Разделение изотопов и применение их в ядерном реакторе

Butex, β,β'-dibutyoxydiethyl ether

A process based on a solvation extraction process using the triether extractant named above. This process has the disadvantage of requiring the use of a salting-out reagent (aluminium nitrate) to increase the nitrate concentration in the aqueous phase to obtain a reasonable distribution ratio. This process was used at Windscale many years ago. This process has been replaced by PUREX.

Pyroprocessing

Pyroprocessing is a generic term for high-temperature methods. Solvents are molten salts (e.g. LiCl+KCl or LiF+CaF2) and molten metals (e.g. cadmium, bismuth, magnesium) rather than water and organic compounds. Electrorefining, distillation, and solvent-solvent extraction are common steps.

These processes are not currently in significant use worldwide, but they have been researched and developed at Argonne National Laboratory and elsewhere.

Advantages

The principles behind them are well understood, and no significant technical barriers exist to their adoption.[17]

Readily applied to high-burnup spent fuel and requires little cooling time, since the operating temperatures are high already.

Does not use solvents containing hydrogen and carbon, which are neutron moderators creating risk of criticality accidents and can absorb the fission product tritium and the activation product carbon-14 in dilute solutions that cannot be separated later.

Alternatively, voloxidation[18] can remove 99% of the tritium from used fuel and recover it in the form of a strong solution suitable for use as a supply of tritium.

More compact than aqueous methods, allowing on-site reprocessing at the reactor site, which avoids transportation of spent fuel and its security issues, instead storing a much smaller volume of fission products on site as high-level waste until decommissioning. For example, the Integral Fast Reactor and Molten Salt Reactor fuel cycles are based on on-site pyroprocessing.

It can separate many or even all actinides at once and produce highly radioactive fuel which is harder to manipulate for theft or making nuclear weapons. (However, the difficulty has been questioned.[19]) In contrast the PUREX process was designed to separate plutonium only for weapons, and it also leaves the minor actinides (americium and curium) behind, producing waste with more long-lived radioactivity.

Most of the radioactivity in roughly 102 to 105 years after the use of the nuclear fuel is produced by the actinides, since there are no fission products with half-lives in this range. These actinides can fuel fast reactors, so extracting and reusing (fissioning) them reduces the long-term radioactivity of the wastes.

Disadvantages

Reprocessing as a whole is not currently (2005) in favor, and places that do reprocess already have PUREX plants constructed. Consequently, there is little demand for new pyrometalurgical systems, although there could be if the Generation IV reactor programs become reality.

The used salt from pyroprocessing is less suitable for conversion into glass than the waste materials produced by the PUREX process.

Electrolysis

PYRO-A and -B for IFR

These processes were developed by Argonne National Laboratory and used in the Integral Fast Reactor project.

PYRO-A is a means of separating actinides (elements within the actinide family, generally heavier than U-235) from non-actinides. The spent fuel is placed in an anode basket which is immersed in a molten salt electrolyte. An electrical current is applied, causing the uranium metal (or sometimes oxide, depending on the spent fuel) to plate out on a solid metal cathode while the other actinides (and the rare earths) can be absorbed into a liquid cadmium cathode. Many of the fission products (such as caesium, zirconium and strontium) remain in the salt.[20][21][22] As alternatives to the molten cadmium electrode it is possible to use a molten bismuth cathode, or a solid aluminium cathode.

As an alternative to electrowinning, the wanted metal can be isolated by using a molten alloy of an electropositive metal and a less reactive metal.

Since the majority of the long term radioactivity, and volume, of spent fuel comes from actinides, removing the actinides produces waste that is more compact, and not nearly as dangerous over the long term. The radioactivity of this waste will then drop to the level of various naturally occurring minerals and ores within a few hundred, rather than thousands of, years.

The mixed actinides produced by pyrometallic processing can be used again as nuclear fuel, as they are virtually all either fissile, or fertile, though many of these materials would require a fast breeder reactor in order to be burned efficiently. In a thermal neutron spectrum, the concentrations of several heavy actinides (curium-242 and plutonium-240) can become quite high, creating fuel that is substantially different from the usual uranium or mixed uranium-plutonium oxides (MOX) that most current reactors were designed to use.

