ML20057F147
| ML20057F147 | |
| Person / Time | |
|---|---|
| Site: | Oconee, Palisades, McGuire  |
| Issue date: | 10/06/1993 |
| From: | Chaffee A Office of Nuclear Reactor Regulation |
| To: | Grimes B Office of Nuclear Reactor Regulation |
| References | |
| OREM-93-037, OREM-93-37, NUDOCS 9310140196 | |
| Download: ML20057F147 (20) | |
Text
~
r October 6, 1993-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 EVENTS BRIEFING SEPTEMBER 29, 1993 - BRIEFING 93-37 On September 29, 1993, we conducted an Operating Reactors Events Briefing'(93-37) to inform senior managers from offices of the Chairman, SECY, NRR, AEvD, RES, and regional offices of selected events that occurred since our last briefing on September 22, 1993. lists the attendees. presents the significant elements of the discussed events. contains reactor scram statistics for the week ending September 26, 1993.
No significant events were identified for input into the NRC Performance Indicator Program.
[ original signed by]
Alfred E.
Chaffee, Chief Events Assessment Branch Division of Operating Reactor Support
Enclosures:
As stated DISTRIBUTION:
Central Files cc w/ enclosures:
PDR See next page LKilgore, SECY EAB R/F KGray RDennig EGoodwin DSkeen RBenedict EBenner TKoshy
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/93 d@/Q/93 OFFICIAL RECORD COPY DvCUMENT NAME:
ORTRANS.KAG (G:\\KAG) lqpp ;;lu: hadW1 uur i
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T. Murley, NRR (12G18)
.A. Hsia (PDIII-1)
F. Miraglia, NRR (12G18)
W. Dean (PDIII-1) l F.
Gillespie, NRR (12G18)
V. Nerses (PDII-3)
J.
Partlow, NRR (12G18)
D. Matthews (PDII-3)
S.
Varga, NRR (14E4)
L. Wiens-(PII-3)
J,. Calvo, NRR (14A4)
G.
Lainas, NRR (14H3)
J. Roe, NRR (13E4)
J.
Zwolinski, NRR (13H24)
E. Adensam, NRR (13E4) l W.
Russell, NRR (12G18) i J.
Wiggins, NRR (7D26) i A.
Thadani, NRR (8E2) l, S.
Rosenberg, NRR (10E4)
C.
Rossi, NRR (9A2)
B.
Boger, NRR (10H3)
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F.
Congel, NRR (10E2) j D.
Crutchfield, NRR (11H21) l W.
Travers, NRR (11B19)
D.
Coe, ACRS (P-315)
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)
E.
A.
Bates, SECY (16G15)
T. Martin, Region I W.
Kane, Region I R.
Cooper, Region I l
S.
Ebneter, Region II E. Merschoff, Region II S. Vias, Region II J. Martin, Region III -
a E.
Greenman, Region III i
J.
Milhoan, Region IV i
B.
Beach, Region IV B.
Paulkenberry, Region V K.
Perkins, Region V bec:
Mr. Sam Newton, Manager i
Events Analysis Department i
Institute of Nuclear Power Operations 700 Galleria Parkway Atlanta, GA 30339-5957 i
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UNITED STATES l
- 'l NUCLEAR REGULATORY COMMISSION
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WASHINGTON, D.C. 20555 0001 October 6, 1993 MEMORANDUM FOR:
Brian K. Grimes, Director Division of Operating Reactor Support i
FROM:
Alfred E. Chaffee, Chief Events Assessment Branch Division of Operating Reactor Support
SUBJECT:
OPERATING REACTORS EVENTS BRIEFING SEPTEMBER 29, 1993 - BRIEFING 93-37 On September 29, 1993, we conducted an Operating Reactors Events Briefing (93-37) to inform senior managers from offices of the Chairman, SECY, NRR, AEOD, RES, and regional offices of selected events that occurred since our last briefing on September 22, 1993. lists the attendees. presents the significant elements of the discussed events. contains reactor scram statistics for the week ending September 26, 1993.
No significant events were' identified for input into the NRC Performance Indicator Program.
f JY Alfred E. Chaffee, Chief i
Events Assessment Branch Division of Operating Reactor Support
Enclosures:
As stated cc w/ enclosures:
See next page
i f
ENCLOSURE 1 LIST OF ATTENDEES OPERATING REACTORS EVENTS FULL BRIEFING (93-37)
SEPTEMBER 29, 1993 l
NAME OFFICE NAME OFFICE A.
CHAFFEE NRR T.
YAMADA NRR R.
E.
BENNER NRR G.
PETRONE NRR i
K. GRAY NRR W.
DEAN NRR i
E.
