RU2366012C2 - Method of irradiated nuclear fuel treatment - Google Patents

Abstract

FIELD: chemistry.

SUBSTANCE: invention relates to field of radio-chemical technology, and can be applied for treatment of irradiated nuclear fuel. Method of INF treatment includes dissolving of fuel, extraction of uranium nitrate and actinides with neutral phosphorus-organic compounds, dissolved in slightly boiling solvent and re-extraction. Dissolving and extraction are carried out in two stages: first dissolving by processing of irradiated nuclear fuel with nitrogen dioxide with addition of stoichiometric quantity of water in order to obtain hydratated nitrate of uranyl and actinides, then actinides are extracted from obtained as result of first stage melt of actinides nitrates, additional extract purification on inorganic sorbet is carried out, and then precipitation re-extraction of actinides by transferring nitrates of uranium and actinides in oxalates or carbonates. In separate operation process of dissolving (processing with nitrogen dioxide) is separated, which results in excluding potentially dangerous contact of organic extragent -TBP with nitric acid in presence of heated by decay irradiated fuel. On stage of extraction time of said components contact is reduced, which also increases safety of process carrying out.

EFFECT: application of precipitation re-extraction allows to carry out the process continuously, because regeneration of extragent occurs under similar conditions as extraction.

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Description

The invention relates to the field of radiochemical technology and can be used for the processing of irradiated nuclear fuel.

Known methods of processing irradiated nuclear fuel using water methods - for example, Purex - process and its modifications (V.M. Vdovenko, Modern Radiochemistry, M .: Atomizdat, 1969, p. 459-468), where neutral neutral mixtures are used organophosphorus compounds (usually tributyl phosphate - TBP) in various diluents. The disadvantages of the method is that in the processing of irradiated fuel, a large number of aqueous solutions are formed. So, for the processing of one ton of irradiated nuclear fuel (SNF), at least 2 m ^{3 of a} solution of nitric acid is necessary and further processing leads to the formation of another 4-6 m ^{3 of} aqueous radioactive waste. These radioactive waste in accordance with modern requirements must be converted to solid form. Recycling waste into a form suitable for long-term storage complicates and increases the cost of the overall recycling process.

Also known are “dry” methods for processing irradiated nuclear fuel, such as gas fluoride and pyrometallurgical ones. (V.M. Vdovenko, Modern Radiochemistry, Moscow: Atomizdat, 1969, p. 488-482). Their disadvantages include carrying out processes at high temperatures.

Known methods for supercritical extraction of metal complexes with carbon dioxide in the presence of complexones, for example tributyl phosphate (Y. Lin, RD Brauer, KELaintz, CMWai, Supercritical Fluid Extraction of Lanthanides and Actinides from Solid Materials with a Fluorinated β-Diketones, Anal. Chem ., 1993, vol. 65, p.2549-2551) or using β-diketones / Y.Lin, CMWai, FMJean, RD Brauer, Supercritical Fluid Extraction of Thorium and Uranium Ions from Solid and Liquid Materials with Fluorinated β-Diketones and Tributyl Phosphate, Environ. Sci. Technol., 1994, vol. 28, No. 6, p. 1190-1193. C. M. Wai, N. G. Smart, C. Phelps, Extraction metals directly from oxides. US Pat. No. 5606724 A / The disadvantage of these methods is the use of a large excess of trialkyl phosphates to ensure complete metal recovery.

Closest to the claimed method is a method of processing irradiated nuclear fuel using a solution of TBP saturated with nitric acid in supercritical carbon dioxide [T.Shimada, S. Oguno, N. Ishihara, Y. Kosaka, Y. Mori A study on the technique of spent fuel reprocessing with supercritical fluid direct extraction method (Suprer DIREX method) // J. Of Nucl.Science and Techn., Suppl.3, p.757-760 (2002)] - prototype.

According to the prototype method, the irradiated fuel is treated with a solution of TBP saturated with nitric acid in supercritical carbon dioxide. In this case, uranium dioxide dissolves, reacts with nitric acid, and uranyl nitrate is formed, which, in turn, is extracted with tributyl phosphate in a medium of carbon dioxide. Thus, the process of dissolution and extraction is carried out simultaneously. Metals are re-extracted from the organic phase according to the existing water scheme.

The disadvantage of the prototype is the use of TBP solution containing high concentrations of nitric acid. Nitric acid, especially in the organic phase, is a strong oxidizing agent, and in contact with irradiated fuel, i.e. at elevated temperatures, there is always a danger of local overheating, which initiates the oxidation of TBP, which can lead to decomposition of the extractant and even to explosion. In addition, large volumes of supercritical carbon dioxide are needed to dissolve the entire amount of uranium, which, in turn, leads to an increase in the size of the apparatus and, accordingly, to a rise in the cost of the process.

The task of the invention is to create a more compact and safer technology compared to the prototype.

