ML20133J794

Corrected Summary of Operating Reactors Events Briefing 96-15 on 961211

ML20133J794
Person / Time
Issue date:	12/19/1996
From:	Chaffee A NRC (Affiliation Not Assigned)
To:	Martin T NRC (Affiliation Not Assigned)
References
OREM-96-015, OREM-96-15, NUDOCS 9701210140
Download: ML20133J794 (24)

Text

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. Dec _ m ber 19, 1996 i

l MEMORANDUM T0: Thomas T. Martin. Director

Division of Reactor Program Management i'

FROM: Alfred E. Chaffee, Chief (original signed by] l 1

Events Assessment and

, Generic Communications Branch  ;

q Division of Reactor Program Management i 1

i

SUBJECT:

OPERATING REACTORS EVENTS BRIEFING j i -ET00ER 30,1996 - BRIEFING 96-W/6' #

1 N C & n 8 6 2 f ly .

j On December 11, 1996, we conducted an Operating Reactors Events Briefing a

(96-15) to inform senior managers from offices of the ACRS, AE0D. RES. NRR and i ragional offices of selected events that occurred since our last briefing on 1 October 30, 1996. Attachment I lists the attendees. Attacheent 2 presents

) the significant elements of the discussed events.

Attachment 3 contains reactor scram statistics for the weeks ending November 3, November 10 November 17, November 24. and December 1. 1996.

Two significant events were identified for input into the NRC Performance

Indicator Program (Attachment 4).

1 i The statements contained in the attached briefing slides represent the best j information currently available to the NRC. Future followup could produce new l

information that may alter the NRC's current view of the events discussed.

Attachments: As stated (4) i cc w/atts:

. See next page

! CONTACT: Kathy Gray. NRR (301) 415-1166

}

DISTRIBUTION: (w/atts)

Central-Files ~

PUBLIC LKilgore. SECY PECB:R/f l

)

DOCUMENT NAME: G:\KAG\0RTRANS To receive a copy of this document, indicate in the bos: "C" = Copy without attachment / enclosure 'E' = Copy with attachment / enclosure *N = No copy omce etce ,, le rece l ecce _e ecce l c/etcs _,Is NAME KGray:kag $ Q, _ SKoenick g Esennerf b ?) eGoodwin Q g AChaffee W DATE 12//D /96 i 12/ /[, /96 12/ g/Sy 12/ Q /96 12//} /96 _l QFFICIAL REC JRD COPY g Q ,dj 150!23 9 7-M 9701210140 961219 NfiC RLE DEillia 00PV PDR ORG NRRA ihts b ' ;

PDR Q E' in ~la *

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l 4

cC:

j F. Miraglia. NRR l0-12G18)

F. Gillespie. NRR (0-12G18)

R. Zimmerman. NRR (0-12G18) i A. Thadani. NRR (0-12G18)

8. Sheron. NRR (0-12G18)

S. Varga. NRR (0-14E4)

J. Zwolinski NRR (0-14H3)

J. Roe. NRR (0-13E4)

E. Adensam. NRR (0-13E4)

G. Lainas. NRR (0-7D26)

G. Holahan. NRR (0-8E2) I M. Virgilio. NRR (0-8E2) '

D. O'Neal. NRR (0-10E4) 1

8. Boger. NRR (0-9E4) I M. Markley. ACRS (T-2E26) l E. Jordan. AE00 'T-4D18) 1 C. Rossi. AE00 (T-4A9) i F. Congel AE00 (T-4D28)

R. Barrett. AEOD (T-4A43) s S. Rubin. AEOD (T 4D28)

M. Harper. AE00 (T-4A9)

W. Leschek. AE00 (T-4A9)

V. McCree. EDO (0-17G21)

J. Gilliland. PA (0-2G4)

D. Morrison. RES (T-10F12)

W. Hill. SECY (0-16G15)

H. Miller. Region I l R. Cooper. Region I S. Ebneter. Region II E. Merschoff. Region II S. Vias. Region II l A. Beach Region III I J. Caldweli Acting. Region III L. Callan. Region IV J. Dyer. Region IV K. Perkins. Region IV/WCF0  ;

G. Fader. INPO J. Zimmer. DOE

Sa l

LIST OF ATTENDEES OPERATING REACTORS EVENTS FULL BRIEFING (96-15) 1 DECEMBER 11, 1996

] NAME OFFICE NAME OFFICE l E. Benner NRR R. Meyer RES

K. Gray NRR G. Marcus NRR R. Dennig NRR E. Kendrick NRR S. Koenick NRR S. Sanchez NRR E. Goodwin NRR C. Rossi AE00

N. Hunemuller NRR M. El-zeftawy ACRS l T. Martin NRR 1

TELEPHONE AT'ENDANCE i (AT ROLL CALL) i Reaions Resident Insoectors Region I {

Region 11 '

Region III Region IV l

Misc.

