Development of a process for treatment and solidification of liquid radioactive concentrates from the Leipunskii Institute of Energy Physics

Treatment of liquid radioactive concentrates (LRCs) from the Leipunskii Institute of Energy Physics using Termoksid-35 ferrocyanide sorbent was studied. After passing LRC with a volume activity of 3.9 × 10⁸ Bq l⁻¹ through a column packed with the sorbent, the volume activity of the filtrate does not exceed 3.7 × 10⁴ Bq l⁻¹. The LRC decontaminated from the major amount of radionuclides is fed to cementation. A formulation of a cement compound with a polyfunctional additive consisting of finely dispersed cement, bentonite clay, biocidal additive, plasticizer, and defoamer was developed. For the storage of a container filter with the spent ferrocyanide sorbent, it is suggested to place it in a 1 m³ metallic container in which the cement compound with the LRC decontaminated from the major fraction of cesium acts as biological protection. The γ-radiation dose rate from the 1 m³ container filled with the cement compound, with the filter with the spent sorbent placed inside, was calculated. A technology for processing of liquid radioactive concentrates from the Leipunskii Institute of Energy Physics was suggested and substantiated.     

Magnesium Potassium Phosphate Compound for Radioactive Waste Immobilization: Phase Composition,Structure, and Physicochemical and Hydrolytic Durability

Low-temperature mineral-like magnesium potassium phosphate (MPP) compounds were synthesized in the course of immobilization of nitric acid solutions containing cesium, strontium, sodium, ammonium, lanthanum, and iron as simulated radioactive waste (RW). The phase composition and structure of the compounds and the distribution of the RW components were studied. The mechanical strength (15 ± 3 MPa), heat resistance (up to 450°С), and radiation resistance (absorbed dose 1 MGy) of the compounds were evaluated in accordance with the existing regulations. The MPP compound exhibits high hydrolytic durability: The differential leach rate of ²³⁹Pu and ¹⁵²Eu on the 28th day, measured in accordance with GOST (State Standard) R 52 126–2003, is 2.1 × 10^–6 and 1.4 × 10^–4 g cm^–2 day^–1, respectively. Introduction of wollastonite into the compound decreases the radionuclide leach rate by a factor of up to 5. The MPP compound shows promise for industrial solidification of liquid RW, including high-level highly saline multicomponent actinidecontaining waste.     

Geopolymeric Agent for Immobilization of Radioactive Ashes after Biomass Burning

Solidification of low-level radioactive wastes obtained after biomass burning was studied. Two solidification modes using Portland cement and geopolymeric binder were tested experimentally. The strength at various hardening times, compacting efficiency, and leaching rate of the resulting monolithic concretes were analyzed. The compacting efficiency in concretes prepared by two different modes is similar. At the same time, geopolymeric binder is solidified in significantly shorter period and its compression strength is several times higher, but its main advantage is chemical immobilization of strontium cations. The leaching rate under the static conditions after 28-day hardening is nearly 10^{- 6} g cm^{- 2} day^{- 1}. Thus, substitution of geopolymerization of the clay component (in general case, aluminosilicate material) for common solidification of low-level wastes using Portland cement is economically promising due to significant energy and resource saving. The geosynthesis can be easily realized as an environmentally safe process, yielding no liquid waste and involving no high-temperature stages with radioactive materials.     

Feasibility of Processing of Accumulated Liquid Radioactive Waste from BN-350 Reactor

It was found that the filtrate yielded by passing bottoms decantates through selective sorbents has the volume activity with respect to ¹³⁷Cs and ⁶⁰Co significantly greater than DK_b. A pretreatment of the bottoms before selective sorption was tested, which consisted in ozonation and subsequent filtration of the resulting precipitate. By the example of the liquid radioactive waste (LRW) decantate with the highest volume activity, it was shown that by combining the pretreatment with selective sorption it was possible to remove ¹³⁷Cs, ⁶⁰Co, and ⁵⁴Mn radionuclides from the decantate to the level under DK_b. Cementation of the slurries in the slurry/cement ratio of 0.3-0.7 was tested. It was proposed that the decantate freed from radionuclides be discharged into the Caspian Sea, and the slurry and secondary wastes yielded by decantate decontamination be included into cement compound.     

