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Radionuclides are removed from low-salt ( < 1 g/liter) low-level ( < 10 -5 Ci/liter, or 370 kBq/liter) radioactive liquid wastes by a technology that includes coagulation and subsequent desalinization on ion-exchange resins in H and OH form [1][2][3]. The operations of deposition and periodic regeneration of the ionites with acid and alkali increase the volume of secondary wastes, which correspondingly increases the cost of solidification and subsequent storage and burial of the wastes [4]. One way to decrease the amount of secondary wastes and lower the cost of the purification process is to use ionites in a salt form [5]. When a cationite is transferred into Na form, the univalent cations, which constitute up to 50% of the total salt content, will not be retained by the ionites.The transfer of an anionite into a salt form will make it possible to eliminate expensive alkali.To determine the possibility of using ionites in a salt form for removing radionuclides from liquid wastes, an apparatus with three sorption columns, filled with the cationites KU-2 ;ind KB-51 in the Na form and AV-17 in the NO 3 form, was developed at the Moscow reprocessing plant. Solutions were fed into the column (load volume 5 liters, feed rate of solution four-column volumes V c per 1 h) after the standard coagulation with iron hydroxide (Tables 1 and 2). Data on purification by the H-OH ionite scheme, are presented for comparison.

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66.064At present, there are several open stores for liquid radioactive wastes in the region of the Mayak organization. The stores for low-activity wastes are certain ponds in the valley of the R. Tech. However, the main hazard is represented by liquid wastes stored in Lake Karachai (pool No. 9), to which medium-activity solutions have been discharged since 1951. The water leaking from pool No. 9 causes a danger of radioactive pollution in water-supply sources and represents a hazard on passage to the open river network [1]. This hazard can be avoided by halting the motion of the polluted water towards the pools or trapping it and purifying it from radionuclides and dissolved salts until it meets discharge standards.These infiltrating waters from pool No. 9 constitute a multicomponent system having total salt contents up to 80 g/liter, 90% of which is accounted for by sodium nitrate. The radioisotopes are mainly 9~ 137Cs, and 6~ The total radioactivity is up to 1.5-105 Bq/liter. There are difficulties in using standard treatment methods [2, 3] for liquid radioactive wastes in these waters because they do not provide the necessary degree of purification, consume large amounts of reagents and energy, and are complicated to implement. It has been shown [4] that electromembrane methods in combination with others are effective for treating radioactive wastes produced at the nuclear fleet installations.We consider that electromembrane technology is best for treating these waters, which includes electrochemical precipitation of the hardness salts and polyvalent metals, electroosmotic concentration, and electrodialysis desalination. Experience exists in treating radioactive wastes by electromembrane methods, which suggests that one should include an electromembrane apparatus of EKhO type intended for the electrochemical precipitation of hydroxides and carbonates, which enables one to eliminate hardness salts and heavy-metal ions from the wastes by means of electrochemical processes. Figure 1 shows the block diagram of a laboratory apparatus handling 2 liter/la for treating liquid wastes whose composition is given in Table 1.The cathode chamber of the EKhO apparatus accumulates hydroxyl ions by the electrolysis of water: H20 + 2e ---H 2 + OH, and the bicarbonate ions become carbonate ones HCO 3 + OH --, CO 3-2 + H20, and sparingly soluble calcium carbonates (SPcacO 3 4.3" 10 -9) and magnesium hydroxide (SPMg(OH) 2 5.5-10 -12) are precipitated along with the carbonate and hydroxides of other heavy metals.

When operating nuclear powered vehicles and industrial nuclear fuel cycle production units, it becomes necessary to deactivate the surfaces of the work areas, equipment, and mechanisms as a preventive operation prior to repair and dismantling, in order to create safe radiation conditions while the work is carried out.The work areas and the external surfaces of the operated equipment are treated with a 1% solution of sulfonol and sodium hexametaphosphate. The internal surfaces of the equipment are washed with oxidizing--reducing solutions using potassium permanganate, nitric and oxalic acids, and complexing agents.The total salt content of the mixtures of alkali and acid solutions is 6 g/liter with a pH of 8-10, a total ~ activity of 1 • i0 -6 to 1 • 10 -8 Ci/liter (the activity level varies depending on the operating conditions of the equipment).Previous investigations [1][2] have indicated the possibility of decontamination using electrodialysis (up to 0.5 g/liter) of the low-level liquid waste with subsequent electroosmotic concentration of the secondary waste.It is well known [3] that ionic forms of radionuclides participate in the electrodialysis process in a similar way to stable saline macrocomponents of solutions. At the same time, radionuclides prone to hydrolysis and polymerization with the formation of colloids screen the ion-transmitting surfaces of the membranes and thereby reduce the decontamination efficiency. Moreover, as the concentration of the salts in the saline chambers increases the solubility of the saline components forming part of the deactivated solutions exerts a considerable influence on the operating efficiency of the electrodialyzer. This principally relates to oxalic and oxyethylidene-diphosphonic acids, sodium hexametaphosphate, etc. For example, sodium oxalate possesses a very low solubility of -33 g/liter, and this makes it impossible to obtain saline waste having a concentration which is optimal for solidification by the cementation method.Taking into account that it is proposed subsequently to concentrate the solution in the saline chambers of the electrodialyzer to a salt content of 200-250 g/liter, the solubility factor can be decisive. In view of this a precontamination stage of the deactivated solutions is required in order to extract the radionuclides in colloidal form, destroy the organic compounds, and separate the low-solubility salts. These aims are most rationally achieved by employing the electrocoagulation method which makes it possible to destroy the organic compounds and separate the precipitates, including the majority of the anions of the deactivated solutions, as a result of oxidation--reduction processes. The subsequent production operation must provide for obtaining decontaminated (salt-free) water and ensure a high content of water-soluble salts in a small volume of the solution to be solidified. However thorough demineralization of the waste (to a concentration of lower than 50 g/liter) with simultaneous waste concentration (to 200-250 g/liter) in one appar...