Immobilization Of Radioactive Waste Research Articles

Immobilization of chloride radioactive waste using a phosphate glass composite material

Immobilization of chloride radioactive waste using a phosphate glass composite material

The mechanical properties and chemical durability of granite wastes based glass-ceramics for immobilization of high-level nuclear wastes

In this study, granite wastes based glass-ceramics containing Eu2O3 as simulated radionuclides is successfully synthesized via solid-state sintering method. The structure, mechanical properties and chemical durability are deeply investigated. The main phase in this glass-ceramics is oxyapatite (Ca2Eu8(SiO4)6O2), which is dispersed homogenously in glass matrix. The density dominates the properties of Vickers hardness and bending strength. The glass-ceramics containing 40 wt % of Eu2O3 has the optimal mechanical properties, with the Vickers hardness, bending strength and density of 5.12 GPa, 79.44 MPa and 98.77 %, respectively. The 28 days leaching experiments show that the glass-ceramics containing 40 wt % of Eu2O3 possess excellent chemical durability. The leaching rate of Al3+, Si4+, Fe3+, Ca2+, Eu3+ are 2.3 × 10−4 g m−2 d−1, 6.0 × 10−4 g m−2 d−1, 1.2 × 10−7 g m−2 d−1, 1.4 × 10−4 g m−2 d−1, 3.6 × 10−8 g m−2 d−1, respectively. This study indicates that granite wastes based glass-ceramics have excellent chemical stability and mechanical properties, which provides a novel idea for the immobilization of radioactive wastes.

Influence of PbO/CuO ratio on phase composition, microstructure, melt wettability and recyclability of copper slag

An estimated 30 % (∼11 million metric tons/year) of copper slag is still landfilled globally. Fayalitic copper slags are well-characterized, and they can be recycled into multiple safe end-products. Conversely, the major type of copper slag produced in Europe is lead-rich and warrants a deep understanding of phases present in it, properties, adverse environmental implications, and potential recycling strategies. The objective of this study is to thoroughly investigate such industrial slags by XRF, XRD, Mössbauer spectroscopy, SEM-EDS, novel automated image analysis using Aphelion, hot stage microscopy (HSM) and assess prospects for their valorization. These industrial slags were found to contain 26.1–50.5 wt% PbO, 3.4–5.4 % As2O5 (both hazardous solid wastes), 5.7–42 % CuO (present partly as Cu2O, an irritant), 3.2–18.2 % Fe2O3, 14–15.9 % SiO2 and oxides of chromium, cobalt (both carcinogenic), zinc and nickel (both toxic to aquatic life). The limitation of Rietveld XRD for glass quantification was overcome by newly-developed automated SEM-image processing which revealed that all three slags consisted of a lead silicate glass (36.3–65.8 vol%). Strikingly, rarely published arsenic- and lead-based compounds like (Ca,Pb)4As2O9-x, (Ca,Cu)2PbO4-x and Pb3Ca2Si3O11 were found when slag's PbO/CuO was 0.6, and Ca3As2O8-Pb2SiO4 formed when PbO/CuO equaled 9. The challenge of toxicity and landfilling can be overcome as these slags have high potential to be used for recovery of copper, cobalt, and lead. Radioactive waste immobilization glasses, gamma-ray shielding concrete and glass-ceramic consumer products may also be prepared from them. HSM revealed that a higher PbO/CuO ratio results in lower melting point and smaller melt contact angle indicating such slag can be useful for manufacturing recycled ceramics but corrosive towards refractories in copper converter.

