ML20154P007

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Summary of Operating Reactor Events 860303 Meeting 86-06 W/Ornl Re Events Occurring Since 860224 Meeting.Followup Review Assignments,Status of Previous Assignments,List of Attendees & Viewgraphs Encl
ML20154P007
Person / Time
Site: Dresden, Mcguire, Palisades, Indian Point, Cooper, Arkansas Nuclear, Three Mile Island, North Anna, Turkey Point, River Bend, Haddam Neck, Farley, McGuire, LaSalle, 05000000
Issue date: 03/12/1986
From: Holahan G
Office of Nuclear Reactor Regulation
To: Harold Denton
Office of Nuclear Reactor Regulation
References
NUDOCS 8603200099
Download: ML20154P007 (25)
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Text

r MAR 12 gggi MEMORANDUM FOR: Harold R. Denton, Director Office of Nuclear Reactor Regulation FROM: Gary M. Holahan, Director Operating Reactors Assessment Staff

SUBJECT:

SUMMARY

OF THE OPERATING REACTORS EVENTS MEETING ON March 3,1986 - MEETING 86-06 On March 3,1986, an Operating Reactor Events meeting (86-06) was held to brief the Office Director, the Division Directors and their representatives on events which occurred since our last meeting on February 24, 1986. The list of attendees is included as Enclosure 1.

The events discussed and the significant elements of these events are presented in Enclosure 2. In addition, the assignment of follow-up review responsibility was discussed. The assignments made during this meeting and the status of previous assignments are presented in Enclosure 3.

Completion dates have been assigned for items in Enclosure 3. Each assignee should review Enclosure 3 with regard to their respective responsibilities.

Note that sevecal assignments are approaching the due date. Please be responsive and advise ORAB (D. Tarnoff, x29526) if the target completion date cannot be met.

Origbal Signed 3r.

Gary M. Holahan, Director Operating Reactors Assessment Staff Enclosures.

As stated cc w/ encl:

See next page DISTRIBUTION Central Files NRC PDR ORAS Rdg ORAS Members fi 0 A D/ S

& noff:dm RWessman GHolahan  !'

.t' j////86 ,2 /fs,,/86 ) / V86 /F g/>g 8603200099 86031D 4 9

PDR ADOCK 05000213 I S PDR

Harold R. Denton cc: D. Eisenhut D. Mcdonald J. Taylor L. Rubenstein C. Heltemes G. Lapinsky T. Murley, Reg. I V. Benaroya J. Nelson Grace, Reg. II J. Thoma -

J. Keppler, Reg. III J. Stolz R. D. Martin, Reg. IV A. Bournia J. B. Martin, Reg. V E. Adensam R. Starostecki, Reg. I F. Akstulewicz R. Walker, Reg. II C. Grimes C. Norelius, Reg. III T. Wambach E. Johnson. Reg. IV A. Thadani D. Kirsch, Reg. V L. Engle H. Thompson F. Miraglia R. Bernero T. Speis W. Russell T. Novak F. Schroeder W. Houston B. Sheron D. Ziemann J. Knight D. Crutchfield G. Lainas V. Benaroya W. Regan D. Vassallo E. Jordan E. Rossi R. Baer E. Weiss R. Hernan S. Showe S. Rubin

ENCLOSURE 1 LIST OF ATTENDEES OPERATING REACTORS EVENTS BRIEFING (86-06)

March 3,1986 NAME DIVISION NAME DIVISION D. Eisenhut NRR S. Long IE/DEPER S. Rubin AE00 K. Eccleston NRR/TOSB

