Defense 10/22/2015

1. IMPACT OF THORIUM BASED MOLTEN SALT
REACTOR ON THE CLOSURE OF THE
NUCLEAR FUEL CYCLE
Missouri S&T
Nuclear Engineering Department
Safwan Jaradat
PhD Candidate
10/22/2015

2. Outline
• Introduction
• Objective
• MCNP Model
 FUJI-U3
 Conclusion
• LFTR model
• Optimization
• Summary and Conclusion
2

3. Introduction
Molten Salt Reactor (MSR)
o selected by the Generation IV International Forum (GIF).
o one of six innovative reactor concepts.
Liquid Fluoride Thorium Reactor (LFTR)
o a type of MSR
o uses 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

4. Historical Overview of MSRs
4
1954 : Aircraft Reactor Experiment
(ARE). Power = 2.5 MWth, at (ORNL)
1964 : MSRE
Power: 8 MWth
1980s : Japan
FUJI project
1971 : MSBR
Stopped-1976
2000s : Gen-IV
LS-VHTR
1956 : TMSR
MacPherson
& his group
2010 : FHR
DOE

5. 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

6. 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

7. 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) !! :p
7

8. 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

9. 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

10. 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 keV
2) Vessel:
1.4*1011 (1/cm2. s)- fast neutrons > 0.8 MeV
7.1*1012 (1/cm2. s)- thermal neutrons < 1.0 eV

11. 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

12. kinf vs. Graphite/U233
12
0.6
0.7
0.8
0.9
1.0
1.1
1.2
1.3
1.0E+02 1.0E+03 1.0E+04 1.0E+05 1.0E+06
k-infinity
Graphite/233U atom density ratio
MCNP
FUJI-U3

13. RadialFluxofThermalNeutronattheCenteroftheCore
13
0.0
2.0
4.0
6.0
8.0
10.0
0.0 0.2 0.4 0.6 0.8 1.0
Thermalneutronflux[1013/cm2.s]
r/Rv
model 1
model 2
FUJI-U3th
1eV 

14. RadialFluxofFastNeutronattheCenteroftheCore
14
0.0
1.0
2.0
3.0
4.0
5.0
6.0
7.0
8.0
0.0 0.2 0.4 0.6 0.8 1.0
Fastneutronflux[1013/cm2.s]
r/Rv
model 1
model 2
FUJI-U3
f
52keV 
Irradiation limit

15. AxialFluxofThermalNeutronattheCenteroftheCore
15
0.0
2.0
4.0
6.0
8.0
10.0
0 0.2 0.4 0.6 0.8 1
Thermalneutronflux[1013/cm2.s]
z/Hv
model 1
model 2
FUJI-U3th
1eV 

16. AxialFluxofFastNeutronattheCenteroftheCore
16
0.0
1.0
2.0
3.0
4.0
5.0
6.0
7.0
8.0
0 0.2 0.4 0.6 0.8 1
Fastneutronflux[1013/cm2.s]
z/Hv
model 1
model 2
FUJI-U3
f
52keV 
Irradiation limit

17. Time Behavior of keff
0.98
0.99
1
1.01
1.02
1.03
1.04
0.0 20.0 40.0 60.0 80.0 100.0
keff
Burnup time (days)
model-1
model-2
Model Time to k=1.01 (days)
Original Fuji model 40
Model 1 (our model) 40
Model 2 (our modified) 41
17

18. 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

19. 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 .

20. 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

21. 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

22. 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

23. Different Fuel Salt Compositions
23
Fuel Salt Composition (mol. %)
7LiF - BeF2 - ThF4 - UF4
Melting
Temperature
(°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

24. 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

25. kinf vs. Graphite/U233 For Compositions
25
0.6
0.8
1
1.2
1.4
1.6
1.8
2
2.2
2.4
1.0E0 1.0E1 1.0E2 1.0E3 1.0E4 1.0E5 1.0E6 1.0E7
kinf
Graphite/233U atom density ratio
2.01% 12.55% 25.11% 50.22% 100.43%

26. kinf vs. time for compositions
26
0.9
1.1
1.3
1.5
1.7
1.9
2.1
2.3
0.0E+00 2.0E+02 4.0E+02 6.0E+02 8.0E+02 1.0E+03 1.2E+03
kinf
Burnup time (days)
2.01% 12.55% 25.11% 50.22% 100.43%

27. 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

28. kinf vs. graphite/U233 of LFTR
0.6
0.7
0.8
0.9
1
1.1
1.2
1.3
1.0E2 1.0E3 1.0E4 1.0E5 1.0E6
kinf
Graphite/U233 atom density ratio
28

29. kinf vs. graphite/U233 of LFTR
29
Temperature
(due to fission)
# Density of Gr (Gr/233U) %
Reduce thermalized
neutrons
Fission rate
Temperature K-infinity  Safety

30. Neutron Energy Spectrum In a Unit Cell
30
0.0E+00
5.0E-05
1.0E-04
1.5E-04
2.0E-04
2.5E-04
3.0E-04
3.5E-04
4.0E-04
1E-9 1E-7 1E-5 1E-3 1E-1 1E+1
Fluxperunitlethargy(ArbitraryUnit)
Energy (MeV)
Fuel Channel
Graphite Moderator
B A
22 eV1.26 eV