ATTACHMENT 1

I <

1 I

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OPERATING REACTORS EVENTS BRIEFING 96-15 LOCATION: 0-10811, WHITE FLINT l WEDNESDAY, DECEMBER 11, 1996 11:00 A.M.

)

MULTIPLE PLANTS CONCERNS ASSOCIATED j WITH THE USE OF HIGH 1 BURNUP FUEL DESIGNS AND I LONGER CORE OPERATING CYCLES i

i 1

4 PRESENTED BY:

EVENTS ASSESSMENT AND GENERIC COMMUNICATIONS BRANCH DIVISION OF REACTOR PROGRAM MANAGEMENT, NRR 2

96-15 MULTIPLE PLANTS CONCERNS ASSOCIATED WITH THE USE OF HIGH BURNUP FUEL DESIGNS AND LONGER CORE OPERATING CYCLES PROBLEM:

SEVERAL POTENTIALLY SIGNIFICANT EVENTS HAVE OCCURRED WHERE A CONTRIBUTING CAUSE WAS THE USE OF HIGH BURNUP FUEL DESIGNS ASSOCIATED WITH LONGER FUEL CYCLES.  !

CAUSE: I INADEQUATE REVIEW 0F THE IMPACT OF LONGER FUEL CYCLES ON FUEL DESIGN, OPERATION CHARACTERISTICS, OPERATING MARGIN, AND ACCIDENT ANALYSIS.

DISCUSSION OF INDIVIDUAL CONCERNS:

e AVG BURNUP HAS INCREASED (mwd /MTU = MEGAWATT-DAYS / METRIC TON URANIUM):

25,000 -> 42,000 mwd /MTU FOR BOILING WATER REACTORS l (BWRs) 36,000 + 46,000 mwd /MTU FOR PRESSURIZED WATER REACTORS (PWRs)

METALLURGICAL, MECHANICAL, AND CHEMICAL EFFECTS:

e FAILURE OF CONTROL RODS TO FULLY INSERT WOLF CREEK: THIMBLE TUBE DISTORTION FROM EXCESSIVE COMPRESSIVE LOADING SOUTH TEXAS PROJECT FOREIGN REACTOR EVENTS CONTACTS: E. BENNER, NRR/DRPM AIT: NO S. K0ENICK, NRR/DRPM

REFERENCE:

MULTIPLE SIGEVENT: NO

!~

~

MULTIPLE 96-15 e DISTINCTIVE CRUD PATTERN ON FUEL

" SPIDER WEB" PATTERN ON FUEL AT CRYSTAL RIVER AND THREE MILE ISLAND CONTRIBUTING FACTORS:

ELEVATED BORON CONCENTRATION AT BEGINNING OF CYCLE LOWER FLOW VELOCITY AT GAP BETWEEN FUEL ASSEMBLIES e FRETTING 0F FUEL RODS GRID-TO-ROD FRETTING FUEL ASSEMBLY VIBRATION INDUCED BY GE0 METRY OF " VANTAGE 5" FUEL e AXIAL POWER OFFSET AN0MALY 0FFSETS VARIED SIGNIFICANTLY FROM RELOAD PREDICTIONS (SEVERAL PWRs)

CONDITIONS CONCENTRATE BORON-LITHIUM COMPOUND IN  ;

CRUD LAYER OF FUEL ASSEMBLY UPPER SPANS. I CHANGES TO OPERATING STRATEGIES - OPERATOR KNOWLEDGE WEAKNESSES e BWRs ALLOW ACCESS TO HIGHER R0D LINES AT REDUCED CORE FLOW.

e PWR BEGINNING-0F-LIFE MODERATOR TEMPERATURE COEFFICIENT (MTC) MAY BE MORE POSITIVE DUE TO HIGH BORON CONCENTRATION.

e END-0F-LIFE MTC MAY BE MORE NEGATIVE CALVERT CLIFFS: REACTIVITY EXCURSION AND SCRAM AT LOW-POWER OPERATION DUE TO OVERFEED OF STEAM GENERATOR.

