添付資料：ワーキンググループにおける報告資料Canada 45,000 128,000 Australia 19,000 — South Africa 18,000 — Egypt 15,000 309,000 Other Countries 505,000 —

添付資料：

ワーキンググループにおける報告資料

検討課題１「トリウム燃料利用に適した

炉心設計・選定とエネルギー収支評価」

水冷却トリウム増殖炉

近年の固体燃料トリウム炉の研究動向

トリウム炉の導入シナリオと発電コスト

京都大学原子炉実験所におけるトリウム利用炉関連炉物理実験

軽水炉へのトリウム燃料適用の検討

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,

6Neutron 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


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


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


Comparative Analysis on Physical Properties of Water Cooled Reactors

12

WG

Sidik Permana,

13

ReprocessingReactorsH2O/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

Rel

ativ

e N

eutr

on S

pect

ra p

erle

thar

gy[#

/cm

2 s]

Energy [eV]

D2OH2O

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�� Low energy region < 1 keV.

H2O coolant� 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

Rel

ativ

e N

eutr

on S

pect

ra p

erle

thar

gy[#

/cm

2 s]

Energy [eV]

MFR=20

0.1

0.5

28.0

8.0

MFR=20

D2O Coolant

Harder Spectrum �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

Rel

ativ

e N

eutr

on S

pect

ra [1

/leth

argy

]

Nuetron Energy [eV]

U-Pu, D2O

Th-233U, D2O

U-Pu, H2O

Th-233U, H2O

10-5

10-4

10-3

10-2 100 102 104 106

Rel

ativ

e N

eutr

on S

pect

ra [1

/leth

argy

]

Nuetron Energy [eV]

U-Pu, D2O

Th-233U, D2O

U-Pu, H2O

Th-233U, H2O

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 ��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 232Th [-]

Moderator to Fuel Ratio (MFR) [-]

6 GWd/t

Eta Value [2]Eta value of U-233 ��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]


Eta value of U-233 ��superior than other fissile materials along the MFR

21

0

2

4

6

8

10

0 1 2 3 4 5

Enric

hmen

t U23

5 or 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

Enric

hmen

t U23

5 or U

233 [%

]


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

Con

vers

ion

Rat

io [-

]


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

Con

vers

ion

Rat

io [-

]


U-Pu, H2O

Th- 233U, H2O

U-Pu, D2O

Th- 233U, D2O

Breeding line

Fissile Content and Conversion Ratio WG

Sidik Permana,


Required Fissile Content Conversion Ratio

22

-2

-1.5

-1

-0.5

0

0.5

1

1.5

0 1 2 3 4 5

Void

Rea

ctiv

ity x

1e-3

[dk/

k/%

vol]


Th- 233U, H2O

Th- 233U, D2O

Void Reactivity Coefficient WG

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


Feasibility Analysis on Water-Cooled Breeder Reactor

23

WG

Sidik Permana,

24


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


25Ref : - 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 �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 coefficient

Void 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 Coolants WG

Sidik Permana,

28

Comparative U-233 and All HM closed Cycle WG

Sidik Permana,

U-233 Closed Cycle All HM Closed Cycle

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

D2O Coolant

29


Sidik Permana,


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


Sidik Permana,


Heavy water coolant for both U-233 only recycling and All HM recycling options �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


Feasibility Analysis on Water-Cooled Breeder Reactor with MA doping

31

WG

Sidik Permana,

32


Fabrication

All HM

Th

Supply

Storage

WG

Sidik Permana,

Fuel Cycle Options

Supply Fuel : Thorium

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


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 RatioEffect of MFR Effect of Power Density

WG

Sidik Permana,

35


Core Design Analysis on Water-Cooled Breeder Reactor

35

WG

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/t183650

Enric

hmen

t 233 U

Conversion R

atio[-]

MFR[-]

Burnup decreases

Required enrichment and Conversion ratio

High burnup � High required enrichment or fissile content, except for very low MFR with less moderator

High burnup � 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/t183650

Void

Rea

ctiv

ity C

oef.

x10e

-3[d

k/k/

%vo

l]

MFR[-]

Burnup increases

50% void

Void Reactivity Coefficient

High burnup �� Less void reactivity coefficient or becomes positive void.

