Transcript 原子炉工学

University of Fukui
Safety of Nuclear Reactors
Professor H. MOCHIZUKI
Research Institute of Nuclear Engineering,
University of Fukui
1
Accident (1/2)
University of Fukui
• Design Basis Accident: DBA
• Assumption of simultaneous double ended break
• Installation of Engineered Safety Features
Emergency Core Cooling System: ECCS
Accumulated Pressurized Coolant Injection
System: APCI
Low Pressure Coolant Injection System: LPCI
High Pressure Coolant Injection System: HPCI
2
Accident (2/2)
University of Fukui
• 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.
3
ECCS
University of Fukui
Turbine
Control
Rod Drive
Containment Air Cooling System
Relief
valve
Sea water
Feed Water System
Residual Heat Removal System (RHR)
Shield Cooling System
Dump
valve
High Pressure Coolant Injection System (HPCI)
Heavy Water Cooling
System
Bypass valve
Low Pressure Coolant Injection System (LPCI)
(APCI)
Containment Spray
System
Condensate
Tank
Reactor Auxiliary
Component Cooling
Water System
Reactor Auxiliary
Component Cooling
Sea-Water System
Reactor Core Isolation Cooling System
(RCIC)
Main System Diagram of Fugen
4
Blow-down experiment
University of Fukui
5
６MW ATR Safety Experimental Facility
University of Fukui
Outlet pipes
(74mmID, 10.5mL, 2 deg.)
EL 9.7 P,T
EL 9.51
EL 11.5
P Pressure transducer
T Thermocouple
P,T
Steam drum
(1525mmID,
4.46mL)
EL 8.45
Shield plug
EL 7.0
EL 5.97
P,T
High power
heater
3.7mL, 6MW
Downcomer
(275.7mmID,
6.8mL)
P,T
P,T
Low power
heaters
3.7mL, 200kW
EL 2.27
P,T
EL 1.64
EL: Elevation in m
ID: Inner diameter
L : Length from a component
to the next arrow.
Main steam
isolation valve
Water drum
(387mmID,
3.9mL)
EL 4.5
(186mmID, 24.6mL)
Inlet ppes
(62.3mmID,
12mL) EL0.9
EL 2.3
Turbine flow meter
Check valves
EL 0.4
Connecting pipe
(186mmID, 15.6mL)
Fig. 6 Schematic of Safety Experiment Loop (SEL)
Pump
6
Water level behavior after a main steam
pipe break
University of Fukui
Drum water level (m)
Drum water level
12
0
Drum water level increase during
downcomer water level decrease
-0.2
9
Device oscillation due to break
6
-0.4
Dowmcomer water level
-0.6
3
-0.8
0
50
0
10
30
20
Time (sec)
40
Downcomer water lwvel (m)
15
0.2
Break
100 mm break at main steam pipe
7
University of Fukui
Simulated fuel bundle
Local peaking is high for the
outer rods due to the
neutronic characteristics
Unit in mm
Location of maximum
axial peaking
Power of cluster
at each zone
(kW/m)
52.01
47.30
Tie rod
14.5 OD
29.72
Heater pin
14.5 OD
49.55
34.69
56.07
51.81
39.63
36.26
Outer
Middle
36.04
27.04
Spacer
Gadlinia pin
14.5 OD
33.12
54.49
Inner
25.23
18.92
V
493
405
T T
1020
Bottom
IV
740
460
T
III
740
II
1234
Tie
plate
I
493
360 260 260 260 260 260 260 260 280 320
T
T
T
T
400
T
Top
Active heated length : 3700 mm
Cross sectional view of 36-heater bundle
T Thermocouple position
Fig. 7 Power distribution of 36-rod high power heater
8
Thermocouple positions
University of Fukui
9
Cladding temperature measured in a same cross
section
of
heater
bundle
University of Fukui
10
Calculation model of pipe break experiment
University of Fukui
Main steam pipe
Main steamisolation valve
Steam drum
Outlet pipes
Upper shield
Max. 6 MW
1 ch.
200kW
5 ch.
0.8
Downcomer
Power
distribution
1.15
1.10
Break
Check
valves
1.05
0.6
Water
drum
Pump
Lower
extension
Inlet pipes
Accumlated
Pressure Coolant
Injection system
Fig. 9 Nodalization scheme for ATR Safety Experiment Loop
11
Comparison between experimental result
and simulation
University of Fukui
12
Improvement of blow-down analysis by
applying statistical method
University of Fukui
Downcomer 100 mm break
Temperature (℃)
600
550
Calculation
500
Experiment
450
400
350
300
250
Scram
200
ECCS operation
150
100
0
5
10
15
20
25
30
35
40
45
50
Time (sec)
13
University of Fukui
Downcomer 150 mm break
500
Calculation
Experiment
450
Temperature (℃)
400
350
300
250
200
150
100
0
5
10
15
20
25
30
35
40
45
50
Time (sec)
14
University of Fukui
Severe accident
15
Heat transfer of melted fuel to material
University of Fukui
16
Heat transfer between melted jet and
materials
University of Fukui
10
4
1000
Nu = 0.0033 Re Pr
m
j
o
Material, D (mm), T ( C)
j
j
NaCl-Sn 10 1100
NaCl-Sn 20 900
NaCl-Sn 20 1000
NaCl-Sn 20 1100
NaCl-Sn 30 900
NaCl-Sn 30 1000
NaCl-Sn 30 1100
Al2O3-SS 10 2200
Al2O3-SS 10 2300
Al2O3-SS 10 2300
m
Nu /Pr
j
100
10
Nu = 0.00123 Re Pr
m
1
j
j
0.1
10
3
10
4
Re
10
j
5
6
10
j
Comparison of Nusselt number between present data and data from
Saito 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).
