Nuclear fuel cycle
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2008-05-29T01:45:48Z
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[[WP:UNDO|Undid]] revision 215621127 by [[Special:Contributions/76.224.5.64|76.224.5.64]] ([[User talk:76.224.5.64|talk]])
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 [[nuclear reprocessing|reprocess]] or dispose of [[used nuclear fuel|spent nuclear fuel]]. If spent fuel is not reprocessed, the fuel cycle is referred to as an ''open fuel cycle'' (or a ''once-through fuel cycle''); if the spent fuel is reprocessed, it is referred to as a ''closed fuel cycle''.
== Fuel cycles ==
===Once-through fuel cycle ===
[[Image:Oncethroughcycle.png|500px|thumb|right|A once through (or open) 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]].<ref>{{cite web |url= http://www.iaea.org/Publications/Magazines/Bulletin/Bull401/article6.html |title= Management of Spent Fuel at Nuclear Power Plants |work= IAEA Bulletin |accessdate= 2008-01-15 |coauthors= Dyck, Peter; Crijns, Martin J.}}</ref> 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===
[[Image:Plutrecyclefuelcycle.png|500px|thumb|right|A fuel cycle in which plutonium is used for fuel]]
Several countries are using the [[reprocessing]] services offered by [[BNFL]] and [[COGEMA]]. Here, the [[fission product]]s, [[minor actinides]], [[activation product]]s, 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 and odd numbers|even]]-[[atomic mass|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 reactor]]s become available, they may be able to burn these, or almost any other [[actinide]] [[isotope]]s.
===Minor actinides recycling===
It has been proposed that in addition to the use of plutonium, the [[minor actinide]]s could be used in a critical power reactor. Tests are already being conducted in which [[americium]] is being used as a fuel. <ref>{{cite web |url=http://www.nea.fr/html/pt/docs/iem/jeju02/session3/SessionIII-04.pdf |format= PDF |title= The Preparation of the EFTTRA-T5 Americium Transmutation Experiment |work= Seventh Information Exchange Meeting on Actinide and Fission Product Partitioning and Transmutation |coauthors= Warin D.; Konings R.J.M; Haas D.; Maritin P.; Bonnerot J-M.; Vambenepe G.;Schram R.P.C.;Kuijper J.C.; Bakker K.; Conrad R. |accessdate= 2008-01-15 |year= 2002 |month= October}}</ref>
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-life|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 [[neutron]]s 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 [[proton]]s (United States and European designs)<ref>{{cite web |url= http://www.neutron.kth.se/publications/conference_papers/W_Gudowski_FR202_1.PDF |title= Why Accelerator-Driven Transmutation of Wastes Enables Future Nuclear Power? |last= Gudowski |first= W. |accessdate= 2008-01-15 |work= XX International Linac Conference |year= 2000 |month=August |format= PDF}}</ref><ref>{{cite web |url=http://www.osti.gov/bridge/product.biblio.jsp?osti_id=10165253 |title= An overview of accelerator-driven transmutation technology| date= 1994-08-01 |last= Heighway |first= E. A. |format= PDF |accessdate= 2008-01-15}}</ref><ref>{{cite web| url= http://www.nea.fr/html/ndd/reports/2002/nea3109.html |title= Accelerator-driven Systems (ADS) and Fast Reactors (FR) in Advanced Nuclear Fuel Cycles | format= PDF |work= Nuclear Energy Agency |accessdate =2008-01-15}}</ref> or [[electron]]s (Japanese design)<ref>{{cite web| url= http://www.kfki.hu/~hordosy/ral/ane2005.pdf |format= PDF | title= Concept of a Small-scale Electron Accelerator Driven System for Nuclear Waste Transmutation Part 2. Investigation of burnup |coauthors= Brolly Á.;Vértes P. |work= ScienceDirect |month= March |year=2005 |accessdate= 2008-01-15}}</ref> 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 [[photon]]s 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 [[nuclear fusion|fusion]] 2.5 MeV
* DT fusion 14 MeV
* Accelerator driven core 200 MeV (lead driven by 1.6 GeV [[proton]]s)
* [[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.
