ML20045B684
| ML20045B684 | |
| Person / Time | |
|---|---|
| Site: | Crystal River, North Anna  |
| Issue date: | 05/05/1993 |
| From: | Chaffee A Office of Nuclear Reactor Regulation |
| To: | Grimes B Office of Nuclear Reactor Regulation |
| References | |
| OREM-93-016, OREM-93-16, NUDOCS 9306180279 | |
| Download: ML20045B684 (16) | |
Text
l May 5, 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 APRIL 28, 1993 - BRIEFING 93-16 On April 28, 1993, we conducted an Operating Reactors Events Briefing (93-16) to inform senior managers from offices of the Commission, ACRS, OE, NRR, and regional offices of seletted
~
events that occurred since our last briefing on April 23, 1993. lists the attendees. presents 1.he significant elements of the discussed events. contains reactor scram statistics for the week ending April 25, 1993.
No significant events were identifi'ed for input into the NRC performance indicator program.
- original signed by Robert L. Dennig for -
Alfred E. Chaffee, Chief Events Assessment Branch Division of Operating Reactor Support
Enclosures:
As stated DISTRIBUTION:
Central Files EGoodwin cc w/ attachments:
PDR DSkeen See next page EAB R/F EBenner KGray JZarzuela RDennig LKilgore, SECY S
/
S EAB/
S EA KGray Re g
EGoodwin AChaffe rf g$/)3/93
[/3/93 T'/j/93
[/s'/9
.,,ff~'s',f,f, 1
fV*
OFFICIAL RECORD COPY DOCUMENT NAME:
ORTRANS.KAG (G:KAG)
REURN "O REGULATOM ClI4 ES y ao~6 9306180279 930505 W g/4'
~~~
PDR ORG NRRD g
6 j
.~
cc:
T. Murley, NRR (12G18)
L. Engle (PDII-2)
F. Miraglia, NRR (12G18)
H. Berkow (PDII-2)
F. Gillespie, NRR (12G18)
H. Silver (PDII-2)
J. Partlow, NRR (12G18)
S. Varga, NRR (14E4)
J. Calvo, NRR (14A4)
G.
Lainas, NRR (14H3)
J. Roe, NRR (13E4)
J.
Zwolinski, NRR (13H24)
M. Virgilio, NRR (13E4)
W. Russell, NRR (12G18)
J. Richardson, NRR (7D26)
A. Thadani, NRR (8E2)
S.
Rosenberg, NRR (10E4)
C.
Rossi, NRR (9A2)
B.
Boger, NRR (10H3)
F.
Congel, NRR (10E2)
D.
Crutchfield, NRR (11H21)
W.
Travers, NRR (11B19)
D.
Coe, ACRS (P-315)
E. Jordan, AEOD (MN-3701)
Acting Director, DSP/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)
A.
Bates, SECY (16G15)
G. Rammling, OCM (16G15)
T. Martin, Region I W. Kane, Region I C.
Hehl, Region I S.
Ebneter, Region II E. Herschoff, Region II S. Vlas, Region II J.
Martin, Region III E. Greenman, Region III J. Milhoan, Region IV B. Beach, Region IV B.
Faulkenberry, Region V K. Perkins, Region V bec:
Mr. Sam Newton, Manager Events Analysis Department Institute of Nuclear Power Operations 700 Galleria Parkway Atlanta, GA 30339-5957
)
i ENCLOSURE 1 LIST OF ATTENDEES OPERATING REACTORS EVENTS FULL BRIEFING (93-16)
APRIL 28, 1993 NAME OFFICE NAME OFFICE A.
BENNER NRR J.
ROE NRR E.
GOODWIN NRR G.
ENG NRR J.
ZARZUELA NRR R.
ZIMMERMAN NRR R.
ECKENRODE NRR G. MARCUS NRR L.
SILVER NRR S. ROSENBERG NRR H.
BERKOW NRR J. RICHARDSON NRR S.
BREWER NRR W.
LEACH OCM/IS J.
