ML20215L328

Summary of Operating Reactors Events Meeting 87-11 on 870420.List of Attendees,Viewgraphs & Input to NRC Performance Indicator Program Encl

ML20215L328
Person / Time
Site:	Calvert Cliffs, Oconee, River Bend, Diablo Canyon, 05000000
Issue date:	05/05/1987
From:	Lanning W Office of Nuclear Reactor Regulation
To:	Rossi C Office of Nuclear Reactor Regulation
References
OREM-87-011, OREM-87-11, NUDOCS 8705120189
Download: ML20215L328 (28)

Text

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gg MEMORANDUM FOR:. Charles E. Rossi, Director a <

Division of Operational Events Assessment .

FROM: -Wayne Lanning, Acting Chief Events Assessment Branch Division of Operational Events Assessment

SUBJECT:

SUPMARY 0F THE OPERATING REACTORS EVENTS

. MEETING ON APRIL 20, 1987 - MEETING 87-11 On April 20, 1987, an Operating Reactors Events meeting (87-11)'was held to brief senior managers from NRR, RES, AEOD and Regional Offices en events which occurred since our last meeting on April _ 6,1987. The list ~of' attendees is included as Enclosure'1.

The events discussed and the significant elements of these events are presented in Enclosure 2. Enclosure 3 provides a summary of those presented events that will be input to NRC's performance indicator program as significant events.

Od; int Siimed By:

Wayne Lanning, . Acting Chief Events. Assessment Branch Division of Operational Events Assessment

Enclosures:

As stated i

cc'w/ Encl.:

See Next Page L

cc w/ Encl.:

See Next Page

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. MEMORANDUM FOR: Charles E." Rossi, Director ..

i ' Division of Operational Events Assessment ,

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SUMMARY

,0F THE. OPERATING REACTORS EVENTS

SUBJECT:

'y .PEETING ON APRIL-20,- 1987J : MEETING 87 . ,

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. On April 20, 1987, an Operating Reactors Events meeting (87-11) was held to -

brief the Offic'e Director, the: Division Directors and their! representatives on 4

' events which occurred sincefour last meeting on April 6, .1987. The list of:

attendees is .inc.luded ,

as Enclosure 1.

Yheeventsdiscu$sedandthe'significantelementsoftheseeventsarepresentedt in Enclosure 2. Enclosure 3 provides a summary of.those presented events'that-will be input:to N,RC's perfomance-indicator program as significant events.

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'\ . Wayne Lanning, Acting Chief i Events Assessment Branch i Division of Operational Events _ Assessment

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Enclosures:

i-As stated 4

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DISTRIBUTION

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3 PEMORANDUM FOR:- Thomas E. Murley, . Director.

0ffice of Nuclear. Reactor.' Regulation

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FROM: . Wayne Lanning, Acting Chief Events Assessment Branch ... .

Division of' Operational Events Assessment- ,

SUB ECT:

SUMMARY

0FLTHE OPERATING REACTORS EVENTS -

PEETING ON 1987 - MEETING 87-

~

dn Apriig20, ,1987, an ~0peratina-Peactor Events meeting (87-11) was held to . .i

~brief the Office. Director, the Division Directors and their representatives on ' ,

events which occurred:since our last meeting on~ April 6,1987. The list _of. <-

attendees is, included as En' closure 1. .

S; _

The events' discussed and the'significant elements of these. events' are presented in Enclosure 2g-; Enclosure 3 provides a summary of:those presented events that will be input to NRC's- performance indicator program as .significant events. -

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Wayne Lanning, Acting Chief _

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. Events Assessment Branch Division of Operational Events Assessment s .

, Enclosures ^

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See Next Page  ; x cc w/ Encl.. .

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!S n NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20565 O j 1,,.....,/ mr: arr MEMORANDUM FOR: Charles E. Rossi, Director Division of Operational Events Assessment

-FROM: Wayne Lanning, Acting Chief .

Events Assessment Branch Division of Operational Events Assessment ,

SUMMARY

OF THE OPERATING REACTORS EVENTS.

MEETING ON APRIL 20, 1987 - MEETING 87-11 On April 20, 1987, an Operating Reactors. Events meeting -(87-1k)J was held to

.brief senior managers from NRR, RES, AEOD and Regional Offices, on e' vents which occurred since our last meeting on April 6,1987. The list'of attendees is included as Enclosure 1.

The events discussed and the significant elements of these events are presented in Enclosure 2. Enclosure 3 provides a summary of those presented events that will be input to NRC's performance indicator program as significant events.

,/

))W Waynd Lanning, Ac ing hief

Events Assessment Division of Operational Events Assessment

Enclosures:

As stated cc w/ Encl.:

See Next Page l

I-Those Listed- . .