0. Tarnoff NRR/0 RAS P. Milano IE/VPB W. Swenson NRR/ORAS E. Doolittle NRR/PWR-A R. Perfetti NRR/PWR-B L. Kelly NRR/0 RAS A. Thadani NRR/PWR-B D. Crutchfield NRR/PWR-B C. Miller NRR/PWR-B R. Jones NRR/PWR-B V. Benaroya NRR/F0B-A M. Caruso NRR/0 RAS V. Hodge IE/EGCB C. Grimes NRR/PWR-B D. Allison IE/EAB T. Wambach NRR/PWR-B G. Murphy ORNL R. Hernan NRR/PPAS R. Tripathi AE0D/R0AB L. Rubenstein NRR/PWR-2 G. Lapinsky NRR/PWR-A E. Leeds AE0D/R0AB D. Vassallo NRR/ DBL W. Regan NRR/PWR-B M. Goodman NRR/PWR-B R. Young IE/DEPER A. Bournia NRR/ DBL /PD-3 F. Akstulewicz NRR/PBR-A D. Mcdonald NRR/PWR-A R. Weller NRR/PWR-B
J. Thoma NRR/PBR-6 J. T. Beard NRR/0 RAS M. Virgilio NRR/0 RAS G. Holahan NRR/0 RAS B. Sheron NRR/DSR0 R. Bernero NRR/ DBL F. Schroeder NRR/DPL-B C. Ader OCM D. Mumanansky OCM M. Chiramal AE00 J. Stolz NRR/DPL-B E. Rossi NRR/DPL-A F. Miraglia NRR/DPL-B R. Baer IE/DEPER E. Jordan IE/DEPER J. Stone IE/VPB

e 4- ENCLOSURE 2 OPERATING REACTOR EVENTS BRIEFING 86-06 MARCH 3, 1986 TURKEY POINT UNIT 3 COMPONENT COOLING WATER SYSTEM VALVES MISPOSITIONED FOR THE LIFE OF PLANT WESTINGHOUSE PLANTS SINGLE FAILURE OF NUCLEAR INSTRUMENTATION INTERLOCK P-10 TMI UNIT 1 FAILURE OF REACTOR TRIP BREAKER SHUNT TRIP FEATURE l

LASALLE UNIT 1 AUTO START OF HPCS EDG DURING TEST OF STANDBY LIQUID CONTROL SYSTEM l

i HADDAM NECK DROPPED FUEL ELEMENT i

PALISADES REACTOR COOLANT SYSTEM VALVE LEAKAGE PROBLEMS 1

j OTHER EVENT OF INTEREST i NORTH ANNA UNIT 2 FAILURE OF MULTIPLE STEAM SAFETY RELIEF VALVES

PALISADES RECENT PLANT STARTUP EXPERIENCE (CYCLE 7) i

TURKEY POINT UNIT 3 - COMPONENT COOLING WATER SYSTEM VALVES MISPOSITIONED FOR THE LIFE OF THE PLANT FEBRUARY 24, 1986 (D. MCDONALD, NRR) 4 PROBLEM:

CCW DISCHARGE VALVE FROM THE RHR HEAT EXCHANGER THROTTLED TO 30% SINCE START OF PLANT OPERATION SAFETY SIGNIFICANCE:

CCW SYSTEM CANNOT MEET DESIGN BASIS FLOW (4000 GPM) WITH ONE PUMP (SINGLE FAILURE)

AFFECTS MULTIPLE SAFETY-RELATED SYSTEMS CIRCUMSTANCES:

UNIT 3 AT 100% POWER AND UNIT 4 IN REFUELING DISCHARGE VALVES THROTTLED IN 1972 DUE TO WESTINGHOUSE GENERIC CONCERN (TUBE VIBRATION IN RHR HEAT EXCHANGERS)

RHR HEAT EXCHANGERS MODIFIED IN 1974 - CCW DISCHARGE VALVES LEFT 30% OPEN ERROR IDENTIFIED DURING LICENSEE SAFETY SYSTEM REVIEW PiiASE 1 (JANUARY 31, 1986)

CCW NOT DECLARED INOPERABLE PENDING INVESTIGATION

! FOLLOWUP:

IDENTIFIED COMPONENTS WHICH COULD BE ISOLATED TO REDUCE j POST-LOCA CCW LOADS, PLANT OPERATION CONTINUED, WESTINGHOUSE AND PUMP MANUFACTURER CONFIRMED FLOW REQUIREMENTS (4000 GPM) AND THAT PUMP RUN0VT IS NOT A PROBLEM (FEBRUARY 21, 1986)

CCW VALVES POSITIONED TO FULL OPEN (FEBRUARY 21, 1986)

TEST PROCEDURE FOR SIMULATED LOCA CONDITIONS TO VERIFY THE ACCEPTABILITY OF SINGLE PUMP OPERATION (MARCH 3, 1986)

