Nuclear fuel cycle
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The nuclear fuel cycle, also called nuclear fuel chain, is the progression of nuclear fuel through a series of differing stages. It consists of steps in the front end, which are the preparation of the fuel, steps in the service period in which the fuel is used during reactor operation, and steps in the back end, which are necessary to safely manage, contain, and either reprocess or dispose of spent nuclear fuel. If spent fuel is not reprocessed, the fuel cycle is referred to as a open fuel cycle (or a once-through fuel cycle). Likewise, if the spent fuel is reprocessed, it is referred to as a closed fuel cycle.
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[edit] Fuel cycles
[edit] 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 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.
[edit] Plutonium cycle
Many countries are using the reprocessing services offered by BNFL and COGEMA. Here, the fission products, 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 plants in many parts of the world. This use of fuel which was created in a reactor closes the cycle.
[edit] 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.[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 sub-critical reactor. Here a beam of either protons (United States and European designs)[3][4] or electrons (Japanese design)[5] 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, half life 18 years) could be left to decay into Pu-240 before being used in fuel in a fast reactor.
[edit] 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 elements). A series of different matrices exist which can control this production of heavy actinides.
[edit] Actinides in an inert matrix
The actinide will be mixed with a metal which will not form more actindies, for instance an alloy of actinides in a solid such as zirconia could be used.
[edit] 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.
[edit] 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.
[edit] Thorium cycle
The thorium fuel cycle has thorium-232 absorbing a neutron under neutron bombardment in either a fast or thermal reactor. The thorium-233 then forms uranium-233 through two beta decays; which in turn is burned as fuel. Hence, like uranium-238, thorium-232 is a fertile material.
As a fuel, uranium-233 is superior to uranium-235 and plutonium-239 from a neutronic standpoint, because of its higher neutron yield per neutron absorbed. Another positive is that thorium dioxide melts around 3300 °C compared to 2800 °C for uranium dioxide. U-233 also keeps its good neutronic properties with high temperatures, better than either U-235 or Pu-239. This stability means high burn-ups and higher operating temperatures, with thermal yields of 50-55%. Also, from the respective position of uranium and thorium on the periodic table, the long-lived minor actinides resulting from fission are in much lower quantity with the thorium cycle, especially compared with the plutonium fuel cycle.
After starting the reactor with some other fissile material (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 normally decays to protactinium-233 and then U-233. The irradiated fuel is then discharged from the reactor, the U-233 extracted, then used in another reactor forming a closed fuel cycle.[6][7]
[edit] 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. [8]
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.
[edit] Front end
 1 Uranium ore - the principal raw material of nuclear fuel |
 2 Yellowcake - the form in which uranium is transported to an enrichment plant |
 3 UF6 - used in enrichment |
 4 Nuclear fuel - a compact, inert, insoluble solid |
[edit] Exploration
A deposit of uranium, such as uraninite, discovered by geophysical techniques, is evaluated and sampled to determine the amounts of uranium materials that are extractable at specified costs from the deposit. Uranium reserves are the amounts of ore that are estimated to be recoverable at stated costs. Uranium in nature consists primarily of two isotopes, U-238 and U-235. The numbers refer to the atomic mass number for each isotope, or the number of protons and neutrons in the atomic nucleus. Naturally occurring uranium consists of approximately 99.28% U-238 and 0.71% U-235. The atomic nucleus of U-235 will nearly always fission when struck by a free neutron, and the isotope is therefore said to be a "fissile" isotope. The nucleus of a U-238 atom on the other hand, rather than undergoing fission when struck by a free neutron, will nearly always absorb the neutron and yield an atom of the isotope U-239. This isotope then undergoes natural radioactive decay to yield Pu-239, which, like U-235, is a fissile isotope. The atoms of U-238 are said to be fertile, because, through neutron irradiation in the core, some eventually yield atoms of fissile Pu-239.
[edit] Mining
Uranium ore can be extracted through conventional mining in open pit and underground methods similar to those used for mining other metals. In situ leach mining methods also are used to mine uranium in the United States. In this technology, uranium is leached from the in-place ore through an array of regularly spaced wells and is then recovered from the leach solution at a surface plant. Uranium ores in the United States typically range from about 0.05 to 0.3% uranium oxide (U3O8). Some uranium deposits developed in other countries are of higher grade and are also larger than deposits mined in the United States. Uranium is also present in very low-grade amounts (50 to 200 parts per million) in some domestic phosphate-bearing deposits of marine origin. Because very large quantities of phosphate-bearing rock are mined for the production of wet-process phosphoric acid used in high analysis fertilizers and other phosphate chemicals, at some phosphate processing plants the uranium, although present in very low concentrations, can be economically recovered from the process stream.