. ~ . . ... - -- - -. .

cc: . T. . Murley - J. Crews, Reg. Y J. Sniezek C. Trammell-J. Taylor H. Schierling E. Jordan G. Knighton E. Beckjord H. Pastis W. Russell, Reg. I B. Youngblood

.J. Nelson Grace, Reg. II S. Stern B. Davis, Reg. III J. Calvo R. D. Partin, Reg. IV S. McNeil J. R. Martin, Reg. V R. Capra W. Kane, Reg. I L. Reyes, Reg. II C. Norelius, Reg. III E. Johnson, Reg. IV D. Kirsch, Reg. V

.R. Starostecki F. Miraglia S. Varga D. Crutchfield B. Boger G. Lainas G. Holahan F. Schroeder L. Shao J. Partlow B. Grimes F. Congel H. Miller E.- Weiss S. Black T. Partin, EDO E. Merschoff i

__x._ . _ .

ENCLOSURE 1 LIST'0F ATTENDEES OPERATING REACTCRS EVENTS BRIEFING (27-11)

APRIL 21, 1987 NAME DIVISION NAME DIVISION C. Rossi NRR A. Thadani NRR W. Lanning NRR R. Scholl NRR K. Heitner NRR M. Caruso NRR G. Lainas NRR C. Schulten NRR P. Baranowsky NRR T. Greene NRR F. Schroeder NRR J. Calvo NRR L. Spessard AE0D H. Pastis NRR J. Youngblood NRR G. Knighton NRR S. Black NRR J. Shapaker:. .NRR T. Martin EDO J. Partlow' NRR T. Chan NRR T. Novak AE00 R. Lob'el NRR J. Thompson NRR

J. Stefano NRR S. Stern NRR J. Rosenthal AEOD J. Carter NRR l E. Weiss AEOD J. Richardson NRR B. Grimes NRR H. Miller NRR L. Shau NRR R. Starostecki NRR E. Jordan AE00 S. Varga NRR C. Berlinger NRR D. Crutchfield NRR B. Boger NRR F. Miraglia NRR W. Minners DRPS G. Holahan NRR i W. Swenson NRR D. Oudinot NRR l

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ENCLOSURE 2 OPERATING REACTORS EVEtlTS BRIEFING 87-11

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APRIL 21, 1987 DIABLO CANYON 2 LOSS OF RESIDUAL HEAT REMOVAL AND RCS THERMAL TRANSIENT OCONEE 1, 2, 8 3 FOULING OF LOW PRESSI'RE SERVICE WATER, LOW PPESSURE INJECTION AND REACT 0P BUILDING COOLING UNITS-UPDATE OCONEE 3 VALVES TO HIGH PRESSURE INJECTION PUMPS ISOLATED OTHER EVENTS OF INTEREST RIVER BEND QA/0C PP0BLEM FOR PROCESS COMPUTER SOFTWARE CALVERT CLIFFS 1 ESF INADVERTENT ACTUATION OF CONTAINMENT SPRAYS

4/21/87 DIABLO CANYON 2 LOSS OF RESIDUAL HEAT REM 0"Al AND RCS THERMAL TRANSIENT PROBLEM:

LOSS OF RHR WHILE IN MODE.5, MID-LOOP OPERATION WITH STEAM GENERATOR MANWAY. REMOVAL IN PROGRESS AND CONTAINMENT ECUIPMENT HATCH REMOVED, CAUSE:-

UNPLANNED DRAINAGE-FROM RCS RCS LEVEL INSTRUMENTATION ERPORS P0TENTIAL PROCEDURAL INADEQUACIES SIGNIFICANCE:

THERMAL TRANSIENT-TO BOILING CONDITIONS'WAS EXPERIENCFD IN THE RCS WHILE BOTH RCS AND CONTAINMENT EOUNDARIES WERE BREACHED P0TENTIAL FAILURE BY LICENSEE TO CLASSIFY EVENT DISCUSSION:

INITIAL CONDITIONS - MODE 5, CONTAINMENT EQUIPMENT HATCH REMOVED,

. STEAM GENERATOR MANWAYS BEING. REMOVED RCS DRAIN D0WN TO MID-LOOP OPERATION COMPLETED ON DAY SHIFT, VORTEXING AT RHR PUMP SUCTION OCCURRED BRIEFLY DURING THIS EVOLllTION UNEXPECTED RCS LEAKAGE COMMENCED AT APPP0XIMATELY 8:30 P.M.

OPERATORS STARTED INVESTIGATING CAUSE RHR PUMP SHUT 0FF BY OPERATORS AT APPR0XIMATELY 9:20 P.M. DUE TO ERRATIC MOTOR CURRENT CONT ' IONS SECOND RHR PUMP STARTED c.i.D PROMPTLY SHUTDOWN DUE TO~SAME CONDITIONS (OPERATORS ATTRIBUTE RHR PUMP BEHAVIOR TO SUSPECTED VORTEXING)

PHR FLOW NOT RESTORED DUE TO UNCERTAINTY REGARDING PRESENCE OF PERSONNEL IN OP NEAR STEAM GENERATORS B0ILING COMMENCED'IN RCS WITHIN APPR0XIMATELY 1 HOUR (N0 RCS TEMPERATURE INDICATION AVAILABLE)

SHIFT FOREMAN DECLARED SIGNIFICANT EVENT (50.72 NOTIFICATION) AT 10:23 P.M. BASED ON EXCEEDING LCO 0F TECHNICAL SPECIFICATIONS

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4/21/87 DIABLO CANYON 2 OPERATORS RESTORED RHR COOLING BY ADDING WATER TO RCS FROM REFUELING WATER STORAGE TANK AND RESTARTING RHR PUMP AT.