Immobilization of radioactive sulphate waste simulate in polymer–cement composite based on recycled expanded polystyrene foam: evaluation of the final waste form resistance for Cs-134 and Co-60 leachability

The present study investigates to incorporate spiked sulfate solution simulate into polymer–cement composite (PCC) based on recycled expanded polystyrene foam. The main aim is to convert the waste stream into leach resistance solid forms able to slow down or even to retard the back release of hazardous radionuclides to the surrounding. Effective parameters e.g. leachant medium, temperature, radioactivity contents, leachant volumes and radionuclides speciation versus the leaching time were studied. The incremental leach rate (Rn, cm/day) and the leach index (Lx) were evaluated for the final waste form after 145 days. The experimental results revealed that the Lx for the all variances can fulfil the waste acceptance criteria for the disposal facility and are above the threshold value of 6. Moreover, leach rate percentages for Cs-134 and Co-60 were not exceeding 2%. The acquired data, based on lab leaching experiments, can recommend the developed PCC under consideration for solidification of radioactive sulphate waste stream safely.

Mechanisms and energetics of calcium aluminosilicate glass dissolution through ab initio molecular dynamics-metadynamics simulations

The dissolution of silicate glasses has implications in diverse fields ranging from the immobilization of radioactive waste to the development of sustainable alternatives to Portland cement. Here, we used ab initio molecular dynamics simulations biased with well-tempered metadynamics to study Si-O-T bridge dissociation in calcium aluminosilicate glasses, crucial for understanding their dissolution. In a departure from the conventional Michalske-Freiman model, our findings reveal a nucleophilic substitution reaction mechanism characterized by a short-lived, 5-fold coordinated Si intermediate or transition state, depending on the Si bridge coordination, with a near-trigonal bipyramidal geometry. We find that the reorganization required for reaching this state causes the activation energy barriers to be dependent on the Si bridge coordination, with Si Q3 species serving as the rate-limiting step in the dissolution reaction. Our findings not only challenge long-standing theoretical models but also pave the way for more accurate and comprehensive frameworks for understanding the dissolution of silicate glasses in various applications.

MgO/KH2PO4 and Curing Moisture Content in MKPC Matrices to Optimize the Immobilization of Pure Al and Al-Mg Alloys.

Magnesium Potassium Phosphate Cements (MKPCs) are considered a good alternative for the immobilization of aluminium radioactive waste. MKPC composition and moisture curing conditions are relevant issues to be evaluated. The corrosion of pure aluminium (A1050) and AlMg alloys (AA5754) with 3.5% of Mg is studied in MKPC systems prepared with different MgO/KH2PO4 (M/P) molar ratios (1, 2, and 3M) and moisture curing conditions (100% Relative Humidity (RH) and isolated in plastic containers (endogenous curing)). The Al corrosion potential (Ecorr) and corrosion kinetic (icorr and Vcorr) are evaluated over 90 days. Additionally, the pore ion evolution, the matrix electrical resistance, the pore structure, and compressive strength are analysed. The corrosion process of Al alloy is affected by the pH and ion content in the pore solution. The pore pH increases from near neutral for the 1M M/P ratio to 9 and 10 for the 2 and 3M M/P ratio, increasing in the same way the corrosion of pure Al (AA1050) and AlMg alloys (AA5754). The effect of Mg content in the alloy (AA5754) becomes more relevant with the increase in the M/P ratio. The presence of phosphate ions in the pore solution inhibits the corrosion process in both Al alloys. The MKPC physicochemical stability improved with the increase in the M/P ratio, higher mechanical strength, and more refined pore structure.

Glass-Containing Matrices Based on Borosilicate Glasses for the Immobilization of Radioactive Wastes

Glass-containing materials are widely considered among the most reliable materials for the immobilization of radioactive waste materials. The present work considers the synthesis of glass–ceramic and glass crystalline composite materials based on borosilicate glasses. The synthesis of glass–ceramic materials was carried out by a gradual temperature decrease, followed by crystallization for several hours. Sintering of crushed samples with crystalline components was carried out as an alternative procedure. Porous glasses were produced from glass melts by quenching. After impregnating the resulting porous materials with aqueous solutions of cesium nitrate, compaction of the glass was carried out to form glass crystalline composites. The thermochemical characteristics of the parent glasses were determined using the differential scanning calorimetry method. The phase composition and structure of the glass-containing materials were determined using X-ray phase analysis, X-ray spectral microanalysis, and Raman spectroscopy.