RECONSTITUTE CCW DESIGN BASIS (SAFETY SYSTEM REVIEW PHASE 2)

GENERIC ASPECTS:

0THER WESTINGHOUSE PLANTS (1972 CONCERN AND CORRECTIONS)

e.

l II. PHASE 1 ASSESSMENT l

e INTRODU'CTION: BASIS FOR SELECT SYSTEMS CHOSEN  ;

l

  • UlTAL SAFETY FUNCTIONS AS DESCRIBED IN F.S.A.R.
  • UITAL SUPPbHT

MAIN STEAM ISOLATION

  • INTAKE COOLING WATER I
  • COMPONENT COOLING WATER EMERGENCY COOLERS EMERGENCY FILTERS
  • EPERGENCY COOLING ,
  • m RGENe - R
  • SUPPORT SYSTEMS
  • DERGENCY FILTERS
  • CONTAlfRENTISOLATION 1.C.W.
  • VITAL D.C.

C.C.W.

VITAL A.C/D.C.

EMERGENCY A.C. POWER l

  • DUAL FUN CT10 4 SYSTEMS i (NORMAL AND EMERGENCY OPERATION) ,

l l

~

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PESIGN BASIS EFFORT PEOPLE: FPL-SITE ENGINEERING MANAGER BECHTEL WESTINGHOUSE PROCESS: RECONSTITUTE THE DESIGN BASIS FOR THE SELECT SYSTEMS THROUGH THE FOLLOWING PROCESS.

  • REVIEW SYSTEM BOUNDARIES
  • KEY SYSTEM DOCUMENT APPLICABILITY EVALUATION
  • REVIEW LICENSING CORRESPONDENCE AND COMMITMENTS
  • REVIEW DESIGN AND ACCIDENT ANALYSES

~

(ASSUMPTIONS,ETC.)

  • ESTABLISH SYSTEM PERFORMANCE CRITERI A USING DESIGN BASIS, ANALYSES,AND COMMITMENTS
  • EVALUATl0N SYSTEMS TESTING
  • VERIFY CONSISTENCY BETWEEN SYSTEM DOCUMENTS AND THE DESIGN BASIS (IE,):

DRAWINGS ,

PROCEDURES TECHNICAL SPECIFICATIONS Q-LIST VENDOR DOCUMENTS

  • RESOLVE INCONSISTENCIES AND MODIFY SYSTEM ,

AS REQUIRED PRODUCT:

  • RECONSTITI.TTED DESIGN BASIS CONSISTENT WITH l i LICENSING COMMITMENTS AND ANALYSES
  • CONSISTENCY BETWEEN DESIGN BASIS AND AS-BUILT l DRAWINGS i
  • VERIFICATION OF SYSTEM PERFORMANCE l

. - - - , _ . , . - - - - . - - - - _ . _ _ , . _ - - . . - _ - , , _ --,,,,,,,,,.,,,,__,,,.,,n.,,,,.,.,_ , , . . _ , , , , _ , . , _ . , . - - , ., , , _ _ ,.

WESTINGHOUSE PLANTS-SINGLE FAILURE OF NUCLEAR INSTRUMENTATION INTERLOCK P-10 FEBRUARY 26, 1986 (G. LAPINSKY, NRR)

PROBLEM: SOME NUCLEAR INSTRUMENTATION AND REACTOR PROTECTION FUNCTIONS

ARE VULNERABLE TO SINGLE FAILURE OF PERMISSIVE P-10 WHEN ONE POWER RANGE CHANNEL IS TRIPPED (PER TECH, SPECS.)

SIGNIFICANCE:

REDUCES MARGIN OF SAFETY FOR THREE ACCIDENTS IN FSAR ANALYSIS-BORON DILUTION, UNCONTROLLED ROD WITHDRAWAL, AND R0D EJECTION FROM 0% POWER GENERIC APPLICABILITY: NEARLY ALL W PLANTS AND STANDARD W TECH.

SPECS. AFFECTED DISCUSSION:

FIRST IDENTIFIED DURING ENGINEERING REVIEW BY ONE UTILITY W SAFETY REVIEW COMMITTEE DETERMINED THAT SITUATION CONSTITUTES "UNREVIEWED SAFETY OUESTION" (10 CFR 50.59) FOR MANY PLANTS TELEPHONE NOTIFICATION FEBRUARY 26, 1986; LETTER i

DATED FEBRUARY 27, 1986 WHEN ONE POWER RANGE CHANNEL IS TRIPPED (AS REQUIRED BY MOST TECH.

