1. ThEC13 Daniel Mathers daniel.p.mathers@nnl.co.uk
The Thorium Fuel Cycle
ThEC13
Daniel Mathers

2. Outline Content: Background
Sustainability, proliferation resistance, economics, radiotoxicity
Advantages and disadvantages
Fuel Cycles

3. Fertile thorium A Thorium Fuel Cycle needs Uranium or Plutonium
Th-232 is the only naturally occurring thorium nuclide
It is a fertile nuclide that generates fissile U-233 on capturing a neutron
Th-232 is fissionable in that it fissions on interacting with fast neutrons > 1 MeV kinetic energy
Fertile conversion occurs with thermal neutron captures:
Th-232 (n,g) Th-233 (b-) Pa-233 (b-) U-233
U-233 has a high thermal fission cross-section and a low thermal neutron capture cross-section
The fission/capture ratio for U-233 is higher than the other major fissile nuclides U-235, Pu-239 and Pu-241
This is very favourable for the neutron multiplication factor and minimises the probability of neutron captures leading to transuranics
A Thorium Fuel Cycle needs Uranium or Plutonium
to initiate a fission reaction

4. Thorium fuel cycles Options for a thermal reactor are:
Once-through fuel cycle with Th-232 as alternative fertile material to U-238 with U-235 or Pu-239 driver
U-233 fissioned in-situ without reprocessing/recycle
Modest reduction in uranium demand and sustainability
Recycle strategy with reprocessing/recycle of U-233
Much improved sustainability analogous to U/Pu breeding cycle
But some technical difficulties to overcome
Options for a fast reactor are:
MSFR and other Gen IV concepts (Sodium cooled fast reactors, ADS systems)
All require U / Pu to initiate fission reaction
Neutron balance in a thermal reactor
For U-235 fissions average number of fission neutrons n ~ 2.4
When a neutron causes a fission it is removed from the system, which accounts for 1 of the 2.4 available neutrons
Unavoidable neutron losses occur due to neutron captures in the fuel, structural materials, moderator, control materials and surface leakage and amount to about 0.7 to 0.8 neutrons in current thermal reactor designs
Leaving about 0.6 to 0.7 neutrons available for fertile captures of U-238 to Pu-239
U-235 thermal reactors are therefore limited to a conversion ratio of ~ and cannot operate as breeder reactors
Thermal breeder
The neutron balance for U-233 is more favourable (because neutron captures in U-233 are very low) and this gives the possibility of a conversion ratio close to 1.0 in a thermal reactor
This is a breeding system
Not quite ideal, but much better than other thermal reactors
Not possible with U/Pu fuel cycle
The possibility of a thermal breeder reactor was the driver for the thorium fuel cycle in the 1950s
With uranium ore expected to be scarce
Fast breeder technology more difficult to realise

5. Sustainability / Inherent Proliferation Resistance
Thorium abundance higher than uranium
Thorium demand lower because no isotopic enrichment
Thorium economically extractable reserves not so well defined
Rate of expansion of thorium fuel cycle will be limited by the slow conversion rate
Inherent proliferation resistance
U-233 is a viable weapons usable material
High U-232 inventory implies high doses unless shielded
Low inherent neutron source suggests that U-233 weapon design may be simplified and potentially more accessible
U-233 fissile quality hardly changes with irradiation

6. Economics and Radiotoxicity
U-233 recycle has lower demand on thorium than uranium because there is no isotopic enrichment process
U-233 recycle potentially reduces the ore procurement cost and eliminates the enrichment cost
Future uranium and thorium market prices unknown
Short term economic barrier presented by need for R&D to demonstrate satisfactory fuel performance
Radiotoxicity
Spent fuel activity/radiotoxicity dominated by fission products for 500 years after discharge
U/Pu long term fuel activity determined by activity of Np, Pu, Am and Cm
Th/U-233 long term fuel activity has only trace quantities of transuranics and therefore lower radiotoxicity after 500 years
However, this only applies to the long term equilibrium condition with self-sustained U-233 recycle
In a practical scenario, the reduction in radiotoxicity is more modest than the long term equilibrium would indicate
It is too soon to say whether the thorium fuel cycle will be economically advantageous
Need to compare radiotoxicity over range of timeframes

7. Advantages of Th fuel cycle
Thorium more abundant than uranium and combined with a breeding cycle is potentially a major energy resource
Low inventories of transuranics and low radiotoxicity after 500 years’ cooling
Almost zero inventory of weapons usable plutonium
Theoretical low cost compared with uranium fuel cycle
ThO2 properties generally favourable compared to UO2 (thermal conductivity; single oxidation state)
ThO2 is potentially a more stable matrix for geological disposal than UO2