10:48 P.M.

REC 0VERY TO MID-LOOP OPERATION ACHIEVED ON SUNDAY APRIL 12, OPERATORS CONDUCTED SPECIAL TEST TO DETEPMINE POINT OF VORTEXING BY LOWERING RCS LEVEL CONTAINMENT AND RCS INTEGRITY REMAINED BREACHED)

NOTE RCS LEVEL INSTRUPENT INDICATION WAS EPPATIC PRIOR TO AND DURING EVEFT, AT NO TIME DURING EVENT DID LEVEL INSTRUMENTATION INDICATE LOWERING OF PCS LEVEL FOLLOWUP:

AUGFENTED INSPECTION TEAM (AIT) INVESTIGATION ~CONTINU!NG CONTACT:

J. CREWS, REG. V l

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April 18, 1987

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2330 Hrs.

4/10/87 RER EVENT

SUMMARY

Plant Status: The Unit was in the seventh day of the first refueling outage following The plant was in aMode shutdown on 4/3/87 at 2352 hrs.

5 with the reactor coolant system temperature being maintained at approximately 87 F. Preparations are in progress to install tho steam generator nossle dams.

The reactor vessel level is being maintained at approximately half loop to support the installation of the nossle dams.

Operational - Residual Heat Removal pump 2-1 is in service Conflauration: providing flow through both RHR heat exchangers (trains are cross-tied).

. - Reactor vessel level uis being maintained by

-(,f.

V .. balancing letdown flow to the VCT with the charging flow back to the primary system (constant VCT level). Letdown is from the RER pump discharge via HCV-133 and charging is by gravity flow from the VCT via the normal charging path (through a non-

, operating centrifugal charging pump).

- Seactor Coolant System boron concentration is approximately 1997 ppm. '

., - The containment equipment hatch and personnel air 4 0 lock are open. The emergency personnel hatch is closed. Various jobs are in progress containment and a continuous purge is in inside progress of 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 is in service and the Conf.lguration: 2-2 RHR pump is available for service. All instrumentation associated with the RNR system is in service.

- Both Safety injection pumps are # cleared" and unavailable for service.

(' - Centrifugal Charging Pump 2-2 is OPERABLE and available for immediate service.

CCP 2-1 and the 4- Tmed $ and ts0Wed V M4snkutd<refteh4 .

ca.a seet tu n m: ass n i>:se test se >e

l positive displacement charging pump are

• . . administratively tagged out but are available for

(. service.

- The Refueling Water Storage Tank is avsilable-as a:

borated -water source with-level at- approximately 97%.

- All four scoumulators have been cleared and drained.

- Boric Acid Storage Tank 2-2 is at 80% level with a boron concentration of 22050 ppab. BAST 2-1 is Boric Acid Transfer pump 2-2 is available t

empty.

for service. The 2-1 transfer pump is cleared.

- Containment Fan Cooler Unita 2-1, 2-2, 2-3 an'd 2-4 are available for service. CFCU 2-5 is cleared.

l CFCU 2-3 is in service running in slow speed.

l - 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 auxiliary transformer banha have been oleared and the Unit is being powered from the Startup transformer bank. Diesel Generators 2-1, 2-2 and.1-3 are all available for servioe. 480 volt 1

(}, bus 2F is cleared for outage related work.

- All core exit thermocouples have been determinated

, in preparation for reactor vessel head removal.

- Post accident monitoring panels 1 and 2 are out of service for human factors related upgrades.

- Plant vent high range radiation monitor RM-29 is out of service. All other required process and area

, radiation monitors are in service.

P r

Shift Turnover: The previous watch had completed draining of the steam generator U-tubes per operating 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 wa's 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 stabilised at 107'0". Work was ongoing to uncouple and backseat the Reactor coolant Pumps. Various 7- containment penetration leak tests were ongoing.

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TIME EVENT DESCRIPTION DATA SOURCE

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4/10/87 1700 Bra Early in the shift, the Shift Foreman SFM/C0 informs the Control Operator that due to Statements the planned work to remove the steam generator primary manways, the reactor vessel level should be maintained below approximately 107'-8". This would assure '

that rater would not be allowed to spill over irto the steam generator lower head area from the RCS loops. Since indicat ions of RHR pump suotion vortexing had been stopped on the previous shift by raisint vessel level to 107'-0", the CO planned to maintain vessel level between 107'-0" to 1078-8" during the shift.

( 1850 Hrs Since assuming the watch at approximately CO Log

~ 1700 hrs, the vessel water level as indicated on the RVLIS system had slowly

(" . risen to the 1078-8". Vessel level is C0 Statement reduced back to 107'-0" by rejecting water back to the RWST via valve 8741.

