(Updated August 2012)
- Thorium is more abundant in nature than uranium.
- It is fertile rather than fissile, and can be used in conjunction with fissile material as nuclear fuel.
- Thorium fuels can breed fissile uranium-233.
The use of thorium as a new primary energy source has been a tantalizing prospect for many years. Extracting its latent energy value in a cost-effective manner remains a challenge, and will require considerable R&D investment.
Nature and sources of thorium
Thorium is a naturally-occurring, slightly radioactive metal discovered in 1828 by the Swedish chemist Jons Jakob Berzelius, who named it after Thor, the Norse god of thunder. It is found in small amounts in most rocks and soils, where it is about three times more abundant than uranium. Soil commonly contains an average of around 6 parts per million (ppm) of thorium.
Thorium exists in nature in a single isotopic form - Th-232 - which decays very slowly (its half-life is about three times the age of the Earth). The decay chains of natural thorium and uranium give rise to minute traces of Th-228, Th-230 and Th-234, but the presence of these in mass terms is negligible.
When pure, thorium is a silvery white metal that retains its lustre for several months. However, when it is contaminated with the oxide, thorium slowly tarnishes in air, becoming grey and eventually black. Thorium oxide (ThO2), also called thoria, has one of the highest melting points of all oxides (3300°C). When heated in air, thorium metal turnings ignite and burn brilliantly with a white light. Because of these properties, thorium has found applications in light bulb elements, lantern mantles, arc-light lamps, welding electrodes and heat-resistant ceramics. Glass containing thorium oxide has a high refractive index and dispersion and is used in high quality lenses for cameras and scientific instruments.
The most common source of thorium is the rare earth phosphate mineral, monazite, which contains up to about 12% thorium phosphate, but 6-7% on average. Monazite is found in igneous and other rocks but the richest concentrations are in placer deposits, concentrated by wave and current action with other heavy minerals. World monazite resources are estimated to be about 12 million tonnes, two-thirds of which are in heavy mineral sands deposits on the south and east coasts of India. There are substantial deposits in several other countries (see Table below). Thorium recovery from monazite usually involves leaching with sodium hydroxide at 140°C followed by a complex process to precipitate pure ThO2.
Thorite (ThSiO4) is another common mineral. A large vein deposit of thorium and rare earth metals is in Idaho.
The 2007 IAEA-NEA publication Uranium 2007: Resources, Production and Demand (often referred to as the 'Red Book') gives a figure of 4.4 million tonnes of total known and estimated resources, but this excludes data from much of the world. Data for reasonably assured and inferred resources recoverable at a cost of $80/kg Th or less are given in the table below. Some of the figures are based on assumptions and surrogate data for mineral sands, not direct geological data in the same way as most mineral resources.
Estimated world thorium resources 1
There is no international or standard classification for thorium resources and identified Th resources do not have the same meaning in terms of classification as identified U resources. Thorium is not a primary exploration target and resources are estimated in relation to uranium and rare earths resources.
||% of total
OECD NEA & IAEA, Uranium 2011: Resources, Production and Demand ("Red Book"), using the lower figures of any range and omitting ‘unknown’ CIS estimate.
Thorium as a nuclear fuel
Thorium (Th-232) is not itself fissile and so is not directly usable in a thermal neutron reactor – in this regard it is very similar to uranium-238. However, it is ‘fertile’ and upon absorbing a neutron will transmute to uranium-233 (U-233)a, which is an excellent fissile fuel material b. Thorium fuel concepts therefore require that Th-232 is first irradiated in a reactor to provide the necessary neutron dosing. The U-233 that is produced can either be chemically separated from the parent thorium fuel and recycled into new fuel, or the U-233 may be usable ‘in-situ’ in the same fuel form.
Thorium fuels therefore need a fissile material as a ‘driver’ so that a chain reaction (and thus supply of surplus neutrons) can be maintained. The only fissile driver options are U-233, U-235 or Pu-239 (none of which is easy to supply).
