Ultraphosphate glasses for immobilization of liquid plutonium-containing wastes

Ultraphosphate glass is proposed for immobilization, taking account of the composition of the plutonium-containing wastes at the Mining-Chemical conglomerate. Some parameters of the process of immobilizing in ultraphosphate glass model wastes with plutonium content up to 150 mg/liter are investigated. A preliminary study is made of the degree of inclusion of plutonium dioxide, the chemical stability of the ultraphosphate glass obtained, and the distribution of plutonium in it. It is shown that almost 2 times more plutonium dioxide is included in ultraphosphate glass than in phosphate glass. With respect to the main physicochemical parameters, specifically, component leachability, ultraphosphate glass can be regarded as promising for immobilizing liquid plutonium-containing wastes. __________ Translated from Atomnaya énergiya, Vol. 100, No. 5, pp. 368–372, May, 2006.     

Investigation of the incorporation of fission product surrogates and process impurities into ceramics

The results of a model experiment on removing fission products and process impurities from a melt consisting of the salts NaCl-2CsCl by precipitation of weakly soluble phosphates are described. x-Ray phase analysis, scanning electron microscopy, emission spectral analysis, and the MSS-1 test are used to study the chemical and phase composition as well as the thermal and chemical stability of a ceramic produced from the phosphate precipitate. It is shown that the main phase of the ceramic is monazite, which incorporates many rare-earth elements and processing impurities. The ceramic obtained possesses high hydrolytic and thermal stability, making it a promising material for immobilizing wastes. __________ Translated from Atomnaya énergiya, Vol. 102, No. 4, pp. 221–224, April, 2007.     

Handling radwastes during rehabilitation of contaminated objects and grounds of the Kirovo-Chepetsk Chemical Works

Possible ways to rehabilitate the objects and grounds of the Kirovo-Chepetsk Chemical Works are examined. The technology for handling radioactive wastes and methods for immobilizing radioactive wastes in repositories are briefly described. Ways to eliminate radwaste repositories containing medium-level wastes with fissile substances and α-emitters are proposed. The experience gained in such work abroad is examined and the problems which must be solved to rehabilitate the objects at the Works are noted.     

Epoxy Compounds for Immobilizing Radioactive Wastes

The radiation resistance of epoxy compounds, solidified by crystalline hardening agents - metaphenyl-enediamine and phthalic anhydride - is investigated. It is shown that under conditions of γ irradiation (E = 1.33 MeV) the temperature of vitrification of the compounds depends on the dose (doses up to 1500 Mrad were investigated) and temperature 20–160°C, and the radiation gas release depends on the vitrification temperature and irradiation. It is shown that epoxy-phthalic anhydride compound is best as a matrix for immobilizing solid radioactive wastes at high temperatures (80–130°C).The experiments showed that the proposed compound can be recommended for immobilizing solid radioactive wastes.__________Translated from Atomnaya Energiya, Vol. 98, No. 5, pp. 348–351, May 2005.     

Change of structure and properties of vitrified radioactive wastes during long-time storage in an experimental repository

Scanning electron microscopy, x-ray phase analysis, and infrared spectroscopy are used to study samples selected from a near-surface zone of glass blocks of vitrified radioactive wastes from nuclear power plants after they have been in storage in an experimental repository for 17 years. It is shown that only a thin surface layer of the glass blocks undergoes any changes due to environmental factors, and quartz-like and alumosilicate embryos are found on this layer. Translated from Atomnaya énergiya, Vol. 105, No. 2, pp. 87–93, August, 2008.     

Characteristics of Cast Stone cementitious waste form for immobilization of secondary wastes from vitrification process

The high-temperature in vitrification process of radioactive wastes could cause radioactive technetium (⁹⁹Tc) in secondary liquid wastes to become volatile. Solidified cementitious waste forms at low temperature were developed to immobilize radioactive secondary waste. This research focuses on the characterization of a cementitious waste form called Cast Stone. Properties including compressive strength, surface area, phase composition, and technetium leaching were measured. The results indicate that technetium diffusivity is affected by simulant type. Additionally, ettringite and AFm (Al₂O₃–Fe₂O₃–mono) main crystalline phases were formed during hydration. The Cast Stone waste form passed the qualification requirements for a secondary waste form, which are compressive strength of 3.45 MPa and technetium diffusivity of 10^?9 cm²/s. Cast Stone was found to be a good candidate for immobilizing secondary waste streams.     

Solidification of radioactive wastes by fusion in basalt

It is shown that fused bagalt is suitable for immobilizing radioactive wastes having high specific activity. It has been found that recrystallization of the vitreous fused basalt phase improves the properties of the material, principally the mechanical strength and the chemical stability. The calculated diffusion coefficients vary from 10^–15 to 10^–17 cm²/sec at temperatures of 30–70°C.Institute of Nuclear Studies, Czechoslovakian Academy of Sciences, Rzhezh near Prague, Czechoslovakia. Translated from Atomnaya Énergiya, Vol. 21, No. 4, pp. 285–289, October, 1966. The scientific research work on this subject is being done under an agreement with the International Atomic Energy Agency.     