8. Void coefficient mitigation
Supplementing a U/Pu recycle strategy
Thorium fuels drive the void coefficient more negative in thermal and fast systems
A positive void coefficient is an undesirable in-core positive feedback effect unless counteracted by other feedback effects
In LWRs a positive void coefficient is usually considered unacceptable and limits the total plutonium load in MOX fuel to <12 w/o
This is a potential restriction with poor fissile quality plutonium
Thorium-plutonium fuel could allow significantly higher total plutonium loads (up to ~18 w/o), giving more flexibility for plutonium re-use in LWRs
I have a yet to be published paper from Cambridge University which indicates that the void coefficient becomes more negative when U-Pu MOX is replaced with Th-Pu MOX. The paper examines Reduced Moderation PWR (RM-PWR), Reduced Moderation BWR (RM-BWR) and SFR and shows a more negative void coefficient in all of them. Mike Thomas has shown the same for PWR, a result which is consistent with AREVA’s work.
THIS BIT IS JUST FOR MY OWN SATISFACTION:
Why does thorium have this effect? Th-232 is a resonant nuclide with a large double resonance centred at about 22 eV (see slide at end of this pack). Voiding the water density in an LWR shifts the neutron spectrum towards this resonance and increases the proportion of slowing down neutrons that are removed by resonant captures. The same effect occurs with U-238 (very large and narrow lowest resonance at 6.5 eV), but the resonance peak in Th-232 is broader and lower than the lowest U-238 resonance, possibly leading to reduced self-shielding in Th-232 and thereby affecting the void coefficient. In other words, because U-238 is more heavily self-shielded, when voidage is introduced all the neutrons in the range of the lowest U-238 resonance are already removed by the resonance, the change in 1/E scattering source has limited impact. In Th-232 there is less self-shielding and therefore the change in 1/E scattering source is felt more strongly.
EXTRACT FROM KIM & DOWNAR PAPER: One of the most important differences in the two isotopes is that the resonance integral of Th232 is about 3 times smaller than that of U238. This has an important implication for the void reactivity in a tight pitch lattice as will be discussed later. It is important to note that despite the large differences in the resonance integral, the effective resonance capture of both fertile fuels is comparable because resonance self -shielding reduces the effective resonance capture of U238. Thorium-based fuel designs also have the potential of enhanced safety characteristics in intermediate spectrum designs. Because of the reduced resonance integral and the higher fast fission threshold in Th232 compared to U238, the void coefficient of the thorium pins is larger and negative in the tight lattice cases
A Possible way to manage plutonium stocks with poorer fissile quality and to allow time for thorium plutonium MOX qualification

9. Radiotoxicity
Thorium-plutonium MOX fuel theoretically could be advantageous for UK plutonium disposition
Detailed assessment by NNL of decay heat load and radiotoxicity per GWye shows there is only a marginal difference between Th-Pu MOX and U-Pu MOX
This is a holistic calculation that accounts for the total decay heat outputs of different scenarios
In the Th-Pu and U-Pu MOX cases, the decay heat is concentrated in the MOX assemblies, whereas in the UO2 reference case it is distributed over a larger number of UO2 assemblies

10. Decay heat Model uses standard PWR 17x17 assembly
The Decay heat is per Gwye, therefore it is a normalised figure i.e. decay heat from assemblies (of these three types) irradiated in a reactor that produces 1GW of energy per year

11. Many technological issues to address - MSR is a long term option
Molten salt reactor
Molten Salt Reactor (MSR)
Generation IV International Project is researching MSR
Gen IV MSR will be a fast spectrum system
Molten salt fuel circulates through core and heat exchangers
On-line reprocessing to remove fission products
Ideally suited to thorium fuel as fuel fabrication is avoided
Equilibrium fuel cycle will have low radiotoxicity (fission products only)
Many technological issues to address - MSR is a long term option

12. Accelerator driven system
Accelerator driven system (ADS)
Sub-critical reactor core
Proton beam provides neutron source in spallation target
Neutron source multiplied by sub-critical core

