University of Fukui

Safety of Nuclear Reactors

Professor H. MOCHIZUKINuclear Reactor Thermal Hydraulics DivisionNuclear Reactor Thermal Hydraulics Division

Research Institute of Nuclear Engineeringf

1

University of Fukui

Accident (1/2)University of Fukui

• Design Basis Accident: DBA• Assumption of simultaneous double ended break

• Installation of Engineered Safety FeaturesEmergency Core Cooling System: ECCSg y g yAccumulated Pressurized Coolant Injection System: APCIyLow Pressure Coolant Injection System: LPCIHigh Pressure Coolant Injection System: HPCIg j y

2

Accident (2/2)University of Fukui

C t d d t l t• Computer codes are used to evaluate temperature behavior of fuel bundle.

• Computer codes should be validated.• Blow-down and ECC injection tests haveBlow down and ECC injection tests have

been conducted using mock-ups.RELAP5/mod3 and TRAC code are• RELAP5/mod3 and TRAC code are developed and validated.

3

ECCSUniversity of Fukui

Control Rod Drive

Turbine

Containment Air Cooling SystemRod Drive Relief

valve

Feed Water System Sea water

Shield Cooling SystemDump valve

Residual Heat Removal System (RHR)

g yHigh Pressure Coolant Injection System (HPCI)

Low Pressure Coolant Injection System (LPCI)

ve

Containment SprayS t

(APCI)

Byp

ass

valv

Heavy Water CoolingSystem

Reactor AuxiliaryComponent CoolingW S

Reactor Core Isolation Cooling System(RCIC)

SystemCondensate Tank

4Main System Diagram of Fugen

Water SystemReactor AuxiliaryComponent CoolingSea-Water System

(RCIC)

Blow-down experimentUniversity of Fukui

5

６MW ATR Safety Experimental FacilityUniversity of Fukui

y p y

Outlet pipes(74mmID, 10.5mL, 2 deg.)( , , g )

P,T

P

TPressure transducerThermocouple

P T

EL 11.5EL 9.7EL 9.51

Downcomer(

EL: Elevation in mID: Inner diameterL : Length from a component

to the next arrow.P,T

P,T Steam drum(1525mmID,4.46mL)

High power

EL 5.97EL 7.0

EL 8.45Shield plug

(275.7mmID,6.8mL)

Low powerheaters

Main steamisolation valve

P,T

P,T

P TWater drum

High powerheater3.7mL, 6MW

EL 2.27

EL 4.5

3.7mL, 200kWP,T (387mmID,3.9mL)

(186mmID, 24.6mL)

EL 1.64

EL 2 3Inlet ppes(62.3mmID,12mL) EL0.9

Check valves

Turbine flow meter

Connecting pipe

EL 0.4

EL 2.3

6

Pump

Fig. 6 Schematic of Safety Experiment Loop (SEL)

Connecting pipe(186mmID, 15.6mL)

Water level behavior after a main steam pipe break

University of Fukuip p

Drum

0

0.2

12

15Drum water level

m)

vel (

m) D/P

-0.2 9

er le

vel (

m

wat

er lw

v

D i ill ti d t b k

Drum water level increase duringdowncomer water level decrease

0 6

-0.4

3

6Dowmcomer water level

um w

ate

ncom

er wDevice oscillation due to break

D/P

-0.8

-0.6

0

3

0 10 20 30 40 50

Dr

Dow

n Reactor core

0 10 20 30 40 50Time (sec)