High MFR � High Void reactivity coefficient

WG

Sidik Permana,

390

0.5

1

1.5

2

2.5

3

0 10 20 30 40 50 60

MFR

[-]

Burnup [GWd/t]

50% void

Feasible area of breeding and negative void reactivity

Feasible Breeding Area

High burnup � Narrow feasible area of breeding and negative void coefficient

Less MFR � 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 fuelRcl : Radius of claddingRco : 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 DaysFuel Core Batches : 3 Batches

Criticality (K-Eff) : Decreases with Reactor Operation Time

Breeding (Conversion Ratio): Increases with Operation Time

At BOC : Breeding � less than unityAt EOC : Breeding � High than unity

Achievable Discharged Fuel Burnup:More than 33 GWd/t

Out In Method :�Less criticality for next recycling step�Conversion ratio starts from less thanunity at BOC and reaches higher than unity at EOC.�It 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-EFF

Conversion Ratio

In-Out Method

700 Days 700 Days 700 Days

Cycle Length : 700 DaysFuel Core Batches : 3 Batches

Criticality (K-Eff) : Decreases with Reactor Operation Time

Breeding (Conversion Ratio): Increases with Operation Time

Out In Method :�It confirmed breeding can be achieved

Breeding (Conversion Ratio): In-Out Method : 1.01Out-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%

Void

Rea

ctiv

ity C

oeffi

cien

t [pc

m]

Burnup[Days]

Out-in Method

-300

-250

-200

-150

-100

-50

0

50

100

0 500 1000 1500 2000

5% void fraction100%

Void

Rea

ctiv

ity C

oeffi

cien

t [pc

m]

Burnup[Days]

In-Out Method

Void reactivity coefficient : �� Always negative during reactor operation� Higher void fraction : less negative at EOC

Void reactivity coefficient : � Voided fraction effect : low for out-in method and higher for in-out method 44

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


Coolant Flow

Cladding Surface

In-Out Method

Temperature [C]

Temperature : �� Maximum Temp. cladding surface : Out-In : Less 400 C, In-out : reaches 400 C

Temperature: � Out-in method : Relatively Higher temperature than In-out method 45

WG

Sidik Permana,

46

Temperature Distribution

300

350

400

450


Out-in Method

In-out Method

Temperature of Cladding Surface [C]

1000

1100

1200

1300

1400

1500

1600

1700

1800


Out-in Method

In-out Method

Temperature of Fuel Center-line [C]

46

WG

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

Pow

er P

eaki

ng [-

]

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 PropertiesUnit

Tinlet 300 °CToutlet 332 °CAverage Thermal Conductivity of TH 2.75 W/m KAverage Thermal Conductivity of Zr- 10.7 W/m KThemal Conductivity of Heavy water 0.483 W/m KSpecific heat capacity of Heavy water 4228 J/Kg KFuel Area 1.66E-04 m2Volume of Fuel 6.12E-04 m3Coolant Area 1.08E-04 m2Max Power density of Core 1.10E+08 W/m3Fluid Density 720 Kg/m3P/D 1.28 -Hydraulic Diameter 0.0118 mHeat Flux clad surface 4.07E+05 W/m2

Parameters UnitMean flow velocity 4.14 m/sRenoylds number 2.80E+04Prandtl number 10.94 -Heat Trannsfer 2.85E+04 W/m2 KFanin Friction 6.11E-03Friction Pressure drop 0.47 barFuel Temperature Drop 526 CTotal Temperature drop 679 CMaximum Coolant Temperature 332 CMaximum Fuel Temperature 1058 C

Thermal hydraulic parameter

Out-in Method

WG

Sidik Permana,

49

Thermal Hydraulic PropertiesUnit

Tinlet 300 °CToutlet 332 °CAverage Thermal Conductivity of TH 2.75 W/m KAverage Thermal Conductivity of Zr- 10.7 W/m KThemal Conductivity of Heavy water 0.483 W/m KSpecific heat capacity of Heavy water 4228 J/Kg KFuel Area 1.66E-04 m2Volume of Fuel 6.12E-04 m3Coolant Area 1.08E-04 m2Max Power density of Core 2.11E+08 W/m3Fluid Density 720 Kg/m3P/D 1.28 -Hydraulic Diameter 0.0118 mHeat Flux clad surface 7.66E+05 W/m2