t

~

i MULTIPLE 96-15 e END-0F-LIFE FLUX SHAPE ARKANSAS NUCLEAR ONE: SCRAMMED DURING POWER ASCENSION WHEN CORE PROTECTION CALCULATORS EXCEEDED AXIAL SHAPE INDEX LIMITS.

t l e IMPACT OF BWR OPERATION AT REDUCED FLOW i -

CLINTON: RECIRC FLOW CONTROL VALVE FAILED FULLY OPEN, CAUSING POWER TO INCREASE TO 109%;

CONTRIBUTING FACTOR WAS OPERATION AT HIGHER FLOW l CONTROL LINE WITH REDUCED CORE FLOW.

1 CORE OPERATING LIMITS e BWR - CRITICAL POWER RATIO POWER UPRATES, MIXED CORES AND OPERATION IN MAXIMUM EXTENDED LOAD LINE LIMIT ANALYSIS REGION HAVE EFFECTIVELY REDUCED OPERATING MARGIN TO MINIMUM CRITICAL POWER RATIO (MCPR).

e PWR CORE OPERATING LIMIT - NUCLEAR ENTHALPY RISE HOT

! CHANNEL FACTOR FACTOR INCREASED DUE TO LOADING 0F HIGHER ENRICHMENT CORES AND REDUCED-LEAKAGE CORES l

e PR0XIMITY TO A THERMAL LIMIT LIMERICK: MAXIMUM AVERAGE PLANAR RATIO THERMAL LIMIT EXCEEDED CONTROL R0D BANKING SEQUENCE WAS NOT APPROPRIATE j FOR HIGH-ENERGY CORE 1 e UNANTICIPATED APPROACH TO A THERMAL LIMIT PECO UNITS EXPERIENCED INSTANCES WHERE LINEAR HEAT GENERATION RATE THERMAL LIMITS WERE UNEXPECTEDLY APPROACHED EFFECT OF INCREASED BURNUP ON THERMAL LIMIT NOT ANTICIPATED

1-l~

(' MULTIPLE .

96-15 e GENERIC BWR MCPR DERIVATION DEFICIENCY IN SOME INSTANCES, APPLICATION OF CYCLE SPECIFIC METHODOLOGY RESULTED IN INCREASED MCPR VALUE OVER THE APPROVED GENERIC METHODOLOGY DEFICIENCIES IN THE DESIGN AND ANALYSIS PROCESS e CORE RELOAD ANALYSIS DEFICIENCIES INAPPROPRIATE OMISSION OF PEAK CLAD TEMPERATURE  !

ANALYSIS AT PRAIRIE ISLAND AT PALO VERDE, PROCEDURES AND DESIGN-BASIS DOCUMENTATION INCONSISTENT WITH RELOAD ANALYSIS ASSUMPTIONS EVOLVED OVER SEVERAL FUEL CYCLES INSUFFICIENT CONSIDERATION OF BWR HIXED CORE DESIGN RESULTS IN UNANTICIPATED POWER OSCILLATIONS (25% -

49%) AT WNP 2 9x9 FUEL ASSEMBLIES OPERATED AT HIGHER THAN l

AVERAGE POWER AND LOWER THAN AVERAGE FLOW e DEFICIENCIES IN MODELING CORE RESPONSE MISPREDICTED SHUTDOWN MARGIN AT LIMERICK AND DUANE ARNOLD ESTIMATED CRITICAL POSITION ERROR AT WNP 2 DUE TO INAPPROPRIATE TRAINING ON NEW CORE MONITORING ,

SOFTWARE VENDOR HISLOADED BURNABLE POIS0N RODS AND USED INCORRECT PATTERN OF GAD 0LINIUM RODS FOR FLUX MAPPING INPUT DATA AT ROBINSON l

a 3

i.