GOODWIN NRR C.
ROSSI NRR R.
DENNIG NRR S. VARGA NRR L. WIENS NRR J. ROE NRR M. VIRGILIO NRR F.
CHERNY RES S. ROSENBERG NRR B. WELLING RI C.
GOODMAN NRR R.
RASMUSSEN RI i
G.
HORNSETH NRR V.
BENAROYA AEOD i
A.
HSIA NRR A. VIETTI-COOK OCM/IS i
S.
FLANDERS NRR K. HART-SECY
(
E. HACKETT NRR l
i TELEPHONE ATTENDANCE-i (AT ROLL CALL) i Recions Resident Inspectors Region I G. Maxwell (McGuire)
Region II T. Cooper (McGuire) l Region III Region IV Region V IIT/AIT Team Leaders
- Misc, P
t i
.i
ENCLOSURE 2 OPERATING REACTORS EVENTS BRIEFING 93-37 l
LOCATION:
10 811, WHITE FLINT l
WEDNESDAY, SEPTEMBER 29, 1993, 11:00 A.M.
PALISADES, UNIT 1 CRACK IN REACTOR COOLANT SYSTEM PRESSURE B0UNDARY 4
MCGUIRE, UNIT 1 SIPHONING 0F REACTOR COOLANT THROUGH STEAM GENERATOR U-TUBES OCONEE, UNIT 1 REACTOR TRIP FROM MAINTENANCE OVERSIGHTS PRESENTED BY:
EVENTS ASSESSMENT BRANCH DIVISION OF OPERATING REACTOR SUPPORT, NRR l
93-37 y
PALISADES, UNIT 1 l
CRACK IN REACTOR COOLANT SYSTEM BOUNDARY SEPTEMBER 16, 1993 PROBLEM UNIS0LABLE LEAK IN 4-INCH PRESSURIZER N0ZZLE.
i CAUSr l
CRACK IN N0ZZLE SAFE-END.
SAFETY SIGNIFICANCE 0
POTENTIAL SMALL-BREAK LOSS-OF-COO! AMT ACCIDENT ATOP PRESSURIZER.
i o
LICENSEE INSPECTION TWO MONTHS EARLIER DID NOT IDENTIFY I
GROWING CRACK.
l o
CIRCUMFERENTIAL PRIMARY WATER STRESS CORROSION CRACKING IN INCONEL 600 WELDS IS RARE.
i
-DISCUSSIOR o
LICENSEE'S IN-SERVICE INSPECTION RADIOGRAPHIC I
EXAMINATION ON 06/19/93 SHOWED AN INDICATION.
AN INDICATION WAS'ALSO IDENTIFIED IN THE ORIGINAL j
CONSTRUCTION'RADI0 GRAPHS.
i CONTACT:
R. BENEDICT, NRR/ DORS /0EAB AIT:
NO
REFERENCES:
10 CFR 50.72 #26075, SIGEVENT:
NO PN39355 AND HORNING REPORT j
DATED 09/20/93 e
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PALISADES, UNIT 1 93-37 o
NRC INDEPENDENT RADIOGRAPHY AND EVALUATION OF j
LICENSEE'S RADIOGRAPHS IDENTIFIED INDICATION AS A THERMAL FATIGUE CRACK.
NRC SIZED THE INDICATION, WITH ULTRASONIC EXAMINATIONS, AS A CRACK ABOUT 1/3 THROUGH WALL AND 1 7/8-INCH LONG.
o LICENSEE'S FOLLOWUP ULTRASONIC EXAMINATION ON 06/25/93 j
CONCLUDED THE INDICATION WAS DUE TO WELD ROOT GE0 METRY.
LICENSEE UNABLE TO IDENTIFY LOADS THAT WOULD DRIVE A FATIGUE CRACK.
o LICENSEE THEN PERFORMED AN ANALYTICAL FLAW EVALUATION PER ASME IWB-3600 USING THE FOLLOWING ASSUMPTIONS:
2 1/2 IN. X 1/3 THROUGH-WALL FLAW ASME XI CRACK GROWTH CURVE FOR FERRITIC MATERIAL i
PROPOSED TRANSIENT LOADS i
o LICENSEE CONCLUDED N0 FURTHER CRACK GROWTH FROM FATIGUE WOULD OCCUR.
o NRC DID NOT INITIALLY DISPUTE LICENSEE'S EVALUATION.
