2009/12/17

1

University of Fukui

Safety Researches

1

Professor H. MOCHIZUKIResearch Institute of Nuclear Engineering,

University of Fukui

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: LPCI

2

Low Pressure Coolant Injection System: LPCIHigh Pressure Coolant Injection System: HPCI

2009/12/17

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University of Fukui

Accident (2/2)

• Computer codes are used to evaluate temperature behavior of fuel bundletemperature 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

3

developed and validated.

University of FukuiECCS

Containment Air Cooling System

Control Rod Drive Relief

valve

Turbine

Feed Water System Sea water

Shield Cooling SystemDump valve

Residual Heat Removal System (RHR)

High Pressure Coolant Injection System (HPCI)

Low Pressure Coolant Injection System (LPCI)

(APCI)

s va

lve

4Main System Diagram of Fugen

Reactor AuxiliaryComponent CoolingWater System

Reactor AuxiliaryComponent CoolingSea-Water System

Reactor Core Isolation Cooling System(RCIC)

Containment SpraySystem

Bypa

ss

Heavy Water CoolingSystem

Condensate Tank

2009/12/17

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University of Fukui

Blow-down experiment

5

University of Fukui６MW ATR Safety Experimental Facility

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

P,T

P Pressure transducer

EL 11.5EL 9.7EL 9.51

Downcomer(275.7mmID,6.8mL)

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

to the next arrow.

T Thermocouple

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)EL 1.64

High powerheater3.7mL, 6MW

EL 2.27

EL 5.97EL 7.0

EL 8.45Shield plug

6

EL 4.5

Pump

Fig. 6 Schematic of Safety Experiment Loop (SEL)

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 2.3

2009/12/17

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University of Fukui

Water level behavior after a main steam pipe break

0.2 15Drum water level

(m)

-0.6

-0.4

-0.2

0

3

6

9

12

Dowmcomer water level

Dru

m w

ater

leve

l (m

)

ownc

omer

wat

er lw

vel (

Device oscillation due to break

Drum water level increase duringdowncomer water level decrease

7

-0.8 00 10 20 30 40 50

Do

Time (sec)

100 mm break at main steam pipeBreak

University of FukuiSimulated fuel bundle

52.0154.49

51 81

56.07Unit in mm

Power of clusterat each zone(kW/m)

Tie rodφ14.5 OD

29.72

27.04

18.92

33.12

47.30

34.69

49.55

36.26

51.81

25.23

36.04

39.63

493 740 740 1234 493

OuterMiddle

InnerHeater pinφ14.5 OD

V IV III II ISpacer Tie

plateGadlinia pinφ14.5 OD

8

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

T T T TT T T T

T Thermocouple position

2009/12/17

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University of FukuiThermocouple positions

H19-1

DT-1270°

0°

90°

Section 1

G22-1E8-1

H1-1

D34-1F17-1

A4-1

D36-4

GT-4

G4-4

270° 90°

Section 4D20-4

F34-4

D16-4C1-4

B18-4

H5-4

B22-4

B10-4A3-4 H24 4

D8-4

H32-4G23-4D2-4

A9-4

A21-4C7-4

C19-4

F6-4E33-4A17-4

C35-4

G31 4

C15-4

0°

180° 180°0°

A4 1

D14-1A31-1

E28-1

A11-1C25-1

G4 4

F14-4B30-4

D28-4

A3 4

F26-4

H24-4

H12-4

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

90°

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

C27-4

E25-4G11-4

A29-4

E13-4

G31-4

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

0°

9

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°

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

180°

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

8 5

180° 180°

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 60

｡

10

Time (sec)Fig. 14 Experimental cladding temperature for 150 mmdowncomer break

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University of FukuiCalculation model of pipe break experiment

Outlet pipes

Main steamisolationvalveMain steam pipe

Steam drum

Checkvalves

Break

Upper shield

0.8

1.15

1.05

1.10

Powerdistribution200kW

5 ch.

