Nuclear power consists of a large number enterprises for various purposes. Raw materials for this industry are extracted from uranium mines. After that, it is delivered to fuel manufacturing enterprises.

Further, the fuel is transported to nuclear power plants, where it enters the reactor core. When the nuclear fuel fulfills its term, it is subject to disposal. It should be noted that hazardous wastes appear not only after fuel processing, but also at any stage - from uranium mining to work in a reactor.

Nuclear fuel

Fuel is of two types. The first is uranium mined in mines, respectively, of natural origin. It contains raw materials that are capable of forming plutonium. The second is a fuel that is artificially created (secondary).

Also, nuclear fuel is divided according to chemical composition: metallic, oxide, carbide, nitride and mixed.

Uranium mining and fuel production

A large share of uranium production is accounted for by just a few countries: Russia, France, Australia, the USA, Canada and South Africa.

Uranium is the main element for fuel in nuclear power plants. To get into the reactor, it goes through several stages of processing. Most often, uranium deposits are located next to gold and copper, so its extraction is carried out with the extraction of precious metals.

In mining, people's health is at great risk because uranium is a toxic material, and the gases that are released during its mining cause various forms of cancer. Although the ore itself contains a very small amount of uranium - from 0.1 to 1 percent. The population that lives near uranium mines is also at great risk.

Enriched uranium is the main fuel for nuclear power plants, but after its use, a huge amount of radioactive waste remains. Despite all its danger, uranium enrichment is an integral process of creating nuclear fuel.

IN natural form uranium is almost impossible to use anywhere. In order to use it, it must be enriched. Gas centrifuges are used for enrichment.

Enriched uranium is used not only in nuclear energy, but also in the production of weapons.

Transportation

At any stage of the fuel cycle there is transportation. It is carried out by all accessible ways A: by land, by sea, by air. This is a big risk and a big danger not only for the environment, but also for humans.

During the transportation of nuclear fuel or its elements, many accidents occur, resulting in the release of radioactive elements. This is one of the many reasons why it is considered unsafe.

Decommissioning of reactors

None of the reactors has been dismantled. Even the infamous Chernobyl The thing is that, according to experts, the cost of dismantling is equal to, or even exceeds, the price of building a new reactor. But no one can say for sure how much money will be needed: the cost was calculated on the basis of the experience of dismantling small stations for research. Experts offer two options:

  1. Place reactors and spent nuclear fuel in burial grounds.
  2. Build sarcophagi over decommissioned reactors.

In the next ten years, about 350 reactors around the world will reach their end of life and must be decommissioned. But since the most suitable method in terms of safety and price has not been invented, this issue is still being resolved.

There are currently 436 reactors operating around the world. Of course, this is a big contribution to the energy system, but it is very unsafe. Studies show that in 15-20 years, nuclear power plants will be able to be replaced by stations that operate on wind energy and solar panels.

Nuclear waste

Great amount nuclear waste formed as a result of the operation of nuclear power plants. Reprocessing nuclear fuel also leaves behind hazardous waste. However, none of the countries found a solution to the problem.

Today, nuclear waste is kept in temporary storage facilities, in pools of water, or buried shallow underground.

Most safe way- this is storage in special storage facilities, but radiation leakage is also possible here, as with other methods.

In fact, nuclear waste has some value, but requires strict adherence to the rules for its storage. And this is the most pressing problem.

An important factor is the time during which the waste is hazardous. Each has its own decay time, during which it is toxic.

Types of nuclear waste

During the operation of any nuclear power plant, its waste enters the environment. This is water for cooling turbines and gaseous waste.

Nuclear waste is divided into three categories:

  1. Low level - clothes of NPP employees, laboratory equipment. Such waste can also come from medical institutions, scientific laboratories. They do not pose a great danger, but require compliance with security measures.
  2. Intermediate level - metal containers in which fuel is transported. Their radiation level is quite high, and those who are close to them must be protected.
  3. High-level - this is spent nuclear fuel and products of its processing. The level of radioactivity is rapidly decreasing. There is very little high-level waste, about 3 percent, but it contains 95 percent of all radioactivity.

Chemical processing of irradiated nuclear fuel is carried out in order to extract plutonium, uranium and other valuable components and purify them from fission products. In the laboratories of nuclear centers in many countries, various methods reprocessing of irradiated fuel, which can be classified as aquatic And non-aqueous. On an experimental scale, such methods as: bismuth-phosphate, trigly, butex, torex, extraction with amines, aqua-fluorine process - water methods; sublimation of fluorides, melting-refining with selective oxidation, electrolysis of salts - non-aqueous methods.

In a number of countries, research and development of so-called dry(anhydrous) methods of chemical regeneration: fluoride (based on the conversion of U and Pu into the gaseous phase of hexafluorides), pyrometallurgical, extraction, in molten salts, etc. Their goal is to provide the most technically and economically efficient industrial regeneration technology while simultaneously solving the problem of processing, conservation and disposal of radioactive waste in the most compact and safest form for storage. It is assumed that dry methods will make it possible to regenerate fuel in the cores of fast breeder reactors with a short exposure of this fuel and with less fuel loss compared to liquid extraction. These methods are also attractive in that the specific volumes of radioactive waste produced are small (mostly in a solid compact form suitable for conservation during regeneration). Most of the installations on which research and development of the above methods were carried out are not currently functioning.

Aqueous processing methods based on the use of liquid countercurrent extraction have been intensively developed. Among them is the water-extraction technology for separating and purifying uranium and plutonium from fission products with tributyl phosphate ( purex process) is recognized as the most effective and is used on all existing industrial enterprises for SNF processing. This method is the only industrially mastered method of chemical processing of uranium oxide fuel spent in NPP reactors.

Extraction of uranium and plutonium with tributyl phosphate technological scheme, called the Purex process, first used in the United States in 1945. to isolate plutonium from irradiated natural uranium metal. This method has various improvements and technological options aimed at reducing the radiation impact on the extractant and achieving a deeper purification of uranium and plutonium from fission products. These improvements made it possible to apply the Purex process to the processing of oxide fuels.

In both wet and dry spent fuel chemical reprocessing processes (and their associated difficulties) the purification, conservation, and removal of gaseous and volatile fission products are very similar, although iodine and tritium capture and removal are simplified in dry processes. Figure 19 shows a diagram of the main stages of preparation and radiochemical processing of spent fuel by the liquid extraction method.

For spent fuel from thermal neutron reactors of the LWR (USA), VVER and RBMK (Russia) types, the optimal holding time in water pools at NPPs is 3-5 years, the minimum is 1 year. For fast breeder reactors, the standard residence time of fuel assemblies in spent fuel pools has not yet been established. In order to obtain a short fuel doubling time, this time should be minimal (no more than a year).

