Thorium Energy: The clean energy source we need!

Engineerblogger
Nov 2011



With the global spotlight on green / renewable energy and on the safety of nuclear power following Japan's tsunami and the problems at the Fukushima Daiichi plant, many countries are looking at spearheading efforts to make the industry and the environment safer.  For instance, the Chinese are investing millions in research into reactors powered by the element Thorium -- a metal, proponents say, as common as lead, and one which, despite some concerns, would lead to power plants with fewer safety issues as well as other benefits.  Thorium-based reactors certainly have advantages, the energy release from Thorium is greater than from Uranium, the by-products from using Thorium are less toxic than from Uranium, and it's much harder to make weapons from those by-products.

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) , which is an excellent fissile fuel material. 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.

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).
  
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:

• Between 1967 and 1988, the AVR (Atom Versuchs Reaktor, Nuclear Test Reactor) experimental pebble bed reactor at Jülich, Germany, operated for over 750 weeks at 15 MWe, about 95% of the time with thorium-based fuel. The fuel used consisted of about 100,000 billiard ball-sized fuel elements. Overall a total of 1360 kg of thorium was used, mixed with high-enriched uranium (HEU). Burn-ups of 150,000 MWd/t were achieved.

• Thorium fuel elements with a 10:1 Th/U (HEU) ratio were irradiated in the 20 MWth Dragon reactor at Winfrith, UK, 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, from 1964 to 1973. The Th/U fuel was used to 'breed and feed', so that the U-233 formed replaced the U-235 at about the same rate, and fuel could be left in the reactor for about six years.

• General Atomics' Peach Bottom high-temperature, graphite-moderated, helium-cooled reactor in the USA operated between 1967 and 1974 at 110 MWth, using high-enriched uranium with thorium.

• In Canada, AECL has more than 50 years experience with thorium-based fuels, including burn-up to 47 GWd/t. Some 25 tests were performed to 1987 in three research reactors and one pre-commercial reactor (NPD), with fuels ranging from ThO2 to that with 30% UO2, though most were with 1-3% UO2, the U being high-enriched.

• In India, the Kamini 30 kWth experimental neutron-source research reactor using U-233, recovered from ThO2 fuel irradiated in another reactor, started up in 1996 near Kalpakkam. The reactor was built adjacent to the 40 MWt Fast Breeder Test Reactor, in which the ThO2 is irradiated.

• In the Netherlands, an aqueous homogenous suspension reactor operated at 1MWth for three years in the mid-1970s. The HEU/Th fuel was circulated in solution and reprocessing occurred continuously to remove fission products, resulting in a high conversion rate to U-233.

There have also been several experiments with fast neutron reactors.

Current thorium fuel cycle research

Several advanced reactors concepts are currently being developed, including:

• High-temperature gas-cooled reactors (HTGRs) of two kinds: pebble bed and with prismatic fuel elements. The Gas Turbine-Modular Helium Reactor (GT-MHR) being developed by General Atomics uses a prismatic fuel and builds on US experience, particularly from the Fort St Vrain reactor. The GT-MHR core can accommodate a wide range of fuel options, including HEU/Th, U-233/Th and Pu/Th. Pebble bed reactor development builds on German work with the AVR and THTR and is under development in China and South Africa c . A pebble bed reactor can potentially use thorium in its fuel pebbles.

• The Radkowsky Thorium Reactor builds on the seed-blanket arrangement of the LWBR concept. (see subsection below)

• The molten salt reactor (MSR) is an advanced breeder concept, in which the coolant is a molten salt, usually a fluoride salt mixture. This is hot, but not under pressure, and does not boil below about 1400°C. Much research has focused on lithium and beryllium additions to the salt mixture. The fuel can be dissolved enriched uranium, thorium or U-233 fluorides, and recent discussion has been on the Liquid Fluoride Thorium Reactor, utilizing U-233 which has been bred in a liquid thorium salt blanket and continuously removed to be added to the core. The MSR was studied in depth in the 1960s, but is now being revived because of the availability of advanced technology for the materials and components. There is now renewed interest in the MSR concept in China, Japan, Russia, France and the USA, and one of the six Generation IV designs selected for further development is the MSR (see also subsection below and information page on Generation IV Nuclear Reactors).

• CANDU-type reactors – AECL is researching the thorium fuel cycle application to Enhanced Candu 6 and ACR-1000 reactors 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. AECL envisages fleets of CANDU reactors with near-self-sufficient equilibrium thorium (SSET) fuel cycles and a few fast breeder reactors to provide plutonium. AECL is also working closely with Third Qinshan Nuclear Power Company (TQNPC), China North Nuclear Fuel Corporation and Nuclear Power Institute of China (NPIC) at Chengdu to 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. (see also Th in PHWR subsection of R&D section in China Fuel Cycle paper)

• Advanced heavy water reactor (AHWR) – India is working on this and, like the Canadian ACR design, the 300 MWe AHWR design is light water cooled. The main part of the core is subcritical with Th/U-233 oxide and Th/Pu-239 oxide, mixed so that the system is self-sustaining in U-233. The initial core will be entirely Th-Pu-239 oxide fuel assemblies, but as U-233 is available, 30 of the fuel pins in each assembly will be Th-U-233 oxide, arranged in concentric rings. It is designed for 100-year plant life and is expected to utilise 65% of the energy of the fuel. About 75% of the power will come from the thorium.

• Fast breeder reactor (FBRs), along with the AHWRs, play an essential role in India's three-stage nuclear power program (see section on India's plans for thorium cycle below). A 500 MWe prototype FBR under construction in Kalpakkam is designed to breed U-233 from thorium.

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.

Much development work is still required before the thorium fuel cycle can be commercialised, 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.

Source: World Nuclear Association





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