IMPACT OF THORIUM BASED MOLTEN SALT REACTOR ON THE CLOSURE OF THE

NUCLEAR FUEL CYCLE

Missouri S&T

Nuclear Engineering Department

Safwan Jaradat

PhD Candidate10/22/2015

Outline• Introduction

• Objective

• MCNP Model FUJI-U3 Conclusion

• LFTR model

• Optimization

• Summary and Conclusion2

Introduction

Molten Salt Reactor (MSR)o selected by the Generation IV International Forum (GIF).oone of six innovative reactor concepts.

Liquid Fluoride Thorium Reactor (LFTR)o a type of MSRouses 232Th and 233U as the fertile and fissile materials,

respectively.o 233U and 232Th are dissolved in a mixed fluoride salt of

lithium and beryllium (FLiBe).

3

Historical Overview of MSRs

4

1954 : Aircraft Reactor Experiment (ARE). Power = 2.5 MWth, at (ORNL)

1964 : MSREPower: 8 MWth

1980s : JapanFUJI project

1971 : MSBRStopped-1976

2000s : Gen-IVLS-VHTR

1956 : TMSRMacPherson& his group

2010 : FHR DOE

Thorium Fuel Cycle

• What is the liquid fuel concepts of MSR?– Moderate melting temperature

at low vapor pressures.– High boiling temperature.– Good thermal properties (fuel = = coolant).– Stability under irradiation.– Good solubility of fissile and

fertile materials.– Less waste production of

isotopes hardly manageable.

The fluoride systems are the most recognized candidates for MSR fuels.

7LiF–BeF2– 232ThF4– 233UF4

Liquid Fluoride Thorium Reactor (LFTR).

5

Thorium Fuel Cycle

• Advantages of Liquid Thorium- Molten Salts

– It cannot meltdown (liquid fuel).

– Core can be emptied in an accident scenario.

– Safety, efficiency, and sustainability.

– Negligible production of Pu & minor actinides.

– Thorium is 3 times as abundant as Uranium.

– Supports online refueling.

6

Objective

To complete feasibility studies of a small commercial Liquid

Fluoride Thorium Reactor (LFTR) focused on neutronic calculations

in order to prescribe core design parameter such as core size, fuel

block pitch (p), fuel channel radius, fuel path, reflector thickness,

fuel salt composition, and power.

Approach: - Things to determine, eg., k-eff, flux, refueling, cycle length, etc.

- How to calculate these things? (MCNP) !! :p7

MCNP Model

Can MCNP gives comparable results to published work?!!

Can MCNP gives comparable results to published work?!!

Can MCNP gives comparable results to published work?!!

Well,

FUJI-U3-(0) model was verified using MCNP and compared

the results.

8

FUJI Reactor

• FUJI is a one kind of molten salt reactors that uses

molten thorium salt liquid fuel, which called Liquid

Fluoride Thorium Reactors (LFTR).

• Where 232Th plays as the fertile material, 233U as the

fissile material, and graphite as the moderator.

9

Core configuration of FUJI-U3-(0):Core 1 Core 2 Core 3

Δr (m) 1.16 0.80 0.40

Δh (m) 1.23 0.70 0.40

Fuel vol.% 0.39 0.27 0.45

Verification of FUJI-U3-(0) Reactor Model

FUJI-U3-(0) Design Conditions:-- Total power: 450 MWth (200 MWe)- Thermal efficiency: 44.4 %- Salt composition: 71.76% LiF – 16.0% BeF2 - 12.0% -ThF4 – 0.24% UF4

- Mean temperature: 630 °C (900 K)- Hastelloy-N: Ni/Mo/Cr/Fe/Nb/Si- Irradiation limits (to achieve 30-year of design life of graphite and avoid the replacement):-

1) Graphite moderator: 4.2*1013 (1/cm2. s)- fast neutrons > 52 keV2) Vessel: 1.4*1011 (1/cm2. s)- fast neutrons > 0.8 MeV 7.1*1012 (1/cm2. s)- thermal neutrons < 1.0 eV

