Magnesium Potassium Phosphate Compound for Immobilization of Radioactive Waste Containing Actinide and Rare Earth Elements

The problem of effective immobilization of liquid radioactive waste (LRW) is key to the successful development of nuclear energy. The possibility of using the magnesium potassium phosphate (MKP) compound for LRW immobilization on the example of nitric acid solutions containing actinides and rare earth elements (REE), including high level waste (HLW) surrogate solution, is considered in the research work. Under the study of phase composition and structure of the MKP compounds that is obtained by the XRD and SEM methods, it was established that the compounds are composed of crystalline phases—analogues of natural phosphate minerals (struvite, metaankoleite). The hydrolytic stability of the compounds was determined according to the semi-dynamic test GOST R 52126-2003. Low leaching rates of radionuclides from the compound are established, including a differential leaching rate of 239Pu and 241Am—3.5 × 10−7 and 5.3 × 10−7 g/(cm2∙day). As a result of the research work, it was concluded that the MKP compound is promising for LRW immobilization and can become an alternative material combining the advantages of easy implementation of the technology, like cementation and the high physical and chemical stability corresponding to a glass-like compound

The Influence of Zeolite (Sokyrnytsya Deposit) on the Physical and Chemical Resistance of a Magnesium Potassium Phosphate Compound for the Immobilization of High-Level Waste

The manuscript presents the results of the development of new material for high-level waste (HLW) management: the magnesium potassium phosphate (MKP) compound. The possibility of using zeolite (Sokyrnytsya deposit) to increase the mechanical, thermal, and hydrolytic resistance of this compound with immobilized HLW was studied. The main component of the used natural zeolite is a mineral of the clinoptilolite–heulandite series, and quartz, microcline, and clay minerals (illite, sepiolite, and smectite) are present as impurities. The compressive strength of the compound, containing at least 4.2 wt % zeolite, is about 25 MPa. Compound containing 28.6 wt % zeolite retains high compressive strength (at least 9.0 MPa), even after heat treatment at 450 °C. The adding of zeolite to the composition of the compound increases its hydrolytic stability, while the leaching rate of the mobile nuclides 137Cs and 90Sr decreases up to one order of values. Differential leaching rate of radionuclides from the compound containing 28.6 wt % zeolite is 2.6 × 10−7 for 137Cs, 2.9 × 10−6 for 90Sr, 1.7 × 10−9 for 239Pu, and 2.9 × 10−9 g/(cm2∙day) for 241Am. Thus, the properties of the resulting compound correspond to the requirements for solidified HLW in Russia

Medium-Temperature Phosphate Glass Composite Material as a Matrix for the Immobilization of High-Level Waste Containing Volatile Radionuclides

The search for matrices and technological solutions for the reliable immobilization of volatile radionuclides and high-level waste (HLW) components is an actual radiochemical problem. Methods of obtaining of sodium alumino-iron phosphate (NAFP) and iron phosphate (FP) glass composite materials synthesized at temperatures of 450–750 °C, their structure and hydrolytic stability were investigated in this paper. The structure of the samples was studied by XRD and SEM-EDS. It was shown that, in the case of FP materials, the phase composition varies depending on the synthesis temperature, while NAFP materials have a complex multiphase composition at all crystallization temperatures. It has been established that the samples of the obtained glass composite materials have a high hydrolytic stability. At the same time, FP material obtained at 650 °C are the most stable, which makes this medium-temperature method of synthesis promising for the immobilization of volatile HLW components

Medium-Temperature Phosphate Glass Composite Material as a Matrix for the Immobilization of High-Level Waste Containing Volatile Radionuclides

The search for matrices and technological solutions for the reliable immobilization of volatile radionuclides and high-level waste (HLW) components is an actual radiochemical problem. Methods of obtaining of sodium alumino-iron phosphate (NAFP) and iron phosphate (FP) glass composite materials synthesized at temperatures of 450–750 °C, their structure and hydrolytic stability were investigated in this paper. The structure of the samples was studied by XRD and SEM-EDS. It was shown that, in the case of FP materials, the phase composition varies depending on the synthesis temperature, while NAFP materials have a complex multiphase composition at all crystallization temperatures. It has been established that the samples of the obtained glass composite materials have a high hydrolytic stability. At the same time, FP material obtained at 650 °C are the most stable, which makes this medium-temperature method of synthesis promising for the immobilization of volatile HLW components

Medium-Temperature Glass-Composite Phosphate Materials for the Immobilization of Chloride Radioactive Waste

Among the many radiochemical problems, the search for new materials and technologies for the immobilization of radioactive waste remains relevant, and the range continues to change and expand. The possibility of immobilizing the spent chloride electrolyte after the pyrochemical processing of the mixed uranium-plutonium spent nuclear fuel of the new fast reactor BREST-OD-300 on lead coolant into glass-composite phosphate materials synthesized at temperatures of 650–750 °C was studied. The structure of the obtained samples was studied using XRD and SEM/EDS methods. It has been shown that the spent electrolyte simulator components create stable mixed pyrophosphate phases in the glass composite structure. The materials were found to have high hydrolytic stability. This indicates the promise of using phosphate glass composites as materials for the reliable immobilization of the spent electrolyte

Magnesium Potassium Phosphate Compound for Immobilization of Radioactive Waste Containing Actinide and Rare Earth Elements

