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FUEL CYCLES Once-through fuel cycle Not a cycle ''per se'', fuel is used once and then sent to storage without further processing save repackaging to provide for better isolation from the , Canada , Sweden , Finland , Spain and South Africa . {Link without Title} Some countries, notably Sweden and Canada, have designed repositories to permit future recovery of the material should the need arise, while others plan for permanent sequestration. Plutonium cycle Many countries are using the reprocessing services offered by BNFL and COGEMA . Here, the Fission Product s, Uranium and Plutonium , are separated for disposal or further use. Already BNFL have started to make MOX Fuel which has been supplied to Nuclear Power Plant s in many parts of the world. This use of fuel which was created in a Reactor closes the cycle. Minor actinides recycling It has been proposed that in addition to the use of plutonium, that the Minor Actinides could be used in a critical power reactor. Already tests are being conducted in which Americium is being used as a fuel. {Link without Title} A number of reactor designs, like the Integral Fast Reactor , have been designed for this rather different fuel cycle. In principle, it should be possible to derive energy from the fission of any Actinide nucleus. With a careful reactor design, all the actinides in the fuel can be consumed, leaving only lighter elements with short half-lives. Whereas this has been done in prototype plants, no such reactor has ever been operated on a large scale, and the first plants with full actinide recovery are expected to be ready for commercial deployment in 2015 at the earliest. However, such schemes would most likely require advanced remote reprocessing methods due to the neutron emitting compounds formed. For instance if Curium is irradiated with Neutron s it will form the very heavy actinides Californium and Fermium which undergo Spontaneous Fission . As a result, the Neutron Emission from a used fuel element which had included curium will be much higher, potentially posing a risk to workers at the back end of the cycle unless all reprocessing is done remotely. This could be seen as a disadvantage, but on the other hand it also makes the nuclear material difficult to steal or divert, making it more resistant to Nuclear Proliferation It so happens that the Neutron Cross-section of many actinides decreases with increasing neutron energy, but the ratio of fission to simple activation ( Neutron Capture ) changes in favour of fission as the neutron energy increases. Thus with a sufficiently high neutron energy, it should be possible to destroy even Curium without the generation of the transcurium metals. This could be very desirable as it would make it significantly easier to reprocess and handle the actinide fuel. One promising alternative from this perspective is an accelerator driven sub-critical reactor. Here a beam of either s will be generated. These high-energy neutrons and photons will then be able to cause the fission of the heavy actinides. Such reactors compare very well to other neutron sources in terms of neutron energy:
As an alternative, the curium-244, half life 18 years) could be left to decay into Pu-240 before being used in fuel in a fast reactor. Fuel or targets for this actinide transmutation To date the nature of the fuel (targets) for actinide transformation has not been chosen. If actinides are transmuted in a Sub Critical Reactor it is likely that the fuel will have to be able to tolerate more thermal cycles than conventional fuel. An accelerator driven sub critical reactor is unlikely to be able to maintain a constant operation period for equally long times as a critical reactor, and each time the accelerator stops then the fuel will cool down. On the other hand, if actinides are destroyed using a fast reactor, such as an Integral Fast Reactor , then the fuel will most likely not be exposed to many more thermal cycles than in a normal power station. Depending on the matrix the process can generate more transuranics from the matrix, this could either be viewed as good (generate more fuel) or can be viewed as bad (generation of more ''radiotoxic'' Transuranic Element s). A series of different matrices exist which can control this production of heavy actinides. =Actinides in an inert matrix The actinide will be mixed with a metal which will not form more actinides, for instance an alloy of actinides in a solid such as Zirconia could be used. =Actinides in a thorium matrix Thorium will on neutron bombardment form Uranium-233 . U-233 is fissile, and has a larger fission cross section than both U-235 and U-238, and thus it is likely to produce very little additional actinides through neutron capture. =Actinides in a uranium matrix If the actinides is incorporated into a uranium-metal or uranium-oxide matrix, then the neutron capture of U-238 is likely to generate new Plutonium-239 . An advantage of mixing the actinides with Uranium and Plutonium is that the large fission cross sections of U-235 and Pu-239 for the less energetic Delayed-neutrons could make the reaction stable enough to be carried out in a critical Fast Reactor , which is likely to be both cheaper and simpler than an accelerator driven system. Thorium cycle In the thorium fuel cycle Thorium-232 absorbs a Neutron in either a fast or thermal reactor. The thorium-233 Beta Decay s to Protactinium -233 and then to Uranium-233 , which in turn is used as fuel. Hence, like Uranium-238 , thorium-232 is a Fertile Material . After starting the reactor with existing U-233 or some other Fissile Material such as U-235 or Pu-239 , a breeding cycle similar to but more efficient than that with U-238 and plutonium can be created. The Th-232 absorbs a neutron to become Th-233 which quickly decays to Protactinium -233. Protactinium-233 in turn decays with a half-life of 27 days to U-233. In some Molten Salt Reactor designs, the Pa-233 is extracted and protected from neutrons (which could transform it to Pa-234 and then to U-234 ), until it has decayed to U-233. This is done in order to improve the Breeding Ratio . Uranium-233 is an excellent reactor fuel. Uranium-233 is superior to uranium-235 and plutonium-239 because it produces more neutrons per neutron absorbed (it has a high "beta" coefficient). Its absorption of neutrons ( Cross-section ) also varies less with temperature and Neutron Energy than plutonium-239 or U-235. This stability means high Burnup s and higher operating temperatures, so the conversion to electricity is more efficient, with thermal yields of 50-55%. When U-233 absorbs a neutron, it either fissions or becomes the next heavier isotope, U-234. The chance of not fissioning on absorption of a Thermal Neutron is about 1/7, less than the corresponding probabilities for U-235 (about 1/6) or for Pu-239 or Pu-241 (about 1/4). U-234, like most even-mass isotopes, is not easily fissionable with slow neutrons, but further neutron capture produces fissile U-235. Further failure to fission on neutron capture will produce Uranium-236 , Neptunium-237 , Pu-238 , and eventually fissile Pu-239 , but production of heavy Transuranic elements is far less than in the uranium / plutonium cycle, because most thorium cycle fuel nuclei fission before attaining the mass of the starting points of the uranium / plutonium cycle. On the other hand, the thorium cycle produces some Protactinium -231 (halflife 33,000 years) via the (n,2n) reaction on Th-232. Because the thorium / uranium-233 cycle produces a smaller amount of long-lived Actinide isotopes, the long-term Radioactivity of the Spent Nuclear Fuel is less. Common Fission Products have half-lives up to 30 years ( Sr-90 , Cs-137 ) or more than 200,000 years ( Tc-99 ), and radioactivity in the period intermediate between these two scales is chiefly from Actinide wastes. Another positive, if a solid-fuel reactor is used, is that Thorium Dioxide melts around 3300 °C compared to 2800 °C for Uranium Dioxide . cycle. {Link without Title} {Link without Title} Current industrial activity Currently the only isotopes used as nuclear fuel are Uranium-235 (U-235), Uranium-238 (U-238) and Plutonium-239 , although the proposed thorium fuel cycle has advantages. Some modern reactors, with minor modifications, can use Thorium . Thorium is approximately three times more abundant in the Earth's crust than all forms of uranium combined. However, there has been little exploration for thorium resources, and thus the proved resource is small. Thorium is more plentiful than Uranium in some countries, notably India . {Link without Title} Heavy Water Reactor s and graphite-moderated reactors can use Natural Uranium , but the vast majority of the world's reactors require Enriched Uranium , in which the ratio of U-235 to U-238 is increased. In civilian reactors the enrichment is increased to as much as 5% U-235 and 95% U-238, but in Naval Reactors there is as much as 93% U-235. The term '' Nuclear Fuel '' is not normally used in respect to Fusion Power , which fuses Isotope s of Hydrogen into Helium to release Energy . FRONT END See Also: Uranium mining |
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