4 o
LICENSEE PRELIMINARY VISUAL EXAMINATION RESULTS INDICATE CRACKING IS INTERGRANULAR AND LOCATED NEAR A REPAIRED AREA.
o NRC SPECIAL EVALUATION TEAM ONSITE.
e OVER-COOLING OCCURRED DURING FINAL STAGES OF C00LDOWN FOLLOWING DISCOVERY OF THE LEAK.
l
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. PALISADES, UNIT 1-93-37 i
FOLLOWUP
.o LICENSEE' EXAMINING OTHER N0ZZLES ON PRESSURIZER.
i 4
o INDEPENDENT. METALLURGICAL EXAMINATIONS BEING PERFORMED THIS WEEK BY LICENSEE AND BY NRR.
o REGION III IS REVIEWING'THE EXCESSIVE C00LDOWN.
1 i
o NRR ALSO EVALUATING C00LDOWN PER ASME XI, APPENDIX E.
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93-37 MCGUIRE, UNIT 1 SIPHONING 0F REACTOR COOLANT THROUGH STEAM GENERATOR U-TUBES AUGUST 30, 1993 PROBLEM REACTOR COOLANT INVENTORY LOST THROUGH STEAM GENERATOR (SG) U-TUBES.
CAUSE NITROGEN (N ) COMING OUT OF SOLUTION ENTRAINED REACTOR 2
COOLANT THROUGH SG U-TUBES.
SAFETY SIGNIFICANCE MINIMAL; COOLANT LOSS WAS SELF-LIMITING.
SEQUENCE OF EVENTS PLANT CONDITIONS PRIOR TO EVENT:
o PRIMARY TEMPERATURE = 100 *F e
PRIMARY PRESSURE = ATMOSPHERIC e
PRESSURIZER LEVEL 0 20%
e REACTOR VESSEL VENTS OPEN AND PRESSURIZER VENTED AUGUST 30, 1993, 7:00 A.M.
CONTROL ROOM RECEIVES CALL FROM STEAM GENERATOR (SG) TRAILER REPORTING THAT WATER FLOWING THROUGH THREE SG TUBES AND OUT OPEN T MANWAY OF SG "1C".
com CONTACT:
E. BENNER, NRR/ DORS /0EAB AIT:
H0
REFERENCE:
10 CFR 50.72 #25990 SIGEVENT:
H0
l MCGUIRE, UNIT 1 93-37 CONTROL ROOM INDICATION SHOWING DECREASING e
PRESSURIZER LEVEL.
FLOW RATE THROUGH TUBES = 17 GPM; FLOW CONTINUED e
FOR = 2 HOURS; TOTAL INVENTORY LOST = 2000 GALLONS.
e LICENSEE ACTIONS:
- LOWERED PRESSURIZER LEVEL TO 10%
- REMOVED FILTER SYSTEM IN B0WL 0F STEAM GENERATOR WHICH MAY HAVE BEEN CREATING A VACUUM e
DURING EVENING 0F 8/30/93: SAME THREE TUBES (OR CLOSE PROXIMITY TUBES) " BURPED" REACTOR COOLANT.
e 8/31/93: AT 3:00 A.M.,
ONE TUBE BEGAN TO STREAM.
LICENSEE LOWERED REACTOR COOLANT LEVEL BELOW BOTTOM 0F PRESSURIZER TO BREAK SIPHON.
DISCUSSION LICENSEE USING T o13 MANWAY AS SUITABLE VENT PATH:
o c
f HOT LEG N0ZZLE DAM OUT COLD LEG N0ZZLE DAM IN THOUGH THIS VENT METHOD IS STANDARD FOR MCGUIRE, THEY HAD NEVER USED THE FLOW PATH THRU THE "C" SG BEFORE.
m HCGUIRE, UNIT 1 3-93-37 LICENSEE BELIEVES N COMING OUT OF SOLUTION IS THE 2
CAUSE OF THE LOSS OF INVENTORY THROUGH SG "1C" U-TUBES:
45 PSI N BLANKET IN VOLUME CONTROL TANK (VCT):
2 N
CONCENTRATION IN RESIDUAL HEAT REMOVAL (RHR) 2 LINES GREATER THAN SATURATION CONCENTRATION AT BULK REACTOR COOLANT TEMPERATURE AND PRESSURE VCT @ 65 "F REACTOR COOLANT SYSTEM LOOP "C" IS WHERE RHR FLOW PATH IS LOCATED SG "1C" HOT LEG RETURN WAS DIRECTLY BELOW TUBES WHICH WERE STREAMING AND " BURPING"
" BURPING" CEASED AFTER LETDOWN WAS ISOLATED e
SIPHONING WILL NOT RESULT IN REDUCED INVENTORY:
BOTTOM 0F TUBESHEET IS AT 83.1" VESSEL FLANGE IS AT 84" REDUCED INVENTORY CUT 0FF IS AT 76" e
LICENSEE MONITORING TUBES WITH A CAMERA.