[edit] Milling
Mined uranium ores normally are processed by grinding the ore materials to a uniform particle size and then treating the ore to extract the uranium by chemical leaching. The milling process commonly yields dry powder-form material consisting of natural uranium, "yellowcake," which is sold on the uranium market as U3O8.
[edit] Uranium conversion
Milled uranium oxide, U3O8, must be converted to uranium hexafluoride, UF6, which is the form required by most commercial uranium enrichment facilities currently in use. A solid at room temperature, uranium hexafluoride can be changed to a gaseous form at moderately higher temperature of 134 °F (57 °C). The uranium hexafluoride conversion product contains only natural, not enriched, uranium.
Triuranium octaoxide (U3O8) is also converted directly to ceramic grade uranium dioxide (UO2) for use in reactors not requiring enriched fuel, such as CANDU. The volumes of material converted directly to UO2 are typically quite small compared to the amounts converted to UF6.
[edit] Enrichment
The concentration of the fissionable isotope, U-235 (0.71% in natural uranium) is less than that required to sustain a nuclear chain reaction in light water reactor cores. Natural UF6 thus must be enriched in the fissionable isotope for it to be used as nuclear fuel. The different levels of enrichment required for a particular nuclear fuel application are specified by the customer: light-water reactor fuel normally is enriched to 3.5% U-235, but uranium enriched to lower concentrations also is required. Enrichment is accomplished using some one or more methods of isotope separation. Gaseous diffusion and gas centrifuge are the commonly used uranium enrichment technologies, but new enrichment technologies are currently being developed.
The bulk (96%) of the byproduct from enrichment is depleted uranium (DU), which can be used for armor, kinetic energy penetrators, radiation shielding and ballast. Still, there are vast quantities of depleted uranium in storage. The United States Department of Energy alone has 470,000 tonnes.[9] About 95% of depleted uranium is stored as uranium hexafluoride (UF6).
[edit] Fabrication
For use as nuclear fuel, enriched uranium hexafluoride is converted into uranium dioxide (UO2) powder that is then processed into pellet form. The pellets are then fired in a high temperature sintering furnace to create hard, ceramic pellets of enriched uranium. The cylindrical pellets then undergo a grinding process to achieve a uniform pellet size. The pellets are stacked, according to each nuclear reactor core's design specifications, into tubes of corrosion-resistant metal alloy. The tubes are sealed to contain the fuel pellets: these tubes are called fuel rods. The finished fuel rods are grouped in special fuel assemblies that are then used to build up the nuclear fuel core of a power reactor.
The metal used for the tubes depends on the design of the reactor. Stainless steel was used in the past, but most reactors now use zirconium. For the most common types of reactors, boiling water reactors (BWR) and pressurized water reactors (PWR), the tubes are assembled into bundles[10] with the tubes spaced precise distances apart. These bundles are then given a unique identification number, which enables them to be tracked from manufacture through use and into disposal.
[edit] Service period
[edit] In-core fuel management
A nuclear reactor core is composed of a few hundred "assemblies", arranged in a regular array of cells, each cell being formed by a fuel or control rod surrounded, in most designs, by a moderator and coolant, which is water in most reactors.
Because of the fission process that consumes the fuels, the old fuel rods must be changed periodically to fresh ones (this period is called a cycle). However, only a part of the assemblies (typically one-third) are removed since the fuel depletion is not spatially uniform. Furthermore, it is not a good policy, for efficiency reasons, to put the new assemblies exactly at the location of the removed ones. Even bundles of the same age may have different burn-up levels, which depends on their previous positions in the core. Thus the available bundles must be arranged in such a way that the yield is maximized, while safety limitations and operational constraints are satisfied; Consequently, reactor operators are faced with the so-called optimal fuel reloading problem, which consists in optimizing the rearrangement of all the assemblies, the old and fresh ones, while still maximizing the reactivity of the reactor core so as to maximise fuel burn-up and minimise fuel-cycle costs.
