ML20056D834

Summary of Operator Reactors Event Meeting 93-18 on 930519

ML20056D834
Person / Time
Site:	Browns Ferry, Indian Point, Wolf Creek, Robinson
Issue date:	05/26/1993
From:	Chaffee A Office of Nuclear Reactor Regulation
To:	Grimes B Office of Nuclear Reactor Regulation
References
OREM-93-018, OREM-93-18, NUDOCS 9308180115
Download: ML20056D834 (20)

Text

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IRY 2 61933 MEMORANDUM FOR: Brian K. Grimes, Director Division of Operating Reactor Support FROM: Alfred E. Chaffee, Chief Events Assessment Branch Division of Operating Reactor Support

SUBJECT:

OPERATING REACTORS EVElITS BRIEFING MAY 19, 1993 - BRIEFlNG 33-18 On May 19, 1993, we conducted an Operating Reactors Events Briefing (93-18) to inform senior managers from offices of the ACRS, OE, NRR, RES, AEOD, and regional offices of selected events that occurred zince our last briefing on May 12, 1993.

Enclosure 1 list.s the attendees. Enclosure 2 presents the significant ele:aents of the discussed events.

Enclosure 3 contains reactor scram statistics for the week ending May 16, 1993. No significant events were identified for input into the NRC performance indicator program.

Alfred E. Chaffee, Chief Events Assessment Branch Division of Operating Reactor Support

Enclosures:

As stated DISTRIBUTION:

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T. Murley, NRR (12G18) B. Mozafari (PDII-1)  !

F. Miraglia, NRR (12G18) Acting PD (PDII-1) '

F. Gillespie, NRR (12G18) N. Conicella (PDI-1)

J. Partlow, NRR (12G18) R. Capra (PDI-1)  ;

S. Varga, NRR (14E4) W. Reckley (PDIV-2)  :

J. Calvo, NRR (14A4) S. Black (PDIV-2) j G. Lainas, NRR (14H3) T. Ross (PDII-4)

J. Roe, NRR (13E4) F.-Hebdon-(PDII-4) {

J. Zwolinski, NRR (13H24)  !

E. Adensam, NRR (13E4) 4 W. Russell, NRR (12G18)

J. Richardson, NRR (7D26)

A. Thadani, NRR (8E2)  ;

S. Rosenberg, NRR (10E4) i C. Rossi, NRR (9A2)  !

B. Boger, NRR (10H3)

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F. Congel, NRR (10E2) .:

D. Crutchfield, NRR (11H21)

W.

D.

Travers, NRR (11B19)

Coe, ACRS (P-315) l E. Jordan, AEOD (MN-3701)

G. Holahan, AEOD (MN-9112)

L. Spessard, AEOD (MN-3701) . }

K. Brockman, AEOD (MN-3206) ';

-S . Rubin, AEOD (MN-5219)  ;

M. Harper, AEOD (MN-9112)

G. Grant, EDO (17G21)

R. Newlin, GPA (2G5) i E. Beckjord, RES (NLS-007) i A. Bates, SECY (16G15)

G. Rammling, OCM (16G15) }

T. .

Martin, Region I  ;

W. Kane, Region I  ;

C. Hehl, Region I S. Ebneter, Region II I E. Merschoff, Region II S. Vias, Region II i J. Martin, Region III E. Greenman, Region III ,

J. Milhoan, Region IV _  !

B. Beach, Region IV ,

B. Faulkenberry, Region V i K. Perkins, Region V -l.

bec: Mr. Sam Newton, Manager Events Analysis Department  ;

Institute of Nuclear Power Operations  ;

~700 Galleria Parkway C Atlanta, GA 30339-5957  :

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MAY 2 61993 -

MEMORANDUM FOR: Brian K. Grimes, Director Division of Operating Reactor Support FROM: Alfred E. Chaffee, Chief I Events Assessment Branch  !

Division of Operating Reactor Support ,

SUBJECT:

OPERATING REACTORS' EVENTS BRIEFING- i MAY 19, 1993 - BRIEFING 93-18  ;

On May 19, 1993, we conducted an Operating Reactors Events Briefing (93-18) to inform senior managers'from offices of the ACRS, OE, NRR, RES, AEOD, and regional offices of. selected events  ;

that-occurred since our last briefing on May 12, 1993. . .  ;

Enclosure 1 lists the attendees. ' Enclosure 2 presents the~ i significant elements of the discussed events. i Enclosure 3 contains reactor scram statistics for the week ending  !

