WO2020040668A1

WO2020040668A1 - Method for reprocessing oxide nuclear fuel

Info

Publication number: WO2020040668A1
Application number: PCT/RU2019/000717
Authority: WO
Inventors: Юрий Павлович ЗАЙКОВ; Владимир Юрьевич ШИШКИН; Вадим Анатольевич КОВРОВ; Альберт Рафаэльевич МУЛЛАБАЕВ; Андрей Викторович СУЗДАЛЬЦЕВ
Original assignee: Акционерное общество "Прорыв"
Priority date: 2018-08-22
Filing date: 2019-10-08
Publication date: 2020-02-27
Also published as: RU2700934C1

Abstract

The invention relates to the field of nuclear power and can be used predominantly in a closed nuclear fuel cycle. The present method comprises reducing components of oxide nuclear fuel in the electrolysis of a lithium chloride melt with the addition of not less than 1 wt% of lithium oxide, by using an oxygen-generating anode made of NiO or TiO₂ doped with lithium oxide in an amount of at least 1 wt%, wherein the electrolysis is performed at an anode current density not exceeding 0.3 A/cm² and a temperature not exceeding 700°C. The method makes it possible to simultaneously shorten the holding time of the fuel before reprocessing and to carry out a stable and lengthy reprocessing process so as to obtain pure metals that are nuclear fuel components.

Description

A method of processing oxide nuclear fuel

The invention relates to nuclear energy, in particular, to methods for processing oxide nuclear fuel, and can be used mainly in a closed nuclear fuel cycle (NFC).

Currently, most nuclear reactors use metal or oxide fuels. To ensure operational safety, fuel is discharged from the reactor when its degree of production is not more than 15%. Basically, this need is caused by a change in the initial operational characteristics of the shells of the fuel elements, the accumulation of fission products and the difficulty of managing chain reactions with constant changes in conditions. Spent as well as rejected nuclear fuel is sent for storage, after which it must be buried or reprocessed.

Current methods for processing aged oxide nuclear fuel include dissolving oxides in nitric acid solutions to form a medium suitable for subsequent extraction of nuclear fuel components, including by electrolysis [1]. The methods are quite well established and are carried out at relatively low temperatures. However, such significant shortcomings as the long-term (3-7 years) fuel exposure before reprocessing and the formation of a large amount of radioactive water during reprocessing make the method unsafe, inefficient and unsuitable for the development of modern nuclear technologies aimed at creating an energy-efficient and resource-saving closed nuclear fuel cycle. One of the main provisions of the NFC is the timely reprocessing and regeneration of spent nuclear fuel.

The disadvantages of the known methods of hydrochemical processing of oxide nuclear fuel, including the need for a long exposure of fuel before reprocessing, can be eliminated by using molten salts, which are more stable when they interact with radioactive materials.

So, there is a known method of processing metal and oxide nuclear fuel, including the electrolysis of a melt based on lithium chloride using anodes containing U, Zr and Pu, which are components of the processed fuel at a temperature of 500 ° C and anode current density of up to 0.5 A / cm ² [ 2-4]. A significant disadvantage of this method is the rather rapid accumulation of a solid mixture containing lithium chloride and oxides in the pores and on the surface of the anodes. This leads to passivation of the anodes, instability of the parameters, a high probability of release of chlorine and poisonous gases at the anode, and a quick stop of electrolysis. In this case, only partial processing of the fuel is achieved to obtain a medium suitable for the subsequent extraction of the components of nuclear fuel. Complete processing of oxide nuclear fuel in this way is possible only with repeated repetition of electrolysis, as well as operations to separate the components of the anode from the solid mixture.

The problems of passivation of the anodes, as well as the release of chlorine and poisonous gases can be partially eliminated by the electrolysis of melts based on lithium chloride using oxygen-producing anodes.

To this end, methods have been developed for processing oxide nuclear fuel, including the electrolysis of a melt based on lithium chloride, in which the processes of lithium electrowinning at the cathode occur, secondary reduction of the oxide components of nuclear fuel by lithium with the formation of metals and lithium oxide, and the evolution of oxygen from lithium oxide to oxygen excretory anode [5-6]. The methods are carried out in the temperature range 450-650 ° C at an anode current density of up to 0.2 A / cm ² , and metals from the number Pt, Au, Ag, Pd, Rh, Ru, Ni, Ni coated with Ag, or a mixture of oxides are used as anodes Bao _. Зз $ go _.b 7МпОз, mainly Pt. As a prototype of the claimed method can be adopted a method of processing oxide nuclear fuel [7]. This method involves the electrolysis of a melt based on lithium chloride with the addition of May 1.0. % lithium oxide, while electrolysis is carried out at an anode current density of 0.3 A / cm ² and a temperature of not higher than 700 ° C, and Pt, Rh, Pd, Ir, Au are used as an anode.