This is a discrete optimization problem, and computationally infeasible by current combinatorial methods, due to the huge number of permutations and the complexity of each computation. Many numerical methods have been proposed for solving it and many commercial software packages have been written to support fuel management. This is an on-going issue in reactor operations as no definitive solution to this problem has been found and operators use a combination of computational and empirical techniques to manage this problem.
[edit] On-load reactors
Some reactor designs, such as RBMKs or CANDU reactors, can be refueled without being shut down. This is achieved through the use of many small pressure tubes, the fuel channels, to contain the fuel and coolant, as opposed to one large pressure vessel as in pressurized water reactor (PWR) or boiling water reactor (BWR) designs. Each tube can be individually isolated and refueled by an operator-controlled fueling machine, typically at a rate of up to 8 channels per day out of roughly 400 in CANDU reactors. On-load refueling allows for the optimal fuel reloading problem to be dealt with continuously, leading to more efficient use of fuel. This increase in efficiency is partially offset by the added complexity of having hundreds of pressure tubes and the fueling machines to service them.
[edit] Back end
[edit] Interim storage
After its operating cycle, the reactor is shut down for refueling. The fuel discharged at that time (spent fuel) is stored either at the reactor site, commonly in a spent fuel pool or, potentially in a common facility away from reactor sites. If on-site pool storage capacity is exceeded, it may be desirable to store the now cooled aged fuel in modular dry storage facilities known as Independent Spent Fuel Storage Installations (ISFSI) at the reactor site or at a facility away from the site. The spent fuel rods are usually stored in water or boric acid, which provides both cooling, the spent fuel continues to generate decay heat as a result of residual radioactive decay, and shielding to protect the environment from residual ionizing radiation, although after several years of cooling they may be moved to dry cask storage.
[edit] Transportation
[edit] Reprocessing
See also used nuclear fuel.
Spent fuel discharged from reactors contains appreciable quantities of fissile (U-235 and Pu-239), fertile (U-238), and other radioactive materials, including reaction poisons, which is why the fuel had to be removed. These fissile and fertile materials can be chemically separated and recovered from the spent fuel. The recovered uranium and plutonium can, if economic and institutional conditions permit, be recycled for use as nuclear fuel. This is currently not done for civilian spent nuclear fuel in the US.
Mixed oxide, or MOX fuel, is a blend of reprocessed uranium and plutonium and depleted uranium which behaves similarly, although not identically, to the enriched uranium feed for which most nuclear reactors were designed. MOX fuel is an alternative to low-enriched uranium (LEU) fuel used in the light water reactors which predominate nuclear power generation.
Currently, plants in Europe are reprocessing spent fuel from utilities in Europe and Japan. Reprocessing of spent commercial-reactor nuclear fuel is currently not permitted in the United States due to the perceived danger of nuclear proliferation. However the recently announced Global Nuclear Energy Partnership would see the U.S. form an international partnership to see spent nuclear fuel reprocessed in a way that renders the plutonium in it usable for nuclear fuel but not for nuclear weapons.
[edit] Partitioning and transmutation
As an alternative to the disposal of the PUREX raffinate in glass or Synroc, the most radiotoxic elements can be removed through advanced reprocessing. After separation the minor actinides and some long lived fission products can be converted to short-lived isotopes by either neutron or photon irradiation. This is called transmutation.
[edit] Waste disposal
A current concern in the nuclear power field is the safe disposal and isolation of either spent fuel from reactors or, if the reprocessing option is used, wastes from reprocessing plants. These materials must be isolated from the biosphere until the radioactivity contained in them has diminished to a safe level. In the U.S., under the Nuclear Waste Policy Act of 1982 as amended, the Department of Energy has responsibility for the development of the waste disposal system for spent nuclear fuel and high-level radioactive waste. Current plans call for the ultimate disposal of the wastes in solid form in a licensed deep, stable geologic structure called a deep geological repository. The Department of Energy chose Yucca Mountain as the location for the repository. However, its opening has been repeatedly delayed.
[edit] See also
- Deep geological repository
- Nuclear reprocessing
- Enrico Fermi
- Global Nuclear Energy Partnership announced February, 2006
- Manhattan Project
- Nuclear physics
- Nuclear power plant
- Nuclear proliferation
- United States Naval reactor
[edit] References
(Reference V. Artisyuk, M. Saito and A. Shmelev, Progress in Nuclear Energy, 2000, 37, 345-350)
