Post on 28-Dec-2015
1
University of Fukui
Safety of Nuclear Reactors
Professor H. MOCHIZUKIResearch Institute of Nuclear Engineering,
University 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: APCI Low Pressure Coolant Injection System: LPCI High 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
Byp
ass
valv
e
Heavy Water CoolingSystem
Turbine
Feed Water System
Condensate Tank
Sea water
5
University of Fukui
Blow-down experiment
6
University of Fukui
6 MW ATR Safety Experimental Facility
D o w nc o m e r(2 75 .7m m ID ,6 .8 m L )
E L 4 .5
P u m p
O u tle t p ip e s(7 4 m m ID , 1 0 .5 m L , 2 d e g . )
Fig . 6 S ch em atic of S a fety E xp erim en t L o op (S E L )
P ,T
E L : E le v a tio n in mID : Inn e r d i a m e te rL : L e n g th f ro m a c om po n e n t
to th e n e x t a r row .
P
T
P re s su re t ra ns d u c e r
T h e rm o c ou p le
L o w p o w e rh e a te r s
3 .7m L , 2 0 0k W
M a in s te a mis o la tio n v a lv e
P ,T
P , T
P ,T
P ,T
P , TS te a m dru m(1 5 25 m m ID ,4 .4 6 m L )
W a te r d ru m(3 87 m m ID ,3 .9 m L )
In le t p p e s(62 .3m m ID ,1 2m L ) E L 0 .9
C h e c k v a l ve s
T u rb i n e f lo w m e te r
(1 8 6 m m ID , 2 4 .6 m L )
C o n n e c ti ng p ip e(1 8 6 m m ID , 1 5 .6 m L )
E L 0 .4
E L 1 1 .5E L 9 .7E L 9 .51
E L 1 .64
H igh p o w e rh e a te r3 .7 m L , 6 M W
E L 2 .2 7
E L 5 .9 7E L 7 .0
E L 8 .4 5S h ie ld p lu g
E L 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
ate
r le
ve
l (m
)
Do
wn
co
me
r w
ate
r lw
vel
(m)
Time (sec)
100 mm break at main steam pipe
Device oscillation due to break
Drum water level increase duringdowncomer water level decrease
Break
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
B15-3
F31-3
D29-3
D3-3
F11-3
H25-3
B27-3
B7-3
180°
0°
H19-1
DT-1270°
0°
90°
Section 1
G22-1E8-1
H1-1
D34-1
F17-1
A4-1
D14-1A31-1
E28-1
A11-1
C25-1
D36-4
GT-4
G4-4
270° 90°
Section 4
D20-4
F34-4
D16-4C1-4
B18-4
H5-4
B22-4
B10-4
F14-4
B30-4
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
H14-2
D30-2
B28-2 H26-2
D10-2
F24-2
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 5
D35-5
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
C17-2
G35-2
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
C8-3
E34-3 A18-3
C20-3C36-3
E26-3
G24-3
C28-3
G12-3A30-3
G23-4D2-4A9-4
A21-4C7-4
C19-4
F6-4E33-4
A17-4
C35-4
C27-4
E25-4
G11-4
A29-4
E13-4
G31-4
C15-4
0°
C22-5D1-5
A8-5
A20-5
E5-5E32-5
A16-5
C18-5G34-5
A36-5
E12-5
G26-5
E24-5
C10-5
A28-5
G14-5C30-5
180° 180°
10
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
W aterdrum
C heckvalves
Inlet pipes
Outlet p ipes
Main steam isolation valveM ain steam pipe
Steam drum
Fig. 9 No dalization scheme for ATR Safety Ex periment L oop
Break
Pum p
U pper shield
Low erextension
0.8
1.15
1.05
1.1 0
0.6
Powerdistribution2 00kW
5 ch.
Max. 6 MW1 ch.