[[Image:Twincyclefuelcycle.png|thumb|center|853px|A pair of fuel cycles in which uranium and plutonium are kept separate from the minor actinides. The minor actinide cycle is kept within the green box.]]
====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 element]]s). 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 decay]]s 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 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]] and [[Liquid fluoride 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 [[breeder reactor#breeding ratio|breeding ratio]].
Uranium-233 is an excellent reactor fuel. Uranium-233 is superior to uranium-235 and plutonium-239 because it produces more neutrons per neutron absorbed (it has a high "beta" coefficient). Its absorption of neutrons ([[cross-section]]) also varies less with [[temperature]] and [[neutron energy]] than plutonium-239 or U-235. This stability suggests potential for high [[burnup]], higher operating temperatures, and therefore more efficient conversion of heat to electricity. <ref>{{cite web |url= http://www.touchbriefings.com/pdf/2402/windham.pdf |title= Thorium – Fuelling a Sustainable Future for Nuclear Power}}</ref>
When U-233 absorbs a neutron, it either fissions or becomes the next heavier isotope, U-234. The chance of not fissioning on absorption of a [[thermal neutron]] is about 1/7 (or even less than 10% according to another source), which is less than the corresponding capture/fission ratios for U-235 (about 1/6) or for Pu-239 or [[Pu-241]] (about 1/4). U-234, like most [[actinide]] [[nuclide]]s with an even number of neutrons, is not easily fissionable with slow neutrons, but further neutron capture produces fissile U-235; if this in turn fails to fission on neutron capture, it will produce [[uranium-236]], [[neptunium-237]], [[Pu-238]], and eventually fissile [[Pu-239]]. Thus production of heavy [[transuranic]] [[nuclide]]s (the [[minor actinides]] other than [[neptunium]]) is far less than in the uranium-238/plutonium-239 cycle, because 98-99% of thorium cycle fuel nuclei would fission before reaching even U-236. On the other hand, the thorium cycle produces some [[protactinium]]-231 (half-life 33,000 years) via the (n,2n) reaction on Th-232.
<ref>{{cite web| url= http://www-nds.iaea.org/reports/indc-ccp-416_p1.pdf| title= Status of Nuclear Data for the Thorium Fuel Cycle |page= 4 |quote= Relative number of nuclei n = N(<sup>231</sup>Pa)/N(<sup>233</sup>U): Fast reactor 0.8{{e|-2}} Thermal reactor 1.9{{e|-3}}... The problem is exacerbated by the fact that the secondary heavy nuclei produced in this cycle possess, as a rule, extremely unpleasant nuclear physics characteristics from the experimentalist’s point of view |year= 1997 |coauthors= Kuz’minov B.D.; Manokhin V.N. |accessdate= 2008-01-15}}</ref><ref>{{cite web |url= http://www.freepatentsonline.com/4344912.html |title= Method of increasing the deterrent to proliferation of nuclear fuels - Patent 4344912 |quote= protactinium-231 which is normally found in the spent fuel rods of a thorium base nuclear reactor |accessdate= 2008-01-15}}</ref><ref>{{cite web |url= http://www.indian-nuclear-society.org.in/conf/2003/1.pdf |title= AN OVERVIEW OF R&D IN FUEL CYCLE ACTIVITIES OF AHWR |quote= the higher energy (n, 2n) reactions encountered by Th-232 during the irradiation in Th-U233 fuel also lead to the formation of long lived Pa-231... Pa-231 is of special concern in HLLW of AHWR because the pentavalent Pa-231 is capable to migrate much more in water/soil compared to other ions... the amount of Pa-231 produced in (Th-U233)O2 MOX fuel is ~ 3 gms/te at 20,000 MWd/t of burn-up... removal of protactinium has to be established using suitable solvents that is yet to be tested |author= Bhattacharjee B. |publisher= Bhabha Atomic Research Centre |accessdate= 2008-01-15}}</ref><ref>{{cite web |url= http://lpsc.in2p3.fr/gpr/english/NEWNRW/NEWNRW.html#foot284 |title= Nuclear Energy With (Almost) No Radioactive Waste? |quote=according to computer simulations done at ISN, this Protactinium dominates the residual toxicity of losses at 10 000 years |month= July |year= 2001 |coauthors= Brissot R.; Heuer D.; Huffer E.; Brun, C. Le; Loiseaux, J-M; Nifenecker H.; Nuttin A.