It is possible – but quite difficult – to design thorium fuels that produce more U-233 in thermal reactors than the fissile material they consume (this is referred to as having a fissile conversion ratio of more than 1.0 and is also called breeding). Thermal breeding with thorium is only really possible using U-233 as the fissile driver, and to achieve this the neutron economy in the reactor has to be very good (ie, low neutron loss through escape or parasitic absorption). The possibility to breed fissile material in slow neutron systems is a unique feature for thorium-based fuels and is not possible with uranium fuels.
Another distinct option for using thorium is as a ‘fertile matrix’ for fuels containing plutonium (and even other transuranic elements like americium). No new plutonium is produced from the thorium component, unlike for uranium fuels, and so the level of net consumption of this metal is rather high.
In fresh thorium fuel, all of the fissions (thus power and neutrons) derive from the driver component. As the fuel operates the U-233 content gradually increases and it contributes more and more to the power output of the fuel. The ultimate energy output from U-233 (and hence indirectly thorium) depends on numerous fuel design parameters, including: fuel burn-up attained, fuel arrangement, neutron energy spectrum and neutron flux (affecting the intermediate product protactinium-233, which is a neutron absorber). The fission of a U-233 nucleus releases about the same amount of energy (200 MeV) as that of U-235.
An important principle in the design of thorium fuel is that of heterogeneous fuel arrangements in which a high fissile (and therefore higher power) fuel zone called the seed region is physically separated from the fertile (low or zero power) thorium part of the fuel – called the blanket. Such an arrangement is far better for supplying surplus neutrons to thorium nuclei so they can convert to fissile U-233, in fact all thermal breeding fuel designs are heterogeneous. This principle applies to all the thorium-capable reactor systems.
Reactors able to use Thorium
There are seven types of reactor into which thorium can be introduced as a nuclear fuel. The first five of these have all entered into operational service at some point. The last two are still conceptual:
Heavy Water Reactors (PHWRs): These are very well suited for thorium fuels due to their combination of: (i) excellent neutron economy (their low parasitic neutron absorption means more neutrons can be absorbed by thorium to produce useful U-233), (ii) slightly faster average neutron energy which favours conversion to U-233, (iii) flexible on-line refueling capability. Furthermore, heavy water reactors (especially Candu) are well established and widely-deployed commercial technology for which there is extensive licensing experience.
Thorium fuel has been tested over the past 50 years at the NRX and NRU reactors at AECL's Chalk River Laboratories, see R&D section below. CANDU reactors have very high flexibility to switch between uranium and thorium based fuels without extensive modification.
There is potential application to Enhanced Candu 6 and ACR-1000 reactors fueled with 5% plutonium (reactor grade) plus thorium. In the closed fuel cycle, the driver fuel required for starting off is progressively replaced with recycled U-233, so that on reaching equilibrium 80% of the energy comes from thorium. Fissile drive fuel could be LEU, plutonium, or recycled uranium from LWR. Fleets of PHWRs with near-self-sufficient equilibrium thorium fuel cycles could be supported by a few fast breeder reactors to provide plutonium.
High-Temperature Gas-Cooled Reactors (HTRs): These are well suited for thorium-based fuels in the form of robust ‘TRISO’ coated particles of thorium mixed with plutonium or enriched uranium, coated with pyrolytic carbon and silicon carbide layers which retain fission gases. The fuel particles are embedded in a graphite matrix that is very stable at high temperatures. Such fuels can be irradiated for very long periods and thus deeply burn to exploit their original fissile charge. Thorium fuels can be designed for both ‘pebble bed’ and ‘prismatic’ HTR fuel varieties.
Boiling (Light) Water Reactors (BWRs): BWR fuel assemblies allow for structure & composition options, such as extra moderation and/or half-length fuel rods. This design flexibility means that well-optimized thorium fuels can be created for BWRs, for example, thorium-plutonium fuels that are tailored for ‘burning’ plutonium. BWRs are a well-understood and licensed reactor design.
Pressurised (Light) Water Reactors (PWRs): Viable thorium fuels can be designed for a PWR, though with less flexibility than for BWRs. Fuel needs to be in heterogeneous arrangements in order to achieve satisfactory fuel burn-up. It is not possible to design thorium-based PWR fuels that convert significant amounts of U-233. Even though PWRs are not the perfect reactor in which to use thorium, they are the industry workhorse and there is a lot of PWR licensing experience. They are a viable early-entry thorium platform.