The fuel delivered from the NPP to the radiochemical plant is reloaded under water from containers to the pool of storage depots, where fuel assemblies are installed in special racks or racks, placed so that in any case the critical mass is not reached and the necessary cooling is provided. The depth of the basins and the thickness of the water layer above the fuel assemblies are calculated in such a way as to create the necessary radiation protection. The pools have a closed circulation system for cooling and water purification and are equipped with air exhausts into a special ventilation cleaning system.

From the pools, fuel assemblies enter the cutting department, which is the most complex complex of a radiochemical plant, equipped with remote-controlled equipment. The cutting of fuel assemblies before fuel dissolution at plants in the USA and Western Europe (except for the Eurochemic plant in Mol, Belgium) is carried out by mechanical means: cutting with the help of special presses, cutting the whole fuel assemblies with milling cutters without disassembling into separate fuel rods, while the end parts are preliminarily cut off (“ idle ends") that do not contain fuel. At the Eurochemic plant in Belgium chemical removal of zirconium fuel claddings was used. The disadvantage of this method is a large amount (8-10 m 3 /t of uranium) of intermediate radioactive waste. Installations are being developed for cutting with a laser beam (Great Britain, France), as well as for disassembling fuel assemblies into separate fuel elements and their cutting. To ensure better solubility, fuel rods are cut into pieces 15-50 mm long. The cut pieces fall into the troughs and enter the boron stainless steel batch solvent tanks. In these tanks, uranium and plutonium are leached (extracted) using heated strong nitric acid. Complete dissolution of oxide fuel occurs in 2-4 hours, metal - in 24 hours.

In France and the United States, continuous dissolution apparatuses of the drum type are being developed. Nuclear safety is achieved by adding neutron absorbers (for example, gadolinium) to the solution or by combining safe geometry and absorber inserts. The solutions are carefully filtered using fine-pored stainless steel filters (pore diameter of the order of 3 µm) or centrifuges. The dissolution of uranium dioxide in nitric acid occurs according to the reaction:

UO 2 + 4HNO 3 → UO 2 (NO 3) 2 + 2NO 2 + 2H 2 O

For more complete dissolution of plutonium, additional operations are introduced. Uranium metal is dissolved in boiling strong nitric acid. For the recombination of nitrogen oxides, oxygen is added to the system and, as a result, nitric acid is obtained, which is again returned to the cycle.

A carefully filtered aqueous solution of uranyl nitrate UO 2 (NO 3) 2 with accompanying soluble fission products is fed to solvent extraction.

The main process of solvent extraction is the distribution of a solute between two immiscible liquids (aqueous and organic phases). Between these phases, according to a well-known law, solutes are distributed in each stage in a certain constant ratio. The ratio of the concentration of a substance in the organic phase to its concentration in water phase under conditions of equilibrium between the phases is called distribution ratio.

With several successive extraction processes, it is possible to concentrate almost 100% of uranium and plutonium nitrates in the organic phase, providing the necessary purification factor from radioactive fission products: 5 10 7 -10 8 for plutonium, 10 6 -10 7 for uranium.

Thus, multistage extraction with an organic solvent makes it possible to have both a high extraction of nuclear fuel from solutions and its deep purification from radioactive fission products. The degree of this purification should allow working with regenerated uranium without biological protection, i.e. its radioactivity should be close to natural radioactivity (~ 0.3 μCi/kg or 1.1·10 4 dispersed/(s kg)). This determines the purification limit to which one should strive in the chemical reprocessing of spent fuel.

Tributyl phosphate (TBP) diluted up to 30% with purified kerosene (N-dodecane) has been successfully used as an organic extractor-solvent. The main advantage of TBP as an extractant is its ability to selectively extract uranium and plutonium from a nitric acid solution. In this case, nitric acid serves as a salting out agent. Nitric acid is easily purified by distillation, which makes it possible to return it to the process and not increase radioactive discharges due to it. The organic phase selectively extracts only uranium and plutonium, leaving almost all fission products in the water-acid phase, in which the highly active waste products of the process are thus concentrated. The organic phase containing uranium and plutonium is washed with nitric acid to remove various contaminants and then sent to the second apparatus, where it contacts with water, which washes uranium and plutonium from TBP, transferring them back to the aqueous phase (re-extraction). This completes the first extraction cycle.

In the second extraction cycle, or the U-Pu separation cycle, the liquid water phase from the first cycle (after concentration in the evaporator) is sent back to the extraction-flush contactor (column). The feed phase (organic extract) is fed into another column, where the uranium is separated from the plutonium by contacting the organic phase with an aqueous solution containing a reducing agent (usually tetravalent uranium is used). The tetravalent plutonium is reduced to the trivalent state, in which it is less susceptible to extraction by TBP and therefore can be removed from the column in the aqueous phase. The solution of plutonium in nitric acid is concentrated, then denitrated and converted into dry powder of plutonium dioxide PuO 2 . uranium is removed from the organic phase in the third column. For complete extraction of the uranium product, two or three additional extraction cycles with an organic solvent are used.

For purification from fission products (especially from ruthenium) and concentration of plutonium, one additional extraction cycle is required, followed by treatment with an anion exchange reagent.

The waste remaining in the nitric acid is evaporated to concentrate and store, purify and return the nitric acid to the process.

The organic solvent (TBP) at the outlet of the extraction process is purified from the remaining uranium. Plutonium and fission products, as well as dissolved substances that ended up in TBP due to chemical and radiochemical damage to the organic phase. The solvent purification process typically includes alkaline and acid washing. After purification, the organic solvent (solvent) is returned to the process.

Extraction cycles at reprocessing plants make it possible to isolate 98.5-99.5% of the uranium and plutonium contained in reprocessed fuel rods and achieve high fission product removal rates. There are difficulties in cleaning working solutions from zirconium, niobium and ruthenium. The radioactive isotope 95 Zr (T 1/2 = 65 days) is formed during the fission of uranium by thermal neutrons with a yield of 6.2%. Decaying, it turns into 95 Nb (T 1/2 = 35 days), which, in turn, turns into stable 95 Mo. These elements, like uranium and plutonium, are also extracted by TBP, forming complex compounds, colloids, and adsorbed on solid materials. 103 Ru (T 1/2 = 39.35 days) and 106 Ru (E 1/2 = 1 year) also have significant yields in the fission of uranium with thermal neutrons (3 and 0.38%, respectively) and an even greater yield in fission with fast neutrons. . To get rid of these "annoying and harmful satellites", a number of processes that complicate and increase the cost of technology are used, including operations for preliminary purification of solutions, the mandatory introduction of two extraction cycles for both uranium and plutonium, additional purification on absorbents, as well as through ion exchange and etc.

In the first extraction cycle, it is possible to almost completely get rid of long-lived isotopes of cesium, strontium, yttrium, as well as rare earth elements. All of them form simple hydrated ions in nitric acid solutions. It does not cause any particular difficulties in cleaning from stable nuclides - products of corrosion of the walls of the apparatus, components of shell alloys.

