ML20132F089
| ML20132F089 | |
| Person / Time | |
|---|---|
| Issue date: | 12/19/1996 |
| From: | Chaffee A NRC (Affiliation Not Assigned) |
| To: | Martin T NRC (Affiliation Not Assigned) |
| References | |
| OREM-96-14, NUDOCS 9612240117 | |
| Download: ML20132F089 (24) | |
Text
.-
s~
Dece m ber 19, 1996 a
MEMORANDUM TO:
Thomas T. Martin Director l
Division of Reactor Program Management FROM:
Alfred E. Chaffee Chief [ original signed by)
Events Assessment and i
Generic Communications Branch Division of Reactor Program Management
SUBJECT:
OPERATING REACTORS EVENTS BRIEFING OCTOBER 30. 1996 - BRIEFING 96-14 On December 11, 1996. we conducted an Operating Reactors Events Briefing (96-15) to inform senior managers from offices of the ACRS. AEOD RES. NRR and regional offices of selected events that occurred since our last briefing on October 30. 1996. Attachment 1 lists the attendees. presents the significant elements of the discussed events. 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).
The statements contained in the attached briefing slides represent the best information currently available to the NRC.
Future followup could produce new information that may alter the NRC's current view of the events discussed.
Attachments: As stated (4) cc w/atts:
See next page CONTACT:
Kathy Gray. NRR (301) 415-1166 DISTRIBUTION:
(w/atts) i J Central M Jes
)
PUBLIC l
LKilgore. SECY t
1 PECB R/F
'DFOS i
l
.3~ 09 4-) CP DOCUMENT NAME:
G:\\KAG\\0RTRANS Ci To receive a copy of this docurnent, Indicate in the boa: "C" = Copy without attachrnent/ enclosure
'E" = Copy with attachmen I s re
- = oc y omca esce
_,,, Ie esca l
esce
,e esca l
c/rece
,l n
NAME KGray:ke9 %Mtif,
SKoenick M' esennerf7B eGoodwin $d, 4 AChaffee N DATE 12/ j(0 /96
()
12/ /6 /96 12/ gI95 ~
12/ p /96 12/ f) /96 OFFIG6AL HEC JMD COPY 9612240117 961219 PDR ORG NRRA PDR j
i cc:
F. Miraglia. NRR (0-12G18)
F. Gillespie. NRR (0-12G18)
R. Zimmerman NRR (0-12G18)
I A. Thadani. NRR (0-12G18)
B. 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)
M. Virgilio. NRR (0-8E2)
D. O'Neal. NRR (0-10E4)
- 8. Boger. NRR (0-9E4)
M. Markley. ACRS (T-2E26)
)
E. Jordan. AE00 (T-4D18)
C. Rossi. AEOD (T-4A9)
F. Congel. AEOD (T-4D28)
R. Berrett. AE00 (T-4A43)
S. Rubin. AEOD (T-4D28)
M. Harper. AEOD (T-4A9)
W. Leschek. AE00 (T-4A9) l V. McCree. E00 (0-17G21)
J. Gilliland. PA (0-2G4)
D. Morrison. RES (T-10F12)
W. Hill. SECY (0-16G15)
H. Miller. Region I R. Cooper. Region I S. Ebneter. Region II E. Merschoff. Region II S. Vias. Region II A. Beach Region III J. Caldwell. Acting. Region III i
L. Callan Region IV J. Dyer. Region IV K. Perkins, Region IV/WCF0 G. Fader. INPO J. Zimmer. DOE
LIST OF ATTENDEES i
OPERATING REACTORS EVENTS FULL BRIEFING (96-15)
DECEMBER 11, 1996 NAME OFFICE NAME OFFICE 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 T. Martin NRR TELEPHONE ATTENDANCE (AT ROLL CALL) 4 Reaions Resident Insoectors Region I Region II 4
Region III Region IV Misc.
l ATTACHMENT 1
OPERATING REACTORS EVENTS BRIEFING 96-15 LOCATION:
0-10B11, WHITE FLINT WEDNESDAY, DECEMBER 11, 1996 11:00 A.M.