PARTLOW NRR D. COE ACRS F. MIRAGLIA NRR W. TROSKOSKI OE TELEPHONE ATTENDANCE (AT ROLL CALL)
Regions Resident Inspectors Region II Region III Region IV Region V IIT/AIT Team Leaders Misc.
4 h
ENCLOSURE'2
.I OPERATING REACTORS EVENTS BRIEFING 93-16 LOCATION:
10 B11,. WHITE FLINT WEDNESDAY, APRIL 28, 1993,-11:00 A.M.
NORTH ANNA, UNIT 2 FEEDWATER FLOW OSCILLATIONS AND-WATER HAMMER CRYSTAL RIVER, UNIT 3 EXCESSIVE RATE OF C00LDOWN FROM DECAY HEAT REMOVAL (DHRF SYSTEM' 4
PRESENTED BY:
EVENTS ASSESSMENT BRANCH DIVISION OF OPERATING REACTOR SUPPORT, NRR
..-...a
93-16 NORTH ANNA, UNIT 2 FEEDWATER FLOW OSCILLATIONS AND WATER HAMMER APRIL 24, 1993 PROBLEM UNIT 2 EXPERIENCED FEEDWATER (FW) FLOW OSCILLATIONS AND A SUBSEQUENT WATER HAMMER.
CAUSE THE CAUSE OF THE FW FLOW OSCILLATIONS WAS A POSITIVE FEEDBACK LOOP THAT EXISTED IN THE FW REG VALVE (FRV)
PNEUMATIC CONTROLLER.
SAFETY SIGNIFICANCE WATER HAMMER CAN CAUSE SEVERE DAMAGE TO PIPING SYSTEMS AND SUPPORTS.
SEQUENCE OF EVENTS 02:55:00 UNIT AT 71% POWER TO CORRECT QUADRANT POWER-TILT
- RATIO, "C" FW FLOW BEGINS TO OSCILLATE.
05:00:00 "B"
FW FLOW BEGINS TO OSCILLATE.
05:17:00 "C"
FW FLOW SWINGS WORSEN.
"B" AND "C" FRVs PLACED IN MANUAL.
05:21:00 OPERATORS GET REPORTS OF FW LINES MOVING.
SENIOR REACTOR OPERATOR (SRO) DISPATCHED TO INVESTIGATE.
CONTACT:
E. BENNER, NRR/ DORS AIT:
NO
REFERENCES:
10 CFR 50.72 #25442 AND SIGEVENT:
TBD PN29318
i 5
NORTH ANNA, UNIT 2 93-16 05:29:51 "C" FRV GOES FULL CLOSED AND CYCLES SEVERAL-TIMES.
05:30:00 SRO ORDERS MANUAL TRIP.
ALL EQUIPMENT PERFORMED AS EXPECTED.
NO FURTHER VIBRATIONS FELT OR HEARD IN CONTROL ROOM.
06:02:00 ALERT WAS DECLARED TO ASSIST IN PERSONNEL ACCOUNTABILITY AND PROVIDE ADDITIONAL TECHNICAL i
STAFFING.
08:16:00 ALERT DOWNGRADED TO UNUSUAL EVENT (UE).
10:45:00 NRC NOTIFIED THAT UE TERMINATED.
12:53:00 REACTOR COOLANT SYSTEM TEMPERATURE <350*F- (HOT-SHUTDOWN).
DISCUSSION e
DAMAGE:
1 SNUBBER 1
1 SPRING CAN v
1 HORIZONTAL SNUBBER v
INSULATION v
"B" FW BYPASS VALVE YOKE BREAKAGE 1
v CONTROL AIR LINES PULLED OUT FROM "B" FRV l
v AND BYPASS VALVE OPERATORS ROOT CAUSE OF EVENT WAS POSITIVE FEEDBACK FROM THE "C" e
I
.FRV PNEUMATIC B0OSTER.
TESTING IDENTIFIED THIS CONDITION DID NOT EXIST ON "A" AND "B" FRVs.
i
~.
4 NORTH ANNA, UNIT 2 93-16 HAD NOT BEEN IDENTIFIED BEFORE BECAUSE OF NO PREVIOUS e
STEADY STATE OPERATION AT THIS POWER' LEVEL.
