safety of nuclear reactors - 茨城大学...3 university of fukui accident (2/2) • computer codes...
TRANSCRIPT
1
University of Fukui
Safety of Nuclear Reactors
Professor H. MOCHIZUKINuclear Reactor Thermal Hydraulics Division
Research Institute of Nuclear EngineeringUniversity of Fukui
2
University of Fukui
Accident (1/2)• Design Basis Accident: DBA• Assumption of simultaneous double ended break
• Installation of Engineered Safety FeaturesEmergency Core Cooling System: ECCSAccumulated Pressurized Coolant Injection System: APCILow Pressure Coolant Injection System: LPCIHigh Pressure Coolant Injection System: HPCI
3
University of Fukui
Accident (2/2)
• Computer codes are used to evaluate temperature behavior of fuel bundle.
• Computer codes should be validated.• Blow-down and ECC injection tests have
been conducted using mock-ups.• RELAP5/mod3 and TRAC code are
developed and validated.
4
University of FukuiECCS
Main System Diagram of Fugen
Containment Air Cooling System
Shield Cooling System
Reactor AuxiliaryComponent CoolingWater System
Reactor AuxiliaryComponent CoolingSea-Water System
Dump valve
Control Rod Drive
Residual Heat Removal System (RHR)
High Pressure Coolant Injection System (HPCI)
Low Pressure Coolant Injection System (LPCI)
Reactor Core Isolation Cooling System(RCIC)
Containment SpraySystem
(APCI)
Relief valve
Bypa
ss v
alve
Heavy Water CoolingSystem
Turbine
Feed Water System
Condensate Tank
Sea water
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University of Fukui
Blow-down experiment
6
University of Fukui6MW ATR Safety Experimental Facility
Downcomer(275.7mmID,6.8mL)
EL 4.5
Pump
Outlet pipes(74mmID, 10.5mL, 2 deg.)
Fig. 6 Schematic of Safety Experiment Loop (SEL)
P,T
EL: Elevation in mID: Inner diameterL : Length from a component
to the next arrow.
P
TPressure transducerThermocouple
Low powerheaters
3.7mL, 200kW
Main steamisolation valve
P,T
P,T
P,T
P,T
P,T Steam drum(1525mmID,4.46mL)
Water drum(387mmID,3.9mL)
Inlet ppes(62.3mmID,12mL) EL0.9
Check valves
Turbine flow meter
(186mmID, 24.6mL)
Connecting pipe(186mmID, 15.6mL)
EL 0.4
EL 11.5EL 9.7EL 9.51
EL 1.64
High powerheater3.7mL, 6MW
EL 2.27
EL 5.97EL 7.0
EL 8.45Shield plug
EL 2.3
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University of Fukui
Water level behavior after a main steam pipe break
-0.8
-0.6
-0.4
-0.2
0
0.2
0
3
6
9
12
15
0 10 20 30 40 50
Drum water level
Dowmcomer water level
Dru
m w
ater
leve
l (m
)
Dow
ncom
er w
ater
lwve
l (m
)
Time (sec)
100 mm break at main steam pipe
Device oscillation due to break
Drum water level increase duringdowncomer water level decrease
Break
D/P
Beak
D/P
Drum
Reactor core
8
University of FukuiSimulated fuel bundle
Tie rod14.5 OD
405 460 360 260 260 260 260 260 260 260 280 320 400
1020 Active heated length : 3700 mm
Fig. 7 Power distribution of 36-rod high power heater
Cross sectional view of 36-heater bundleBottom
Top
29.72
27.04
18.92
33.12
47.30
52.01
34.69
49.55
54.49
36.26
51.81
56.07
25.23
36.04
39.63
493 740 740 1234 493
Unit in mm
OuterMiddle
InnerHeater pin14.5 OD
Power of clusterat each zone(kW/m)
V IV III II I
T T T TT T T T
T Thermocouple position