2010 Hrs An Engineer enters the containment to Engineer's begin draining containment penetration 45 Statement in preparation to perform Surveillance -

i

, Test Procedure V-645. STP V-845 is the Security

• Local Leak Rate Test (LLRT) for that Computer penetration. The penetration was cleared by the Operations Department on 4/9/87 to allow the leak rate test to be completed (CR 00006713). The penetration serves the Reactor Coolant Pump seal leakoff return line to the Volume Control Tank.

! 2041- The Engineer enters the regenerative heat Engineer's exchangsr room and opens CVCS-314 as part Statement of the : procedure to drain the penetzation prior to beginning the leak rate test.

The Engineer verifies flow through the YCT on P-250 drain valve and then Trend Recorder exits .the containment to log onto another SWP while the peaetration is draining. Since the (s clearan,:e request for the job was approved the previous day, the Shift P

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Foremati was unaware that the diraining of the per etration was ongoing at this time. '

( Due to a leaking boundary . valve associrited with.the clearance, a drain path is created betwaan the VCT sud ~*

the RCDT. VCT level _immediately begins to decreas.e.

2051 Hrs contro3 Room operators note the downward trend VCT on P-250 in VCT level and increase letdown Trend Recorder from the primary system to stabilize VCT level ty further opening HCV-133. Due to the ir. crease in letdown flow, reactor vessel level begins to slowly decrease.

2054 Hra While trying to determine the cause of operator's the level decrease in the VOT, Control' Operator contacts the AO at the the Statement Auxiliary Building control board determine if any unusual evolutions are to P-250 Alarm occurring. The AO reports that the RCDT Printout level has increased. While they are on ,

the phene, the RCDT pump starts on high level. The C0 requests that the A0 estimate the flow rate into the RCDT and

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report bach $o the Control Room. .

.v Control ; Room receives notification that SFM Log the.RCP4 have been uncoupled and are backseated.

Due to the apparent loss of inventory 2124 H kgg 26s 3) fromthe primary VCT on P-250 system. Operators g , j isolate the charging and letdown flow Trend Recorder

• paths.

The loss of letdown flow to r,he VCT>causes VCT level to rapidly decrease. RVLIS on P-250 Level decrease in the primary system Trend Recorder stops (107'-4").

Auxiliary Building A0 reports the Control room that the estimated to leakage C0 statement into the RCDT is approximately 30 syn.

')hj' 2125 Hrs Operato:es notice amps on the 2-2 RER pump A /0 M,1I WY l,y beginning to fluctuate. The pump is shut P-250 Alarm down a:'ter starting the 2-1 pump. Ampe Printout also fluotuate on the 2-1 pump and it is SIM Statement secured almost immediately. Operators are dispatched

( ., to vent the pumps and seal coolers on both RHR pumps.

4 90*d SBEC EP5 6 *g58 3*u*H*s (1 Et*:80 486L'87/PO

.Due to the unexpected RHR pump cavitation

• SFH Statement

(- or vartexinsi operatora suspect validity of the JRVLIS ' indication.-- the operator -is An sent into containment to verify level indication on Tygon" Tube.

Outage Coordinator is requested to. verify SFM Statement the at-stus of the work on Steam Generator manway.s.

Operators continue work to locate the source of the leakage. SFH Statement 2138 Hrs Operators close LCV-1120 to stop inventory loss from the VCT. Co Los

'- isolatos the VCT from the RCDT.

This valve The level decrease in the VCT stops.

2147 Hrs The Engineer performing the leak rate Security test re-enters the containment to Computer continue the local leak rate test.

2200 Hrs The valves

! [7-vent penetration associated with the being drained Engineer I are opened. Statement

'l After opening the valves, the Engineer goes to find a Decon Tech to assist with the leth rate test.

2203 Bra Contro3 Room is notified that the venting on the 2-1 RHR pump has been completed. co Los 2221' Brs Control Room has received notification SFH that the Tyson Tube level is between Statement kh ./.b;- 106'-9" and 107'-0". The Control l Alb Operator throttled 'the discharge and P-250 Alarm

! 8 s/ started RER 2-1. Pump was vented before Printout i and during the re-start. Pump amps are

! awinging by about 20 amps.

Pump is immediately shut down.

2223 Hrs Shift Foreman declares a Significant Event. SFH Log s'

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L . 2225 Hrs Operaters re-cpen LCV-112C in an attempt t VCT on P-250 to localise the source of the leakage.

~(. VCT level again begins. to decrease.

Trend Recorder I Operatcra close LCV-112C and VCT level CO Statemen\Y stabilises.

2226 Hrs The Engineer performing the leak rate Engineer test finds a large amount of water on the Statement 91' elevation of the containment and believes that the water is associated with his draining of the penetration. He notifies Rad Protection personnel of the spill and isolates the vent valves from the penetration.

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

(; 2233 3rs HP Tech on 115' elevation notes HP Tech

.,, background on friskers exceeding.the X10 Statement soale. Continuous Air Monitors on the 140' elsvation are alarming.

2238 Hrs Operations personnel believe steam is CO Log being generated in the head as indicated by a slow trend up on the RVLIS SFM Statement

• indication. The Control Operator is notifieil that th; Steam Generator primary manways have not been removed. Valves 8805 A and B are opened to establish makeup ,o the remotor vessel.