CORRECTIVE ACTIONS:
DETUNE FRVs TO ELIMINATE POSITIVE FEEDBACK v
MONITOR FW FLOW UPON POWER ASCENSION v
REPLACE FRVs~WITH DIFFERENT DESIGN DURING NEXT v
REFUELING OUTAGE FOLLOWUP ADDITIONAL INSPECTORS LOOKED INTO EVENT:
e TWO REGION BASED INSPECTORS ONE REGION BASED SECTION CHIEF v
ONE NRR METALLURGIST v
{
LICENSEE HAS PROVIDED LIST OFJACTIONS TO BE PERFORMED e
PRIOR T0'STARTUP.
' I.
i
't 93-16 CRYSTAL RIVER, UNIT 3 EXCESSIVE RATE 0F C00LDOWN FROM DECAY HEAT REMOVAL (DHR) SYSTEM MARCH 5, 1993 PROBLEM OVER ONE HOUR, C00LDOWN RATE WAS MORE THAN ALLOWED BY TECHNICAL SPECIFICATIONS.
l
- CAUSE MALFUNCTION OF VALVE THAT CONTROLS COOLING WATER FLOW TO DECAY HEAT REMOVAL (DHR) COOLER, TRAIN "A" (DCV-177).
SAFETY SIGNIFICANCE e
THE C00LDOWN RATE OF REACTOR COOLANT SYSTEM-(RCS), IF T00 FAST, CAN AFFECT THE FRACTURE TOUGHNESS PROPERTIES:
0F REACTOR VESSEL.
e RCS C00LDOWN RATE LIMIT IN TECH SPEC IS 25*F PER 1/2 HOUR FROM 150 - 280*F.
C00LDOWN RATE B0UNDED BY AN ESTIMATION OF 128*F IN 51 MINUTES.
SEQUENCE OF EVENTS e
PLANT IN MODE 4- (HOT SHUTDOWN) COOLING DOWN FOR SCHEDULED OUTAGE BY STEAM GENERATORS AT COLD LEG TEMP.
)
0F 263*F AND PRESSURE OF 200 PSIG.
j CONTACT:
J. ZARZUELA, NRR/ DORS AIT:
NO
REFERENCE:
10 CFR 50.72 #25194 AND SIGEVENT:
TBD LER 50-302/93-001
9 CRYSTAL RIVER, UNIT 3 93-16 e
OPERATORS SWITCHING TO DHR COOLING MODE.
1239 DHR TRAIN "A" PUT IN SERVICE.
COOLER OUTLET TEMP.
229'F.
1245 RCS PUMPS SECURED.
CONTROL ROOM OPERATORS WERE UNABLE TO CONTROL RATE OF COOLING USING DCV-177.
AUX OPERATOR WAS DISPATCHED TO CLOSE THE VALVE LOCALLY.
1322 WHILE SHIFTING DCV-177 FROM AUTOMATIC TO MANUAL, THE VALVE OPENED, CAUSING ADDITIONAL RCS_C00LDOWN.
l AUX OPERATOR IMMEDIATELY CLOSED THE VALVE.
OVER THE NEXT SEVERAL MINUTES THE VALVE DRIFTED OPEN'AND ADDITIONAL COOLING WAS EXPERIENCED BY THE RCS.
1331 OPERATORS SWITCHED FROM TRAIN "A" T0."B" 0F DHR.
TRAIN "B" WAS AT AMBIENT TEMPERATURE - 70*F.
COLD WATER FLOWED TO THE DOWNCOMER OF THE VESSEL, BUT ONLY UNTIL RCS COOLANT FILLED DHR "B" COOLER.
SHORTLY AFTERWARDS, REACTOR COOLANT TEMPERATURE WAS STABILIZED AT 220*F.
4
CRYSTAL RIVER, UNIT 3 93-16 DISCUSSION SAFETY MARGIN TO STRUCTURAL INTEGRITY.OF REACTOR VESSEL e
CALCULATED BY B&W WAS 1.7.
THE TECH SPEC LIMIT-OF 25'F PER 1/2 HOUR IS EQUIVALENT e
TO A SAFETY MARGIN OF 2.0.
