Разделение изотопов и применение их в ядерном реакторе

'QUADRISO Particle

RBMK fuel

RBMK reactor fuel was used in Soviet designed and built RBMK type reactors. This is a low enriched uranium oxide fuel. The fuel elements in an RBMK are 3 m long each, and two of these sit back-to-back on each fuel channel, pressure tube. Reprocessed uranium from Russian VVER reactor spent fuel is used to fabricate RBMK fuel. Following the Chernobyl accident, the enrichment of fuel was changed from 2.0% to 2.4%, to compensate for control rod modifications and the introduction of additional absorbers.

CerMet fuel

CerMet fuel consists of ceramic fuel particles (usually uranium oxide) embedded in a metal matrix. It is hypothesized that this type of fuel is what is used in United States Navy reactors. This fuel has high heat transport characteristics and can withstand a large amount of expansion.

Plate type fuel

ATR Core The Advanced Test Reactor at Idaho National Laboratory uses plate type fuel in a clover leaf arrangement. The blue glow around the core is known as Cherenkov radiation.

Plate type fuel has fallen out of favor over the years. Plate type fuel is commonly composed of enriched uranium sandwiched between metal cladding. Plate type fuel is used in several research reactors where a high neutron flux is desired, for uses such as material irradiation studies or isotope production, without the high temperatures seen in ceramic, cylindrical fuel. It is currently used in the Advanced Test Reactor (ATR) at Idaho National Laboratory.

Soium bonded fuel

Sodium bonded fuel consists of fuel that has liquid sodium in the gap between the fuel slug (or pellet) and the cladding. This fuel type is often used for sodium cooled liquid metal fast reactors. It has been used in EBR-I, EBR-II, and the FFTF. The fuel slug may be metallic or ceramic. The sodium bonding is used to reduce the temperature of the fuel.

Spent nuclear fuel

Main article: Spent nuclear fuel

Used nuclear fuel is a complex mixture of the fission products, uranium, plutonium and the transplutonium metals. In fuel which has been used at high temperature in power reactors it is common for the fuel to be heterogeneous; often the fuel will contain nanoparticles of platinum group metals such as palladium. Also the fuel may well have cracked, swelled and been used close to its melting point. Despite the fact that the used fuel can be cracked, it is very insoluble in water, and is able to retain the vast majority of the actinides and fission products within the uranium dioxide crystal lattice.

Oxide fuel under accident conditions

Main article: Nuclear fuel response to reactor accidents

Two main modes of release exist, the fission products can be vaporised or small particles of the fuel can be dispersed.

Fuel behavior and post irradiation examination (PIE)

Materials in a high radiation environment (such as a reactor) can undergo unique behaviors such as swelling [4] and non-thermal creep. If there are nuclear reactions within the material (such as what happens in the fuel), the stoichiometry will also change slowly over time. These behaviors can lead to new material properties, cracking, and fission gas release:

Fission gas release

As the fuel is degraded or heated the more volatile fission products which are trapped within the uranium dioxide may become free. For example see J.Y. Colle, J.P. Hiernaut, D. Papaioannou, C. Ronchi, A. Sasahara, Journal of Nuclear Materials, 2006, 348, 229.

Fuel cracking

As the fuel expands on heating, the core of the pellet expands more than the rim which may lead to cracking. Because of the thermal stress thus formed the fuel cracks, the cracks tend to go from the center to the edge in a star-shaped pattern.

In order to better understand and control these changes in materials, these behaviors are studied. A common experiment to do this is post irradiation examination, in which fuel will be examined after it is put through reactor-like conditions . Due to the intensely radioactive nature of the used fuel this is done in a hot cell. A combination of nondestructive and destructive methods of PIE are common.

The PIE is used to check that the fuel is both safe and effective. After major accidents, the core is normally subject to PIE in order to find out what happened. One site where PIE is done is the ITU which is the EU center for the study of highly radioactive materials.

In addition to the effects of radiation and the fission products on materials, scientists also need to consider the temperature of materials in a reactor, and in particular, the fuel. Too high a fuel temperature can compromise the fuel, and therefore it is important to control the temperature in order to control the fission chain reaction.