100 mm break at main steam pipeBreak

7

Beak

Simulated fuel bundleUniversity of Fukui

Location of maximum axial peaking

Local peaking is high for the outer rods due to the

47.30

52.01

49.55

54.49

51.81

56.07

39.63

Unit in mm

Outer

Power of clusterat each zone(kW/m)

axial peakingneutronic characteristics

Tie rod14.5 OD

29.72

27.04

18.92

33.12 34.6936.26

25.23

36.04 Middle

InnerHeater pin14.5 OD

Spacer TieGadlinia pin

493 740 740 1234 493V IV III II I plate

Gadlinia pin14.5 OD

405 460 360 260 260 260 260 260 260 260 280 320 400

T

T T T TT T T T

1020 Active heated length : 3700 mm

Cross sectional view of 36-heater bundleBottom

Top

T Thermocouple position

8

Fig. 7 Power distribution of 36-rod high power heater

Thermocouple positionsUniversity of Fukui

p pH19-1

0°

Section 1

E8 1D34-1

F17 1

D36-4Section 4

D20-4

F34-4 B18-4

D8 4

A21-4C7-4

C19-4

A17 4

C35-4

0°

DT-1270° 90°

G22-1E8-1

H1-1

F17-1

A4-1

D14-1A31-1

A11-1C25-1

GT-4

G4-4

270° 90°

D16-4C1-4

H5-4

B22-4

B10-4

F14-4B30-4

A3-4 H24-4

H12-4

D8-4

H32-4G23-4D2-4

A9-4F6-4E33-4

A17-4

E25-4G11-4

E13-4

G31-4

C15-4

180° 180°0°

E28-1 D28-4

F26-4

B20-2

Section 2

D18-2H34-2

B36-2 D19-5Section 5D35-5

D7 5 B21 5

A19-2

A7-2 C21-2

A35-2

C27-4A29-4

A20-5

C18 5G34-5

A36-5

0°

E1-2

F12 2

270° 90°

D22-2

B2-2F32-2

B8-2

B16-2

E1-2

H14-2

D10-2F24-2

G6-2

90°HT-5270°

F33-5B17-5

D7-5 B21-5

B9-5F6-5

D15-5H31-5 H4-5 B3-5

H11-5

A2-5H23-5

A7 2 C21 2

A1-2C9-2

E23-2

C17-2G35-2

F5-2

C3-2

E11-2

A15-2E31-2

G13 2

C22-5D1-5

A8-5

E5-5E32-5A16-5

C18-5G34-5

E24-5

C10-5

G14-5

B19-3Section 3

B35-3

180°0°

F12-2D30-2

B28-2 H26-2

E20-6

0°

Section 6H35-6

180°

F13-5

B29-5F25-5

D27-5

G13-2

C29-2

G25-2

A27-2

C20-3C36-3

E12-5

G26-5A28-5

C30-5

FT-3270° 90°

D21-3

H33-3 H6-3

D17-3

G5-3

D9-3

F23-3C2-3

B15-3F31-3 D3-3

F11 3

B7-3

AT-6270° 90°G10-6

B32-6E16-6

F3-6

C6-6

H7-6

A23-6

E14 3

F4-3A10-3

G32-3

C16-3

A22-3B1-3

C8-3E34-3 A18-3

G24-3

9

H13-3

D29-3

F11-3H25-3

B27-3

Fig. 8　Thermocouple positions on high power heater rods

C13-6

F29-6 D26-6

E14-3

E26-3C28-3

G12-3A30-3

180° 180°

Cladding temperature measured in a same cross section of heater bundle

University of Fukui

500Data at section-II

300

400

500

｡

0

100

200

0 10 20 30 40 50 60Time (sec)

Fig. 14 Experimental cladding temperature for 150 mmdowncomer break

10

Calculation model of pipe break experimentUniversity of Fukui

p p p

Main steamisolationvalveMain steam pipe

Outlet pipesSteam drum

Upper shield

Powerdi t ib ti200kW

Max. 6 MW1 ch. Downcomer

Break

0.8

1.15

1.10

distribution200kW5 ch.