13. Disadvantages of Th fuel cycle
Th-232 needs to be converted to U-233 using neutrons from another source
Neutrons are expensive to produce
The conversion rate is very low, so the time taken to build up usable amounts of U-233 are very long
Reprocessing thorium fuel is less straightforward than with the uranium-plutonium fuel cycle
The THOREX process has been demonstrated at small scale, but will require R&D to develop it to commercial readiness
U-233 recycle is complicated by presence of ppm quantities of U-232 (radiologically significant for fuel fabrication operations at ppb levels)
U-233 is weapons useable material with a low fissile mass and low spontaneous neutron source
U-233 classified by IAEA in same category as High Enriched Uranium (HEU) with a Significant Quantity in terms of Safeguards defined as 8 kg compared with 32 kg for HEU

14. R&D requirements Fuel materials properties Fuel irradiation behaviour
THOREX reprocessing
Waste management /disposal
U-233 fuel fabrication
Systems development
Scenario modelling
Systems development
Engineering design
Materials
Liquid salt chemistry and properties
MSR fuel and fuel cycle
Systems integration and assessment
Safety

15. Fuel cycle scenario modelling
Fuel cycle simulation computer programs are used to assess the impacts that different fuel cycle scenarios may have on:
Uranium or Thorium ore requirements,
Time and resources needed to create sufficient fertile material to start a Thorium ‘only’ reactor
Ability to start a sustainable fast reactor fleet,
Time at which feed of natural uranium is no longer required
Packing density and inventory of a geological repository
The practicalities of handling fresh nuclear fuel
Processing of spent nuclear fuel
Requirements for high level waste immobilisation technologies
Fuel Cycle codes (e.g. NNL’s ORION code) are capable of assessing the impacts of the alternative U, Th based scenarios 15

16. Fuel cycle
Building up a fleet with the aim of reducing dependency on U/Pu will take time.
Reactor doubling time is an important consideration
Some contention that alternative systems might give a different result
But these underlying equations give confidence that the same limitations will apply to all workable systems

17. Relevance to thorium Long doubling times are relevant to:
Initial build-up of U-233 inventory to get thorium fuel cycle to equilibrium
For practical systems this timescale is very long and this will govern strategic analysis of transition to thorium fuel cycle using enriched uranium or plutonium/transuranic fuels
Important for strategic assessments to account for impact of transition effects
Subsequent expansion of thorium reactor fleet and rate at which thorium systems can expand to meet increasing demand
This is a message I’ve been preaching for a long time, that transition effects in the fuel cycle are important.

18. Breeding ratio The breeding ratio (BR) is defined as:
Mass of fissile material produced by fertile neutron captures
Mass of fissile material consumed
EXAMPLE:
1 GWth breeder reactor operating at 90% load factor would consume approximately 330 kg of fissile material per year – equivalent to 1 kg per full power day
If a breeder reactor produces 1.3 kg of new fissile material by fertile captures per full power day, the breeding ratio is 1.3/1.0 = 1.3 and the breeding gain (BG) defined as BG = BR-1 is ( )/1.0 = +0.3
This is the standard textbook definition.

19. TD (full power days) = [MC + MO]/g
Doubling time
This is the time in which a breeder reactor would take to generate enough surplus fissile material to start off an identical reactor system
The doubling time (TD) is the time needed to replace the total fissile inventory of the core MC (kg) plus the out of core fissile inventory MO (kg)
For a system which consumes m kg of fissile material and has a net gain g kg of fissile material per full power day, the doubling time is:
TD (full power days) = [MC + MO]/g
= [MC + MO]/[(BR-1).m]
= [MC + MO]/BG.m
GOVERNING PARAMETERS:
m is governed by the thermal power output only – 1 kg per full power day for 1 GWth output
[MC + MO] and BG are dependent on the specific reactor design
[MC + MO] typically a few thousand kg
Large positive BG very difficult to achieve and 0.3 to 0.4 is about the highest claimed for any system
This is again the standard textbook definition.
TD (full power days) = [MC + MO]/BG.m
The bottom line is that the numerator is several thousand kg and the denominator is 1 kg multiplied by BG < 0.4
Therefore doubling time is at best several thousand full power days and in practice longer still
This applies generally to thermal and fast breeders and irrespective of the fissile nuclide (U-233, U-235 or Pu-239)