MULTIPLE PLANTS CONCERNS ASSOCIATED WITH THE USE OF HIGH BURNUP FUEL DESIGNS AND LONGER CORE OPERATING CYCLES PRESENTED BY:
EVENTS ASSESSMENT AND GENERIC COMMUNICATIONS BRANCH DIVISION OF REACTOR PROGRAM MANAGEMENT, NRR ATTAOfBT 2
l
96-15 MULTIPLE PLANTS CONCERNS ASSOCIATED WITH THE USE OF HIGH BURNUP FUEL DESIGNS AND LONGER CORE OPERATING CYCLES PROBLEM:
i SEVERAL P0TENTIALLY SIGNIFICANT EVENTS HAVE OCCURRED WHERE A CONTRIBUTING CAUSE WAS THE USE OF HIGH BURNUP FUEL l
DESIGNS ASSOCIATED WITH LONGER FUEL CYCLES.
l CAUSE:
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-1 DAYS / METRIC TON. URANIUM):
25,000 + 42,000 mwd /MTU FOR BOILING WATER REACTORS i
(BWRs) 36,000 + 46,000 mwd /MTU FOR PRESSURIZED WATER 1
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: NQ S. K0ENICK, NRR/DRPM
REFERENCE:
MULTIPLE SIGEVENT: MQ
j MULTIPLE 96-15 e DISTINCTIVE CRUD PATTERN ON FUEL i
" SPIDER WEB" PATTERN ON FUEL AT CRYSTAL RIVER AND THREE MILE ISLAND CONTRIBUTING FACTORS:
ELEVATED BORON CONCENTRATION AT BEGINNING 0F CYCLE LOWER FLOW VELOCITY AT GAP BETWEEN FUEL i
ASSEMBLIES e FREIIING OF FUEL RODS GRID-TO-ROD FRETTING i
i FUEL ASSEMBLY VIBRATION INDUCED BY GE0 METRY OF " VANTAGE 5" FUEL e AXIAL POWER OFFSET ANOMALY 0FFSETS VARIED SIGNIFICANTO' FROM RELOAD PREDICTIONS (SEVERAL PWRs) i CONDITIONS CONCENTRATE BORON-LITHIUM COMPOUND IN CRUD LAYER OF FUEL ASSEMBLY UPPER SPANS.
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 (KTC) MAY BE MORE POSITIVE DUE TO HIGH BORON CONCENTRATION.
o END-0F-LIFE MTC MAY BE MORE NEGATIVE CALVERT CLIFFS: REACTIVITY EXCURSION AND SCRAM AT LOW-POWER OPERATION DUE TO OVERFEED OF STEAM GENERATOR.
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.
e IMPACT OF BWR OPERATION AT REDUCED FLOW CLINTON: RECIRC FLOW CONTROL VALVE FAILED FULLY OPEN, CAUSING POWER TO INCREASE TO 109%;
CONTRIBUTING FACTOR WAS OPERATION AT HIGHER FLOW CONTROL LINE WITH REDUCED CORE FLOW.
CORE OPERATING LIMITS e BWR - CRITICAL POWER RATIO POWER UPRATES, MIXED CORES AND OPERATION IN MAXIMUM EXTENDED LOAD LINE LIMIT ANALYSIS REGION HAVE EFFECTIVELY R. EDUCED OPERATING MARGIN TO MINIMUM I
CRITICAL POWER RATIO (MCPR).