Spacer Tieplate
Gadlinia pin14.5 OD
Location of maximum axial peaking
Local peaking is high for the outer rods due to the neutronic characteristics
9
University of FukuiThermocouple positions
180° 180°
B19-3
FT-3
H13-3
270° 90°
Section 3
D21-3
B35-3
H33-3 H6-3
D17-3
G5-3
D9-3
F23-3C2-3
B15-3F31-3
D29-3
D3-3
F11-3H25-3
B27-3
B7-3
180°
0°
H19-1
DT-1270°
0°
90°
Section 1
G22-1E8-1
H1-1
D34-1F17-1
A4-1
D14-1A31-1
E28-1
A11-1C25-1
D36-4
GT-4
G4-4
270° 90°
Section 4D20-4
F34-4
D16-4C1-4
B18-4
H5-4
B22-4
B10-4
F14-4B30-4
D28-4
A3-4
F26-4
H24-4
H12-4
D8-4
H32-4
0°
B20-2
E1-2
F12-2
270° 90°
Section 2
D22-2
B2-2F32-2
B8-2
D18-2
B16-2
H34-2
E1-2
H14-2D30-2
B28-2 H26-2
D10-2F24-2
B36-2
G6-2
Fig. 8 Thermocouple positions on high power heater rods
E20-6
AT-6
C13-6
270°
0°
90°
Section 6
G10-6
H35-6
B32-6E16-6
F29-6
F3-6
D26-6
C6-6
H7-6
A23-6
90°
180°
D19-5
HT-5
F13-5
270°
Section 5D35-5
F33-5B17-5
D7-5 B21-5
B9-5F6-5
D15-5H31-5 H4-5
B29-5
B3-5
H11-5
A2-5H23-5
F25-5
D27-5
A19-2
A7-2 C21-2
A1-2C9-2
E23-2
A35-2
C17-2G35-2
F5-2
C3-2
E11-2
A15-2E31-2
G13-2
C29-2
G25-2
A27-2
E14-3
F4-3A10-3
G32-3
C16-3
A22-3B1-3
C8-3E34-3 A18-3
C20-3C36-3
E26-3
G24-3
C28-3
G12-3A30-3
G23-4D2-4A9-4
A21-4C7-4
C19-4
F6-4E33-4A17-4
C35-4
C27-4
E25-4G11-4
A29-4
E13-4
G31-4
C15-4
0°
C22-5D1-5
A8-5
A20-5
E5-5E32-5A16-5
C18-5G34-5
A36-5
E12-5
G26-5
E24-5
C10-5
A28-5
G14-5C30-5
180° 180°
0°
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University of Fukui
Cladding temperature measured in a same cross section of heater bundle
Data at section-II
0
100
200
300
400
500
0 10 20 30 40 50 60Time (sec)
。
Fig. 14 Experimental cladding temperature for 150 mmdowncomer break
11
University of FukuiCalculation model of pipe break experiment
Waterdrum
Checkvalves
Inlet pipes
Outlet pipes
Main steamisolationvalveMain steam pipe
Steam drum
Fig. 9 Nodalization scheme for ATR Safety Experiment Loop
Break
Pump
Upper shield
Lowerextension
0.8
1.15
1.05
1.10
0.6
Powerdistribution200kW
5 ch.
Max. 6 MW1 ch.
AccumlatedPressure CoolantInjection system
Downcomer
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University of Fukui
-5
0
5
10
15
20
0 10 20 30 40 50 60
Calculatin
Experiment
Time (sec)
0
100
200
300
400
500
600
0 10 20 30 40 50 60Time (sec)
。
CalculationExperiment
Fig.16 Behavior of cladding temperature after 100 mm downcomer break
Section-II
Comparison between experimental result and simulation
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University of Fukui
100
150
200
250
300
350
400
450
500
550
600
0 5 10 15 20 25 30 35 40 45 50
Calculation
Experiment
Time (sec)
Temperature (℃
)
Improvement of blow-down analysis by applying statistical method
Scram ECCS operation
Downcomer 100 mm break
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University of Fukui
100
150
200
250
300
350
400
450
500
0 5 10 15 20 25 30 35 40 45 50
CalculationExperiment
Time (sec)
Tem
pera
ture
(℃)
Downcomer 150 mm break
Improvement of blow-down analysis by applying statistical method
H. Mochizuki, A Validation of ATR LOCA Thermal-hydraulic Code with a Statistical Approach, Journal of Nuclear Science and Technology, 37, 8 (2000), pp.697-709.