}}</ref> Because the thorium/uranium-233 cycle produces a smaller amount of long-lived [[actinide]] isotopes, the long-term [[radioactivity]] of the [[spent nuclear fuel]] is less. Common [[fission products]] have half-lives up to 30 years ([[strontium-90]], [[caesium-137]]) or more than 200,000 years ([[technetium-99]]), and radioactivity in the period intermediate between these two scales is chiefly from [[actinide]] wastes. Another positive, if a solid-fuel reactor is used, is that [[thorium dioxide]] melts around 3,300 °C compared to 2,800 °C for [[uranium dioxide]] cycle. <ref>{{cite web| url= http://npc.sarov.ru/english/digest/142004/appendix9.html |title= Perspectives of the Thorium Fuel Cycle |work= NUCLEAR TECHNOLOGIES AND NON-PROLIFERATION POLICIES |publisher= Analytical Center for Non-Proliferation |year= 2004 |accessdate= 2008-01-15}}</ref>
===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 [[crust (geology)|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]].<ref>{{cite web| url= http://www.npcil.nic.in/nupower_vol11_1-3/chidambaram.htm |title= Towards an Energy Independent India |year= 1997 | author= Dr. Chidambaram R. |work= Nu-Power |publisher= Nuclear Power Corporation of India Limited |accessdate= 2008-01-15}}</ref>
[[Heavy water reactor]]s 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 [[nuclear marine propulsion|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 [[isotope]]s of [[hydrogen]] into [[helium]] to release [[energy]].
==Front end==
{{main|Uranium mining}}
<gallery>
Image:Uranium ore square.jpg|'''1 Uranium ore''' - the principal raw material of nuclear fuel
Image:Yellowcake.jpg|'''2 Yellowcake''' - the form in which uranium is transported to an enrichment plant
Image:UF6 square.jpg|'''3 UF<small>6</small>''' - used in enrichment
Image:Nuclear fuel pellets.jpeg|'''4 Nuclear fuel''' - a compact, inert, insoluble solid
</gallery>
===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 [[proton]]s and [[neutron]]s 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.
===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 (U<sub>3</sub>O<sub>8</sub>). 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 [[fertilizer]]s 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 U<sub>3</sub>O<sub>8</sub>.
===Uranium conversion===
Milled uranium oxide, U<sub>3</sub>O<sub>8</sub>, must be converted to [[uranium hexafluoride]], UF<sub>6</sub>, 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]] (U<sub>3</sub>O<sub>8</sub>) is also converted directly to [[ceramic]] grade [[uranium dioxide]] (UO<sub>2</sub>) for use in reactors not requiring enriched fuel, such as [[CANDU]]. The volumes of material converted directly to UO<sub>2</sub> are typically quite small compared to the amounts converted to UF<sub>6</sub>.
===Enrichment===<!-- This section is linked from [[Nuclear proliferation]] -->
{{main|enriched uranium}}
[[Image:Nuclear Fuel Cycle.png|330px|thumb|'''Nuclear fuel cycle''' begins when [[uranium]] is mined, enriched and manufactured to nuclear fuel (1) which is delivered to a [[nuclear power plant]]. After usage in the power plant the spent fuel is delivered to a reprocessing plant (if fuel is recycled) (2) or to a final repository (if no recycling is done) (3) for geological disposition. In [[nuclear reprocessing|reprocessing]] 95% of spent fuel can be recycled to be returned to usage in a [[nuclear power plant]] (4).]]
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 UF<sub>6</sub> 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 penetrator]]s, [[radiation shielding]] and [[Sailing ballast|ballast]]. Still, there are vast quantities of depleted uranium in storage. The [[United States Department of Energy]] alone has 470,000 [[tonne]]s.<ref>{{cite web |url= http://web.ead.anl.gov/uranium/faq/storage/faq16.cfm |title= How much depleted uranium hexafluoride is stored in the United States? |work= Depleted UF6 Management Information Network |accessdate= 2008-01-15}}</ref> About 95% of depleted uranium is stored as [[uranium hexafluoride]] (UF<sub>6</sub>).