Parameters UnitMean flow velocity 7.79 m/sRenoylds number 5.28E+04Prandtl number 10.94 -Heat Trannsfer 5.38E+04 W/m2 KFanin Friction 5.22E-03Friction Pressure drop 1.43 barFuel Temperature Drop 1008 CTotal Temperature drop 1303 CMaximum Coolant Temperature 332 CMaximum Fuel Temperature 1704 C

Thermal hydraulic parameter

In-Out Method

WG

Sidik Permana,

50

Conclusion1. 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,

113-8656

1

� Proceedings of ICAPP ’10, San Diego, CA, USA, June 13-17, 2010

� Proceedings of The 18th International Conference on Nuclear Engineering (ICONE18) , Xi'an, China., May 17-21, 2010

� Proceedings of GLOBAL 2009, "The Nuclear Fuel Cycle: Sustainable Options & Industrial Perspectives“, Paris France, Sep. 6-11, 2009

Thorium Based Fuel Cycle Options for PWRsMichael Todosow and Gilad RaitsesBrookhaven National Laboratory,

• 17 17 PWR

1.U233-Th U-233

2.TRU (Np, Pu, Am) -Th U-233 TRU59GWd/t ALWR5 +2

3.U-TRU-Th U-TRU ”2”TRU-Th

•1. U-233 LWBR, FBR, MSR, Fusion-

Fission Hybrid, ADS pre breeder TRU TRUU233 10,000

2. 1 Pu Am, TRUMOX

3. 2 MOXU-235 Pu U-233

2

(Paper from ICAPP’10)

Investigation on the feasibility of thorium breeder reactor in a BWRYoshitaka FUNAHASHI, Yoichiro SHIMAZU, Tadashi NARABAYASHI, Masashi TSUJIGraduate school of Engineering, Hokkaido University

• 8 8 Th-U233 BWR

•–– U-233

––

H/F moderator to fuel number ratio= 0.32.8

•• Sidik / (MFR)=0.3 MFR

BWR MFR

• 8 8 MFR 1.5 9 9

3


Usage of Thorium Based Nuclear Fuel in VVER reactors

• VVER-1000

• Th fissile U Pu U-233

•– U Pu

– U-233

– Fissile 5

4


• Th fissile• Pa-233

• Pa-233 U-234U-233

• U-233U-233

ACHIEVING RESOURCE SUSTAINABILITY IN CHINA THROUGH THE THORIUM FUEL CYCLE IN THE CANDU REACTORPeter G. Boczar* et.al.Atomic Energy of Canada Limited (AECL)

• CANDU

•– 8 ThO2 U LEU CANFLEX

CANDU

– U LEU Th U-233

•– Pu Th-Pu U

•– U-233 Th Pu

1.0 CANDU

•• 20MWd/kg U-233

FBR Pu fissile CANDU FBRFBR fissile U-233/Th

•MFR

• CANDUPu

5

(Paper from ICONE18, Xi’an)

6

~1.25 wt% LEU~1.70 wt% LEU

ThO2

ThO2

7

0

1

2

3

4

5

6

0 0.1 0.2 0.3 0.4 0.5 0.6

U

U [wt%]

1.7%LEU-Th

1.25%Breakeven line

8 U/ThU U/Th U

U

Comparison of Thorium-Plutonium fuel and MOX fuel for PWRsKlara Insulander Björk a and Valentin Fhager Thor Energy A/S

• PWR (U, Pu)O2 (Th, Pu)O2

– MOX (U, Pu)O2 PWR55GWd/t (Th, Pu)O2

– (Th, Pu)O2 PWR

• Pu 21% (MOX 15%)

• MOX

•••• Pu

– (Th, Pu)O2 MOX

–

8

(Paper from Global2009)

Loss of Power to Recirculation Pumps in the VVER-1000 Reactor with Thorium Power, Ltd. Thorium Seed and Blanket Fuel Assembliesby Alexei Morozov, Michael Montgomery, Andrey Mushakov(Thorium Power-USA)

• VVER-1000 seed/blanket

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9

• U-Zr seedThO2/UO2 Zr

• seed/blanket

VVER-1000 UO2

• U-Zr seedVVER-1000

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• Thorium Fuel Cycle - an Alternative Options for LWRs

by Juraj Breza, Peter Čudrnák (VUJE, Slovak Univ of Technology-Slovakia), Petr Dařílek (VUJE-Slovakia), Vladimír Nečas (Univ of Technology-Slovakia)