SPECS.), THE FAILURE OF A P-10 BISTABLE IN ANOTHER CHANNEL WILL CAUSE CERTAIN FUNCTIONS NOT TO BE AUTOMATICALLY REINSTATED WHEN POWER IS REDUCED BELOW 10%:

POWER TO SOURCE RANGE DETECTORS INTERMEDIATE RANGE HIGH FLUX TRIP POWER RANGE LOW-LEVEL TRIP FOLLOW-UP:

, WESTINGHOUSE HAS NOTIFIED ALL W REACTOR OWNERS BY TELEPHONE AND

FOLLOW-UP LETTER NRR ACTIONS
1. CONFIP'1 THAT W HAS PROPERLY DEFINED THE SCOPE OF THE PROBLEM IN TERMS OF POTENTIAL HARDWARE INTERACTIONS ,

AND EFFECTS ON FSAR EVENTS ANALYSIS

2. CONFIRM W INTERIM RECOMMENDATIONS
3. DETERMINE ADEQUACY OF DESIGN
4. DETERMINE ADEQUACY OF TECHNICAL SPECIFICATIONS
5. PREPARE IE INFORMATION NOTICE l

. _ . _ . _ _ _ . ._ _ ._ . . . - _ _ ~ . _ ._

I TMI UNIT 1 - FAILURE OF REACTOR TRIP BREAKER SHUNT TRIP FEATURE

FEBRUARY 26, 1986 (J, THOMA, NRR) i PROBLEM
ONE DC REACTOR TRIP BREAKER (RTB) (GE TYPE AK-2-25) i FAILED TO OPEN DURING SHUNT TRIP TEST SIGNIFICANCE: POTENTIAL FAILURE OF REACTOR PROTECTION SYSTEM CIRCUMSTANCES:

HISTORY i -

SEPTEMBER 26, 1985 - MECHANICAL FAILURE OF RTB (CB-11

AC BREAKER) CAUSED SHUNT RELAY TO BURN OUT JANUARY 1.4, 1986 - RTB CB-11 FAILED SHUNT TRIP TEST DUE TO IMFROPERLY INSTALLED RELAY BY CONTRACTOR FEBRUARY 26, 1986 PLANT AT 100% POWER CONDUCTING MONTHLY RTB SURVEILLANCE RTB CB-2 (DC BREAKER) SUCCESSFULLY TRIPPED ON UNDERVOLTAGE i

i -

RTB CB-2 FAILED TO OPEN ON SHUNT TRIP TEST l

ALL REDUNDANT RTBs TESTED SATISFACTORILY FINDINGS CB-2 FAILURE ATTRIBUTED TO MISALIGNED SECONDARY BREAKER i

CONTACTS THAT PROVIDE CONTROL POWER TO THE SHUNT TRIP DEVICE MOUNTED ON THE BREAKER CUBICLE CB-2 CONTACT MOUNTING BOARD WAS REALIGNED AND BREAKER TESTED SATISFACTORILY LICENSEE AND RESIDENT INSPECTOR VISUALLY INSPECTED ALIGNMENT ON THE OTHER RTBs - NO OTHER PROBLEMS IDENTIFIED 3 FOLLOWUP:

PROBLEM APPEARS TO HAVE BEEN THE RESULT OF " CUSTOM ALIGNMENT" l OF CONTROL CONTACTS OF THE BREAKERS BY THE CONTRACTOR WHEN SHUNT TRIP FEATURES WERE INSTALLED, DURING NEXT REFUELING OUTAGE, LICENSEE INTENDS TO STANDARDIZE TOLERANCES ON ALIGNMENTS, PROPOSED IE INFORMATION NOTICE BEING DRAFTED i

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l LASALLE UNIT 1 - AUTO START OF HPCS EDG DURING TEST OF STANDBY LIQUID CONTROL SYSTEM l

FEBRUARY 26, 1986 (A, BOURNIA, NRR)  !