Fast Neutron Reactors (FNRs): Thorium can serve as a fuel component for reactors operating with a fast neutron spectrum – in which a wider range of heavy nuclides are fissionable and may potentially drive a thorium fuel. There is, however, no relative advantage in using thorium instead of depleted uranium (DU) as a fertile fuel matrix in these reactor systems due to a higher fast-fission rate for U-238 and the fission contribution from residual U-235 in this material. Also, there is a huge amount of surplus DU available for use when more FNRs are commercially available, so thorium has little or no competitive edge in these systems.
Molten Salt Reactors (MSRs): These reactors are still at the design stage but will be very well suited for using thorium as a fuel. The unique fluid fuel incorporates thorium and uranium (U-233 and/or U-235) fluorides as part of a salt mixture that melts in the range 400-600ºC, and this liquid serves as both heat transfer fluid and the matrix for the fissioning fuel. The fluid circulates through a core region and then through a chemical processing circuit that removes various fission products (poisons) and/or the valuable U-233. Certain MSR designs [c] will be designed specifically for thorium fuels to produce useful amounts of U-233 – eventually leading to the self-sustaining use of thorium as an energy source.
Accelerator Driven Reactors (ADS): The sub-critical ADS system is an unconventional concept that is potentially ‘thorium capable’. Spallation neutrons are produced d when high-energy protons from an accelerator strike a heavy target like lead. These neutrons are directed at a region containing a thorium fuel, eg, Th-plutonium which reacts producing heat as in a conventional reactor. The system remains subcritical ie, unable to sustain a chain reaction without the proton beam. Difficulties lie with the reliability of high-energy accelerators and also with economics due to their high power consumption. (See also information page on Accelerator-Driven Nuclear Energy)
A key finding from thorium fuel studies to date is that it is not economically viable to use low-enriched uranium (LEU - with a U-235 content of up to 20%) as a fissile driver with thorium fuels, unless the fuel burn-up can be taken to very high levels – well beyond those currently attainable in LWRs with zirconium cladding.
With regard to proliferation significance, thorium-based power reactor fuels would be very poor source for fissile material usable in the illicit manufacture of an explosive device. U-233 contained in spent thorium fuel contains U-232 which decays to produce very radioactive daughter nuclides and these create a strong gamma radiation field. This confers proliferation resistance by creating significant handling problems and by greatly boosting the detectability (traceability) and ability to safeguard this material.
Prior Thorium Fuelled Electricity Generation
There have been several significant demonstrations of the use of thorium-based fuels to generate electricity in several reactor types. Many of these early trials were able to use high-enriched uranium (HEU) as the fissile ‘driver’ component, and this would not be considered today.
The 300 MWe Thorium High Temperature Reactor (THTR) in Germany, a HTR, operated with thorium-HEU fuel between 1983 and 1989. Over half of its 674,000 pebbles contained Th-HEU fuel particles (the rest graphite moderator and some neutron absorbers). These were continuously moved through the reactor as it operated, and on average each fuel pebble passed six times through the core.
The 40 MWe Peach Bottom HTR in the USA was a demonstration thorium-fuelled reactor that ran from 1967-74 . It used a thorium-HEU fuel in the form of microspheres of mixed thorium-uranium carbide coated with pyrolytic carbon. These were embedded in annular graphite segments (not pebbles). This reactor produced 33 billion kWh over 1349 equivalent full-power days with a capacity factor of 74%.
The 330 MWe Fort St Vrain HTR in Colorado, USA, was a larger-scale commercial successor to the Peach Bottom reactor and ran from 1976-89. It also used thorium-HEU fuel in the form of microspheres of mixed thorium-uranium carbide coated with silicon oxide and pyrolytic carbon to retain fission products. These were embedded in graphite ‘compacts’ that were arranged in hexagonal columns ('prisms'). Almost 25 tonnes of thorium was used in fuel for the reactor, much of which attained a burn-up of about 170 GWd/t.