– VVER-440 Pu Th-Pu Th-Pu-U233

– Th-Pu SF Pu UOXPu

• Suggestions on Development of the Thorium Fuel Cycle in China

by Yongming Hu (Tsinghua Univ-China), Xiuan Shi (China National Nuclear Corporation-China), Zhiwei Zhou (Tsinghua Univ-China)

• Irradiation of Thorium - Plutonium Mixed Oxide Fuel to 37.7 GWd/tonne in the Obrigheim Pressurised Water Reactor (KWO)

by J. Somers, D. Papaoiannou (JRC-ITU-Germany), D. Sommer (EnBW Kernkraft GmbH-Germany)

• Feasibility Study for Thorium Fuel

by Tomas Lefvert (Vattenfall AB-Sweden), Øystein Asphjell (Thor Energy AS-Norway)10


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京都⼤学原⼦炉実験所におけるトリウム利⽤炉関連炉物理実験

Nov. 18, 2010

京都⼤学原⼦炉実験所宇根崎博信

H. Unesaki / Thorium WG @ Osaka University, Nov. 18, 2010 2

KUCA Experiments for Advanced Nuclear Reactor System

To provide fundamental scientific information for the development of advanced nuclear reactor, the authors are performing a series of experimental studies at the Kyoto University Critical Assembly(KUCA) facility.

The following specific topics are of present interest; 1) benchmark experiments for high burnup next generation reactor fuel, 2) basic experiments on thorium fueled reactor, and

3) experiments on ADS using high-energy proton accelerator and subcritical cores

These experiments are mainly aimed at verification and validation of current methodology for nuclear characteristics design, and also aimed at development of experimental techniques.


Thorium : Motivations

Thorium has recently regained a growing interest from nuclear society. This is due to the attractive potential of thorium-based fuel cycle, such as its rich natural resource (recently with relation to rare-earth ore waste), less possibility of generating TRU wastes and excellent non-proliferation characteristics.

For the reliable design of thorium-based systems, the accuracy of neutron cross section, especially that of Th-232, will be of primary importance. Compared to the uranium-plutonium fuel cycle, less attention have been paid to the validation of nuclear data related to thorium fuel cycle.


Impact of Nuclear Library Difference on Neutronic Characteristics of Thorium-loaded Light Water Reactor Fuel

PWR UO2, next generation fuel benchmark (2001)

Enrich. 7% : Up to 70GWd/t UO2, (235U,Th)O2

Light Water Moderator

Fuel Pellet

Zr Cladding

Cell Pitch : 1.265cm

Pellet diameter : 0.824cm

Cladding outer diameter : 0.952cm

Nuclide UO2 (U,Th)O2 235U 1.5122E-3* 1.5122E-3 238U 2.1477E-2 ---

232Th --- 2.2032E-2 16O 4.5945E-2 4.5945E-2

Zr-natural 4.3107E-2 4.3107E-2 H2O 3.3315E-2 3.3315E-2

*in barn/cm; read as 1.512210-3

• Neutronic calculation was performed using the SRAC code system with its build-in 107-group cross section set generated from JENDL-3.3, ENDF/B-VI.8 or JEFF3.0 libraries.

• Cell burnup calculations with 2-dimensional collision probability method as transport solver were performed up to discharge burnup of 70GWd/t.

• The linear power density was set to 179W/cm.


Results : k-infinity

0.8

0.9

1

1.1

1.2

1.3

1.4

0 10 20 30 40 50 60 70

Burnup (GWd/t

JENDL-3.3ENDF/B-VI.8JEFF3.0

0.8

0.9

1

1.1

1.2

1.3

1.4

1.5

0 10 20 30 40 50 60 70

Burnup (GWd

JENDL-3.3ENDF/B-VI.8JEFF3.0

UO2 (U,Th)O2


K-inf difference (relative to JENDL-3.3 results)

-1.2

-1

-0.8

-0.6

-0.4

-0.2

0

0.2

0.4

0.6

0.8

0 10 20 30 40 50 60 70

Burnup (GWd/t)

ENDF/B-VI.8JEFF3.0

-1

-0.8

-0.6

-0.4

-0.2

0

0.2

0.4

0.6

0.8

1

0 10 20 30 40 50 60 70

Burnup (GWd/t)