15
University of Fukui
7
Evaporator Super-heater
[13]
[1]~[7]
[47]15
10-2
8 9
12
1
5
6
-1
[8] IHX
Air cooler
[28]
[29]
[30]
High-pressureplenum
Pump
Upper plenum
[31]
[32]
[33][15]
[16]
[34][35]
[17]
[14]to turbine
11[10]
[11]
[12]
[9]
Pump
3
Link 1-Link 6: 1st to 6th layer (Inner driver)Link 7: 7th & 8th layer (Outer driver)Link 8: 9th to 11th layer (Blanket)Link 9: Center CRLink 10: CRsLink 11: Bypass
[24]
[18]
[19]
[46]
[23]
19
[49]
1922
-4
2021
24
[42]
[43]
[44] [45]
23
[25]
[26]
[27]
13
[48]
1816
-3
14
18
[37]
[38]
[39] [40]
17
[20]
[21]
[22]
[36]
[41]to B-loop
to C-loop
A-loopB-loop
C-loop From B&C loops
14 16
4
15
2
2 3020
#2#3#4#5
25
#1
1
13
1
-1
Low-pressure turbine
Deaerator
Feedwaterpump
Condenser
High-pressure feedwater heaters Low-pressure feedwater heaters
Low-pressure side
1
1
16
16
14
14
Low-pressureplenum
Application of stochastic method to FBR analysis
H. Mochizuki, Development of the plant dynamics analysis code NETFLOW++, Nuclear Engineering and Design, 240, (2010), pp. 577-587.
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University of Fukui
Application of stochastic method to FBR
Cool-down velocity of plenum temperature (℃/s)
Freq
uenc
y(-)
Tem
pera
ture
(℃
)
Time (sec)
Tem
pera
ture
(℃
)
Time (sec)
Measured at MonjuAnalysis
Measured at MonjuAnalysis
Result:The measured temperature transient has been included in the group of calculated curves. Most non-safety side value could be evaluated taking into account various statistical errors.
[1] N. Kikuchi and H. Mochizuki, Application of statistical method for FBR plant transient computation, FR’13, Paris (2013-March)
Statistics if measured Nu number Evolution of plenum temperature during turbine trip
Method:Plant parameters are investigated by 10,000 trials of the Monte-Carlo calculation for 43 factors which can affect on the plenum temperature.
Freq
uenc
y(-)
Pe number: 1000
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University of FukuiSevere accident
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University of FukuiHeat transfer of melted fuel to material
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University of Fukui
Heat transfer between melted jet and materials
0.1
1
10
100
1000
104
103 104 105 106
NaCl-Sn 10 1100NaCl-Sn 20 900NaCl-Sn 20 1000NaCl-Sn 20 1100NaCl-Sn 30 900NaCl-Sn 30 1000NaCl-Sn 30 1100Al2O3-SS 10 2200Al2O3-SS 10 2300Al2O3-SS 10 2300
Rej
Nu m
/Pr j
Material, Dj(mm), T
j(oC)
Num = 0.0033 Re
j Pr
j
Comparison of Nusselt number between present data and data fromSaito et al.1) and Mochizuki2).1)Saito, et al., Nuclear Engineering and Design, 132 (1991)2)Mochizuki, Accident Management and Simulation Symposium, Jackson Hole, (1997).