May 16, 1993. No significant events were identified for input i into the NRC performance indicator program. /

[ l Alfred E. Chaffee, Chief Events Assessment Branch-Division'of_ Operating-Reactor-Support '

Enclosures:

As stated cc w/ enclosures:

See next page +

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ENCLOSURE 1 -

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LIST OF ATTENDEES

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OPERATING REACTORS EVENTS FULL BRIEFING-(93-18) ,

MAY 19, 1993 .j NAME OFFICE NAME OFFICE A. CHAFFEE NRR G. ZECH NRR I T. KOSHY NRR R. ECKENRODE NRR ,

D. SKEEN NRR R. ZIMMERMAN NRP  !

K. GRAY NRR B. BOGER NRR  :

R. DENNIG NRR S. VARGA NRR .

E. GOODWIN NRR S. LONG NRR l M. CARUSO NRR F. MIRAGLIA NRR  !

R. JONES NRR M. VIRGILIO NRR i N. CONICELLA NRR E. ADENSAM NRR  !

J. ROE NRR B. MOZAFARI NRR  !

W. RECKLEY NRR R. ROBINSON RES S. BREWER NRR J. KAUFFMAN AEOD . ~

M. WOHL NRR J. IBARRA AEOD l S. BLACK NRR D. COE ACRS  :

S. CONGEL NRR J. ROSENTHAL AEOD l

T. KIM NRR J. LIEBERMAN OE +

J. WILLIAMS NRR G. HOLAHAN AEOD f

TELEPHONE ATTENDANCE ,

(AT ROLL CALL)

Reaions Resident Inspectors Region I H. B. Robinson Region II Indian Point 3 Region III ,

Region IV l Region V '

IIT/AIT Team Leaders Misc.

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ENCLOSURE-2 ^

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l OPERATING REACTORS EVENTS BRIEFING 93-18 ,

5 LOCATION: 10 B11, WHITE FLINT l WEDNESDAY, MAY 19, 1993, 11:00 A.M.

t H. B. ROBINSON ENTRY INTO REDUCED INVENTORY -

l INDIAN-POINT, UNIT 3 OPERATION WITHOUT RELIABLE LEVEL. I WOLF CREEK- INSTRUMENTS BROWNS FERRY, UNIT.2 MISCOMMUNICATION DURING REACTOR ,

~

VESSEL INSTRUMENTATION LINE EXCESS 1 CHECK VALVE TESTING l o

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i PRESENTED BY: EVENTS ASSESSMENT BRANCH  !

DIVISION OF-OPERATING' REACTOR l SUPPORT, NRR l

'_____________l

93-18 H. B. ROBINSON INDIAN POINT, UNIT 3 WOLF CREEK ENTRY INTO REDUCED INVENTORY OPERATION '

WITHOUT RELIABLE LEVEL INSTRUMENTS CAUSES o NITR0 GEN OVERPRESSURE IN REACTOR COOLANT SYSTEM (RCS).

o UNDETECTED LOOP SEALS IN LEVEL SENSING LINES.

SAFETY SIGNIFICANCE INADEQUATE RCS LEVEL CONTROL IN REDUCED INVENTORY OPERATION COULD LEAD TO LOSS OF SHUTDOWN COOLING AND POTENTIAL CORE UNC0VERY. ,

H. B. ROBINSON SEPTEMBER 12, 1992 CAUSE INCORRECT PROCEDURE ALLOWED NITROGEN OVERPRESSURE OF 5.0 PSIG IN RCS INSTEAD OF 0.5 PSIG. ,

SEQUENCE OF EVENTS SEPTEMBER 11, 1992 e "B" REACTOR COOLANT PUMP (RCP) #1 SEAL WAS TO BE REPLACED; THIS~ REQUIRED RCS LEVEL BE REDUCED TO 25" BELOW VESSEL FLANGE (-25").