ENDF/B-VI.8JEFF3.0

UO2 (U,Th)O2

1.21%k/kk’@BOC 1.13%k/kk’

@EOC


TRU Production

Concentrations of TRU elements obtained using JENDL-3.3, ENDF/B-VI.8 and JEFF3.0 and their relative difference to JENDL-3.3

UO2Element JENDL-3.3 ENDF/B-VI.8 JEFF3.0 ENDF/B-VI.8 JEFF3.0

Np total 3.12E-05 3.01E-05 3.15E-05 -3.38 0.89Pu total 3.92E-04 3.92E-04 3.96E-04 0.07 0.94Am total 9.42E-06 9.91E-06 9.61E-06 5.2 1.97Cm total 4.97E-06 5.29E-06 5.16E-06 6.27 3.71TRU total 4.38E-04 4.38E-04 4.42E-04 0 0.99Np+Am+Cm 4.56E-05 4.53E-05 4.62E-05 -0.55 1.43

(U,Th)O2

Element JENDL-3.3 ENDF/B-VI.8 JEFF3.0 ENDF/B-VI.8 JEFF3.0

Np total 2.79E-05 2.63E-05 2.74E-05 -5.73 -1.65Pu total 1.78E-05 1.72E-05 1.83E-05 -3.49 2.97Am total 4.28E-08 4.42E-08 4.63E-08 3.36 8.14Cm total 1.52E-08 1.55E-08 1.67E-08 1.93 10.01TRU total 4.57E-05 4.35E-05 4.58E-05 -4.85 0.16Np+Am+Cm 2.79E-05 2.63E-05 2.75E-05 -5.71 -1.63

Number Density （1024atms/cc）Relative Difference to

JENDL-3.3（%）

Number Density （1024atms/cc）Relative Difference to

JENDL-3.3（%）


TRU Concentration Difference

-8.00

-6.00

-4.00

-2.00

0.00

2.00

4.00

6.00

8.00

10.00

12.00

UO2, ENDF/B-VI.8 UO2, JEFF3.0 (U,Th)O2, ENDF/B-VI.8 (U,Th)O2, JEFF3.0

Rel

ativ

e di

ffere

nce

to J

EN

DL-

3.3

(%)

Np total Pu total

Am total Cm total

TRU total Np+Am+Cm total


Experiments on U-Th System at KUCA

From this point of view, a series of critical experiments on thorium fueled thermal spectrum cores are being performed at KUCA in order to accumulate experimental information on thermal spectrum systems containing thorium. (Th+Graphite) test zone + U driver core

Th/Graphite ratio varied H/U ratio fixed at polyethylene moderated U driver Polyethylene reflected

(Th+U) core Polyethylene moderated / reflected Th/U ratio varied H/U ratio varied


Kyoto University Critical Assembly (KUCA)

First Criticality: Aug. 1974 / Max. power 100W (1kW for limited time) The only critical assembly owned by university in Japan Multiple core type critical assembly

one Light water moderated core + two Solid material(polyethylene, graphite) moderated cores

Research Subjects Basic Reactor Physics Experiment Nuclear Criticality Safety Neutron Field for Development of Neutron Detector Thorium Fueled Reactor Accelerator Driven Reactor (ADS) Advanced LWR Reactor Laboratory Course for undergraduate and graduate students


KUCA Building

A core

Pulsed Neutron Generator

B Core

Proton beam line from FFAG acc.

C Core

• Three cores in reactor room• Wide variety of material

composition and core geometry• Combined use of accelerator and

subcritical core


Solid Moderated Core & Fuel

Fuel Plate

PolyethylenePlate

} Unit Cell

Core RegionApprox. 40cm

Upper ReflectorApprox. 50cm

Lower ReflectorApprox. 50cm

Aluminum Sheath

Material PlatesFuel & Moderator

Elements

Approx. 150 cm

Fuel Element

Reflector Element

Control Rods

Core Assembly


(Th+C) zone cores : example of core configuration

(Th+C) Test Zone

U Driver Fuel

Control Rod

Polyethylene Reflector

•3x3 test zone: Th metal plate + Graphite plate mixed•Number of U driver depends on test zone configuration

Th II’ (C/Th=48)