MULTIPLE 96-15 EMERGENCY CORE COOLING SYSTEM (ECCS) EVALUATION MODELS l (LARGE-BREAK LOSS OF COOLANT ACCIDENT (LBLOCA))

l e 10CFR50.46 ACCEPTANCE CRITERIA

! ECCS COOLING PERFORMANCE MUST BE CALCULATED IN ACCORDANCE WITH AN ACCEPTABLE EVALUATION MODEL:

j -

PEAK CLADDING TEMPERATURE (PCT) < 2200 F l

MAX CLADDING OXIDATION < 0.17

• TOTAL CLADDING THICKNESS MAX HYDR 0 GEN GENERATION < 0.01

• HYP0THETICAL i

AMOUNT OF ENTIRE VOLUME OF CLADDING REACTED C00LABLE GE0 METRY LONG-TERM COOLING REPORTING REQUIREMENTS ANNUAL REPORT OF CHANGES AND/0R ERRORS TO COMMISSION 30 DAY REPORT FOR SIGNIFICANT CHANGES / ERRORS (0NE ,

TIME OR CUMULATIVE RESULTING IN 50 F)  !

e EXPERIENCE WITH SIEMENS LBLOCA ECCS EVALUATION MODEL  ;

JULY 1986: STAFF APPROVES SIEMENS (FORMERLY EXXON) l LBLOCA ECCS MODEL AUGUST 1991: MODEL WAS MODIFIED TO CORRECT FOR NONPHYSICAL BEHAVIOR AUGUST 1995: STAFF NOTED SIGNIFICANT CHANGES TO PCT FROM 1991 MODEL CHANGES JANUARY 1996 TO PRESENT: MEETINGS AND CORRESPONDENCE BETWEEN VENDOR, AFFECTED UTILITIES, AND STAFF WITH RESPECT TO MODEL CHANGES NOVEMBER 29, 1996: NRC ISSUES LETTER TO SIEMENS CONTAINING SAFETY EVALUATION THAT CONCLUDED 1991 MODEL UNACCEPTABLE AND 1986 MODEL NEEDED TO BE CORRECTED.

l' MULTIPLE 96-15

e BASIS FOR REJECTION OF MODEL CHANGES

[ 1) NONPHYSICAL BEHAVIOR OF HEAT TRANSFER COEFFICIENT (HTC) VS. REFLOOD RATE

2) INSUFFICIENT DATA IN RANGE OF CONCERN T0 1 DEMONSTRATE CONSERVATISM l 3) USE OF NEW QUENCH TIME METHODOLOGY NOT SUPPORTED BY .

l DATA COMPARISONS PWR REFLOOD: COOLANT INJECTION THROUGH DOWNCOMER

! INTO LOWER PLENUM AND SUBSEQUENTLY TO THE BOTTOM 0F i CORE HTC MODELS ARE BASED ON APPLICABLE EXPERIMENTAL i DATA, INCLUDING WESTINGHOUSE FLECHT TESTS l

AFFECTED PLANTS: PALISADES, HARRIS, MILLSTONE 2, j ST. LUCIE, COMANCHE PEAK, ROBINSON, AND KEWAUNEE l

PLANTS EXCEEDING 2200 F: COMANCHE PEAK AND i

ROBINSON ADJUSTED PEAKING FACTORS AND KEWAUNEE l (CURRENTLY SHUTDOWN) STILL EVALUATING e GENERAL ELECTRIC LBLOCA MODEL DURING AUDIT CONDUCTED BY HOPE CREEK, IDENTIFIED i

WEAKNESS IN VENDOR'S TRACKING 0F ERRORS / CHANGES EARLY 1996: TWO CHANGES REPORTED TO LICENSEE l

DURING AUDIT, LICENSEE DISCOVERED 3 MORE CHANGES i DATING BACK TO 1990 i -

CUMULATIVE AFFECTS EXCEEDED SIGNIFICANT CHANGE REPORTING CRITERIA (50 F)

LICENSEE VERIFIED CALCULATED PCT REMAINED BELOW 2200 F e LICENSEES HAVE IN SOME CASES RELIED ON VENDOR SUBMITTED REPORTS T0' SATISFY REQUIREMENTS OF 50.46

~

MULTIPLE 96-15 NRC ACTIONS:

e TWO ACTION PLANS:

(1) HIGH BURNUP FUEL: ASSESS FUEL PERFORMANCE FOR HIGH BURNUP FUEL AND EVALUATE ADEQUACY OF STANDARD REVIEW PLAN LICENSING ACCEPTANCE CRITERIA (2) CORE PERFORMANCE: ASSESS IMPACT OF RELOAD CORE DESIGN ACTIVITIES ON PLANT SAFETY THROUGH INSPECTIONS OF FUEL VENDORS, EVALUATION OF LICENSEE RELOAD ANALYSES, INDEPENDENT EVALUATION OF CORE PERFORMANCE INFORMATION, WITH REGIONAL TRAINING AND INTERACTION PLAN 10 LICENSEE INSPECTIONS (WITH FIVE VENDOR FOLLOWUPS) FOR FY97-98.