Waterdrum

Checkvalves

Pump

1.05

0.6

Inlet pipes

pLowerextension

AccumlatedPressure Coolant

11Fig. 9 Nodalization scheme for ATR Safety Experiment Loop

essu e Coo a tInjection system

Comparison between experimental result and simulation

University of Fukui

20

and simulation

10

15

20Calculatin

Experiment

-5

0

5

0 10 20 30 40 50 60Time (sec)

5

500

600Calculation Section-II

200

300

400

500

｡

Experiment

0

100

200

0 10 20 30 40 50 60

12

Time (sec)Fig.16 Behavior of cladding temperature after 100 mm downcomer break

Improvement of blow-down analysis by applying statistical method

University of Fukui

600

pp y gDowncomer 100 mm break

500

550 Calculation

Experiment

350

400

450

ture (℃

)

250

300

350

Temperat

150

200

250

ScramECCS operation

1000 5 10 15 20 25 30 35 40 45 50

Time (sec)

13

( )

Improvement of blow-down analysis by applying statistical method

University of Fukui

500

Calculation

Downcomer 150 mm break

pp y g

400

450 Experiment

300

350

ture

(℃)

250

300

Tem

pera

t

150

200

100

150

0 5 10 15 20 25 30 35 40 45 50

14

Time (sec)

Application of stochastic method to FBR analysis

University of Fukui

7 8 9

Air cooler[31]

[32]

[33][15]

[14]to turbine

[19]

13

[48]

14[37]

[38][20]

[36] A-loopB-loop15

y

[13]

[47]15

10-2

1

5

6

-1

[28]

[29]

[30]

Upper plenum

[17]

turbine

Pump

3

[18]

[46][48]

1816

-3

[22]1

16

14

Evaporator Super-heater

[1]～[7]

12

[8] IHXHigh-pressureplenum

[16]

[34][35]11

[10]

[11]

[12]

[9]

3

[42]

18

[39] [40]

17[21]

[41]to B-loopC-loop From B&C loops

４2 Low-pressure

plenum

Pump[24]

[23]

19

[49]

1922

-4

2021[ ]

[43][25]

[41]

to C-loop

C loop p

14 16Low-pressure turbine

Low-pressure side

16

p

19

24

[44] [45]

[27]

-1Deaerator

1 14

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: CRs

[44] [45]

23[26]

2 3020

#2#3#4#5

25

#1

1

13

1Feedwater

Condenser

15

Link 10: CRsLink 11: Bypass

Feedwaterpump

High-pressure feedwater heaters Low-pressure feedwater heaters

Application of stochastic method to FBRUniversity of Fukui

℃） Method：

Plant parameters are

Pe number: 1000

mpe

ratu

re （

Measured at Monju

Analysis

Plant parameters are investigated by 10,000 trials of the Monte-Carlo calculation for 43 factorsqu

ency

（-）

Tem

Time (sec)

calculation for 43 factors which can affect on the plenum temperature.

Freq

℃）

( )

Result：The measured temperature

Statistics if measured Nu number Evolution of plenum temperature during turbine trip

quen

cy（-）

erat

ure （℃ Measured at Monju

Analysis

The measured temperature transient has been included in the group of calculated curves Most non safety

Freq

Tem

p curves. Most non-safety side value could be evaluated taking into account various statistical

16

Cool-down velocity of plenum temperature (℃/s)

Time (sec)account various statistical errors.

[1] N. Kikuchi and H. Mochizuki, Application of statistical method for FBR plant transient computation, FR’13, Paris (2013-March)

Severe accidentUniversity of Fukui

17

Heat transfer of melted fuel to materialUniversity of Fukui

18

Heat transfer between melted jet and materials

University of Fukui

104

1000

10

Nu = 0 0033 Re Pr

100NaCl-Sn 10 1100NaCl-Sn 20 900NaCl-Sn 20 1000/P

r j

Material, Dj(mm), T

j(oC)

Num

= 0.0033 RejPr

j

1

10 NaCl-Sn 20 1100NaCl-Sn 30 900NaCl-Sn 30 1000NaCl-Sn 30 1100

Nu m

/

Num = 0.00123 Re

j Pr

j

0.1

1Al2O3-SS 10 2200Al2O3-SS 10 2300Al2O3-SS 10 2300

j j

103 104 105 106Re

jComparison of Nusselt number between present data and data fromSaito et al.1) and Mochizuki2).1)Saito et al Nuclear Engineering and Design 132 (1991)

19

1)Saito, et al., Nuclear Engineering and Design, 132 (1991)2)Mochizuki, Accident Management and Simulation Symposium, Jackson Hole, (1997).