2243 !!ra HP Foroman is notified by the Control EP Tech Room of a possible containment evacuation Statement due to the problems with the IUDL system.

HP Forenan then enters the contaisment to begin evacuation of unnecessary personnol.

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' - 2250 Hrs The Cet Trol Room is notified by personnel #

( inside Engineer was' of containment that the leak path -Statement the it.ontified as being associated with

. Inah rate test and that the- leakage gU was ist. lated (valve CVCS-314).

i 2251 Bra Reactor vessel level is indicating 4,IJ, & h,lN. 3 ft 0D 1 approximately 1108 Operators start RHR RVLIS on P-250

'/ pump i:-2. Pump amps Trend Recorder v' immedia,tely after the . pump fluotuate slightly stabilise. start, but P-250 Alarm Printer Short1r following the pump start, RHR RHR Discharge pump discharge board recorder rises to temperature on the control Temp Recorder 220 F. Within 6 minutes, approximately the pump discharge temperature has dropped to less than 2C 0 F.

2253 Bra Operatcra note minor cavitation on indication of the running RHR pump. BFH Log i

Valve 8980 (RWST to RER

! partially auction) is CO Log opened to increase makeup to the reactor vessel. Pump amps stabilise.

0-) .

2258 Hrs Control room receives notification of Q,g steam venting from a ruptured on the tygon tube SFM Log reactor head vent.

i Containment Evacuation alarm initiated at SFM Log the direction of the Shift Foreman.

f . .

2310 3rs Shift Foreman requests that the operator HP Tech /CO insidevant head of conatinment which is supplying isolate the thereactor Statements leak i.n the tygon hose. steam HP Tech and operator the le.sk. descend to head area and isolate No visible condensation or water i.: noted in the area.

! 2313 Hra

' Control room notified that the reactor CO Log vessel head vent has been isolated.

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232n Hrs Control Room is notified by HP Fersonnel SFM Statement

'r that the containment airborne is greater' "

(' than 1 MPC and is high in Iodine. C0 Log Operat ors place the containment Iodine Remova.1 fans into service to attempt t'o reduc < airborne activity.

232! Hrs Shift Foreman goes to the containment Security persor nel hatch to verify the status of Computer the ocntainment evacuation. While there, he in informed by the RP Foreman of the SFH/RP Foreman water leakage from the steam generator Statements manway s.

2342 Hrs Level in the pressurizer has reached CO Los approximately 40%. Operators begin divert ing letdown flow to the LHUT to reduce level and minimize the leakage from the steam generator manways.

0044 Hrs Operators open valve 8741 to begin CO Log pumping primary system water back to the RWST to further reduce level in the primary system. Letdown divert to the

(,. LHUT is secured at this time.

0102 Bra Operators stop rejecting water back to CO Log /

the RWST. Valve 8741 closed. RVLIS RVLIS on P-250 indicating approximately 114'-0". Trend Recorder 0320 Hrs Reactor vessel level is reduced to half- CO Log loop asd leakage from the steam generator '

manways is stopped.

approximately 108'-4".

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4/21/87' OCONEE 1, 2, AND 3 - FOULING OF LOW PRESSURE SERVICE WATER, 1.0W PRESSl!PE INJECTION AND REACTOP PUILDING COOLING UNITS llPDATE PROBLEM:

REDUCED HEAT TRANSFER CAPABILITY OF THE LOW PRESSURE INJECTION (LPI) AND REACTOR BUILDING COOLING UNIT 1RBCU) HEAT EXCHANGERS LIMITS REACTOR OPERATING POWEP CAUSE: LAKE SEDIMENT DEPOSITED IN HEAT EXCHANGERS SIGNIFICAPCE:

SAFETY EQUIPMENT INCAPABLE OF PERFORMING THE DESIGN HEAT REJECTION SAFETY FUNCTION DISCUSSION:

THE LOW PRESSURE SERVICE WATER SYSTEM TAKES RAW LAKE WATER TO C00L THE LPI AND RBCU HEAT EXCHANGERS LICENSEE ANALYZED THE TEST DATA AND CORRELATED IT TO EMERGENCY CONDITIONS THE-ANALYSIS SHOWED THAT UNDER LOCA CONDITIONS, THE PEAT EXCHANGERS WOULD NOT PERF0PM AT DESIGN RATING J

UPDATE:

ON APRIL 10, 1987, NRR ISSUED A CONFIRMATORY ORDER WHICH REQUIRED THAT THE HEAT EXCHANGERS BE CLEANED AND TESTED REACTOR OPERATION AT LOWER POWER LEVELS WITHIN THE HEAT TRANSFER CAPABILITY OF THE HEAT EXCHANGERS THE REACTOR HIGH FLUX TRIP SETPOINTS TO BE REDUCED CONTACT:

H. PASTIS, NRR

- -_ _ _ j

4/21/87 OCONEE 3 VALVES TO HIGH PRESSURE INJECTTON PUMPS ISOLATED

-PROBLEM:

BOTH MOTOR OPERATED SUCTION VALVES FROM THE B0 RATED WATER STORAGE TANK TO THE HIGH. PRESSURE INJECTION PUMPS WERE IN THEIR NORMAL Cl.0 SED POSITION BUT THE MOTOR SUPPLY BREAKERS WERE OPEN.