Cement mortar-degraded spinney waste composite as a matrix for immobilizing some low and intermediate level radioactive wastes: Consistency under frost attack

The increasing amounts of spinning waste fibers generated from cotton fabrication are problematic subject. Simultaneous shortage in the landfill disposal space is also the most problem associated with dumping of these wastes. Cement mortar composite was developed by hydrating mortar components using the waste slurry obtained from wet oxidative degradation of these spinney wastes. The consistency of obtained composite was determined under freeze–thaw events. Frost resistance was assessed for the mortar composite specimens by evaluating its compressive strength, apparent porosity and mass loss at the end of each period of freeze–thaw up to 45 cycles. Scanning electron microscopy, infrared spectroscopy and X-ray diffraction analyses were performed for samples subjected to frost attack aiming at evaluating the cement mortar in the presence of degraded spinney waste. The cement mortar composite exhibits acceptable resistance and durability against the freeze–thaw treatment that could be chosen in radioactive waste management as immobilizing agent for some low and intermediate level radioactive wastes.     

Mathematical model for the selection of nuclear-safe matrix compositions for immobilization of fissioning materials

A mathematical model for making a preliminary selection of nuclear-safe matrix compositions for immobilizing fissioning materials was presented. Software was developed for IBM PCs possible to perform calculations of nucleaf-safe matrix compositions. The maximum concentration of fissioning materials in some, matrix compositions which are currently in use and in some promising compositions for immobilization of radioactive wastes was calculated. 2 figures, 3 tables. 12 references. State Scientific Center of the Russian Federation—A. A. Bochvar All-Russia Scientific-Research Institute of Standardization in Machine Engineering. Translated from Atomnaya énergiya. Vol. 88, No. 3, pp. 189–193, March, 2000.     

Gd1.6Nd0.4Zr2O7烧绿石的快速合成及其组织结构研究

为探索Gd₂Zr₂O₇烧绿石快速固化高放废物中锕系核素的新途径,实验用高温高压固相反应法在3～4GPa压力、1573～1673K温度范围内合成了Gd_1.6Nd_0.4Zr₂O₇烧绿石固化体,并利用X射线衍射仪、扫描电镜对样品进行了分析。结果表明：高温高压固相反应法可在极短时间（15min）内合成完全固溶的Gd_1.6Nd_0.4Zr₂O₇立方烧绿石固化体,较常用制备方法（一般合成时间不低于48h）快近200倍;用该技术合成的样品在常温常压下的相转变温度及压力得以显著提高,烧绿石相更趋稳定;样品晶格常数随Nd含量的增加及合成温度的升高而逐渐增大,随合成压力的增加而逐渐减小。这种快速高效的合成方法为未来开展高放核素的工业固化提供了一种新的技术途径和基本数据参考。     

Immobilization of Radioactive Wastes by Embedding in Phosphate Ceramic

This work is devoted to the development of new composite materials for immobilizing medium- and low-level radioactive wastes. The most promising existing methods of solidification are low-temperature processes because they do not require complicated equipment and are cost-effective. As a result of the higher cost-effectiveness, less inert binder is needed in the overall volume of the solidified compound. Thus, using a complex phosphate, making it possible to increase the concentration of toxic products while maintaining reliable immobilization of the latter, is certainly a topical problem.Two advantages of phosphate ceramic are that high concentrations of anions which adversely affect the longevity of cement monoliths can be included and the spectrum of types of wastes that can be handled, including sol residues of combustion and mercury-containing wastes, is wide.     

Principles for Finding Matrices for Immobilizing High-Level Wastes

The principles for selecting matrices for immobilizing high-level wastes are examined. It is concluded that actinides should be separated and embeeded in the matrix separately from fission products. The minimum significant specific activity, regulated by NRB-99, cannot be used to estimate the storage time. It is suggested that such values be established as a function of the leach rate of radionuclides and matrix materials from matrices. For this, a standard leaching method using pulverized materials must be specified. Two criteria are proposed for comparing matrix materials.     

A new method for monitoring the radiation damage in nuclear waste containers using ion channelling

Immobilizing nuclear wastes has been one of the most important challenges in nuclear technology. A method to quantify and monitor the radiation damage to waste immobilizing crystalline materials like zircon is proposed. This method will make use of proton/ion channelling measurements of the crystalline containment sample or test crystalline sample placed in the crystalline or amorphous containment of nuclear waste for a long time from years to a few decades and the mathematical method to determine the structure collapse rate of the containment material using channelling measurements. Implementation procedure of this method/technique for radiation damage measurement in nuclear waste container materials is described.     

Conditioning of Radioactive Ash Residue in a Wave of Solid-Phase Exothermal Reactions

The possibility is studied of using the heat of exothermal reactions in the solid phase for conditioning the ash residue produced when solid radioactive wastes are burned. It is shown that solidifcation of the ash wastes can be performed without input of energy from external heat sources with the final glassy product containing 50–60 mass % ash residue and meeting the requirements for solidified radioactive wastes. The conditioning is performed in a special crucible–container, intended not only for performing the process but also for subsequent storage or burial of the final product. 5 figures, 2 tables, 10 references.     