Num = 0.00123 Re
j Pr
j
20
University of Fukui
Fuel melt experiment using BTF in Canada
21
University of FukuiFuel melt experiment using CABRI
22
University of Fukui
Source term analysis codesGeneral codes
NRC codes ORIGEN-2, MARCH-2, MERGE, CORSOR, TRAP-MELT, CORCON, VANESA, NAUA-4, SPARC, ICEDF
IDCOR codes MAAP, FPRAT, RETAIN
NRC code (2nd Gen.) MELCOR
Precise analysis codes
Core melt SCDAP, ELOCA, MELPROG, SIMMER
Debris-concrete reaction CORCON
Hydrogen burning HECTOR, CSQ Sandia, HMS BURN
FP discharge FASTGRASS, VICTORIA
FP behavior in heat transport system
TRAP-MELT
FP discharge during debris-concrete reaction
VANESA
FP behavior in containment CONTAIN, NAUA, QUICK, MAROS, CORRAL-II
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University of Fukui
CONATIN code
(1)(2)(3)
(4)(5)(6)
(7)(8)
(9)(10)
(11)
(12)
(14)
(13)
Filter Blower
Stack
Air
Air
Annulus
Containment spray In case of containment bypassContainmentrecirculation
system
Steam release pool
Water flowGas flow
24
University of Fukui
Fluid- structure interaction analysis during hydrogen detonation
25
University of Fukui
Analysis of Chernobyl Accident- Investigation of Root Cause -
H. Mochizuki, Analysis of the Chernobyl Accident from 1:19:00 to the First Power Excursion, Nuclear Engineering and Design, 237, (2007), pp.300-307.
26
University of Fukui
Schematic of Chernobyl NPP1. Core2. Fuel channels3. Outlet pipes4. Drum separator5. Steam header6. Downcomers7. MCP8. Distribution group headers
9. Inlet pipes10. Fuel failure detection equipment11. Top shield12. Side shield13. Bottom shield14. Spent fuel storage15. Fuel reload machine16. Crane
Electrical power 1,000 MWThermal power 3,200 MWCoolant flow rate 37,500 t/hSteam flow rate 5,400 t/h (Turbine)Steam flow rate 400 t/h (Reheater) Pressure in DS 7 MPaInlet coolant temp. 270 0COutlet coolant temp. 284 0CFuel 1.8%UO2Number of fuel channels 1,693
27
University of Fukui
Elevation Plan
28
University of Fukui
Above the Core of Ignarina NPP
29
University of Fukui
Core and Re-fueling Machine
30
University of Fukui
Control Room
31
University of Fukui
Configuration of inlet valve
1
2
3 41. Isolation and flow control valve2.Ball-type flow meter 3.Inlet pipe4.Distribution group header
32
University of Fukui
Drum Separator
33
University of Fukui
Configuration of Fuel Channnel
Effective core region
S.S.
Diffusion welding
Zr-2.5%Nb Roll region
Electron beem welding (EBW)
Fuel assembly
Spacers
Connecting rod 20
0 m
m
Diffusion welding
Position: -0.018m
-8.283
-8.483
-8.969
-15.933
(-16.478)
-16.671
EBW -16.433
(-16.588)S.S.
80
72
77 -16.633
Zr-2.5%Nb
Welding
(-8.335)
-12.451
-14.192
34
University of Fukui
Heat Removal by Moderation
Pressure tube
Graphiteblocks
Graphite ring
Gap of 1.5mmCoolant
φ88mmφ114mm
φ111mm
φ91mm Heat generated in graphite blocks is removed by coolant
Maximum graphite temperature is 720℃at rated power
35
University of Fukui
RBMK & VVER
Russia
Lithuania
Finland
Germany
Ukraine
36
University of Fukui
Objective of the Experiment
• Power generation after the reactor scram for several tens of seconds in order to supply power to main components.
• There is enough amount of vapor in drum separators to generate electricity.
• But they closed the isolation valve.• They tried to generate power by the inertia
of the turbine system.