l l

1 PROBLEM: AUTO START OF IB HPCS EDG DURING SLCS TEST SIGNIFICANCE:

i

  • POTENTIAL FOR UNDESIRABLE INTERACTION BETWEEN ATWS-MITIGATING SYSTEM AND ECCS. (WATER INJECTION WHEN LOWER LEVEL IS DESIRABLE) j DISCUSSION:

UNIT 1 IN REFUELING MODE; TESTING SLCS BY ;UJECTING CLEAN WATER INTO VESSEL

~

SLCS MAY NOT HAVE BEEN VENTED FOLLOWING MAINL" NANCE REACTOR VESSEL LEVEL INSTRUMENTATION TAP IS COMMON TO SLC FLOW PATH PIPING PRESSURE OF SLCS INJECTION FLOW UPSET REACTOR VE'SEL LEVEL i

INSTRUMENTATION SPURIOUS ECCS INITIATION AUTO START OF IB HPCS EDG; HPCS WAS OUT OF SERVICE AND THEREFCRE NO INJECTION FOLLOW-UP:

! REGION III IS FOLLOWING LICENSEE ACTION

NRR REVIEWING DESIGN DETAILS TO DETERMINE P0TENTIAL FOR SLCS i

l INJECTION RESULTING IN HPCS INITIATION DURING ATWS i

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HADDAM NECK-DROPPED FUEL ELEMENT FEBRUARY 26, 1986 (F AKSTULEWICZ, NRR)

PROBLEM: DURING REMOVAL OF UPPER CORE STRUCTURE A FUEL ASSEMBLY WAS LIFTED OUT OF THE CORE, SUBSEQUENTLY DROPPING BACK ONTO THE CORE.

SIGNIFICANCE:

POTENTIAL RELEASE OF RADI0 ACTIVITY. POTENTIAL DAMAGE TO OTHER FUEL ELEMENTS, DISCUSSION:

LICENSEE POSTULATES THAT A BURR OR OBSTRUCTION DEVELOPED BETWEEN THE FUEL ASSEMBLY LOCATING PIN AND THE ALIGNMENT HOLES IN THE FUEL ASSEMBLY UPPER N0ZZLES PROHIBITING MOVEMENT.

  • THE DROPPED ELEMENT HIT THE CORE BAFFLE SUPPORT PLATE (DENT FOUND)

AND CAME TO REST WITH ITS BASE ON THE TOP 0F THE REACTOR CORE AND ITS TOP 2 - 3 FEET BELOW THE VESSEL FLANGE,

  • REMOTE SURVEILLANCE IDENTIFIED SEVERAL DAMAGED FUEL RODS - NO SIGN OF BREACHED CLADDING - A DROPPED ASSEMBLY SLIGHTLY DEFORMED, NO INDICATION OF NOBLE GAS RELEASE AND NO CHANGE IN PRIMARY COOLANT RADI0 ACTIVITY LEVELS, FOLLOWUP:

ASSEMBLY REMOVED FROM REACTOR VESSEL ON MARCH 2, 1986 AND IS PRESENTLY STORED IN THE REFUELING POOL, WESTINGHOUSE (REFUELING VENDOR) AND BABC0CK & WILC0X (FUEL VENDOR)

ASSISTING LICENSEE IN DETERMINING SPECIAL PRECAUTIONS REQUIRED FOR PERMANENT SAFE STORAGE OF DAMAGED ASSEMBLY, REGION I CLOSELY MONITORING LICENSEE'S REFUELING / ACTIVITIES,

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PALISADES - REACTOR COOLANT SYSTEM VALVE LEAKAGE PROBLEMS FEBRUARY 20, 1986 (R. M. YOUNG, IE)

PROBLEM: SAFETY INJECTION AND CVCS VALVE LEAKAGE SIGNIFICANCE:

CHRONIC LEAKAGE PROBLEMS WITH THE POTENTIAL FOR OVERPRESSURIZATION OF LOW PRESSURE SYSTEMS DISCUSSION:

APPR0XIMATELY 3.5 GPM RCS LEAK ABOUT 2.5 GPM THROUGH TWO SI TANK ISOLATION CHECK VALVES AND CONTROL VALVES ABOUT 1.0 GPM THROUGH DIVERTER VALVE IN CVCS LETDOWN LINE FOLLOW-UP:

MANIPULATION OF LETDOWN SYSTEM (2-20-86) RESULTED IN BLOWN PACKING ON MANUAL VALVE IN CVCS AND AN ADDITIONAL 0.36 GPM LEAKAGE

' RESIDENT INSPECTOR FOLLOWING LICENSEE ACTIVITY TO REPAIR LEAKING COMPONENTS

- 1A LOOP CHECK VALVE RESEATED; LEAK TERMINATED EFFORTS TO RESEAT 2B LOOP CHECK VALVE UNSUCCESSFUL; HOWEVER PRESSURE CONTROL VALVE, MANUAL VALVE ISOLATED; LEAK RATE REDUCED TO ABOUT 0.4 GPM; LEAK CONTINUES REGION III 0FFICE HAS REQUESTED THAT LICENSEE REPAIR REMAINDER OF LEAKING COMPONENTS DURING NEXT SHUTDOWN COMPUTER SEARCH OF 10 CFR 50.72 REPORTS IDENTIFIED 19 RCS LEAKAGE RELATED OCCURRENCES SINCE MARCH 1985 l

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NORTH ANNA, UNIT 2 - FAILURE OF MULTIPLE MAIN STEAM SAFETY RELIEF VALVES FEBRUARY 21, 1986 (L. ENGLE, NRR)

PROBLEM:

SURVEILLANCE TESTING IDENTIFIED 8 0F 15 MAIN STEAM CODE SAFETY VALVES FAILED TEST ON-SITE VERSUS OFF-SITE TEST METHODS MAY CAUSE DISCREP-ANCIES IN THE CALIBRATION AND DETERMINATION OF VALVE SETPOINTS SIGNIFICANCE:

LOSS OF MAIN STEAM SYSTEM OVERPRESSURE PROTECTION POTENTIAL GENERIC APPLICABILITY DISCUSSION:

IN MODE 4, MAIN STEAM CODE SAFETY VALVES TESTED WITH AIR PNEUMATIC DEVICE AND AT HOT CONDITION IN MAIN STEAM LINES TEST RESULTS IDENTIFIED 8 OUT OF 15 VALVES FAILED TO LIFT DURING TEST (BEYOND 1% TOLERANCE)

VALVES BEING SENT TO 0FF-SITE TEST FACILITY FOR FURTHER TESTING FOLLOW-UP:

LICENSEE ACTIONS SENDING 10 VALVES TO AN INDEPENDENT TEST FACILITY FOR DETERMINATION OF AS-FOUND SETPOINT AND CALIBRATION 8 VALVES WHICH FAILED TO LIFT WITH PNEUMATIC ASSIST METHOD 1 VALVE WHICH LIFTED PROPERLY AFTER ADJUSTMENT 1 VALVE WHICH WAS FOUND TO BE ACCEPTABLE AS-IS EVALUATING SAFETY CONSEQUENCES AND IMPLICATIONS OF THE EVENT NRC REGION II FOLLOWING UP ON OFFSITE TESTING 0F VALVES 4

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NORTH ANNA UNIT 2 MAIN STEAM SAFETY VALVES PNEUMATIC ASSIST TEST METHOD RECOMMENDED BY VALVE MANUFACTURER FOR IN-SITU TESTING DEVICE UTILIZES AN AIR CYLINDER AND PISTON ARRANGEMENT WHICH IS CLAMPED TO THE VALVE STEM AND WHICH COUNTERACTS THE VALVE SPRING FORCE WHEN AIR IS SUPPLIED LIMITED BY STEAM GENERATOR PRESSURE AND AIR SUPPLY TO VALVE LIFT PRESSURES LESS THAN ABOUT 1160 PSIG TEST METHODS USED DURING PREVIOUS OUTAGES UNIT 1 UNIT 2 MOST RECENT REFUELING OUTAGE PNEUMATIC ASSIST PNEUMATIC ASSIST PREVIOUS REFUELING OUTAGE PNEUMATIC ASSIST FULL STEAM PRESSURE AT INDEPENDENT TEST FACILITY ALL OTHER OUTAGE PNEUMATIC ASSIST PNEUMATIC ASSIST

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FAILED TO LIFT AT THIS PRESSURE

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PALISADES - RECENT PLANT STARTUP EXPERIENCE (CYCLE 7)

(T. WAMBACH, NRR)

PRESENT STATUS: CLOSED ON-LINE AT 12:55 P.M. MARCH 3, 1986. HOLDING AT 25% POWER TO RUN TESTS.

