Nuclear fuel cycle

From Free net encyclopedia

Template:TOCleft The nuclear fuel cycle, also called nuclear fuel chain, 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 radioactive waste.

Contents 1 Different fuel cycles 1.1 Once-through fuel cycle 1.2 Plutonium cycle 1.3 Minor actinides recycle 1.4 The Thorium fuel cycle 1.5 Current industrial activity 2 Front end 2.1 Exploration 2.2 Mining 2.3 Milling 2.4 Uranium conversion 2.5 Enrichment 2.6 Fabrication 3 Service period 3.1 Transport of Radioactive Materials 3.2 In-core fuel management 3.3 On-Load Reactors 4 Back end 4.1 Interim Storage 4.2 Reprocessing 4.3 Waste disposal 5 See also 6 External links

Different fuel cycles

Once-through fuel cycle

Technically 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:U.S.; 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.

Image:Oncethroughcycle.jpg

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 power reactors in many parts of the world. This use of fuel which was created in a reactor closes the cycle. Image:Plutrecyclefuelcycle.jpg

Minor actinides recycle

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] But it is important to note that neutron bombardment in even a fast reactor is not a suitable method for 'burning' all the transplutonics. 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.

A number of reactor designs (for example, 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. No such reactor has ever been operated on a large scale.

It is vital for transmutation of the transplutonium metals the neutron energy should be high. Even the neutron energy in a fast breeder reactor (which can be used as a fast burner if operated with a different fuel) might not be high enough. One alternative to a critical reactor where neutrons are generated by the fission of actinide nuclei is an accelerator driven sub-critical reactor. Here a beam of either protons (US 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. It so happens that the neutron cross section of many actinides decreases with increasing neutron energy, but the ratio of fission to simple activation (ng reactions) changes in favour of fission as the neutron energy increases.

Depending on the neutron source the energy will differ.

• 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

Hence it should be possible to destroy even curium without the generation of the transcurium metals if the neutron energy is high, as an alternative the curium (^{244Cm, half life 18 years) could be left to decay into 240Pu before being used in fuel in a fast reactor. (Reference V. Artisyuk, M. Saito and A. Shmelev, Progress in Nuclear Energy, 2000, 37, 345-350)}

Image:Twincyclefuelcycle.jpg

It is likely that the fuel will have to be able to tolerate more thermal cycles than conventional fuel, this is because if the accelerator is likely stop working on a regular basis. Each time the accelerator stops then the fuel will cool down, it is normal in many conventional power reactors to run the plant at full power for weeks or months at a time, rather than switching it on and off each day.

• Fuel or targets for this actinide transmutation

To date the nature of the fuel (targets) for actinide transformation has not been chosen.

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 matrixs 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 actindies, for instance a solid solution of an actinide in a solid such as zirconia could be used.

• • Actinides in thorium oxide

The actinide oxide when mixed with thorium oxide will on neutron bombardment form ^{233U (While is fissile), it is likely that the 233U on further neutron bombardment would undergo fission and it is unlikely that the transuranium elements will be generated from the matrix.}

• • Actinides in uranium oxide

This is likely to lead to the generation of new ^239Pu.

The Thorium fuel cycle

The Thorium fuel cycle has thorium absorbing a slow neutron (in a reactor) to ultimately form Uranium-233; which in turn is burned as fuel. Hence like uranium-238 it is a fertile material.

As a fuel, U-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 oxide 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. Finally all of the mineable thorium is potentially usable in a reactor, compared with the 0.7% of natural uranium, so some 40 times the amount of energy per unit mass might be available.

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.

References: Thorium Fuel Links and Perspectives of the Thorium Fuel Cycle

Current industrial activity

Currently the only isotopes used as nuclear fuel are Uranium U²³⁵, Uranium U²³⁸ and Plutonium Pu²³⁹, although the proposed thorium fuel cycle has advantages. Some modern reactors, with minor modifications, can use thorium, which is more plentiful than uranium.

Heavy-water reactors and graphite-moderated reactors can use uranium as it is mined and refined, but the vast majority of the world's reactors require that the ratio of Uranium-235 (U²³⁵) to Uranium-238 (U²³⁸) be increased. In civilian reactors the enrichment is increased to as much as 5% U²³⁵ and 95% U²³⁸, but in naval reactors there is as much as 93% U²³⁵.

The term nuclear fuel is not normally used in respect to fusion power, which fuses isotopes of hydrogen into helium to release energy.