NO SIGNIFICANT RISK FOR REACTOR VESSEL EXPECTED AS LONG-AS MARGIN IS AB0VE 1.0.
i MALFUNCTIONING OF DCV-177, A BUTTERFLY VALVE WITH A e
PNEUMATIC ACTUATOR, WAS DUE TO:
DEBRIS FOUND IN A SMALL ORIFICE OF ACTUATOR'S q
1 CONTROLLER, APPARENTLY CAUSED' VALVE TO OPEN MORE THAN DEMANDED WHILE IN AUTOMATIC MODE.
INSTRUCTION STEPS POSTED ON THE VALVE ACTUATOR ADDRESSING MANUAL OPERATION CAUSED THE-VALVE TO GO FULL OPEN DUE TO VENTING THE' ACTUATOR.
THESE INSTRUCTIONS, PROVIDED WITH THE VALVE'ARE INAPPROPRIATE FOR SPECIFIC USE IN THIS SYSTEM.
I 1
i l
ta,
4 CRYSTAL' RIVER, UNIT 3-93-16 BOTH AUTOMATIC AND MANUAL CONTROL FUNCTIONED SIMULTANEOUSLY.
THIS CAUSED A PROTECTIVE PIN LOCATED IN THE MANUAL SHAFT TO BREAK.
THIS WAS THE CAUSE OF THE VALVE DRIFTING OPEN OBSERVED AFTERWARDS.
FOLLOWUP NRR/EMCB REVIEWED CALCULATIONS OF SAFETY MARGIN FOR-e STRUCTURAL' INTEGRITY OF THE VESSEL PRESENTED BY LICENSEE AND FOUND THEM ACCEPTABLE.
LICENSEE WILL MODIFY INSTRUCTIONS POSTED IN THE. VALVE e
FOR SWITCHING FROM AUTO TO MANUAL.
j NEED FOR INFORMATION NOTICE CONCERNING VALVE-ACTUATION e
IS BEING EVALUATED.
4 I
v w =
_v.,
+ -,,,
.4-m 1
BRIEFING 93-16 CRYSTAL RIVER, UNIT 3 l
(
3/4.4.9 PRESSURE / TEMPERATURE LIMITS REACTOR COOLANT SYSTEM t!MITING CON,DI_ TION FOR OPERATI1N 3.4.9.1 The Reactor Coolant System (except the pressurizer) temperature and pressure shall be limited in accordance with the limit lines shown on Figures 3.4-2, 3.4-3, and.3.4-4 d ring heatup, cooldown, and inservice leak and hydrostatic testing with:
A maximum heatup of 50*F in any one hour period, a.
b.
For the temperature ranges specified below, the cooldown rates should be as specified:
1.
T > 280*F s 50*F in any 1/2 hour period it.
150*F < T s 280'F 125'F in any 1/2 hour period 111.
T s 150*F s 10*F in any 1/2 hour period and A maximum temperature change of.less than or equal to 5*F in any c.
one hour period during hydrostatic testing operations above system design pressure.
APPLICABILITY:
At all times.
ACTION:
With any of the above limits exceeded, restore the temperature and/or-pressure to within the limits within 30 minutes; perform an enaineering evaluation to
__ determine the effects of the out-of-limft condition on the fracture toughness the_ Reactor Coolant System; determine that the Reactor Coolant properties nf System remains acceptable for continued operation or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and reduce RCS T., and pressure to less than 200*F and 500 psig, respectively, within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
CRYSTAL RIVER UNIT 3 3/4 4-24 Amendment No. 52,133, FEB 7 1991 I
BRIEFING 93-16'-
=
CRYSTAL RIVER, UNIT 3-Der Heat % - m3.s e tsas Yea, R 9
V no
]
n,,
n e,
l A tae, T<x*:
j' P
/1 :
l one 4
i t
t 0"*
I b /.
