Steam Generator Tube Rupture · 2012-12-05 · Objectives 1. Discuss why operator intervention is...
Transcript of Steam Generator Tube Rupture · 2012-12-05 · Objectives 1. Discuss why operator intervention is...
Steam Generator Tube Rupturep
Chapter 4.6
Objectives1. Discuss why operator intervention is
necessary to limit or prevent radiological releases during a Steam Generator Tube Rupture (SGTR) event.
2. Discuss the primary-side and secondary-
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side indications of an SGTR in the control room.
3. Discuss how the affected SG may be identified either prior to or following the reactor/turbine trip.
Objectives (Cont)4. List the initial actions taken by the operator
once the affected SG has been identified.
5. Discuss the actions required to stop the primary-secondary leakage.
6. Discuss the problems associated with the
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6 scuss e p ob e s assoc a ed efollowing: Secondary-to-primary leakage, SG Overfill.
7. List the principal systems/components affected by a loss of offsite power (LOOP).
Objectives (Cont)8. Discuss how plant cooldown and pressure
control are accomplished with an SGTR and LOOP.
9. Discuss what affect the following events had on the SGTR transient at the Ginna Plant:
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Tripping of the reactor Coolant Pumps,
Failure of pressurizer power-Operated relief valve,
Automatic operation of letdown valves,
Pressurizer relief tank failure, and
Steam generator Safety Valve Failure.
SGTR
• Most frequent occurring major accident.
• Provides a direct release path for primary coolant to environment via SG and its safety/relief valves (Containment Bypass).
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y ( yp )
• Timely operator involvement is required to prevent SG overfill and limit radiological releases.
Figure 4.6-1 Closeup View of SGTR
Past Steam Generator Tube Rupture Accidents at Pressurized Water Reactors
PlantDate Leak Rate (gpm) Cause
Point Beach Unit 1 February 26, 1975 125 wastage
Surry Unit 2 September 15, 1976 330 PWSCC in U-bend
Doel Unit 2 June 25, 1979 135 PWSCC in U-bend
Prairie Island 1 October 2, 1979 390 loose parts
Ginna Unit 1 January 25, 1982 760 loose parts and tube wear
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Fort Calhoun May 16, 1984 112 ODSCC at a crevice
North Anna Unit 1 July 15, 1987 637 high cycle fatigue in a U-bend
McGuire Unit 1 March 7, 1989 500 ODSCC in the free span
Mihama Unit 2 February 9, 1991 700 high cycle fatigue
Palo Verde Unit 2 March 14, 1993 240 ODSCC
Indian Point Unit 2 February 15, 2000 150 PWSCC in U-bend
Industry Improvements
• Secondary Side Inspections
• Improved Steam Generator Designs
• Water Chemistry Control
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• More Reliable Eddy Current Techniques
Primary-Side Indications of SGTR• Decreasing Prz level
• Decreasing Prz Pressure.
• Increasing charging flow.
These indication would also occur for a LOCA,how could the operator differentiate between
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how could the operator differentiate betweenthe two?• Lack of degraded containment conditions.• Abnormal secondary side indications.
1 tube
10 tubes
PR
ES
SU
RE
(PS
IG)
1000
1250
1500
1750
2000
2250
2500
RCS pressure drop is one indication of SGTR or LOCA
Trip from 0T T or If reducing Turb Load, Low Prz Press
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RC
SP
0
250
500
750
0 2 4TIME (MINS)
6 8 10
Figure 4.6-2(a) Initial Pressurizer Pressure Response4.6-35
Trip Trip
indication of SGTR or LOCA.
40
60
80
100
1 tube
RE
SS
UR
EL
EV
EL
(%S
PA
N)
Prz level drop is another indication of SGTR or LOCA.
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Figure 4.6-2(b) Initial Pressurizer Level Response4.6-35
8 10
0
2010 tubes
RC
SP
R
0 2 4
TIME (MINS)
6
LOCA.