===Fabrication===
{{main|Nuclear fuel}}
For use as nuclear fuel, enriched uranium hexafluoride is converted into [[uranium dioxide]] (UO<sub>2</sub>) 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 reactor]]s (BWR) and [[pressurized water reactor]]s (PWR), the tubes are assembled into bundles<ref>{{cite web |url= http://www.pplweb.com/NR/rdonlyres/F63D7386-A57E-46C6-90A5-857D513B0254/0/seic_plantguide.pdf |format= PDF |title= Susquehanna Nuclear Energy Guide |publisher= PPL Corporation |accessdate= 2008-01-15}}</ref> 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]] (UF<sub>6</sub>) 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, [[radiation shielding|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 cask]]s are used which are designed to maintain integrity under normal transportation conditions and during hypothetical accident conditions.
=== 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 [[neutron moderator|moderator]] and [[coolant]], which is water in most reactors.
Because of the [[nuclear fission|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 [[permutation]]s and the complexity of each computation. Many [[numerical method]]s 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 [[computation]]al and [[empirical]] techniques to manage this problem.
==== The study of used fuel ====
{{main|Post irradiation examination}}
Used nuclear fuel is studied in [[Post irradiation examination]], where used fuel is examined to know more about the processes that occur in fuel during use, and how these might alter the outcome of an accident. For example, during normal use, the fuel expands due to thermal expansion, which can cause cracking. Most [[nuclear fuel]] is uranium dioxide, this is a [[cubic]] solid which has a structure similar to that of [[calcium fluoride]], in used fuel the solid state structure of most of the solid remains the same as that of pure cubic uranium dioxide. SIMFUEL is the name given to the simulated spent fuel which is made by mixing finely ground metal oxides, grinding as a slurry, spray drying it before heating in hydrogen/argon to 1700 <sup>o</sup>C.<ref>A good report on the microstructure of used fuel is Lucuta PG ''et al'' (1991) ''J Nuclear Materials'' '''178''':48-60</ref> In SIMFUEL, 4.1% of the volume of the solid was in the form of metal [[nanoparticle]]s which are made of [[molybdenum]], [[ruthenium]], [[rhodium]] and [[palladium]]. Most of these metal particles are of the ε phase ([[hexagonal]]) of Mo-Ru-Rh-Pd alloy, while smaller amounts of the α ([[cubic]]) and σ ([[tetragonal]]) phases of these metals were found in the SIMFUEL. Also present within the SIMFUEL was a cubic [[perovskite]] phase which is a [[barium]] [[strontium]] [[zirconium|zirconate]] (Ba<sub>x</sub>Sr<sub>1-x</sub>ZrO<sub>3</sub>).
[[Image:UO2lattice.jpg|right|thumb|250px|The solid state structure of uranium dioxide, the oxygen atoms are in red and the uranium atoms in green]]
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 (UO<sub>2</sub>) can be oxidised to an oxygen rich hyperstoichiometric oxide (UO<sub>2+x</sub>) which can be further oxidised to U<sub>4</sub>O<sub>9</sub>, U<sub>3</sub>O<sub>7</sub>, U<sub>3</sub>O<sub>8</sub> and UO<sub>3</sub>.2H<sub>2</sub>O.