37
University of Fukui
Report in Dec. 1986
38
University of Fukui
0500
10001500
2000250030003500
25:00
:00:00
25:01
:00:00
25:13
:05:00
25:23
:10:00
26:00
:28:00
26:01
:00:00
26:01
:23:04
26:01
:23:40
Trend of the Reactor PowerTh
erm
al P
ower
(MW
)
Scheduled power level for experiment
30MW
Power excursion
secminhourday
20-30% of rated power
200MW
39
University of FukuiTime Chart Presented by USSR
再循環流量
給水流量 蒸気ドラム圧力出力
安全棒挿入開始Safety rod insertion
Feedwater flowrate Power
Recirculation flowrate
DS pressure
40
University of Fukui• T. Wakabayashi, H. Mochizuki, et al., Analysis of the Chernobyl Reactor Accident (I) Nuclear and
Thermal Hydraulic Characteristics and Follow-up Calculation of the Accident, J. Atomic Energy Society of Japan, 28, 12 (1986), pp.1153-1164.
• T. Wakabayashi, H. Mochizuki, et al., Analysis of the Chernobyl Reactor Accident (I) Nuclear and Thermal Hydraulic Characteristics and Follow-up Calculation of the Accident, Nuclear Engineering and Design, 103, (1987), pp.151-164.
• Requirement from the Nuclear Safety Committee in Japan
Feed water
Water level
Drum pressure
Recirculation flow rate
Neutron flux
Result in the Past Analysis (1/2)
41
University of Fukui
Result in the Past Analysis (2/2)
Power at 200 MW
Power at 48,000 MW
Power just before the accident was twice as large as the report. Why???
Result of FATRACcode is transferred, and initial steady calculation was conducted.
Timing of peak was different. Why???
???
42
University of Fukui
Possible Trigger of the Accident
• Positive scram due to flaw of scram rods• Pump cavitation• Pump coast-down• Opening of turbine bypass valve
(6.96MPa)
43
University of Fukui
Calculation Model by NETFLOW++ Code
7 MPa
8.2 m Flow control valve
Flow meter
EL:25.6m
Check valve
EL: 0
Suction header
Check valveThrottling-regulating valve
EL:21.3EL:20.0
EL:11.8EL:11.6
-1-21
[1]
[2]
EL:33.65
EL:14.85
EL:5.9
200 (3200) MWt
2
-21
[9]
[10]
EL:7.6
[3]
93
[11]
[4][5][6][7]
11213141516171
91
96
3
92
Drum separators2 for one loopL: 30mID: 2.6mt: 0.105 m
Feed water for one DSWf =115 t/h, 140 oC(1453 t/h, 177-190 oC)
[8]DowncomersOD: 0.325 mID: 0.295 mL: 23 - 33.5 mN: 12 (for each DS)
OD: 1.020mID: 0.9mL: 21m
Cooling pump :2 for one DSH: 200 m1000 rpm5500 kWGD2=1500 kg m25250 t/h ×2(8000 m3/h for one pump)
Pressure headerOD: 1.040mID: 0.9mL: 18.5m
Distribution group headersOD: 0.325mID: 0.295mL: 24mN: 16
OD: 0.828mID: 0.752mL= 36m
OD: 0.828mID: 0.752mL= 34m
Feeder pipesL: 22.5 - 32.5 mOD: 0.057 mID : 0.050 mEL: 6.3
Fuel channelOD: 0.088 mID: 0.080mN: 1661ch. Outlet pipe
L: 12.7 - 23mOD: 0.076mID : 0.068m
EL: 9.3
1
EL: 12.15
EL: 9.6
Feed water pipe
Main steam pipe
0.08
0.091
0.088
0.111
Graphite ring
Fuel channel
0.02
94
95
44
University of FukuiTrigger of the Accident
P.S.W. Chan and A.R. DasterNuclear Science and Engineering,103, 289-293 (1989).
• Positive scram
Andriushchenko, N.N. et al., Simulation of reactivity and neutron fields change, Int. Conf. of Nuclear Accident and the Future of Energy, Paris, France, (1991).
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University of Fukui
Trigger of the Accident (cont.)
8.0
1.0
5.0
1.5
Negative reactivity
Positive reactivityGraphite displacer
Scram rod (24rods)inserted by AZ-5 button
Fuel 2×3.5m
Graphite block
Water Column
46
University of Fukui
Simulation from 1:19:00 to First PeakData acquired by SKALA
5
6
7
8
9
10
-600
-400
-200
0
200
400
0 60 120 180 240 300
Flowrate (m3/s)P (MPa)Flowrate (calc.)Pressure (calc.)