CONTACT: T. K0 SHY, NRR/0EAB AIT: N/A l

REFERENCES:

IR 50-261/92-27, SIGEVENT: TBD l IR 50-286/93-09, AND l IR 50-482/93-03 l l

-i

.. MULTIPLE PLANTS 93-18

' SEPTEMBER 12, 1992 l e RESIDENT INSPECTORS DISCUSSED WITH ACTING OPERATION j MANAGER (AOM) PROBLEMS EXPERIENCED AT OTHER :l UTILITIES CONCERNING VESSEL WATER ~ LEVE

L. PROCEDURE

i USED TO VENT RCS FOR DRAINDOWN WAS SHOWN TO THE ,

RESIDENT INSPECTORS. j e DECISION MADE BY SHIFT SUPERVISOR (SS) TO DRAINDOWN l

~

WITH 5.0 PSIG NITROGEN OVERPRESSURE TO MINIMIZE j

. CORROSION PRODUCT FORMATION, AOM NOT INFORMED j (ALLOWED BY PROCEDURE GP-008). j e SENIOR REACTOR OPERATOR EXPRESSED CONCERN THAT  !

NITROGEN OVERPRESSURE WOULD CAUSE LEVEL ~

INSTRUMENTATION INACCURACIES. SS STATES THAT LEVEL INACCURACIES HAVE NOT BEEN SEEN IN THE PAST WHEN i USING THIS PROCEDURE. l 1

e SENIOR RESIDENT INSPECTOR (SRI) NOTED DRAINDOWN  !

WITH NITROGEN OVERPRESSURE AND RE-ITERATED LEVEL l INDICATION ACCURACY CONCERN.

• LICENSEE VENTED.. REACTOR COOLANT SYSTEM TO j ATMOSPHERE, TYGON TUBE LEVEL MEASUREMENT DROPPED l 10 FEET DUE TO DEPRESSURIZATION. j

' SEPTEMBER 13, 1992 j e RCP SEAL WORK REQUIRED FURTHER DRAINDOWN TO JUST- l AB0VE -36" (ENTRY POINT FOR MID-LOOP' OPERATING  !

CONDITION). DECISION:MADE TO MAINTAIN RCS: LEVEL AB0VE ENTRY POINT TO AVOID PROCEDURAL CONTROLS AND

- i EQUIPMENT REQUIREMENTS OF MID-LOOP OPERATION.  !

. 1 MULTIPLE PLANTS 93-18 e INSPECTORS NOTED THAT TYGON TUBES READ =2" BELOW LEVEL INSTRUMENT USED FOR THE DRAINDOWN (LI-404).

LOWEST LEVEL SEEN ON LI-404 WAS -34". INSTRUMENTS i (LI-403, LI-404) WERE NOT CALIBRATED.  ;

DISCUSSION SEPTEMBER 12, 1992 e WITH RCS PRESSURIZED, LEVEL INDICATION WOULD NOT COME ON SCALE UNTIL LEVEL WAS AT -120". MID-LOOP IS AT -72".  ;

• PREVIOUS REVISIONS OF GP-008 SPECIFIED MAXIMUM NITROGEN OVERPRESSURE OF 0.5 PSIG. CHANGE TO 5 PSIG APPEARS TO BE A TYP0 GRAPHICAL ERROR.

• IN 92-16, SUPPLEMENT 1, " LOSS OF FLOW FROM THE RESIDUAL HEAT REMOVAL PUMP DURING REFUELING CAVITY DRAINDOWN," HAD BEEN RECEIVED BY THE LICENSEE'S OPERATING EXPERIENCE FEEDBACK GROUP, BUT HAD NOT BEEN REVIEWED FOR APPLICABILITY. i e TRAINING RECORDS INDICATED THAT OPERATIONS PERSONNEL HAD READ WESTINGHOUSE DESCRIPTION OF IN 92-16, SUPPLEMENT 1. SS HAD NO RECOLLECTION OF CONTENT OF IN.

SEPTEMBER 13, 1992

• LICENSEE DID NOT WANT TO ENTER MID-LOOP CONDITIONS DUE TO OUTAGE SCHEDULE CONSTRAINTS. l e LICENSEE MAY HAVE INADVERTENTLY ENTERED MID-LOOP CONDITIONS DUE TO USE OF NONCONSERVATIVE LEVEL INDICATION.