Washing of uranyl nitrate and plutonium nitrate from TBP and removal of residual fission products and decomposition products of TBP is carried out using aqueous solutions of sodium hydroxide, soda, nitric acid and other reagents or by steam distillation. With the help of centrifugal extractors, very short contact and phase separation times are achieved, which contributes to the radiolysis stability of TBP when exposed to intense irradiation.

The final stage of the fuel cycle of the nuclear power industry - the chemical processing of spent nuclear fuel - against the background of the rapid growth in the pace of construction of nuclear power plants, turned out to be the most lagging behind the level of industrial and technological development of other stages of the nuclear fuel cycle. This is due to the fact that the cost of uranium extracted from irradiated fuel still far exceeds its cost in mining, extraction and enrichment. Plutonium has so far found use only in the form of MOX, a fuel that is produced in France.

Technical data on the main radiochemical plants foreign countries are given in Table 19. In Russia, SFAs are reprocessed at the Mayak Production Association (PO).

Table 19

Technical data of SNF reprocessing plants

*) - at the end of 1976, the NFS company announced the final rejection of further operation and reconstruction of its plant due to the seismicity of the West Valley region and the upcoming high costs (~ 600 million dollars). Since 1977, work on the chemical processing of NPP fuel in the USA has been stopped, and radiochemical plants have been mothballed for an indefinite period. However, research and development work continued. Construction of federal long-term SFA storage facilities is underway. Currently Government program development of nuclear energy in the United States provides for a return to industrial processing spent fuel.

**) - the Eurochemic plant in Mol was dismantled in 1979.

***) - in Germany, for a number of years, there have been heated discussions about the admissibility, for reasons of safety and security environment, construction of radiochemical plants and long-term storage facilities for radioactive waste in the country. Until 2007, the Government of Germany did not make a decision.

Like any other production, fuel processing represents a certain environmental hazard. Peculiarities technological process, in terms of environmental education hazardous waste production can be considered on the example of a large plant designed by KEWA for the processing of oxide fuel from PWR and BWR reactors in West Germany. Its productivity is 1400 tons of uranium per year (about 5 tons per day). The standard content of plutonium in SFAs of PWR and BWR reactors does not exceed 0.8%, and fission products - 3% of the mass of the fuel rod (2.3·10 6 Ci/t). Most of the fuel is expected to be delivered to the plant in 120 ton containers. The holding time in the reactor pools is 3 years. It is supposed to use dry unloading. The assemblies are placed in the pools on special racks. Two pools of 700 tons of uranium each are designed for the maximum amount of fuel supplies. The generated heat will be removed by cooling units.

At the first stage of processing, fuel assemblies will be cut with shears into pieces 20-50 mm long, and then the fuel will be dissolved in boiling nitric acid. The resulting gaseous fission products will be discharged to an off-gas cleaning plant. Iodine is supposed to be captured by a filter made of inorganic material containing silver. To capture krypton, a low-temperature distillation method has been designed. The pieces of shells remaining after the dissolution of the fuel will be sent directly to the solid waste storage, and finely dispersed (~ 1 μm) insoluble particles will be filtered out and the clarified solution will be fed for extraction.

The designed extraction scheme provides for the following main technological purex processes. In three extraction cycles, uranium, plutonium and fission products are isolated from the solution. In the first cycle, using several stages of pulse columns, fission products are separated, and uranium and plutonium are also separated. In the second and third extraction cycles, extraction purification of solutions of uranyl and plutonium nitrates is carried out, which then enter the intermediate storage. The technological scheme includes auxiliary processes of acid regeneration, extractant purification, preparation of chemical reagent solutions and gaseous waste purification. The final purification of uranium takes place in silica gel columns. The high 235 U solution is then converted directly at the plant into UF 4 suitable for intermediate storage, which is used as needed to produce UF 6 . The heavily depleted uranium solution is evaporated, followed by the production of UO 3 , which is stored at the plant until it is sent to permanent storage.

Plutonium nitrate is immediately converted to dioxide after extraction. This product can then be sent to a fuel fabrication facility or a central plutonium storage facility.

For intermediate storage of high-level solid waste (pieces of shells, sediments), special storage facilities are designed. In the future, these wastes will be cemented and sent for permanent storage. Other non-combustible wastes will be treated in a similar way after their preliminary cleaning and grinding. combustible solid waste will be incinerated, and the residues will be cemented and stored in metal containers. Stainless steel tanks will be used for temporary storage of high level liquid waste. After a significant decrease in activity, the liquid waste will solidify and vitrify. Liquid waste of medium activity (after extraction of organic components and free acids) will be concentrated and temporarily stored in liquid form. Liquid waste of low activity will be separated by distillation, concentration and chemical treatment into a fraction that can be safely discharged into the environment, and a distillation residue of medium activity. The 85 Kr liquefied during the gaseous waste treatment process will be stored in sealed cylinders. After a significant decrease in activity during the temporary storage period, all waste will be sent to a permanent storage located in the workings of the salt mine. The number of personnel of the plant is 1000 people. Some significant technical indicators of the plant are given in Table 20.

Table 20

Specifications project plant for spent nuclear fuel reprocessing

The construction of such a plant costs several billion dollars, the price of processing is several hundred dollars per kilogram of uranium. It is clear that the proceeds from the sale of uranium and plutonium extracted during fuel reprocessing, under such conditions, will cover only part of the costs of reprocessing itself, neutralization and disposal of waste. Therefore, the reprocessing of fuel from thermal neutron reactors should not be considered as a possible source of income and profit, but rather as a necessary production process that ensures the neutralization and removal of radioactive waste, as well as the conservation and increase in raw materials through the use of unburned uranium and plutonium formed during fuel irradiation. .

France is the most actively involved in fuel reprocessing among Western countries at the radiochemical plant in Cape Ag. Moreover, this plant processes not only French fuel, but also from other countries (Japan, Germany).

Prospects for reprocessing in the future are also associated with the reprocessing of uranium-plutonium fuel from fast reactors.

Along with the development of industrial technologies for reprocessing irradiated fuel at pilot and pilot plants and plants in various countries laboratory research is being carried out aimed at improving individual stages in the technology of the purex process, searching for and testing new extractants and developing new fuel processing processes. In the future, the task is to develop a technology for reprocessing irradiated fuel that provides:

· removal of actinides from high-level waste, which will reduce the time during which the waste remains hazardous from 25·10 4 to 10 3 years;

Reducing the amount of waste from fuel processing by 20 times compared to modern technology based on the Purex process;

Extraction of noble metals such as palladium, rhodium and ruthenium.

In all countries, with the exception of the United States, scientific research is carried out in centers owned by government bodies management and control over the use of atomic energy. In the US, part of the research is transferred to private firms under government contracts (under the US Department of Energy).