(Th+Graphite) Zone Cores

Core ID C/ThB3/8"P50EU(3)-Th I 96B3/8"P50EU(3)-Th II' 48B3/8"P50EU(3)-Th III 24B3/8"P50EU(3)-Th IV' 12B3/8"P50EU(3)-Th V 6B3/8"P50EU(3)-Th 0

•C/Th atom ratio in test zone varied•Polyethylene-moderated U driver fuel (3/8”P50EU fuel) used

0

0.05

0.1

0.15

0.2

0.25

10-3 10-2 10-1 100 101 102 103 104 105 106 107

Th I (C/Th=96)Th II' (C/Th=48)Th III (C/Th=24)Th IV' (C/Th=12)Th V (C/Th=6)Th Lump (C/Th=0)

Flu

x pe

r Le

thar

gy (

norm

aliz

ed to

uni

ty)

Energy(eV)


1/4” Polyethylene1/16” Enriched Uranium1/8” Polyethylene

1/16” Enriched Uranium

1/8” Thorium1/8” Polyethylene1/16” Enriched Uranium1/4” Polyethylene

Unit Cell

Upper Polyethylene ReflectorApprox. 50cm

Active Core Region: 17 Unit CellsApprox. 45.4cm

Lower Polyethylene ReflectorApprox. 50cm

Polyethylene Reflector ElementsFuel Elements

Control / Safety Rods

Fuel Element (6/8”P17EU-Th-EU-EU fuel)Core (Horizontal View,

B6/8”P17EU-Th-EU-EU(3) Core)

(Th+U) Cores : core configuration


Core ID Unit Cell ID

H/235U ratio

232Th/235

U ratioSpectrum Index*

CoreVolume (liter)

B4/8"P24EU-Th-EU-EU(5) ETEE4 138 12.7 0.184 56.8

B6/8"P24EU-Th-EU-EU(3) ETEE6 211 12.7 0.242 48.8

B3/8"P48EU16Th(3) EU16Th 316 12.7 0.313 58.5

B3/8"P45EU18Th(3) EU18Th 316 15.2 0.309 65.9

B3/8"P30EU-Th-EU(5) ETE3 155 19.0 0.191 93.4

B4/8"P17EU-Th-EU(5) ETE4 207 19.0 0.230 81.2

B6/8"P17EU-Th-EU(5) ETE6 316 19.0 0.297 89.1

(Th+U) Cores : Core Specifications

Spectrum Index (E)

dE

EE1eV

(E)dE

EE10MeV


0

0.05

0.1

0.15

0.2

10-4 10-3 10-2 10-1 100 101 102 103 104 105 106 107

ETE3 (Sp.Indx.=0.191)ETE4 (Sp.Indx.=0.230)ETE6 (Sp.Indx.=0.297)

Cel

l Ave

rage

d S

pect

rum

(S

um F

lux=

1)

Energy (eV)

0

0.05

0.1

0.15

0.2

10-4 10-3 10-2 10-1 100 101 102 103 104 105 106 107

ETEE4 (Sp.Indx.=0.184)ETEE6 (Sp.Indx.=0.242)EU16Th (Sp.Indx.=0.313)

Cel

l Ave

rage

d S

pect

rum

(S

um F

lux=

1)

Energy (eV)

0

0.05

0.1

0.15

0.2

10-4 10-3 10-2 10-1 100 101 102 103 104 105 106 107

EU18Th (Sp.Indx.=0.309)

Cel

l Ave

rage

d S

pect

rum

(S

um F

lux=

1)

Energy (eV)

(Th+U) Cores:Cell-Averaged Neutron Spectrum of fuel cell


Criticality Analysis

•Cont. Energy Monte Carlo : MVP code (JAERI) on PC Linux

•JENDL-3.2, JENDL-3.3, JEFF3.0, ENDF/B-VI.8

(analysis with ENDF/B-VII.0, JEFF3.1 and JENDL-4.0 ongoing)

•3,000,000 - 6,000,000 active histories

•statistical error (1-sigma) = 0.03%-0.05% for k-effective


(Th+C) Zone Cores : C/E

Significant reduction of overestimationaverage C/E 1.011 (JENDL-3.2) --> 1.005 (JENDL-3.3)

