1

Water-Cooled Thorium Breeder Reactors

The 3rd Working Group on Thorium Fuel Utilization

in Light Water and Fast Reactors

WG

Sidik Permana

Nuclear Nonproliferation Science and technology Center (NPSTC)

Japan Atomic Energy Agency (JAEA)

1

2

Possible trends for the contribution of different sources to

total energy supply in the next centuries

Reference : Presented paper,

Contribution of Coal and

Nuclear to Sustainable Energy

Supply: Perspectives and

Problems, President’s meeting of

G8 countries, Brazil, China,

India and South Africa, Moscow

2006

Estimated Energy Sources

2

WG

Sidik Permana,

3

Estimated Energy SourcesWG

Sidik Permana,

4

Estimated Energy SourcesWG

Sidik Permana,

5

Reasonably Assured Reserves (RAR) and Estimated Additional Reserves (EAR) of thorium comes from

OECD/NEA, Nuclear Energy, "Trends in Nuclear Fuel Cycle", Paris, France (2001):

Country RAR Th (tonnes) EAR Th (tonnes)

Brazil 606,000 700,000

Turkey 380,000 500,000

India 319,000 —

United States 137,000 295,000

Norway 132,000 132,000

Greenland 54,000 32,000

Canada 45,000 128,000

Australia 19,000 —

South Africa 18,000 —

Egypt 15,000 309,000

Other Countries 505,000 —

World Total 2,230,000 2,130,000

Resources of Nuclear FuelWG

Sidik Permana,

6

Neutron regeneration ratio of each nuclide as a function of neutron energy

Refference : L. Michael and G. Otto, 1998

233U

239Pu

Possible Breeding of Each

Fissile Material

6

WG

Sidik Permana,

7

235U 236U 237U 238U

239Np

238Pu 239Pu 240Pu 241Pu 242Pu

241Am

242mAm

243Am

242gAm

242Cm 243Cm 244Cm 245Cm 246Cm

237Np

236Pu

236gNp

232U 233U 234U

231Pa

230Th 232Th

233Pa

IT

-decay

EC

(n, ) (n,2n)

decay

Th

Pa

U

Np

Pu

Am

Cm

Th-U Cycle

U-Pu Cycle

Nuclide Chain Mechanism

7

WG

Sidik Permana,

8

Water-Cooled Thorium Breeder Reactors

Content of Presentation

8

1. MOX fuel behavior on Water coolant reactors

2. Comparative analysis on physical properties of

water coolant reactor for different fuel

3. Feasibility analysis on water-cooled breeder

reactor

4. Feasibility analysis on water-cooled breeder

reactor with MA doping as supply fuel

5. Core design analysis on water-cooled breeder

reactor

WG

Sidik Permana,

9

Water-Cooled Thorium Breeder Reactors

MOX fuel properties of Water Cooled

Reactors

9

WG

Sidik Permana,

10

MOX Fuel Behavior [1]WG

Sidik Permana,

Na

Void

Ref : - Hibi and Sekimoto / Journal of nucl. Science and technol, Vol. 42, No. 2, p. 153–160 (2005)

H2O

D2O Na

MOX Fuel

Neutron Spectrum for diff. coolants Plutonium Composition [wt %]

Hard Spectrum : Void>Na>D2O>H2O Fissile Content (X/HM<2):Na<H2O<D2O

11

WG

Sidik Permana,

Ref : - Hibi and Sekimoto / Journal of nucl. Science and technol, Vol. 42, No. 2, p. 153–160 (2005)

H2O

D2O

Na

H2O

D2O

Na

MOX Fuel Behavior [2]

High BR : Na>D2O>H2O Less Void (X/HM<2):Na<H2O<D2O

Breeding Ratio Profile Void Reactivity Coefficient

12

Water-Cooled Thorium Breeder Reactors

Comparative Analysis on Physical

Properties of Water Cooled Reactors

12

WG

Sidik Permana,

13

ReprocessingReactors

H2O/D2O

Coolants

Fabrication

U-233/All HM

Pu/All HM

U

Th

Supply

Storage

WG

Sidik Permana,

Fuel Cycle Options

Supply Fuel : Natural Uranium or Thorium

Recycled Fuel : Plutonium or U-233 or All Heavy Metals (HM)