To solve this problem, a method for reprocessing spent nuclear fuel is proposed, which includes dissolving fuel, extracting uranium nitrates and actinides with neutral organophosphorus compounds dissolved in a low-boiling solvent and stripping. The method is characterized in that the dissolution and extraction are carried out in two stages. At the first stage, the fuel is dissolved by treating the irradiated nuclear fuel with nitrogen dioxide with the addition of a stoichiometric amount of water to produce hydrated uranyl nitrates and actinides. Then actinides are extracted from the actinide nitrate obtained as a result of the first stage of melt, an additional purification of the extract is carried out on an inorganic sorbent, and then precipitating reextraction of actinides is carried out by converting uranium nitrates and actinides into oxalates or carbonates.

The drawing shows a diagram of the process.

At the first stage, the irradiated fuel is treated with nitrogen dioxide with the addition of a stoichiometric (based on the production of uranyl nitrate hexahydrate) amount of water. In this process, actinides are converted to nitrates, and a significant part of the fission products remains in the sediment. After removing the excess gaseous nitrogen oxides and radioactive gases, a melting of actinide nitrates is obtained with a small concentration of excess nitric acid (not more than 0.1 mol of acid per 1 mol of uranium).

At the second stage, during continuous extraction with reverse TBP dissolved in supercritical carbon dioxide or a low boiling solvent (for example, a fluorocarbon diluent) and continuous extraction of the extract for 2 hours, up to 99.9% of uranium, plutonium, and neptunium are extracted. After the extraction is completed, the SNF residue containing fission products is washed for 20 minutes with a pure solvent, which is also sent for reextraction of uranium, plutonium and neptunium. In contrast to the prototype method, extraction takes place in the absence of nitric acid and without nitrogen oxides, which increases the safety of the process — the possibility of local overheating and oxidation is practically excluded. The process is much faster, since extraction takes place from the liquid phase. Therefore, the volume of the apparatus is reduced, which dramatically reduces costs.

As a result, two products are formed:

1. A solution of solvates of nitrates of uranium, plutonium and neptunium with TBP in a solvent that is fed to reextraction.

2. The remaining dry residue, consisting of nitrates and oxides of PD.

At the end of the extraction and washing with a pure solvent, the latter is pumped out of the apparatus by the compressor into the receiver (collector) to a residual pressure of 1 atm. After that, the apparatus is depressurized and the undissolved residue containing fission products (PD) is sent for curing. The conversion of actinide oxides to nitrates is known, but the fact that most of the fission products remain undissolved is an unobvious, but very useful fact. The insoluble residue containing a significant portion of the active nuclides can be separated in the first stage, which drastically reduces the radiation load on the extractant.

At the third stage, the solution of uranyl, plutonium, and neptunium solvates with TBP obtained as a result of extraction is fed by a compressor to a sorption column for additional purification of the extract from Cs, Sr, REE, and TPE. As the sorbent can be used, for example, porous zirconium oxide in a mixture with silica gel. The spent absorber is sent for curing together with PD from the first stage. Inorganic sorbents for the purification of sub- and supercritical media have not previously been used.

In a fourth step, the purified extract is sent to a solid phase stripping column of Pu, Np and U. The stripping is carried out on oxalic acid or ammonium carbonate with stirring. The resulting oxalate or carbonate containing up to 99.9% of the initial uranium, plutonium and neptunium is washed with a solvent. This product is sent for heat treatment to obtain a solid solution of dioxides of uranium, plutonium and neptunium for subsequent storage.

Thus, as a result of the proposed method, the irradiated fuel is divided into a solid mixture of actinide oxides and solid waste - fission products. Actinide oxides contain less than 0.1% of fission products, and solid waste - less than 0.1% of actinides (uranium, plutonium, neptunium).

Compared with the prototype method, the dissolution process (treatment with nitrogen dioxide) was isolated in a separate operation, which eliminates the potentially dangerous contact of the organic extractant - TBP with nitric acid in the presence of irradiated fuel heated by the decay. At the extraction stage, the contact time of the same components is reduced, which further increases the safety of the process. The use of precipitation reextraction allows the process to be carried out continuously, because regeneration of the extractant takes place under the same conditions as extraction.

The following examples illustrate the applicability of this method.

Example 1 (prototype)

A VVER-1000 fuel simulator oxidized to U ₃ O ₈ was treated with a solution of TBP-HNO ₃ in supercritical carbon dioxide at 120 atm and 40 ° C. At a molar ratio of U: TBP = 1: 6, in three hours 14% of uranium was extracted with the purification coefficients from strontium 620, from neodymium 140, from cerium 2000.

Example 2 (prototype)

Fragments of tablets of irradiated VVER fuel after three years of exposure were treated with a solution of TBP-HNO ₃ in supercritical carbon dioxide at 300 atm and 60 ° C. At a molar ratio of U: TBP = 1: 18, 98% of uranium was extracted in three hours with a cesium removal coefficient of 320, europium -140, and americium 180.