17
Fuel melt experiment using BTF in Canada
University of Fukui
18
Fuel melt experiment using CABRI
University of Fukui
19
Source term analysis codes
University of Fukui
General
codes
Precise
analysis
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
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- VANESA
concrete reaction
FP behavior in containment
CONTAIN, NAUA, QUICK, MAROS,
CORRAL-II
20
CONATIN code
University of Fukui
(13)
Containment spray
Air
In case of containment bypass
Containment
Air
(14)
recirculation
system
(11)
(12)
Annulus
(10)
(9)
(7)
(8)
(6)
(5)
(4)
(3)
Stack
(2)
Filter Blower
Water flow
Gas flow
(1)
Steam release pool
21
Fluid- structure interaction analysis during
hydrogen detonation
University of Fukui
22
University of Fukui
Analysis of Chernobyl Accident
- Investigation of Root Cause -
23
Schematic of Chernobyl NPP
University of Fukui
1. Core
2. Fuel channels
3. Outlet pipes
4. Drum separator
5. Steam header
6. Downcomers
7. MCP
8. Distribution
group headers
9. Inlet pipes
10. Fuel failure
detection
equipment
11. Top shield
12. Side shield
13. Bottom shield
14. Spent fuel
storage
15. Fuel reload
machine
16. Crane
Electrical power
1,000 MW
Thermal power
3,200 MW
Coolant flow rate 37,500 t/h
Steam flow rate
5,400 t/h (Turbine)
Steam flow rate
400 t/h (Reheater)
Pressure in DS
7 MPa
Inlet coolant temp.
270 0C
Outlet coolant temp. 284 0C
Fuel
1.8%UO2
Number of fuel channels
1,693
24
Elevation Plan
University of Fukui
25
Above the Core of Ignarina NPP
University of Fukui
26
Core and Re-fueling Machine
University of Fukui
27
Control Room
University of Fukui
28
Configuration of inlet valve
University of Fukui
1
2
3
4
1. Isolation and flow control valve
2．Ball-type flow meter
3．Inlet pipe
4．Distribution group header
29
Drum Separator
University of Fukui
30
Configuration of Fuel Channnel
University of Fukui
Position: -0.018m
-8.283
S.S.
(-8.335)
200 mm
Diffusion welding
Zr-2.5%Nb
Electron beem
welding (EBW)
Roll region
(Width
:
-8.483
50mm)
Zr-2.5%Nb
-8.969
Spacers
Effective
core region
Fuel assembly
-12.451
Connecting rod
-14.192
-15.933
80
EBW
Diffusion welding
S.S.
-16.433
72
77
(-16.478)
(-16.588)
-16.633
Welding
-16.671
31
Heat Removal by Moderation
University of Fukui
Pressure tube
φ91mm
φ114mm
φ88mm
φ111mm
Graphite ring
Maximum graphite
temperature is 720℃
at rated power
Heat generated in graphite
blocks is removed by coolant
Graphite
blocks
Coolant
Gap of 1.5mm
32
RBMK & VVER
University of Fukui
Finland
Russia
Lithuania
Germany
Ukraine
33
Objective of the Experiment
University of Fukui
• 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.
34
Report in Dec. 1986
University of Fukui
35
Trend of the Reactor Power
University of Fukui
Power
excursion
3000
2500
2000
1500
1000
Scheduled power level for experiment
500
:0
0
20-30% of
rated power
200MW
30MW
25 :00
:0 :00
1
25 :00
:1 :00
3:
25 05
:2 :00
3:
26 10
:0 :00
0:
26 28
:0 :00
1:
26 00
:0 :00
1:
26 23
:0 :04
1:
23
:4
0
0
25
Thermal Power (MW)
3500
sec
min
hour
day
36
Time Chart Presented by USSR
University of Fukui
37
Result in the Past Analysis (1/2)
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
•
Recirculation flow rate
Drum pressure
Water level
Feed water
Neutron flux
38
Result in the Past Analysis (2/2)
University of Fukui
Power at 48,000 MW
Timing of peak
was different.
Why???
Power just before the
accident was twice as
large as the report.
Why???
Power at 200 MW
？？？
Result of FATRAC
code is transferred, and initial steady calculation was conducted.