Assessment of <Superscript>137</Superscript>Cs decontamination of concrete by the reagent method

The particulars of radioactive contamination of concrete wastes by ¹³⁷Cs were studied. x-Ray phase analysis and chemical analysis show that clayey materials, including Al₂O₃, Fe₂O₃, K₂O, and MgO, on which ¹³⁷Cs sorption is possible, were present in the concrete wastes. The content and form in which ¹³⁷Cs was found in radioactive concrete wastes from nuclear power facilities as well as in model samples were determined. When the wastes were treated with nitric acid the binder dissolved and a polydisperse suspension formed. The suspension consisted of three phases: solution, fine suspension, and rapidly settling precipitate of heavy filler particles. x-Ray phase analysis was performed and the ¹³⁷Cs mass ratio and distribution in the phases were determined. The possibility of decontaminating the concrete by a reagent method was evaluated.     

秦山第二核电厂放射性废物管理进展

通过对秦山第二核电厂商业运行后三废系统的运行、废物流排放管理以及低、中放固体废物的管理情况进行比较,并与国内其它同类电厂的运行情况作了比较,介绍了秦山第二核电厂在三废管理方面已取得的进展,分析了目前存在的问题,提出了为实现秦山第二核电厂低、中放固体废物的最少化而采取的有效措施.     

Vitrification of a surrogate for high-level wastes from the Savannah River facility (USA) in a commercial cold-crucible facility

Experiments on vitrifying a surrogate for SB2 pulp from the Savannah River facility (USA) in a cold crucible with inner diameter 418 mm in a commercial facility for vitrifying medium-level wastes at the Moscow Scientific and Industrial Association Radon are performed. Borosilicate glass materials, containing 50 mass % oxides of the wastes, including a magnetitic spinel phase in amounts not exceeding 15 vol. %, are obtained. The maximum mass average velocity of the slurry load and the melt output reach 40 and 16 kg/h, respectively. This corresponds to maximum specific vitrified wastes production capacity 2830 kg/(m²·day). The specific energy consumption for obtaining the glass product is about 10 kW·h/kg, which is approximately half the level for reprocessing slurries in crucibles with half the diameter. The chemical stability of the glass materials is 10–50 times higher than that of the materials made from glasses recommended by the US Environmental Protection Agency. __________ Translated from Atomnaya énergiya, Vol. 102, No. 5, pp. 296–299, May, 2007.     

Vitrification of a simulator of Savannah River site (USA) wastes with high iron and aluminum content on bench and commercial facilities with a cold crucible

Bench and commercial-facility experiments have been performed on cold-crucible vitrification of a simulator of high-level wastes from the Savannah River site (USA). The wastes contained up to 29 mass% Fe₂O₃ and 26 mass% Al₂O₃. The specific product flow reached 1700 and 2450 kg/(m²·day) with specific energy consumption 14–16 and 9–10 kW·h/kg, respectively. The crucibles did not undergo any appreciable corrosion during the period of the work performed and are reusable. The product consisted of a borosilicate matrix, containing up to 10 vol.% crystalline phase of spinel. The method of induction melting in a cold crucible is especially effective for crucibles with a large diameter, since the specific productivity increases and the specific energy consumption on the vitrification of high-level wastes decreases. __________ Translated from Atomnaya énergiya, Vol. 104, No. 5, pp. 291–295, May, 2008.     

Ecological aspects of liquid radwaste confinement in the severnyi deep repository

The results of studies of the effect of low-level wastes on the ecosystem of the reservoir bed at the Severnyi site of the Mining-Chemical Works are presented. The technogenic action of the wastes is greatest on stratal liquid in the region of the injection wells. In all other sections of the site, the changes of the radiochemical, chemical, and microbiological parameters were either temporary or minimal, which shows that the effect of the wastes on the subterranean ecosystem is negligible. Microbial communities present in stratal water participate in the transformation of the macrocomponents of low-level wastes and in gas generation. The formation of an individual gas phase as a result of biogenic processes under the conditions of the repository is local and does not disrupt the site’s operating regime. The activity of denitrifying micro-organisms decreases the toxicity of the macrocomponents of the low-level wastes.     

Handling of the Bottom Residues of a Nuclear Power Plant

The following technological operations are suggested for reprocessing bottom residues from nuclear power plants: separation of radionuclides using oxidation, filtration, and selective absorption, solidification and long-term storage of secondary radioactive wastes (cement compound from filtration stage and spent sorbent in filters); concentration and obtaining dry salts from bottom residues from which radionuclides have been removed. Laboratory and stand tests have been performed, showing that radionuclides can be removed from the bottom residues of nuclear power plants to a level below the ASA_sat according to NRB-96. This treatment decreases the volume of radioactive wastes by approximately a factor of 100. The dried purified bottom residues, which are commercial danger class III wastes, are shipped to storage sites used for industrial wastes. Calculations of the material flows are performed and the site arrangement of the wastes is given for the bottom residues from the Kursk nuclear power plant. 1 figure, 2 tables, 12 references.