Acid Transfer pump 2-2 was available for servicei The 2-1 transfer pump was cleared.

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- Containment Fan Cooler Units 2-1, 2-2, 2-3 and 2-4 were available for service. CFCU 2-5 was cleared.* CFCU 2-3 was in service running in slow speed.

- All four steam generators have a secondary side water level of approximately 73% wide range with the 10% atmospheric dumps open to atmosphere.

- The main and auxil'iary transformer banks have been cleared

• and the Unit was being powered from the Startup transformer bank. Diesel Generators 2-1,s 2-2 and 1-3 were all available _ for service. 480 volt bus 2F was cleared

• for outage related work.

- All core exit thermocouples have'been de-terminated in preparation for reactor vessel head removal.

- Post accident m'nitoring o panels 1 and 2 were out of service for human

_ factors related upgrades.

Plant vent high rar.ge radiation monitor RM-29 was out of service. All other required process and area radiation monitors were.in service.

Shift Turnover: The previous watch had completed draining of the steam generator-U-tubes per op'erating procedure OP A-2:II. During the draining, it was noted that the U-tubes began to drain once vessel level, as indicated on the RVLIS system, reached'107'-3". It was reported that once the U-tubes had drained, level dropped to 106'-6" where signs of RHR pump cavitation were noted.

Once level had been restored to 107'-0", indications of pump cavitation or vortexing stopped and level was stabilized at 107'0". Work was ongoing to uncouple and backseat.the Reactor Coolant Pumps. Various containment penetration leak tests were ongoing.

• -Tagged out and isolated for maintenance / testing.

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TIME EVENT-DESCRIPTION DATA SOURCE'

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4/10/87 _

1700-hrs Early in the shift, the Shift Foreman informs the Control Operator. SFM/CO:

'that due to the planned work to remove the steam generator primary Statements -

a manways, the. reactor vessel level should be maintained below approximately.107'-8". This would _ assure that water would not be 3- allowed to spill over into the steam generator _ lower head area from the RCS loops. Since indications of'RHR pump. suction vortexing i had been stopped on.the previous shift by. raising vessel level to

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i 107'-0", the C0 planned to maintain vessel level between 107'-0"

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and 107?-8" during the. shift.

1850 Hrs Since'aisuming the watch at'approximately 1700 hrs, the vessel C0 Log '

4 water level as indicated.on.thei RVLIS system ha'd slowly risen to i the 107'-8". . Vessel}1evel is reduced back,to 107'-0" by rejecting C0 Statement, 1 water back'to the RWST via' valve 8741..

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~2010 Hrs ..

An Engineer enters the containment to' begin. draining contain- Engineer's

ment penetration 45 in preparation to perform Surveillance _ Statement i

Test Procedure V-645. STP V-645 is the Local Leak Rate Test 1 (LLRT) for that penetration. .The penetration was cleared

• by _

Security 1 the Operations Department 'on 4/9/87 to allow the leak rate test Computer

} to be completed (CR 00005713). The penetration serves the Reactor Coolant Pump seal leakoff return line to the Volume j Control Tank.

l 2043 Hrs The Engineer enters the~ regenerative heat exchanger room and opens Engineer's CVCS-314 as part of the procedure to drain the penetration prior Statement to beginning the leak rate test. The Engineer verifies flow through the drain valve and then exits the containment to log VCT on P-250'

onto another SWP while the penetration is draining. Since the Trend Recorder '

, clearance request for the job was approved the previous day,

} the Shift Foreman was unaware that the draining of the penetration i was ongoing at this time. Due to a leaking boundary valve associated with the clearance, a drain path is created between the VCT and the RCDT. VCT level immediately begins to decrease.

j 2051 Hrs Control Room operators note the downward trend in VCT level'and VCT on P-250 j increase letdown from the primary system to stabilize VCT_ level Trend Recorder 1 by further opening HCV-133. Due to the. increase in. letdown flow, j reactor vessel level begins to slowly decrease.

! 2054 Hrs While trying to determine the cause of the level decrease in the- Operator's

!- VCT, the Control Operator contacts the A0 at the Auxiliary Statement l Building control board to determine if any unusual evolutions

are occurring. The A0 reports that the RCDT level has increased. P-250 Alarm i While they are on the phone, the RCDT pump starts on high level.

l The C0 requests that the A0 estimate the flow rate into the RCDT j and report back to the Control Room.

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, TIME -

EVENT DESCRIPTION' - DATA SOURCE

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Control Room receives notification that the'RCP's have been SFM Log uncoupled and are backseated. '

i 2124 Hrs Due.to'the apparent loss of inventory from the primary system, VCT on P-250 Operators isolate the charging =and letdown flow paths. The

loss-of. letdown flow to the VCT causes VCT level to rapidly -

decrease. Level decrease in the primary system stops. (107'-4")

Auxiliary Building A0 reports to the Control Room that the C0 Statement estimated leakage into the' RCDT is approximately 30 gpm.

2125 Hrs ' Operators notice' amps on the 2-2 RHR pump beginning to' fluctuate. P-250 Alarm The pump is shut down after starting the 2-1 pump. Amps also Printout fluctuate on the 2-1 pump and it is secured almost imediately.

Operators are dispatched to vent the pumps and seal coolers on SFM Statement both RHR pumps.

Due to the unexpected RHR pump cavitation or vortexing, operators SMF Statement.

suspect the validity of the RVLIS indication. An operator is '

sent into containment to-verify level indication on Tygon Tube'.