20. Application to MSR THERMAL SPECTRUM MSR FAST SPECTRUM MSR
Based on simplistic scale-up of ORNL Molten Salt Reactor Experiment:
1.0 GWth; m = 1.0 kg/full power day; MC = 1500 kg U-233; MO = 3000 kg U-233; BG = (estimated)
TD = [MC + MO]/BG.m = (4500/0.06 x 1.0) = full power days (200 full power years)
Probable scope for optimisation, but doubling time still likely to be very long
FAST SPECTRUM MSR
Based on Delpech/Merle-Lucotte et al TMSR-NM (non-moderated thorium molten salt reactor) core:
2.5 GWth; m = 2.5 kg/full power day; MC+MO = 5700 kg; BG = +0.12
TD = [MC + MO]/BG.m = (5700/0.12 x 2.5) = full power days (52 full power years)
These are just illustrative examples. Thermal spectrum MSR has a reduced core mass, but then BG is lower as well and the result is just marginal breeding gain and very long doubling time. Fast spectrum MSR benefits from a larger breeding gain, but the core mass is higher.
It is possible for a higher BG if you reprocess the salt more often (less fission products absorbing neutrons used to Breed), or add more Pu in place of U-233 into the core.

21. Hypothetical profile of installed capacity versus time for a breeder system
Point A: Fast reactors first become available
Segment AB: Fast reactors deployed using separated plutonium stock
Point B: Initial deployment using available plutonium stocks ends
Segment BC: Fast reactors deployed limited by doubling time
Point C: Maximum capacity achieved
Segment CD: Steady state operation in self sufficient mode
Point D: Fast reactors converted to burners
Segment DE: Fissile inventory drawn down to zero during planned phase-out
ORION scenarios are necessarily tied to a specific set of reactor systems
This leads to possible challenge that alternative systems might give a different result
But the underlying equations give confidence that the same limitations will apply to all workable systems
A kind of a 2nd Law of Thermodynamics for breeder reactors

22. Reactor parameters Parameter Value Units Unit size 1.6 GWe
Initial core fissile loading
10.0
tHM
Dwell time
4.0
years
Recycle time
5.0
Net breeding gain (in breeding mode)
+0.3 -
Net breeding gain (in self-sufficient mode)
0.0
Net breeding gain (in burner mode)
-0.40
Earliest fast reactor deployment
2040
Maximum fast reactor capacity
22.4

23. Generating capacity - transition from LWRs to fast reactors
This assumes a Breeding Gain of or 5.8%
If you want to increase the energy output from your FR fleet you could use a different design FR with a higher BG; through either using higher density fuels (e.g. metal or nitride), employing radial reflector regions (more of a proliferation risk higher levels of fissile Pu in fuel), shorter dwell time (less Pu held up i.e. reduce the total dwell time through 5yrs cooling + 2yrs for repro and fabrication)
75 GWe target installed capacity
FRs introduced at same rate as LWRs retire
LWRs fuelled with UO2

24. Generating capacity - transition from LWRs to fast reactors
The gap is plugged through reducing the total dwell time through 5yrs cooling + 2yrs for repro and fabrication)
You need a pre-existing LWR programme of similar size to produce sufficient Pu to fuel Breeding in a FR programme.
FR’s need 10Te Pu to start + 5Te Pu for core reloads – you can produce this in around 60 yrs lifetime of a LWR.
In this case you can see how the stockpile of Pu from a LWR programme is able to build a FR programme which becomes self sufficient through breeding.
It takes
75 GWe target installed capacity
FRs introduced at same rate as LWRs retire
FR Fuel dwell time is reduced

25. Generating capacity - transition from LWRs to fast reactors - Scenario (b)
You need a pre-existing LWR programme of similar size to produce sufficient Pu to fuel Breeding in a FR programme.
FR’s need 10Te Pu to start + 5Te Pu for core reloads – you can produce this in around 60 yrs lifetime of a LWR.
In this case you can see how the stockpile of Pu from a LWR programme is able to build a FR programme which becomes self sufficient through breeding.
Thorium reactor fleet power increases at a rate of 5GWe over 10 years assuming system as indicated in Doubling time example
10GWe of Th breeding FR’s introduced ~2040
FR breeders fuelled only with U-233 introduced ~2045

26. Conclusions Thorium is a valuable strategic alternative to uranium
Sustainability remains one of the main drivers
Radiotoxicity benefit is real, but modest
Long term equilibrium radiotoxicity a simplistic measure
Inherent proliferation resistance not proven for thorium
Economics of thorium not known at present
Minimum year timeframe for commercial deployment (thermal systems) and longer timeframes for fast reactors
Significant R&D programme required to progress technical maturity