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, HULTIPLE i

PLANTS 93-18 j INDIAN POINT UNIT 3 MARCH 19, 1993 1 CAUSE LOOP SEAL IN LEVEL INSTRUMENT REFERENCE LEG.

DISCUSSION l e THE ULTRASONIC LEVEL MONITORING (ULMS) INSTALLED IN THE l PAST OUTAGE FAILED. THE REPLACEMENT WAS BENCH  :

CALIBRATED, BUT REQUIRED GOING INTO MID-LOOP FOR FIELD l CALIBRATION. THEREFORE NOT DECLARED OPERABLE.

j

-o- A TEMPORARY TYGON TUBING (HL) CONNECTED TO HOT LEG WAS l ADDED AND PLACED IN SERVICE TO MONITOR LEVEL DOWN T0 t TOP OF HOT LEG (APPROX. 64 FEET). -.

o INTERMEDIATE LEG INDICATION (ILI) ALS0-- AVAILABLE SHOWED 3 TO-7 INCHES LOWER THAN THE TEMPORARY HOT-LEG LEVEL  !

INDICATOR WHILE-RCS WAS DRAINING. THE READING 1

STABILIZED WITHIN 6 INCHES AT 64 FEET. (DUE TO LOOP -!

SEALS IN ILI INSTRUMENT LINE). (THE LICENSEE'S ACCEPTANCE VALUE F 3 6 INCHES FOR THE HL-IL INDICATION j!

DIFFERENCE.)

l o THE HL INDICATOR WAS ISOLATED AT 64 FEET AND DRAIN DOWN- j CONTINUED. ~

]

o~ RCS WAS THEN DRAINED DOWN TO 62 FEET 10 INCHES (MID-LOOP). THE TWO INSTRUMENTS- (ULMS AND ILI) SHOWED A  !

DIFFERENCE OF 3 INCHES WITH ULMS LEVEL LESS THAN'ILI -

j LEVEL. (THE LICENSEE'S ACCEPTANCE VALUE WAS 1 INCH FOR THE ULMS-ILI INDICATION DIFFERENCE.)- ,

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MULTIPLE PLANTS 93-18 o THE LICENSEE ISOLATED THE ILI INSTRUMENT LINE, DRAINED 1 IT AND RETURNED TO SERVICE.

o THE IL INDICATOR INDICATED 5 INCHES LESS THAN ULMS ,

READING.

o WHILE THE LICENSEE WAS PREPARING TO ISOLATE THE ILI AGAIN, THE RESIDENT INSPECTORS QUESTIONED THE OPERABILITY OF LEVEL INSTRUMENTATION.

o THE LICENSEE MANAGEMENT CAME TO KNOW 0F THE LEVEL DISCREPANCIES AND DIRECTED TO RAISE THE RCS LEVEL UNTIL TWO LEVEL INDICATIONS WERE AVAILABLE.

o THE OPERATOR CONDUCTED A WALKDOWN AND LOCATED ANOTHER LOOP SEAL IN THE ILI LINE AND DRAINED IT.

o THE ILI READINGS AGREED WITH OTHER INDICATIONS. ,

o IT APPEARS THAT RESIDUAL HEAT REMOVAL OPERATING PARAMETERS WERE NORMAL AND THEREFORE THE RCS LEVEL MUST HAVE REMAINED AB0VE MID-LOOP.

l WOLF CREEK MARCH 9, 1993  !

PROBLEM LEVEL INSTRUMENT DEVIATIONS FROM UNDETECTED LGOP SEAL IN  !

THE REFERENCE LEG.

l I

MULTIPLE PLANTS 93-18 DISCUSSION i o ON MARCH 9, 1993, LICENSEE STARTED RCS DRAIN DOWN TO l

REMOVE VESSEL HEAD. l i

o FOUR HOURS INTO DRAIN DOWN, OPERATORS NOTICED 30-INCH l LEVEL DIFFERENCE BETWEEN TYG0N TUBE AND LEVEL TRANSMITTERS.  !

o WHEN A LOOP SEAL IN THE TYG0N HOSE WAS REMOVED THE DIFFERENCE IN LEVEL BECAME 16 INCHES.

o DRAINING CONTINUED UNTIL 18 INCHES BELOW REACTOR 1 FLANGE. (BASED ON LOWEST INDICATED LEVEL) o TROUBLE SHOOTING CONTINUED WHILE RCS LEVEL WAS -

MAINTAINED WITHIN 3 FEET BELOW FLANGE TO AVOID ENTERING  !