FOLLOWUP RESIDENT INSPECTORS WILL ISSUE SPECIAL INSPECTION REPORT.
i
BRIEFING 93-37 McGuire Reactor Coolant System STEAM VESSEL HEAD VENT GENERATOR
/
O REACTOR VESSEL VESSEL FLANGE @ 84" NOZZLE BOTTOM OF TUBESHEET @ 83.1" - -
DAM REDUCED INVENTORY @ 76"
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93-37 OCONEE, UNIT 1 REACTOR TRIP FROM MAINTENANCE OVERSIGHTS AUGUST 23, 1993 PROBLEM PLANT TRIP FROM FULL POWER DUE TO MAINTENANCE OVERSIGHTS.
CAUSE INADEQUATE MAINTENANCE PRACTICES / PROCEDURES AND WEAK POST MAINTENANCE TESTING.
SAFETY SIGNIFICANCE o
UNNECESSARY CHALLENGES TO AUXILIARY FEEDWATER AND ENGINEERED SAFEGUARDS SYSTEMS DUE TO MAINTENANCE WEAKNESS.
i o
POTENTIAL MISSILE IN REACTOR BUILDING FROM THE RESTART OF REACTOR COOLANT PUMPS WHILE C0ASTING DOWN.
DISCUSSION o
DURING A TEST OF THE VITAL I&C POWER SYSTEM, A LOSS OF POWER TO ONE 125voc AND ONE 120VAC PANEL OCCURRED.
IT RESULTED IN THE FOLLOWING:
REACTOR TRIP FROM LOSS OF POWER TO ELECTRO-HYDRAULIC CONTROL CIRCUIT ENGINEERED SAFEGUARDS ANALOG CHANNEL 1A TRIP.
(N0 ACTUATION SINCE IT IS 2/4 LOGIC)
CONTACT:
T. K0 SHY, NRR/ DORS /0EAB AIT:
NO
REFERENCE:
10 CFR 50.72 #25961 SIGEVENT:
NO
~
OCONEE, UNIT 1 93-37 LOST CONTROL VOLTAGE FOR ONE 6900v BUS AND CONCURRENT LOSS OF TWO REACTOR COOLANT PUMPS (RCPs).
LOSS OF DC PREVENTED FAST TRANSFER TO THE ALTERNATE POWER SOURCE.
LOSS OF AREA RADIATION MONITOR INDICATION IN CONTROL ROOM LOSS OF INDICATION FOR TRAIN 1A INADEQUATE CORE COOLING MONITOR o
THE TESTING REQUIRED REMOVING ONE OF THE TWO DC SOURCES.
o WHEN ONE SOURCE WAS REMOVED, THE BACK UP SOURCE DID NOT ASSUME THE LOAD DUE TO WIRING ERROR.
o THE POLARITY OF THE DC SOURCE WAS REVERSED WHILE REINSTALLING THE BREAKER AFTER TESTING DURING THE PREVIOUS OUTAGE.
THIS CONDITION WAS NOT DETECTED DURING POST MAINTENANCE TESTING.
o BOTH RCP BREAKERS REMAINED CLOSED WHEN THE BUS WAS DEENERGIZED.
WHEN DC POWER WAS REC 0VERED, RCP POWER SHIFTED TO THE START-UP SOURCE.
THE PUMPS THEN TRIPPED ON OVERCURRENT CAUSED BY LOW VOLTAGE DURING SIMULTANE0US START OF BOTH RCPs.
o-MAIN FEEDWATER FAILED TO MAINTAIN STEAM GENERATOR LEVEL DUE TO THE PRESENCE OF AN INCORRECT CARD.
o THE INCORRECT CARD WITH A SPEED LIMITING CIRCUIT, WHICH l
LIMITED PUMP OUTPUT, WAS INSTALLED IN THE PREVIOUS REFUELING OUTAGE.
OCONEE, UNIT 1 93-37 FOLLOWUP e
NRC CONFERENCE CALL TO LICENSEE MANAGEMENT THAT ALSO ADDRESSED A SUBSEQUENT TRIP AT UNIT 2.
e ANOTHER NRC MANAGEMENT MEETING WITH THE LICENSEE IS PLANNED TO ADDRESS RECENT MAINTENANCE DEFICIENCIES IDENTIFIED AT THE OCONEE SITES INCLUDING THIS EVENT.