27. Acknowledgements Kevin Hesketh Robbie Gregg Mike Thomas Chris Grove
Richard Stainsby

28. Further information

29. Thorium history
In the 1950s through to the 1980s, there were thorium research programmes for:
Pressurised water reactors (PWR)
Shippingport breeder core
Germany-Brazil collaboration
High temperature gas reactors (HTR)
DRAGON (UK), Fort St Vrain (USA), Peach Bottom (USA), AVR (Germany)
Molten salt reactors (MSR)
Molten Salt Reactor Experiment (USA)
The common driver for all these plants was to decouple nuclear expansion from uranium availability

30. Why did thorium research stall?
Thorium cycle requires neutrons from uranium or plutonium fissions to get started
U/Pu fuel cycle already established
Large barrier to entry for a new system
Technological issues
THOREX reprocessing and fabrication of U-233 fuels

31. India/Lightbridge India Lightbridge
Synergistic fuel cycle involving fast reactor and Advanced Heavy Water Reactors (AHWR)
Fast reactor will breed U-233 in a thorium blanket
U-233 will be recycled into AHWR fuel
Lightbridge
Seed/blanket assembly design for PWRs
Low enriched uranium (LEU) seed region provides spare neutrons
ThO2 blanket breeds U-233
Seed and blanket regions have different in-core dwell times

32. Pu/Th MOX
AREVA are investigating PuO2/ThO2 MOX fuel for the eventual disposition of PWR MOX fuel assemblies
PWR MOX fuel currently not reprocessed in France
Held in long term storage pending eventual recycle in SFR fleet
Requirement to cover all contingency that SFR fleet is not built
Recycle of Pu from MOX fuel preferred over disposal
PuO2/ThO2 MOX is presumed to be another option with potential advantage of low development cost and high stability as a final waste form
Thor Energy undertaking PuO2/ThO2 MOX fuel qualification programme through a international consortium

33. Th-232 radiative capture cross-section

34. U-238 radiative capture cross-section

35. Decay heat and radiotoxicity
Thorium-plutonium MOX fuel theoretically could be advantageous for UK plutonium disposition
Detailed assessment by NNL of decay heat load and radiotoxicity per GWye shows there is only a marginal difference between Th-Pu MOX and U-Pu MOX
This is a holistic calculation that accounts for the total decay heat outputs of different scenarios
In the Th-Pu and U-Pu MOX cases, the decay heat is concentrated in the MOX assemblies, whereas in the UO2 reference case it is distributed over a larger number of UO2 assemblies

36. Core fissile inventory MC
The fissile inventory of the core depends on a number of factors:
Minimum critical mass for the system
Thermal power output
Specific rating of the core in MW/tonne
Refuelling interval
Refuelling strategy – single batch or multiple batch core loading
KEY POINTS:
The minimum critical mass can range over 3 orders of magnitude for different configurations (for example from 5 kg for a HEU research reactor core to several 1000 kg for a typical 1 GWe power plant)
Workable designs typically nearer the upper end of the mass range and therefore MC is practically constrained to a few x 1000 kg
Very important distinction between MC and m, which are orders of magnitude different for any practical system
My main point is that engineering constraints force the core fissile mass to be at least 1000 kg. While it is possible to have a much smaller critical mass, if you want a large thermal output you need to spread the thermal load over a larger mass of fuel, otherwise the fuel will fail. The graph on the next slide confirms this for all current and projected commercial power reactors.

37. Illustration that large power reactors have a large fissile inventory based on survey or world reactors

38. Out of core fissile inventory MO
For a conventional solid fuel reactor, this is the inventory in spent fuel awaiting reprocessing or being reprocessed, plus the inventory of fuel under fabrication, which depends on:
Spent fuel cooling time tc
Reprocessing time tr
Fuel fabrication time tf
For a fuel dwell time T, MO scales with MC:
MO = MC x (tc+tr+tf)/ T
For a liquid fuel system such as Molten Salt Reactor (MSR), there is an out-of-core inventory, which is the mass of fuel circulating through the heat exchangers
This is the standard textbook definition.
Out of core fissile inventory MO
For a conventional solid fuel reactor, this is the inventory in spent fuel awaiting reprocessing or being reprocessed, plus the inventory of fuel under fabrication, which depends on:
Spent fuel cooling time tc
Reprocessing time tr
Fuel fabrication time tf
For a fuel dwell time T, MO scales with MC:
MO = MC x (tc+tr+tf)/ T
For a liquid fuel system such as Molten Salt Reactor (MSR), there is an out-of-core inventory, which is the mass of fuel circulating through the heat exchangers