REDUCED INVENTORY.

o THE LEVELS AGREED WITHIN HALF INCH WHEN WATER IN I REFERENCE LEG FOR THE TRANSMITTERS WAS DRAINED.  !

I o LICENSEE ATTRIBUTED THE CAUSE OF THE LEVEL DIFFERENCE l TO WATER DROPLETS COLLECTED IN THE REFERENCE LEG FROM AN EARLIER OPENING 0F MID-LOOP LEVEL INSTRUMENT ROOT VALVE. 4

)

o ON MARCH 21, 1993 WITH THE VESSEL DEFUELED, THE RCS LEVEL WENT BELOW MID-LOOP FROM MAINTENANCE' ACTIVITY ON MOVs. THIS RESULTED FROM A TAGGING CONTROL ERROR. l l

4

. MULTI?LE PLANTS. -

7- 93-18. j FOLLOWUP (ALL EVENTS) i o REGIONS HAVE' ISSUED EWFORCEMENT ACTIONS. ESCALATED  :

ENFORCEMENT IS PLANNED AT INDIAN POINT, UNIT 3.

a t

o OEAB IS CONSIDERING THESE EVENTS FOR'A SUPPLEMENT TO i IN 92-16. .

i o STAFF ~ EVALUATING THE NEED FOR REQUIREMENTS ON DIVERSE i LEVEL INSTRUMENTS FOR MID-LOOP OPERATION. j a

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93-18 BROWNS FERRY, UNIT 2 MISCOMMbHICATION DURING REACTOR VESSEL .

INSTRUMENTATION LINE EXCESS CHECK VALVE TESTING MAY 11, 1993 PROBLEM REACTOR VESSEL EXCEEDED EXPECTED PRESSURE DURING TESTING 0F THE INSTRUMENTATION LINE EXCESS FLOW CHECK VALVES (MAR 0TTA VALVES).

CAUSE DUE TO MISCOMMUNICATION BETWEEN AN I&C TECHNICIAN AND A REACTOR OPERATOR (RO), A PRESSURE INDICATOR THAT HAD BEEN ISOLATED FOR TESTING WAS USED BY AN R0, RESULTING IN OPERATOR ACTION THAT INCREASED REACTOR PRESSURE.

SAFETY SIGNIFICANCE LACK 0F CLEAR COMMUNICATION BETWEEN CONTROL ROOM PERSONNEL 4

AND FIELD TECHNICIANS PERFORMING SURVEILLANCE TESTING MAY l UNNECESSARILY CHALLENGE PLANT SYSTEMS. ,

DISCUSSION o DURING A REFUELING OUTAGE, TFE ' '49EE HAD JUST.

COMPLETED A REACTOR VESSEL HYD' < IF4 '4ITH THE VESSEL FILLED TO THE FLANGE AND PRESSURALED TO 1034 PSIG AND INVENTORY TEMPERATURE AT 199'F.

4 CONTACTS: D. SKEEN, NRR/0EAB AIT: NO ,

T. ROSS, NRR/PDII-4 '

REFERENCE:

10 CFR 50.72 #25507 SIGEVENT: TBD I

BROWNS FERRY, UNIT 2 93-18

-o PRESSURE WAS REDUCED TO 1000 PSIG. LEVEL AND PRESSURE. j WERE BEING CONTROLLED BY THE CONTROL ROD DRIVE SYSTEM l AND REACTOR WATER CLEANUP SYSTEM. l 1

o DURING THE EVENING SHIFT, I&C TECHNICIANS TESTED THE ]

GROUP "A" REACTOR VESSEL INSTRUMENTATION LINE.MAROTTA  !