Concept of autonomous processing of liquid radioactive waste from ship nuclear power installations

The concept of autonomous processing of liquid radioactive waste from ship nuclear power installations is substantiated, and its implementation in the form of a module membrane sorption installation consisting of micro- and ultrafiltration, reverse-osmosis, and ion-exchange modules is suggested. Data on treatment of liquid radioactive waste of complex physicochemical composition using this installation are reported.     

Treatment of Locally Accumulated Liquid Radioactive Waste at the Shelter To Remove Transuranium Elements

Liquid radioactive waste (LRW) from the Shelter yielded by reactions of atmospheric precipitations, condensate, and dust-suppressing solutions with structural materials and fuel-containing masses are alkali-carbonate solutions containing fission products (¹³⁷Cs, ⁹⁰Sr), organic compounds, uranium, and transuranium elements. This waste is unsuitable for treatment by the standard procedure utilized at the Chernobyl NPP, as the content of organic compounds and transuranium elements in the Shelter LRW exceeds that acceptable for evaporators. For pretreatment of the Shelter LRW to remove organic compounds of uranium and transuranium elements, a procedure was developed and tested at the Shelter, based on the use of hydrolysis lignin as sorbent and specially designed equipment.     

Recovery of <Superscript>137</Superscript>Cs and <Superscript>90</Superscript>Sr from wastewater by sorption on finely dispersed minerals under static conditions

The natural minerals clinoptilolite and tripoli were used for sorption treatment of liquid radioactive waste (LRW) to remove ¹³⁷Cs and ⁹⁰Sr. The efficiency of sorption recovery of these radionuclides with finely dispersed mineral sorbents under static conditions was studied in relation to the sorption time, pH, size of mineral granules, sorbent amount, salt content and chemical composition of solutions, and number of successive sorption steps. The distribution coefficient of radionuclides between the sorbent and aqueous phase and the sorption capacity of sorbents for radionuclides were determined. It was found that treatment of real salt-containing LRW from the Leipunskii Institute of Energy Physics, State Scientific Center of the Russian Federation, with the natural sorbents decreased their activity by 2–3 orders of magnitude owing to recovery of ¹³⁷Cs and ⁹⁰Sr.     

Membrane–sorption technology for reprocessing liquid radioactive waste from nuclear reactors

The need for developing liquid radioactive waste (LRW) reprocessing technologies oriented on autonomous reprocessing at research and production enterprises that are equipped with nuclear reactors and/or use radionuclides in the research and production activity is substantiated. Membrane–sorption technology of LRW reprocessing is suggested as a universal solution. An installation implementing this technology in the modular version, ensuring its low-cost transportation from one facility to another, is described. The developed modular membrane–sorption installation ensures efficient treatment of radioactively contaminated waters with the decontamination factor of 10³–10⁴, irrespective of the physicochemical forms of the radionuclides (ionic, colloidal, molecular), with obtaining LRW concentrates for the subsequent cementation.     

Geoconcrete monolith as stable matrix material for immobilization of radioactive wastes

Immobilization of the intermediate-level radioactive waste (RAW) in a new type of binding agents, geoconcrete monoliths, was studied. The experiments were performed using liquid radioactive wastes of various salinities formed in the course of operation of nuclear power plants and various sorbents saturated with radionuclides, in particular, with ¹³⁷Cs. The prolonged operation life of geoconcretes is determined by the chemical and mineralogical composition of the hydrated neogenic compounds, which are analogs of water-resistant rock-forming silicate materials. A procedure for synthesis of a geoconcrete matrix with immobilization of biologically important radionuclides was developed, and the properties of geoconcrete monolith (GCM) were studied. The pilot experiments with real RAW were performed. The strength of the GCM samples exceeds by a factor of 2–4 the level prescribed by the regulations, and the leaching rate of ¹³⁷Cs is lower than the prescribed level by 2–3 orders of magnitude.     

Calculations of gamma ray incineration of 90Sr and 137Cs

The feasibility of the γ-ray incineration method, using the (γ, n) reaction, was examined for the long-lived radioactive waste ⁹⁰Sr and ¹³⁷Cs. The incineration of ⁹⁰Sr and ¹³⁷Cs as well as the buildup of long-lived radioactive by-products were calculated by changing the γ-ray flux as a parameter. The effect of alternative irradiation modes was also studied.     