The disadvantages of this method are the use of relatively expensive anodes, which are prone to anodic dissolution and to interact with the oxide components of the melt and the processed oxide nuclear fuel. This leads to instability of the electrolysis parameters, the need to stop the electrolysis process to replace the anode. In addition, dissolution of the anode leads to contamination of the reduced metals with the components of the anodes.

Thus, from the prior art it follows that implemented on an industrial scale methods can stably process oxide nuclear fuel, aged for 3-7 years in special storage facilities, while the developed known methods [2-7] can reduce the exposure time of oxide nuclear fuel, but do not allow for stable and complete processing of it.

The objective of the present invention is to develop a method for processing spent and rejected oxide nuclear fuel, which can simultaneously reduce the fuel exposure before reprocessing and carry out a stable and lengthy reprocessing to produce pure metals - components of nuclear fuel.

For this, a method for processing oxide nuclear fuel is proposed, which, like the prototype method, includes the recovery of components of oxide nuclear fuel during the electrolysis of molten lithium chloride with the addition of lithium oxide in an amount of at least May 1. % using an oxygen-releasing anode, while electrolysis is carried out at the anode current density not higher than 0.3 A / cm ² and temperature not higher than 700 ° C. The claimed method is characterized in that during electrolysis, anodes of NiO or TiO _{2 are used} , doped with lithium oxide Li ₂ 0 in an amount of not less than 1 wt.%.

In contrast to the prototype method, in which the electrolysis of molten lithium chloride with the addition of lithium oxide in an amount of at least May 1. % at an anode current density of not higher than 0.3 A / cm ² and a temperature of not higher than 700 ° C using an oxygen-releasing anode from Pt, Rh, Pd, Ir, Au, in the claimed method, NiO anodes are used as an oxygen-releasing anode or TU ₂ doped with lithium oxide in an amount of not less than 1 wt.%.

In this case, lithium is released during electrolysis at the cathode, which reacts with oxide components to form metals and oxygen-containing ions. Recent discharged at the anode of the NiO-Li ₂ 0 or Ti0 ₂ -Li ₂ 0 to molecular oxygen.

The use of anodes of NiO or TU ₂ doped with lithium oxide in an amount of at least May 1. %, leads to the fact that during the electrolysis of the molten lithium chloride with additives of at least May 1. % Of lithium oxide, an anode current density not higher than 0.3 A / cm ² and a temperature of not higher than 700 ° C at the anodes of the NiO-Li ₂ 0 or Ti0 ₂ -Li ₂ 0 occurs the formation of new compounds with lithium oxide as in the prior art. Moreover, even with an increase in the anode current density above 0.3 A / cm ^2, such anodes are not prone to anodic dissolution; the limitation of the anode density to 0.3 A / cm ^{2 is} caused only by difficulties in the melt. Additive to the melt on May 1. % lithium oxide is sufficient for oxygen evolution at the anodes without significant diffusion difficulties. This is achieved with a higher content, however, an increase in the content of lithium oxide in the melt more than May 3. % is not recommended because negatively affects the removal of oxygen-containing ions from the cathode space and the completeness of recovery of the components of oxide nuclear fuel. Because oxide doped with lithium anodes NiO-Li ₂ 0 or Ti0 ₂ -Li ₂ 0, with the electrolysis conditions necessary for the reduction of lithium at the cathode and the processing of oxide nuclear fuel are not subjected to anodic or chemical oxidation, the resource of use of the anodes is significantly increased, which leads to the possibility of complete processing of the components of oxide nuclear fuel without stopping the electrolysis. The fact that the electrolysis of lithium chloride melt with the addition of lithium oxide is recommended to be carried out at a temperature not exceeding 700 ° C is due to a sharp increase in the vapor pressure of the components of the melt.

Doping of NiO and TIG with lithium oxide L12O in an amount of at least May 1. % allows, firstly, to reduce the electrical resistance of these oxide mixtures, and secondly, eliminates the occurrence of unsteady side processes of uneven incorporation of lithium oxide in the crystal lattices of NiO and T1O2. Both factors are essential for the stable operation of the anodes of NiO-L ^ O and Tyb-L ^ O during electrolysis. It should be noted that the increase in lithium oxide in the anodes of NiO and T1O2 over May 2. % is inefficient, and can lead to destruction of the anodes.