The owners of the patent RU 2560119:

The invention relates to means for processing spent nuclear fuel (SNF). In the claimed method, oxide spent nuclear fuel pellets destroyed during cutting of fuel rods are subjected to dissolution by heating in an aqueous solution of iron(III) nitrate at a molar ratio of iron to uranium in the fuel equal to 1.5-2.0:1, the resulting precipitate of the basic salt of iron with undissolved fission products of nuclear fuel are separated by filtration, and uranyl peroxide is precipitated from the resulting weakly acidic solution by successively feeding the disodium salt of ethylenediaminetetraacetic acid into the solution with stirring. Next, the resulting heterogeneous system is kept for at least 30 minutes, and after separation and washing with acid and water, the precipitate of uranyl peroxide is subjected to solid-phase reduction when heated by treating it with an alkaline solution of hydrazine hydrate in water at a 2-3-fold molar excess of hydrazine relative to uranium, followed by separation obtained hydrated uranium dioxide UO 2 ·2H 2 O, washing it with a solution of HNO 3 with a concentration of 0.1 mol/l, water and drying. In this case, the precipitate of basic iron salts with fission products, the mother liquor of the peroxide precipitation stage with the remains of fission products, the waste of alkaline and washing solutions are sent to the waste collector for their subsequent processing. The technical result is to increase environmental safety and reduce the amount of waste. 8 w.p. f-ly.

The invention relates to the field of nuclear energy, in particular to the reprocessing of spent nuclear fuel (SNF), and can be used in the technological scheme of reprocessing, including MOX fuel, since the extraction of the remaining amounts of U and Pu from SNF for the preparation of new fuel is the main the task of a closed nuclear fuel cycle, to which the country's nuclear power industry is oriented. Currently, it is relevant to create and optimize new, low-waste, environmentally safe and economically viable technologies that would ensure the reprocessing of spent nuclear fuel from both operating and 3rd and 4th generation fast neutron reactors operating on mixed oxide uranium-plutonium fuel (MOX fuel). ).

Known methods of processing SNF using fluorine or fluorine-containing chemical compounds. The resulting volatile fluorine compounds of the nuclear fuel components pass into the gas phase and are distilled off. During fluorination, uranium dioxide is converted into UF 6 , which evaporates relatively easily, in contrast to plutonium, which has a lower volatility. Usually, when SNF is reprocessed in this way, SNF is fluorinated, extracting from it not all of the uranium contained in it, but only its required amount, thus separating it from the rest of the processed fuel. After that, the evaporation mode is changed and a certain amount of plutonium contained in it is also removed from the SNF residue in the form of vapors.

[RF patent No. 2230130, S22V 60/02, publ. 01/19/1976]

The disadvantage of this technology is that this method of SNF processing uses gaseous, aggressive and environmentally toxic chemical compounds. Thus, the technology is environmentally unsafe.

One of the methods close in essence to the claimed method is known method, declared in US Pat. RF No. 2403634, (G21C 19/44, publ. 11/10/2010), according to which SNF regeneration includes the stage of fuel dissolution in a solution of nitric acid, the stage of electrolytic valence control, with the reduction of Pu to the trivalent state and the preservation of the pentavalent state of Np, the stage of extraction of the hexavalent uranium extracting agent in an organic solvent; an oxalic acid precipitation step resulting in co-precipitation of minor actinides and fission products remaining in the nitric acid solution as an oxalate precipitate; a chlorination step to convert the oxalate precipitate to chlorides by adding hydrochloric acid to the oxalate precipitate; a dehydration step to produce synthetic anhydrous chlorides by dehydration of the chlorides in a stream of argon gas; and a molten salt electrolysis step of dissolving anhydrous chlorides in the molten salt and accumulating uranium, plutonium and minor actinides at the cathode by electrolysis.

The disadvantage of this method of SNF processing is its multi-stage nature and complexity in implementation, since it includes electrochemical stages, which are energy-consuming, require special equipment and the process at high temperature, especially when working with molten salts.

There is also a method according to which spent nuclear fuel is processed purely pyrochemically using a salt melt of uranium or plutonium, after which the separated components of nuclear fuel are reused. In the pyrochemical processing of SNF, its induction heating in a crucible and its cooling by supplying a coolant to the crucible are used.

[RF patent No. 2226725, G21C 19/46, publ. 01/19/2009]

Pyrometallurgical technologies do not lead to the formation of large amounts of liquid radioactive waste (LRW), and also provide compact placement of equipment, but they are very energy intensive and technologically complex.

SNF processing methods also include:

(1) a method involving the oxidation of uranium with gaseous chlorine, nitrogen oxides, sulfur dioxide in a dipolar aprotic solvent or mixtures thereof with a chlorine-containing compound [RF patent No. 2238600, G21F 9/28, publ. 04/27/2004];

(2) a method of dissolving materials containing metallic uranium, including the oxidation of metallic uranium with a mixture of tributyl phosphate-kerosene containing nitric acid [US patent No. 3288568, G21F 9/28, publ. 12/10/1966];

(3) a process for dissolving uranium, which involves oxidizing uranium metal with a solution of bromine in ethyl acetate with heat.

The disadvantages of these methods include the increased fire hazard of systems and the limited scope of their use.

A widely used SNF reprocessing technology is the Purex process (which we took as a prototype), in which SNF containing uranium, plutonium and fission products (FP) of nuclear fuel is dissolved in strongly acidic nitric acid solutions when heated to 60-80°C. The actinides are then removed from the nitric acid solution by an organic phase containing tributyl phosphate in kerosene or another organic solvent. This is followed by technological stages associated with the separation of uranium and plutonium and their purification from PD. The Purex process is described, for example, in The Chemistry of the Actinide and Transactinide Elements, 3rd Edition, Edited by Lester R. Morss, Norman M. Edelstein and Jean Fuger. 2006, Springer, pp. 841-844.

The specified SNF reprocessing process is multi-stage and is based on the use of environmentally hazardous media:

(1) nitric acid (6-8 mol/l) as a SNF solvent at 60-80°C and forming aggressive gaseous products during reactions with its participation;

(2) since the acidity of the solution after completion of dissolution is about 3.5 mol/l in nitric acid, this inevitably leads to the use of extraction to extract U(Pu) with organic solvents;

(3) the use of organic solvents, toxic, combustible, highly flammable, explosive and often unstable to radiation, leads to the formation of large volumes of waste together with aqueous LRW (up to 7-12 tons per 1 ton of processed SNF).

The objective of the present invention is to create an innovative, low-waste, environmentally safe and economically viable technology for spent nuclear fuel reprocessing.