DS water level (mm)Feed water flowrate (kg/s)Reactor power (calc.)DS water level (calc.)
Flow
rate
(m3 /s
), Pr
essu
re (M
Pa)
DS
wat
er le
vel (
mm
), Fe
edw
ater
flow
rate
(kg/
s)
Time (sec) Push AZ-5 button
Close stop valve:Turbine trip
Trend of parameters for one loop from 1:19:00 on 26 April 19861:19:00
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University of Fukui
Behavior of Steam Quality
-0.02
-0.01
0
0.01
0.02
0.03
0.04
0.05
0 50 100 150 200 250 300
TopCenter
Tim e (sec)
Ther
mal
equ
ilibriu
m s
team
qua
lity,
x (-
) Turbine tripRCP trip
Push AZ-5 button
Water
Two-phase
48
University of Fukui
Void Characteristic
0
0.2
0.4
0.6
0.8
1
0 0.2 0.4 0.6 0.8 1
Measured
Correlation
Void fraction, α (-)
Thermal equilibrium steam quality, x (-)
Pressure 7MPa 2Void fraction increase
Void fraction increase
49
University of FukuiNuclear Characteristics
-1.8 10-5
-1.6 10-5
-1.4 10-5
-1.2 10-5
-1 10-5
-8 10-6
0 500 1000 1500 2000 2500
Δk/k/℃
T (℃)
0
0.0001
0.0002
0.0003
0.0004
0.0005
0 20 40 60 80 100
Δk/k/%Void
Void fraction (%)
Doppler Void
50
University of FukuiPeak Power and its Reactivity
0
5 104
1 105
1.5 105
2 105
2.5 105
3 105
3.5 105
270 275 280 285 290
Power reported by USSR (MW)NETFLOW
Pow
er (M
W)
Time (sec)
-3
-2
-1
0
1
2
3
270 275 280 285 290
TotalScram (input)DopplerVoid
Rea
ctiv
ity ($
)
Time (sec)1:23:30
Push AZ-5 button
51
University of Fukui
Relationship between Peak Powerand Peak Positive Reactivity
0.1
1
10
100
0.75 0.8 0.85 0.9 0.95Peak positive reactivity ($)
Peak
val
ue o
f firs
t pow
er p
eak
mul
tiple
s fu
ll po
wer
52
University of FukuiJust after the Accident
53
University of Fukui
Control Room and Corium beneath the Core
54
University of Fukui
55
University of Fukui
Before earthquake (14:46) & After Tsunami(15:45) at Fukushima-1 on 11 Mar. 2011
After the Tsunami
Area damaged by the Tsunami
Before the earthquake
Seawater pumps
Intake of seawater Heavy-oil
tanks
56
University of Fukui
Severe accident at Fukushima-1After Tsunami
StackSuppression pool
DC powerOff-site power
ECCS
Seawater pumps
○
○
Cooling pump for suppression pool
原子炉圧力容器
Containment vessel
Fuel pool
×Loss of off-site power
Diesel generators
×
××
×
Operators injected water into the reactorcore by RCIC. After the loss-off-all-AC-power, water injection was impossible. Damaged
by Tsunami
RCIC pump
57
University of FukuiHand calculation to estimate uncovery
• Assumption: Reactor diameter=4m, water level from top of fuel=4m
• Water inventory ≒ 50t• Latent heat at 7MPa≒1500kJ/kg• Initial heat generation rate of Unit-1(460,000kW)
≒460,000/0.3 ≒ 1,500,000kW (70% of heat is released to seawater)Decay heat ratio at around 1000 sec.≒ 2%Heat generation rate by decay heat≒1500000×0.02=30,000kW
• Mass evaporated for 1000 sec. : M(kg)M=30,000×1000÷1500
=20000kg=20t
58
University of Fukui
Real water level and detected water level
V
V
Low pressure(Spent fuel pool)Volume of steam expands 1600 times of water at 0.1 MPa.