A unique thorium-fuelled Light Water Breeder Reactor operated from 1977 to 1982 at Shippingport in the USA  – it used uranium-233 as the fissile driver in special fuel assemblies having independently movable ‘seed’ regions. The reactor core was housed in a reconfigured early PWR. It operated at 60 MWe (236 MWt) with an availability factor of 86% producing over 2.1 billion kWh. Post-operation inspections revealed that 1.39% more fissile fuel was present at the end of core life, proving that breeding had occurred.
* The core of the Shippingport demonstration LWBR consisted of an array of seed and blanket modules surrounded by an outer reflector region. In the seed and blanket regions, the fuel pellets contained a mixture of thorium-232 oxide (ThO2) and uranium oxide (UO2) that was over 98% enriched in U-233. The proportion by weight of UO2 was around 5-6% in the seed region, and about 1.5-3% in the blanket region. The reflector region contained only thorium oxide at the beginning of the core life. U-233 was used because at the time it was believed that U-235 would not release enough neutrons per fission and Pu-239 would parasitically capture too many neutrons to allow breeding in a PWR.
Indian heavy water reactors (PHWRs) have for a long time used thorium-bearing fuel bundles for power flattening in some fuel channels – especially in initial cores when special reactivity control measures are needed.
Other Thorium Energy R&D – Past & Present
Research into the use of thorium as a nuclear fuel has been taking place for over 40 years, though with much less intensity than that for uranium or uranium-plutonium fuels. Basic development work has been conducted in Germany, India, Canada, Japan, China, Netherlands, Belgium, Norway, Russia, Brazil, the UK & the USA. Test irradiations have been conducted on a number of different thorium-based fuel forms.
Noteworthy studies and experiments involving thorium fuel include:
Heavy Water Reactors: Thorium-based fuels for the ‘Candu’ PHWR system have been designed and tested in Canada for more than 50 years, including burn-up to 47 GWd/t. Dozens of test irradiations have been performed on fuels including: ThO2, mixed ThO2-UO2, (both LEU and HEU), and mixed ThO2-PuO2, (both reactor- and weapons-grade). R&D into thorium fuel use in CANDU reactors continues to be pursued by Canadian and Chinese groups. In China, INET has been looking at a wide range of fuel cycle options including thorium, especially for the Qinshan Phase III PHWR units, where there has been demonstrated use of 8 thorium oxide fuel pins in the middle of a Canflex fuel bundle with low-enriched uranium. The fuels have performed well in terms of their material properties.
Closed thorium fuel cycles have been designed  in which PHWRs play a key role due to their fuelling flexibility: thoria-based HWR fuels can incorporate recycled U-233, residual plutonium and uranium from used LWR fuel, and also minor actinide components in waste-reduction strategies.
In July 2009 a second phase agreement was signed among AECL, the Third Qinshan Nuclear Power Company (TQNPC), China North Nuclear Fuel Corporation and the Nuclear Power Institute of China to jointly develop and demonstrate the use of thorium fuel and to study the commercial and technical feasibility of its full-scale use in Candu units such as at Qinshan. This was supported in December 2009 by an expert panel appointed by CNNC. The panel also noted the ability of Candu reactors to re-use uranium recycled from light water reactor fuel, and unanimously recommended that China consider building two new Candu units to take advantage of the design's unique capabilities in utilizing alternative fuels. The expert panel comprised representatives from China’s leading nuclear academic, government, industry and R&D organizations. In particular it confirmed that thorium use in the Enhanced Candu 6 reactor design is “technically practical and feasible”, and cited the design’s “enhanced safety and good economics” as reasons it could be deployed in China in the near term.
India’s nuclear developers have designed an Advanced Heavy Water Reactor (AHWR) specifically as a means for ‘burning’ thorium – this will be the final phase of their 3-phase nuclear energy infrastructure plan (see below). The reactor will operate with a power of 300 MWe using thorium-plutonium or thorium-U-233 seed fuel in mixed oxide form. It is heavy water moderated (& light water cooled) and is capable of self-sustaining U-233 production. In each assembly 30 of the fuel pins will be Th-U-233 oxide, arranged in concentric rings. About 75% of the power will come from the thorium. Construction of the pilot AHWR may start in 2012.