CAUSE: NUMEROUS PERSONNEL ERRORS SIGNIFICANCE HIGH PRESSURE INJECTION SYSTEM COULD NOT HAVE RESPONDED TO AN ENGINEERED SAFEGUARDS SIGNAL DISCUSSION:

OCONEE UNIT 3 WAS HEATING UP WHEN VALVES VERE DISCOVERED INOPERABLE FOR TWENTY HOURS SITl'ATION CORRECTED IMMEDI ATELY UPON IDENTIFICATION FOLLOWUP:

ENFORCEMENT CONFERENCE HAS BEEN SCHEDULED l

CONTACT:

H. PASTIS, NRR

4/21/87-OTHER EVENT OF' INTEREST.

RIVER BEND - QA/0C PROBLEM FOR PROCESS COMPUTER SOFTWARE PROBLEM: RIVER BEND STATION EXCEEDED LICENSED THEPMAL LIMIT CAllSE: INADEQUATE QA/0C BY VENDOR ON SOFTWAPE'FOR-PROCESS. CONTROL COMPUTER SIGNIFICANCE:

QA/QC 0N SOFTWARE FOR PLANT COMPUTER ERROR IN SOFTWARE OF A NON-SAFETY PELATED COMPUTER IMPACTS POWER LEVEL DISCUSSION:-

ON.MAY 13, 1987, LICENSEE INSTALLED VENDOR SUPPLIED SOFTWARE 0N PLANT COMPUTER

- VEND 0F, GE, STATED SOFTWARE ONLY' IMPACTS B0P PARAMETERS LICENSEE HAD BENCH CHECKED SOFTWARE AGAINST VEND 0P SPECIFICATIONS ALERT OPERATOR NOTICED COMPUTER' THERMAL POWER READ 0VTS NOT CCRRECTLY RESPONDING TO CHANGES IN RECIRCULATION FLOW LICENSEE REDUCED POWER, SilBSTITUTED OLD SOFTWARE - PP0BLEM RESOLVED.

' THERMAL LIMIT EXCEEDED FOR MAXIMUM 0F FOUR HOURS

- PEAK THERMAL POWER 101,15 LICENSE LIMIT FOLLOWUP:

LICENSEE /GE PERFORMING IN DEPTH ROOT CAUSE ANALYSIS REGION IV FOLLOWING UP WITH LICENSEE, WILL ISSUE INSPECTION REPORT VENDOR INSPECTIONS FOLLOWING UP WITH GE INFORMATION NOTICE ON MAXIMUM LICENSED POWER LEVEL UNDER REVIEW BY NRR CONTACT:

S. STERN, NRR

4/21/87 OTHER EVENT OF INTEREST CALVERT CLIFFS 1 - ESF INADVERTENT ACTUATION OF CONTAINMENT SPPAYS PROBLEM:

ESF INADVERTENT ACTUATION OF CONTAINMENT SPRAYS CAUSE:

OPERATOR ERROR SIGNIFICANCE:

INADVERTENT SAFETY SYSTEM ACTUATION RESULTING FROM GPEPATOR ACTIONS ON WRONG UNIT POTENTIAL FOR ELECTRICAL EQUIPMENT DEGRADATION DUE TO WETTING BY BORATED WATER (BORIC ACID INTRUSION)

PERSONNEL CONTAMINATION

- DISCllSSION:

UNIT 1 IN MODE 5, ELECTRICAL TERMINAL B0XES OPENED FOR EQ MAINTENANCE, AND CONTAINMENT SPRAY CONTAINMENT ISOLATION (CS/CI)

BLOCK VALVE DE-ENERGIZED (FAIL OPEN) FOR EQ MAINTENANCE OPERATOR ON 12-HOUR SHIFT; UNIT 2 (4:00-8:00 A.M.), UNIT 1 (8:00 A.M.-4:00 P.M.)

OPERATOR ERRCNE00 SLY WENT TO UNIT 2 PENETRATION ROOM INSTEAD OF UNIT 1 PENETRATION ROOM FOR CLOSURE OF CS/CI VALVE FOLLOW-UP:

LICENSEE WILL: ANALYZE POTENTIAL SAFETY IMPLICATION OF OPENED TERMINAL B0XES TO BORIC ACID DEGRADATION INVESTIGATE THE HUMAN ERROR AND CONTRIBUTING FACTORS INVOLVED FOR THE WRONG UNIT EVENT INVESTIGATE POSSIBLE UNIT 1/ UNIT 2 LABELING INADE0tlACIES FOR EQUIPMENT REGION I TO FOLLOW UP ON CORRECTIVE ACTION MEASURES TAKEN BY LICENSEE, INCLUDING LATE NOTIFICATION OF 50,72 CODE SECTION ON REPORTABLE EVENTS FOR OPERATION OF ESF SYSTEMS CONTACT:

J. THOMPSON, NRR

i; .;

REA:iDR SCRAM SUVARY WEEK ENDINS 04/12/87

!.-PLANT SPECIFIC DATA IATE SITE LNIT F0sER E 3 CcUEE COMPLI- VT3 YT3 Yi; CAI D i 43IiE EELCd 70TAL 151 !!!