CONTINUE VENDOR INSPECTIONS (18 MONTH INTERVAL) e INTERACTION WITH INDUSTRY GROUPS e EVENT FOLLOWUP OF INDIVIDUAL EVENTS, i.e., MCPR, STUCK RODS e ISSUANCE OF SEVERAL GENERIC COMMUNICATIONS e CONDUCTED CORE PERFORMANCE WORKSHOP FOR INDUSTRY (10/96) e VENDOR INSPECTION BRANCH WILL LEAD INSPECTION AT SIEMENS (2/97)

1 BRIEFING 96-15 Figure 1 BEHAVIOR OF HEAT TRANSFER COEFFICIENT WITH RESPECT TO REFLOOD VELOCITY nonphysical behavior

,o / modeled by Siemens

= /

e >

o 7 ~N - ;

O ,/' -

N 5g m , -

j -

FLECHT data l cW m=

$N 753 eg I-1.0 1.77 Reflood Velocity (in/sec) 1: FLECHT (Full length emergency cooling heat transfer) experimental data ,

FIGURE 1

REACTOR ;CRM Reporting Period: 11/25/96 to 12/01/%

YTD YTD ABOVE BELOW YTD E.AI.1 PLANT & UNIT POWER M flyg [0MPLICATIONS j,1} jj] TOTAL 11/27/96 V0GTLE 1 100 SA Maintenance Error NO 2 0 2 11/27/96 CRAND GULF 1 0 Equipnent Failure NO 1 1 2 KITAGIINT 3 Note: Year To Date (YTC) Totals Include Events Within The Calendar Year Indicated By The End Date of The specified Reporting Period Page:1 12/12/96 ETS-10

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

4 I

1 4 .

) -

COMPARISON OF lEEELY SCRAM STATISTICS WITH INDUSTRY AVERAGES PERIOD ENDING 12/01/ %

1 i NUMBER 1996 3 95 1994 1993 1992 0F WEEKLY WEEKLY WEEKLY WEEKLY WEEKLY SCRAM CAUSE SCRAMS AVERAGE AVERAGE AVERAGE AVERAGE AVERAGE (YTD) j POWER GREATER THAN OR EQUAL 70 15%

i j EQUIPMENT FAILURE O 1.58 1.83 1.52 1.83 2.62 I

DESIGN / INSTALLATION ERROR 0 0.10 0.12 0.08 0.04 -

I OPERATING ERROR 0 0.06 0.15 0.21 0.27 0.31 MAINTENANCE ERROR 1 0.52 0.38 0.54 0.52 0.50

, EXTERNAL 0 0.15 0.21 0.17 0.13 -

OTHER 0 0.08 0.06 - 0.02 -

?

J

! $4 total 1 2.49 2.75 2.52 2.81 3.43 i,

POWER LESS THAN 15%

EQUIPMENT FAILURE 1 0.21 0.10 0.27 0.38 0.42 4 DESIGN / INSTALLATION ERROR 0 0.00 -

0.02 - -

4 OPERATING ERROR 0 0.08 0.13 0.08 0.13 0.15

! MAINTENANCE ERROR 0 0.06 0.08 -

0.02 0.08

) EXTERNAL 0 0.00 - - 0.04 -

j OTHER 0 0.00 - - -

I Subtotal 1 0.35 0.31 0.37 0.57 0.65

, TOTAL  ? 2.84 3.06 2.89 3.38 4.~8 I 1996 1995 1994 1993 1992 NO. OF WEEKLY WEEKLY WEEKLY WEEKLY WEEKLY

, SCRAM TYPE SCRANS AVERACE AVERAGE AVERAGE AVERAGE AVERAGE (YTD)

TOTAL AUTOMATIC SCRAMS 1 1.77 1.92 2.19 2.44 3.06 TOTAL MANUAL SCRAMS 0 1.06 1.13 0.69 0.94 1.02 3

TOTALS MAY DIFFER BECAUSE OF ROUNDING OFF i

l 1

Page: 1 12/12/96 ETS-14

REACTOR SCRAM Reporting Period: 11/18/% to 11/24/%

YTD YTD ABOVE BELOW YTD 11.4 TOTAL QA71 PLANT & UNIT POWER ILi [AV.51 COMPT! CATIONS 113 3 0 3 11/22/96 DIA8t0 CANYON 1 100 SA Equipment Falture NO period Note: Year To Oce (YTD) Totals Irclude Events Within The Catendar Year Indicated By The End Date Of The Specified Reporting 11/25/'6 ETS 10 Page:1