The problem is solved by using a new method of spent nuclear fuel processing, which is characterized by the fact that oxide spent nuclear fuel pellets destroyed during cutting of fuel rods are subjected to dissolution when heated in an aqueous solution of iron(III) nitrate at a molar ratio of iron to uranium in the fuel equal to 1.5-2, 0:1, the resulting precipitate of the basic salt of iron with undissolved fission products of nuclear fuel is separated by filtration, and uranyl peroxide is precipitated from the resulting weakly acidic solution containing mainly uranyl nitrate by sequentially feeding into the solution with stirring the disodium salt of ethylenediaminetetraacetic acid in a molar excess with respect to uranium, equal to 10%, and 30% hydrogen peroxide solution, taken in a 1.5-2-fold molar excess relative to uranium, at a temperature not exceeding 20°C, the resulting heterogeneous system is kept for at least 30 minutes and after separation and washing with acid and water, the precipitate of uranyl peroxide is subjected to solid-phase reduction when heated by treating it with an alkaline solution of hydrazine hydrate in water at a 2-3-fold molar excess of hydrazine relative to uranium, followed by separation of the resulting hydrated uranium dioxide UO 2 2H 2 O, washing it with a solution of HNO 3 with a concentration of 0.1 mol / l, water and drying, while the precipitate of basic iron salts with fission products, the mother liquor of the peroxide precipitation stage with the remains of fission products, waste alkaline and washing solutions are sent to the waste collector for their subsequent processing.

Typically, the dissolution of SNF is carried out in the temperature range of 60-90°C for no more than 5-10 hours using an aqueous solution of iron(III) nitrate with a pH of 0.2 to 1.0.

It is advisable to wash the isolated uranyl peroxide with a solution of HNO 3 with a concentration of 0.05 mol/l, and its solid-phase reduction should be carried out with a 10% aqueous solution of hydrazine hydrate at pH 10 at 60-90°C for 10-15 hours.

Preferably, the drying of the hydrated uranium dioxide is carried out at 60-90°C.

It is possible to conduct the process in two serially connected bifunctional apparatuses, the design of which provides for the presence of a filtration unit and the possibility of changing the spatial orientation of the apparatuses by 180°, the first of which is used for dissolving and collecting process waste, and the second for uranium peroxide precipitation, its solid-phase reduction and isolation target product.

The technical result of the method is achieved by the fact that at all stages of spent nuclear fuel processing, the fuel components (UO 2 with a content of up to 5 wt.% 239 Pu) - U (Pu), dissolving (iron nitrate), precipitating (hydrogen peroxide) and reducing reagents are in different phases suitable for their further separation. At the stage of dissolution, uranium goes into solution, and the bulk of the dissolving reagent is released in the form of a solid compound. At the stage of peroxide precipitation and its solid-phase reductive transformation into uranium dioxide, the target product is in solid form and is easily separated from the liquid phase.

The proposed method is carried out as follows.

The tablets of uranium dioxide (UO 2 containing up to 5 wt.% 239 Pu) destroyed during cutting of fuel rods are immersed in water containing iron(III) nitrate and dissolved when heated to 60-90°C. The resulting solution containing U(Pu) and the pulp of the basic iron salt formed during dissolution are separated. After removal of the solution with U(Pu), the precipitate of the main iron salt—iron salt with PD—Mo, Tc, and Ru (~95%) and partly Nd, Zr, and Pd (~50%)—remains in the waste collector.

Hydrogen peroxide is added to the separated solution with U(Pu), and uranyl peroxide is precipitated at room temperature, with which plutonium is also co-precipitated, the purification factor of the target product from PD is about 1000. PD and Fe(III) nitrate are sent to a waste collector with a precipitate of basic salt. The solution from washing the precipitate of the mixed peroxide is also sent to the waste collector. Further, the solid-phase reduction of the formed peroxide is carried out after the introduction of hydrazine hydrate under stirring with a stream of nitrogen at 80-90°C and hydrated U(Pu) dioxide is obtained. The separated alkaline solution is transported to a waste collector. The precipitate of dioxide is washed with a small volume of 0.1 M HNO 3 , then with distilled water, which are also sent to the waste collector. The resulting target product is dried in a stream of heated nitrogen at 60-90°C and unloaded from the apparatus.

Weakly acidic and slightly alkaline aqueous solutions-wastes, which are collected during the processing of SNF in the waste collector, are removed by evaporation, and the iron contained in them is precipitated in the form of hydroxide together with cations of 2-, 3-, and 4-valent PD. The solid product of iron compounds with PD included in their phase is the only waste in the proposed method of SNF processing. The evaporated water can be condensed and returned, if necessary, to the process.

SNF processing can be carried out in a bifunctional special apparatus (s), the design of which provides for the presence of a filtration unit (UF), a jacket capable of supplying a coolant and carrying out the dissolution process at a temperature of ≤90°C in the reaction mixture, and the ability to change the spatial orientation by 180° device.

The process is carried out, as a rule, in two series-connected bifunctional devices as follows.

When the filtration unit of the device is located in the upper part, the device is designed to dissolve SNF. The resulting solution containing U(Pu) and the slurry of basic iron salt formed upon dissolution of SNF are separated. To do this, the device is turned 180°, while the UV is in the bottom. Filtration is carried out by applying excess pressure to the internal volume of the apparatus, or by connecting it to a vacuum line. After filtration and removal of the solution with U(Pu), the device with a precipitate of iron salt and PD (Mo, Tc and Ru (~95%) and partially Nd, Zr and Pd (~50%)) is turned by 180° to the position where UV is located in the upper part, and then the device performs the function of a collection of waste solutions.

The filtered solution with U(Pu) is fed into the second apparatus of the same design in a position where the UV is located at the top of the apparatus. Hydrogen peroxide is added to the solution, and U(Pu) peroxide is precipitated at room temperature. Having completed the deposition, the device is turned over 180° and a filtration separation is carried out through the bottom of the apparatus. The resulting peroxide remains on the filter in the apparatus, and the mother liquor with dissolved PD (purification factor of about 1000) and residual Fe(III) nitrate is sent to the first apparatus with a basic salt precipitate, which has become a waste collector.

The device is inverted to position with UV at the top and the peroxide precipitate from the filter in the apparatus is washed off with a small amount of water containing hydrazine hydrate to form a slurry in which the peroxide is converted to hydrated U(Pu) dioxide at 80-90°C by solid phase reduction with hydrazine.

Having completed the solid-phase reduction and having obtained hydrated U(Pu) dioxide, the apparatus is transferred to a position in which it performs the filtering function. The separated alkaline solution is sent to the first apparatus with a sediment of basic salt, which has become a waste collector. The precipitate of dioxide is washed with a small volume of 0.1 M HNO 3 , then with distilled water, which are also sent to the waste collector. The device with the precipitate of hydrated U(Pu)O 2 ·nH 2 O is turned by 180° to positions where the UV is located at the top. Next, the target product is dried in the apparatus at 60-90°C by supplying a stream of nitrogen, and upon completion of drying, the preparation is unloaded from the apparatus.

The following examples illustrate the efficiency of using aqueous weakly acidic solutions of Fe(III) nitrate (chloride) for dissolving oxide SNF with simultaneous separation of U(Pu) at this stage from a part of PD, followed by their separation from PD residues during peroxide precipitation of U(Pu) from the resulting solution . Further solid-phase reductive transformation of peroxide, first into hydrated and then into crystalline U(Pu) dioxide, increases the efficiency of the proposed method.

