Breeder Reactors

Even more intriguing is the possibility of actually producing more fuel than is consumed, using the so-called "breeder reactor". In this reactor, which is based on fast neutrons, more fissile material is produced through the conversion of 238U into 239Pu than is consumed through fission (Fig. 8.19). As fast neutrons are desired, no moderator is necessary in this type of reactor. Because it does not appreciably slow down neutrons and is an excellent heat-transfer material, liquid sodium is typically used as the coolant. Several fast breeder reactors have been constructed over the years in a number of countries. The best known are France's "Phénix" and "Superphénix" reactors. Phénix is an experimental reactor with a power capacity of 250 MW. Its successor, Superphénix, was a commercial-scale reactor that was connected to the grid and had a capacity of 1300 MW, but it was closed down in 1997, after some technical problems, but mainly for administrative and political concerns [80]. In Russia, Japan and India, breeder reactors are well under development with several units under construction. The United States, which pioneered this field with the construction of the first breeder reactor in 1951 at the Argonne National Laboratory [81], is now reconsidering this technology for future power plants. In fact, the United States initiated in 2000 the Generation IV International Forum (GIF) with nine other countries aimed at determining and developing the most promising generation IV reactor designs that can provide future worldwide needs for electricity, and also produce hydrogen and other products. Among the six selected systems, three are fast-neutron reactors which are all operated in a close fuel cycle, meaning that the fuel is recycled [79].

an electron is ejected while a neutron is transformed into a proton ((3" radiation)

an electron is ejected while a neutron is transformed into a proton ((3" radiation)

Half-life: 24 min Half-life: 2 days

Figure 8.19 Uranium 238 fertilization by neutron capture.

Half-life: 24 min Half-life: 2 days

Figure 8.19 Uranium 238 fertilization by neutron capture.

About 4.6 million tonnes of proven uranium reserves are estimated to be exploitable at costs below $130 kg-1. At a current worldwide demand of 60000 t per year, this represents about 75 years of consumption. Based on geological surveys, potential resources exploitable at a cost of $130 kg-1 uranium are estimated at an equivalent of 280 years of consumption [82]. With present low prices of around $20-30 kg-1 and still considerable reserves, there is nevertheless little incentive to open new mines or to start explorations in the quest of uranium. However, if nuclear power were to supply a much larger part of our energy demand, in the long term the transition to breeder reactors is necessary. All the uranium could then be used as fuel, and not only the 0.7% of 235U contained in natural uranium, multiplying the reserves by a factor of roughly 100 and fulfilling our energy needs for a least thousand years. Eventually, the 4 billion tonnes of uranium contained in the oceans in the form of carbonates could also be exploited. Although very dilute (3-4 mg m-3) and with an estimated extraction price of up to $1000 kg-1, it could become a viable source because the cost of uranium is only a minor component of the price of electricity generated by breeder reactors. This would leave us with an almost unlimited source of energy [83].

Besides 238U, another "fertile" isotope which is susceptible to be transformed into fissionable atoms exists in nature, namely thorium 232 (232Th). Almost 30 isotopes of thorium are known, but only 232Th (which has a half-life of 14 billion years) is present naturally. When bombarded with neutrons, 232Th becomes 233Th, which eventually decays to

233U. 233U

is a fissionable material with similar properties to 235U, and can be used as nuclear fuel. Thorium is considered to be about three times more abundant than uranium, and so the potential energy available from its exploitation is therefore tremendous. The use of thorium for electricity generation has been studied and successfully demonstrated in several reactor prototypes, including the high-temperature gas-cooled reactor (HTGR) and the molten salt reactor (MSR). The MSR was first considered by the United States in the early 1950s for aircraft propulsion. Using a 232Th-233U fuel cycle, MSR is now one of the six systems selected by the Generation IV International Forum on future nuclear reactors. Countries such as India, with limited deposits of uranium but large thorium resources, are particularly interested in such a technology. The use of thorium fuels, compared to uranium fuels, also produce much less plutonium and other actinides, so that induced radiotoxicity is considerably reduced. This is due to the fact that 232Th has two protons and four neutrons less than 238U, which makes it less probable for 232Th through a succession of neutron captures to be transformed into long-lived toxic transuranic actinides such as neptunium, americium, curium, or plutonium. From the viewpoint of decreased nuclear waste production, the thorium-based fuel cycle is thus more desirable.

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