FUJI-U3 Design parameters:- Reactor vessel: Diameter / Height (inner): 5.40 m/5.34 m Thickness: 0.05 m- Core: Diameter / Height : 4.72 m/4.66 m Fuel volume fraction (av.): 36 vol.%- Fuel path: Width: 0.038 m Fuel volume fraction 100 vol.%- Reflector: Thickness: 0.3 m Graphite volume fraction: 100 vol.%- Fuel salt: volume in reactor: 33.6 m3

volume in primary loop: 38.8 m3

- Inventory in primary loop: 233U : 1.133t* Th : 56.4t* Graphite : 163.1t- Hexagonal graphite: p=0.19 m

Verification of FUJI-U3-(0) Reactor Model

kinf vs. Graphite/U233

12

1.0E+02 1.0E+03 1.0E+04 1.0E+05 1.0E+060.6

0.7

0.8

0.9

1.0

1.1

1.2

1.3 MCNPFUJI-U3

Graphite/233U atom density ratio

k-in

finity

Radial Flux of Thermal Neutron at the Center of the Core

13

0.0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1.00.0

2.0

4.0

6.0

8.0

10.0

model 1model 2FUJI-U3

r/Rv

Ther

mal

neu

tron

flux

[101

3 /c

m2

.s]

th 1eV

Radial Flux of Fast Neutron at the Center of the Core

14

0.0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1.00.0

1.0

2.0

3.0

4.0

5.0

6.0

7.0

8.0

model 1model 2FUJI-U3

r/Rv

Fast

neu

tron

flux

[101

3 /c

m2

.s]

f 52keV

Irradiation limit

Axial Flux of Thermal Neutron at the Center of the Core

15

0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 10.0

2.0

4.0

6.0

8.0

10.0model 1model 2FUJI-U3

z/Hv

Ther

mal

neu

tron

flux

[101

3 /c

m2

.s]

th 1eV

Axial Flux of Fast Neutron at the Center of the Core

16

0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 10.0

1.0

2.0

3.0

4.0

5.0

6.0

7.0

8.0model 1model 2FUJI-U3

z/Hv

Fast

neu

tron

flux

[101

3 /c

m2

.s]

f 52keV

Irradiation limit

Time Behavior of keff

0.0 10.0 20.0 30.0 40.0 50.0 60.0 70.0 80.0 90.0 100.00.98

0.99

1

1.01

1.02

1.03

1.04model-1model-2

Burnup time (days)

keff

Model Time to k=1.01 (days)

Original Fuji model 40

Model 1 (our model) 40

Model 2 (our modified) 41

17

Compare Results

Model Keff CR αT

[1/K] (×10-5

)

ϕG[1/cm2s]>52KeV(×1013)

ϕv[1/cm2s]

>0.8 MeV(×1011)

<1.0 eV(×1012)

FUJI-U3 1.027 1.034 -3.10 4.10 1.34 2.46

Model-1 1.032 1.04 -5.01 3.53 0.80 3.13

Model-2 1.034 1.04 -5.06 3.46 0.88 3.37

18

Conclusion

19

• A verification for FUJI-U3-(0) was conducted.

• MCNP code was used to study the reactor physics characteristics for the

FUJI-U3.

• The results were comparable with each other.

• Based on the that, MCNP was found to be a reliable code to model a small

liquid fluoride thorium reactor LFTR .

LFTR Model

How did we choose starting specification?!!

Based on FUJI, but not FUJI because:

- Simple single-region core.

- Small size.

- Hexagonal fuel block.

- Refueling process.

- MCNP.

Why small size?- Ease of construction and factory fabrication.

- Ease of transportation and shipment globally.

- For use where large reactors are not ideal, e.g, micro-grids.

20

LFTR’s Design Strategy

A series of survey calculations were conducted using MCNP6 to obtain the

conceptual core.

The calculations started by determining the candidate fuel composition with a

(233U/232Th)% that would achieve the minimal change of reactivity.