A powdered sample of uranium dioxide (238+235 UO 2 ) was preliminarily calcined at 850°C in an argon atmosphere with 20% hydrogen content for 8 hours.

Tablets or powder of ceramic nuclear fuel containing uranium and 5 wt.% plutonium, weighing 132 g, are immersed in an aqueous solution of iron (III) nitrate with a volume of 1 l with a pH of at least 0.2 at a concentration of Fe (NO 3) 3 in water from 50 up to 300 g / l and dissolve when heated to 60-90 ° C at a molar ratio of Fe (III) to fuel as 1.5 to 1.

The pH value and the uranium content in the solution are controlled and the dissolution of the tablets is continued until the uranium content does not change in successive samples. As a result of the dissolution process, a solution containing predominantly uranyl nitrate and having a pH value of ≤ 2 and a precipitate of basic iron salt are obtained. It takes no more than 5-7 hours for the quantitative dissolution of the samples taken.

The resulting nitrate solution is separated from the pulp by filtration, for example, using a cermet filter. The sediment of the basic iron salt remaining on the filter is washed with water and sent to the waste collector along with the washing water.

To a slightly acidic solution of the separated uranyl nitrate at a temperature of ≤20°C, add 60 ml of a 10% solution of disubstituted sodium salt of EDTA (Trilon-B), stir for 10 minutes. A white complex compound of uranyl precipitates in solution.

With stirring, to the resulting suspension is added in portions of 50 ml with an interval of 1-1.5 min 300 ml of a 30% solution of hydrogen peroxide (H 2 O 2) also at a temperature of ≤20 ° C to obtain uranyl peroxide, with which also quantitatively plutonium co-precipitates.

The precipitate of uranyl peroxide is separated by filtration from the mother liquor, which is sent to the waste collector. The precipitate is washed with 0.25 l of 0.05 M HNO 3 , the wash solution is sent to the waste collector.

The washed precipitate of uranyl peroxide is first transferred into suspension with a 10% aqueous alkaline solution of hydrazine hydrate in water, the solution having a pH value of ~10.

With stirring and heating the suspension to 80°C, uranyl peroxide transforms into hydrated UO 2 ·H 2 O dioxide during the solid phase reduction of U(VI) with hydrazine to U(IV).

Control over the process of reduction of U(VI) to U(IV) is carried out by periodic sampling of suspensions containing no more than 50 mg of solid suspension. The precipitate is dissolved in a mixture of 4M HCl with 0.1M HF, the first spectrum of the solution is recorded. The solution is then treated with amalgam and a second spectrum of this solution is recorded. In this case, all uranium in solution must be completely reduced to U(IV). Thus, if the first and second spectra coincide, then the process of solid-phase reduction is completed. Otherwise, the procedure for converting peroxide to uranium dioxide is continued. The process is completed in 10-15 hours.

The resulting hydrated uranium dioxide is separated by filtration from the alkaline solution (volume ~0.6 l), the solution is sent to the waste collector. The precipitate of hydrated uranium dioxide is washed on the filter with 0.25 l of 0.1 M HNO 3 to neutralize the alkali remaining in the precipitate volume, then with the same volume of water to remove traces of acid from the precipitate volume with pH control of the last wash water. Washing solutions are sent to the waste collector.

The results of analyzes of the mother liquor and uranium peroxide indicate that the degree of precipitation of uranium is not less than 99.5%, and the iron content in the separated peroxide does not exceed 0.02 wt.%.

The precipitate of uranium peroxide, washed from traces of alkali, is dried, for example, with a stream of nitrogen heated to 60-90°C, and unloaded from the apparatus in the form of a powder.

The result is not less than 131.3 g of uranium dioxide.

In the slightly alkaline aqueous solutions collected in the waste collector, iron residues are released in the form of amorphous hydroxide. The heterogeneous suspension is evaporated, and almost complete removal of water is achieved. Wet or dry solid product, which is mainly iron compounds, is the only waste in the claimed method of processing ceramic oxide fuel using solutions of iron(III) nitrate.

The proposed method makes it possible to simplify the processing of spent nuclear fuel and exclude the formation of LRW in comparison with the Purex process.

New essential and distinctive features of the proposed method (in comparison with the prototype) are:

The use of aqueous weakly acidic solutions of Fe(III) nitrate for dissolving oxide SNF, which were not previously used for this. Without a significant deterioration in the dissolving power, iron nitrate can be replaced by Fe(III) chloride;

Unlike the prototype, there is no special stage with the introduction of ferrous sulfate into the system to restore Pu(IV) to Pu(III). In the claimed method, when dissolving oxide uranium and mixed fuel, uranium (IV) is oxidized by Fe (III) to uranium (VI), and the resulting Fe (II) cations reduce Pu (IV) to Pu (III), and the actinides quantitatively pass into solution in the form of their nitrates;

In the claimed method, it is not required to introduce acid to dissolve SNF, since the medium used has an acidity due to the hydrolysis of iron(III) nitrate, and, depending on its concentration from 50 to 300 g/l, the pH value ranges from 1 to 0.3;

In the proposed method, after dissolving the fuel, the acidity of the resulting solutions will be ≤0.1 M (for uranium 100-300 g/l), while in the Purex process, strongly acidic ~3M HNO 3 solutions are formed, which inevitably leads to extraction and formation a large amount of organic and aqueous LRW;

Low acidity after dissolution of SNF according to the claimed method makes it possible to abandon the extraction extraction of fuel components with organic solutions, to simplify the organization of the SNF processing process and to eliminate LRW in comparison with the Purex process technology;

In the proposed method, the process of fuel dissolution is completed by obtaining a solution containing U(Pu) and a precipitate of the main salt of iron, in an amount of ~50% of the initial content of iron(III) nitrate;

Fission products, such as Mo, Tc, and Ru (~95%) and partly from Nd, Zr, and Pd (~50%), are separated from uranium already at the stage of SNF dissolution and are concentrated in the formed precipitate of the basic iron salt. This is also an advantage of the proposed method of SNF dissolution in comparison with the Purex process;

In the weakly acidic solutions used, the structural materials of the fuel rod cladding and the phases formed from the FP in the SNF matrix in the form of light metallic (Ru, Rh, Mo, Tc, Nb) and gray ceramic inclusions (Rb, Cs, Ba, Zr, Mo) do not dissolve. Therefore, weakly acid ones will be less contaminated with dissolved shell components and PD, in contrast to 6–8 M HNO 3 in the Purex process;

Acidity ≤0.1 M obtained solutions with a concentration of uranium 100-300 g/l is optimal for the deposition of peroxides of uranium(VI) and plutonium(IV). Hydrogen peroxide is preferred because it converts uranium to the U(VI) state, which is required for quantitative precipitation;

Precipitation of U(Pu) peroxide from solution results in the quantitative separation of U from almost all PD and iron residues present in the solution (purification factor ~1000);

A new and original solution in the proposed method is the process of solid-phase reduction in an aqueous suspension of U(Pu) peroxide with hydrazine hydrate at 90°C to hydrated U(Pu)O 2 ×nH 2 O, followed by drying the target product at 60-90°C and unloading from the apparatus

Weakly acidic and slightly alkaline aqueous waste solutions accumulated during SNF processing in the waste collector are removed during evaporation, and the iron contained in them precipitates in the form of hydroxide together with cations of 2-, 3-, and 4-valent PD. The solid product of iron compounds with included in their phase PD is the only waste in the proposed method of processing oxide SNF.