1.0000

1.0050

1.0100

1.0150

0 20 40 60 80 100 120

C/Th ratio

C/E, JENDL-3.2C/E, JENDL-3.3


(Th+C) Zone Cores : C/E

• Difference between libraries increase with decreasing C/Th ratio

1.0000

1.0050

1.0100

0 20 40 60 80 100 120

C/Th ratio

C/E, JENDL-3.3C/E, ENDF/B-VI.8C/E, JEFF3.0


KUCA (Th+U) Cores : C/E Values

1.000

1.005

1.010

1.015

1.020

0.15 0.20 0.25 0.30 0.35

Spectrum Index

Th232/U235=12.7, ENDF/B-VI.8Th232/U235=15.2, ENDF/B-VI.8Th232/U235=19.0, ENDF/B-VI.8

1.000

1.005

1.010

1.015

1.020

0.15 0.20 0.25 0.30 0.35

Spectrum Index

Th232/U235=12.7, JEFF3.0Th232/U235=15.2, JEFF3.0Th232/U235=19.0, JEFF3.0

1.000

1.005

1.010

1.015

1.020

0.15 0.20 0.25 0.30 0.35

Spectrum Index

Th232/U235=12.7, JENDL-3.3Th232/U235=15.2, JENDL-3.3Th232/U235=19.0, JENDL-3.3

JENDL-3.3

ENDF/B-VI.8

JEFF3.0


(Th+U) cores : Impact of Th-232 evaluation difference to k-eff (1)

JENDL-3.3 + Th-232 from ENDF/B-VI or JENDL-3.3K-eff difference

-0.50%

-0.40%

-0.30%

-0.20%

-0.10%

0.00%

0.10%

0.15 0.20 0.25 0.30 0.35

Spectrum Index

Th232/U235=12.7, (Th232=B6) - J33Th232/U235=15.2, (Th232=B6) - J33Th232/U235=19.0, (Th232=b6) - J33


(Th+U) cores : Impact of Th-232 evaluation difference to k-eff (2)

JENDL-3.3 + Th-232 from JEFF3.0 or JENDL-3.3K-eff difference

-0.10%

0.00%

0.10%

0.20%

0.30%

0.40%

0.50%

0.15 0.20 0.25 0.30 0.35

Spectrum Index

Th232/U235=12.7, (Th232=F3) - J33Th232/U235=15.2, (Th232=F3) - J33Th232/U235=19.0, (Th232=F3) - J33


(Th+U) Cores : C/E Values

C/E Values of KUCA Uranium Fueled Cores and Uranium / Thorium Fueled Cores by JENDL-3.3

0.995

1.000

1.005

1.010

1.015

0.00 0.05 0.10 0.15 0.20 0.25 0.30 0.35

Spectrum Index

Enrich.=93%, JENDL-3.3Enrich.=9.6%, JENDL-3.3Enrich.=5.4%, JENDL-3.3Th232/U235=12.7, JENDL-3.3Th232/U235=15.2, JENDL-3.3Th232/U235=19.0, JENDL-3.3

•Cont. Energy Monte Carlo : MVP code (JAERI) , 1,000,000 - 3,000,000 histories•statistical error (1-sigma) = 0.05%-0.08% for k-effective


k-inf difference due to 232Th

all data except 232Th = JENDL-3.3

-0.5%-0.4%-0.3%-0.2%-0.1%0.0%0.1%0.2%0.3%0.4%0.5%

0.15 0.20 0.25 0.30 0.35

Spectrum Index

Th232=ENDF/B-VI.8

Th232=JEFF3.0


Conclusions on Thorium experiments

prediction accuracy of thorium fueled thermal systems have been improved by the use of recent data libraries such as JENDL-3.3 and ENDF/B-VI, but is still inferior to that of the conventional uranium fueled systems.

Considerable discrepancy between the 232Th cross section evaluations exist and has been shown to have considerable impact on nuclear characteristics of thorium fueled thermal systems.