Assessment of Applicability of Bentonite Clay-Based Ceramics for Immobilization of Radioactive Waste from Pyrochemical Processing of Spent Nuclear Fuel

Assessment of Applicability of Bentonite Clay-Based Ceramics for Immobilization of Radioactive Waste from Pyrochemical Processing of Spent Nuclear Fuel

Application of alkali-activated cements for immobilization of dry low-level radioactive waste containing copper ferrocyanide

Summary: The possibility of using alkali-activated slag-Portland cement for immobilization of dry radioactive waste containing copper ferrocyanide they been confirmed. Because of optimization, the areas of existence of compositions that provide the criterion requirements for strength have been established. Introduction of magnetite additives in the amount of 14...15 wt. % (factor X1), zeolite in the amount of 6.5...7.5 wt. % (factor X2) and dry radioactive waste on the basis of copper ferrocyanide in the amount of 10...12 wt. % (factor X3) provides criterion requirements for compressive strength. The maximum value of strength – 13.2 MPa on the 28th day of hardening is characteristic for the above-mentioned quantities of additives, and it is 1.32 times higher than the standard level of strength. Introduction of zeolite for 6.5...7.5 wt. % allows to reduce the mass and density (1.07 times), radioactivity (1.09 times) of composites. Because of modelling of compound compositions, especially in the expected reactions, the factor X3 shows a weakening factor contributing to the reduction of values of output parameters. Therefore, the introduction of dry radioactive waste into the alkaline slag-Portland cement matrix containing copper ferrocyanide is limited to no more than 12 wt. %. The processes occurring in the volume of the material explain the reduction of mass, density and radioactivity of the compounds. The energy released during radioactive decay of cesium, strontium and other radionuclides is absorbed by magnetite and converted into heat. Heat promotes the removal of physically bound and partially chemically bound water from the structure of tobermorite-like low-basic calcium hydrosilicates, hydrogranates, alkaline-alkaline-earth zeolite-type hydroaluminosilicates, and copper ferrocyanide hydrate shell. However, the passage of radiolysis does not affect the kinetics of strength gain of the compounds, but contributes to the increase of their compressive strength by 1.87 times compared to the strength of the compounds on 7 days of curing

BEHAVIOUR OF MAGNESIUM POTASSIUM PHOSPHATE CEMENT UNDER LEACHING

The effectiveness of the use of magnesium potassium phosphate cements (MKPC) as protective matrices for radioactive waste was investigated in this work from the point of view of the main elements leaching. Fly ash and blast-furnace slag were used as MKPC fillers. Leaching tests were performed on monolithic MKPC samples according to ANSI/ANS 16.1 test. Based on the obtained data, the cumulative leaching fractions, leaching mechanisms, effective diffusion coefficients and leaching indices K, Mg, P, B, Si, Al, and Ca were determined. The values of the calculated leaching indices (>10) of the constituent elements indicate that the characteristics of the studied MKPC matrix are acceptable for radioactive waste immobilization.

Influence of CeO2 on glass formation, structural and thermal properties of sodium niobium phosphate glass