1. A method for reprocessing spent nuclear fuel, characterized in that the tablets of oxide spent nuclear fuel destroyed during cutting of fuel rods are subjected to dissolution when heated in an aqueous solution of iron(III) nitrate at a molar ratio of iron to uranium in the fuel equal to 1.5-2.0 :1, the resulting precipitate of the basic iron salt with undissolved fission products of nuclear fuel is separated by filtration, and uranyl peroxide is precipitated from the resulting weakly acidic solution containing mainly uranyl nitrate by successively adding disodium salt of ethylenediaminetetraacetic acid to the solution with stirring in a molar excess with respect to uranium equal to 10% and 30% hydrogen peroxide solution, taken in a 1.5-2-fold molar excess with respect to uranium, at a temperature not exceeding 20 ° C, the resulting heterogeneous system is kept for at least 30 minutes and after separation and washing with acid and water the precipitate of uranyl peroxide is subjected to solid-phase reduction when heated by treating it with an alkaline solution of hydrazine hydrate in water at a 2-3-fold molar excess of hydrazine relative to uranium, followed by separation of the resulting hydrated uranium dioxide UO 2 2H 2 O, washing it with a solution of HNO 3 with with a concentration of 0.1 mol/l, water and drying, while the precipitate of basic iron salts with fission products, the mother liquor of the peroxide precipitation stage with residues of fission products, waste alkaline and washing solutions are sent to the waste collector for their subsequent processing.

2. The method of processing spent nuclear fuel according to claim 1, characterized in that the dissolution of spent nuclear fuel is carried out at 60-90°C.

3. The method of processing spent nuclear fuel according to claim 1, characterized in that an aqueous solution of iron (III) nitrate with a pH value of 0.2 to 1.0 is used to dissolve the fuel.

4. The method of processing spent nuclear fuel according to claim 1, characterized in that the dissolution of spent nuclear fuel is carried out for no more than 5-10 hours.

5. A method for processing spent nuclear fuel according to claim 1, characterized in that the precipitate of uranyl peroxide is washed with a solution of HNO 3 with a concentration of 0.05 mol/l.

6. The method of processing spent nuclear fuel according to claim 1, characterized in that solid-phase reduction is carried out with a 10% aqueous solution of hydrazine hydrate at pH 10.

7. The method of processing spent nuclear fuel according to claim 1, characterized in that solid-phase reduction is carried out at 60-90°C for 10-15 hours.

8. The method of processing spent nuclear fuel according to claim 1, characterized in that the drying of hydrated uranium dioxide is carried out at 60-90°C.

9. The method of processing spent nuclear fuel according to any one of paragraphs. 1-8, characterized in that the process is carried out in two serially connected bifunctional apparatuses, the design of which provides for the presence of a filtration unit and the possibility of changing the spatial orientation of the apparatuses by 180 °, the first of which is used to dissolve and collect process waste, and the second to precipitate peroxide uranyl, its solid-phase reduction and isolation of the target product.

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The population of the planet, as well as its need for energy, is only growing every year, along with the prices of gas and oil, the processing of which, by the way, has its sad and irreversible consequences for the ecology of the earth. And nuclear energy today does not have a worthy alternative either in terms of such parameters as profitability, or in terms of the ability to meet the world's energy needs.

Despite the fact that such statements sound very abstract, in practice, the rejection of atomic energy will mean a sharp rise in the price of such things that are necessary for everyone, such as food, clothing, medicine, comfortable Appliances, education, medicine, the ability to move freely around the world and much more. In such a situation, the best solution is to focus on making nuclear energy as safe and efficient as possible.

Not everyone knows this fact: fresh nuclear fuel does not pose any danger to humans. Prior to the widespread introduction of manufacturing automation, uranium dioxide fuel pellets were hammered into assembly rods by hand. The radioactivity of the fuel increases several million times after irradiation in a nuclear reactor. It is at this point that it becomes dangerous for humans and the environment.

Like any production nuclear power plants generate waste. At the same time, the amount of waste produced by nuclear power plants is much less compared to other industries, but due to its high environmental hazard, they require special handling. And here it is necessary to clarify some confusion between the concepts of RW (radioactive waste) and SNF (spent nuclear fuel), which often occurs in the media.

According to the Russian classification, SNF refers to spent fuel elements removed from the reactor. Let us trace the path along which the natural uranium mined in the mines is converted into SNF. As we know, natural uranium consists of the isotopes uranium-235 and uranium-238. Modern nuclear power plants operate on uranium - 235. But due to the low content of the 235 isotope (only 0.7%), for use as nuclear fuel, uranium extracted from the bowels of the earth has to be enriched up to several percent. Uranium used in reactors is placed in fuel elements (TVEL), from which fuel assemblies are assembled in the form of hexagonal rods. They are immersed in the reactor until the critical mass is reached. Before starting the reactor, the fuel rods contain 95% uranium-238 and 5% uranium-235. As a result of the operation of the reactor, instead of uranium-235, fission products - radioactive isotopes - appear. The rods are removed, but already as spent nuclear fuel.

SNF has a rich resource potential. First, the radioisotopes of spent fuel, which can be chemically recovered, have a wide range of medical and scientific applications. And not only for medical purposes - the platinum group metals formed in the reactor during the fission of uranium turn out to be cheaper than the same metals obtained from ore. Secondly, the spent fuel contains uranium-238, which is considered throughout the world as the main fuel element for nuclear power plants of the future. Thus, reprocessed spent nuclear fuel becomes not only the richest source for obtaining fresh nuclear fuel, but also solves the environmental problems of uranium deposits: it makes no sense to develop uranium mines, because already at this moment Russia has accumulated 22,000 tons of SNF. At the same time, the content of radioactive elements in SNF, which cannot be reprocessed and need reliable isolation from the environment, is only 3%. For reference: processing 50 tons of spent nuclear fuel saves 1.6 billion cubic meters natural gas or 1.2 million tons of oil.

Radioactive waste (RW) also contains radioisotopes. The difference lies in the fact that it is not possible to extract them, or the costs of extracting them are not economically feasible. Currently, depending on the type of radioactive waste, there are several ways to handle radioactive waste. The sequence of actions is as follows: to begin with, the volume of radioactive waste is reduced. At the same time, for solid RW, pressing or incineration is used, for liquid RW - coagulation and evaporation, processing through mechanical or ion-exchange filters. After processing using special fabric or fiber filters, the volume of gaseous radioactive waste is reduced. The next stage is immobilization, that is, the placement of RW in a solid matrix of cement, bitumen, glass, ceramics or other materials that reduce the likelihood of RW release into the environment. The resulting masses are placed in special containers and then stored. The final stage is the transfer of containers with radioactive waste to the repository.

