Once-through nuclear fuel cycle
Not a cycle per se, fuel is used once and then sent to storage without further processing save additional packaging to provide for better isolation from the biosphere. This method is favored by six countries: the United States, Canada, Sweden, Finland, Spain and South Africa.[1] 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 in a geological repository like Yucca Mountain in the United States.
Plutonium cycle
Several countries are using the reprocessing services offered by BNFL and COGEMA. Here, the fission products, minor actinides, activation products, and reprocessed uranium are separated from the reactor-grade plutonium, which can then be fabricated into MOX fuel. Because the proportion of the non-fissile even-mass isotopes of plutonium rises with each pass through the cycle, there are currently no plans to reuse plutonium from used MOX fuel for a third pass in a thermal reactor. However, if fast reactors become available, they may be able to burn these, or almost any other actinide isotopes.
Minor actinides recycling
It has been proposed that in addition to the use of plutonium, the minor actinides could be used in a critical power reactor. Tests are already being conducted in which americium is being used as a fuel.[2]
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 neutrons 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 subcritical reactor. Here a beam of either protons (United States and European designs)[3][4][5] or electrons (Japanese design)[6] is directed into a target. In the case of protons, very fast neutrons will spall off the target, while in the case of the electrons, very high energy photons 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:
- Thermal 0 to 100 eV
- Epithermal 100 eV to 100 KeV
- Fast (from nuclear fission) 100 KeV to 3 MeV
- DD fusion 2.5 MeV
- DT fusion 14 MeV
- Accelerator driven core 200 MeV (lead driven by 1.6 GeV protons)
- Muon-catalyzed fusion 7 GeV.
As an alternative, the curium-244, with a half life of 18 years, could be left to decay into plutonium-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 Subcritical 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 elements). A series of different matrices exist which can control this production of heavy actinides.
Fissile nuclei, like Uranium-235, Plutonium-239 and Uranium-233 respond well to delayed neutrons and are thus important to keep a critical reactor stable, and this limits the amount of minor actinides that can be destroyed in a critical reactor. As a consequence it is important that the chosen matrix allows the reactor to keep the ratio of fissile to non-fissile nuclei high, as this enables it to destroy the long lived actinides safely. In contrast, the power output of a sub-critical reactor is limited by the intensity of the driving particle accelerator, and thus it need not contain any uranium or plutonium at all. In such a system it may be preferable to have an inert matrix that doesn't produce additional long-lived isotopes.
Actinides in an inert matrix
The actinides 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 are 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.
Mixed matrix
It is also possible to create a matrix made from a mix of the above mentioned materials. This is most commonly done in fast reactors where one may wish to keep the breeding ratio of new fuel high enough to keep powering the reactor, but still low enough that the generated actinides can be safely destroyed without transporting them to another site. One way to do this is to use fuel where actinides and uranium is mixed with inert zirconium, producing fuel elements with the desired properties.
Thorium cycle
In the thorium fuel cycle thorium-232 absorbs a neutron in either a fast or thermal reactor. The thorium-233 beta decays 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[7] 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 which is low compared to fast reactors.
Thorium is at least 4-5 times more abundant in nature than all of uranium isotopes combined; thorium is fairly evenly spread around Earth with a lot of countries[8] having huge supplies of it; preparation of thorium fuel does not require difficult [7] and expensive enrichment process; thorium fuel cycle creates mainly Uranium-233 contaminated with Uranium-232 which makes it harder to use in a nuclear weapon; elimination of at least the transuranic portion of the nuclear waste problem is possible in MSR and other breeder reactor designs.
One of the earliest efforts to use a thorium fuel cycle took place at Oak Ridge National Laboratory in the 1960s. An experimental reactor was built based on molten salt reactor technology to study the feasibility of such an approach, using thorium fluoride salt kept hot enough to be liquid, thus eliminating the need for fabricating fuel elements. This effort culminated in the Molten-Salt Reactor Experiment that used 232Th as the fertile material and 233U as the fissile fuel. Due to a lack of funding, the MSR program was discontinued in 1976.
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 uranium (and 550 times more abundant than uranium-235). 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.[9]
Heavy water reactors 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 isotopes of hydrogen into helium to release energy.
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