)
e PWR CORE OPERATING LIMIT - NUCLEAR ENTHALPY RISE HOT 4
CHANNEL FACTOR FACTOR INCREASED DUE TO LOADING 0F HIGHER ENRICHMENT CORES AND REDUCED-LEAKAGE CORES i
e PR0XIMITY TO A THERMAL LIMIT LIMERICK: MAXIMUM AVERAGE PLANAR RATIO THERMAL LIMIT EXCEEDED CONTROL R0D BANKING SEQUENCE WAS NOT APPROPRIATE FOR HIGH-ENERGY CORE 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
/0ITICIPATED
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 MIXED CORE DESIGN RESULTS IN UNANTICIPATED POWER OSCILLATIONS (25% -
49%) AT WNP 2 9x9 FUEL ASSEMBLIES OPERATED AT HIGHER THAN 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 POISON RODS AND USED INCORRECT PATTERN OF GAD 0LINIUM RODS FOR FLUX MAPPING INPUT DATA AT ROBINSON
3 MULTIPLE 96-15 i
EMERGENCY CORE COOLING SYSTEM (ECCS) EVALUATION MODELS (LARGE-BREAK LOSS OF COOLANT ACCIDENT (LBLOCA))
e 10CFR50.46 ACCEPTANCE CRITERIA ECCS COOLING PERFORMANCE MUST BE CALCULATED IN ACCORDANCE WITH AN ACCEPTABLE EVALUATION MODEL:
PEAK CLADDING TEMPERATURE (PCT) < 2200 F MAX CLADDING OXIDATION < 0.17
- TOTAL CLADDING i
THICKNESS MAX HYDR 0 GEN GENERATION < 0.01
- HYPOTHETICAL AMOUNT OF ENTIRE VOLUME OF CLADDING REACTED C00LABLE GE0 METRY LONG-TERM COOLING REPORTING REQUIREMENTS ANNUAL REPORT OF CHANGES AND/0R ERRORS TO i
COMMISSION 30 DAY REPORT FOR SIGNIFICANT CHANGES / ERRORS (ONE l
TIME OR CUMULATIVE RESULTING IN 50 F) e EXPERIENCE WITH SIEMENS LBLOCA ECCS EVALUATION MODEL JULY 1986:
STAFF APPROVES SIEMENS (FORMERLY EXXON)
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.
MULTIPLE 96-15 e BASIS FOR REJECTION OF MODEL CHANGES 1)
NONPHYSICAL BEHAVIOR OF HEAT TRANSFER COEFFICIENT i
(HTC) VS. REFLOOD RATE 2)
INSUFFICIENT DATA IN RANGE OF CONCERN TO DEMONSTRATE CONSERVATISM 3)
USE OF NEW QUENCH TIME METHODOLOGY NOT SUPPORTED BY l
DATA COMPARISONS PWR REFLOOD:
COOLANT INJECTION THROUGH D0WNCOMER INTO LOWER PLENUM AND SUBSEQUENTLY TO THE BOTTOM 0F CORE l
HTC MODELS ARE BASED ON APPLICABLE EXPERIMENTAL l
DATA, INCLUDING WESTINGHOUSE FLECHT TESTS AFFECTED PLANTS:
PALISADES, HARRIS, MILLSTONE 2,
)
ST. LUCIE, COMANCHE PEAK, ROBINSON, AND KEWAUNEE PLANTS EXCEEDING 2200 F:
COMANCHE PEAK AND ROBINSON ADJUSTED PEAKING FACTORS AND KEWAUNEE (CURRENTLY SHUTDOWN) STILL EVALUATING l
e GENERAL ELECTRIC L8LOCA MODEL DURING AUDIT CONDUCTED BY HOPE CREEK, IDENTIFIED WEAKNESS IN VENDOR'S TRACKING 0F ERRORS / CHANGES EARLY 1996:
TWO CHANGES REPORTED TO LICENSEE DURING AUDIT, LICENSEE DISCOVERED 3 MORE CHANGES DATING BACK TO 1990 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 TO SATISFY REQUIREMENTS OF 50.46 4
l MULTIPLE 96-15 NRC ACTIONS:
e TWO ACTION PLANS:
2 (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 J
INSPECTIONS OF FUEL VENDORS, EVALUATION OF LICENSEE RELOAD ANALYSES, INDEPENDENT EVALUATION OF CORE i
PERFORMANCE INFORMATION, WITH REGIONAL TRAINING AND
{
INTERACTION PLAN 10 LICENSEE INSPECTIONS (WITH FIVE VENDOR l
FOLLOWUPS) FOR FY97-98.