Physical Parameters : Neutron Spectrum, Eta-value

Investigated Parameters : Required Enrichment, Breeding

and void reactivity coefficient

Others

140.001

0.01

0.1

1

10

100

0.001 0.1 10 1000 105

107

Re

lati

ve N

eu

tro

n S

pe

ctr

a p

erl

eth

arg

y[#

/cm

2s

]

Energy [eV]

D2O

H2O

MFR : Moderator to Fuel Ratio

- D2O harder than H2O

- Very small thermal peak of

D2O, shifts to higher energy.

Neutron Spectrum :

Light water Coolant (H2O) and

Heavy Water Coolant (D2O) at

MFR =2

WG

Sidik Permana,

Neutron Spectrum[2]

15

WG

Sidik Permana,

Neutron Spectrum[4]

H2O coolant , MFR: 0.1 – 1.6

à Low energy region < 1 keV.

H2O coolant

à High energy region > 1 keV.

Ref : - S. Permana et al. / Journal of nucl. Science and technol, Vol. 44, No. 7, p. 946–957 (2007)

16

WG

Sidik Permana,

Neutron Spectrum[4]

0.01

0.1

1

10

100

0.01 1 100 104

106

Rela

tive

Neu

tro

n S

pec

tra p

erl

eth

arg

y[#

/cm

2s

]

Energy [eV]

MFR=20

0.1

0.5

28.0

8.0

MFR=20

D2O Coolant

Harder

Spectrum è

Less MFR

Neutron Spectrum :

Heavy Water Coolant (D2O)

for several MFR

D2O coolant : MFR = 0.1 - 20

17

WG

Sidik Permana,

10-5

10-4

10-3

10-2

100

102

104

106

Re

lati

ve

Ne

utr

on

Sp

ec

tra

[1

/le

tha

rgy

]

Nuetron Energy [eV]

U-Pu, D2O

Th-233U, D2O

U-Pu, H2O

Th-233U, H2

O

10-5

10-4

10-3

10-2

100

102

104

106

Re

lati

ve

Ne

utr

on

Sp

ec

tra

[1

/le

tha

rgy

]

Nuetron Energy [eV]

U-Pu, D2O

Th-233U, D2O

U-Pu, H2O

Th-233U, H2

O

Neutron Spectrum[3]

MFR=2

H2O and D2O Coolants for different fuels

Thorium fuel shows softer than U-Pu

fuel for both water coolants

18

WG

Sidik Permana,

Eta Value [1]

H2O coolant, MFR : 0 – 1.6

D2O coolant, MFR : 0 – 30

Eta value of U-233 à

superior than other

fissile materials

19

1.6

1.8

2

2.2

2.4

0

0.02

0.04

0.06

0.08

0.1

0 0.5 1 1.5 2

Eta

[-]

Eta

of 2

32T

h [-]

Moderator to Fuel Ratio (MFR) [-]

6 GWd/t

Eta Value [2]

Eta value of U-233 à

superior than other

fissile materials and

almost constant along

the MFR

WG

Sidik Permana,

D2O coolant, MFR : 0.1 – 2

20

1.6

1.7

1.8

1.9

2

2.1

2.2

2.3

0 0.5 1 1.5 2 2.5 3 3.5 4

Eta

-Valu

e[-

]

MFR[-]

233U-D2O

WG

Sidik Permana,

Eta Value [3]

H2O and D2O Coolants for different fuels

Eta value of U-233 à

superior than other fissile

materials along the MFR

21

0

2

4

6

8

10

0 1 2 3 4 5

En

ric

hm

en

t U

235 o

r U

233 [

%]

Moderator to Fuel Ratio[-]

U-Pu, H2O

Th- 233U, H2O

U-Pu, D2O

Th- 233U, D2O

0

2

4

6

8

10

0 1 2 3 4 5

En

ric

hm

en

t U

235 o

r U

233 [

%]

Moderator to Fuel Ratio[-]

