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. 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.
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.
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) or electrons (Japanese design) 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:
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.
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.
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.
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.
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.
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. About 95% of depleted uranium is stored as uranium hexafluoride (UF6).
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 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.
Since nuclear materials are radioactive, it is important to ensure that radiation exposure of both those involved in the transport of such materials and the general public along transport routes is limited. Packaging for nuclear materials includes, where appropriate, shielding to reduce potential radiation exposures. In the case of some materials, such as fresh uranium fuel assemblies, the radiation levels are negligible and no shielding is required. Other materials, such as spent fuel and high-level waste, are highly radioactive and require special handling. To limit the risk in transporting highly radioactive materials, containers known as spent nuclear fuel shipping casks are used which are designed to maintain integrity under normal transportation conditions and during hypothetical accident conditions.
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.
Uranium dioxide is very insoluble in water, but after oxidation it can be converted to uranium trioxide or another uranium(VI) compound which is much more soluble. It is important to understand that uranium dioxide (UO2) can be oxidised to an oxygen rich hyperstoichiometric oxide (UO2+x) which can be further oxidised to U4O9, U3O7, U3O8 and UO3.2H2O.
Because used fuel contains alpha emitters (plutonium and the minor actinides), the effect of adding an alpha emitter (238Pu) to uranium dioxide on the leaching rate of the oxide has been investigated. For the crushed oxide, adding 238Pu tended to increase the rate of leaching, but the difference in the leaching rate between 0.1 and 10% 238Pu was very small.
The concentration of carbonate in the water which is in contact with the used fuel has a considerable effect on the rate of corrosion, because uranium(VI) forms soluble anionic carbonate complexes such as [UO2(CO3)2]2- and [UO2(CO3)3]4-. When carbonate ions are absent, and the water is not strongly acidic, the hexavalent uranium compounds which form on oxidation of uranium dioxide often form insoluble hydrated uranium trioxide phases.
By ‘sputtering’, using uranium metal and an argon/oxygen gas mixture, thin films of uranium dioxide can be deposited upon gold surfaces. These gold surfaces modified with uranium dioxide have been used for both cyclic voltammetry and AC impedance experiments, and these offer an insight into the likely leaching behaviour of uranium dioxide.
The nuclear chemistry associated with the nuclear fuel cycle can be divided into two main areas, one area is concerned with operation under the intended conditions while the other area is concerned with maloperation conditions where some alteration from the normal operating conditions has occurred or (more rarely) an accident is occurring.
The releases of radioactivity from normal operations are the small planned releases from uranium ore processing, enrichment, power reactors, reprocessing plants and waste stores. These can be in a different chemical/physical form to the releases which could occur under accident conditions. In addition the isotope signature of a hypothetical accident may be very different to that of a planned normal operational discharge of radioactivity to the environment.
It is important to note that just because a radioisotope is released it does not mean it will enter a human and then cause harm. For instance the migration of radioactivity can altered by the binding of the radioisotope to the surfaces of soil particles. For example caesium binds tightly to clay minerals such as illite and montmorillonite hence it remains in the upper layers of soil where it can be accessed by plants with shallow roots (such as grass). Hence grass and mushrooms can carry a considerable amount of 137Cs which can be transferred to humans through the food chain. But 137Cs is not able to migrate quickly through most soils and thus is unlikely to contaminate well water. It is important to note that colloids of soil minterals can migrate through soil so simple binding of a metal to the surfaces of soil particles does not fix the metal totally.
According to Jiří Hála's text book the distribution coefficient Kd is the ratio of the soil's radioactivity (Bq g-1) to that of the soil water (Bq ml-1). If the radioactivity is tightly bonded to by the minerals in the soil then less radioactivity can be absorbed by crops and grass growing on the soil.
One of the best countermeasures in dairy farming against 137Cs is to mix up the soil by deeply ploughing the soil. This has the effect of putting the 137Cs out of reach of the shallow roots of the grass, hence the level of radioactivity in the grass will be lowered. Also after a nuclear war or serious accident the removal of top few cm of soil and its burial in a shallow trench will reduce the long term gamma dose to humans due to 137Cs as the gamma photons will be attenuated by their passage through the soil.
Even after the radioactive element arrives at the roots of the plant, the metal may be rejected by the biochemistry of the plant. The details of the uptake of 90Sr and 137Cs into sunflowers grown under hydroponic conditions has been reported. The caesium was found in the leaf veins, in the stem and in the apical leaves. It was found that 12% of the caesium entered the plant, and 20% of the strontium. This paper also reports details of the effect of potassium, ammonium and calcium ions on the uptake of the radioisotopes.
In livestock farming an important countermeasure against 137Cs is to feed to animals a little prussian blue. This iron potassium cyanide compound acts as a ion-exchanger. The cyanide is so tightly bonded to the iron that it is safe for a human to eat several grams of prussian blue per day. The prussian blue reduces the biological half life (different from the nuclear half life) of the caesium. The physical or nuclear half life of 137Cs is about 30 years. This is a constant which can not be changed but the biological half life is not a constant. It will change according to the nature and habits of the organism for which it is expressed. Caesium in humans normally has a biological half life of between one and four months. An added advantage of the prussian blue is that the caesium which is stripped from the animal in the droppings is in a form which is not available to plants. Hence it prevents the caesium from being recycled. The form of prussian blue required for the treatment of humans or animals is a special grade. Attempts to use the pigment grade used in paints have not been successful. Note that a good source of data on the subject of caesium in Chernobyl fallout exists at , this is the Ukrainian Research Institute for Agricultural Radiology.
The IAEA assume that under normal operation the coolant of a water cooled reactor will contain some radioactivity but during a reactor accident the coolant radioactivity level may rise. The IAEA state that under a series of different conditions different amounts of the core inventory can be released from the fuel, the four conditions the IAEA consider are normal operation, a spike in coolant activity due to a sudden shutdown/loss of pressure (core remains covered with water), a cladding failure resulting in the release of the activity in the fuel/cladding gap (this could be due to the fuel being uncovered by the loss of water for 15-30 minutes where the cladding reached a temperature of 650-1250 oC) or a melting of the core (the fuel will have to be uncovered for at least 30 minutes, and the cladding would reach a temperature in excess of 1650 oC).
Based upon the assumption that a PWR contains 300 tons of water, and that the activity of the fuel of a 1 GWe reactor is as the IAEA predict, then the coolant activity after an accident such as the three mile island accident where a core is uncovered and then recovered with water then the resulting activity of the coolant can be predicted.
It is normal to allow used fuel to stand after the irradiation to allow the shortlived and radiotoxic iodine isotopes to decay away, in one experiment in the USA fresh fuel which had not been allowed to decay was reprocessed (the Green run ) to investigate the effects of a large iodine release from the reprocessing of short cooled fuel. It is normal in reprocessing plants to scrub the off gases from the dissolver to prevent the emission of iodine. In addition to the emission of iodine the noble gases and tritium are released from the fuel when it is dissolved, it has been proposed that by voloxidation (heating the fuel in a furnace under oxidizing conditions) the majority of the tritium can be recovered from the fuel.
A paper was written on the radioactivity found in oysters found in the Irish Sea, these were found by gamma spectrscopy to contain 141Ce, 144Ce, 103Ru, 106Ru, 137Cs, 95Zr and 95Nb. In addition a zinc activation product (65Zn) was found, this is thought to be due to the corrosion of magnox fuel cladding in cooling ponds. It is likely that the modern releases of all these isotopes from Windscale is smaller.
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.
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