CONTINUE VENDOR INSPECTIONS (18 MONTH INTERVAL) i e
INTERACTION WITH INDUSTRY GROUPS e
EVENT FOLLOWUP 0F INDIVIDUAL EVENTS, i.e., MCPR, STUCK RODS e
ISSUANCE OF SEVERAL GENERIC COMMUNICATIONS i
e CONDUCTED CORE PERFORMANCE WORKSHOP FOR INDUSTRY (10/96) 1 e
VENDOR INSPECTION BRANCH WILL LEAD INSPECTION AT SIEMENS (2/97)
BRIEFING 96-15 Figure 1 BEHAVIOR OF HEAT TRANSFER COEFFICIENT WITH RESPECT TO REFLOOD VELOCITY E
,e nonphysical behavior j
modeled by Siemens E
w N
[
/
\\
u e^
FLECHT data l u_
j E&
r as c Ab E3 t
I 53 1.0 1.77 i
Reflood Velocity (in/sec)
I 1: FLECHT (Full length emergency cooling heat transfer) experimental data i
FIGURE 1
. -. - _. ~. - _...,
.... -... _... -......... ~.. -. -..... -....
REACTOR SCRM Reporting Period: 11/25/96 to 12/01/96 ITD YTD A80VE BELOW YTD Q&Il PLMT & UNIT 25BER M
SE3[
COMPLICATIONS S
TOTAL 11/27/96 V0GTLE 1 100 SA Maintenance Error No 2
0 2
11/27/96 GRAND GULF 1 0
Equipment failure NO 1
1 2
)
I i
)
1 l
l i
l l
i i
1 1
l l'
i i
E ATTA0MNT 3 No2d: Year To Date (Y10) Totals include Events Within The Calendar Year Indicated By The Erd Date Of The Specified Reporting Period 3 ETS-10 Fage:1 12/12/96 2
.I i
- ~.
COMPARISON OF E EKLY SCRAM STATISTICS WITN INDUSTRY AVERAGES l
FERIOD EN0 LNG 12/01/96 NUMBER 1996 1995 1994-1993 1992 OF WEEKLY WEEKLY WEEKLY WEEKLY WEEKLY SCRAM CAUSE SCRAMS AVERAGE AVERAGE
.tVERAGE AVERAGE AVERAGE (YTD) j POWER GREATER THAN OR EQUAL 70 15%
i EQU:PMENT FAILURE O
1.56 1.83 1.52
- 1. 83 2.62 DESIGN / INSTALLATION ERROR 0
0.10 0.12 0.08 0.04 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 f
Subtotal 1
2.49 2.75 2.52 2.81 3.43 POWER LESS THAN 15%
EQUIPMENT FAILURE 1
0.21 0.10 0.27 0.38 0.42 DESIGN / INSTALLATION ERROR 0
0.00 0.02 OPERATlWG ERROR 0
0.08 0.13 0.08 0.13 0.15 MAINTENANCE ERROP 0
0.06 0.08 0.02 0.08 EXTERNAL 0
0.00 0.04 OTHER 0
0.00 Subtotal 1
0.35 0.31 0.37 0.57 0.65 TOTAL 2
2.84 3.06 2.89 3.38 A.*8 1996 1995 1994 1993 1992 No. OF WEEKLY WEEKLY WEEKLY WEEKLY WEEKLY SCRAM TYPE SCRAMS AVERAGE 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 TOTALS MAY DIFFER BECAUSE OF ROUNDING OFF ETS-16 Page: 1 12/12/96
3 REACTOR SCRAM Reporting Period 11/18/96 to 11/24/%
YTD YTD ABOVE BELOW YTD DAIX PLANT & UNIT Engj{
IJ3
$83[
COMPLICATIONS JH 13 19,LA1 11/22/96 DIABLO CANYON 1 100 SA Equipment Failure NO 3
0 3
Year To Date (YTD) Totals include Events Within The Calender Year Indicated By The End Date of The specified Reporting Period tote:
11/25!96 ETS 10 Page:1
1 t
e COMPARISON OF E EKLY SCRAM STATISTICS WITH IMOUSTRY AVERAGES PERIOD ENDING 11/24/96
~
NUMBER 1996 1995 1994 1993 1992 0F WEEKLY WEEKLY WEEKLY WEEKLY WEEKLY
$ CRAM CAUSE SCRAMS AVERAGE AVERAGE AVERAGE AVERAGE AVERAGE (YTC) l POWER GREATER TNAN OR EQUAL 70 15%
EQUIPMENT FAILURE 1
1.62 1.E3 1.52 1.E3 2.62 i
DESIGN / INSTALLATION ERROR 0
0.11 0.12 0.08 0.04 OPERATING ERROR 0
0.06 0.15 0.21 0.27 0.31 i
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 0
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 MAINTENANCE ERROR 0