Because used fuel contains alpha emitters (plutonium and the [[minor actinides]]), the effect of adding an alpha emitter (<sup>238</sup>Pu) to uranium dioxide on the leaching rate of the oxide has been investigated. For the crushed oxide, adding <sup>238</sup>Pu tended to increase the rate of leaching, but the difference in the leaching rate between 0.1 and 10% <sup>238</sup>Pu was very small.<ref>V.V. Rondinella VV ''et al'' (2000) ''Radiochimica Acta'' '''88''':527-531</ref>
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 [UO<sub>2</sub>(CO<sub>3</sub>)<sub>2</sub>]<sup>2-</sup> and [UO<sub>2</sub>(CO<sub>3</sub>)<sub>3</sub>]<sup>4-</sup>. 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.<ref>For a review of the corrosion of uranium dioxide in a waste store which explains much of the chemistry, see Shoesmith DW (2000) ''J Nuclear Materials'' '''282''':1-31</ref>
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.<ref>Miserque F ''et al'' (2001) ''J Nuclear Materials'' '''298''':280-90</ref>
==== Fuel cladding interactions ====
The study of the nuclear fuel cycle includes the study of the behaviour of nuclear materials both under normal conditions and under accident conditions. For example, there has been much work on how [[uranium dioxide]] based fuel interacts with the [[zirconium]] alloy tubing used to cover it. During use, the fuel swells due to [[thermal expansion]] and then starts to react with the surface of the zirconium alloy, forming a new layer which contains both fuel and zirconium (from the cladding). Then, on the fuel side of this mixed layer, there is a layer of fuel which has a higher [[caesium]] to [[uranium]] ratio than most of the fuel. This is because [[xenon]] isotopes are formed as [[fission products]] that diffuse out of the lattice of the fuel into voids such as the narrow gap between the fuel and the cladding. After diffusing into these voids, it decays to caesium isotopes. Because of the thermal gradient which exists in the fuel during use, the volatile fission products tend to be driven from the centre of the pellet to the rim area.<ref>Further reading on fuel cladding interactions: Tanaka K ''et al'' (2006) ''J Nuclear Materials'' '''357''':58-68</ref> Below is a graph of the temperature of uranium metal, uranium nitride and [[uranium dioxide]] as a function of distance from the centre of a 20 mm diameter pellet with a rim temperature of 200 <sup>o</sup>C. It is important to note that the uranium dioxide (because of its poor thermal conductivity) will overheat at the centre of the pellet, while the more thermally conductive other forms of uranium remain below their melting points.
[[Image:Rim200pd1000rad1000fueltemp.jpg|center|thumb|450px|Temperature profile for a 20 mm diameter fuel pellet with a power density of 1000 W per cubic meter. The fuels other than uranium dioxide are not compromised.]]
==== Normal and abnormal conditions ====
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 <sup>137</sup>Cs which can be transferred to humans through the food chain. But <sup>137</sup>Cs is not able to migrate quickly through most soils and thus is unlikely to contaminate [[Water well|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 K<sub>d</sub> is the ratio of the soil's radioactivity (Bq g<sup>-1</sup>) to that of the soil water (Bq ml<sup>-1</sup>). 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.
* [[Cs-137]] K<sub>d</sub> = 1000
* [[plutonium|Pu-239]] K<sub>d</sub> = 10000 to 100000
* [[strontium|Sr-90]] K<sub>d</sub> = 80 to 150
* [[Iodine|I-131]] K<sub>d</sub> = 0.007 to 50
One of the best countermeasures in dairy farming against <sup>137</sup>Cs is to mix up the soil by deeply ploughing the soil. This has the effect of putting the <sup>137</sup>Cs 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 <sup>137</sup>Cs 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 <sup>90</sup>Sr and <sup>137</sup>Cs into [[sunflower]]s grown under [[hydroponic]] conditions has been reported.<ref>P. Soudek, Š. Valenová, Z. Vavříková and T. Vaněk, ''Journal of Environmental Radioactivity'', 2006, '''88''', 236-250</ref> 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 <sup>137</sup>Cs 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 [[half-life|nuclear half life]]) of the caesium. The physical or nuclear half life of <sup>137</sup>Cs 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 [[feces|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 [[paint]]s have not been successful. Note that a good source of data on the subject of [[caesium]] in [[Chernobyl]] fallout exists at [http://www.uiar.org.ua/Eng/index.htm], this is the ''Ukrainian Research Institute for Agricultural Radiology''.