碱激发和复合激发下建筑垃圾砖粉活性研究

为提高建筑垃圾废砖再生利用率,采用砂浆力学测试手段研究了建筑垃圾砖粉活性和碱激发、复合激发对建筑垃圾砖粉活性的影响,并借助扫描电镜和热分析方法对建筑垃圾砖粉、建筑垃圾复合粉体材料的颗粒形貌、水化产物组成等进行了研究。结果表明:建筑垃圾砖粉活性较小,当掺量大于20%时,砂浆强度随其掺量的增加直线下降;不同碱激发剂对建筑垃圾砖粉有不同的激发效果,Ca(OH)2激发效果最好,NaOH次之,Na2SiO3·9H2O激发效果最差;复合形成的建筑垃圾复合粉体材料具有较好的活性,当其掺量不超过40%时,砂浆28d抗压强度高达50 MPa。微观分析结果表明:建筑垃圾复合粉体材料比砖粉具有更好的颗粒级配,在其水化过程中降低了稳定性较差的Ca(OH)2含量,提高了水泥石密实度,是一种经济、环保的新材料。     

Cation exchange applications of synthetic tobermorite for the immobilization and solidification of cesium and strontium in cement matrix

Immobilization and solidification of hazardous cations like Cs¹³⁷ and Sr⁹⁰ are required while handling the radioactive waste of nuclear power plants. Efforts are on to find a fail proof method of safe disposal of nuclear wastes. In this context, various materials like borosilicate glass, zeolites, cements and synthetic rocks have been tried by several workers. This communication deals with the synthesis, characterization, cesium uptake capacity and leaching behaviour of synthetic alumina-substituted calcium silicate hydroxy hydrate, which are close to that obtained for the natural mineral,11 Å tobermorite. The synthetic mineral show cation selectivity for Cs⁺ in presence of500–1000 times concentrated solutions of Na⁺, K⁺, Mg⁺, Ca²⁺, Ba²⁺ and Sr²⁺. Although the ordinary portland cement (OPC) which is often used in waste management operations alone holds negligible amounts of Cs⁺ and Sr²⁺, the addition of alumina-substituted tobermorite to OPC enhances the retention power of cement matrix by drastically lowering the leach rate of cations     

Studies on Erythromycin Melibionate — A New Water Soluble Salt of Ebythrowycin

Abstract

A new water soluble salt of erythromycin, erythromycin melibionate was prepared and some physicochemical and biological properties of the salt was studied. The salt was found to be soluble in water as well as in various organic solvents and was partitioned well in different organic solvent - water systems.

The potency of the new salt determined by microbiological assay was found to be 560, mcg/mg. The in vitro antimicrobial spectrum of the new salt was similar to that of the erythromycin     

Development of a technology for protecting groundwater from <Superscript>137</Superscript>Cs on the territory of the BN-350 reactor facility

The ability of clinoptilolite from the Chankanai deposit to take up ¹³⁷Cs from groundwater and liquid radioactive waste from the BN-350 reactor facility on the territory of TOO MAEK-Kazatomprom in Kazakhstan was studied. The results of semicommercial tests demonstrate the efficiency of using clinoptilolite as a reagent material of the geochemical barrier for preventing the radionuclide migration into groundwater in case of leakage of tanks for storage of liquid radioactive waste. A design of the geochemical barrier was suggested.     

Processes for treatment of liquid radioactive waste containing seawater

The main sources of formation of liquid radioactive waste (LRW) containing seawater are determined, and the main problems arising in management of such waste are analyzed. Sorption methods for removing long-lived Cs and Sr radionuclides from highly mineralized (>1 g L^–1) LRW are determined. The main physicochemical and sorption characteristics, advantages, and drawbacks of candidate sorbents for removing Cs and Sr radionuclides are described. Examples of using SRM and VS-5 chemical reaction sorption materials developed for removing Sr from LRW with the mineralization of up to 60 g L^–1 are given. The results of studying composite materials based on BaSiO₃ and resorcinol–formaldehyde resins, intended for removing Cs and Sr radionuclides from seawater, are analyzed. Composite sorbents of such type efficiently remove Cs and Sr radionuclides from seawater. Processes developed by the authors and brought into practice at various plants of the Far East for treatment of multicomponent LRW formed in the course of operation, repair, and decommissioning of nuclear-powered surface ships and submarines are described.     