COMPARISON OF nEEKLY SCRAM STATISTICS WITM INDUSTRY AVERAGES PERIOD ENo!NG ,

11/24/96 NUMBER 1996 1995 1994 1993 1992 0F WEEKLY WEEKLY WEEKLY WEEKLY WEEKLY SCRAM CAUSE SCRAMS AVERAGE AVERAGE AVERAGE AVERAGE AVERAGE (YTO)

POWER GREATER THAN OR EQUAL TO 15%

EQUIPMENT FAILURE 1 1,62 1.83 1.52 1.83 2.62 DESIGN /thSTALLATION ERROR 0 0.11 0.12 0.08 0.04 -

OPERATING ERROR 0 0.06 0.15 0.21 0.27 0.31 MAINTENANCE ERROR 0 0.51 0.38 0.54 0.52 0.50 EXTERNAL 0 0.15 0.21 0.17 0.13 -

OTHER 0 0.09 0.06 -

0.02 -

Subtotal 1 2.54 2.75 2.52 2.81 3.43 POWER LESS THAN 15%

EQUIPMENT FAILURE O 0.19 0.10 0.27 0.38 0.42 DESIGN / INSTALLATION ERROR 0 0.00 - 0.02 - -

OPERATING ERROR 0 0.09 0.13 0.08 0.13 0.15 1 MAINTENANCE ERROR 0 0.06 0.08 -

0.02 0.08 l

EXTERNAL 0 0.00 - -

0.04 -

OTHER 0 0.00 - - -

Subtota! 0 0.34 0.31 0.37 0.57 0.65 TOTAL 1 2.88 3.06 2.89 3.38 4.08 1996 1995 1 994 1993 1992 NO. OF WEEKLY WEEKLY WEEKLY WEEKLY WEEKLY SCRAM TYPE SCRAMS AVERAGE AVERAGE AVERAGE AVERAGE AVERAGE (YTO)

TOTAL AUTOMATIC SCRAMS 1 1.79 1.92 2.19 2.44 3.06 TOTAL MANUAL SCRAMS 0 1.09 1.13 0.69 0.94 1.02 TOTALS MAY DIFFER BECAUSE OF ROUWolWG OFF Page: 1 11/25/96 ETS-14

_  ;.__ __ _ _ _ . _ _ _ . . _ . _ . . . _- _ _ _ m ._ _ . _ _ . _ _ . _ . _ _.

l l

REACTCR SCRAM Reporting Period: 11/11/96 to 11/17/96 YTD YTD s

A80VE SELOW YT l

i 9111 PLANT & UNIT M M M COMPLICATIONS $ M M

! 11/12/96 MONTICELLO 1 100 SM Maintenance Error NO 3 0 l,

l 11/12/96 NORTH ANNA 2 100 SA Equipment Faiture NO 1 0 1

11/16/96 SEGUOYAN 1 36 SM Equipment Falture NO 2 0 ,

11/17/96 CALVERT CLIFFS 2 100 SA Equipment Failure NO 2 0  ;

I I

I e

Note: Year To Date (YTD) Totals include Events Within The Calender Year Indicated By The End Date Of The Specified Reporting Pericx ETS 10 Page:1 11/20/9e-

l i

e

COMPARISON OF W EKLY SCRAM STATISTICS WITN INDUSTRY AVERAGES a

] PERIOD ENDING j 11/17/ % j i

NUMBER 1996 1995 1994 1993 1992 f

]

0F WEEKLY WEKLY EEKLY WEEKLY WEEKLY SCRMI CAUSE SCRAMS AVERAGE AVERAGE AVERAGE AVERAGE AVERAGE .

l 3 (YTD) 1 POWER GREATER THAN OR EQUAL TO ISS l l

EQUIPMENT FAILURE 3 1.63 1.83 1.52 1.83 2.6s DESIGN /INSTALIATION ERROR 0 0.11 0.12 0.08 0.04 - 1 i

OPERATING ERROR 0 0.07 0.15 0.21 0.27 0.31 l MAINTENANCE ERROR 1 0.52 0.38 0.54 0.52 0.50 l j EXTERNAL 0 0.15 0.21 0.17 0.13 -

l OTHER 0 0.09 0.06 -

0.02 -

)

l Subtotal 4 2.57 2.75 2.52 2.81 3.43 I i

POWER LESS THAN 15%

EQUIPMENT FAILURE O 0.20 0.10 0.27 0.38 0.42 DESIGN / INSTALLATION ERROR 0 0.00 -

0.02 - -

OPERATING ERROR 0 0.09 0.13 0.08 0.13 0.15 MAINTENANCE ERROR 0 0.07 0.08 - 0.02 0.08 EXTERNAL 0 0.00 - -

0.04 -

OTHER 0 0.00 - - - -

SQtotal 0 0.36 0.31 0.37 0.57 0.65 l

TOTAL '