Max. 6 MW1 ch. Downcomer

11

Waterdrum

Inlet pipes

Fig. 9 Nodalization scheme for ATR Safety Experiment Loop

PumpLowerextension

0.6

AccumlatedPressure CoolantInjection system

University of Fukui

15

20Calculatin

Experiment

Comparison between experimental result and simulation

-5

0

5

10

0 10 20 30 40 50 60Time (sec)

400

500

600

｡

CalculationExperiment

Section-II

12

0

100

200

300

400

0 10 20 30 40 50 60Time (sec)

Fig.16 Behavior of cladding temperature after 100 mm downcomer break

2009/12/17

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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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2009/12/17

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University of Fukui

Heat transfer between melted jet and materials

1000

104

1

10

100

1000

NaCl-Sn 10 1100NaCl-Sn 20 900NaCl-Sn 20 1000NaCl-Sn 20 1100NaCl-Sn 30 900NaCl-Sn 30 1000NaCl-Sn 30 1100Al2O3-SS 10 2200

Nu m

/Pr j

Material, Dj(mm), T

j(oC)

Num = 0.0033 Re

j Pr

j

Num = 0.00123 Re

j Pr

j

15

0.1

103 104 105 106

Al2O3-SS 10 2300Al2O3-SS 10 2300

Rej

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).

University of Fukui

Fuel melt experiment using BTF in Canada

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2009/12/17

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University of FukuiFuel melt experiment using CABRI

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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 TRAP-MELT

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transport systemFP discharge during debris-concrete reaction

VANESA

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

2009/12/17

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University of Fukui

CONATIN code

(11)(14)

(13)

Air

AirContainment spray In case of containment bypass

Containmentrecirculation

(7)(8)

(9)(10)

(11)

(12)

Filter Blower

Stack

Annulus

recirculationsystem

19

(1)(2)(3)

(4)(5)(6)

Steam release pool

Water flowGas flow

University of Fukui

Fluid- structure interaction analysis during hydrogen detonation

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2009/12/17

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University of Fukui

Analysis of Chernobyl Accidenty y- Investigation of Root Cause -

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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 storage

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 0C

15. Fuel reload machine16. Crane

Fuel 1.8%UO2Number of fuel channels 1,693

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2009/12/17

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University of Fukui

Elevation Plan

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University of Fukui

Above the Core of Ignarina NPP

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2009/12/17

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University of Fukui

Core and Re-fueling Machine

25

University of Fukui

Control Room

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2009/12/17

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University of Fukui

Configuration of inlet valve

1

2

27

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

University of Fukui

Drum Separator

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2009/12/17

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University of Fukui

Configuration of Fuel Channnel

S.S.

Diffusion welding

Zr-2.5%Nb Roll region

Electron beem ( )

200

mm

Position: -0.018m

-8.283

-8.483

(-8.335)

Effective core region

welding (EBW)

Fuel assembly

Spacers

Connecting rod

-8.969 Zr-2.5%Nb

-12.451

-14.192

29

Diffusion welding

-15.933

(-16.478)

-16.671

EBW -16.433

(-16.588)S.S.

φ80

φ72

φ77 -16.633

Welding

University of Fukui

Heat Removal by Moderation

Pressure tube Graphite ring Maximum graphite temperature is 720℃

Graphiteblocks

φ88mmφ114mm

φ111mm

φ91mm Heat generated in graphite blocks is removed by coolant

at rated power

30

Gap of 1.5mmCoolant

2009/12/17

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University of Fukui

RBMK & VVER

Fi l d

Russia

Lithuania

Finland

Germany

Uk i

31

Ukraine

University of Fukui

Objective of the Experiment

• Power generation after the reactor scram f l t f d i d tfor 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

32

• But they closed the isolation valve.• They tried to generate power by the inertia

of the turbine system.

2009/12/17

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University of Fukui

Report in Dec. 1986

33

University of Fukui

2500

3000

3500

Trend of the Reactor Power

W)

Power excursion

0

500

1000

1500

2000

2500

Ther

mal

Pow

er (M

W

Scheduled power level for experiment

30MW

20-30% of rated power

200MW

34

0

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

30MW

secminhourday

2009/12/17

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University of FukuiTime Chart Presented by USSR

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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

Result in the Past Analysis (1/2)

Water level

Drum pressure

Recirculation flow rate

36

Feed water

Neutron flux

2009/12/17

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University of Fukui

Result in the Past Analysis (2/2)

Power at 48,000 MW

Ti i f k

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

Timing of peak was different. Why???

37

Power at 200 MW

Result of FATRACcode is transferred, and initial steady calculation was conducted.