Thus, the claimed method allows for the complete processing of oxide nuclear fuel without stopping the electrolysis and the need to replace the anode.

The technical result achieved by the claimed method is to increase the resource use of anodes in the complete processing of oxide nuclear fuel.

The invention is illustrated in the drawings, where FIG. 1 shows photographs of the anode of lithium oxide doped NiO-L O before and after electrolysis of the melt; in FIG. 2 - reduced uranium; in FIG. 3 - x-ray of reduced uranium.

The experimental testing of the method was carried out using the example of processing of uranium oxide UO2, which is one of the main components of oxide nuclear fuel. For this, a laboratory electrolyzer was constructed, which is a crucible of MgO oxide, in which was loaded with pre-prepared reagents: lithium chloride weighing 120.2 g and lithium oxide weighing 1.22 g (May 1.0.%). The reagent crucible was placed in a quartz tube, the upper part of which was closed with a special fluoroplastic cap with gas ducts and holes for electrodes and thermocouples. The operations of preparing materials, manufacturing electrodes, and mounting them in a fluoroplastic lid were performed in air, after which an electrolyzer was mounted in a “dry box”, placing a reagent crucible at the bottom of a quartz tube and closing the tube with a lid with electrodes. A cylinder weighing 25.6 g from NiO-L O doped on May 1 was used as the anode. % lithium oxide, as a cathode - a molybdenum wire, at the end of which a molybdenum container with UO2 oxide weighing 12.7 g was suspended. The process parameters were controlled using a (lithium) reference electrode.

The laboratory electrolyzer in a closed quartz tube was placed in a resistance furnace and heated to a temperature of 625 ° C. After melting the mixture of lithium chloride and lithium oxide, the electrodes were immersed in the melt and the molten mixture was electrolyzed at an anode current density of 0.3 A / cm ² and a cathodic current density of at least 0.5 A / cm ² . During electrolysis, the anode overvoltage was recorded and the end of the process was determined by its sharp increase. During a long (12 hours) electrolysis, the anode overvoltage was 0.35 ± 0.04 V, and the voltage between the anode and cathode was 3.7 ± 0.06 V, which indicates the stability of the electrolysis parameters.

At the end of electrolysis, the electrodes were removed from the melt and the electrolyzer with a quartz tube was cooled to room temperature. After cooling, the anode and the obtained uranium powder were washed from the melt, weighed, and analyzed. The mass of the anode from NiO-LbO did not change, and the mass of the obtained uranium was 11.2 g, which corresponds to the complete reduction of UO2 oxide. In this case, impurities of Ni and other elements in uranium were not detected. A similar result was obtained during the processing of 1yr oxide by electrolysis of a lithium chloride melt with the addition of May 2.23. % LhO and using an anode Tyb-bO. This is due to the similarity of the main regularities of the anode process at the anodes of lithium oxide doped NiO-LhO and Tyr-LiO in the electrolysis of molten lithium chloride with the addition of lithium oxide.

In order to optimize the properties of the melt and its electrolysis parameters, alkali and / or alkaline-earth metal chlorides and / or oxides can be introduced into the melt based on lithium chloride with the addition of lithium oxide, which do not affect the overall recovery mechanism of the components of oxide nuclear fuel.

In most experiments on the processing of uranium oxide using anodes of doped with lithium oxide NiO-LbO or T1O2-L12O at an anode current density of up to 0.3 A / cm ² and a temperature of 400 to 700 ° C in a lithium chloride melt with the addition of lithium oxide on May 1. %, as well as in lithium chloride melts with the addition of chlorides and / or oxides of alkali and / or alkaline earth metals, a complete reduction of uranium oxide was achieved with stable electrolysis and the absence of anode consumption.

Thus, the claimed method of processing oxide nuclear fuel allows both to reduce the exposure of fuel before processing and to carry out a stable and lengthy processing process to produce pure metals - components of nuclear fuel.

Sources of information:

TOlander D. Theoretical foundations of fuel elements of nuclear reactors, M., 1982);

2. Journal of Nuclear Materials, 2011, Vol. 414, pp. 194-199;

3. Journal of Nuclear Materials, 2014, Vol. 452, pp. 517-525;

Claims

Claim

A method of processing oxide nuclear fuel, including the recovery of components of oxide nuclear fuel during the electrolysis of molten lithium chloride with the addition of lithium oxide in an amount of at least May 1. % using an oxygen-emitting anode, while the electrolysis is carried out at an anode current density of not higher than 0.3 A / cm ² and a temperature of not higher than 700 ° C, characterized in that the electrodes use anodes of NiO or Tyr doped with lithium oxide in an amount not less than 1 wt.%.