Thorium continues to be a tanatalising possibility for use in nuclear power reactors, though for many years India has been the only sponsor of major research efforts to use it. Other endeavours include the development of the Radkowsky Thorium Reactor concept being carried out by US company Thorium Power (now Lightbridge Corporation) with Russian collaboration.
In mid-2009, AECL signed agreements with three Chinese entities to develop and demonstrate the use of thorium fuel in the Candu reactors at Qinshan in China. Another mid-2009 agreement, between Areva and Lightbridge Corporation, was for assessing the use of thorium fuel in Areva’s EPR, drawing upon earlier research. Thorium can also be used in Generation IV and other advanced nuclear fuel cycle systems.
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-232 (Th-232) decays very slowly (its half-life is about three times the age of the Earth) but other thorium isotopes occur in its and in uranium’s decay chains. Most of these are short-lived and hence much more radioactive than Th-232, though on a mass basis they are 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.
|(Reasonably assured and inferred resources recoverable at
up to $80/kg Th)
|Country||Tonnes||% of total|
Thorium as a nuclear fuel
Thorium, as well as uranium, can be used as a nuclear fuel. Although not fissile itself, Th-232 will absorb slow neutrons to produce uranium-233 (U-233)a, which is fissile (and long-lived). The irradiated fuel can then be unloaded from the reactor, the U-233 separated from the thorium, and fed back into another reactor as part of a closed fuel cycle. Alternatively, U-233 can be bred from thorium in a blanket, the U-233 separated, and then fed into the core.
In one significant respect U-233 is better than uranium-235 and plutonium-239, because of its higher neutron yield per neutron absorbed. Given a start with some other fissile material (U-233, U-235 or Pu-239) as a driver, a breeding cycle similar to but more efficient than that with U-238 and plutonium (in normal, slow neutron reactors) can be set up. (The driver fuels provide all the neutrons initially, but are progressively supplemented by U-233 as it forms from the thorium.) However, there are also features of the neutron economy which counter this advantage. In particular the intermediate product protactinium-233 (Pa-233) is a neutron absorber which diminishes U-233 yield.
Over the last 40 years there has been interest in utilising thorium as a nuclear fuel since it is more abundant in the Earth’s crust than uranium. Also, all of the mined thorium is potentially useable in a reactor, compared with the 0.7% of natural uranium in today’s reactorsb, so some 40 times the amount of energy per unit mass might theoretically be available (without recourse to fast neutron reactors). But this relative advantage vanishes if fast neutron reactors are used for uranium.
Thorium R&D history
The use of thorium-based fuel cycles has been studied for about 40 years, but on a much smaller scale than uranium or uranium/plutonium cycles. Basic research and development has been conducted in Germany, India, Japan, Russia, the UK and the USA. Test reactor irradiation of thorium fuel to high burn-ups has also been conducted and several test reactors have either been partially or completely loaded with thorium-based fuel.
Noteworthy experiments involving thorium fuel include the following, the first three being high-temperature gas-cooled reactors:
There have also been several experiments with fast neutron reactors.
Much experience has been gained in thorium-based fuel in power reactors around the world, some using high-enriched uranium (HEU) as the main fuel:
Light Water Breeder Reactor
The Light Water Breeder Reactor (LWBR) concept is a major potential application for conventional pressurised water reactors (PWRs) and was successfully demonstrated at the Shippingport reactor in the USA2. Shippingport commenced commercial operation in December 1957 as the first large-scale nuclear power reactor to be operated solely for electricity production. In 1965 the Atomic Energy Commission began designing a uranium-233/thorium core for the reactor and in 1976, the Energy Research and Development Administration (now the Department of Energy) established the Advanced Water Breeder Applications programme to evaluate the LWBR concept for commercial-scale applications. Shippingport operated as an LWBR between August 1977 and October 1982, when the station was finally shut down. During this period, the demonstration LWBR operated for over 29,000 effective full power hours with an availability factor of 76% and had a gross electrical output of over 2.1 billion kilowatt hours. Following operation, inspection of the core found that 1.39% more fissile fuel was present at the end of core life than at the beginning, 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.
Current thorium fuel cycle research
Several advanced reactors concepts are currently being developed, including:
Radkowsky Thorium Reactor
The work at Shippingport (see section above on the Light Water Breeder Reactor) was developed by Alvin Radkowsky, who was the chief scientist of the United States Navy’s nuclear propulsion programme from 1950 to 1972 and headed the team that built the Shippingport plant. The Radkowsky Thorium Reactor (RTR) addresses the aspects of the thorium fuel cycle that are considered sensitive from the point of view of weapons proliferation. In particular the RTR avoids the need to separate U-233.