Fuel melt experiment using BTF in CanadaUniversity of Fukui

20

Fuel melt experiment using CABRIUniversity of Fukui

p g

21

Source term analysis codesUniversity of Fukui

General codes

NRC codes ORIGEN-2, MARCH-2, MERGE, CORSOR, TRAP-MELT, CORCON, VANESA NAUA 4 SPARC ICEDFVANESA, NAUA-4, SPARC, ICEDF

IDCOR codes MAAP, FPRAT, RETAIN

NRC code (2nd Gen.) MELCOR( )Precise analysis

d

Core melt SCDAP, ELOCA, MELPROG, SIMMER

Debris-concrete reaction CORCONcodes

Hydrogen burning HECTOR, CSQ Sandia, HMS BURN

FP discharge FASTGRASS, VICTORIA

FP behavior in heat transport system

TRAP-MELT

FP discharge during debris- VANESAFP discharge during debris-concrete reaction

VANESA

FP behavior in containment CONTAIN, NAUA, QUICK, MAROS, CORRAL II

22

CORRAL-II

CONATIN codeUniversity of Fukui

(13) AirContainment spray In case of containment bypass

(11)(14)

Air

ypContainmentrecirculation

system

(10)

(12)

Stack

Annulus

system

(7)(8)

(9)(10) Stack

(5)(6)

(7)(8) Filter Blower

Water flow

(1)(2)(3)

(4)(5)( )

Gas flow

23

( )(2)

Steam release pool

Fluid- structure interaction analysis during hydrogen detonation

University of Fukuiy g

24

University of Fukui

Analysis of Chernobyl AccidentInvestigation of Root Cause- Investigation of Root Cause -

25

Schematic of Chernobyl NPPUniversity of Fukui

1. Core2. Fuel channels3. Outlet pipes4. Drum separator5 St h d

9. Inlet pipes10. Fuel failure detection equipment11 Top shield

Electrical power 1,000 MWThermal power 3,200 MWCoolant flow rate 37,500 t/hSteam flow rate 5,400 t/h (Turbine)5. Steam header

6. Downcomers7. MCP8. Distribution group headers

11. Top shield12. Side shield13. Bottom shield14. Spent fuel storage15. Fuel reload machine16 Crane

,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

16. Crane

26

Elevation PlanUniversity of Fukui

27

Above the Core of Ignarina NPPUniversity of Fukui

28

Core and Re-fueling MachineUniversity of Fukui

29

Control RoomUniversity of Fukui

30

Configuration of inlet valveUniversity of Fukui

1

2

3 41. Isolation and flow control valve2．Ball-type flow meter 3．Inlet pipe4 Distribution group header

31

4．Distribution group header

Drum SeparatorUniversity of Fukui

32

Configuration of Fuel Channnel

University of Fukui

S.S.

Position: -0.018m

-8.283 (-8.335)

Diffusion welding

Zr-2.5%Nb Roll region

Electron beem welding (EBW)

200

mm

-8.483

Effective

Spacers

-8.969 Zr-2.5%Nb

core regionFuel assembly

Connecting rod

-12.451

14 192

-15.933

80

-14.192

Diffusion welding

(-16.478) EBW -16.433

(-16.588)S.S.

72

77 -16.633

33

-16.671 Welding

Heat Removal by ModerationUniversity of Fukui

P t b G hit iPressure tube Graphite ring Maximum graphite temperature is 720℃at rated power

φ88mmφ114mmφ91mm Heat generated in graphite

blocks is removed by coolant

p

Graphiteblocks

φφ111mm

Gap of 1.5mmCoolant

34

RBMK & VVERUniversity of Fukui

Finland

Russia

Lith iLithuania

Germany

Ukraine

35

Objective of the ExperimentUniversity 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• 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• They tried to generate power by the inertia

of the turbine system.

