Thorium

(Updated 11 November 2011) 

• 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 resources¹Â 

 

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).

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.
  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 [2]. 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 [3] – 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. The fuels have performed well in terms of their material properties.

Closed thorium fuel cycles have been designed [4] 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.

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 [5]. 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) [6]. 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-thorium 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.*