VALVES AND COMMENCED TESTING THE GROUP "B" VALVES AT j 21:53.

o CONTROL ROOM STAFF DISCUSSED THE. HYDRO TEST AND THE i MAR 0TTA VALVE TEST DURING THE PRE-SHIFT BRIEFING. THEY l BELIEVED THAT ONLY ONE. LEVEL / PRESSURE TRANSMITTER WOULD BE TAKEN OUT OF SERVICE AT A TIME. i i

o THE R0 ASSIGNED TO MONITOR REACTOR PRESSURE (HYDRO.RO) l THOUGHT HE WOULD BE INFORMED BY THE WORK DESK OPERATOR l WHEN THE INDICATOR HE WAS MONITORING WAS REMOVED FROM i L SERVICE FOR TESTING.  ;

i o A CONTRACTOR I&C TECHNICIAN IN THE CONTROL ROOM. l NOTIFIED THE. WORK DESK R0 THAT-GROUP "B" j INSTRUMENTATION WAS GOING TO-BE TAKEN OUT OF SERVICE. l THE R0-QUESTIONED THE TECHNICIAN AND'WAS TOLD.THAT-ONLY- l ONE INSTRUMENT AT A TIME WOULD-BE REMOVED FROM SERVICE j

.AND WAS SHOWN A STEP IN THE PROCEDURE INDICATING-THAT 1 LEVEL TRANSMITTER 2-LT-3-52 WOULD BE REMOVED.FIRST.

o -ALL OF THE GROUP "B" INSTRUMENTS WERE REMOVED FROM  !

SERVICE-BY CLOSING THE ISOLATION. VALVES FOR EACH-  !

INDIVIDUAL TRANSMITTER, INCLUDING THE ONE USED BY.THE j HYDRO R0. - 1 l

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BROWNS FERRY, UNIT 2 93-18 o THE HYDR 0 R0 SAW PRESSURE DROPPING AND THROTTLED THE REACTOR WATER CLEANUP LETDOWN VALVE TO MAINTAIN REACTOR PRESSURE. BOTH THE OPERATOR AND THE SHIFT SUPERVISOR NOTED THAT IT TOOK EXCESSIVE FLOW REDUCTION TO CONTROL l PRESSURE.

l i

o TWO MINUTES LATER, THE PLANT RECEIVED AN ALTERNATE R0D l INSERTION (ARI) SIGNAL, AND REACTOR RECIRCULATION PUMP l TRIP (THESE SIGNALS ACTUATE AT 1118 PSIG). THE ARI SIGNAL ALUWED AIR PRESSURE TO THE SCRAM DISCHARGE VALVES TO VENT OFF, CLOSING THE VALVES AND RESULTING IN l A SCRAM SIGNAL.

o ALL CONTROL RODS WERE FULLY INSERTED PRIOR TO THE SCRAM. ALL PLANT SYSTEMS RESPONDED NORMALLY. THE.  ;

REACTOR WAS DEPRESSURIZED T0 600 PSIG AS A RESULT OF )

THE SCRAM. i FOLLOWUP o THE MAIN STEAM RELIEF VALVES (SETPOINT 1105 PSIG) DID NOT LIFT DURING THE EVENT. A REPORT FROM THE VENDOR l INDICATES THAT THE VALVES ARE SET UP FOR NORMAL OPERATING PRESSURE (1020 PSIG) AND TEMPERATURE (550*F).

AT 200*F, THE SETPOINT WOULD BE 3.6% HIGHER (1145 PSIG)

THAN THE CALIBRATED SETPOINT.

o THE LICENSEE IS REMOVING ONE OF THE MSRVs AND WILL TEST  ;

IT TO ENSURE THAT THE VALVE WOULD OPERATE AS STATED BY l THE VENDOR. i e THE OPERATING CREW WAS RELIEVED FROM DUTY FOR ONE DAY WHILE THE EVENT WAS INVESTIGATED.

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.: BROWNS FERRY, UNIT 2 . 93-18  !

o A HUMAN. PERFORMANCE INSPECTION TEAM FROM AE0D WAS '

DISPATCHED-T0 THE. SITE. THE TEAM CONCLUDED THAT ,

SEVERAL FACTORS CONTRIBUTED TO THE EVENT:  ;

i v THE TEST PROCEDURE WAS INCONSISTENT  ;

i v THERE WAS A LACK OF CLEAR COMMUNICATION BETWEEN THE a TECHNICIANS AND THE WORK DESK OPERATOR  !

v THERE WAS A LACK.0F UNDERSTANDING OF THE l INSTRUMENTS BEING ISOLATED '

l NSTRUMEN S N HE C N OL R OM W S A WE KNESS IN  !