Another pyrochemical process, the PYRO-B process, has been developed for the processing and recycling of fuel from a transmuter reactor ( a fast breeder reactor designed to convert transuranic nuclear waste into fission products ). A typical transmuter fuel is free from uranium and contains recovered transuranics in an inert matrix such as metallic zirconium. In the PYRO-B processing of such fuel, an electrorefining step is used to separate the residual transuranic elements from the fission products and recycle the transuranics to the reactor for fissioning. Newly-generated technetium and iodine are extracted for incorporation into transmutation targets, and the other fission products are sent to waste.

Voloxidation

 

Voloxidation (for volumetric oxidation) involves heating oxide fuel with oxygen, sometimes with alternating oxidation and reduction, or alternating oxidation by ozone to uranium trioxide with decomposition by heating back to triuranium octoxide.[18] A major purpose is to capture tritium as tritiated water vapor before further processing where it would be difficult to retain the tritium. Other volatile elements leave the fuel and must be recovered, especially iodine, technetium, and carbon-14. Voloxidation also breaks up the fuel or increases its surface area to enhance penetration of reagents in following reprocessing steps.

Volatilization in isolation

Simply heating spent oxide fuel in an inert atmosphere or vacuum at a temperature between 700°C and 1000°C as a first reprocessing step can remove several volatile elements, including caesium whose isotope cesium-137 emits about half of the heat produced by the spent fuel over the following 100 years of cooling (however, most of the other half is from strontium-90 which remains). The estimated overall mass balance for 20,000 grams of processed fuel with 2,000 grams of cladding is: Input Residue Zeolite

Tritium is not mentioned in this paper.

Fluoride volatility

Main article: Fluoride volatility

Blue elements have volatile fluorides or are already volatile; green elements do not but have volatile chlorides; red elements have neither, but the elements themselves or their oxides are volatile at very high temperatures. Yields at 100,1,2,3 years after fission, not considering later neutron capture, fraction of 100% not 200%. Beta decay Kr-85→Rb, Sr-90→Zr, Ru-106→Pd, Sb-125→Te, Cs-137→Ba, Ce-144→Nd, Sm-151→Eu, Eu-155→Gd visible.

In the fluoride volatility process, fluorine is reacted with the fuel. Fluorine is so much more reactive than even oxygen that small particles of ground oxide fuel will burst into flame when dropped into a chamber full of fluorine. This is known as flame fluorination; the heat produced helps the reaction proceed. Most of the uranium, which makes up the bulk of the fuel, is converted to uranium hexafluoride, the form of uranium used in uranium enrichment, which has a very low boiling point. Technetium, the main long-lived fission product, is also efficiently converted to its volatile hexafluoride. A few other elements also form similarly volatile hexafluorides, pentafluorides, or heptafluorides. The volatile fluorides can be separated from excess fluorine by condensation, then separated from each other by fractional distillation or selective reduction. Uranium hexafluoride and technetium hexafluoride have very similar boiling points and vapor pressures, which makes complete separation more difficult.

Many of the fission products volatilized are the same ones volatilized in non-fluorinated, higher-temperature volatilization, such as iodine, tellurium and molybdenum; notable differences are that technetium is volatilized, but caesium is not.

Some transuranium elements such as plutonium, neptunium and americium can form volatile fluorides, but these compounds are not stable when the fluorine partial pressure is decreased.[27] Most of the plutonium and some of the uranium will initially remain in ash which drops to the bottom of the flame fluorinator. The plutonium-uranium ratio in the ash may even approximate the composition needed for fast neutron reactor fuel. Further fluorination of the ash can remove all the uranium, neptunium, and plutonium as volatile fluorides; however, some other minor actinides may not form volatile fluorides and instead remain with the alkaline fission products. Some noble metals may not form fluorides at all, but remain in metallic form; however ruthenium hexafluoride is relatively stable and volatile.

Distillation of the residue at higher temperatures can separate lower-boiling transition metal fluorides and alkali metal (Cs, Rb) fluorides from higher-boiling lanthanide and alkaline earth metal (Sr, Ba) and yttrium fluorides. The temperatures involved are much higher, but can be lowered somewhat by distilling in a vacuum. If a carrier salt like lithium fluoride or sodium fluoride is being used as a solvent, high-temperature distillation is a way to separate the carrier salt for reuse.

Molten salt reactor designs carry out fluoride volatility reprocessing continuously or at frequent intervals. The goal is to return actinides to the molten fuel mixture for eventual fission, while removing fission products that are neutron poisons, or that can be more securely stored outside the reactor core while awaiting eventual transfer to permanent storage.