Radkowsky proposed the use of a heterogenous seed-blanket fuel assembly geometry, which separates the uranium (or plutonium) part of the fuel (the seed) from the thorium part of the fuel (the blanket). In the blanket part, U-233 is generated and fissioned, while the seed part supplies neutrons to the blanket. Either uranium enriched to 20% U-235 or plutonium can be used in the seed region3. One method of increasing the proliferation resistance of the design is to include some U-238 in the thorium blanket. Any uranium chemically separated from it (for the U-233 ) would not be useable for weapons. Used blanket fuel would also contain U-232, which decays rapidly and has very gamma-active daughters creating significant problems in handling the bred U-233 and hence conferring proliferation resistance. Plutonium produced in the seed will have a high proportion of Pu-238, generating a lot of heat and making it even more unsuitable for weapons than normal reactor-grade plutonium. Radkowsky’s designs are currently being developed by Thorium Power (now Lightbridge Corporation)d, based in McLean, Virginia.
Since 1994, Thorium Power Ltd has been involved in a Russian programme to develop a thorium-uranium fuel, which more recently has moved to have a particular emphasis on utilization of weapons-grade plutonium in a thorium-plutonium fuel. The program is based at Moscow’s Kurchatov Institute and receives US government funding to design fuel for Russian VVER-1000 reactors. The design has a demountable centre portion and blanket arrangement, with the plutonium in the centre and the thorium (with uranium) around ite. The blanket material remains in the reactor for nine years but the centre portion is burned for only three years (as in a normal VVER). Design of the seed fuel rods in the centre portion draws on extensive experience of Russian navy reactors.
The thorium-plutonium fuel claims four advantages over the use of mixed uranium-plutonium oxide (MOX) fuel: increased proliferation resistance; compatibility with existing reactors – which will need minimal modification to be able to burn it; the fuel can be made in existing plants in Russia; and a lot more plutonium can be put into a single fuel assembly than with MOX fuel, so that three times as much can be disposed of as when using MOX. The used fuel amounts to about half the volume of MOX and is even less likely to allow recovery of weapons-useable material than used MOX fuel, since less fissile plutonium remains in it. With an estimated 150 tonnes of surplus weapons plutonium in Russia, the thorium-plutonium project would not necessarily cut across existing plans to make MOX fuel.
Liquid Fluoride Thorium Reactor
A quite different concept is the Liquid Fluoride Thorium Reactor (LFTR), utilizing U-233 which has been bred in a liquid thorium salt blanket.
The core consists of fissile U-233 tetrafluoride in molten fluoride salts of lithium and beryllium at some 700°C and at low pressure within a graphite structure that serves as a moderator and neutron reflector. Fission products dissolve in the salt and are removed progressively – xenon bubbles out, others are captured chemically. Actinides are less-readily formed than in fuel with atomic mass >235, and those that do form stay in the fuel until they are transmuted and eventually fissioned.
The blanket contains a mixture of thorium tetrafluoride in a fluoride salt containing lithium and beryllium, made molten by the heat of the core. Newly-formed U-233 forms soluble uranium tetrafluoride (UF4), which is converted to gaseous uranium hexafluoride (UF6) by bubbling fluorine gas through the blanket solution (which does not chemically affect the less-reactive thorium tetrafluoride). Uranium hexafluoride comes out of solution, is captured, then is reduced back to soluble UF4 by hydrogen gas in a reduction column, and finally is directed to the core to serve as fissile fuel.
The LFTR is not a fast reactor, but with some moderation by the graphite is epithermal (intermediate neutron speed). 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.
India’s plans for thorium cycle
With about six times more thorium than 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:
This Indian programme has moved from aiming to be sustained simply with thorium to one ‘driven’ with the addition of further fissile uranium and plutonium, 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.
Another option for the third stage, while continuing with the PHWR and FBR stages, is the use of subcritical accelerator driven systems.
Thorium and accelerator driven systems
In an accelerator driven system (ADS), high-energy neutrons are produced through the spallationf reaction of high-energy protons from an accelerator striking heavy target nuclei (lead, lead-bismuth or other material). These neutrons can be directed to a subcritical reactor containing thorium, where the neutrons breed U-233 and promote the fission of it. There is therefore the possibility of sustaining a fission reaction which can readily be turned off, and used either for power generation or destruction of actinides resulting from the U/Pu fuel cycle. The use of thorium instead of uranium reduces the quantity of actinides that are produced. (See also information page on Accelerator-Driven Nuclear Energy.)
Developing a thorium-based fuel cycle
Despite the thorium fuel cycle having a number of attractive features, development has always run into difficulties.
The main attractive features are:
The problems include:
Much development work is still required before the thorium fuel cycle can be commercialised, and the effort required seems unlikely while (or where) abundant uranium is available. In this respect, recent international moves to bring India into the ambit of international trade might result in the country ceasing to persist with the thorium cycle, as it now has ready access to traded uranium and conventional reactor designs.
Nevertheless, the thorium fuel cycle, with its potential for breeding fuel without the need for fast neutron reactors, holds considerable potential in the long-term. It is a significant factor in the long-term sustainability of nuclear energy.