CONFIGURATION CONTROL -

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, BRIEFING 93-18 BROWNS FERRY 2 C0hTROL ROOM COMPONEh75 AFTECTED BY SURVEILLANCE INSTRUCTION PERFORMANCE ,

GROUP UNID DESCRIPTION IE ATION 2-LI-3-5SA REACTOR WATER IIVEL EMERGENCY SYSTEMS RANGE Panel 2-9-5 A 2-PI-3-74A REACTOR PRESSURE A Panel 2-9-3 2-LI-3-58AA REACTOR WATER LEVEL EMERGENCY SYSTDiS RANGE Panel 2-9-5 2-LI-3-53 REACTOR WATER LEVEL NORMAL RANGE LEVEL A Panel 2-9-5 2-XR-3-53 RX VESSEL LEVEL (NOTE 1) Panel 2-9-5 2-PR-3-53 RX PRESS (NOTE 1) Panel 2-9-5 2-LI-3-55 REACTOR WATER LEVEL SSUIDOWN PLOOD UP RANGE Panel 2-9-3 2-PI-3-54 REACTOR PRESSURE WIDE RANGE PRESS A Panel 2-9-5 B 2-11-3-206 REACTOR WATER LEVEL NORMAL RANGE LEVEL C Panel 2-9-5 2-PI-3-207 REACTOR PRESSURE WIDE RANGE PRESS C _ Panel 2-9-5 Panel 2-9-5 _)

~

$( 2-PI-3-207A 2-PDI-3-51 REACTOR PRESSURE RECIRC JET PUMP EEAD Panel 2-9-3 2-LI-3-52 REACTOR WATER LEVEL POST ACCIDENT RANGE Panel 2-9-3 LA-3-53 REACTOR WATER LEVEL ABNORMAL 2-LA-3-53 Panel 2-9-5 (2-KA-55-5A, Window B) '

2-PI-3-74B REACTOR PRESSURE B Panel 2-9-3 C 2-PR-3-53 RX PRESS Panel 2-9-5 .

2-LI-3-58B REACTOR WATER LEVEL EMERGENCY SYSTEMS RANGE Panel 2-9-5 l 2-LI-3-5BBB REACTOR WATER LEVEL EMERGENCY SYSTE515 RANGE Panel 2-9-5 l 2-LI-3-60 REACTOR WATER LEVEL NOR%L RANGE LEVEL B Panel 2-9-5 2-PI-3-61 RX PRESSURE WIDE RANGE PRESS B Panel 2-9-5 D 2-LI-3-62A REACTOR WATER LEVEL POST ACCIDEh7 RANGE Panel 2-9-3 2-LR-3-62 RX WATER LEVEL POST ACCIDEh7 RANGE Panel 2-9-3 LA-3-53 REACTOR WATER LEVEL AENORMAL 2-LA-3-53 Panel 2-9-5 i (2-KA-55-5A Window 8) l 2-IR-3-53 RX VESSEL LEVEL (NOTE 2) Panel 2-9-5  !

2-PR-3-53 RX PRESS (NOTE 2) Panel 2-9-5

• - Pats 50u Iumuro( osca ov Ameo R o.

.- ENCLOSURE 3 e

REACTDR SCRAM Reporting Period: 05/10/93 to 05/16/93 i

YTD VfD  ;

ABOVE BELOW VfD EiT[ PLANT E UNIT POWE R TYPE CAUSE COMPi t C ATI ONS 15} 15} TOTAL '

05/11/93 BYRON 2 97 SA Equipment f ailure - NO 1 0 1 05/14/93 NAlCH 1 5 SA Operating Error WO 1 1 2 05/16/93 HOPE CREEK 1 60 SA Equipment f ailure WO 1 0 1 1

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i Coter ' Year To Date (YTD) Totals Include Events Within The Calendar Year Indicated By The End Date of The specified Reporting Period [