The temperature of the fuel varies as a function of the distance from the center to the rim. At distance x from the center the temperature (Tx) is described by the equation where ρ is the power density (W m−3) and Kf is the thermal conductivity.

Tx = TRim + ρ  (rpellet2 - x2) (4 Kf)-1

To explain this for a series of fuel pellets being used with a rim temperature of 200 °C (typical for a BWR) with different diameters and power densities of 250 MW·m−3 have been modeled using the above equation. Note that these fuel pellets are rather large; it is normal to use oxide pellets which are about 10 mm in diameter.

Nuclear Fuel Temperature Profiles

Temperature profile for a 20 mm diameter fuel pellet with a power density of 250 MW per cubic meter. Note the central temperature is very different for the different fuel solids. 

Temperature profile for a 26 mm diameter fuel pellet with a power density of 250 MW per cubic meter. 

Temperature profile for a 32 mm diameter fuel pellet with a power density of 250 MW per cubic meter.

 

Temperature profile for a 20 mm diameter fuel pellet with a power density of 500 MW per cubic meter. Because the melting point of uranium dioxide is about 3300 K, it is clear that uranium oxide fuel is overheating at the center. 

Temperature profile for a 20 mm diameter fuel pellet with a power density of 1000 MW per cubic meter. The fuels other than uranium dioxide are not compromised.

Reference Radiochemistry and Nuclear Chemistry, G. Choppin, J-O Liljenzin and J. Rydberg, 3rd Ed, 2002, Butterworth-Heinemann, ISBN 0-7506-7463-6

Radioisotope decay fuels

Radioisotope battery

Main article: atomic battery

The terms atomic battery, nuclear battery and radioisotope battery are used interchangely to describe a device which uses the radioactive decay to generate electricity. These systems use radioisotopes that produce low energy beta particles or sometimes alpha particles of varying energies. Low energy beta particles are needed to prevent the production of high energy penetrating Bremsstrahlung radiation that would require heavy shielding. Radioisotopes such as tritium, nickel-63, promethium-147, and technetium-99 have been tested. Plutonium-238, curium-242, curium-244 and strontium-90 have been used.

There are two main categories of atomic batteries: thermal and non-thermal. The non-thermal atomic batteries, which have many different designs, exploit charged alpha and beta particles. These designs include the direct charging generators, betavoltaics, the optoelectric nuclear battery, and the radioisotope piezoelectric generator. The thermal atomic batteries on the other hand, convert the heat from the radioactive decay to electricity. These designs include thermionic converter, thermophotovoltaic cells, alkali-metal thermal to electric converter, and the most common design, the radioisotope thermoelectric generator.

Radioisotope thermoelectric generators

A radioisotope thermoelectric generator (RTG) is a simple electrical generator which converts heat into electricity from a radioisotope using an array of thermocouples.

238Pu has become the most widely used fuel for RTGs. In the form of plutonium dioxide it has a half-life of 87.7 years, reasonable energy density and

exceptionally low gamma and neutron radiation levels. Some Russian terrestrial RTGs have used 90Sr; this isotope has a shorter half-life and a much

lower energy density, but is cheaper. Early RTGs, first built in 1958 by the U.S. Atomic Energy Commission, have used 210Po. This fuel provides

phenomenally huge energy density, (a single gram of polonium-210 generates 140 watts thermal) but has limited use because of its very short half-life and gamma production and has been phased out of use in this application.

Radioisotope heater units (RHU)

Photo of a disassembled RHU

Radioisotope heater units normally provide about 1 watt of heat each, derived from the decay of a few grams of plutonium-238. This heat is given off continuously for several decades.

 

Their function is to provide highly localised heating of sensitive equipment (such as electronics in outer space). The Cassini-Huygens orbiter to Saturn contains 82 of these units (in addition to its 3 main RTG's for power generation). The Huygens probe to Titan contains 35 devices.