T hot
T cold
MP
ER
AT
UR
E(°
F)
300
400
500
600
700
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Figure 4.6-3 RCS Temperature Following Reactor Trip4.6-37
6 8 10
TE
TIME (MINS)
0 2 4
0
100
200
Secondary-Side Indications of SGTR
• Chemistry Samples.
• Condenser Off Gas RM.
• SGBD Rad Monitor Alarms.
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• Main Steam Line RM’s
• SF/FF mismatch on affected SG.
• SG level rise on affected SG.
AFFECTED SG
INTACT SG AFFECTED SG
VE
L(%
NA
RR
OW
RA
NG
E)
100
80
60
40
TRIP
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Figure 4.6-4(a) Steam Generator Response Following Reactor Trip4.6-39
TIME (MINS)SG
WA
TE
RL
EV 20
0
1086420
SG
PR
ES
SU
RE
(PS
IG)
250
500
750
1000
1250
15
TIME (MINS)
S
Figure 4.6-4(b) Steam Generator Response Following Reactor Trip4.6-39
2 4 6 8 10
0
250
0
AM
FL
OW
(%N
OM
INA
L)
INTACT SGAFFECTED SG
100
80
60
40
20
16
ST
EA
Figure 4.6-4(c) Steam Generator Response Following Reactor Trip4.6-39
20
0
1086420
TIME (MINS)
Initial Operator Actions
• Isolate the affected SG.
1. Isolate Feedwater & AFW.
2. Close MSIV.
• These actions help minimize radiological
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releases & prevent SG overfill.
• Isolating the affected SG also allows use of the non-affected SG for cooling down RCS.
Recovery Actions
• Following isolation of the effected SG the intent is to match RCS pressure with affected SG pressure to stop the leak.
• 1st step cooldown RCS using intact SG’s.
2nd t d i RCS i P
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• 2nd step depressurize RCS using Prz sprays or PORV if RCPs unavailable.
• 3rd Step terminate SI to prevent re-pressurization of RCS.
RE
SS
UR
E(P
SIG
)
1000
1200
1400
1600
1800
2000
2200
(1)
(2)
(4) (10)
BREAK FLOW(N) = # of FAILED TUBES
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0 20 40 60 80 100 120 140
FLOW (LB/SEC)
160 180 200
RC
SP
Figure 4.6-5 Equilibrium Break Flow4.6-41
600
800 SI FLOW
0
200
400
500
600
700
MP
ER
AT
UR
E(°
F)
300
400
T hot
T cold
RCS COOLDOWN INITIATED
Cooldown RCS to less than SaturationTemp for affected SG pressure.
i.e. SG Press 800-899 Core Exit TC 470o
Saturation for 899# is ~532o
20
10 20 30 40 50 60TIME (MINS)
TE
Figure 4.6-6 RCS Response - Offsite Power Available4.6-43
0
0
100
200 This will allow depressurization of RCSto SG press and stop the leak with the RCS sub-cooled.
TRIP
Cooldown Initiated
Prz Spray Initiated
SI TerminatedSI Flow=Break Flow
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Prz Spray Terminated
Break FlowStopped
100
125
150
75
W(L
BM
/SE
C)
SI FLOW
As Primary is depressurizedBreak flow decreases and SIFlow increases refilling Pzr.
