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One of the main properties which guarantee safety of the storage, shipment, and burial of solidified radioactive wastes is their water resistance, which is determined mainly by the quantity of radionuclides which are transferred into a water medium when the water comes into contact with the compound. From this standpoint, the element cesium which forms easily soluble compounds is most dangerous. The content of this element in some types of wastes, for example, in salt ~oncentrates from nuclear power plants, is equal to 90% and higher of the total radioactivity. Simple cementing of liquid radioactive wastes containing cesium compounds does not fix the cesium reliably, and cesium can be almost completely leached out of small samples (8 cm 3 cubes) within several days on contact with water.The leaching of cesium from cement compounds can be decreased by two methods: preprocessing of the wastes in order to form insoluble cesium ferrocyanide or using sorption additives. The latter variant is cheaper and technologically simpler. Different types of clays, especially bentonites, have good sorption properties with respect to cesium [1][2][3][4]. However, introducing clay into a cement compound and at the same time improving its water resistance can change other properties of the compound. One such property, determining the safety of the handling of the wastes, especially at the stages of reloading and shipment, is mechanical strength. Another important parameter of cementing, which can be affected by the presence of clay in the mixture, is the plasticity of the cement test compound, characterized by the degree of cracking. Lower plasticity can lead to difficulties in preparing a homogeneous mixture and on loading it from the mixer into containers.Our objective in the present paper is to examine these questions. The main investigations were performed for salt concentrates from a nuclear power plant with a VVI~R reactor (Novovoronezh nuclear power plant) and a RBMK reactor (Leningrad and Kursk nuclear power plants). The water resistance was determined by the standard procedure (GOST 29114-91. Radioactive wastes. Method for measuring the chemical stability of solidified radioactive wastes by means of prolonged leaching). A comparative assessment of different materials was made according to the rate of leaching of radionuclides and the amount of radioactivity which has transferred into the water phase over a definite time interval.The results of tests on samples obtained by cementing salt concentrates from a nuclear power plant with a RBMK reactor are presented in Table 1. In these and subsequent tests Portland cement and slag Portland cement were used as binders and bentonite clay was used as a sorption additive. Adding to the binder bentonite clay in amounts of 5-10% by mass decreases the leaching of the radionuclides by a factor of 10, irrespective of the type of binder and the degree of filling of the compound within the limits investigated.The water resistance of cement compounds with salt concentrates from a nuclear pow...

Cementing, i.e., incorporating liquid radioactive wastes into inorganic binders (Portland cement, Portland blastfurnace cement, metallurgical slags), is the simplest and cheapest method of conditioning. However, this method also has 11 drawbacks, one of which is that it is impossible to decrease or even increase the volume when cementing liquid radioactive wastes with salt content of up to 200 g/liter [1]. In this case, the salt content in the solidified wastes is 5-7%.Preliminary investigations on model compositions have shown that cementing can produce quite hard materials, containing up to 26 and 19% dry residue of salt concentrates from nuclear power plants with RBMK and VVI~R reactors, respectively. The hardness under compression of such materials is much higher than 100 kg(force)/cm 2 [2].The determination of the water resistance of cemented materials with a high salt content on 137Cs-marked samples has shown that after 14 days the rate of leaching of t37Cs from the samples containing 15% dry residue from the wastes from a nuclear power plant with a RBMK reactor is 1"10 -4 g/(cm2'day), and the leach rate from samples with 33% dry residue of the wastes from a nuclear power plant with a VVER reactor under the same conditions is I. 10 -3 g/(cm2"day). These preliminary results have shown that it is in principle possible to obtain cement materials with a higher salt content than previously thought [I].To confirm the possibility of obtaining cement compounds with a high salt content, investigations were performed with real concentrates. For this, salt concentrates from nuclear power plants with a VVI~R (Novovoronezh, Kalinin, and Zaporozh'e) and RBMK (Kursk) reactors and wastes from the Scientific and Industrial Association "Radon," a large fraction of which consist of distillation residues of liquid wastes from scientific-research institutes, were performed.As one can see from the data in Table I, the mass of the dry residue from the RBMK nuclear power plant (Kursk) and the Scientific and Industrial Association "Radon" consist mainly of sodium nitrate and the wastes from the VVER nuclear power plant contain, together with sodium nitrate, a large quantity of sodium borates, whose content in the wastes from different stations is different.To ensure a high degree of filling of the cement materials, the salt content of the liquid radioactive wastes must be increased by additionally concentrating them. This is achieved by obtaining high-salt concentrates on the UGU-500 deepevaporation plant (Novovoronezh and Zaporozh'e nuclear power plants) or evaporation of real solutions under laboratory

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During the operation of a single power-generating unit of a nuclear power plant with a VVt~R reactor, approximately 300 m3/yr of liquid radioactive wastes with a salt content of 200 g/liter are produced in the course of the technological purification of contaminated waters. In accordance with the concept adopted, the wastes must be solidified with production of bitumen or cement compounds. Their volume, however, decreases by only a factor of 2. In this connection it appears promising to reprocess high-salt-content solutions directly at the nuclear power plant and to extract valuable components from them.To this end, we assessed the use of electrodialysis process for separating acids and alkalides from the regenerates of the water purification system (WPS). This method has the advantage that the apparatus is relatively cheap to build and automate and acids and alkalides can be obtained for reuse, and the volume of the wastes is decreased in the process.The investigations were performed on two types of salt systems: boron-containing alkali regenerates of the WPS-2 setup and nitric-acid regenerates of the WPS-5 setup. The objective of the reprocessing of regenerates of the first type was to extract boric acid and potassium hydroxide, and the objective for the second type of regenerates was to obtain nitric acid and sodium hydroxide. Figure 1 shows the sorption-electrodialysis scheme of the apparatus for reprocessing boron-containing alkali regenerates. This scheme is assembled at the Novovoronezh nuclear power plant. The initial solutions to be reprocessed consisted of desorbate and washing waters from the ion-exchange filters of the WPS-2 setup. A I~DU-1-400 commercial electrodialysis apparatus with a total area of -130 m 2, equipped with bipolar and catonite membranes, was used for reprocessing the regenerates on the apparatus. The characteristics of the heterogeneous ionite membranes, employed in the electrodialysis unit, and also the MB-2 membranes for comparison are presented in Table 1. It follows from the data in Table 1 that the electrical resistance of the MB-3I bipolar membranes is lower than that of the MB-2 membranes. This is explained by the presence of ionogenic phosphoric-acid groups, which are strong catalysts, which increase the rate of constant of the transfer of H + and OH-in the bipolar membrane by a factor of 107, in the MB-3I material [1]. In this connection, the MB-3I membranes could be expected to yield boric acid and potassium hydroxide with less consumption of electricity than in [2].The potassium ions are transferred from the desorbate through the cationite membranes into the washing waters until a pH of 5.2-5.8 is obtained in the solution being desalinized (desorbate). This pH corresponds to a K + content of 30-70 mg/liter in the boric acid solution. The process of transferring potassium ions and the production of boric acid was conducted in a regime with repeated circulation of the solutions through the corresponding chambers of the apparatus.The results of the operation of the EDU-1-400...