U-Pu, H2O

Th- 233U, H2O

U-Pu, D2O

Th- 233U, D2O

0.5

0.6

0.7

0.8

0.9

1

1.1

1.2

0 1 2 3 4 5

Co

nv

ers

ion

Ra

tio

[-]

Moderator to Fuel Ratio[-]

U-Pu, H2O

Th- 233U, H2O

U-Pu, D2O

Th- 233U, D2O

Breeding line

0.5

0.6

0.7

0.8

0.9

1

1.1

1.2

0 1 2 3 4 5

Co

nv

ers

ion

Ra

tio

[-]

Moderator to Fuel Ratio[-]

U-Pu, H2O

Th- 233U, H2O

U-Pu, D2O

Th- 233U, D2O

Breeding line

Fissile Content and Conversion RatioWG

Sidik Permana,

H2O and D2O Coolants for different fuels

Required Fissile Content Conversion Ratio

22

-2

-1.5

-1

-0.5

0

0.5

1

1.5

0 1 2 3 4 5

Vo

id R

ea

cti

vit

y x

1e

-3[d

k/k

/%v

ol]

Moderator to Fuel Ratio[-]

Th- 233U, H2O

Th- 233U, D2O

Void Reactivity CoefficientWG

Sidik Permana,

Negative Void Coefficient

-MOX_D2O : Always Positive

-MOX_H2O : Negative 1<MFR<3.4)

-Th_U233_D2O : Negative for MFR>0.4

-Th_U233_H2O : Negative for MFR<1.4

36 GWd/t

23

Water-Cooled Thorium Breeder Reactors

Feasibility Analysis on Water-Cooled

Breeder Reactor

23

WG

Sidik Permana,

24

ReprocessingReactors

H2O/D2O

Coolants

Fabrication

U-233/All HM

Th

Supply

Storage

WG

Sidik Permana,

Fuel Cycle Options

Supply Fuel : Thorium

Recycled Fuel : U-233 or All Heavy Metals (HM)

Physical Parameters : Neutron Spectrum, Eta-value

Investigated Parameters : Required Enrichment, Breeding

and void reactivity coefficient

25

Ref : - S. Permana et al. / Journal of nucl. Science and technol, Vol. 44, No. 7, p. 946–957 (2007)

- Global, 2007

Negative void

coefficient

Feasible area of breeding and

negative void coefficient

H2O

coolant

Void coefficient profile

Light water coolant à

Shows A feasible design area for

breeding and negative void reactivity

coefficient

Feasible Breeding and Negative Void Reac.WG

Sidik Permana,

26

Ref : - S. Permana et al. / Journal of nucl. Sciece and technol, Vol. 45, No. 7, p. 589–600 (2008)

- Global, 2007

A feasible design of breeding and

negative void coefficientVoid coefficient Profile

2O

coolant

Feasible Breeding and Negative Void Reac.WG

Sidik Permana,

27

Ref :

- S. Permana et al. / Journal of nucl. Sciece and technol, Vol. 45, No. 7, p. 589–600 (2008)

- S. Permana et al. / Journal of nucl. Sciece and technol, Vol. 44, No. 7, p. 946–957 (2007)

- Global, 2007

Output : 3GWtCore Hight : 3.7m

D2O-cooled breeder reactor

- MFR=1

- Pellet power density of 140W/cc

- Burn-up of 50 GWd/t.

Wider feasible

window of breeding

for D2O-cooled

2O

coolant

2O

coolant

Comparative H2O and D2O CoolantsWG

Sidik Permana,

28

Comparative U-233 and All HM closed CycleWG

Sidik Permana,

U-233 Closed Cycle All HM Closed Cycle

Ref :

- S. Permana et al. / Annals of Nuclear Energy, Accepted, 2010

D2O Coolant

29

Comparative U-233 and All HM closed CycleWG

Sidik Permana,

Ref :

- S. Permana et al. / Annals of Nuclear Energy, Accepted, 2010

D2O Coolant

Conversion ratio :

U-233 closed > HM closed

for MFR > 0.5

Breeding ratio :

U-233 closed : MFR < 1.5

All HM closed : MFR > 1.2

Void coefficient :

U-233 closed less void

coefficient than All HM closed

30

Comparative U-233 and All HM closed CycleWG

Sidik Permana,

Ref :

- S. Permana et al. / Annals of Nuclear Energy, Accepted, 2010

Heavy water coolant for both

U-233 only recycling and All

HM recycling options à

Shows A feasible design area

for breeding and negative void

reactivity coefficient

U-233 only recycling case

obtains wider feasible design

area for breeding and negative

void reactivity coefficient than

All HM closed cycle.