Example 3

After three years of exposure, the VVER-1000 simulated irradiated fuel was treated with calculated quantities of liquid nitrogen dioxide and water taken in a molar ratio of NO ₂ / UO ₂ = 8/1 and H ₂ O / UO ₂ = 6/1 at a temperature of 105 ° C for 4 hours.

After removal of the vapor-gas phase and cooling to a temperature of 25 ° С, extraction was carried out with a solution of 12% TBP in liquid carbon dioxide, with a molar ratio U: TBP = 1: 6. In three hours, the recovery of uranium and actinides was 98%. The undissolved residue, containing 95% of the activity of the fission products, is sent for curing.

An extract containing uranium and actinides was passed through an inorganic sorbent. In this case, the additional purification coefficient amounted to 800 cesium, 430 europium and americium, and the total gamma activity purification coefficient was more than 8000.

An extract containing uranium and actinides was passed through a layer of ammonium carbonate to reextract and regenerate the extractant. After regeneration, the extractant was returned to the extraction stage.

The resulting actinide carbonate is thermally decomposed by a known method to actinide oxides.

Example 4

Fragments of a tablet of irradiated nuclear fuel RBMK after three years of exposure were treated with calculated quantities of liquid nitrogen dioxide and water taken in a molar ratio of NO ₂ / UO ₂ = 8/1 and H ₂ O / UO ₂ = 6/1 at a temperature of 105 ° C for 4 hours.

After removal of the vapor-gas phase and cooling, extraction was carried out with a solution of 30% TBP in supercritical carbon dioxide at 300 atm and 60 ° C. The extraction of uranium and actinides was 99% with a molar ratio of U: TBP = 1: 6.

An extract containing uranium and actinides was passed through a layer of a mixture of zirconium oxide and silica gel for additional purification from suspensions. After passing through a layer of sorbent in the extract, less than 0.05% of the initial activity of cesium (purification coefficient 2000) and 0.03% of the initial activity of americium (purification coefficient 3000) remained. After purification, the extract, dissolved in supercritical carbon dioxide, was passed through a layer of oxalic acid to reextract and regenerate the extractant. In the process of re-extraction, 99.9% of extracted uranium and 99.5% of extracted plutonium and neptunium were recovered. After regeneration, the extractant was returned to the extraction stage.

The resulting oxalate of uranium and actinides is thermally decomposed by a known method to actinide oxides.

Example 5

Fragments of a tablet of irradiated nuclear fuel RBMK after three years of exposure were treated with calculated quantities of liquid nitrogen dioxide and water taken in a molar ratio of NO ₂ / UO ₂ = 8/1 and H ₂ O / UO ₂ = 6/1 at a temperature of 105 ° C for 4 hours.

After removal of the vapor-gas phase and cooling, extraction was carried out with a solution of 30% TBP in liquid Freon-125 at 30 atm and 60 ° C. The recovery of uranium and neptunium and plutonium was 99.9%.

An extract containing uranium and actinides was passed through a layer of a mixture of zirconium oxide and silica gel for additional purification from suspensions. After passing through a layer of sorbent in the extract dissolved in liquid Freon-125, less than 0.03% of the initial activity of cesium (purification coefficient 3300), and 0.03% of the initial activity of americium (purification coefficient 3300) and 0.02% of the initial activity remained europium (purification coefficient over 3000). After purification, the extract, dissolved in liquid Freon-125, was passed through a layer of oxalic acid to reextract and regenerate the extractant. In the process of re-extraction, 99.5% of extracted uranium and 99.7% of extracted plutonium and neptunium were recovered. After regeneration, the extractant was returned to the extraction stage.

The resulting oxalate of uranium and actinides is thermally decomposed by a known method to actinide oxides.

The above examples show that actinides can be extracted and separated from fission products with coefficients of at least 1000 by the above method. Compared with the prototype, the method accelerates and makes safer the dissolution and extraction process, leads to a reduction in the total volume of the apparatus at the same productivity as the prototype. This gives a significant gain in economic efficiency, because in radiochemical processes, the main costs are associated with the volume of capital construction.

Claims (4)

1. A method of processing irradiated nuclear fuel, including dissolving fuel, extracting uranium nitrates and actinides with neutral organophosphorus compounds dissolved in a low boiling solvent and stripping, characterized in that the dissolution and extraction are carried out in two stages: first, dissolution by treatment of the irradiated nuclear fuel with nitrogen dioxide with by adding a stoichiometric amount of water to obtain uranyl nitrate hydrates, nitrates of other actinides, then actinides are extracted from the obtained as a result of the first stage of the actinide nitrate melt, additional purification of the extract is carried out on an inorganic sorbent, and then actinides are re-extracted by converting uranium nitrates and actinides to oxalates or carbonates. 2. The method according to claim 1, characterized in that neutral organic phosphorus compounds, for example tributyl phosphate, are used as extractant. 3. The method according to claim 1, characterized in that carbon dioxide or organofluorine compounds, for example freon-125, are used as solvent for the extractant. 4. The method according to claim 1, characterized in that the porous zirconium oxide mixed with silica gel is used as an inorganic sorbent.