36

Report in Dec. 1986University of Fukui

37

Trend of the Reactor PowerUniversity of Fukui

3500 Power excursion

2500

3000(M

W)

excursion

1500

2000

al P

ower

500

1000

Ther

ma

Scheduled power level for experiment20-30% of rated power

200MW0

00 00 00 00 00 00 04 40

30MW

sec

200MW

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 sec

minhourday

38

2 2 2 2 2 2 2 2 day

Time Chart Presented by USSRUniversity of Fukui

y

再循環流量Recirculation flowrate

給水流量 蒸気ドラム圧力出力Feedwater flowrate PowerDS pressure給水流量 蒸気ドラム圧力DS pressure

39

安全棒挿入開始Safety rod insertion

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 NuclearThermal 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

Water level

Drum pressure

Feed water

Water level

Neutron flux

40

Result in the Past Analysis (2/2)University of Fukui

Power at 48,000 MWTiming of peak

diff t

P j t b f th

was different. Why???

Power just before the accident was twice as large as the report. Why???

Power at 200 MW

Why???

？？？

Result of FATRAC

41

code is transferred, and initial steady calculation was conducted.

Possible Trigger of the AccidentUniversity of Fukui

• Positive scram due to flaw of scram rods• Pump cavitation• Pump coast down• Pump coast-down• Opening of turbine bypass valve

(6.96MPa)

42

Calculation Model by NETFLOW++ CodeUniversity of Fukui

y

7 MPaMain steam pipe

7 MPa

-1-21

EL:33.65-21

[11]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)

Fuel channelOD 0 088

Feed water pipe

Suction header

200 (3200) MWt[8]DowncomersOD: 0.325 mID: 0.295 mL: 23 - 33.5 mN: 12 (for each DS)

OD: 0.088 mID: 0.080mN: 1661ch. Outlet pipe

L: 12.7 - 23mOD: 0.076mID : 0.068m

8.2 m Flow control valve

EL:25.6m

Check valve

Suction header

EL:21.3EL:20.0

EL:14.852[10]

91

OD: 1.020mID: 0.9mL: 21m

Cooling pump :2 for one DSH: 200 m1000 rpm

Distribution group headersOD: 0.325mID: 0.295mL: 24mN: 16

1

EL: 12.15

Flow meterCheck valve

EL:11.8EL:11.6

[1]EL:5.9

EL:7.6

96

3

1000 rpm5500 kWGD2=1500 kg m25250 t/h ×2(8000 m3/h for one pump)

Pressure headerOD: 1.040mID: 0.9mL: 18.5m

EL: 9.60.080.088 95

EL: 0

Check valveTh ttli l ti l

[ ]

[2]

[9]

[3]

93

[4][5][6][7]

1121314151617192

OD: 0.828mID: 0.752mL= 36m

OD: 0.828mID 0 752

Feeder pipesL: 22.5 - 32.5 mOD: 0.057 mID : 0.050 mEL: 6.3

EL: 9.3

0.091

0.111

Graphite ring

0.02

94

43

Throttling-regulating valveL= 36m ID: 0.752mL= 34m

Graphite ring

Fuel channel

Trigger of the AccidentUniversity of Fukui

gg• Positive scram

P.S.W. Chan and A.R. DasterNuclear Science and EngineeringNuclear 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, P i F (1991)Paris, France, (1991).

44

Trigger of the Accident (cont.)University of Fukui

Scram rod (24rods)inserted by AZ-5 button

1.0Negative reactivity

8.0 5.0 Graphite block

1.5 Positive reactivityG hit di lGraphite displacer

Fuel 2×3.5mWater Column

45

Simulation from 1:19:00 to First PeakUniversity of Fukui

Data acquired by SKALA

10 400Flowrate (m3/s) DS water level (mm)F d t fl t (k / ) g/

s)

9

10

200

400( )P (MPa)Flowrate (calc.)Pressure (calc.)