According to scientists, the most effective method of disposal of radioactive waste today is in stable geological formations of the earth's crust. This method provides an effective insulating barrier for a period of tens of thousands to millions of years. Published in the electronic bulletin of the European Atomic Society, the results of joint research by the Subatech laboratory in France and the SCK-CEN research center in Belgium showed that the period during which blocks with nuclear waste can maintain their integrity exceeds 100 thousand years. The researchers came to this conclusion after making probabilistic estimates of the possible dissolution of buried nuclear waste from open and closed fuel cycles over various periods of time.

At the recent international scientific and practical conference"Safety, Efficiency and Economics of Nuclear Power", also discussed the pressing problems of SNF management. In Russia, SNF is currently stored and reprocessed by the Mayak Production Association (Ozersk, Chelyabinsk region) and the Mining and Chemical Combine (Zheleznogorsk, Krasnoyarsk region), which are part of the nuclear and radiation safety complex of Rosatom State Corporation. Advisor to the State Corporation "Rosatom" I.V. Gusakov-Stanyukovich spoke about the departmental "Program for creating infrastructure and handling spent nuclear fuel for 2011-2020 and for the period up to 2030." According to him, today, out of the available 22,000 tons of SNF, most of it is at nuclear power plants. At the same time, the amount that is exported for storage during the year is less than the nuclear power plant manages to produce during this time. And if SNF from those plants that use VVER-type reactors (pressure-cooled power reactor) is transported for storage at FSUE MCC or for processing at FSUE PO Mayak, then the main problem Currently, it is spent fuel from RBMK reactors (high power channel reactor), the amount of which is 12.5 thousand tons. Recently, a dry storage facility for RBMK SNF at the Mining and Chemical Combine began to operate, and in the spring of 2012 the first train with SNF from the Leningrad NPP arrived there. In the future, conditioned SNF from the Leningrad, Kursk and Smolensk NPPs will be sent to the Mining and Chemical Combine, and non-conditioned SNF - to the Mayak Production Association.

By 2018, the implementation of the program for creating infrastructure and handling spent nuclear fuel will increase the volume of annual removal of spent nuclear fuel from NPP sites, which will exceed the annual amount of spent nuclear fuel by 1.5 times. And by 2030, all 100% of the SNF from the RBMK-1000 and VVER-1000 reactors will be placed for long-term centralized storage at the MCC site, after which the main specialization of the MCC will be the production of MOX fuel. As for the plans for spent nuclear fuel from the VVER-440 and BN-600 reactors, as well as transport and research reactors, Mayak will deal with the processing of these spent nuclear fuel. The exception will be the Bilibino NPP, whose spent nuclear fuel cannot be transported to centralized reprocessing facilities because of its geographical remoteness, so it will be disposed of on site.

MOSCOW, November 20 - RIA Novosti. Mining and Chemical Combine, an enterprise of the state corporation Rosatom (GKhK, Zheleznogorsk, Krasnoyarsk Territory), has begun pilot processing of spent nuclear fuel (SNF) from Russian NPPs using unique technologies that do not create risks for the environment, on an industrial scale such a "green" processing will start at MCC after 2020.

Earlier, the MCC isotope-chemical plant built the world's most modern start-up complex of the Pilot Demonstration Center (ODC) for radiochemical processing of SNF from nuclear power plant reactors, which will use the latest, environmentally friendly technologies of the so-called generation 3+. The start-up complex will make it possible to work out technological regimes for SNF reprocessing on a semi-industrial scale. In the future, on the basis of the ODC, it is planned to create a large-scale RT-2 plant for the regeneration of spent nuclear fuel.

A feature of the technologies that will be used at the ODC will be complete absence liquid low-level radioactive waste. Thus, Russian specialists will have a unique opportunity for the first time in the world to prove in practice that recycling nuclear materials possible without harming the environment. According to experts, no other country except Russia possesses these technologies now. The construction of the center was technologically the most complex project ever recent history GCC.

The first ever spent fuel assembly of the VVER-1000 reactor from the Balakovo NPP, which was stored at the plant for 23 years, was placed in one of the "hot cells" of the ODC - a box for remotely controlled work with highly radioactive substances, the corporate publication of the Russian nuclear industry newspaper reported on Monday "Country Rosatom".

“We are starting to work out the modes (processing of spent nuclear fuel). Now the main thing is to qualitatively work out the technology that will be in the basic scheme of the RT-2 plant,” explained Igor Seelev, director of the isotope-chemical plant of the MCC, as quoted by the newspaper.

"Green" technologies

First, the so-called thermochemical opening and fragmentation of the spent fuel assembly is carried out. Then begins voloxidation (from the English volume oxidation, volumetric oxidation) - an operation that distinguishes generation 3+ of spent nuclear fuel processing from the previous generation. This technology makes it possible to distill radioactive tritium and iodine-129 into the gas phase and prevent the formation of liquid radioactive waste after dissolving the contents of the fuel assembly fragments.

After voloxidation, the fuel is sent for dissolution and extraction. Uranium and plutonium are separated and returned to the fuel cycle in the form of uranium and plutonium dioxide, from which it is planned to produce mixed oxide uranium-plutonium MOX fuel for fast neutron reactors and REMIX fuel for thermal neutron reactors that form the basis of modern nuclear energy.

The fission products are conditioned, vitrified and packaged in a protective container. Liquid radioactive waste does not remain.

After working out new technology SNF reprocessing will be scaled up to be used in the second, full-scale stage of the OFC, which will become the industrial basis for the closed nuclear fuel cycle (CFFC). Now the construction of the building and the second stage of the ODC is being completed. It is expected that the experimental demonstration center will start working on an industrial scale after 2020, and in 2021 the Mining and Chemical Combine expects to process tens of tons of spent fuel from VVER-1000 reactors, Strana Rosatom reported, citing CEO enterprises of Peter Gavrilov.

In the nuclear fuel cycle, it is believed that due to the expanded reproduction of nuclear "fuel", the fuel base of nuclear energy will be significantly expanded, and it will also be possible to reduce the volume of radioactive waste due to the "burning" of dangerous radionuclides. Russia, according to experts, ranks first in the world in the technologies for building fast neutron reactors, which are necessary for the implementation of the CNFC.

The Federal State Unitary Enterprise "Mining and Chemical Combine" has the status of a federal nuclear organization. MCC is the key enterprise of Rosatom in creating technological complex closed nuclear fuel cycle based on innovative technologies new generation. For the first time in the world, the Mining and Chemical Combine concentrates three high-tech processing units at once - the storage of spent nuclear fuel from nuclear power plant reactors, its processing and the production of new nuclear MOX fuel for fast neutron reactors.