Widely changing parameters, including core size, hexagonal graphite pitch (p),

fuel channel radius, fuel path, reflector graphite thickness, and expected power

level, etc.

The calculations ended with a full-scale reactor core with a power of 150 MWth.

21

k-Infinity Geometry and Calculations

Different fuel compositions of different (233U/232Th) % were examined in order to

find the proper ratio that would achieve the minimum change of reactivity.

A single fuel rod was modeled with specular reflectors to eliminate the leakage of

neutrons.

The fuel channel is a cylindrical bore through a hexagonal graphite moderator prism.

22

Different Fuel Salt Compositions

23

Fuel Salt Composition (mol. %)7LiF - BeF2 - ThF4 - UF4

Melting Temperatur

e (°C)

Density (g/cc)at T=900K

Atom Ratios (233U/232Th) × 100%

60.00 – 38.00 – 1.00 – 1.00 442 2.197 100.43

63.00 – 35.50 – 1.00 – 0.50 456 2.140 50.22

65.00 – 30.00 – 4.00 – 1.00 448 2.548 25.11

65.00 – 30.50 – 4.00 – 0.50 453 2.492 12.55

71.76 – 16.00 – 12.0 – 0.24 457 3.330 2.01

Different Fuel Salt Composition

• It is desirable for these kinds of reactors to have relatively small mole

fractions of 233U to keep the physical properties of the diluents under control.

• The difficulty in conducting experiments to get the physical and chemical

information for every fuel composition.

• The densities were calculated using the rule of additivity of molar volumes.

• Carefully transformed the molar ratios into weight fractions to be used in the

MCNP material card.

24

kinf vs. Graphite/U233 For Compositions

25

1.0E1 1.0E2 1.0E3 1.0E4 1.0E5 1.0E6 1.0E70.6

0.8

1

1.2

1.4

1.6

1.8

2

2.2

2.4

2.01% 12.55% 25.11%50.22% 100.43%

Graphite/233U atom density ratio

kinf

kinf vs. time for compositions

26

0.0E+00 2.0E+02 4.0E+02 6.0E+02 8.0E+02 1.0E+03 1.2E+030.9

1.1

1.3

1.5

1.7

1.9

2.1

2.3

2.01% 12.55% 25.11%50.22% 100.43%

Burnup time (days)

kinf

Full-Scale of a Small LFTR

Small LFTR Design Conditions:-- Total power: 150 MWth (50 - 66 MWe)- Thermal efficiency: (33.0 % - 44.0 %)- Salt composition: 71.76% LiF – 16.0% BeF2 - 12.0% -

ThF4 – 0.24% UF4

- Mean temperature: 630 °C (900 K)- Hastelloy-N: Ni/Mo/Cr/Fe/Nb/Si

LFTR Design parameters:- Reactor vessel: Diameter / Height (inner): 3.30 m/3.10 m Thickness: 0.05 m-Core: Diameter / Height : 2.80 m/2.60 m Number of fuel channels: 91 Fuel volume fraction (av.): 17 vol.%- Fuel path: Width: 0.07 m- Reflector: Thickness: 0.23 m- Hexagonal graphite: p=0.26 m- Flow-hole radius: r=variable

kinf vs. graphite/U233 of LFTR

1.0E2 1.0E3 1.0E4 1.0E5 1.0E60.6

0.7

0.8

0.9

1

1.1

1.2

1.3

Graphite/U233 atom density ratio

kinf

28

kinf vs. graphite/U233 of LFTR

29

Temperature(due to fission) # Density of Gr (Gr/233U) %

Reduce thermalizedneutrons Fission rate

Temperature K-infinity Safety

Neutron Energy Spectrum In a Unit Cell

30

1E-9 1E-8 1E-7 1E-6 1E-5 1E-4 1E-3 1E-2 1E-1 1E+0 1E+10.0E+00

5.0E-05

1.0E-04

1.5E-04

2.0E-04

2.5E-04

3.0E-04

3.5E-04

4.0E-04Fuel ChannelGraphite Moderator

Energy (MeV)

Flux

per

uni

t let

harg

y (A

rbitr

ary

Uni

t)

B A

22 eV1.26 eV

MCNP6 TiersIn the “Burn” card there are three built-in “Tiers” of fission products available to the user.