Feed water flowrate (kg/s)Reactor power (calc.)DS water level (calc.)

MP

a)

flow

rate

(k

8 0

Pre

ssur

e (

edw

ater

f

7 -200

(m3 /s

), P

(mm

), Fe

6 -400

Flow

rate

ater

leve

l

Close stop valve:Turbine trip

5 -6000 60 120 180 240 300 DS

wa

Time (sec) Push AZ-5 button

Turbine trip

1:19:00

46

Trend of parameters for one loop from 1:19:00 on 26 April 1986

Behavior of Steam QualityUniversity of Fukui

0.05

0 03

0.04 TopCenter

ty, x

(-) Turbine trip

RCP trip

0.02

0.03

stea

m q

ualit

0

0.01

equi

libriu

m s

Two-phase

-0.01

Ther

mal

e

Water

-0.020 50 100 150 200 250 300

Tim e (sec) Push AZ-5 button

47

Void CharacteristicUniversity of Fukui

1

0.8-)

Void fraction

0.6Measuredon

, α (-

Pressure 7MPa 2Void fraction increase

0.4Correlation

id fracti

Void fraction

0.2

Vo

Void fraction increase

00 0.2 0.4 0.6 0.8 1

48

Thermal equilibrium steam quality, x (-)

Nuclear CharacteristicsUniversity of Fukui

Doppler Void

-1 10-5

-8 10-6

0.0004

0.0005

pp

-1.2 10-5

1 10

Δk/k/℃ 0.0003

k/k/%Void

-1.6 10-5

-1.4 10-5

Δ

0.0001

0.0002Δk

-1.8 10-5

0 500 1000 1500 2000 2500

T (℃)

00 20 40 60 80 100

Void fraction (%)

49

Peak Power and its ReactivityUniversity of Fukui

y

3.5 105

Power reported by USSR (MW)

3

T t l

5

2.5 105

3 105Power reported by USSR (MW)NETFLOW

)

1

2TotalScram (input)DopplerVoid

$)

5

1.5 105

2 105

Pow

er (M

W)

-1

0

Rea

ctiv

ity ($

0

5 104

1 105

-2Push AZ-5 button

0270 275 280 285 290

Time (sec)

-3270 275 280 285 290

Time (sec)1:23:30

50

Relationship between Peak Powerand Peak Positive Reactivity

University of Fukuiy

100

ers

full

pow

e

10

ak m

ultip

les

1pow

er p

ea

1

alue

of f

irst

0.10.75 0.8 0.85 0.9 0.95

Peak

va

51

Peak positive reactivity ($)

Just after the AccidentUniversity of Fukui

52

Control Room and Corium beneath the CoreUniversity of Fukui

53

University of Fukui

54

Before earthquake (14:46) & After Tsunami(15:45) at Fukushima-1 on 11 Mar. 2011

University of Fukui(15:45) at Fukushima 1 on 11 Mar. 2011

B f thBefore the earthquake

After the Tsunami

Area damaged by the Tsunami

Seawater pumps

Intake of seawater Heavy-oilpumps seawater Heavy-oil

tanks

55

Severe accident at Fukushima-1After Tsunami

University of FukuiAfter Tsunami

Operators injected water into the reactorcore by RCIC. After the loss-off-all-AC-

Fuel pool

core by RCIC. After the loss off all ACpower, water injection was impossible. Damaged

by Tsunami

DC powerOff-site powerContainment

vessel○

○

原子炉圧力容器

vessel

×Loss of off-site power

ECCS

○

Diesel generators

×RCIC pump

Stack

g

××

×pu p

56

Suppression poolSeawater pumps

Cooling pump for suppression pool

Hand calculation to estimate uncoveryUniversity of Fukui

y

• Assumption: Reactor diameter=4m, water level from top of f l 4fuel=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% f h t i l d t≒460,000/0.3 ≒ 1,500,000kW (70% of heat is released to seawater）Decay heat ratio at around 1000 sec ≒ ２%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

57

=20000kg=20t

Real water level and detected water levelUniversity of Fukui

Low pressure(Spent fuel pool)

High pressure (Core)Volume of steam ( p p )

Volume of steam expands 1600 times of water at 0.1 MPa.

expands 21 times of water at 7 MPa.

