ML20151Q813

Summary of Operating Reactors Events Meeting 88-031 on 880802.Viewgraphs & List of Attendees Encl

ML20151Q813
Person / Time
Site:	Arkansas Nuclear, Prairie Island, Brunswick, 05000000
Issue date:	08/03/1988
From:	Lanning W Office of Nuclear Reactor Regulation
To:	Rossi C Office of Nuclear Reactor Regulation
References
OREM-88-031, OREM-88-31, NUDOCS 8808110296
Download: ML20151Q813 (17)

Text

.

5008 q MEMORANDUM FOR:

Charles E. Rossi, Director Division of Operational Events Assessment Office of Nuclear Reactor Regulation FROM:

Wayne Lanning, Chief Events Assessment Branch Division of Operational Events Assessment Office of Nuclear Reactor Regulation

SUBJECT:

THE OPERATING REACTORS EVENTS MEETING August 2, 1988 - MEETING 88-31 On August 2, 1988 an Operating Reactors Events meeting (88-31) was held to brief senior managers from NRR, OSP, AE00, Commission Staff, and Regional Of fices on events which occurred since our last meeting on July 26, 1988.

The list of attendees is included as Enclosure 1.

The events discussed and the significant elements of these events are presented in Encicsure 2. presents one event suggested for long-term followup and a summary of reactor scrams.

One significant event was identified for input to NRC's Performance Indicator Program.

Wayne Lanning, Chief Events Assessment Branch Division of Operational Events Assessment Office of Nuclear Reactor Regulation

Enclosures:

As stated cc w/Enclo.:

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F. Miraglia, 12G-18 J. Forsyth, INP0 E. Jordan, AE00 E. Sylvester, 14A-18 E. Beckjord, NL-007 E. Adensam, 14H-12 W. Russell, R1

0. Oilanni, 13H-20 B. Davis, Rlli M. Virgilio, 13H-14 J. N. Grace, R11 G. Dick, 13G-16 R. D. Martin, RIV J. Calvo, 130-17 J. B. Martin, RV W. Kane, R1 L. Reyes, Ril E. Greenman, R111 J. Callan, RIV

0. Kirsch, RV S. Varga, 14E-4

0. Crutchfield, 13A-2 B. Boger, 14A-2 G. Holahan, 13H-4 G. Lainas, 14H-3 L. Shao, BE-2 J. Partlow, 70-24 B. Grimes, 9A-2 F. Congel, 10E-4 E. Weiss, AE00 T. Martin, 12G-18 J. Guttmann, SECY

~h A. Thadani, 7E-4 S. Rubin, AE00 R. Barrett, 10E-2

ENCLOSURE 1

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LIST OF ATTENDEES OPERATING REACTORS EVENTS BRIEFING (88-31)

August 2. 1988 NAME ORGANIZATION NAME ORGANIZATION R. Scholl NRR/00EA M.Virgilio NRR/0RSP W. Troskoski DEDR0 F. Paultiz OSP/TVA T.P. Gwynn OCM/LZ M.S. Callahan GPA/CA J. Guttmann SECY C. Harbuck NRR/0RSP J. Kelly OSP/TVA M.L. Reardon NRR/00EA E.A. Reeves NRR/P02-1 B. Buckley NRR/P02-1 T. Greene NRR/00EA H. Berkow NRR/P02-3 S. Varga NRR/0RP

0. Oilanni hRR/003-1 G. Holahan NRR/0RSP L. Rubenstein NRR/0RSP J. Calvo NRR/0RSP W. Minners RES/0RPS

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0. Crutchfield NRR/0RSP D.C. Fischer NRR/00EA C.E.Rossi NRR/00EA B. Boger NRR/AORI J.E. Rosenthal AEOD C. Haughney NRR/0RIS F. Miraglia NRR/ADP

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ENCLOSURE 2 I

OPERATING REACTORS EVENTS BRIEFING 88-31 EVENTS ASSESSMENT BRANCH LOCATION:

12-B-11 WHITE FLINT TUESDAY, AUGUST 2, 1988, 11:00 A.M.