Outage Coordinator is requested to verify the status of the work SFM Statement on Steam Generator manways.

Operators continue work to locate the source of the leakage. SFM Statement 2138 Hrs Operators close LCV-112C to stop inventory loss from the VCT. CO Log This valve isolates the VCT from the RCDT. The level decrease in the VCT stops. 9

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2147 Hrs The Engineer performing the leak rate test re-enters the Security containment to continue the local leak rate test. Computer j 2200 Hrs The vent' valves associated with the penetration being drained Engineer are opened. After opening the valves, the Engineer goes to Statement find a Decon Tech to assist with the leak rate test.

i 2203 Hrs Control Room is notified that the venting on the.2-1 RHR pump CO Log has been completed.

. 2221' Hrs Control Room has received notification that the Tygon Tube level SFM Statement is between 106'-9" and 107'-0". The Control Operator throttled the discharge and started RHR 2-1. Pump was vented before and P-250 Alarm-I during the re-start. Pump amps are swinging by about 20 amps. Printout Pump.is immediately shut down.

2223 Hrs Shift Foreman declares a.Significant Event. SFM Log 8

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j 2258' Hrs Control room receives notification of. steam venting from a '

SFM Log- -

ruptured tygon tube on.the reactor head vent.

s Containment Evacuation' alarm initiated at the direction of. SFM Log-the Shift Foreman.- > >

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. }2310 Hrs'ShiftForemanrequeststhattheoperatorinside:ofcontainment; HP Tech /C0~

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' isolate the reactor head vent which is supplying the steam'. Statements-leak in.the tygon hose. HP Tech and operator descend to head' area and isolate the leak. No visible, condensation or water.

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2313 Hrs -' Control room notified that the reactor vessel head vent has. CO. Log.

been isolated.. .

'2322 Hrs Control Room is notified by HP Personnel that the' containment-SFM Statement 1 airborne is greater than 1 MPC'and is high'in' Iodine. Operators- ~

! place the containment Iodine Removal fans.into' service to attempt CO Log to reduce airborne activity.

2323_ Hrs Shift Foreman goes to the' containment. personnel hatch to .

Security 2

verify the status.of the containment evacuation. While there,- Computer 4

he is informed by the RP Foreman of the water : leakage from the - SFM/RP Foreman

. steam generator manways. Statements i . . .

1 2342 Hrs ' Level in the pressurizer has reached approximately;40%.- 'C0 Log -

, Operators begin diverting letdown flow.to the LHUT to reduce

! level and minimize the leakage from the' steam generator 4

manways.

1 0044 Hrs Operators open valve 8741 to-begin pumping primary-system- ' C0 Log .

water back to the RWST to further reduce level'in the-

.! primary system.:. Letdown divert'to the LHUT lis secured at this time.

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-0102 Hrs Operators stop rejecting; water back'to the RWST. : Valve 8741 C0' Log /-

closed. RYLIS-indicating approximately.114'-0".

.RVLIS on P-250 L

, . Trend Recorder

~0320 Hrs Reactorvessellevelkis-redu6edtohalf-lohpand' leakage .

CO. Log.

from the steam generator manways is stopped M RVLIS indication-

, 4 at approximately 108'-4".

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TIME EVENT DESCRIPTION DATA SOURCE 2225 Hrs Operators re-open LCV-112C in an attempt to localize the source VCT on P-250 of the leakage. VCT level again begins to decrease. Operators Trend Recorder close LCV-112C and VCT level stabilizes. C0 Statement 2226 Hrs The Engineer performing the leak rate test finds a large amount Engineer of water on the 91' elevation of the containment and believes Statement that the water is associated with his draining of the penetration.

He notifies the Rad Protection personnel of the spill and isolates the vent valves from the penetration.

2230 Hrs HP Tech on the 140' elevation of containment notices airborne HP Tech activity levels increasing and begins taking air samples to Statement locate the source. Rad Protection personnel begin evacuation of workers off of the 115' elevation due to elevated airborne readings.

2233 Hrs HP Tech on 115' elevation notes background on friskers exceeding HP Tech the X10 scale. Continuous Air Monitors on the 140' elevation Statement are alarming.

2238 Hrs Operations personnel believe steam is being generated in the head Co Log as indicated by a slow trend up on the RVLIS indication. The Control Operator is notified that the Steam Generator Primary SFM Statement manways have not been removed. Valves 8805 A and B are opened to establish makeup to the reactor vessel.

2243 Hrs HP Foreman is notified by the Control Room of a possible HP Tech containment evacuation due to the problems with the RHR system. Statement HP Foreman then enters the containment to begin evacuation of unnecessary personnel.

2250 Hrs The Control Room is notified by personnel inside of containment Engineer that the leak path was identified as being associated with the Statement leak rate test and that the leakage was isolated (valve CVCS-314).

, 2251 Hrs Reactor vessel level is indicating approximately 110'. RVLIS on P-250 Operators start RHR pump 2-2. Pump amps fluctuate slightly Trend Recorder immediately after the pump start, but stabilize. P-250 Alarm Printer Shortly following the pump start, RHR pump discharge temperature RHR Discharge on the control board recorder rises to approximately 220 F. Temp Recorder Within 5 minutes, the pump discharge temperature has dropped to less than 200 F.

2253 Hrs Operators note minor indication of cavitation on the running SFM Log RHR pump. Valve 8980 (RWST to RHR suctien) is partially opened to increase makeup to the reactor vessel. Pump amps stabilize. C0 Log 1

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