0.06 0.08 0.02 0.08 EXTERNAL 0
0.00 0.04 OTHER 0
0.00 Subtotal 0
0.34 0.31 0.37 0.57 0.65 TOTAL 1
2.88 3.06 2.89 3.38 4.08 1996 1955 1994 1993 1992 NO. OF WEEKLY WEEKLY WEErLY WEEKLY WEEKLY SCRAM TYPE SCRAMS AVERAGE AVERAGE AVERAGE AVERAGE AVERAGE (YTD)
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 ROUNDING OFF ETS-16 Page: 1 11/25/96
t REACTOP SCRM Reportins Period: 11/11/96 to 11/17/96 YTD YTD A80VE BELOW YTD 9311 PLANT & UNIT M
TLP[
$33[
COMPLICAfl0NS jll jil igL.L 11/12/%
MONTICELLO 1 100 SM nalntenance Error No 3
0 3
11/12/96 NonTH ANNA 2 100 SA Equipment Failure NO 1
0 1
11/16/96 SEQUDYAN 1 36 SM Equipment Failure NO 2
0 2
11/17/96 CALVERT CLIFFS 2 100 SA Equipment Failure NO 2
0 2
Noto: Year To Date (YTD) Totals Include Events Within The Calendar Year Incicated By The End Date Of The Specified Reporting Period E75 10 Page:1 11/20/96
m
~
s COMPARISON OF lEEKLY SCRAM STATISTICS WITH INDUSTRY AVERAGES PERIOD EkolNG 11/17/96 NUMBER 1996 1995 1994 1993 1992 of WEEKLY WEEKLY WEEKLY WEEKLY WEEKLY SCRAM CAUSE SCRAMS AVERAGE AVERAGE AVERAGE AVERAGE AVERAGE (YTD)
POWER GREATER THAN OR EQUAL To 15%
EDUIPMENT FAILURE 3
1.63 1.83 1.52 1.83 2.o2 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 1
0.52 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 4
2.57 2.75 2.52 2.81 3.43 POWER LESS THAN 15%
EQUIPMENT FAILURE O
0.20 0.10 0.27 0.th 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 0.04 EXTERNAL 0
0.00 OTHER 0
0.00 1
subtotal 0
0.36 0.31 0.37 0.57 0.65 TOTAL 4
2.93 3.06 2.89 3.38 4.08 1996 1995 1994 1993 1992 No. OF WEEKLY WEEKLY WEEKLY WEEKLY WEEKLY
} CRAM 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 2
1.11 1.13 0.69 0.94 1.02 TOTALS MAY DIFFER BECAUSE OF ROUNolNG OFF ETS-10 Page: 1 11/20/96
.__m._
_ ~ _. _.. _.
_.._...m.....___.m..._m..__
m _
_-.m_.m
~~.-._.._.__m 1
-,e i
1
- j. ~.,
j REACTOR SCRAN l
Reporting Period: 11/04/96 to 11/10/96 l
YTD YTD ABOVE BELOW YTD E
Pmi & WIT g3ggg ynpf GAjat COMPLICATIONS jj) ji3 IgLA1 l
11/05/96 NINE MILE Po!NT 1 100 SA Equipreent Failure NO 2
0 2
l
)
l l
I I
I l
l l
!I' l
i l
l' l
l l.
t i
4 i
J j
tcta: Year To Date (YTD) Totals Include Events Within The Calendar Year Indicated By The End Date of The Specified Reporting Period ]
)
f.
ETS-10 Page:1 11/12/96 l 1..
l 4
1 i
~-
. =
t COMPARISON OF WEEKLY SCRAM STATISTICS WITH INDUSTRY AVERAGES PERIOD ENDING 11/10/96 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 EXTERNAL 0
0.16 0.21 0.17 0.13 OTHER 0
0.09 0.06 0.02 Setotal 1
2.54 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 0.04 EXTERNAL 0
0.00 OTHER 0
0.00
$stotet 0
0.36 0.31 0.37 0.57 0.65 TOTAL 1
2.90 3.06 2.89 3.38 4.08 1996 1995 1994 1993 1992 NO. OF WEEKLY WEEKLY WEEKLY WEEKLY WEEKLY SCRAM TYPE SCRAMS AVERAGE AVERAGE AVERAGE AVERAGE AVERAGE (YTD)
TOTAL AUTOMATIC SCRAMS 1
1.80 1.92 2.19 2.44 3.06 TOTAL MANUAL SCRANS 0
1.09 1.13 0.69 0.94 1.02 TOTALS MAY DIFFER BECAUSE OF ROUNDING OFF ETS-16 Page: 1 11/12/96
- -. - _ -. - _..... ~. - _ ~.
