The article highlights the problem of reprocessing of liquid radioactive waste (LRW) from the Ukrainian nuclear power plants with VVER reactors. The main method of these LRW treatment is distillation (evaporation) and ion exchange sorption. The final products of LRW processing by distillation are distillation residues, spent sorbents and sludges, and salt melt with significant radioactivity accumulated in large volumes, which do not meet the long-term storage and disposal criteria. So, it is necessary to develop a new, more efficient LRW treatment scheme that can solve the problems of large volumes of LRW and transfer waste to a state suitable for long-term storage and disposal. One of the important and challenging problems of LRW treatment is the presence of organic components (ethylenediaminetetraacetic acid (EDTA), oxalic acid, synthetic surfactants (SPAR) in the waste, since these substances complicate removal of radionuclides from the LRW. The results of investigation of ozonolysis conditions of LRW organic components are presented on the example of a model solution with imitated solutions of disodium salt of ethylenediaminetetraacetic acid and ethane diacid. It was established, that ozonation of organic components of LRW results in decreasing of pH value, which causes reduction of their destruction efficiency. Control of the pH values at the initial level (11−12 units) by permanent alkalifying can increase the ozonolysis efficiency of LRW organic components in 30%. EDTA is better destructed by ozone than ethane diacid. Destruction of LRW organic components progresses in two stages: the first is fast and second is low. The maximum of EDTA destruction degree, in the experimental conditions, was 86%; destruction degree of ethane diacid — 51%; maximum of LRW model solution organic components (EDTA, ethane diacid, synthetic surface-active reagent) destruction degree was 67%. Due to co-precipitation and adsorption during the solution ozonizing, decrease in Mn concentration in LRW model solution reaches 94.3%, 137Cs activity decreases by 26%, and 90Sr — 15.7%, concentration of Co decreases only by 6%.
The article presents the general pattern of the combined process of oxidative decomposition of organic components of simulated nuclear power plant (NPP) drain water and sorption interaction of the imitators of main dose-forming radionuclides (Cs – radiolabel for 137Cs; stable isotopes of Co, Sr, Mn salts) on natural bentonites from the Cherkasy deposit in presence of sorption-reagent compounds — iron (II) and manganese (II) salts. Hydroxides, oxyhydroxides and oxides of Fe and Mn formed during ozonation are predominantly localized on the surface of bentonite. The chemical composition of the main elements of bentonite after drain water ozonation with the addition of iron and manganese salts remains almost the same as that of natural bentonite. The phase composition of bentonite is presented by the main rock-forming mineral montmorillonite and secondary mineral quartz. The iron-containing phases of the ozonised bentonite are Fe(II)- Fe(III) layered double hydroxides (Green Rust), goethite α-FeOOH and magnetite Fe3O4, and the manganese-containing phases are hausemannite Mn3O4, manganese oxide (II) and manganese oxyhydroxide MnO(OH)2. The iron- and manganese-containing phases deposited on the bentonite surface during ozonation are predominantly weakly crystallized or amorphized structures. At the concentration of salts of iron (50 mg/dm3) and manganese (100 mg/dm3) in the drain water, the specific surface area of bentonites with the formed layer of iron and manganese hydroxides, (oxy)hydroxides and oxides increases compared to natural bentonite (34.2 m2/g) and equals to 55 and 51 m2/g, respectively. The degree of radionuclide removal during ozonationof the simulated solution with the initial concentration of cations (Fe2+ — 5 mg/dm3; Mn2+ —10 mg/dm3; Ca2+ — 5 mg/dm3) in the presence of natural bentonite is 137Cs — 78% ± 2%, Sr2+ —97.55% ± 1%, Co2+ — 96.5% ± 1%, Mn2+ — 99.7% ± 0.5%. To preserve the efficiency of 137Cs and Co2+ radionuclide removal, the initial concentration of cations in the solution can be increased to the following values: Fe2+ — 50 mg/dm3, Mn2+ — 100 mg/dm3, Ca2+ — 50 mg/dm3, and to: Fe2+ —500 mg/dm3, Mn2+ — 1,000 mg/dm3, Ca2+ — 500 mg/dm3 for Sr2+ and Mn2+ removal.