The default one is Tier 1 with the main common 12 fission products, Tier 2 has 87 fission

products, and in Tier 3 all isotopes contained in the fission product.

0.0 100.0 200.0 300.0 400.0 500.0 600.0 700.0 800.00.95

1

1.05

1.1

1.15

1.2

1.25Tier-1Tier-2Tier-3

Burnup time (Days)

kinf

31

Time Behavior of keff of LFTR

0.0 20.0 40.0 60.0 80.0 100.0 120.0 140.0 160.0 180.0 200.00.94

0.96

0.98

1

1.02

1.04

1.06

1.08

Burnup time (days)

keff

32

Radial Flux of Thermal Neutron at the Center of the Core

0.0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1.00.0

0.5

1.0

1.5

2.0

2.5

r/Rv

Ther

mal

neu

tron

flux

[101

4 /c

m2

.s]

th 1eV

33

Radial Flux of Fast Neutron at the Center of the Core

0.0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1.00.0

0.4

0.8

1.2

1.6

2.0

r/Rv

Fast

neu

tron

flux

[101

4 /c

m2

.s] f 52keV

34

Axial Flux Distribution of Thermal Neutrons

35

0.0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1.0

-175

-125

-75

-25

25

75

125

175

x1=0.5 cm x2=13.6 cm x3=104.1 cm x4=116.2 cm x5=174.5 cm

Normalized axial flux

Heig

ht (c

m)

Graphite GraphiteFuel Fuel

Hastelloy-N

x5 x3 x4 x1

x2

Axial Flux Distribution of Fast Neutrons

36

0.0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1.0

-175

-125

-75

-25

25

75

125

175

x1=0.5 cm x2=13.6 cm x3=104.1 cm x4=116.2 cm x5=174.5 cm

Normalized axial flux

Heig

ht (c

m) x5 x4 x3 x2

x1

Thermal Flux Distribution ϕth < 1 eV

37

Max/Avg= 1.87

Thermal Flux Distribution ϕth < 1 eV

38

Fast Flux Distribution ϕf > 52 keV

39

Max/Avg= 2.78

Fast Flux Distribution ϕf > 52 keV

40

Total Flux Distribution ϕtotal

41

Max/Avg= 1.68

Total Flux Distribution ϕtotal

42

Burn-up Calculations of LFTR

43

0 200 400 600 800 1,000 1,200 1,400 1,600 1,800 2,0000.95

0.97

0.99

1.01

1.03

1.05

1.07

1.09

Time (days)

keff

300 d 510 d 530 d 540 d

25 kg of 233U 27 kg of 233U 29 kg of 233UFed 233U as

7LiF – 233UF4

(73 - 27) mol%

Frozen eutectic salt

Removed

He

Kr

Xe

Removed FP gases

Removed FP gases

Removed FP gases

300 d 810 d 1340 d 1880 d

Phase Diagram Equilibria of Binary LiF-UF4

44Reference: C. F. Weaver et al., "phase equilibria in molten salt breeder reactor fuels", ORNL-2896, Des 27 1960.

Time Behavior of LFTR Characteristics

OperationPeriod(EFPD)

Keff CR Fission/Fertile %

αT

[1/K] (×10-5

)

0290

1.0711.002

0.00.77

0.0201 -2.83

300800

1.0701.004

1.240.84

0.0227 -2.39

8101330

1.0701.003

1.140.81 0.0244 -1.58

13401880

1.0711.001

1.130.78 0.0260 -2.79

45

Production Paths of Fissile 233U

46

Time Behavior of Conversion Ratio

0 200 400 600 800 1000 1200 1400 1600 1800 20000

0.2

0.4

0.6

0.8

1

1.2

1.4

Burnup time (Days)