. -. - - _ =. -.. - ~.. -,
.. ~.. -.. -...
e i
t REACTOR SCRAM i
Reporting Period 10/28/96 to 11/03/ %
i YTD YTD i
l ABOVE BELOW YTD
{
9171 PLANT & UNIT ggg!
IL.1 CAUE COMPLICATIONS jil lig g 1 j.
10/29/96 se0WNS FERRY 2 100 SA Equipment Falture No 2
0 2
11/02/96 NOPE CREEK 1 29 SM Equipment Failure NO 1
0 1
t 5
?
l S
1 I
l a
v i
i I
i i
hots: Year To Date (YTD) Totals Include Events Within The Calender Year Indicated By The End Date Of The Specified Reporting Period ETS-10 Page:1 11/06/96
a t
COMPARISON OF WEEKLY SCRAM STATISTICS WITH INDUSTRY AVERAGES
+
PERIOD ENDING 11/03/96 WUMBER 1996 1995 1994 1993 1992 0F WEEKLY WEEKLY WEEKLY WEEKLY WEEKLY i
SCRAM CAUSE SCRAMS AVERAGE AVERAGE Av'ERAGE AVERAGE AVERAGE (YTD)
POWER GREATER THAN OR EQUAL TO 15%
1 EQUIPMENT FAILURE 2
1.61 1.83 1.52 1.53 2.62 DESIGN / INSTALLATION ERROR 0
0.11 0.12 0.08 0.04 1
DPERATING 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 O
0.20 0.10 0.27 0.38 0.42 DESIGN / INSTALLATION ERROR 0
0.00 0.02 1
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 2
2.92 3.06 2.89 3.38 4.08 1996 1995 1994 1993 1992 NO. OF WEEKLY WEEKLY WEEKLY WEEKLY WEEKLY SCRAM TYPE SCRAMS AVERAGE AVERAGE AVERAGE AVERAGE AVERAGE (YTD)
TOTAL AUT0J.ATIC SCRAMS 1
1.82 1.92 2.19 2.44 3.06 TOTAL MANUAL SCRAMS 1
1.11 1.13 0.69 0.94 1.02 TOTALS MAY DIFFER BECAUSE OF ROUNDING OFF l
ETS-14 Page: 1 11/06/96
l 5
,, NOTES 1
1 l
1.
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.
PERSONNEL 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 l
(LIGHTNING), SYSTEM DESIGN, OR UNKNOWN CAUSE.
OEAB SCRAM DATA Manual and Automatic Scrams for 1987 ------------------ 435 Manual and Automatic Scrams for 1988 ------------------ 291 M nual 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 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
. h e
es e
OPERATING REACTOR PLANTS SIGNIFICANT EV C TS SORT > Event Date QUERT> Event Type SIG & Close Out Date >= 10/22/% & Close out Date <= 11/19/%
DATE OF 50.72 OR CLOSE0UT PLANT 8 UNIT EVENT NUMBER DESCRIPTION OF EVENT SIGNIFICANCE SRIEFING PRESENTER RECORD HOPE CREEK 1 03/14/%
.O Progressentic weaknesses resulted in violations and Reactor Protection System KOT2 ALAS M.
HIGHLIGHT imposition of civil penalty. Classification based l
on repeated failures to plan appropriate testing of equipment following maintenance.
SECUOTAH 2 10/11/9631138 Scram with complications - impulse pressure switch Safety-Related cooling System
%-13 KDENICK S.
HIGHLIGHT failure and auxiliary feedwater inoperable.
i i
I e
I i-
[
I s
t L
l t
i ETS-11 Page:1 12/12/96 ATTAGRNT 4
. -.