Fusion fuels

Fusion fuels include tritium (3H) and deuterium (2H) as well as helium-3 (3He). Many other elements can be fused together, but the larger electrical charge of their nuclei means that much higher temperatures are required. Only the fusion of the lightest elements is seriously considered as a future energy source. Although the energy density of fusion fuel is even higher than fission fuel, and fusion reactions sustained for a few minutes have been achieved, utilizing fusion fuel as a net energy source remains a theoretical possibility.[1]

First generation fusion fuel

Deuterium and tritium are both considered first-generation fusion fuels; they are the easiest to fuse, because the electrical charge on their nuclei is the lowest of all elements. The three most commonly cited nuclear reactions that could be used to generate energy are:

2H + 3H  n (14.07 MeV) + 4He (3.52 MeV)

2H + 2H  n (2.45 MeV) + 3He (0.82 MeV)

2H + 2H  p (3.02 MeV) + 3H (1.01 MeV)

Second generation fusion fuel

Second generation fuels require either higher confinement temperatures or longer confinement time than those required of first generation fusion fuels, but generate fewer neutrons. Neutrons are an unwanted byproduct of fusion reactions in an energy generation context, because they are absorbed by the walls of a fusion chamber, making them radioactive. They cannot be confined by magnetic fields, because they are not electrically charged. This group consists of deuterium and helium-3. The products are all charged particles, but there may be significant side reactions leading to the production of neutrons.

2H + 3He  p (14.68 MeV) + 4He (3.67 MeV)

Third generation fusion fuel

Main article: Aneutronic fusion

Third generation fusion fuels produce only charged particles in the primary reactions, and side reactions are relatively unimportant. Since a very small amount of neutrons is produced, there would be little induced radioactivity in the walls of the fusion chamber. This is often seen as the end goal of fusion research. 3He has the highest Maxwellian reactivity of any 3rd generation fusion fuel. However, there are no significant natural sources of this substance on Earth.

3He + 3He  2p + 4He (12.86 MeV)

Another potential aneutronic fusion reaction is the proton-boron reaction:

p + 11B → 34He

Under reasonable assumptions, side reactions will result in about 0.1% of the fusion power being carried by neutrons. With 123 keV, the optimum temperature for this reaction is nearly ten times higher than that for the pure hydrogen reactions, the energy confinement must be 500 times better than that required for the D-T reaction, and the power density will be 2500 times lower than for D-T.

 

 

 

Nuclear reprocessing uses chemical procedures to separate the useful components (especially the remaining uranium and the newly-created plutonium) from the fission products and other radioactive waste in spent nuclear fuel obtained from nuclear reactors. Reprocessing serves multiple purposes, whose relative importance has changed over time. Originally reprocessing was used solely to extract plutonium for producing nuclear weapons. With the commercialization of nuclear power, the reprocessed plutonium was recycled back into MOX nuclear fuel for thermal reactors. The reprocessed uranium, which constitutes the bulk of the spent fuel material, can in principle also be re-used as fuel, but that is only economic when uranium prices are high. Finally, the breeder reactor can employ not only the recycled plutonium and uranium in spent fuel, but all the actinides, closing the nuclear fuel cycle and potentially multiplying the energy extracted from natural uranium by more than 60 times.[2] Nuclear reprocessing also reduces the volume of high-level nuclear waste and its radiotoxicity, allowing separate management (destruction or storage) of nuclear waste components.

Despite the energy and waste disposal benefits obtainable through nuclear reprocessing, reprocessing has been politically controversial because of the potential to contribute to nuclear proliferation, the potential vulnerability to nuclear terrorism, and because of its high cost compared to the once-through fuel cycle.

Contents

1 Separated components and disposition

2 History

3 Separation technologies

3.1 Water and organic solvents

3.1.1 PUREX

3.1.2 Modifications of PUREX

3.1.2.1 UREX

3.1.2.2 TRUEX

3.1.2.3 DIAMEX

3.1.2.4 SANEX

3.1.2.5 UNEX

3.1.3 Electrochemical methods

3.1.4 Obsolete methods

3.1.4.1 Bismuth phosphate

3.1.4.2 Hexone or Redox

3.1.4.3 Butex, β,β'-dibutyoxydiethyl ether

3.2 Pyroprocessing

3.2.1 Electrolysis

3.2.1.1 PYRO-A and -B for IFR

3.2.2 Voloxidation

3.2.3 Volatilization in isolation

3.2.4 Fluoride volatility

3.2.5 Chloride volatility and solubility

4 Economics

5 List of sites

6 See also

The potentially useful components dealt with in nuclear reprocessing comprise specific actinides (plutonium, uranium, and some minor actinides). The lighter elements components include fission products, activation products, and cladding.material disposition