Chloride volatility and solubility

Many of the elements that form volatile high-valence fluorides will also form volatile high-valence chlorides. Chlorination and distillation is another possible method for separation. The sequence of separation may differ usefully from the sequence for fluorides; for example, zirconium tetrachloride and tin tetrachloride have relatively low boiling points of 331°C and 114.1°C. Chlorination has even been proposed as a method for removing zirconium fuel cladding,[18] instead of mechanical decladding.

Chlorides are likely to be easier than fluorides to later convert back to other compounds, such as oxides.

Chlorides remaining after volatilization may also be separated by solubility in water. Chlorides of alkaline elements like americium, curium, lanthanides, strontium, caesium are more soluble than those of uranium, neptunium, plutonium, and zirconium.

Economics

The relative economics of reprocessing-waste disposal and interim storage-direct disposal has been the focus of much debate over the past ten years. Studies[28] Template:Others? have modeled the total fuel cycle costs of a reprocessing-recycling system based on one-time recycling of plutonium in existing thermal reactors (as opposed to the proposed breeder reactor cycle) and compare this to the total costs of an open fuel cycle with direct disposal. The range of results produced by these studies is very wide, but all are agreed that under current (2005) economic conditions the reprocessing-recycle option is the more costly.

If reprocessing is undertaken only to reduce the radioactivity level of spent fuel it should be taken into account that spent nuclear fuel becomes less radioactive over time. After 40 years its radioactivity drops by 99.9%,[29] though it still takes over a thousand years for the level of radioactivity to approach that of natural uranium.[30] However the level of transuranic elements, including plutonium-239, remains high for over 100,000 years, so if not reused as nuclear fuel, then those elements need secure disposal because of nuclear proliferation reasons as well as radiation hazard.

 

Enriched uranium is a kind of uranium in which the percent composition of uranium-235 has been increased through the process of isotope separation. Natural uranium is 99.284% 238U isotope, with 235U only constituting about 0.711% of its weight. 235U is the only isotope existing in nature (in any appreciable amount) that is fissile with thermal neutrons.

Enriched uranium is a critical component for both civil nuclear power generation and military nuclear weapons. The International Atomic Energy Agency attempts to monitor and control enriched uranium supplies and processes in its efforts to ensure nuclear power generation safety and curb nuclear weapons proliferation.

During the Manhattan Project enriched uranium was given the codename oralloy, a shortened version of Oak Ridge alloy, after the location of the plants where the uranium was enriched. The term oralloy is still occasionally used to refer to enriched uranium. There are about 2,000 tonnes (t, Mg) of highly enriched uranium in the world,[1] produced mostly for nuclear weapons, naval propulsion, and smaller quantities for research reactors.

The 238U remaining after enrichment is known as depleted uranium (DU), and is considerably less radioactive than even natural uranium, though still very dense and extremely hazardous in granulated form—such granules are a natural by-product of the shearing action that makes it useful for armour-penetrating weapons, radiation shielding, while other uses such as for highly dense concrete are currently being explored.[citation needed] At present, 95% of the world's stocks of depleted uranium remain in secure storage.

Contents

1 Grades

1.1 Slightly enriched uranium (SEU)

1.2 Reprocessed uranium

1.3 Low-enriched uranium (LEU)

1.4 Highly enriched uranium (HEU)

2 Enrichment methods

2.1 Diffusion techniques

2.1.1 Gaseous diffusion

2.1.2 Thermal diffusion

2.2 Centrifuge techniques

2.2.1 Gas centrifuge

2.2.2 Zippe centrifuge

2.3 Laser techniques

2.3.1 Atomic vapor laser isotope separation (AVLIS)

2.3.2 Molecular laser isotope separation (MLIS)

2.3.3 Separation of Isotopes by Laser Excitation (SILEX)

2.4 Other techniques

2.4.1 Aerodynamic processes

2.4.2 Electromagnetic isotope separation

2.4.3 Chemical methods

2.4.4 Plasma separation

3 Separative work unit

4 Downblending

5 Global enrichment facilities

Grades

Slightly enriched uranium (SEU)

A drum of yellowcake (a mixture of uranium precipitates)

Slightly enriched uranium (SEU) has a 235U concentration of 0.9% to 2%. This new grade is being used to replace natural uranium (NU) fuel in some heavy water reactors like the CANDU. Costs are lowered because less uranium and fewer bundles are needed to fuel the reactor. This in turn reduces the quantity of used fuel and its subsequent waste management costs.

