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

5

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

7

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°

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

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

12

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

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)

Temperature (℃

)

Improvement of blow-down analysis by applying statistical method

Scram ECCS operation

Downcomer 100 mm break

14

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.

16

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

17

University of FukuiSevere accident

18

University of FukuiHeat transfer of melted fuel to material

19

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

23

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).

45

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

47

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.

60

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)