===== Release of radioactivity from fuel during normal use and accidents=====
The IAEA assume that under normal operation the coolant of a water cooled reactor will contain some radioactivity<ref>page 169 Generic Assessment Procedures for Determining Protective Actions During a Reactor Accident, IAEA-TECDOC-955, 1997</ref> 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 <sup>o</sup>C) 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 <sup>o</sup>C).<ref>page 173 Generic Assessment Procedures for Determining Protective Actions During a Reactor Accident, IAEA-TECDOC-955, 1997</ref>
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,<ref>page 171 Generic Assessment Procedures for Determining Protective Actions During a Reactor Accident, IAEA-TECDOC-955, 1997</ref> 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.
===== Releases from reprocessing under normal conditions =====
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]][http://www.osti.gov/energycitations/product.biblio.jsp?osti_id=7296321][http://www10.antenna.nl/wise/index.html?http://www10.antenna.nl/wise/381/3733.html][http://archive.tri-cityherald.com/thyroid/history.html]) 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 gas]]es 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.[http://www.ornl.gov/~webworks/cppr/y2001/pres/123514.pdf]
A paper was written on the radioactivity found in [[oyster]]s found in the [[Irish Sea]],<ref>A. Preston, J.W.R. Dutton and B.R. Harvey, ''Nature'', 1968, '''218''', 689-690.</ref> these were found by gamma spectrscopy to contain <sup>141</sup>Ce, <sup>144</sup>Ce, <sup>103</sup>Ru, <sup>106</sup>Ru, <sup>137</sup>Cs, <sup>95</sup>Zr and <sup>95</sup>Nb. In addition a zinc activation product (<sup>65</sup>Zn) was found, this is thought to be due to the corrosion of [[magnox]] fuel cladding in [[cooling pond]]s. It is likely that the modern releases of all these isotopes from Windscale is smaller.
=== On-load reactors ===
Some reactor designs, such as [[RBMK]]s or [[CANDU reactor]]s, can be refueled 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 [[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 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 fueling 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 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]].
===Transportation===
{{main|Spent nuclear fuel shipping cask}}
===Reprocessing===
{{main|Nuclear reprocessing}}
{{seealso|Used nuclear fuel}}
[[Image:Sellafield-1515b.jpg|330px|thumb|The [[Sellafield]] reprocessing plant]]
Spent fuel discharged from reactors contains appreciable quantities of fissile (U-235 and Pu-239), fertile (U-238), and other [[radioactive]] materials, including [[nuclear poison|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 weapon]]s.
===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 product]]s can be converted to short-lived [[isotope]]s by either [[neutron]] or [[photon]] irradiation. This is called [[Nuclear transmutation|transmutation]].
===Waste disposal===
{{main|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. In the U.S., under the [[Nuclear Waste Policy Act]] of 1982 as amended, the [[United States Department of Energy|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.
== See also ==
*[[Synthesis of noble metals]]
*[[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]]
==References==
{{reflist|2}}
(''Reference'' V. Artisyuk, M. Saito and A. Shmelev, ''Progress in Nuclear Energy'', 2000, '''37''', 345-350)
==External links==
*[http://news.bbc.co.uk/2/shared/spl/hi/sci_nat/05/nuclear_fuel/html/mining.stm BBC - The Nuclear Fuel Cycle]
*[http://www.nephy.chalmers.se/staff-pages/demaz/new/teaching/supervision/sara_caprioli_presentation.pdf Chalmers University of Technology - In-Core Fuel Management] (PDF)
*[http://www.hyke.org/preprint/2003/08/085.ps HYKE - Assembly Distribution Optimality Condition] (PS)
*[http://www.nuclearfiles.org/menu/key-issues/nuclear-weapons/issues/proliferation/fuel-cycle/ Nuclear Files - The Nuclear Fuel Cycle]
*[http://www.wise-uranium.org/nfce.html WISE - Nuclear Fuel Energy Balance Calculator]
*[http://www.world-nuclear.org/info/inf69.htm World Nuclear Association - Reprocessing Facilities]
*[http://alsos.wlu.edu/qsearch.aspx?browse=science/Nuclear+Fuel+Cycle Annotated bibliography on the nuclear fuel cycle from the Alsos Digital Library for Nuclear Issues]
*[http://thoriumenergy.blogspot.com Thorium Energy] Blog, discussion forum and document repository
[[Category:Nuclear chemistry]]
[[Category:Nuclear reprocessing]]
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