Natural zeolites are abundant and inexpensive resources. They are crystalline hydrated aluminosilicates with a framework structure that has pores and channels occupied by water, alkali, and alkaline earth cations. Having high cation exchange capacity, acting as a molecular sieve, natural zeolites have been widely utilized in recent decades as adsorbents in separation and purification processes. Modification of natural zeolite increases its adsorption capacity of environmental pollutants, in particular, radionuclides from low-level liquid radioactive waste. The article presents results obtained from a study of the chemical composition of the structural elements and ion exchange complexes of natural, acid-modified and alkali-modified zeolites from the Sokyrnytske deposit. The main rock-forming mineral of the Sokyrnytske zeolite is clinoptilolite. The zeolite was modified by a 5.5 M HCl solution for 2 hours at 100 oC using a backflow condenser. The ratio of solid to liquid phases was 1:2. For alkaline modification, 1.4 M NaOH solution was used. The exposure time was 2.75 hours. The conditions and phase ratio were similar to those in acidity modification. By composition of the ion exchange complex, natural zeolite belongs to potassium-calcium-sodium (K > Ca > Na) clinoptilolites. In the process of acidity and alkaline modification of the natural zeolite, redistribution of the exchangeable cations is observed and the content of structural cations in the clinoptilolite lattice changes. In alkali-modified zeolite, the content of exchangeable Na and Ca cations increases, and the content of K and structural Al cations decreases. In the acid-modified zeolite, the number of exchangeable Na, Mg, Ca, K cations decreases. At the same time, the content of Fe and Al decreases and the relative amount of Si in the lattice increases. The Si/Al ratio increases in the following succession: natural zeolite → alkali-modified zeolite → acid-modified zeolite. The specific surface area of the modified zeolites increases compared to the natural ones. The largest increase is observed for the acid-modified zeolite. The textural characteristics and mineral composition of the studied samples indicate that the natural, acid-modified and alkali-modified zeolites from the Sokyrnytske deposit may be used for removal of radionuclides from low-level liquid radioactive waste.
The lack of scientifically substantiated requirements, comprehensively developed and approved in a prescribed manner, for the usage of clays as a barrier material poses risks for the safe disposal of radioactive waste in facilities at the ‘Vector’ site for the period of their operation and closure. The bentonite clay from Ukraine’s largest Cherkasy deposit of bentonite and palygorskite clays is considered the most durable as the main component of the insulating (underlying) screens of radioactive waste disposal facilities. The main properties and compositional features of the Cherkasy natural bentonite clay (Dashukovskaya site, layer II) and its variety such as alkaline earth bentonite (activated soda bentonite), which provide isolation of radioactive waste in disposal, are considered. It is shown that the Cherkasy field has good waterproofing and barrier properties, including a high sorption capacity with respect to 90Sr and 137Cs, which is one of the main characteristics that ensure the safe disposal of radioactive waste. The alkaline earth bentonite absorbs 90Sr and 137Cs more efficiently than natural bentonite does. However, 90Sr is sorbed in larger quantities than 137Cs on both types of bentonite. With increasing time of interaction with an aqueous solution, both types demonstrate a redistribution of the mobile (exchangeable) and immobile (non-replaceable) forms of radionuclides. The contribution of the stationary form that does not participate in migration processes also increases. A comprehensive analysis of the bentonite clays of the Cherkasy deposit was carried out, taking into account the significance of recoverable reserves and the potential for improving the technical and economic parameters of clays. Thus, the Cherkasy bentonite clays can be recommended as an additional anti-migration engineering barrier for ground/near-surface facilities for the disposal of radioactive waste. When choosing the type of bentonite clay for use as a barrier in a radioactive waste disposal facility, one could take into account the data published in the article, but the question of applying the bentonite clays of the Cherkasy deposit to ensure the safe disposal of radioactive waste remains to be further studied.
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