High-Temperature Gas-Cooled Reactors: Thorium fuel was used in HTRs prior to the successful demonstration reactors described above. The UK operated the 20 MWth Dragon HTR from 1964 to 1973 for 741 full power days. Dragon was run as an OECD/Euratom cooperation project, involving Austria, Denmark, Sweden, Norway and Switzerland in addition to the UK. This reactor used thorium-HEU fuel elements in a 'breed and feed' mode in which the U-233 formed during operation replaced the consumption of U-235 at about the same rate. The fuel could be left in the reactor for about six years.
Germany operated the Atom Versuchs Reaktor (AVR) at Jülich for over 750 weeks between 1967 and 1988. This was a small pebble bed reactor that operated at 15 MWe, mainly with thorium-HEU fuel. About 1360 kg of thorium was used in some 100,000 pebbles. Burn-ups of 150 GWd/t were achieved.
Pebble bed reactor development builds on German work with the AVR and THTR and is under development in China (HTR-10, and HTR-PM).
Light Water Reactors: The feasibility of using thorium fuels in a PWR was studied in considerable detail during a collaborative project between Germany and Brazil in the 1980s . The vision was to design fuel strategies that used materials effectively – recycling of plutonium and U-233 was seen to be logical. The study showed that appreciable conversion to U-233 could be obtained with various thorium fuels, and that useful uranium savings could be achieved. The program terminated in 1988 for non-technical reasons. It did not reach its later stages which would have involved trial irradiations of thorium-plutonium fuels in the Angra-1 PWR in Brazil, although preliminary Th-fuel irradiation experiments were performed in Germany. Most findings from this study remain relevant today.
Thorium-plutonium oxide (Th-MOX) fuels for LWRs are being developed by Norwegian proponents with a view that these are the most readily achievable option for tapping energy from thorium. This is because such fuel is usable in existing reactors (with minimal modification) and the fuel can be made in existing uranium-MOX plants, using existing technology and licensing experience. A thorium-MOX fuel irradiation experiment will get underway in the Halden fuel testing reactor in 2012.
The so-called Radkowsky Thorium Reactor is a specific, heterogeneous ‘seed & blanket’ thorium fuel concept, originally designed for Russian-type LWRs (VVERs) . Enriched uranium (20% U-235) or plutonium is used in a seed region at the centre of a fuel assembly, with this fuel being in a unique metallic form. The central seed portion is demountable from the blanket material which remains in the reactor for nine years e, but the centre seed portion is burned for only three years (as in a normal VVER). Design of the seed fuel rods in the centre portion draws on experience of Russian naval reactors.
The European Framework Program has supported a number of relevant research activities into thorium fuel use in LWRs. Three distinct trial irradiations have been performed on thorium-plutonium fuels, including a test pin loaded in the Obrigheim PWR over 2002-06 during which it achieved about 38 GWd/t burnup.
A small amount of thorium-plutonium fuel was irradiated in the 60 MWe Lingen BWR in Germany in the early 1970s. The fuel contained 2.6 % of high fissile-grade plutonium (86% Pu-239) and the fuel achieved about 20 GWd/t burnup. The experiment was not representative of commercial fuel, however the experiment allowed for fundamental data collection and benchmarking of codes for this fuel material.
Molten Salt Reactors: The Oak Ridge National Laboratory (USA) designed and built a thorium-based demonstration MSR using U-233 as the main fissile driver. The reactor ran over 1965-69 and operated at powers up to 7.4 MWt. The lithium-beryllium salt worked at 600-700ºC and ambient pressure. The R&D program demonstrated the feasibility of this system and highlighted some unique corrosion and operational issues that need to be addressed if constructing a larger pilot MSR.
There is significant renewed interest in developing thorium-fuelled MSRs. Projects are (or have recently been) underway in China, Japan, Russia, France and the USA.
It is notable that the MSR is one of the six ‘Generation IV’ reactor designs selected as worthy of further development (see information page on Generation IV Nuclear Reactors). The thorium-fuelled MSR variant is sometimes referred to at the Liquid Fluoride Thorium Reactor (LFTR). See subsection below.