NRC CONDUCTED A SPECIAL INSPECTION AND REVIEWED LICENSEE'S CORRECTIVE ACTIONS.
o I&C BRANCH HAD EVALUATED AND ACCEPTED THE EFFECTS FROM THE LOSS OF ONE DC BUS AT OCONEE (BULLETIN 79-27 AUDIT).
ENCLOSURE 3 REACTOR SCRAM Reportirg Period: 09/20/93 to 09/26/93 I
YTD YTD ABWE RELOW YTD
- pATJ, PLAWT & (MIT POWER ILP1 CAUSE CW PLICAff0k$
j$
11}
M 09/20/93 EAINT LUCIE 1 63 SM External NO 2
0 2
09/22/93 5AINT LUCIE 1 11 SM Externet No 2
1 3
09/22/93
$EABROOK 1 100 SA Essipment Failure No 5
0 5
09/24/93 flT2 PATRICK 1 18 SA maintenance Error wo 3
1 4
D9/25/93 CATAAA 2 100 SA Maintenance Error no 1
0 1
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tota:
Tear To Cate (YTD) Totals include Events Within The Calendar Year Indicated By The End Date of The specified Reporting Period ETE 10 p*D*:1 09s'29/93
h l
i
' *~
_j COMPARISON OF EEKLY SCRAM STA11STICS WITN INDUSTRY AVERAGES j
PERIOD ENDING-09/26/93 EMBER
'1993 1992 1991*
1990* '
1989*
i OF WEEKLY
' WEKLT W EKLY K EELY WEEKLY I
SCRAM CAUSE SCRAMS AVERAGE AVERAGE AVERAGE AVERAGE AVERAGE (YTD)
POWER GREATER THAN OR EQUAL TO 15%
~!
EQUIPMENT FAILURE
- 1 1.8 2.6 2.9 3.4 3.1 DESIGN /lWSTALLATION ERROR
- 0 0.1 OPERATING ERROR
- 0 0.3 0.2 0.6 0.5 1.0
}
MAINTEhAhCE ERROR
- 2 0.6 0.4
{
EXTERNAt*
1 0.2 OTHER*
0 0.0 0.2 0.1 f
l
$ 4 total 4
3.0 3.4 3.5 3.9 4.2
(
POWER LESS THAN 151 I
EQUIPMENT FAILURE
- 0 0.3 0.4 0.3 0.4 0.3 DESIGN /INSTALLAfl0N ERROR
- O 0.0 l
OPERATING ERROR
- 0 0.2 0.1 0.2 0.1 0.3 f
MAINTENAkCE ERROR
- O 0.0 0.1 ErfERhAL*
1 0.1 OTHER*
0 0.0 0.1 l
i
$4 total 1
0.6 0.7 0.5 0.5 0.6 l
-f TOTAL 5
3.6 4.1 4.0 4.4 4.8
-i i
i 1993 1992 1991 1990
'1989 I
NO. OF W EKLY EECLY WEEKLY WEKLY WEKLY j
StaAM YYPE SCRAMS AVERAGE AVERAGE AVERAGE AVERAGE AVERAGE (YTD) 1 TOTAL AUTOMATIC SCRAMS 3
2.6 3.1 ~
3.3 3.2 3.9 i
TOTAL MANUAL SCRAMS 2
1.0 1.0 0.7 1.2 0.9 I
1 l
TOTALS MAY O!FFER BECAUSE OF ROUNDING OFF
}
- Detailed breakdown not in database for 1991 and earlier
- EXTEthAL cause included in fau!PMENT FAILURE
- MAINTEhANCE (RROR eM DESIGN /INSTALLATIDW ERROR causes included in OPERATING ERROR
- OTHER cause inclu$ed in EQUIPMENT FAILURE 1991 and 1990-1 l
l 6
!i 09/29/93 f
ETS-16' Pope 1 4
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4 9
I.
PLANT SPECIFIC DATA BASED ON INITIAL REVIEW OF 50.72 REPORTS FOR THE l
WEEK OF INTEREST.
PERIOD IS MIDNIGHT SUNDAY THROUGH MIDNIGHT SUNDAY.
i SCRA)G 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.
3.
PERSONELL RELATED PROBLEMS INCLUDE HUMAN ERROR, PROCEDURAL DEFICIENCIES, f
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 i
Manual and Automatic Scrans for 1987 ------------------ 435 Manual and Automatic Scrams for 1988 ------------------ 291 Manual and Automatic Scrams for 1989 ------------------ 252 Manual and Automatic Scrans for 1990 ------------------ 226 Manual and Automatic Scrams for 1991 ------------------ 206 l
Manual and Automatic Scrams for 1992 ------------------ 212 Manual and Automatic Scrams for 1993 --(YTD 09/26/93)-- 136
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