31

Water-Cooled Thorium Breeder Reactors

Feasibility Analysis on Water-Cooled

Breeder Reactor with MA doping

31

WG

Sidik Permana,

32

ReprocessingReactors

H2O/D2O

Coolants

Fabrication

All HM

Th

Supply

Storage

WG

Sidik Permana,

Fuel Cycle Options

Supply Fuel : Thorium

Recycled Fuel : U-233 or All Heavy Metals (HM)

Investigated Parameters : Required Enrichment, Breeding

and void reactivity coefficient

MA

33

Conversion Ratio

By doped MA, the breeding capabilities

are improved. Higher breeding for Higher

MA doped.

- Conversion ratio monotonically

decreases with increasing burnup.

- Breeding can be achieved for burnup <

45 GWd/t.

WG

Sidik Permana,

Effect of MA content (%) Effect of Burnup

34

- Breeding capability monotonically

decreases with increasing MFR.

- Breeding can be achieved for MFR < 1.1

- Breeding capability monotonically

increases with decreasing power

density (PD) of fuel pellet.

-Breeding can be achieved for fuel

pellet PD ≤ 200 W/cc.

Conversion Ratio

Effect of MFR Effect of Power Density

WG

Sidik Permana,

35

Water-Cooled Thorium Breeder Reactors

Core Design Analysis on Water-Cooled

Breeder Reactor

35

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Sidik Permana,

36

-Refueling modes : Once Through

methods.

- Three batches core configuration

systems.

- Preliminary Thermal Hydraulic

Analysis.

Evaluate the reactor core performances and fuel management by using

core burnup of SRAC COREBN calculations which adopted 2-dimensional

hexagonal model as the core fuel configuration.

- Basic Reactor : Heavy water

cooled thorium reactors

- Core configuration : Refers to

optimum result of equilibrium

cycle iterative calculation systems

(ECICS)

Objectives and Evaluation

- Fuel Breeding Capability and Reactor Criticality

- Negative void reactivity coefficient. -

Thermal Hydraulic Properties comparable to PWR.

WG

Sidik Permana,

371

2

3

4

5

6

7

8

0.8

0.9

1

1.1

1.2

1.3

1.4

1.5

0 0.2 0.4 0.6 0.8 1 1.2 1.4 1.6 1.8 2

6 GWd/t

18

36

50

En

rich

me

nt

233U

Co

nve

rsio

n R

atio

[-]

MFR[-]

Burnup

decreases

Required enrichment and Conversion ratio

High burnup à High

required enrichment or

fissile content, except

for very low MFR with

less moderator

High burnup à less

conversion ratio

(higher consumption

ratio of fissile)

WG

Sidik Permana,

38-40

-30

-20

-10

0

10

0 0.2 0.4 0.6 0.8 1 1.2 1.4 1.6 1.8 2

6 GWd/t

18

36

50

Vo

id R

eac

tiv

ity C

oef.

x1

0e

-3[d

k/k

/%v

ol]

MFR[-]

Burnup

increases

50% void

Void Reactivity Coefficient

High burnup à Less

void reactivity

coefficient or becomes

positive void.