04/06/97 FARLEY  ! OA EQUIFMENT/SRM N0 3 1 4 04/06/37 FERN! 2 31 A FER3CNhEL NO '3 0 -3 04/07/S7 EALEM 2 90 A -UNrNCWN NO 2 1 3 04/09/37 CATAWiA 1 100 A E901FMENT/ VALVE YES 3 0- 3 04/09/87 CATAWBA '1 0M EGUIPMENT/RPI NO ~ 3 1 4 04/09/87 SAINTLUCIE 2 100 A PERSCMMEL NO 3 0' 3 04/10/B7DREitEN 3 100 A EQUIPMENT / PIPE NO 2 0 2 04/10/97VOSTLE 1 30 A EQUIFMENT/VALVENO 2 6 8 04/11/97V05TLE 1 21 A EQUIFMENT/ VAL'lE NC 3 6 9 04/12/67 PALISA:ES 1 75 M EQUIPMENT / PIPE NO- 1 0 1 04/12/97 MILLSTONE I 66 A EGUIFMENT/ PUMP NO ~3 1 4 02/12/37 PARRIS 1 65 M FI'!:NNEL NO 7 2 9-04/!2!37 MILLSTChE  ! 10 A PERICUEL NO 3 2 i

SUMMARY

OF COMPLICATIONS SITE liNIT COMPLICATIONS CATAWBA 1 BORCN DILUTICN MITISATION SYSTEM INITIATED ESF AFTER SCRAM

__ _ _ _ ._ ~ _ _ . . . - , . . . . .

y II. COMPARISON OF WEEKLY ETATISTICS-WITH INDUSTRY AVERAGES SCRAMS FOR WEEK ENDING 04/12/87 ECRf41 CAUSE POWER NUMBER -1997 1986 1985 CF WEEELY WIEKLY WEEKLY SCRAM 3(5) AVERAGE AVERAGE AVERAGE YTD (3)(4) (8) (9)

• • FOWER 15%

EOUIP. RELATED >15% 6 3.9 4.3 5.4 PERS. RELATED(6) >15% 3 1.7 1.8 2.0

-OTHER(7) >15% 1 1.0 0.4 0.6

• • Subtotal **

10 6.5 6.5 8.0 t* POWER <15%

EQUIP. RELATED <15% 2 'i . 2 1.4 1.3 PERS. RELATED <15% 1 1.O O.8 O.9 OTHER <15% 0 O.4 O.2 0.I tk Gubtotal i4 3 2 . 6 .' 2.4 2.1

, 4t1 fatal 414

• 13 9 .' 1 9.9 10.4 MANUAL VS AUTO SCRAMS TYPE i NUMBER 1987 1986 1985 OF WEEKLY WEEKLY NEEKLY SCRAMS AVEPAGE AVERAGE AVERAGE YTD MANUAL SCRAMS 3 1.5 1.O 1.O AUTOMATIC SCRAMS 10 7.5 7.9 9.4

gp ]P' r .E .

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REACTOR.iCRAM 5'.i? MARY ~

[? "

!di [ WEEKENDING 04/19/67

,  !. PLANT SPECIFIC DATA i

id iO CATE. $1TE . d ti!T,?0'4ER RFS CAUSE

-C3.9.!- YT3 . YTD' : fTD

, ./ CATIONS ??!'iE iELOW -TCTAL g i- 151 15%

W . .

3;

1. '04/13/97 WATERFORD. '3 .100 A EQUIPMENT /CEAC NO .3 01

..  ; .04/13/97 PERRY  ! 10 M ' E2VIFMENT/ FIFE NO 1- 2 3

$ ~04/14/97 SA!MT LUCIE 1 0A FERSONNEL - ' NO . 0 2- 2

'* e - 04/14/97 HMRIS -  ! 87 M . FERSONNEL . N3 3 2 -10 04/15/97 WATERFORD 3 - 50 A PEFSONNEL~ NO- /4 0 4

e. '. 04/15/97 MC6UIRE 1 ' 100 A UNtNCH NO 2 0 2 W. M' " 04/16/87 HALDAM NECK l'!;'100 A UNKNGWN NO 1 .O. 1 3

' rh - .

04/15/97 MILLSTONE 2 100 A EQUIPMENT /BKR NO 2 0 2

~ *

~g , 04/16/27.!US20EHANNA 2 100 A UNENEWN- NO 1 0 1

,t 04/16/87 PALD VERSE 2 100 A UNKNOWN - NO 0 1 2# !Y 04/17/97 INI!AN FDIh? 3 . 100 A UNKNCWN N0' 3 1 4

'4. .

-04/19/57 WOLF MEEX- '! 10 A - dKG41 N0 3' 0- 3 04/19!67 N*.Riii ANNA 1,  !! A EGU!F?ENT/C ?E M 'l 0 1

.,p

' '[

f 1*

, r. /

.. i O

+ f l s -' .