3 150
~.
i j
,1 t
=
3 i
.,eo e
i t
.i I
e so 3a0 tso too wne.ene
' 1. DHR "A" Put in Service
- 2. RCPs Secured-
- 3. DCV-177 Went Full Open
- 4. DCV-177 Closed
- 5. DHR "B" Put in Service f
- e w -~ -
,e s
v s
,,u_
v,a
,w
-~
ENCLOSURE 3 J
REACfDR SCRAM
=
Reporting Period: 04/19/93 to 04/25/93 YTD YTD ABOVE BELOW YTD (Ali PLANT & UNIT POWER TIP1 CAUSE COMPLICATIONS 11}
113
' TOTAL 04/20/93 FITZPATRICK 100 SA Equipnent Falture NO 2
0 2
04/20/93 FERMI 2 1
SA Equipment Falture NO 1
1 2
04/23/93 LASALLE 1 99 SM Equipnent f atture wo 2
0 2
04/24/93 NORTM ANkA 2 70 SM Equipment f ailure NO 2
0 2
D4/25/93 PEACH BOTTOM 2 15 SM Equipment Falture No
~ 3 0
3 04/25/93 VOCTLE 1 0
SM Equipment failure NO O
1 1
bote: Year To Date (YTD) Totals include Events Within The Calendar Year Indicated By The End Date of 1he Specified Reportlne Period E75-10 Fage:1 05/03/93 o
COMPARISDN OF WEEKLY SCRAM STATISTICS WITH INDUSTRY AVERAGES PERIOD ENDIhG 04/25/93 NUMBER 1993 1992 1991*
1990*
1989*
OF WEEKLY WEEKLY WEEKLY WEEKLY WEEKLY SCRAM CAUSE SCRAMS AVERAGE AVERAGE AVERAGE AVERAGE AVERAGE (YTC)
POWER GREATER THAN OR EDUAL TO 15%
EQUIPMENT FAILURE
- 4 2.4 2.6 2.9 3.4 3.1 DESIGN / INSTALLATION ERROR
- 0 0.1 OPERATING ERROR
- O 0.4 0.2 0.6 0.5 1.0 MAINTENANCE ERROR
- 0 0.5 0.4 EXTERNAL
- 0 0.2 CTHER*
0 0.0 0.2 0.1 Subtotal 4
3.6 3.4 3.5 3.9 4.2 POWER LESS THAN 15%
EQUIPMENT FAILURE
- 2 0.4 0.4 0.3 0.4 0.3 DESIGN /INSTALLATIDN ERROR
- 0 0.0 OPERATING ERROR
- 0 0.1 0.1 0.2 0.1 0.3 MAINTENANCE ERROR
- 0 0.0 0.1 EXTERNAL
- O 0.0 OTHER*
0 0.0 0.1 Subtotal 2
0.5 0.7 0.5 0.5 0.6 TOTAL 6
4.1 4.1 4.0 4.4 4.8 1993 1992 1991 1990 1989 No. OF WEEKLY WEEKLY WEEKLY WEEKLY WEEKLY SCRAW TYPE SCRAMS AVERAGE AVERAGE AVERAGE AVERAGE AVERAGE (YTD)
TOTAL AUTOHATIC SCRAMS 2
2.8 3.1 3.3 3.2 3.9 TOTAL KANUAL SCRAMS 4
1.3 1.0 0.7 1.2 0.9 TOTALS KAY DIFFER BECAUSE OF ROUNDING OFF
- Detsited breakdown not in database for 1991 and earlier EXTERNAL cause included in EQUIPMENT FAILURE MAINTENANCE ERROR and DESIGN /INSTALLAfl0N ERROR causes inctoded in OPERATING ERROR
- OTHER cause included in EQUIPMENT FAILURE 1991 and 1990 l
s ETS-14 Page: 1 D5/03/93
o.
Il0.TIR 1.
PLANT SPECIFIC DATA BASED ON INITIAL REVIEW OF 50.72 REPORTS FOR THE PERIOD OF INTEREST.
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.
2.
COMPLICATIONS: RECOVERY COMPLICATED BY EQUIPMENT FAILURES OR PERSONNEL ERRORS UNRELATED TO CAUSE OF SCRAM.
3.
SA = Scram Automatic; SM = Scram Manual 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 --(YTD 04/25/93)-- 67 P
i 7
4