plutonium, minor actinides, reprocessed uranium fission in fast, fusion, or subcritical reactor

reprocessed uranium, cladding, filters less stringent storage as intermediate-level waste

long-lived fission and activation products nuclear transmutation or geological repository

medium-lived fission products 137Cs and 90Sr medium-term storage as high-level waste

useful radionuclides and noble metals industrial and medical uses

History

The first large-scale nuclear reactors were built during World War II. These reactors were designed for the production of plutonium for use in nuclear weapons. The only reprocessing required, therefore, was the extraction of the plutonium (free of fission-product contamination) from the spent natural uranium fuel. In 1943, several methods were proposed for separating the relatively small quantity of plutonium from the uranium and fission products. The first method selected, a precipitation process called the Bismuth Phosphate process, was developed and tested at the Oak Ridge National Laboratory (ORNL) in the 1943-1945 period to produce quantities of plutonium for evaluation and use in weapons programs. ORNL produced the first macroscopic quantities (grams) of separated plutonium with these processes.

The Bismuth Phosphate process was first operated on a large scale at the Hanford Site, in the latter part of 1944. It was successful for plutonium separation in the emergency situation existing then, but it had a significant weakness: the inability to recover uranium.

The first successful solvent extraction process for the recovery of pure uranium and plutonium was developed at ORNL in 1949. The PUREX process is the current method of extraction. Separation plants were also constructed at Savannah River Site and a smaller plant at West Valley, New York which closed by 1972 because of its inability to meet new regulatory requirements.

Reprocessing of civilian fuel has long been employed in Europe, at the COGEMA La Hague site in France, the Sellafield site in the United Kingdom, the Mayak Chemical Combine in Russia, and at sites such as the Tokai plant in Japan, the Tarapur plant in India, and briefly at the West Valley Reprocessing Plant in the United States.

In October 1976, fear of nuclear weapons proliferation (especially after India demonstrated nuclear weapons capabilities using reprocessing technology) led President Gerald Ford to issue a Presidential directive to indefinitely suspend the commercial reprocessing and recycling of plutonium in the U.S. On April 7, 1977 , President Jimmy Carter banned the reprocessing of commercial reactor spent nuclear fuel. The key issue driving this policy was the serious threat of nuclear weapons proliferation by diversion of plutonium from the civilian fuel cycle, and to encourage other nations to follow the USA lead.[4] . After that, only countries that already had large investments in reprocessing infrastructure continued to reprocess spent nuclear fuel. President Reagan lifted the ban in 1981, but did not provide the substantial subsidy that would have been necessary to start up commercial reprocessing.[5]

In March 1999, the U.S. Department of Energy (DOE) reversed its own policy and signed a contract with a consortium of Duke Energy, COGEMA, and Stone & Webster (DCS) to design and operate a Mixed Oxide (MOX) fuel fabrication facility. Site preparation at the Savannah River Site (South Carolina) began in October 2005.[6][7]

Separation technologies

Water and organic solvents

PUREX

Main article: PUREX

PUREX, the current standard method, is an acronym standing for Plutonium and Uranium Recovery by EXtraction. The PUREX process is a liquid-liquid extraction method used to reprocess spent nuclear fuel, in order to extract uranium and plutonium, independent of each other, from the fission products. This is the most developed and widely used process in the industry at present. When used on fuel from commercial power reactors the plutonium extracted typically contains too much Pu-240 to be useful in a nuclear weapon. However, reactors that are capable of refuelling frequently can be used to produce weapon-grade plutonium, which can later be recovered using PUREX. Because of this, PUREX chemicals are monitored.[citation needed]

Modifications of PUREX

UREX

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

The UREX process is a PUREX process which has been modified to prevent the plutonium from being extracted. This can be done by adding a plutonium reductant before the first metal extraction step. In the UREX process, ~99.9% of the uranium and >95% of technetium are separated from each other and the other fission products and actinides. The key is the addition of acetohydroxamic acid (AHA) to the extraction and scrub sections of the process. The addition of AHA greatly diminishes the extractability of plutonium and neptunium, providing greater proliferation resistance than with the plutonium extraction stage of the PUREX process.