Reprocessed uranium

Main article: Reprocessed uranium

Reprocessed uranium (RpU or RU) is a product of nuclear fuel cycles involving nuclear reprocessing of spent fuel. RpU recovered from light water reactor (LWR) spent fuel typically contains slightly more U-235 than natural uranium, and therefore could be used to fuel reactors that customarily use natural uranium as fuel. However, it also contains the undesirable isotope uranium-236 which undergoes neutron capture, wasting neutrons (and requiring higher U-235 enrichment) and creating neptunium-237 which would be one of the more mobile and troublesome radionuclides in deep geological repository disposal of nuclear waste.

Low-enriched uranium (LEU)

Low-enriched uranium (LEU) has a lower than 20% concentration of 235U. For use in commercial light water reactors (LWR), the most prevalent power reactors in the world, uranium is enriched to 3 to 5% 235U. Fresh LEU used in research reactors is usually enriched 12% to 19.75% U-235, the latter concentration being used to replace HEU fuels when converting to LEU.

Highly enriched uranium (HEU)

A billet of highly enriched uranium metal

Highly enriched uranium (HEU) has a greater than 20% concentration of 235U or 233U. The fissile uranium in nuclear weapons usually contains 85% or more of 235U known as weapon(s)-grade, though for a crude, inefficient weapon 20% is sufficient (called weapon(s)-usable)[2][3]; some argue that even less is sufficient[citation needed], but then the critical mass for unmoderated fast neutrons rapidly increases, approaching infinity at 6%235U.[4] For criticality experiments, enrichment of uranium to over 97% has been accomplished.[5]

The very first uranium bomb, Little Boy in 1945, used 64 kilograms of 80% enriched uranium. Wrapping the weapon's fissile core in a neutron reflector (which is standard on all nuclear explosives) can dramatically reduce the critical mass. Because the core was surrounded by a good neutron reflector, at explosion it comprised almost 2.5 critical masses. Neutron reflectors and compressing the fissile core via implosion allow nuclear weapon designs that use less than what would be one bare-sphere critical mass at normal density. The presence of too much of the 238U isotope inhibits the runaway nuclear chain reaction that is responsible for the weapon's power. The critical mass for 85% highly enriched uranium is about 50 kilograms (110 lb), which at normal density would be a sphere about 17 centimetres (6.7 in) in diameter.

Later US nuclear weapons usually use plutonium-239 in the primary stage, but the secondary stage which is compressed by the primary nuclear explosion often uses HEU with enrichment between 40% and 80%[6] along with the fusion fuel lithium deuteride. For the secondary of a large nuclear weapon, the higher critical mass of less-enriched uranium can be an advantage as it allows the core at explosion time to contain a larger amount of fuel. The 238U is not fissile but still fissionable by fusion neutrons.

HEU is also used in fast neutron reactors, whose cores require about 20% or more of fissile material, as well as in naval reactors, where it often contains at least 50% 235U, but typically does not exceed 90%. The Fermi-1 commercial fast reactor prototype used HEU with 26.5% 235U. Significant quantities of HEU are used in the production of medical isotopes, for example molybdenum-99 for technetium-99m generators.[7]

Enrichment methods

Isotope separation is difficult because two isotopes of the same elements have very nearly identical chemical properties, and can only be separated gradually using small mass differences. (235U is only 1.26% lighter than 238U.) This problem is compounded by the fact that uranium is rarely separated in its atomic form, but instead as a compound (235UF6 is only 0.852% lighter than 238UF6.) A cascade of identical stages produces successively higher concentrations of 235U. Each stage passes a slightly more concentrated product to the next stage and returns a slightly less concentrated residue to the previous stage.

There are currently two generic commercial methods employed internationally for enrichment: gaseous diffusion (referred to as first generation) and gas centrifuge (second generation) which consumes only 6% as much energy as gaseous diffusion. Later generation methods will become established because they will be more efficient in terms of the energy input for the same degree of enrichment and the next method of enrichment to be commercialized will be referred to as third generation. Some work is being done that would use nuclear resonance; however there is no reliable evidence that any nuclear resonance processes have been scaled up to production.

Diffusion techniques

Gaseous diffusion

Main article: Gaseous diffusion

Gaseous diffusion is a technology used to produce enriched uranium by forcing gaseous uranium hexafluoride (hex) through semi-permeable membranes. This produces a slight separation between the molecules containing 235U and 238U. Throughout the Cold War, gaseous diffusion played a major role as a uranium enrichment technique, and continues to account for about 33% of enriched production[8] but is now an obsolete technology that is steadily being replaced by the later generations of technology as the diffusion plants reach their ends-of-life.