I E15 10 rage:1 05/21/93 -

E-O ,

COMPARISON OF WEEKLY SCRAM $7ATISTICS WITH INDUSTRY AVERAGES -

o PERIOD ENDING 05/16/93 WUMBER 1993 1992 1991* 1990* 1989*

OF WEEKLY WEEKLY WEEKLY WEEKLY WEEKLY SCRAM CAUSE SCRAMS AVERAGE AVERAGE AVERAGE AVERACE AVERAGE (YTD)

POWER CRE ATER THAN OR EDUAL 1015%

EQUIPMENT FAILURE

• 2 2.1 2.6 2.9 3.4 3.1 DESIGN /1NSTALLATION ERROR

• 0 0.1 - - - -

OPERATING !RROR* O 0.3 0.2 0.6 0.5 1.0 MAINTENANCE ERROR

• 0 0.5 0.4 - - -

EKTERNAL* 0 0.2 - - - -

OTHER* 0 0.0 0.2 - -

0.1  ;

Subtetat 2 3.2 3.4 3.5 3.9 4.2 l

PcruER LESS THAN 15%

EQUIPMENT FAILURE

• 0 0.4 0.4 0.3 0.4 0.3 '

DESIGN / INSTALLATION ERROR

• O 0.0 - - - + -

OPERATING ERROR

• 1 0.2 0.1 0.2 0.1 0.3 MAINTENANCE ERROR

• O 0.0 0.1 - - -

EXTERNAL

• 0 0.0 - - - -

t OTHER* O 0.0 0.1 - - - '

Subtotat 1 0.6 0.7 0.5 0.5 0.6 i

TOTAL 3 3.8 4.1 4.0 4.4 4.8 1993 1992 1991 1990 1989 NO. OF WEEKLY WEEKLY WEEKLY WEEKLY- WEEKLY SCRAM TYPE SCRAMS AVERAGE AVERAGE AVERAGE AVERAGE AVERAGE (YTD)

TOTAL AUTOMATIC SCRAMS 3 2.6 3.1 3.3 3.2 3.9 ,

TOTAL MANUAL SCRAMS 0 1.2 1.0 0.7 1.2 0.9 i

TOTALS MAY DIFFER EECAUSE OF ROUNDING OFF l

• Detailed breakdown not in database for 1991 and earlier i

- EXTERNAL cause included in EQUIPMENT FAILURE -

I

- MAINTENANCE ERROR and DESIGN / INSTALLATION ERROR causes included in OPERATING ERROR

- OTHER cause included in EQUIPMENT FAILURE 1991 and 1990 ,

I 15 Page: 1 05/21/93 l

2, . . PLANT SPECIFIC DATA BASED ON INITIAL REVIEW OF 50.72 REPORTS FOR THE WEEK OF INTEREST. PERIOD IS MIDNIGHT SUNDAY THROUGH MIDNIGHT SUNDAY.

SCRAMS 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. THERE ARE 111 REACTORS HOLDING AN OPERATING LICENSE.

2. PERSONELL RELATED PROBLEMS INCLUDE HUMAN ERROR, PROCEDURAL DEFICIENCIES,.

AND MANUAL STEAM GENERATOR LEVEL CONTROL PROBLEMS.

3. COMPLICATIONS: RECOVERY COMPLICATED BY EQUIPMENT FAILURES OR PERSONNEL ERRORS UNRELATED TO CAUSE OF SCRAM.

4. "OTHER" INCLUDES AUTOMATIC SCRAMS ATTRIBUTED TO ENVIRONMENTAL CAUSES (LIGHTNING), SYSTEM DESIGN, OR UNKNOWN CAUSE.

OEAB SCRAM DATA Manual and Automatic Scrans for 1987 ------------------ 435 Manual and Automatic Scrams for 1988 ------------------ 291 Manual and Automatic Scrams for 1989 ------------------~252 Manual and Automatic Scrams for 1990 ------------------ 226 Manual 2nd Automatic Scrams for. 1991 ------------------ 206 Manua7 and Automatic Scrans for 1992 ------------------ 212 ~

Manual and Automatic Scrams for 1993 --(YTD 05/16/93)-- 73 4

m wwn, e-*