TRUEX

Adding a second extraction agent, octyl(phenyl)-N, N-dibutyl carbamoylmethyl phosphine oxide(CMPO) in combination with tributylphosphate, (TBP), the PUREX process can be turned into the TRUEX (TRansUranic EXtraction) process. TRUEX was invented in the USA by Argonne National Laboratory and is designed to remove the transuranic metals (Am/Cm) from waste. The idea is that by lowering the alpha activity of the waste, the majority of the waste can then be disposed of with greater ease. In common with PUREX this process operates by a solvation mechanism.

DIAMEX

As an alternative to TRUEX, an extraction process using a malondiamide has been devised. The DIAMEX (DIAMideEXtraction) process has the advantage of avoiding the formation of organic waste which contains elements other than carbon, hydrogen, nitrogen, and oxygen. Such an organic waste can be burned without the formation of acidic gases which could contribute to acid rain. The DIAMEX process is being worked on in Europe by the French CEA. The process is sufficiently mature that an industrial plant could be constructed with the existing knowledge of the process. In common with PUREX this process operates by a solvation mechanism.

SANEX

Selective ActiNide EXtraction. As part of the management of minor actinides it has been proposed that the lanthanides and trivalent minor actinides should be removed from the PUREX raffinate by a process such as DIAMEX or TRUEX. In order to allow the actinides such as americium to be either reused in industrial sources or used as fuel, the lanthanides must be removed. The lanthanides have large neutron cross sections and hence they would poison a neutron driven nuclear reaction. To date the extraction system for the SANEX process has not been defined, but currently several different research groups are working towards a process. For instance the French CEA is working on a bis-triazinyl pyridine (BTP) based process.[8] Other

UNEX

The UNiversal EXtraction process was developed in Russia and the Czech Republic; it is designed to completely remove the most troublesome radioisotopes (Sr, Cs and minor actinides) from the raffinate remaining after the extraction of uranium and plutonium from used nuclear fuel.[9][10] The chemistry is based upon the interaction of caesium and strontium with polyethylene glycol)[11] and a cobalt carborane anion (known as chlorinated cobalt dicarbollide). The actinides are extracted by CMPO, and the diluent is a polar aromatic such as nitrobenzene. Other dilents such as meta-nitrobenzotrifluoride and phenyl trifluoromethyl sulfone[12] have been suggested as well.

Electrochemical methods

An exotic method using electrochemistry and ion exchange in ammonium carbonate has been reported.

Obsolete methods

Bismuth phosphate

The bismuth phosphate process is an obsolete process that adds significant unnecessary material to the final radioactive waste. The bismuth phosphate process has been replaced by solvent extraction processes. The bismuth phosphate process was designed to extract plutonium from aluminium-clad nuclear fuel rods, containing uranium. The fuel was declad by boiling it in caustic soda. After decladding, the uranium metal was dissolved in nitric acid.

The plutonium at this point is in the +4 oxidation state. It was then precipitated out of the solution by the addition of bismuth nitrate and phosphoric acid to form the bismuth phosphate. The plutonium was coprecipitated with this. The supernatant liquid (containing many of the fission products) was separated from the solid. The precipitate was then dissolved in nitric acid before the addition of an oxidant such as potassium permanganate which converted the plutonium to PuO22+ (Pu VI), then a dichromate salt was added to maintain the plutonium in the +6 oxidation state.

The bismuth phosphate was next re-precipitated leaving the plutonium in solution. Then an iron (II) salt such as ferrous sulfate was added, and the plutonium re-precipitated again using a bismuth phosphate carrier precipitate. Then lanthanum salts and fluoride were added to create solid lanthanum fluoride which acted as a carrier for the plutonium. This was converted to the oxide by the action of an alkali. The lanthanum plutonium oxide was next collected and extracted with nitric acid to form plutonium nitrate.[14]

Hexone or Redox

This is a liquid-liquid extraction process which uses methyl isobutyl ketone as the extractant. The extraction is by a solvation mechanism. 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 (D value). Also, hexone is degraded by concentrated nitric acid. This process has been replaced by the PUREX process.[15][16]

Pu4+ + 4NO3- + 2S --> [Pu(NO3)4S2]