31. MCNP6 Tiers
In 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.
0.95
1
1.05
1.1
1.15
1.2
1.25
0.0 200.0 400.0 600.0 800.0
kinf
Burnup time (Days)
Tier-1
Tier-2
Tier-3
31

32. Time Behavior of keff of LFTR
0.98
0.99
1
1.01
1.02
1.03
1.04
1.05
1.06
1.07
1.08
0.0 50.0 100.0 150.0 200.0
keff
Burnup time (days)
32

33. RadialFluxofThermalNeutronattheCenteroftheCore
0.0
0.5
1.0
1.5
2.0
2.5
0.0 0.2 0.4 0.6 0.8 1.0
Thermalneutronflux[1014/cm2.s]
r/Rv
th
1eV 
33

34. RadialFluxofFastNeutronattheCenteroftheCore
0.0
0.4
0.8
1.2
1.6
2.0
0.0 0.2 0.4 0.6 0.8 1.0
Fastneutronflux[1014/cm2.s]
r/Rv
f
52keV 
34

35. AxialFluxDistributionofThermalNeutrons
35
-175
-125
-75
-25
25
75
125
175
0.0 0.2 0.4 0.6 0.8 1.0
Height(cm)
Normalized axial flux
x1=0.5 cm x2=13.6 cm x3=104.1 cm x4=116.2 cm x5=174.5 cm
Graphite GraphiteFuel Fuel
Hastelloy-N
x5 x3 x4 x1
x2

36. AxialFluxDistributionofFastNeutrons
36
-175
-125
-75
-25
25
75
125
175
0.0 0.2 0.4 0.6 0.8 1.0
Height(cm)
Normalized axial flux
x1=0.5 cm x2=13.6 cm x3=104.1 cm x4=116.2 cm x5=174.5 cm
x5 x4 x3 x2
x1

37. Thermal Flux Distributionϕth < 1 eV
37
Max/Avg= 1.87

38. Thermal Flux Distributionϕth < 1 eV
38

39. Fast Flux Distribution ϕf > 52 keV
39
Max/Avg= 2.78

40. Fast Flux Distribution ϕf > 52 keV
40

41. Total Flux Distribution ϕtotal
41
Max/Avg= 1.68

42. Total Flux Distribution ϕtotal
42

43. Burn-up Calculations of LFTR
43
0.99
1
1.01
1.02
1.03
1.04
1.05
1.06
1.07
1.08
1.09
0 500 1,000 1,500 2,000
keff
Time (days)
300 d 510 d 530 d 540 d
25 kg of 233U 27 kg of 233U 29 kg of 233U
Fed 233U as
7LiF – 233UF4
(73 - 27) mol%
Frozen
eutectic salt
Removed
He
Kr
Xe
300 d 810 d 1340 d 1880 d

44. Phase Diagram Equilibria of Binary LiF-UF4
44
Reference: C. F. Weaver et al., "phase equilibria in molten salt breeder reactor fuels", ORNL-2896, Des 27 1960.

45. Time Behavior of LFTR Characteristics
Operation
Period
(EFPD)
Keff CR
Fission/Fertile
%
αT
[1/K] (×10
-5
)
0
290
1.071
1.002
0.0
0.77
0.0201 -2.83
300
800
1.070
1.004
1.24
0.84
0.0227 -2.39
810
1330
1.070
1.003
1.14
0.81
0.0244
-1.58
1340
1880
1.071
1.001
1.13
0.78
0.0260
-2.79
45

46. Production Paths of Fissile 233U
46

47. Time Behavior of Conversion Ratio
0
0.2
0.4
0.6
0.8
1
1.2
1.4
0 500 1000 1500 2000
ConversionRatio
Burnup time (Days)
20 days
47

48. U233 Fission XS Vs. Th232 Absorption XS
48

49. 233Pa Mass Production With Burnup Time
49
0
1000
2000
3000
4000
5000
6000
7000
0
0.2
0.4
0.6
0.8
1
1.2
1.4
0 500 1000 1500 2000
233Pamass(gm)
ConversionRatio
Burnup time (days)
CR Mass Pa233

50. MaterialBalanceofLFTR For5 YearsOperation
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

51. 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

52. 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

53. Optimized LFTR Core
53

54. Keff vs. Time
54
0.98
0.99
1
1.01
1.02
1.03
1.04
1.05
1.06
1.07
1.08
0.0 50.0 100.0 150.0 200.0 250.0 300.0 350.0
keff
Burnup time (days)
Optimization of LFTR
LFTR

55. Thermal Neutron Flux
55
0.0
0.5
1.0
1.5
2.0
2.5
0.0 0.2 0.4 0.6 0.8 1.0
Thermalneutronflux[1014/cm2.s]
r/Rv
Optimization of LFTR
LFTR

56. Fast Neutron Flux
56
0.0
0.2
0.4
0.6
0.8
1.0
1.2
1.4
1.6
1.8
0.0 0.2 0.4 0.6 0.8 1.0
Fastneutronflux[1014/cm2.s]
r/Rv
Optimization of LFTR
LFTR

57. Total Neutron Flux
57
0.0
1.0
2.0
3.0
4.0
5.0
6.0
0.0 0.2 0.4 0.6 0.8 1.0
Totalneutronflux[1014/cm2.s]
r/Rv
Optimization of LFTR
LFTR

58. 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

59. 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

60. 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. 61