An aqueous homogenous suspension reactor operated in the Netherlands at 1 MWth for three years using thorium in the mid-1970s. The thorium-HEU fuel was circulated in solution with continuous reprocessing outside the core to remove fission products, resulting in a high conversion rate to U-233.
Accelerator-Driven Reactors: A number of groups have investigated how a thorium-fuelled accelerator-driven reactor (ADS) may work and appear. Perhaps most notable is the ‘ADTR’ design patented by a UK group. This reactor operates very close to criticality and therefore requires a relatively small proton beam to drive the spallation neutron source. Earlier proposals for ADS reactors required high-energy and high-current proton beams which are energy-intensive to produce, and for which operational reliability is a problem.
Research Reactor ‘Kamini’: India has been operating a low-power U-233 fuelled reactor at Kalpakkam since 1996 – this is a 30 kWth experimental facility using U-233 in aluminium plates (a typical fuel-form for research reactors). Kamini is water cooled with a beryllia neutron reflector. The total mass of U-233 in the core is around 600 grams. It is noteworthy for being the only U-233 fuelled reactor in the world, though it does not in itself directly support thorium fuel R&D. The reactor is adjacent to the 40 MWt Fast Breeder Test Reactor in which ThO2 is irradiated, producing the U-233 for Kamini.
Fast breeder reactors (FBRs) play an ancillary role in India's three-stage nuclear power program (see subsection on India's plans for thorium cycle below) but do not themselves use thorium.
Liquid Fluoride Thorium Reactor
A development of the MSR concept is the Liquid Fluoride Thorium Reactor (LFTR), utilizing U-233 which has been bred in a liquid thorium salt blanket.*
* The molten salt in the core circuit consists of lithium, beryllium and fissile U-233 fluorides. It operates at some 700°C and circulates at low pressure within a graphite structure that serves as a moderator and neutron reflector. Most fission products dissolve or suspend in the salt and some of these are removed progressively in an adjacent radiochemical processing unit. Actinides are less-readily formed than in fuel with atomic mass greater than 235. The blanket circuit contains a significant amount of thorium tetrafluoride in the molten Li-Be fluoride salt. Newly-formed U-233 forms soluble uranium tetrafluoride (UF4), which is converted to gaseous uranium hexafluoride (UF6) by bubbling fluorine gas through the salt (which does not chemically affect the less-reactive thorium tetrafluoride). The volatile uranium hexafluoride is captured, reduced back to soluble UF4 by hydrogen gas, and finally is directed to the core to serve as fissile fuel.
Safety is achieved with a freeze plug which if power is cut allows the fuel to drain into subcritical geometry in a catch basin. There is also a negative temperature coefficient of reactivity due to expansion of the fuel.
The China Academy of Sciences in January 2011 launched an R&D program on LFTR, known there as the thorium-breeding molten-salt reactor (Th-MSR or TMSR), and claimed to have the world's largest national effort on it, hoping to obtain full intellectual property rights on the technology. The TMSR Research Centre apparently has a 5 MWe MSR prototype under construction at Shanghai Institute of Applied Physics (SINAP, under the Academy) with 2015 target operation. The US Department of Energy is collaborating with the Academy on the program.
India's plans for thorium cycle
With huge resources of easily-accessible thorium and relatively little uranium, India has made utilization of thorium for large-scale energy production a major goal in its nuclear power programme, utilising a three-stage concept:
- Pressurised heavy water reactors (PHWRs) fuelled by natural uranium, plus light water reactors, producing plutonium.
- Fast breeder reactors (FBRs) using plutonium-based fuel to breed U-233 from thorium. The blanket around the core will have uranium as well as thorium, so that further plutonium (particularly Pu-239) is produced as well as the U-233.
- Advanced heavy water reactors (AHWRs) burn the U-233 and this plutonium with thorium, getting about 75% of their power from the thorium. The used fuel will then be reprocessed to recover fissile materials for recycling.
This Indian programme has moved from aiming to be sustained simply with thorium to one 'driven' with the addition of further fissile plutonium from the FBR fleet, to give greater efficiency. In 2009, despite the relaxation of trade restrictions on uranium, India reaffirmed its intention to proceed with developing the thorium cycle.