Front end

Template:Main

1 Uranium ore - the principal raw material of nuclear fuel

2 Yellowcake - the form in which uranium is transported to an enrichement plant

3 UF6 - used in enrichment

4 Nuclear fuel - a compact, inert, insoluble solid

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, U²³⁸ and U²³⁵. 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% U²³⁸ and 0.71% U²³⁵. The atomic nucleus of U²³⁵ 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 U²³⁸ 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 U²³⁹. This isotope then undergoes natural radioactive decay to yield Pu²³⁹, which, like U²³⁵, is a fissile isotope. The atoms of U²³⁸ are said to be fertile, because, through neutron irradiation in the core, some eventually yield atoms of fissile Pu²³⁹.

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.

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 temperature of 134°F (57°C). The UF₆ conversion product contains only natural, not enriched, uranium.

U₃O₈ is also converted to ceramic grade 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.

Image:UF6solid.jpg

Enrichment

Template:Main. Image:Nuclear Fuel Cycle.png

The concentration of the fissionable isotope, U²³⁵ (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 up to about 5% U²³⁵, 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), for which there are few applications; the United States Department of Energy alone has 470,000 tonnes in store [7].

Fabrication

Template:Main 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.

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 (BWRs and PWRs) the tubes are assembled into bundles (see picture in [8]) 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.

Service period

Transport of Radioactive Materials

Transport is an integral part of the nuclear fuel cycle. There are nuclear power reactors in operation in several countries but uranium mining is viable in only a few areas. Also, in the course of over forty years of operation by the nuclear industry, a number of specialized facilities have been developed in various locations around the world to provide fuel cycle services and there is a need to transport nuclear materials to and from these facilities. Most transports of nuclear fuel material occur between different stages of the cycle, but occasionally a material may be transported between similar facilities. With some exceptions, nuclear fuel cycle materials are transported in solid form, the exception being uranium hexafluoride (UF₆) which is considered a gas. Most of the material used in nuclear fuel is transported several times during the cycle. Transports are frequently international, and are often over large distances. Nuclear materials are generally transported by specialized transport companies.

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.

In-core fuel management

The core of a reactor 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 (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 fourth) 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.

On-Load Reactors

Some reactor designs, such as CANDUs or RBMKs, can be refuelled without being shut down. This is achieved through the use of many small pressure tubes to contain the fuel and coolant, as opposed to one large pressure vessel as in PWR or BWR designs. Each tube can be individually isolated and refuelled by an operator-controlled fuelling machine, typically at a rate of up to 8 channels per day (out of roughly 400) in CANDU reactors. On-Load refuelling allows for the problem of 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 fuelling machines to service them.

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, 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, 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 a period of cooling they may be moved to dry cask storage.

Reprocessing

Template:Main See also used nuclear fuel. Image:Sellafield-1515b.jpg

Spent fuel discharged from reactors contains appreciable quantities of fissile (U²³⁵, Pu²³⁹), fertile (U²³⁸), and other radioactive materials, including reaction poisons (the reason 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.

Mixed oxide, or MOX fuel, is a blend of reprocessed uranium and plutonium and depleted uranium which behaves similarly (though 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 not permitted in the United States due to nonproliferation considerations. 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.

Waste disposal

Template:Main

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 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 as per above.

See also

• Nuclear fuel
• Deep geological repository
• Enrico Fermi
• Global Nuclear Energy Partnership announced February, 2006
• Manhattan Project
• Nuclear fission
• Nuclear reactor
• Nuclear physics
• Nuclear power plant
• Nuclear proliferation
• Spent nuclear fuel shipping cask
• United States Naval reactor

External links

• Annotated bibliography on the nuclear fuel cycle from the Alsos Digital Library
• Uranium.Info publishing uranium price since 1968.
• Reprocessing Facilities
• NFCC - Nuclear Fuel Cycle Consulting
• In-core fuel management
• Nuclear Fuel Energy Balance Calculator designed by: the WISE Uranium Project
• An optimality condition for the assembly distribution in a nuclear reactor
• Nuclear Files.org Information regarding the nuclear fuel cycle, including the technology, relation to NPT, proliferation, and controls.
• BBC: The Nuclear Fuel Cyclede:Brennstoffkreislauf

es:Ciclo del combustible nuclear fa:چرخه سوخت هسته‌ای fr:Cycle du combustible nucléaire ru:Ядерное топливо fi:Ydinpolttoainekierto