PROBLEMS DURING THIS STARTUP:

CONTAINMENT VALVE LEAKS SI AND CVCS VALVE LEAKAGE

  • CONTAINMENT RAD. MONITORS STUCK CONTROL R0D LIFTED S/G SAFETY VALVES VACUUM AND CHEMISTRY
  • CONTINUING DURING OPERATION

REACTOR SCRAM SUl1 MARY WEE!'. ENDING 03/02/86 I. PLANT SPECIFIC DATA DATE SITE UNIT POWER RPS CAUSE COMPLI- YTD CATIONE TOTAL 02/27/E6 COOPER 1 70 A EDUIP/MFWPS ND 1 02/28/E6 INDIAN PCINT 3 100.A EQUIo/TURB GEN NO 1 02/28/86 FARLEY 1 100 A EQUIP / ROD DROP ND 1 03/01/86 RIVER BEND 1 38 A EQUIP /MFWP NO 4 O

II. COMPARISDN OF WEEELY CTATIETICS WITH INDUcTRY AVER ^ GEE SCRAMS FOR WEEK ENDING 03/02/86 SCRAM CAUSE PCWER NUMBER NUMBER 19ES OF NORMALIZED (6) WEEKLY SCRAMS (5) TO 1985 AVERAGE

'3) (4)

  • POWER '15%

EQUIP. RELATED >15% 4 4.0 5.4 (68%)

PERS. RELATED(7) >15% 0 0.0 2.0 (25%)

OTHER(8) >15% 0 0.0 0.6 ( 7%)

  • Subtotal ** .

4 4.0 8.0

  • FOWER <15%

_EOUIP. RELATED <15% 0 0.0 1.3 (54%)

FERS. RELATED <15% 0 0.0 0.9 (3E%)

OTHER <15% 0 O.O O.2 ( 8%)

4* Eubtotal **

O O.0 2.4

      • Total 4tt 4 4.0 10.4 MANUAL VS AUTO SCRAMS TYPE NUMBER NUMBER 1985 OF NORMALIZED WEEKLY SCRAMS TO 1985 AVERAEE MANUAL SCRAMS O O.0 1.0 AUTOMATIC SCRAMS 4 4.0 9.4 l

NOTES

1. PLANT SPECIFIC DATA BASED ON INITIAL REVIEW 0F 50.72 REPORTS FOR THE WEEK 0F INTEREST. PERIOD IS MIDNIGHT SUN. THROUGH MIDNIGHT SUN. SCRAMS ARE DEFINED AS REACTOR PROTECTIVE ACTUATIONS WHICH RESULT IN R0D MOTION, AND EXCLUDE PLANNED TESTS OR SCRAMS AS PART OF PLANNED SHUTDOWN IN ACCORDANCE WITH A PLANT PROCEDURE.
2. REC 0VERY COMPLICATED BY EQUIPMENT FAILURES OR PERSONNEL ERRORS UNRELATED TO CAUSE OF SCRAM ,
3. 1985 INFORMATION DERIVED FROM RECENT ORAS PRELIMINARY STUDY OF UNPLANNED REACTOR TRIPS IN 1985. WEEKLY DATA DETERMINED BY TAKING TOTAL TRIPS IN A GIVEN CATEGORY & DIVIDING BY 52 WEEKS / YEAR
4. IN 1985, THERE WERE AN ESTIMATED TOTAL OF 541 AUTOMATIC &

MANUAL UNPLANNED REACTOR TRIPS AT 93 REACTORS (HOLDING FULL POWER LICENSES). THIS YIELDS AN AVERAGE RATE OF 5.8 TRIPS PER REACTOR PER YEAR & AN AVERAGE RATE OF 10.4 TRIPS PER WEEK FOR ALL REACTORS.