V
High pressure (Core)Volume of steam expands 21 times of water at 7 MPa.
Cv
燃料
燃料
Mochizuki, H. et al., Core coolability of an ATR by heavy water moderator om situation beyo9nd design basis accidents, Nuclear Engineering and Design, 144 (1993), pp293-303.
Experiment
59
University of Fukui
Natural circulation after the loss of AC power and sea water pump
Secondary sodium
Primarysodium
IHX
Primary pump
Core
Air cooler(AC)
Evaporator(EV)
Turbine Generator
Feed water pump
Sea water cooler
Condenser
Secondary pump
Super heater(SH)
[1] H. Mochizuki, Plant Behavior of a Fast Breeder Reactor under Loss of AC Power for Long Period, Nuclear Engineering and Design, 245, (2012), pp.19-27.
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University of Fukui
7
Evaporator
Super-heater
[13]
[1]~[7]
[47] 1510
-2
8 9
12
1
5
6
-1
[8] IHX
Air cooler
[28]
[29]
[30]
High-pressureplenum
Pump
Upper plenum
[31]
[32]
[33][15]
[16]
[34][35]
[17]
[14]to turbine
11[10]
[11]
[12]
[9]
Pump
3
Link 1-Link 6: 1st to 6th layer (Inner driver)Link 7: 7th & 8th layer (Outer driver)Link 8: 9th to 11th layer (Blanket)Link 9: Center CRLink 10: CRsLink 11: Bypass
[24]
[18]
[19]
[46]
[23]
19
[49]
1922
-4
2021
24
[42]
[43]
[44] [45]23
[25]
[26]
[27]
13
[48]
1816
-3
14
18
[37]
[38]
[39] [40]17
[20]
[21]
[22]
[36]
[41]to B-loop
to C-loop
A-loopB-loop
C-loop From B&C loops
14 16
4
15
2
2 3020#2#3#4#5
25#1
1
13
1
-1
Low-pressure turbine
Deaerator
Feedwater pump
Condenser
High-pressure feedwater heaters Low-pressure feedwater heaters
Low-pressure side
1
1
16
16
14
14
Low-pressureplenum
Analytical Model of Whole Heat Transport Systems
H. Mochizuki, Development of the plant dynamics analysis code NETFLOW++, Nuclear Engineering and Design, 240, (2010), pp. 577-587.
61
University of Fukui
Station Blackout Event of “Monju”
200
300
400
500
600
700
0
100
200
300
400
500
0 1 2 3 4 5 6 7
Inner core Ch.1 outletInner core Ch.2 outletInner core Ch.3 outletInner core Ch.4 outletInner core Ch.5 outlet
Inner core Ch.6 outletOuter core outletBlanket outletUpper plenumHigh pressure plenum
A loopB loopC loop
Flow
rate
(kg/
s)
Tem
pera
ture
(o C)
Time (day)
200
300
400
500
600
700
0
100
200
300
400
500
0 1000 2000 3000 4000 5000
Inner core Ch.1 outletInner core Ch.2 outletInner core Ch.3 outletInner core Ch.4 outletInner core Ch.5 outlet
Inner core Ch.6 outletOuter core outletBlanket outletUpper plenumHigh pressure plenum
A loopB loopC loop
Flow
rate
(kg/
s)
Tem
pera
ture
(o C)
Time (sec)
62
University of FukuiStation Blackout at 1000 sec.
200
300
400
500
600
700
800
900
1000
0
100
200
300
400
500
0 1000 2000 3000 4000 5000
Inner core Ch.1 outletInner core Ch.2 outletInner core Ch.3 outletInner core Ch.4 outletInner core Ch.5 outlet
Inner core Ch.6 outletOuter core outletBlanket outletUpper plenumHigh pressure plenum
A loopB loopC loop
Flow
rate
(kg/
s)
Tem
pera
ture
(o C)
Time (sec)