ExperimentExperiment

ＶＶ

Ｃｖ

Ｖ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,

58

燃料

, g g g ,144 (1993), pp293-303.

Natural circulation after the loss of AC power and sea water pump

University of Fukuip p

Secondary sodium

Primary T bi G t

Super heater(SH)

Primarysodium

Primary pump

Air cooler（AC)

Turbine Generator

Condenser

Secondary pump

Evaporator(EV)

Sea water cooler

IHXCore Feed water pump

59[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.

Analytical Model of Whole Heat Transport SystemsUniversity of Fukui

7 8 9

Air cooler[31]

[32]

[33][15]

Link 1-Link 6: 1st to 6th layer (Inner driver)Link 7: 7th & 8th layer (Outer driver)Link 8: 9th to 11th layer (Blanket)

1314[37]

[38][20]

[36] A-loopB-loop15

[13]

[47]15

10-2

16

-1

[29]

[30]

Upper plenum

[33]

[17]

[14]to turbine

Pump

Link 9: Center CRLink 10: CRsLink 11: Bypass

[18]

[19]

[46][48]

1816

-3

[ ][ ]

[22]1

16

14

Evaporator

Super-heater[1]～[7]

12

5

[8] IHX

[28]

[ ]

High- [16]

[34][35]11

[10]

[11]

[9]

3

[18]

18

[39] [40]17[21]

[22]

４[ ] [ ]

pressureplenum

Pump

[12]

[24]

192021[42]

[43][25]

[41]to B-loop

to C-loop

C-loop From B&C loops

14 16

４2

Low-pressure

Low-pressure side

Low-pressureplenum

[23][49]

1922

-4

[27]

turbine

1

16

14

24

[44] [45]23[26]

#2#3#4#5 #1

13-1Deaerat

or Condenser

60

2 3020 251 1Feedwater pump

High-pressure feedwater heaters Low-pressure feedwater heaters

Station Blackout Event of “Monju”University of Fukui

700 500Inner core Ch.1 outletInner core Ch.2 outletInner core Ch 3 outlet

Inner core Ch.6 outletOuter core outletBlanket outlet

500

600

300

400Inner core Ch.3 outletInner core Ch.4 outletInner core Ch.5 outlet

Blanket outletUpper plenumHigh pressure plenum

te (k

g/s)

atur

e (o C

)

700 500300

400

100

200

A loopB loop

Flow

rat

Tem

pera

600 400

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

s)o C)

200 00 1000 2000 3000 4000 5000

C loop

Time (sec)

400

500

200

300A loopB loopC loop

Flow

rate

(kg/

s

Tem

pera

ture

(o

200

300

0

100

0 1 2 3 4 5 6 7

61

0 1 2 3 4 5 6 7Time (day)

Station Blackout at 1000 sec.University of Fukui

1000 500Inner core Ch.1 outletInner core Ch 2 outlet

Inner core Ch.6 outletOuter core outlet

800

900

400

Inner core Ch.2 outletInner core Ch.3 outletInner core Ch.4 outletInner core Ch.5 outlet

Outer core outletBlanket outletUpper plenumHigh pressure plenum

600

700 300A loopB loopC loop at

e (k

g/s)

ratu

re (o C

)

500 200

C loop

Flow

ra

Tem

per

300

400100

200 00 1000 2000 3000 4000 5000

Time (sec)

62