SI Flow=Break Flow Cooldown Initiated
SI Termination
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10 20 30 40
TIME (MINS)
50 60
Figure 4.6-8 SI Flow and Break Flow4.6-47
0
25
50FL
OW
BREAK FLOW
0
SI Flow Break Flow Cooldown Initiated
Cooldown Stopped
40
60
80
100
UR
IZE
RL
EV
EL
(%s
pan
)SERIES COOLDOWN
andDEPRESSURIZATION
CONCURRENTCOOLDOWN and
DEPRESSURIZATION
SI Termination
SI Termination:RCS Sub Cooling >25o
SG FF >720gpm or 1 Intact SG >7%
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TIME(MINS)
40 50 60
20
PR
ES
SU
0
0 10 20 30
Figure 4.6-10 Pressurizer Level Response - RCS Cooldown and Depressurization4.6-51
gpRCS Press stable or increasingPzr level >5%
2000
22002400
AFFECTED SG PRESSURE
PR
ES
SU
RE
(PS
IG)
600
800
1000
1200
1400
1600
1800RCS PRESSURE
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0 10 20 30 40 50 60TIME (MINS)
Figure 4.6-7(a) Multiple Tube Failure Response4.6-45
AFFECTED SG PRESSURE
0200
400
When several tubes fail RCS pressure can drop below SG pressure during recovery. This can result in secondary – to – primary leakage which will dilute RCS boron concentration thus effecting SDM.
80
100
AFFECTED SG
INTACT SG
LE
VE
L(%
NA
RR
OW
RA
NG
E)
20
40
60
25
Figure 4.6-79(b) Multiple Tube Failure Response4.6-45
0 10 20 30 40 50 60
TIME (MINS)
SG
WA
TE
R
0
20
40
60
80
100
LE
VE
L(%
Na
rro
wR
an
ge
)
AFFECTEDSG
INTACTSG
Affected SG level recovers faster andhigher for = feedflow
100% Narrow Range is not a Solid SG
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50 60
0
20
SG
WA
TE
RL
Figure 4.6-12 Steam Generator Levels4.6-55
TIME (MINS)0 10 20 30 40
Break Flow Stopped7%
SGTR with LOOP• No RCPs – natural circulation cooldown
required, slows recovery.
• No Steam Dumps to condenser – must use SG PORVs.
• Prz Sprays unavailable must use Prz
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• Prz Sprays unavailable – must use Prz PORVs or Aux Spray.
• Overall this results in increased potential for rad releases and SG overfill
500
750
1000
12501300
PR
ES
SU
RE
(psi
g)
OFFSITE POWER AVAILABLE
LOSS OF SITE POWER
Steam Dumps 1092#Higher SG pressure SG PORV (1125#), or Safeties (1st @ 1170#). Can results in rad release from affected SG
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0
250
SG
P
Figure 4.6-13 Steam Generator Pressure Following Reactor TripWith and Without Offsite Power
4.6-57
4
TIME (MINS)
0 2 6 8 10
Higher no-load Tave (1125#)Slower Pressure Decrease
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Trip
T HOT
T COLD
EM
PE
RA
TU
RE
(0F
)
400
300
700
600
500
RCPs Trip resulting in Natural Circulation
30
6 8 10TIME (mins)
T
Figure 4.6-15 RCS Temperature Following Reactor Trip Without Offsite Power4.6-61
200
100
0
0 2 4
40
60
80
100
110
PF
LO
W(%
No
min
al)
Without RCPs running upperhead becomes inactive.Stays hot and prone to voidingduring subsequent cooldown
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10
TIME (min)
0
20
40
LO
OP
Figure 4.6-16 Natural Circulation Flow Following Loss of Offsite Power4.6-63
0 2 4 6 8
300
400
500
600
700
MP
ER
AT
UR
E(°
F)
THOT
TCOLD
Cooldown done with PORV on SG vs Steam Dumps
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0
0
10 20 30 40 50 60
100
200
TE
M
TIME (MINS)
Figure 4.6-17 Intact RCS Temperature, Without Offsite Power4.6-65
When the affected SG is isolatedThat loop may stagnate.
CS
PR
ES
SU
RE
(PS
IG)
1000
1250
1500
1740
2000
2250
PRZR PORV OPENED
Cooldown SGPORV
SI Termination
33Figure 4.6-18 RCS Pressure Response, Without Offsite Power
4.6-67
RC
TIME (MINS)
0 10 20 30 40 50 60
0
250
500
750 PRZR PORV CLOSED
During DepressurizationHead may void causing rapidIncrease in pressurizer level.