A 500 MWe prototype FBR under construction in Kalpakkam is designed to produce plutonium to enable AHWRs to breed U-233 from thorium. India is focusing and prioritizing the construction and commissioning of its sodium-cooled fast reactor fleet in which it will breed the required plutonium. This will take another 15 – 20 years and so it will still be some time before India is using thorium energy to a significant extent.
Developing a thorium-based fuel cycle
Thorium fuel cycles offer attractive features, including lower levels of waste generation, less transuranic elements in that waste, and providing a diversification option for nuclear fuel supply. Also, the use of thorium in most reactor types leads to significant extra safety margins. Despite these merits, the commercialization of thorium fuels faces some significant hurdles in terms of building an economic case to undertake the necessary development work.
A great deal of testing, analysis and licensing and qualification work is required before any thorium fuel can enter into service. This is expensive and will not eventuate without a clear business case and government support - abundant uranium is available.
Other impediments to the development of thorium fuel cycle are the higher cost of fuel fabrication* and the cost of reprocessing to provide the fissile plutonium driver material.
* The high cost of fuel fabrication is due partly to the high level of radioactivity that builds up in U-233 chemically separated from the irradiated thorium fuel. Separated U-233 is always contaminated with traces of U-232 which decays (with a 69-year half-life) to daughter nuclides such as thallium-208 that are high-energy gamma emitters. Although this confers proliferation resistance to the fuel cycle by making U-233 hard to handle and easy to detect, it results in increased costs. There are similar problems in recycling thorium itself due to highly radioactive Th-228 (an alpha emitter with two-year half life) present.
Nevertheless, the thorium fuel cycle offers enormous energy security benefits in the long-term – due to its potential for being a self-sustaining fuel without the need for fast neutron reactors. It is therefore an important and potentially viable technology that seems able to contribute to building credible, long-term nuclear energy scenarios.
In 2010 the UK’s National Nuclear Laboratory (NNL) published a paper on the thorium cycle, concluding for the short to medium term:
"NNL believes that the thorium fuel cycle does not currently have a role to play in the UK context, other than its potential application for plutonium management in the medium to long term and depending on the indigenous thorium reserves, is likely to have only a limited role internationally for some years ahead. The technology is innovative, although technically immature and currently not of interest to the utilities, representing significant financial investment and risk without notable benefits. In many cases, the benefits of the thorium fuel cycle have been over-stated."
Weapons and non-proliferation
The thorium fuel cycle is sometimes promoted as having excellent non-proliferation credentials. This is true, but some history and physics bears noting.
The USA produced about 2 tonnes of U-233 from thorium during the ‘Cold War’, at various levels of chemical and isotopic purity, in plutonium production reactors. It is possible to use U-233 in a nuclear weapon, and in 1955 the USA detonated a device with a plutonium-U-233 composite pit, in Operation Teapot. Yield was less than anticipated, at 22 kilotons. In 1998 India detonated a very small device based on U-233 called Shakti V. However, the production of U-233 inevitably also yields U-232 which is a strong gamma-emitter, as are some decay products, making the material extremely difficult to handle and also easy to detect.