39
Possible Trigger of the Accident
University of Fukui
• Positive scram due to flaw of scram rods
• Pump cavitation
• Pump coast-down
40
Calculation Model by NETFLOW++ Code
University of Fukui
Main steam pipe
7 MPa
200 (3200) MWt
Drum separators
2 for one loop
L: 30m
ID: 2.6m
t: 0.105 m
Fuel channel
OD: 0.088 m
ID: 0.080m
N: 1661ch.
[11]
EL:33.65
[8]Downcomers
OD: 0.325 m
ID: 0.295 m
L: 23 - 33.5 m
N: 12 (for each DS)
EL:25.6m
Distribution group headers
OD: 0.325m
EL: 12.15
ID: 0.295m
[10]
L: 24m
1
N: 16
Flow control valve
8.2 m
-1
-21
Outlet pipe
L: 12.7 - 23m
OD: 0.076m
ID : 0.068m
EL:21.3
EL:20.0
Feed water pipe
Feed water for one DS
Wf =115 t/h, 140 oC
(1453 t/h, 177-190 oC)
Suction header
OD: 1.020m
ID: 0.9m
91
L: 21m
2
EL:14.85
3
EL:7.6
Cooling pump :2 for one DS
H: 200 m
1000 rpm
5500 kW
GD2=1500 kg m2
5250 t/h ×2
(8000 m3/h for one pump)
Check valve
96
Pressure header
OD: 1.040m
ID: 0.9m
L: 18.5m
Flow meter
EL:11.8
EL:11.6
EL: 9.6
0.08
0.088
0.091
EL:5.9
95
[1]
71 61 51 41 31 21 11
[9]
92
0.02
0.111
[7] [6] [5] [4] [3] [2]
EL: 6.3
Graphite ring
Fuel channel
Feeder pipes
L: 22.5 - 32.5 m
OD: 0.057 m
ID : 0.050 m
93
EL: 9.3
OD: 0.828m
ID: 0.752m
L= 36m
EL: 0
94
OD: 0.828m
ID: 0.752m
L= 34m
Check valve
Throttling-regulating valve
41
Trigger of the Accident
University of Fukui
• Positive scram
P.S.W. Chan and A.R. Daster
Nuclear Science and Engineering,
103, 289-293 (1989).
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).
42
Trigger of the Accident (cont.)
University of Fukui
Scram rod (24rods)
inserted by AZ-5 button
1.0
8.0
5.0
1.5
Fuel 2×3.5m
Negative reactivity
Graphite block
Positive reactivity
Graphite displacer
Water Column
43
Simulation from 1:19:00 to First Peak
University of Fukui
DS water level (mm)
Feed water flowrate (kg/s)
Reactor power (calc.)
DS water level (calc.)
3
Flowrate (m3/s), Pressure (MPa)
10
9
Flowrate (m /s)
P (MPa)
Flowrate (calc.)
Pressure (calc.)
400
200
8
0
7
-200
6
-400
Close stop valve:
Turbine trip
5
0
1:19:00
60
120
-600
180
240
300
Push AZ-5 button
Time (sec)
Trend of parameters for one loop from 1:19:00 on 26 April 1986
DS water level (mm), Feedwater flowrate(kg/s)
Data acquired by SKALA
44
Behavior of Steam Quality
University of Fukui
Thermal equilibrium steam quality, x (-)
0.05
0.04
Turbine trip
RCP trip
Top
Center
0.03
0.02
Two-phase
0.01
0
Water
-0.01
-0.02
0
50
100
150
Time (sec)
200
250
300
Push AZ-5 button
45
Void Characteristic
University of Fukui
1
Void fraction, α (-)
0.8
Da 2Void fraction
increase
0.6
Pressure 7MPa
Measured
Correlation
0.4
Da1 Void fraction
increase
0.2
0
0
0.2
0.4
0.6
0.8
Thermal equilibrium steam quality, x (-)
1
46
Nuclear Characteristics
University of Fukui
Void
-8 10-6
0.0005
-1 10-5
0.0004
Δk/k/%Void
Δk/k/℃
Doppler
-1.2 10-5
-1.4 10-5
0.0003
0.0002
0.0001
-1.6 10-5
0
-1.8 10-5
0
500
1000
1500
T (℃)
2000
2500
0
20
40
60
Void fraction (%)
80
100
47
Peak Power and its Reactivity
University of Fukui
5
3.5 10
5
3 10
3
Power reported by USSR (MW)
NETFLOW
Total
Scram (input)
Doppler
Void
2
5
2.5 10
Reactivity ($)
Power (MW)
1
5
2 10
5
1.5 10
0
-1
5
1 10
4
-2
5 10
0
270
Push AZ-5 button
275
280
Time (sec)
285
290
-3
270
1:23:30
275
280
285
290
Time (sec)
48
Relationship between Peak Power
and Peak Positive Reactivity
Peak value of first power peak multiples full power
University of Fukui
100
10
1
0.1
0.75
0.8
0.85
0.9
0.95
Peak positive reactivity ($)
49
Just after the Accident
University of Fukui
50
Control Room and Corium beneath the Core
University of Fukui
51