？？？

University of Fukui

Possible Trigger of the Accident

• Positive scram due to flaw of scram rods• Positive scram due to flaw of scram rods• Pump cavitation• Pump coast-down

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2009/12/17

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University of FukuiCalculation Model by NETFLOW++ Code

7 MPa

[11]Drum separators2 for one loopL: 30mID: 2 6m

Feed water for one DSWf =115 t/h, 140 oC(1453 t/h, 177-190 oC)

Feed water pipe

Main steam pipe

8.2 m Flow control valve

Flow meter

EL:25.6m

Check valve

Suction header

EL:21.3EL:20.0

EL:11.8

-1-21

EL:33.65

EL:14.85

200 (3200) MWt

2

-21

[10]

EL 7 6

91

96

3

ID: 2.6mt: 0.105 m

( )

[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 m2Pressure header

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

Fuel channelOD: 0.088 mID: 0.080mN: 1661ch. Outlet pipe

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

1

EL: 12.15

39

EL: 0

Check valveThrottling-regulating valve

EL:11.6

[1]

[2]

EL:5.9

[9]

EL:7.6

[3]

93

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

11213141516171

3

92

5250 t/h×2(8000 m3/h for one pump)

OD: 1.040mID: 0.9mL: 18.5m

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

EL: 9.3

EL: 9.60.08

0.091

0.088

0.111

Graphite ring

Fuel channel

0.02

94

95

University of FukuiTrigger of the Accident

P S W Chan and A R Daster

• Positive scram

P.S.W. Chan and A.R. DasterNuclear 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).

40

2009/12/17

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University of Fukui

Trigger of the Accident (cont.)

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

8.0

1.0

5.0

1 5

Negative reactivity

inserted by AZ 5 button

Graphite block

41

1.5 Positive reactivityGraphite displacer

Fuel 2×3.5mWater Column

University of Fukui

Simulation from 1:19:00 to First PeakData acquired by SKALA

10 400Flowrate (m3/s)P (MPa)Flowrate (calc.)Pressure (calc )

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

a) wra

te(k

g/s)

7

8

9

-200

0

200Pressure (calc.) DS water level (calc.)

ate

(m3 /s

), P

ress

ure

(MP

vel (

mm

), Fe

edw

ater

flow

42

5

6

-600

-400

0 60 120 180 240 300

Flow

r

DS

wat

er le

v

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

2009/12/17

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University of Fukui

Behavior of Steam Quality

0.04

0.05

TopCenter

(-) Turbine trip

0

0.01

0.02

0.03

Center

mal

equ

ilibriu

m s

team

qua

lity,

x

RCP trip

Water

Two-phase

43

-0.02

-0.01

0 50 100 150 200 250 300 Tim e (sec)

Ther

m

Push AZ-5 button

University of Fukui

Void Characteristic

0.8

1

0.4

0.6Measured

Correlation

Void fraction, α (-)

Pressure 7MPa

44

0

0.2

0 0.2 0.4 0.6 0.8 1Thermal equilibrium steam quality, x (-)

2009/12/17

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University of FukuiNuclear Characteristics

-8 10-6 0.0005

Doppler Void

-1.6 10-5

-1.4 10-5

-1.2 10-5

-1 10-5

Δk/k/℃

0.0001

0.0002

0.0003

0.0004

Δk/k/%Void

45

-1.8 10-5

0 500 1000 1500 2000 2500

T (℃)

00 20 40 60 80 100

Void fraction (%)

University of FukuiPeak Power and its Reactivity

3.5 105 3

1 105

1.5 105

2 105

2.5 105

3 105

3.5 10

Power reported by USSR (MW)NETFLOW

Pow

er (M

W)

-1

0

1

2

3

TotalScram (input)DopplerVoid

Rea

ctiv

ity ($

)

46

0

5 104

270 275 280 285 290Time (sec)

-3

-2

270 275 280 285 290

Time (sec)1:23:30

Push AZ-5 button

2009/12/17

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University of FukuiRelationship between Peak Power

and Peak Positive Reactivity100

ll po

wer

1

10

first

pow

er p

eak

mul

tiple

s fu

47

0.10.75 0.8 0.85 0.9 0.95

Peak positive reactivity ($)

Peak

val

ue o

f f

University of FukuiJust after the Accident

48

2009/12/17

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University of Fukui

Control Room and Corium beneath the Core

49