Cerium is a well-known surrogate to trivalent and tetravalent actinides and has been investigated to explore the structural and thermal characteristics of simulated radwaste-loaded material systems. The solubility and distribution of CeO2 into a sodium niobium phosphate glass have been explored with the objective of developing a host matrix for radioactive waste immobilization. Niobium phosphate glasses of (100-x) (40Na2O–20Nb2O5–40P2O5)-xCeO2 (x = 0, 1, 3, 5, 8, and 10 mol%) compositions are prepared by melt-quenching method and are characterized by X-ray diffractometry (XRD), differential scanning calorimetry (DSC), Fourier transform infrared spectroscopy (FTIR), Raman spectroscopy and scanning electron microscopy (SEM). This study summarises the influence of cerium ion incorporation on the structure, thermal stability, and formation of crystalline phases of the niobium phosphate glass. The various glass stability parameters, such as Hruby (KH), Weinberg (KW), and Lu–Liu (KLL) parameters are estimated to assess the stability of the glass melt against devitrification in the supercooled region. The spectroscopic analysis ascertains the presence of metaphosphate and pyrophosphate networks as primary structural building units in the glass. However, the Raman analysis demonstrates the influence of CeO2 on the network linkages arising from the corner-sharing of NbO6 octahedrons and Nb–O–P bonding in the glass. The CeO2 loading is optimized up to 8 mol% into the base 40Na2O–20Nb2O5–40P2O5 glass matrix without precipitation of any crystalline phase. The glasses are heat treated to obtain glass-ceramics embedded with the crystalline phases, namely, monazite CePO4, NbOPO4, NaNbO3, and niobium phosphate bronze Na6Nb8(PO4)5O15. The microanalysis of these glass ceramics shows that the CeO2 is more aggregated inside the niobium phosphate bronze framework than in the bulk at a higher temperature. It is also observed that the surface crystallization process is a prevalent mechanism for the crystallization of glasses on heating at crystallization onset temperature.

Matrices for radioactive waste immobilization: a review

Nuclear energy is considered a clean, reliable, and an inexhaustible energy source for power generation. Nuclear power is harnessed from nuclear fission reactions in a dedicated power plant. The by-products (produced in the nuclear power plant) are radioactive and pose a threat to the environment. The safe disposal of nuclear waste is vital to ensure the sustainable use of the nuclear energy. The immobilization of radioactive waste before final disposal is essential for the interim storage and transportation. This review summarizes the recent work on glass, ceramics, and glass–ceramics matrices to immobilize high-level waste. The synthesis methods, leaching behavior, and radiation resistance of matrices are discussed briefly.

Structure and crystallizability of K2O-B2O3-SiO2 and K2O-B2O3-GeO2 glasses: Effect of composition and heat treatment mode

The stability and crystallizability of glasses K2O-B2O3-GeO2 and K2O-B2O3-SiO2 systems becomes an important consideration when selecting materials for optical purposes and the immobilization of radioactive wastes. To determine the crystallizability of boron-containing glasses, thermal treatments were carried out at 500 and 680 °C and the formed phases were identified. The structural features of the materials before and after treatment were studied by vibration spectroscopy. Glasses with B2O3/SiO2 = 2/5 and B2O3/GeO2 = 4/3 ratio proved to be the most resistant to heating. It was observed that the ratio of B/Si and B/Ge has a significant effect on crystallizability at any temperature. Heat treatment at 680 °C results in the formation of potassium borates, silicates, and borosilicates in the borosilicate system, while the crystallization of germanates and potassium borates is characteristic of the borogermanate system. Structural differences between the crystallization products of these two systems are shown to be due to the ability of germanium to form highly oriented structural units. With an increase in the B/Ge ratio, the crystallization of borogermanate glasses is shown to lead to a decrease in the proportion of crystalline borates. Conversely, during the crystallization of the borosilicate system, an increase in the B/Si ratio in the initial glass leads to an increase in their content.

Insight into immobilization mechanism of nuclear waste management by granite: Impact of waste chemical species