The problem of effective immobilization of liquid radioactive waste (LRW) is key to the successful development of nuclear energy. The possibility of using the magnesium potassium phosphate (MKP) compound for LRW immobilization on the example of nitric acid solutions containing actinides and rare earth elements (REE), including high level waste (HLW) surrogate solution, is considered in the research work. Under the study of phase composition and structure of the MKP compounds that is obtained by the XRD and SEM methods, it was established that the compounds are composed of crystalline phases—analogues of natural phosphate minerals (struvite, metaankoleite). The hydrolytic stability of the compounds was determined according to the semi-dynamic test GOST R 52126-2003. Low leaching rates of radionuclides from the compound are established, including a differential leaching rate of 239Pu and 241Am—3.5 × 10−7 and 5.3 × 10−7 g/(cm2∙day). As a result of the research work, it was concluded that the MKP compound is promising for LRW immobilization and can become an alternative material combining the advantages of easy implementation of the technology, like cementation and the high physical and chemical stability corresponding to a glass-like compound

Solidification of high level waste using magnesium potassium phosphate compound

Compound samples based on the mineral-like magnesium potassium phosphate matrix MgKPO4 × 6H2O were synthesized by solidification of high level waste surrogate. Phase composition and structure of synthesized samples were studied by XRD and SEM methods. Compressive strength of the compounds is 12 ± 3 MPa. Coefficient of thermal expansion of the samples in the range 250–550 °C is (11.6 ± 0.3) × 10−6 1/°C, and coefficient of thermal conductivity in the range 20–500°С is 0.5 W/(m × K). Differential leaching rate of elements from the compound, g/(cm2 × day): Mg - 6.7 × 10−6, K - 3.0 × 10−4, P - 1.2 × 10−4, 137Cs - 4.6 × 10- 7; 90Sr - 9.6 × 10−7; 239Pu - 3.7 × 10−9, 241Am - 9.6 × 10−10. Leaching mechanism of radionuclides from the samples at the first 1–2 weeks of the leaching test is determined by dissolution (137Cs), wash off (90Sr) or diffusion (239Pu and 241Am) from the compound surface, and when the tests continue to 90–91 days - by surface layer depletion of compound. Since the composition and physico-chemical properties of the compound after irradiation with an electron beam (absorbed dose of 1 MGy) are constant the radiation resistance of compound was established. Keywords: Magnesium potassium phosphate compound, High level waste, Immobilization, Leaching rate, Leaching mechanism, Radiation resistanc

Tritium Labeling and Phase Distribution of 18-Crown-6 and Its Derivatives for Further Reprocessing of Radium Waste

To date, the world has accumulated a large amount of long-lived radioactive materials that need to be disposed of or reprocessed. Such materials include nuclear legacy objects containing 226Ra, which is an important material for obtaining a wide range of isotopes for nuclear medicine via irradiation in reactors, cyclotrons, and electron accelerators. For the selective recovery of 226Ra from waste materials, crown-ether (CE) 18-crown-6 (18C6) or its derivatives can be used, which, however, have not been widely studied for these purposes. In our work, the key property of 18C6 and its derivatives, the phase distribution, was studied using tritium labeling. The possibility of introducing a tritium label into CEs molecules using thermal activation of tritium has been demonstrated; a high specific activity of the obtained compounds was achieved (from 18 to 108 TBq/mol). Methods for chromatographic purification of the studied CEs were developed. The distribution of 18C6 and its derivatives between various organic solvents and water was studied in detail for the first time. Subsequently, the obtained data will allow us to choose conditions for the selective recovery of 226Ra from aged sources

Addition of Zn during the phosphine-based synthesis of indium phospide quantum dots: doping and surface passivation

Zinc-doped InP(Zn) colloidal quantum dots (QDs) with narrow size distribution and low defect concentration were grown for the first time via a novel phosphine synthetic route and over a wide range of Zn doping. We report the influence of Zn on the optical properties of the obtained quantum dots. We propose a mechanism for the introduction of Zn in the QDs and show that the incorporation of Zn atoms into the InP lattice leads to the formation of Zn acceptor levels and a luminescence tail in the red region of the spectra. Using photochemical etching with HF, we confirmed that the Zn dopant atoms are situated inside the InP nanoparticles. Moreover, doping with Zn is accompanied with the coverage of the QDs by a zinc shell. During the synthesis Zn myristate covers the QD nucleus and inhibits the particle growth. At the same time the zinc shell leads to an increase of the luminescence quantum yield through the reduction of phosphorous dangling bonds. A scenario for the growth of the colloidal InP(Zn) QDs was proposed and discussed

Recovery of 177Lu from Irradiated HfO2 Targets for Nuclear Medicine Purposes

A new method of production of one of the most widely used isotopes in nuclear medicine, 177Lu, with high chemical purity was developed; this method includes irradiation of the HfO2 target with bremsstrahlung photons. The irradiated target was dissolved in HF and then diluted and placed onto a column filled with LN resin. Quantitative sorption of 177Lu could be observed during this process. The column later was rinsed with the mixture of 0.1 M HF and 1 M HNO3 and then 2 M HNO3 to remove impurities. Quantitative desorption of 177Lu was achieved by using 6 M HNO3. The developed method of 177Lu production ensures high purification of this isotope from macroquantities of hafnium and zirconium and radioactive impurities of carrier-free yttrium. The content of 177mLu in 177Lu in photonuclear production was determined. Due to high chemical and radionuclide purity, 177Lu obtained by the developed method can be used in nuclear medicine