Questions
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Ginna SGTR
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January 25, 1982
RX
M
STBY AUX FEED PUMP
MOTOR DRIVENAUX FEED PUMP
HHSI
LPSI
SG
STBY AUX FEED PUMP
MOTOR DRIVENAUX. FEED PUMP
SV
ATMOS
MSIV
4 SAFETYVALVES
STOPVALVES
GOVERNORVALVES
TURB/GEN
ATMOS
4 SAFETYVALVES
MSIVBLOCKVALVE
TOPRT
PORV
M
HHSI
CIRCULATINGWATERPUMPS
SGPZR
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AUX. FEED PUMP
CHARGING
HI-STRENGTH CONCRETESTEEL LINED CONTAINMENT
cvcsRCP
MAIN FEEDWATERPUMP
TURBINE DRIVENAUX. FEED PUMP
MAIN FEEDWATERPUMP
Fig 4.6-19 GINNA OVERVIEW
PRESSURIZER
M RHR
SISTE
SEAL RETURNRELIEF
PZRSPRAYVALVESPT
LTACC.
M SIS
RUPTUREDISC
TE LT
DEMINERALIZEDWATER SPRAY
TE
PCV435
AUXILIARY SPRAY
LETDOWNRELIEF
VENTHEADER
PT
HEATER CONTROL
TE
M
PCV 431C
PCV 430
PCV434
M
TE
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M
MM
RCPM
S/G
STEAM GENERATOR
CHARGING
TE
CVCSLETDOWN LINE SIS
M
LP SISACC.
MSIS
M
AUXILIARYCHARGING
TE
TE
PT
M LOOP A
RHR HEATREMOVAL LOOP
SISLP SIS
M
TE
LOOP BRCP
Figure 4.6-20 Ginna RCS Piping and Instrumentation
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47% Normal Pzr Level
Increase Charging and Load reduction stabilize Press and Level decreaseThis indicates leak ~750gpm
R T i l
39
Tube Rupture at ~09:25
Plant Computer Data Available
Rx Trip on lowPress at 1900#
BREAK FLOW(B S/G at 1050 psig)
3 SAFETYINJECTIONPUMPS
SU
RE
(ps
ig)
1200
1300
1400
1500
1600
1700
40
0800
200 400 600
FLOW (gpm)
800 1000
900
1000
1100PR
ES
S
Figure 4.6-23 Ginna SGTR - SI and Break Flow4.6-77
SI pumps recover RCS pressure to about1350# and with S/G at about 1050 there is~400 gpm going into “B” SG
"B" S/GISOL.
RCPTRIP
RXTRIP
PZRPORV
LOOP "B" COLD LEGTEMPERATURE
LOOP "A" COLD LEGTEMPERATURE
ESTIMATED
600
560
520
480
440
400
360
320TE
MP
ER
AT
UR
E(0
F)
Cooling downon “A” SG
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START"A" RCP
TIMEPM
9:00 10 20 30 40 50 10:00 10 20 30 40 50 11:00 10 20 30 40 50 12:00AMAMAM
280
240
Figure 4.6-24 Ginna SGTR - Cold-leg Temperature4.6-79
At 9:40 “B” SG is isolatedB coolant loop stagnates.B loop reverse flow out theRuptured tube
Attempt to depressurizeUsing PORV and it sticksOpen on 4th cycle and isIsolated. Rapidly filling Pzr.
