Nuclear fuel cycle

The nuclear fuel cycle consists of front end steps that lead to the preparation of uranium for use as fuel for reactor operation and back end steps that are necessary to safely manage, prepare, and dispose of the highly radioactive spent nuclear fuel.

A number of reactor designs (for example, the Integral Fast Reactor) would make possible a 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. No such reactor has ever been operated on a large scale.

Some modern reactors, with minor modifications, can also consume thorium, which is more plentiful than uranium.

Front end

Exploration

A deposit of uranium, 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, 238U and 235U. 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 percent 238U and 0.71 percent 235U. The atomic nucleus of 235U 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 238U 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 239U. This isotope then undergoes natural radioactive decay to yield 239Pu, which, like 235U, is a fissile isotope. The atoms of 238U are said to be fertile, because, through neutron irradiation in the core, some eventually yield atoms of fissile 239Pu.

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 percent 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.

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.

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, UF6 can be changed to a gaseous form at moderately higher temperatures. The UF6 conversion product contains only natural, not enriched, uranium.

Enrichment

The concentration of the fissionable isotope, 235U (0.71 percent 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 (see also enriched uranium). The different levels of enrichment required for a particular nuclear fuel application are specified by the customer: light-water reactor fuel normally is enriched up to about 5 percent 235U, 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. The gaseous diffusion process consists of passing the natural UF6 gas feed under high pressure through a series of diffusion barriers (semiporous membranes) that permit passage of the lighter 235UF6 atoms at a faster rate than the heavier 238UF6 atoms. This differential treatment, applied across a large number of diffusion "stages," progressively raises the product stream concentration of 235U relative to 238U. In the gaseous diffusion technology, the separation achieved per diffusion stage is relatively low, and a large number of stages is required to achieve the desired level of isotope enrichment. Because this technology requires a large capital outlay for facilities and it consumes large amounts of electrical energy, it is relatively cost intensive. In the gas centrifuge process, the natural UF6 gas is spun at high speed in a series of cylinders. This acts to separate the 235UF6 and 238UF6 atoms based on their slightly different atomic masses. Gas centrifuge technology involves relatively high capital costs for the specialized equipment required, but its power costs are below those for the gaseous diffusion technology.

New enrichment technologies currently being developed are the atomic vapor laser isotope separation (AVLIS) and the molecular laser isotope separation (MLIS). Each laser-based enrichment process can achieve higher initial enrichment (isotope separation) factors than the diffusion or centrifuge processes can achieve. Both AVLIS and MLIS will be capable of operating at high material throughput rates.

Fabrication

For use as nuclear fuel, enriched UF6 is converted into uranium dioxide (UO2) powder which 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 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.

Back end

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 or, potentially, in a common facility away from reactor sites. If on-site pool storage capacity is exceeded, it may be desirable to store 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, which provides both cooling (the spent fuel continues to generate heat as a result of residual radioactive decay) and shielding (to protect the environment from residual ionizing radiation).

Reprocessing

(See also nuclear reprocessing)

Spent fuel discharged from light-water reactors contains appreciable quantities of fissile (U-235, Pu-239), fertile (U-238), and other radioactive materials. 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. Currently, plants in Europe are reprocessing spent fuel from utilities in Europe and Japan. Chemical processing of the spent fuel material to recover the remaining fractions of fissionable products, 235U and 239Pu, for use in fresh fuel assemblies is technically feasible. Reprocessing of spent commercial-reactor nuclear fuel is not permitted in the United States due to nonproliferation considerations.

Waste disposal

Main article: Radioactive waste

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. 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 licensed deep, stable geologic structures.

One method for making the waste from power reactors less likely to cause an ill effect to humans, and to make the disposal cheaper is to reprocess. Most reprocessing uses the PUREX process which is based on the extraction of uranium and plutonium from nitric acid using a mixture of tributyl phosphate and a hydrocarbon solvent. The hydrocarbon solvent is present as a diluant for the tributyl phosphate (TBP).

The PUREX process can be modified to make a UREX (URanium EXtraction) process which could be used to save space inside high level nuclear waste dumps (Yucca Mountain) by removing the uranium which makes up the vast majority of the mass and volume of used fuel.

Also by adding a second extraction agent (CMPO) the PUREX process can be turned into the TRUEX process this is a process which was invented in the USA, and is designed to remove the transplutonium metals (Am/Cm) from waste. The idea is that by lowering the alpha activity of the waste, the majority of the waste can then be disposed of with greater ease.

See also

• nuclear fission
• nuclear reactor
• power plant
• electricity generation
• nuclear physics
• Enrico Fermi
• Manhattan Project
• United States Naval reactor
• technology assessment
• nuclear proliferation
• Nuclearenergyweb's Nuclear Fuel Cycle page (http://www.geocities.com/nuclearenergyweb/fuelcycle.html)fi:Ydinpolttoainekierto