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

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

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

14

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

36

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

45

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