2.93 3.06 2.89 3.38 4.08 '

1996 1995 1994 1993 1992 j NO. OF WEEKLY WEEKLY WEEKLY WEEKLY WEEKLY SCRAM TYPE SCRAMS AVERAGE AVERAGE AVERAGE AVERAGE AVERAGE (YTD)

TOTAL AUTOMATIC SCRAMS 2 1.80 1.92 2.19 2.44 3.06 TOTAL MANUAL SCRAMS 7 1.11 1.13 0.69 0.94 1.02 TOTALS MAY DIFFER BECAUSE OF ROUNDING OFF ETS-14 Page: 1 11/20/96

. REACTOR SCRAM Reporting Period: 11/04/96 to 11/10/%

YYD YTD ABOVE BELOW YTD '

9.A T.l PLANT & UWit PowR LyPI P [Ay1( COMPtlCATIONS 11% lig TOTA 11/05/96 NINE MILE PolNT 1 100 SA Equipnent Failure No 2 0 2l l

l l

i tota: Year To Date (YTD) Totals Include Events Within The Calendar Year Indicated By The End Date of The Specified Reporting period ETS 10 Page:1 11/12/96

9 COMPARISON OF WEEKLY SCRAM STATISTICS WITH INDUSTRY AVERAGES PERIOD EN0fMG _

11/10/ %

NUMBER 1996 1995 1994 1993 1992 0F WEEKLY WEEKLY WEEKLY WEEKLY WEEKLY SCRAM CAUSE SCRAMS AVERAGE AVERAGE AVERAGE AVERAGE AVERAGE (YTD)

POWER GREATER THAN OR EQUAL TO 15%

EQUIPMENT FAILURE 1 1.60 1.83 1.52 1.83 2.62 DESIGN / INSTALLATION ERROR 0 0.11 0.12 0.08 0.04 -

OPERATING ERROR 0 0.07 0.15 0.21 0.27 0.31 MAINTENANCE ERROR 0 0.51 0.38 0.54 0.52 0.50 EXTERKAL 0 0.16 0.21 0.17 0.13 -

OTHER 0 0.09 0.06 - 0.02 -

$4 total 1 2.54 2.75 2.52 2.81 3.43 POWER LESS THAN 15%

EQUIPMENT FAILURE O 0.20 0.10 0.27 0.38 0.42 DESIGN / INSTALLATION ERROR 0 0.00 -

0.02 - -

OPERATING ERROR 0 0.09 0.13 0.08 0.13 0.15 MAINTENANCE ERROR 0 0.07 0.08 -

0.02 0.08 EXTERNAL 0 0.00 - -

0.04 -

OTHER 0 0.00 - - -

Subtotal 0 0.36 0.31 0.37 0.57 0.65 TOTAL 1 2.90 3.06 2.89 3.38 4.08 1996 19h 1994 1993 1992 No. OF WEEKLY WEEKLY WEEKLY WEEKLY WEEKLY SCRAM TYPE SCRAMS AVERACE AVERACE AVERAGE AVERAGE AVERAGE (YTD)

TOTAL AUTOMATIC SCRAMS 1 1.80 1.92 2.19 2.44 3.06 TOTAL MANUAL SCRAMS 0 1.09 1.13 0.69 0.94 1.02 TOTALS MAY DIFFER SECAUSE OF ROUNOING OFF ETS-14 Page: 1 11/12/96

m ._.. ._._._ - _.. _ __. __ _ _ _ _. .. - _ . ._ . . _ , . __.-. .. _ , . . _.m_. ~._......-..,.____1_...

l

.. 1 1

4

~.