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19 tc

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• II. COMPARISON OF WEEKLY STATISTICS WITH INDUSTRY AVERAGES SCRAMS FOR WEEK ENDING s 04/19/87~

SCRAM CAUSE POWER NUME.ER 1987 19S6 1925

• OF WEEf:LY WEEi'LY WFI; _Y ECRAME (3 ) AVERAGE AVERAGE A'.'ERAGE YTD (3) (4) (S)(9)

• • POWER >157.

EQUIP. RELATED > 15 *'. 3 3.7 4.0 5.4 FERS. RELATED(6) >15% 2 1.7 1.2 2.0 OTHER(7) >15% 6 1.3 0.4 0.6

• • Subtotal ** '

11 6.7 6.5 8.0 t* POWER <15%

EQUIP. RELATED <15% 1 1.2 1.4 1.3 PERS. RELATED <15% 1 1. 0 0.8 0.9 OTHER '15% o O.4 0.2 O.2 kt Subtet=1 tv 2 2.S 2. 4 Z.'

444 Total 161 13 9.3 8.9 10.4 MANUAL VS AUTO SCRAMS TYPE NUMBER 1987 1986 1985 OF WEEV.LY WEEKLY WEEKLY ECRAMS AVERAGE AVERAGE A'/ERAGE YTD MANUAL SCRAMS 2 1.5 1.0 1.0 AUTOMATIC SCRAMS 11 7.8 7.9 9.4

[. . . . . . . . . _

s

, 3 NOTES

1. PLANT SPECIFIC DATA PASED ON INITIAL REVIEW OF 50.72 REFORTS FOR THE WEEK OF INTEREST. PERIOD IS MIDNIGHT Sl!NDAY TPROUGH MIDNIGHT SUNDAY SCRANS ARE DEFINED AS REACTOR PROTECTIVE ACTUATIONS WHICH RESULT IN ROD MOTION, AND EXCLUDE PLANNED TESTS OR SCRAMS AS PART OF PLANNED SHUTDOWN IN' ACCORDANCE WITH A PLANT PROCEDURE.

2. RECOVERY COMPLICATED BY EQUIPMENT FAILURES OR PERSONNEL ERRORS UNRELATED TO CAUSE OF SCRAM.

3. 1986 INFORMATION DERIVED FROM ORAS STUDY OF UNPLANNED REACTOR TRIPS IN 1986. WEEKLY DATA DETERMINED BY TAKING TOTAL TRIPS IN A GIVEN CATEGORY AND DIVIDING BY 52 WEEKS / YEAR.

A. IN 1986, THERE WERE AN ESTIMATED TOTAL OF 461 AUTOMATIC AND MANUAL UNPLANNED REACTOR TRIPS AT 104 REACTORS (HOLDING OPERATING LICEN$ES). THIS YIELDS AN AVERAGE RATE OF 4.4 TRIPS PER REACTOR PER YEAR AND AN AVERAGE RATE OF 8.8 TRIPS PER WEEK FOR ALL REACTORS.

5. BASED ON 107 REACTORS HOLDING AN OPERATING LICENSE.

6. PERSONNEL RELATED PROBLEMS INCLUDE HUMAN ERROR, PROCEDURAL DEFICIENCIES, AND MANUAL STEAM GENERATOR LEVEL CONTROL PROBLEMS.

1 I

i 7. "0THER" INCLUDES AUTOMATIC SCRAMS ATTRIBUTED TO ENVIRONMENTAL j CAUSES (LIGHTNING), SYSTEM DESIGN, OR UNKNOWN CAUSE.

l

8. 1985 INFORMATION DERIVED FROM AN ORAS STUDY OF UNPLANNED REACTOR TRIPS IN 1985. WEEKLY DATA DETERMINED BY TAKING TOTAL TRIPS IN A GIVEN CATEGORY AND DIVIDING BY 52 WEEKS / YEAR.

9. IN 1985, THERE WERE AN ESTIMATED TOTAL OF 541 AUTOMATIC AND MANUAL UNPLANNED REACTOR TRIPS AT 93 REACTORS (HOLDING FULL POWER LICENSES). THIS YIELDS AN AVERAGE RATE OF 5.8 TRIPS PER REACTOR

, YEAR AND AN AVERAGE RATE OF 10.4 TRIPS PER WEEK FOR ALL REACTORS.

7,

. .,_ 1

.cy a,_

ENCLOSURE 3 Pa;a N2. 'l 04/21/E7.

SIGNIFICANT EVENTS FREQUENCY PERFCRMANCE INDICATOR No. 3 PLANTNAME EVENT EVENT DESCRIPTION . GTR CAUSE DATE

- 5IA!LO CAMON 2 - 04/11!374C5THERNLTRU:S!ENTCLETOLCISOFRH?WHILEBOTHRCS ~ 1 E2 'FMENT FAILURE & FERSONNEL EEROR .

AND CCNTAIMENT BOUNDARIES BREACHED

, OCONEE 3- ' 04/11/97 YALVES TO HPI INJECTION PURFS FOUND ISOLATED DURING 2 PER$0h 2 ERROR HEATUP e

e s, .

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