a. Neutron absorption by Th-232 produces Th-233 which beta-decays (with a half-life of about 22 minutes) to protactinium-233 (Pa-233) – and this decays to U-233 by further beta decay (with a half-life of 27 days). Some of the bred-in U-233 is converted to U-234 by further neutron absorption. U-234 is an unwanted parasitic neutron absorber. It converts to fissile U-235 (the naturally occuring fissile isotope of uranium) and this somewhat compensates for this neutronic penalty. In fuel cycles involving the multi-recycle of thorium-U-233 fuels, the build up of U-234 can be appreciable. [Back]
b. A U-233 nucleus yields more neutrons, on average, when it fissions (splits) than either a uranium-235 or plutonium-239 nucleus. In other words, for every thermal neutron absorbed in a U-233 fuel there are a greater number of neutrons produced and released into the surrounding fuel. This gives better neutron economy in the reactor system.. [Back]
c. MSRs using thorium will likely have a distinct ‘blanket’ circuit which is optimised for producing U-233 from dissolved thorium. Neutron moderation is tailored by the amount of graphite in the core (aiming for an epithermal spectrum). This uranium can be selectively removed as uranium hexafluoride (UF6) by bubbling fluorine gas through the salt. After conversion it can be directed to the core as fissile fuel. [Back]
d. Spallation is the process where nucleons are ejected from a heavy nucleus being hit by a high energy particle. In this case, a high-enery proton beam directed at a heavy target expels a number of spallation particles, including neutrons. [Back]
e. Blanket fuel is designed to reach 100 GWd/t burn-up. Together, the seed and blanket have the same geometry as a normal VVER-100 fuel assembly (331 rods in a hexagonal array 235 mm wide). [Back]
1. Data taken from Uranium 2007: Resources, Production and Demand, Nuclear Energy Agency (June 2008), NEA#6345 (ISBN 9789264047662). The 2009 figures are largely unchanged. Australian data from Thorium, in Australian Atlas of Minerals Resources, Mines & Processing Centres, Geoscience Australia (see below under General sources) [Back]
2. 2. K.P. Steward, “Final Summary Report on the Peach Bottom End-of-Life Program”, General Atomics Report GA-A14404, (1978)
3. (i) W.J. Babyak, L.B. Freeman, H.F. Raab, “LWBR: A successful demonstration completed” Nuclear News, Sept 1988, pp114-116 (1988), (ii) J.C. Clayton, “The Shippingport Pressurized Water Reactor and Light Water Breeder Reactor” Westinghouse Bettis Atomic Power Laboratory WAPD-T-3007 (October 1993). [Back]
4. (i) S. Şahin, etal, “CANDU Reactor as Minor Actinide / Thorium Burner with Uniform Power Density in the Fuel Bundle” Ann.Nuc.Energy. 35, 690-703 (2008), (ii) J. Yu, K, Wang, R. Sollychin, etal, “Thorium Fuel Cycle of a Thorium-Based Advanced Nuclear Energy System” Prog.Nucl.Energy. 45, 71-84 (2004) [Back]
5. “German Brazilian Program of Research and Development on Thorium Utilization in PWRs”, Final Report, Kernforschungsanlage Jülich, 1988. [Back]
6. A. Galperin, A. Radkowsky and M. Todosow, A Competitive Thorium Fuel Cycle for Pressurized Water Reactors of Current Technology, Proceedings of three International Atomic Energy Agency meetings held in Vienna in 1997, 1998 and 1999, IAEA TECDOC 1319: Thorium fuel utilization: Options and trends, IAEA-TECDOC-1319. [Back]
Thorium based fuel options for the generation of electricity: Developments in the 1990s, IAEA-TECDOC-1155, International Atomic Energy Agency, May 2000
Thorium, in Australian Atlas of Minerals Resources, Mines & Processing Centres (www.australianminesatlas.gov.au), Geoscience Australia (2009)
Taesin Chung, The role of thorium in nuclear energy, Uranium Industry Annual 1996, Energy Information Administration, DOE/EIA-0478(96) p.ix-xvii (April 1997)
M. Benedict, T H Pigford and H W Levi, Nuclear Chemical Engineering (2nd Ed.), Chapter 6: Thorium, , p.283-317, 1981, McGraw-Hill(ISBN: 0070045313)
Kazimi M.S. 2003, Thorium Fuel for Nuclear Energy, American Scientist (Sept-Oct 2003)
W.J. Babyak, L.B. Freeman, H.F. Raab, “LWBR: A successful demonstration completed” Nuclear News, Sept 1988, pp114-116 (1988)
12th Indian Nuclear Society Annual Conference 2001 conference proceedings, vol 2 (lead paper)
Several papers and articles related to the Radkowsky thorium fuel concept are available on the Lightbridge (formerly Thorium Power) website (www.ltbridge.com)
Robert Hargraves and Ralph Moir, Liquid Fluoride Thorium Reactors, American Scientist, Vol. 98, No. 4, P. 304 (July-August 2010)
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Accelerator-Driven Nuclear Energy
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