AccumlatedPressure Coola ntInjection system
D ow ncom er
12
University of Fukui
-5
0
5
10
15
20
0 10 20 30 40 50 60
CalculatinExperiment
Time (sec)
0
100
200
300
400
500
600
0 10 20 30 40 50 60Time (sec)
。
1
2
3
4
5
6
7
8
9
10
11
12
13
14
15
16
17
18
19
20
21
22
23
24
25
26
27
28
29
30
31
32
33
34
35
36
37
38
39
40
41
42
43
44
45
46
47
48
50
26
TR-F32-2TR-C20-3TR-G32-3
CalculationExperiment
Fig.16 Behavior of cladding temperature after 100 mm downcomer break
Section-II
Comparison between experimental result and simulation
13
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)
Tem
pera
ture
(℃
)
Improvement of blow-down analysis by applying statistical method
ScramECCS 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
Calculation
Experiment
Time (sec)
Tem
pera
ture
(℃
)
Downcomer 150 mm break
15
University of FukuiSevere accident
16
University of FukuiHeat transfer of melted fuel to material
17
University of Fukui
Heat transfer between melted jet and materials
0.1
1
10
100
1000
104
103 104 105 106
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
Rej
Nu
m/P
r 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
18
University of Fukui
Fuel melt experiment using BTF in Canada
19
University of FukuiFuel melt experiment using CABRI
20
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
21
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 bypass
Containmentrecirculation
system
Steam release pool
Water flowGas flow
22
University of Fukui
Fluid- structure interaction analysis during hydrogen detonation
23
University of Fukui
Analysis of Chernobyl Accident - Investigation of Root Cause -
24
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
25
University of Fukui
Elevation Plan
26
University of Fukui
Above the Core of Ignarina NPP
27
University of Fukui
Core and Re-fueling Machine
28
University of Fukui
Control Room
29
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
30
University of Fukui
Drum Separator
31
University of Fukui
Configuration of Fuel Channnel
Effective core region
S.S.
Diffusion welding
Zr-2.5%Nb Roll region
(Width :
50mm)
Electron beem welding (EBW)
Fuel assembly
Spacers
Connecting rod
200
mm
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
32
University of Fukui
Heat Removal by Moderation
Pressure tube
Graphite blocks
Graphite ring
Gap of 1.5mm
Coolant
φ88mmφ114mm
φ111mm
φ91mmHeat generated in graphite blocks is removed by coolant
Maximum graphite temperature is 720 ℃at rated power
33
University of Fukui
RBMK & VVER
Russia
Lithuania
Finland
Germany
Ukraine
34
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.
35
University of Fukui
Report in Dec. 1986
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University of Fukui
0
500
1000
1500
2000
2500
3000
3500
25:0
0:00
:00
25:0
1:00
:00
25:1
3:05
:00
25:2
3:10
:00
26:0
0:28
:00
26:0
1:00
:00
26:0
1:23
:04
26:0
1:23
:40
Trend of the Reactor PowerT
herm
al P
ower
(M
W)
Scheduled power level for experiment
30MW
Power excursion
secminhourday
20-30% of rated power
200MW
37
University of FukuiTime Chart Presented by USSR
38
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)
39
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???
???
40
University of Fukui
Possible Trigger of the Accident
• Positive scram due to flaw of scram rods
• Pump cavitation
• Pump coast-down
41
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 DS
Wf =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
42
University of Fukui
Trigger 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).
43
University of Fukui
Trigger of the Accident (cont.)
8.0
1.0
5.0
1.5
Negative reactivity
Positive reactivity
Graphite displacer
Scram rod (24rods)inserted by AZ-5 button
Fuel 2×3.5m
Graphite block
Water Column
44
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.)
Flo
wra
te (
m3/s
), P
ress
ure
(M
Pa
)
DS
wat
er le
vel (
mm
), F
eed
wa
ter
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 1986
1:19:00
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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
Time (sec)
Th
erm
al e
qu
ilibri
um
ste
am
qua
lity, x (
-)
Turbine tripRCP trip
Push AZ-5 button
Water
Two-phase
46
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
47
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 100Δk/k/%Void
Void fraction (%)
Doppler Void
48
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
Po
wer
(M
W)
Time (sec)
-3
-2
-1
0
1
2
3
270 275 280 285 290
TotalScram (input)DopplerVoid
Rea
ctiv
ity (
$)
Time (sec)1:23:30
Push AZ-5 button
49
University of FukuiRelationship between Peak Power
and Peak Positive Reactivity
0.1
1
10
100
0.75 0.8 0.85 0.9 0.95
Peak positive reactivity ($)
Pe
ak v
alue
of f
irst p
ower
pea
k m
ulti
ples
full
pow
er
50
University of FukuiJust after the Accident
51
University of Fukui
Control Room and Corium beneath the Core