PRAIRIE ISLAND 1 & 2 REACTOR PROTECTIVE SYSTEM CIRCulT DESIGN DEFICIENCY s:- LMARAZ UNIT 1 STRESS CORROSION CRACKING STEAM GENERATOR TUBES (UPDATE)

BRUNShlCK UNIT 2 ASCO PRESSURE SWITCHES ARKANSAS UNIT 2 REACTOR COOLANT PUMP SEAL LEAK 8

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88-31

'g PRAIRIE ISLAND UNITS 1 & 2 REACTOR PROTECTIVE SYSTEM CIRCUIT DESIGN DEFICIENCY JULY 26, 1988 PROBLEM OVERPOWER DELTA TEMPERATURE (OPAT) AND OVER TEMPERATURE DELTA TEMPERATURE (OTAT) TRIP SETPOINTS DO NOT COMPENSATE PROPERLY FOR HIGH AXIAL FLUX DIFFERENCE (AFD).

CAUSE A DESIGN FLAW PREVENTS THE NEWER CIRCUlT MODULES FROM FUNCTIONING AS ORIGINALLY INTENDED.

SAFETY SIGNIFICANCE CRITICAL HEAT FLUX LIMITS COULD BE EXCEEDED IF AFD IS NOT ADEQUATELY COMPENSATED FOR IN RPS VARIABLE TRIP SETPOINTS.

DISCUSSION hO UNIT 1 WAS IN COASTDOWN AT END OF CORE LIFE-84% POWER.

O UNIT 2 WAS AT 100% POWER.

O UNIT 1 LOWERED POWER IN ORDER TO AFFECT MINOR REPAIR TO A RCP.

AFTER THE REPAIRS WERE COMPLETED, REACTOR POWER WAS TEMPORARILY LIMITED TO 48% DUE TO XENON BUILDUP.

O WHILE WAITING FOR XENON DECAY / BURNOUT, THE CORE DEVELOPED AN AXIAL FLUX DIFFERENCE (AFD).

O AS THE AFD INCREASED THE OPERATORS EXPECTED TO OBSERVE A DECREASE IN THE SETPOINTS FOR OPAT AND OTAT INSTRUMENTS.

O 3 OUT OF 4 INSTRUMENT CHANNELS RESPONDED AS EXPECTED.

THE ANOMALOUS CHANNEL HAD A FOXBORO 62H STYLE "C" CONTROLLER INSTALLED (THE THREE OPERABLE CHANNELS HAD STYLE "B" CONTROLLERS).

O TESTING ON UNIT 2 REVEALED THAT ALL STYLE "C" MODULES FAILED TO RESPOND PROPERLY (3 OUT OF 4 CHANNELS).

O THE LICENSEE ENTERED THE APPROPRIATE LCO.

O UNIT 1 WAS SHUTDOWN AND UNIT 2 WAS RAMPED DOWN.

UNIT l'S STYLE "B" MODULES WERE TRANSFERRED TO THE UNIT 2 INSTRUMENT CHANNELS SO THAT UNIT 2 COULD RESUME FULL POWER OPERATION.

CONTACT:

R. KARSCH

REFERENCE:

50.72 # 12973

PRAIRIE ISLAND UNITS 1 S 2 88-31 I

O WESTINGHOUSE HAD PREVIOUSLY ADVISED THE LICENSEE THAT STYLE "C'I MODULES WERE AN ACCEPTABLE REPLACEMENT FOR STYLE "B" MODULES.

O THIS PROBLEM OCCURRED AT KEWAUNEE IN 1974 AND WAS SOLVED BY A SIMPLE MODIFICATION TO THE MODULE'S WIRING.

THIS CHANGE WAS NOT COORDINATED WITli EITHER THE NRC OR WESTINGHOUSE.

O PRAIRIE ISLAND MODIFIED THElR STYLE "C" MODULES AND RESTARTED UNIT 1 FOR THE REMAINDER OF THEIR COASTDOWN.

Rlli MONITORED MODIFICATION AND POST-MAtNTENANCE TEST.

O WESTINGHOUSE WAS NOTIFIED BY THE LICENSEE.

WESTINGHOUSE WILL INFORM THE AFFECTED FLANTS TO TAKE CORRECTIVE ACTION (MAY BE AS FEW AS 3 OR AS MANY AS 6 PLANTS).

'O THE PRAIRIE ISLAND LICENSEE HAS INDICATED THAT THEY WILL ISSUE A PART 21 NOTICE AND NOTIFY INPO.

FOLLOWUP

, POSSIBLE INFORMATION NOTICE.