Microbiological evaluation of the condition of cement compounds with radioactive wastes after long-term storage in near-surface repositories

Analysis of the core material taken by check drilling of a monolith of cemented radioactive waste in near-surface repositories operated for 15–45 years revealed the presence of damaged areas in the cement matrix. An association of microorganisms capable to cause degradation of silicate materials was found in cement compounds with radioactive waste. The bacterial genera, the metabolite concentrations, and the character of the biogenic acid damage of the cement matrix were determined. It was concluded that the safety level of the longterm storage of conditioned radioactive waste forms can decrease because of disturbances of the immobilization properties of the cement matrix as a result of biogenic damage of its microstructure. Biocidal additives were suggested for preventing degradation processes in the modern stages of radioactive waste cementation.     

Physicochemical Processes Occurring in Long-Term Storage of Liquid Radioactive Waste in Deep Underground Collector Beds

Interaction under hydrothermal conditions (pressure 3 MPa; temperature 80-170°C; contact time up to 2500 h) of intermediate-level acidic waste with bed rock of the underground repository for liquid radioactive waste is studied. Transformation of rock-forming minerals under these conditions is accompanied by increasing absorption of ¹³⁷Cs, ⁹⁰Sr, ²³⁹Pu, ²⁴¹Am, and ²³⁸U.     

Leaching of<Superscript>60</Superscript>Co and<Superscript>137</Superscript>Cs from spent ion exchange resins in cement-bentonite clay matrix

The leaching rate of⁶⁰Co and¹³⁷Cs from the spent cation exchange resins in cement-bentonite matrix has been studied. The solidification matrix was a standard Portland cement mixed with 290–350 (kg/m³) spent cation exchange resins, with or without 2–5% of bentonite clay. The leaching rates from the cementbentonite matrix for⁶⁰Co : (4,2–7,0) × 10⁻⁵ (cm/d) and¹³⁷Cs : (3,2–6,6) × 10⁻⁴ (cm/d), after 125 days were measured. From the leaching data the apparent diffusivity of cobalt and cesium in cement-bentonite clay matrix with a waste load of 290–350 (kg/m³) spent cation exchange resins, was measured for⁶⁰Co : (1,1−4,0) × 10⁻⁶ (cm²/d) and¹³⁷Cs : (0,5–2,6) x× 10⁻⁴ (cm²/d), after 125 days. The results presented in this paper are part of the results obtained in a 20-year mortar and concrete testing project which will influence the design of radio-active waste management for a future Serbian radioactive waste disposal centre.     

Phase Composition,Structure, and Hydrolytic Durability of a Uranium-Containing Magnesium Potassium Phosphate Compound

Uranium-containing samples of magnesium potassium phosphate (MPP) compound were synthesized using a nitric acid uranium solution. Uranium is incorporated in the MPP compound in the form of potassium uranyl phosphate with the structure of metaankoleite natural mineral, K(UO₂)PO₄·3H₂O. The differential and integral uranium leach rates, determined in accordance with GOST (State Standard) R 52 126–2003 on the 28th day of contact of the compound with water, are 1.7 × 10^–6 and 2.7 × 10^–6 g cm^–2 day^–1, respectively, and the degree of leaching is 0.014%. High hydrolytic durability of the compound with respect to uranium leaching reduces the risk of release of uranium isotopes from radioactive waste into the environment.     

Synthetic calcium aluminosilicate as a matrix for radioactive waste immobilization

The dependence of the total leach rate of monoliths with deionized water at 90°? on the keeping time (MCC-1 test) was measured. The monoliths were formed by cold pressing (25°C, 500 MPa, 2 min) of an air-dry powder of synthetic calcium aluminosilicate (SCAS) with added 10, 20, or 30 wt % simulated radioactive waste (SRW) of the composition typical of radioactive and nonradioactive wastes from a closed nuclear cycle. The leach rate of the SCAS-SRW monolith is 6.7 × 10^-7, 7.2 × 10^-7, and 8.3 × 10^-7 g cm^-2 day^-1 at an SRW content of 10, 20, and 30 wt %, respectively.