SI Termination Criteria exists
Tube Leak increasing“B” S/G Pressure
2 Chg Pps no letdownMaintains RSC ~40#>SGBreak Flow continues
Break FlowReverses
42
Heat removal changed to PORV
Letdown relief open
43
LE
VE
L(%
)"B" S/G LEVEL
"B" S/G PRESSURE
100
90
80
70
60
50
40
30
20
1100
1000
900
800
700
600
500
400
300
PR
ES
SU
RE
(ps
ig)
Control break-flow by keeping RCS pressure ~25# <SG press
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RCS PRESSURE
12:00NOON1/25/82
1 2 3 4 5 6 7 8 9 10 11 1 2 3 4 5 6 7 8 9 10 11 1 2 3 4 5 6 7 8 9 10 1112:00NOON1/26/82
12:00MIDNIGHT
12:00MIDNIGHT TIME
20
10
0
300
200
100
Figure 4.6-27 Ginna SGTR - Long-Term Cooldown and Depressurization4.6-85
Ginna Lessons Learned• Tripping RCPs slowed down and
complicated recovery actions.• PORV failure resulted in sec-to-pri leakage,
RCS steam void formation, and Prz overfill.• Opening of letdown iso valves resulted in
overpressure of PRT
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overpressure of PRT.• PRT rupture disc burst resulting in release of
Reactor Coolant to Containment.• Failure of SG Safety resulted in Rad releases
& complicated recovery actions (lower SG pressure.
ObjectivesObj-1 Discuss why operator intervention is necessary to limit or prevent radiological releases during a Steam Generator Tube Rupture (SGTR) event.
Without operator action the affected SG and associated
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Without operator action the affected SG and associated steam line will fill water solid. S/G PORVs and Safety valves are not designed to pass water and may fail open. The steam lines may fail due to the weight of water in the lines. These failures could result in a containment bypass event releasing raditation directly to the environment.
Obj-2 Discuss the primary-side and secondary-side indications of an SGTR in the control room.
• Primary– Decreasing pressurizer level
– Decreasing RCS pressure
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• Secondary– Condenser Offgas rad-monitor
– SGBD rad-monitor
– Main Steam rad-monitor
– SG Steam flow / feed flow mismatch
– SG water level deviation alarms
Obj-3 Discuss how the affected SG may be identified either prior to or following the reactor/turbine trip.
• Feedflow / Steam Flow mismatch• Water level in affected SG returns on scale before
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other SGs
• Water level in affected SG higher than other S/Gs
• SGBD rad-monitor (if not common)
• Main Steam Line rad-monitor
Obj-4 List the initial actions taken by the operator once the affected SG has been identified.
• Isolate the affected SG.1. Isolate Feedwater & AFW.2. Close MSIV.
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2. Close MSIV.3. Raise Setpoint on Affected SG ASDV
• These actions help minimize radiological releases & prevent SG overfill.
• Isolating the affected SG also allows use of the non-affected SG for cooling down RCS.
Obj-5 Discuss the actions required to stop the primary-secondary leakage.Isolate the affected SG
• Cooldown RCS to saturation temp below affected SG pressure
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• Depressurize RCS down to SG Pressure
• When pressurizer level is restored and SI termination criteria are meant stop SI pumps.
• Pressurizer level will slowly decrease as RCS depressurizes to SG Press and break flow stops
Obj-6 Discuss the problems associated with the following: Secondary-to-primary leakage, SG Overfill.
• Secondary to Primary Leakage– Dilution of primary (SD Margin)
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• SG Overfill– Safety Valves and PORVs are not designed to
pass water and may fail.
– SG Lines fill with water to the MSIV and main steam line may fail.
Obj-7 List the principal systems & components affected by a loss of offsite power (LOOP).
• No RCPs – natural circulation cooldown required, slows recovery.
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q , y
• No Steam Dumps to condenser – must use SG PORVs.
• Prz Sprays unavailable – must use Prz PORVs or Aux Spray (backup).
Obj-8 Discuss how plant cooldown and pressure control are accomplished with an SGTR and LOOP.
• Cooldown using SG PORV vs Stm Dumps
• Depressurize using PORV vs Normal or
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p gAux Spray
OBJ-9 Discuss what affect the following events had on the SGTR transient at the Ginna Plant:
Tripping of the reactor Coolant Pumps,
Failure of pressurizer power-Operated relief valve,
Automatic operation of letdown valves,
Pressurizer relief tank failure, and
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,
Steam generator Safety Valve Failure.
The End
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