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ALMARAZ UNIT 1 STRESS CORROSION CRACKING SG-TUBES JULY 13, 1988 PROBLEM PRIMARY TO THE SECONDARY SIDE LEAK WAS DETECTED IN STEAM GENERATOR "A";

i RAPID STRESS CORROSION CRACKING IN SG TUBES HAS BEEN FOUND.

CAUSE CAUSTIC STRESS CORROSION CRACKING,

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THE UNIT IS A 900 MWE, W-TYPE, 3-LOOP PWR IN SPAIN, o

STEAM GENERATOR TYPE D3, MANUFACTURED IN PENSACOLA (USA),

o DURING REFUEllNG OUTAGE IN FEBRUARY 1988, 60% 0F STEAM GENERATOR TUBES WERE INSPECTED AND NOTHING ABNORMAL WAS DETECTED, y'

NEW FULL FLOW CONDENSATE POLISHING SYSTEM PUT INTO OPERATION THIS

CYCLE, o

BETWEEN APRIL AND JULY 1988, THE CONCENTRATION OF S0DIUM IONS WAS ABOVE 20 PPB AND REACHED 60 PPB, o

WESTINGHOUSE RECOMMENDS AN UPPER LIMIT OF 20 PPB FOR S0DIUM CONCENTRATION, o

SODIUM ATTRIBUTED TO POLISHING SYSTEM (S0DIUM HYDROX1DE IS USED TO CLEAN RESINS, SOME NA+ REMAINED IN THE DEMINERAll2ERS AND LATER FLUSHED OUT WHENEVER THE DEMINERAll2ERS WERE PLACED IN SERVICE).

o ON JULY 13, 1988, A PRIMARY TO SECONDARY LEAK WAS DETECTED (1 GPM)

AND PLANT WAS SHUTDOWN, c

lhMEDIATE INSPECTION OF ALL SG TUBES WAS INITIATED USING STANDARD B0BBIN PROBE-EDDY CURRENT AND ROTATING BOBBIN PROBE (FOR TUBES WITH INDICATION OF DEGRADATION),

o THE RESULTS OF THE INSPECTION TO DATE ARE AS FOLLOWS:

CONTACT:

R. CID AND J. GUILLEN

AtMkRZAUNIT1 85-31 STEAM GENERATOR DEFECTIVE TUBES "A"

106 "B"

235 "C"

1G0 DEFECTIVE:

THOSE TUBES FOUND WITH GREATER THAN 20% TUBE WALL DEGRADATION, o

TWO 10BE SAMPLES WERE REMOVED FROM THE "A" SG (THE LEAKING TUBE AND ANOTHER WITH 90% TUBE WALL DEGRADATION),

o LEAKING TUBE 1 CRACK, AXIALY ORIENTED, 1" LENGTH 3/8" AB0VE TUBE SUPPORT PLATE CRUD WAS DETECTED AROUND THE TUBE, 3" HEIGHT o

SECOND TUBE 3 CRACKS, AX1ALY ORIENTED,

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LOCATED JUST AT THE SUPPORT PLATE CRUD WAS NOT DETECTED ONE SAMPLE HAS BEEN SENT TO W (PITTSBURGH) AND THE OTHER

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TO CIEMAT (SPAIN) FOR INDEPENDENT TESTING, o

THE CHARACTERISTIC 0F THE DEFECTS FOUND ARE AS FOLLOWS:

ALL CRACKS ARE AXIALLY ORIENTED (1-5 CRACKS),

MOST ARE LOCATED AT THE FIRST SUPPORT PLATE, HOT LEG, THE CRACKS ORIGINATED ON THE SECONDARY SIDE, MOST OF THE CRACKS (90%) HAVE A DEPTH BETWEEN 60% AND 80%

OF THE TUBE WALL, o

THE CAUSE OF THE CRACKING IS ATTRIBUTED TO LARGE CONCENTRATION NA+

IN THE SECONDARY COOLANT, THE NA+ IS BELIEVED TO HAVE BEEN DEPOSITED IN THE HARDENED POROUS CRUD AND IN THE GAP BETWEEN THE TUBES AND THE SUPPORT PLATE.

o POLISHING SYSTEM TO REMAIN OFF-LINE UNTil PROBLEM FULLY UNDERSTOOD.

