It hasn't been pursued because those countries that developed enrichment technology already had the means to make a fissile form, 235U, from uranium ore. That was an easier go than breeding 233U from 232Th, for which you need an operating reactor anyway. So it was easier to stick with the uranium fuel cycle than switch over to 232Th-233U. As the article notes, you have other complications, either the need to use Pu and U in a mixture to juice up the neutron population, or one heck of a heavy duty accelerator. Speaking as one who spent a considerable amount of time dealing with accelerator physics in an "earlier life", I can tell you that the headaches associated with keeping a high beam current accelerator running are in some ways worse than dealing with fission products.
Thorium
The use of thorium-based fuel cycles has been studied for about 30 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 burnups 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 experimental pebble bed reactor at Julich, 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). Maximum burnups 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 (HTGR) in the USA operated between 1967 and 1974 at 110 MWth, using high-enriched uranium with thorium.
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 has operated at 1MWth for three years. The HEU/Th fuel is circulated in solution and reprocessing occurs continuously to remove fission products, resulting in a high conversion rate to U-233.
There have been several experiments with fast neutron reactors.
Power 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:
The 300 MWe THTR reactor in Germany was developed from the AVR and operated between 1983 and 1989 with 674,000 pebbles, over half containing Th/HEU fuel (the rest graphite moderator and some neutron absorbers). These were continuously recycled on load and on average the fuel passed six times through the core. Fuel fabrication was on an industrial scale.
The Fort St Vrain reactor was the only commercial thorium-fuelled nuclear plant in the USA, also developed from the AVR in Germany, and operated 1976 - 1989. It was a high-temperature (700°C), graphite-moderated, helium-cooled reactor with a Th/HEU fuel designed to operate at 842 MWth (330 MWe). The fuel was in microspheres of thorium carbide and Th/U-235 carbide coated with silicon oxide and pyrolytic carbon to retain fission products. It was arranged in hexagonal columns (’prisms’) rather than as pebbles. Almost 25 tonnes of thorium was used in fuel for the reactor, and this achieved 170,000 MWd/t burn-up.
Thorium-based fuel for Pressurised Water Reactors (PWRs) was investigated at the Shippingport reactor in the USA using both U-235 and plutonium as the initial fissile material. It was concluded that thorium would not significantly affect operating strategies or core margins. The light water breeder reactor (LWBR) concept was also successfully tested here from 1977 to 1982 with thorium and U-233 fuel clad with Zircaloy using the ‘seed/blanket’ concept.
The 60 MWe Lingen Boiling Water Reactor (BWR) in Germany utilised Th/Pu-based fuel test elements.
India
In India, both Kakrapar-1 and -2 units are loaded with 500 kg of thorium fuel in order to improve their operation when newly-started. Kakrapar-1 was the first reactor in the world to use thorium, rather than depleted uranium, to achieve power flattening across the reactor core. In 1995, Kakrapar-1 achieved about 300 days of full power operation and Kakrapar-2 about 100 days utilising thorium fuel. The use of thorium-based fuel was planned in Kaiga-1 and -2 and Rajasthan-3 and -4 (Rawatbhata) reactors.
With about six times more thorium than uranium, India has made utilisation of thorium for large-scale energy production a major goal in its nuclear power program, utilising a three-stage concept:
Pressurised Heavy Water Reactors (PHWRs, elsewhere known as CANDUs) fuelled by natural uranium, plus light water reactors, produce plutonium.
Fast Breeder Reactors (FBRs) use this 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 (ideally high-fissile Pu) is produced as well as the U-233. Then
Advanced Heavy Water Reactors burn the U-233 and this plutonium with thorium, getting about 75% of their power from the thorium.
The spent fuel will then be reprocessed to recover fissile materials for recycling.
This Indian program 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.
Another option for the third stage, while continuing with the PHWR and FBR programs, is the subcritical Accelerator-Driven Systems (ADS).
http://www.world-nuclear.org/info/inf62.html