• REACT 0k SCRAM Reporting Period: 10/28/% to 11/03/%

YTD YTD A80VE BELOW YTD Q&Il PLANT & UNIT M M gegl[ CrutPLICATIONS jj} jj} IQ[Al 10/29/ % sa0WNS FERRY 2 100 SA Equipment failure No 2 0 2 11/02/M NOPE CREEK 1 29 SM Equipment Failure No 1 0 1 ll l

I Note: Year To Date (YTD) Totals include Events Within The Calendar Year Indicated By The Erd Date Of The Specified Reporting Period Page:1 11/06/ %

ETS 10

e-

.. l 1

COMPARISON OF WEEKLY SCRAM STATISTICS WITH INDUSTRY AVERAGES PERIOD EN0!NG 11/03/ %

NUMSER 1996 1995 1994 1993 1992 0F WEEKLY WEEKLY WEEKLY WEEKLY WEEKLY SCRAM CAUSE SCRAMS AVERAGE AVERAGE AVERAGE AVERAGE AVERAE (YTD) l POWER GREATER THAN OR EQUAL TO 15%

EQUIPMENT FAILURE 2 1.61 1.83 1.52 1.83 2.9 DESIGN / INSTALLATION ERROR 0 0.11 0.12 0.08 0.04 -

OPERATING ERROR 0 0.07 0.15 0.21 0.27 0.31 MAINTENANCE ERROR 0 0.52 0.38 0.54 0.52 0.50 EXTERNAL 0 0.16 0.21 0.17 0.13 -

OTHER 0 0.09 0.06 - 0.02 -

Subtotal 2 2.56 2.75 2.52 2.81 3.43 POWER LESS THAN 15%

EQUIPMENT FAILURE 0 0.20 0.10 0.27 0.38 0.42 DESIGN / INSTALLATION ERROR 0 0.00 0.02 - -

OPERATING ERROR 0 0.09 0.13 0.08 0.13 0.15 MAINTENANCE ERROR 0 0.07 0.08 - 0.02 0.08 EXTERNAL 0 0.00 - - 0.04 -

OTHER 0 0.00 - - - -

Subtotal 0 0.36 0.31 0.37 0.57 0.65 TOTAL 7 2.92 3.06 2.89 3.38 4.08 1996 1995 1994 1993 1992 No. OF WEEKLY WEEKLY WEEKLY WEEKLY WEEKLY jg AM TYPE. SCRAMS AVERAGE AVERACE AVERAGE AVERAGE AVERAGE (YTD) j 3.06 I TOTAL AUTOMATIC SCRAMS 1 1.82 1.92 2.19 2.44 TOTAL MANUAL SCRAMS 1 1.11 1.13 0.69 0.94 1.02 TOTALS MAY O!FFER BECAUSE OF ROUNDING OFF l

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1. PLANT SPECIFIC DATA BASED ON INITIAL REVIEW O? 50.72 REPORTS FOR THE

! WEEK OF INTEREST. PERIOD IS MIDNIGHT SUNDAY THROUGH MIDNIGHT SUNDAY.

1 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 l OPERATING LICENSE.

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

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

j 4. "OTHER" INCLUDES AUTOMATIC SCRAMS ATTRIBUTED TO ENVIRONMENTAL CAUSES i

(LIGHTNING), SYSTEM DESIGN, OR UNKNOWN CAUSE.

OEAB SCRAM DATA Manual and Automatic Scrams 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 and Automatic Scrams for 1991 ------------------ 206 Manual and Automatic Scrams for 1992 ------------------ 212 Manual and Automatic Scrams for 1993 ------------------ 175 i Manual and Automatic Scrams for 1994 ------------------ 150 Manual and Automatic Scrams for 1995 ------------------ 159 Manual and Automatic Scrams for 1996 --(YTD 12/01/96)-- 136 1

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OPERATING REACTOR PLA:TS SIGNIFICLNT EVENTS SORT >- Event Date QUERY > Event Type SIG & Close Out Date >= 10/22/% & Close Out Date <= 11/19/96 DATE OF 50.72 OR PL CT & UNIT EVE NT NUM8ER DESCRIPil0W OF EVENT CLOSEOUT SIGNIFICANCE BRigFING PRESENTER RECORO HCPE CIEEE 1 03/14/96 0 Progransnatic weaknesses resulted in violations and Reactor Protection System KOTZALAS M. HIGHLIGHT imposition of civit penalty. Classification based on repeated failures to plan agropriate testing of equipment following maintenance.

SEQUOTAM 2 10/11/9631138 Scram with couplications - inpolse pressure switch safety-Related Cooling System  %-13 K0ENICK S. HIGHLIGHT failure and auxiliary feedwater inoperable.

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