Analysis based on JENDL-4 in progress


Future Work

As for the extension of the current studies, various new activities are being conducted or planned as follows; critical experiment on thorium fueled cores for

expanding the variety of core characteristics

basic experiment on thorium-loaded ADS coresusing FFAG proton accelerator and KUCA subcritical cores

scenario studies on introduction of thorium fuel cycle based on fuel cycle modeling using system dynamics


KUCA A-core & FFAG Accelerator

KUCA A core150 MeV

Proton

Beam

Line

FFAG Accelerator


Thorium-loaded core (keff = 0.03625±0.00001 w/o source)

- 5 by 5 fuel assembly with 80 Th plate & No moderator

- Cubic core (~10x10x10 inch) with 180.8 kg thorium loading

11 12 13 14 15 16 17 18 19

い Th Th Th Th Th H e1 Th Th Fuel(80Th)

ろ Th Th Th Th Th P roton B eam

は H e4 Th Th Th Th Th H e# He3 D etector

に Th Th Th Th Th

ほ Th Th Th Th Th

へ H e3 H e2

と

ち

1" Al Plate

1cm Al Sheath Bottom

1" Al Plate

1/8" Th Plate x 4EA

25.4 cm(10")

1.04 cm

0.00 cm

1.00 cm

3.54 cm

69.91 cm

57.21 cm

54.67 cm

82.61 cm

AlPipe

1.50 cm

Φ=2.20cm

50.00cm

1" Al Plate

5.13 cm

7.67 cm

11.27 cm

1/2" Poly + 1/8" Poly

152.4 cm

KUCA A-core (5 by 5 F.A.) Configuration of F.A. with 80 Th fuel plate

In Foil(50x50x1 mm)

Thorium-loaded ADS Exp. (1)


Thorium-loaded core with graphite moderator

(keff = 0.02703±0.00001 w/o external source)

- 7 by 7 fuel assembly with 48 Th plate & 12 graphite plate

- Hexahedron (~14x14x12 inch) with 212.7 kg thorium loading

- Graphite loading with same volume of thorium fuel

Configuration of F.A. with 48 Th fuel plate

1" Al Plate

1" Al Plate

1/8" Th Plate x 4EA

AlPipe

30.48 cm(12")

1.04 cm

1.50 cm

Φ=2.20cm

0.00 cm

1.00 cm

70.86 cm

53.08 cm

55.62 cm

88.64 cm

50.00cm

1/2" Graphite Plate

3.54 cm

1cm Al Sheath Bottom

1" Al Plate

6.08 cm

152.4 cm

11 12 13 14 15 16 17 18 19

い TG ' TG ' TG ' TG ' TG ' TG ' TG ' TG ' Th+G r Fuel(48Th)

ろ TG ' TG ' TG ' TG ' TG ' TG ' TG ' P roton B eam

は TG ' TG ' TG ' TG ' TG ' TG ' TG '

に TG ' TG ' TG ' TG ' TG ' TG ' TG '

ほ TG ' TG ' TG ' TG ' TG ' TG ' TG '

へ TG ' TG ' TG ' TG ' TG ' TG ' TG '

と TG ' TG ' TG ' TG ' TG ' TG ' TG '

ち

KUCA A-core (7 by 7 F.A.)

In Foil(50x50x1 mm)

Thorium-loaded ADS Exp. (2)


Thorium-loaded ADS Exp. - configurations


In (n,n') Reaction Rate Distribution Pulsed Neutron Method

Preliminary Results

• Detailed analysis of experiment results ongoing• Very low keff & low beam intensity at present; limited information gained so far• Increased accuracy & more evidence on Th-232 reactions by enhanced proton

beam is expected

Thank you for your attention.Thank you for your attention.

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Small PWR using Coated Particle FueL of Thorium and Plutonium, J. Nucl Sci. Technol., Vol.44, No.8, pp. 1045-1051(2007)

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A new innovative spherical cermet nuclear fuel element to achieve ultra-long core life for use in grid-appropriate LWRs)

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Small PWR “PFPWR50” using Cermet Fuel of Th-Pu Particles, ICAPP09, Paper ID-9363

Fuel composition ThO2:PuO2=9:1Pu LWR(50000MWD/t)

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Investigation on the feasibility of thorium breeder reactor in a BWRICAPP2010, Paper ID-10086

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Documents

添付資料： ワーキンググループにおける報告資料Canada 45,000 128,000 Australia 19,000 — South Africa 18,000 — Egypt 15,000 309,000 Other Countries 505,000 —

添付資料：ワーキンググループにおける報告資料Canada 45,000 128,000 Australia 19,000 — South Africa 18,000 — Egypt 15,000 309,000 Other Countries 505,000 —