FOLLOWUP o

CORRECTIVE ACTION IS TO PLUG THOSE TUBES FOUND WITH GREATER THAN 20% TUBE WALL DEGRADATION (501 TUBES),

o WESTINGHOUSE AND CIEMAT WILL PROVIDE METALLURGICAL REPORT ON THElR EXAMINATION IN TWO WEEKS TO CONFlRM ROOT CAUSE, o

ACCIDENT ANALYSES WILL BE REANALYZED,

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ASCO PRESSURE SWilCHES JULY 25, 1988 PROBLEM POTENTIAL GEtiERIC PROBLEM WITH ASCO PRESSURE SWITCHES MODEL TG13A41.

CAUSE UNKN0hN SAFETY SIGNIFICANCE SWITCHES USED IN SAFETY SYSTEM (HPCS AND RCIC).

DISCUSSION i

o DISCOVERED WHEN RHR SHUTDOWN COOLING COULD NOT BE ESTABLISHED

(!NBOARD SUCTIOfi VALVE WOULD NOT OPEN).

0 FOLLOWING CYCLING SWITCH FUNCTIONED PROPERLY.

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ABOUT 44 SWITCHES PER UNIT INVOLVED OF WHICH 10 SWITCHES ARE SAFETY-RELATED.

o LICENSEE REQUIRES BY TECH. SPEC. TO CHECK SAFl?TY SYSTEM ONCE PER

MONIH, o

LICENSEE PLANNING 10 DISASSEMBLE SWITCHES TO DETERMINE FAILURE o

NKNOWN IF SIMILAR TO PROBLEM AT YANKEE-ROWE (DEC. '87) MODEL j

TL10A22 AND TM10A22.

FOLLOWUP o

REGION !! AND RVlB (VENDOR BRANCH) CONTINUING TO FOLLOW.

0 EAB TO DEIERMINE IF AN INFORMATION NOTICE IS WARRANTED.

1 CONTACT:

T. GREENE

REFERENCE:

MORNING REPORT 07/26/88 AND 50.72 # 12955

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DISK DEFORMATION RESULTING FR0!1 QUALIFICATION TEST O

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88-31 4

ARKANSAS UNIT 2 REACTOR COOLANT PUMP SEAL LEAK AUGUST J, 1988 P

PROBLEM LEAKAGE OF REACTOR COOLANT PUMP SEAL PACKAGE.

CAUSE MIDDLE SEAL PRESSURE SENSING LINE SHEARED.

SAFETY SIGNIFICANCE REACTOR COOLANT LOSS TO CONTAINMENT.

DISCUSSION O

ON AUGUST 1, 1988 AT 16:50 CDT HIGH/ LOW CONTROLLED BLEEDOFF ALARM ON "A" RCP.

O DECREASING PRESSURIZER LEVEL, INCREASED SUMP LEVEL, INCREASED CHARGING FLOW, HIGH CONTAINMENT RADIATION.

D REACTOR MANUAL TRIP AT 17:00 CDT.

O LEAK RATE INCREASED FROM APPROXIMATELY 2 GPM TO 35 GPM BEFORE COOLDOWN.

O UNUSUAL EVENT DECLARED BY LICENSEE AT 17:25 CDT BASED ON LEAK RATE GREATER THAN 10 GPM.

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BY 19:28 CDT REACTOR COOLED TO 532 F, DEPRESSURIZED TO 1550 PSIG LEAK INCREASED TO 40 PSIG.

O LEAK RATE DECREASED BY CONTINUED RCS COOLING AND DEPRESSURIZATION.

O CONTAINMENT ENTRY BY 00:38 CDT AUGUST 2, 1988.

O BROKEN SENSING LINE APPEARS TO HAVE CAUSED FAILURE OF MIDDLE, UPPER AND VAPOR SEALS FROM LACK OF FLOW.

O LOWER SEAL APPEARS TO HAVE FAILED FROM HIGH AP.

O RC PUMPS MANUFACTURED BY BYRON JACKSON, CONTROLLED LEAKAGE DESIGN WITHOUT SEAL INJECTION.

FOLLOWUP

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LICENSEE AND NRR CONTINUING TO INVESTIGATE CAUSE OF PUMP SEAL FAILURE.

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CONTACT:

W.

JENSEN RFFERFNCE:

50.72 #S 13055 AND 13056

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