Thermal diffusion

Thermal diffusion utilizes the transfer of heat across a thin liquid or gas to accomplish isotope separation. The process exploits the fact that the lighter 235U gas molecules will diffuse toward a hot surface, and the heavier 238U gas molecules will diffuse toward a cold surface. The S-50 plant at Oak Ridge, Tennessee was used during World War II to prepare feed material for the EMIS process. It was abandoned in favor of gaseous diffusion.

Centrifuge techniques

Gas centrifuge

Main article: Gas centrifuge

A cascade of gas centrifuges at a U.S. enrichment plant

The gas centrifuge process uses a large number of rotating cylinders in series and parallel formations. Each cylinder's rotation creates a strong centrifugal force so that the heavier gas molecules containing 238U move toward the outside of the cylinder and the lighter gas molecules rich in 235U collect closer to the center. It requires much less energy to achieve the same separation than the older gaseous diffusion process, which it has largely replaced and so is the current method of choice and is termed second generation. It has a separation factor per stage of 1.3 relative to gaseous diffusion of 1.005,[8] which translates to about one-fiftieth of the energy requirements. Gas centrifuge techniques produce about 54% of the world's enriched uranium.

Zippe centrifuge

Diagram of the principles of a Zippe-type gas centrifuge with U-238 represented in dark blue and U-235 represented in light blue

The Zippe centrifuge is an improvement on the standard gas centrifuge, the primary difference being the use of heat. The bottom of the rotating cylinder is heated, producing convection currents that move the 235U up the cylinder, where it can be collected by scoops. This improved centrifuge design is used commercially by Urenco to produce nuclear fuel and was used by Pakistan in their nuclear weapons program.

Laser techniques

Laser processes promise lower energy inputs, lower capital costs and lower tails assays, hence significant economic advantages. Several laser processes have been investigated or are under development. None of the laser processes below are yet ready for commercial use, though SILEX is well advanced and expected to begin commercial production in 2012.(see here: 30 April 2008) and May 2010 Investor Presentation

Atomic vapor laser isotope separation (AVLIS)

Atomic vapor laser isotope separation employs specially tuned lasers[9] to separate isotopes of uranium using selective ionization of hyperfine transitions. The technique uses lasers which are tuned to frequencies that ionize 235U atoms and no others. The positively charged 235U ions are then attracted to a negatively charged plate and collected

Molecular laser isotope separation (MLIS)

Molecular laser isotope separation uses an infrared laser directed at UF6, exciting molecules that contain a 235U atom. A second laser frees a fluorine atom, leaving uranium pentafluoride which then precipitates out of the gas.

Separation of Isotopes by Laser Excitation (SILEX)

Separation of isotopes by laser excitation is an Australian development that also uses UF6. After a protracted development process involving U.S. enrichment company USEC acquiring and then relinquishing commercialization rights to the technology, GE Hitachi Nuclear Energy (GEH) signed a commercialization agreement with Silex Systems in 2006 (see here). GEH has since begun construction of a demonstration test loop and announced plans to build an initial commercial facility. (see here: 30 April 2008). Details of the process are restricted by intergovernmental agreements between USA and Australia and the commercial entities. SILEX has been indicated to be an order of magnitude more efficient than existing production techniques but again, the exact figure is classified.

Other techniques

Aerodynamic processes

Schematic diagram of an aerodynamic nozzle. Many thousands of these small foils would be combined in an enrichment unit.

Aerodynamic enrichment processes include the Becker jet nozzle techniques developed by E. W. Becker and associates using the LIGA process and the vortex tube separation process. These aerodynamic separation processes depend upon diffusion driven by pressure gradients, as does the gas centrifuge. In effect, aerodynamic processes can be considered as non-rotating centrifuges. Enhancement of the centrifugal forces is achieved by dilution of UF6 with hydrogen or helium as a carrier gas achieving a much higher flow velocity for the gas than could be obtained using pure uranium hexafluoride. The Uranium Enrichment Corporation of South Africa (UCOR) developed and deployed the Helikon vortex separation process based on the vortex tube and a demonstration plant was built in Brazil by NUCLEI, a consortium led by Industrias Nucleares do Brasil that used the separation nozzle process. However both methods have high energy consumption and substantial requirements for removal of waste heat; neither is currently in use.