Conv

ersi

on R

atio

20 days

47

U233 Fission XS Vs. Th232 Absorption XS

48

233Pa Mass Production With Burnup Time

49

0 200 400 600 800 1000 1200 1400 1600 1800 20000

0.2

0.4

0.6

0.8

1

1.2

1.4

0

1000

2000

3000

4000

5000

6000

7000

CR Mass Pa233

Burnup time (days)

Conv

ersi

on R

atio 233Pa mass (gm

)

Material Balance of LFTR For 5 Years Operation

Th232

(ton)Ufis+233Pa

(ton)Pu(g)

MA(g)

All FP(kg)

Gas FP(kg)

Initial inventory 7.644 0.154 --- --- --- ---

Total net feed --- 0.081 --- --- --- ---

Total demand 7.644 0.235 --- --- --- ---

Final remain 7.380 0.172 7.63 34.5 294.3 ---

Net production - 0.264 - 0.063 7.63 34.5 294.3 7.1

50

Fuel Salt Composition to the End of Run

Burnup Up(days)

LiF(mol%)

BeF2(mol%)

ThF4(mol%)

UF4(mol%)

Other elements

0 71.76 16.0 12.0 0.24 0.0

300 71.80 16.0 11.91 0.26 0.03

810 71.81 15.96 11.78 0.28 0.17

1340 71.81 15.93 11.65 0.29 0.32

1880 71.88 15.95 11.55 0.26 0.36

51

In order to increase the cycle length of burnup, the radii of the fuel rods

at the outer rings of the LFTR core were increased while keeping the total

mass/volume of the fuel inside the core fixed. Thus, the radii of the fuel rods

at the inner rings of the core were decreased. A lot of scenarios with different

radii were conducted.

52

Optimization

Optimized LFTR Core

53

Keff vs. Time

54

0.0 50.0 100.0 150.0 200.0 250.0 300.0 350.00.94

0.96

0.98

1

1.02

1.04

1.06

1.08

1.1

Optimization of LFTRLFTR

Burnup time (days)

keff

Thermal Neutron Flux

55

0.0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1.00.0

0.5

1.0

1.5

2.0

2.5

Optimization of LFTRLFTR

r/Rv

Ther

mal

neu

tron

flux

[101

4 /c

m2

. s]

Fast Neutron Flux

56

0.0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1.00.0

0.2

0.4

0.6

0.8

1.0

1.2

1.4

1.6

1.8

Optimization of LFTRLFTR

r/Rv

Fast

neu

tron

flux

[101

4 /c

m2

.s]

Total Neutron Flux

57

0.0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1.00.0

1.0

2.0

3.0

4.0

5.0

6.0

Optimization of LFTRLFTR

r/Rv

Tota

l neu

tron

flux

[101

4 /c

m2

.s]

Summary and Conclusion

In this dissertation, a complete feasibility studies of a

conceptual small thermal commercial liquid fluoride thorium

reactor LFTR design, has been demonstrated. The core performance

and the burnup analysis were obtained using MCNP6 code. The

results were promising and the main outcomes obtained are as

follows:

• The reactor can be operated for five years at a thermal power

level of 150 MWth together with a load factor of 100% with an

initial inventory of fissile material 233U of 0.154 (ton).

58

Summary and Conclusion

• The total net feed of 233U-fissile was 0.081 (ton). At the end of

reactor operation, 0.172 (ton) was the final remain of fissile

material.

• The average fuel conversion ratio CR was 0.78.

• The temperature coefficient of reactivity at the beginning of

operation (t=0) was -2.83×10-5 / T.

59

Summary and Conclusion

• The reactor produced 7.63 (g) of Pu for a 5 years of operation.

• 89.84% of the produced Pu was 238Pu (with a half-life 87.7 years).

• The production of minor actinide (MA) was 34.5 (g) with mostly

237Np and 238Np, and no Am or Cm were produced during the

burnup time.

• The first cycle length of burnup was increased 40 days by

optimized the reactor core.60

61