High MFR à High Void

reactivity coefficient

WG

Sidik Permana,

390

0.5

1

1.5

2

2.5

3

0 10 20 30 40 50 60

MF

R[-

]

Burnup [GWd/t]

50% void

Feasible area of breeding

and negative void reactivity

Feasible Breeding Area

High burnup à Narrow

feasible area of

breeding and negative

void coefficient

Less MFR à preferable

to have better breeding,

however, its limited by

negative void reactivity

limitation

WG

Sidik Permana,

40

Unit

Power MWt

Core Height (no reflector) cm

Core Radius cm

Reflector width cm

Fuel pellet Diameter cm

Fuel Pin Diameter cm

MFR (without cladding) -

Pin Pitch Gap cm

Pin Pitch cm

P/D -

U-233 Enrichment %

Cycle Length Days

Achievable burnup GWd/t

Refueling Scheme batch

0.4

1.86

1.282

Parameters

3411

370

179

700

38

3

24

1.31

1.45

6.8

1

Basic Parameters WG

Sidik Permana,

41

Calculation Schemes

Rf

Rcl

Rco

Rf

Rcl

Rco

Fuel Pin

Fuel Assembly

Rf : Radius of fuel

Rcl : Radius of cladding

Rco : Radius of coolant

Core 1

Core 2

Core 3Out-In

Mode

Fresh

Fuel

Core 1

Core 2

Core 3

Core 1

Core 2

Core 3Out-In

Mode

Fresh

Fuel

Core 1

Core 2

Core 3In-Out

Mode

Fresh

Fuel

Core 1

Core 2

Core 3

Core 1

Core 2

Core 3In-Out

Mode

Fresh

Fuel

Out-In Shuffling Scheme

In-Out Shuffling Scheme

41

WG

Sidik Permana,

42

Criticality and Breeding Profiles

0.95

1

1.05

1.1

0 500 1000 1500 2000Operation Time [Days]

K-EFF

Conversion Ratio

Out-in Method

700 Days 700 Days 700 Days

Cycle Length : 700 Days

Fuel Core Batches : 3 Batches

Criticality (K-Eff) : Decreases with

Reactor Operation Time

Breeding (Conversion Ratio):

Increases with Operation Time

At BOC : Breeding à less than unity

At EOC : Breeding à High than unity

Achievable Discharged Fuel Burnup:

More than 33 GWd/t

Out In Method :

àLess criticality for next recycling step

àConversion ratio starts from less than

unity at BOC and reaches higher than

unity at EOC.

àIt confirmed breeding can be achieved

42

WG

Sidik Permana,

43

Criticality and Breeding Profiles

0.95

1

1.05

1.1

0 500 1000 1500 2000Operation Time [Days]

K-E

FF

Conversion Ratio

In-Out Method

700 Days 700 Days 700 Days

Cycle Length : 700 Days

Fuel Core Batches : 3 Batches

Criticality (K-Eff) : Decreases with

Reactor Operation Time

Breeding (Conversion Ratio):

Increases with Operation Time

Out In Method :

àIt confirmed breeding can be achieved

Breeding (Conversion Ratio):

In-Out Method : 1.01

Out-In Method : 1.02

WG

Sidik Permana,

44

Void Reactivity Coefficient

-300

-250

-200

-150

-100

-50

0

50

100

0 500 1000 1500 2000

5% void fraction100%

Vo

id R

ea

cti

vit

y C

oe

ffic

ien

t [p

cm

]

Burnup[Days]

Out-in Method

-300

-250

-200

-150

-100

-50

0

50

100

0 500 1000 1500 2000

5% void fraction100%

Vo

id R

ea

cti

vit

y C

oe

ffic

ien

t [p

cm

]

Burnup[Days]

In-Out Method

Void reactivity coefficient :

à Always negative during reactor operation

à Higher void fraction : less negative at EOC

Void reactivity coefficient :

à Voided fraction effect : low for out-in

method and higher for in-out method44

WG

Sidik Permana,

45

Temperature Distribution

250

300

350

400

450

0 50 100 150 200 250 300 350 400Axial Direction from bottom to top [cm]

Coolant Flow

Cladding Surface

Out-In Method

Temperature [C]

250

300

350

400

450

0 50 100 150 200 250 300 350 400Axial Direction from bottom to top [cm]

Coolant Flow

Cladding Surface

In-Out Method

Temperature [C]

Temperature :

à Maximum Temp. cladding surface : Out-

In : Less 400 C, In-out : reaches 400 C

Temperature:

à Out-in method : Relatively Higher

temperature than In-out method45

WG

Sidik Permana,

46

Temperature Distribution

300

350

400

450

0 50 100 150 200 250 300 350 400Axial Direction from bottom to top [cm]