- 5. BASED ON 93 REACTORS HOLDING A FULL POWER LICENSE, AS OF 2/1/86

6. NORMALIZED VALUES ALLOW COMPARIS0N TO 1985 DATA BY MULTIPLYING ACTUAL 1986 VALUE BY: 93 REACTORS IN 1985 NUMBER OF REACTORS REPORTING
7. PERSONNEL RELATED PROBLEMS INCLUDE HUMAN ERROR, PROCEDURAL DEFICIENCIES, AND MANUAL STEAM GENERATOR LEVEL CONTROL PROBLEMS
8. "0THER" INCLUDES AUTOMATIC SCRAMS ATTRIBUTED TO ENVIRONMENTAL i CAUSES (LIGHTNING), SYSTEM DESIGN, OR UNKNOWN CAUSE. l l

i

ENCLOSURE 3 Page No. 1 0FERATING REACTORS EVENTS MEETING FOLLOWUP ITEMS AS OF MEETING E6-06 ON MARCH 3, 1986 (IN ASCENDINS MEETING DATE, NSSS VENDOR, FACILITY ORDER)

MEETING FACILITY RESF0NSISLE TASK DESCRIPT!DN SCHEDULE CLOSED DATE COMMENTS NUMBER / NSSS VENCORt DIVISION! COMPLET. BY DOCUMENT, MEETINS EVENT DESCRIP. INDIVIDUAL DATE(S) MEETINS,ETC.

CATE B5-13 TURNEY POINT 3 PSB /MILHCAN J . REVIEW ADEDUACY OF GOVERNOR 03/30/86 OPEN / /

08/13/E5 W /FCST-TRIF / DESIGN ON TURPINE DRIVEN AFW 01/05/86 LOSS De AFW 10/13/05

} PUMPS E6-03 AUI,NSAS 1 NRR /MIRAGLIA F REVIEW 54S15 FOR FACILITY 03/31/86 CPEN / /

0!/27/96 PW / DESIGN / ORIGINAL IESIGN MODIFICATION / /

DEFICIENCY IN AND EASIS FOR FURTHEP / /

I EMERSENCf MODIFICATION FROPOSED BY FEE 0 WATER SYSTEM LICENSEE.

I S6-03 AFKANSAS 1 IE /J0FDAM E IDENTIFY WHAT IS ACTUAL DESIGN 03/31/86 0 FEN / /

i 01/27/E: EW / DeSIEN / IN E1W PLANTS //

t IEFICIENCf IN / /

EMERGENCY

FEE N TER SYSTEM 86-03 TMI 1 NCR /H2LAHAN 5 CONS!CER SUPFLEMENTAL IE NOTICE 03/31/06 CPEN / /

01/27/06 PW / FAFT!AL !E /J0PtAN E ON FECENT TRIP EFEAkER PROBLEMS //

LOSS OF hMI / /

56-03 MC591RE DHFS/ZIEMANN D REVIEW NOVEMBER 1, 1985 M:5UIRE 03/31/26 0 FEN / /

1 01/27/26 W / STAE.T-UP / EVENT AND DETERMINE IF FAILURE //

I WITHIEEFiADED TO REFAIR VALVE MOTOR OFERATOR / /

HFS! SYSTEM FRIOR TO START-UF WAS IN l

V10LAT!0N CF TECH SPECS FE2UIREMENT .

26-04 DFE5 DEN 3 NR: /EERhERG E DISCUSS WITH BWRS OWNERS GROUP 03/31/26 0 FEN / /

02/10/86 GE / FIRE IN / THE FROBABILITY OF FDLYURETHANE //

I DRYbELL IGNITING DURINS OPERATION AND //

EIFANSION BAF A5FENDIX R COMPL!ANCE ISSUE E6-06 TURKEY F0 INT 3 NRR /HDLAHAN 6. SCHEDULE TURKEY POINT FOLLOWUP 03/15/86 OPEN / / PRESENTATION 03/03/86 W / COMPONENT / PRESENTATION BEFORE 03/18/86 // SCHEDULED FOR C00 LINS WATER // 03/10/86 SYSTEM VALVES MISF0 SIT!0NED I FOR LIFE OF PLANT I

60-06 W PLANTS IE / JORDAN E REVIEW FEFORTABILITY OF ITEMS 03/20/86 OPEN / /

I 03/03/96 W / SINGLE / SUCH AS SINGLE FliLUPE OF P-10 //

FAILURE OF INTERLOCR / /

NUCLEAR INSTRUMENTATION INTERLOCK F-10 i

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