Out-in Method

In-out Method

Temperature of Cladding Surface [C]

1000

1100

1200

1300

1400

1500

1600

1700

1800

0 50 100 150 200 250 300 350 400Axial Direction from bottom to top [cm]

Out-in Method

In-out Method

Temperature of Fuel Center-line [C]

46

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Sidik Permana,

47

Power Peaking Profile

0.5

1

1.5

2

2.5

0 400 800 1200 1600 2000 2400

Out-In MethodIn-Out Method

Po

wer

Peakin

g [

-]

Burnup[Days]

Power peaking : Ratio of maximum

power density to the average power

density

- Out-In Method : decreases for the

next recycling step

- In-Out Method : increases for nex

recycling step

- Out-in method : less than 1.5

- In-Out Method : reaches more than

2

Power peaking profile shows the

maximum different of power at a

certain location to the average total

power distribution.

WG

Sidik Permana,

48

Thermal Hydraulic Properties

Unit

Tinlet 300 °C

Toutlet 332 °C

Average Thermal Conductivity of TH 2.75 W/m K

Average Thermal Conductivity of Zr- 10.7 W/m K

Themal Conductivity of Heavy water 0.483 W/m K

Specific heat capacity of Heavy water 4228 J/Kg K

Fuel Area 1.66E-04 m2

Volume of Fuel 6.12E-04 m3

Coolant Area 1.08E-04 m2

Max Power density of Core 1.10E+08 W/m3

Fluid Density 720 Kg/m3

P/D 1.28 -

Hydraulic Diameter 0.0118 m

Heat Flux clad surface 4.07E+05 W/m2

Parameters Unit

Mean flow velocity 4.14 m/s

Renoylds number 2.80E+04

Prandtl number 10.94 -

Heat Trannsfer 2.85E+04 W/m2 K

Fanin Friction 6.11E-03

Friction Pressure drop 0.47 bar

Fuel Temperature Drop 526 C

Total Temperature drop 679 C

Maximum Coolant Temperature 332 C

Maximum Fuel Temperature 1058 C

Thermal hydraulic parameter

Out-in Method

WG

Sidik Permana,

49

Thermal Hydraulic Properties

Unit

Tinlet 300 °C

Toutlet 332 °C

Average Thermal Conductivity of TH 2.75 W/m K

Average Thermal Conductivity of Zr- 10.7 W/m K

Themal Conductivity of Heavy water 0.483 W/m K

Specific heat capacity of Heavy water 4228 J/Kg K

Fuel Area 1.66E-04 m2

Volume of Fuel 6.12E-04 m3

Coolant Area 1.08E-04 m2

Max Power density of Core 2.11E+08 W/m3

Fluid Density 720 Kg/m3

P/D 1.28 -

Hydraulic Diameter 0.0118 m

Heat Flux clad surface 7.66E+05 W/m2

Parameters Unit

Mean flow velocity 7.79 m/s

Renoylds number 5.28E+04

Prandtl number 10.94 -

Heat Trannsfer 5.38E+04 W/m2 K

Fanin Friction 5.22E-03

Friction Pressure drop 1.43 bar

Fuel Temperature Drop 1008 C

Total Temperature drop 1303 C

Maximum Coolant Temperature 332 C

Maximum Fuel Temperature 1704 C

Thermal hydraulic parameter

In-Out Method

WG

Sidik Permana,

50

Conclusion

1. Core burnup calculations have confirmed that breeding is

feasible for water cooled thorium reactor system. It also

confirmed that negative void reactivity coefficients are

obtained during reactor operation.

2. Fuel breeding capabilities have been shown 1.01 (Out-In

method) and 1.02 (In-Out method) at the end of cycle.

3. Thermal hydraulic parameters show the comparable

result with conventional reactors and have the large

margin to the limitation of thermal hydraulic properties.

4. Reactor core optimization for neutronic and thermal

hydraulic aspects should be done for future investigation

WG

Sidik Permana,

51

END OF PRESENTATION

Thank You

Questions or

comments?

WG

Sidik Permana,