Nuclear waste immobilization is one of the ways to alleviate the harm of radioactive waste produced in the nuclear field to human beings and the environment. In the current study, granite was utilized as a substrate material for the immobilization of simulated radioactive waste. Based on the obtained results, the relationship between the waste chemical species and the physicochemical properties, such as solid-solubility of waste, infrared, density and leaching data, was systematically studied as well. The findings indicate that the granite substrate used for the immobilization of the simulated waste (Nd2O3) exhibits a solid-solubility limit of 27 wt%. When the added content of waste is below 27 wt%, all Nd elements are immobilized into the glassy phase. However, when the waste content exceeds 27 wt%, in addition to being immobilized in the glassy phase, the presence of Nd2O3 reacted with Ca3Si3O9 to form Ca2Nd8(SiO4)6O2. As the Nd2O3 content increased from 20 wt% to 25 wt%, the microstructures related to Si-, Al-, and Fe-are more involved in the immobilization of waste. However, the appearance of Ca2Nd8(SiO4)6O2 and Nd2O3 makes the density and Vickers hardness of the matrix not increase further. The maximum density and hardness values are 3.17 g cm−3 and 7.88 GPa, respectively. The leaching results show that the cured body has good anti-leaching performance (LRNd = ∼10−6 g cm−2 d−1). Finally, the comparison of vacancy formation energy of Nd in Nd2O3 and Ca2Nd8(SiO4)6O2 proved that the formation of Ca2Nd8(SiO4)6O2 is beneficial to the immobilization of simulated waste. This study provides a certain reference for the selection of nuclear waste solidification substrates. This kind of granite material from nature has the potential advantages of low cost and better immobilization effect on waste, and may be more suitable for the treatment of radioactive waste in the future.

Study of the behavior of cesium and strontium in co-melting basalt and silica gel containing cesium or strontium nitrates for the development of a potential material for radioactive waste immobilization

The work studied the distribution of cesium and strontium in cast stone matrices obtained by co-melting basalt and silica gel (hereinafter SiO2) containing cesium or strontium nitrates at 1623 K during 5 h. The melt cooling rate was ∼2 °C/min.Co-melting basalt and SiO2 containing CsNO3 or Sr(NO3)2 at a basalt/SiO2 initial mass ratio of 5 : 1 produced cast stone matrices, in which the main phases were Cr–Fe spinel and glass. At the same time co-melting basalt and SiO2 containing CsNO3 or Sr(NO3)2 at an initial mass ratio of 2 : 1 yielded matrices which, apart from Cr–Fe spinel and glass, included pollucite, clinopyroxene, and Ti-containing magnetite phases for Cs-containing system, magnetite and olivine phases for Sr-containing system.For Cs-containing system at a basalt/SiO2 initial mass ratio of 5 : 1 the glass phase showed almost an even distribution of cesium, whereas the spinel phase was virtually free of cesium. At a basalt/SiO2 initial mass ratio of 2 : 1 the major amount of cesium was found in the pollucite phase, where it mainly occurred in parts containing SiO2. Some amount of cesium was found in the glass phase. The other crystal phases found in the matrix were virtually free of cesium.For Sr-containing system at a basalt/SiO2 initial mass ratio of 5 : 1 the process yielded two types of glass: one rich in Ca and the other in Sr. Strontium was mainly found in the glass phases. The spinel phase was virtually free of strontium. At a basalt/SiO2 initial mass ratio of 2 : 1 strontium was mainly found in the glass phases. The magnetite, olivine, and Cr–Fe spinel phases were virtually free of strontium.

POSSIBILITES OF ISOTOPES PRODUCTION OF 153Sm, 175Yb, 186Re AT THE ELECTRONIC ACCELERATOR

Photonuclear technology for producing osteotropic isotopes 153Sm (Т1/2 = 1.9 days, Eβ = 0.8 MeV, Eγ = 103.2 keV), 175Yb (Т1/2 = 4.2 days, Eβ = 0.5 MeV, Eγ = 396.3 keV) and 186Re (T1/2 = 3.8 days, Eβ = 1.1 MeV, Eγ = 137.2 keV) with using nanoparticles (50…80 nm) of oxides of these elements and the Szilard-Chalmers reaction to increase specific activity have been developed at the NSC KIPT electron accelerator. Such medical isotopes are not produced in Ukraine. The overall electron accelerator isotope yield when bremsstrahlung irradiated these samples with a maximum energy of 40 MeV and a current of 250 µA for 186Re and 13.5 MeV and a current of 500 µA for 153Sm and 175Yb has been shown to have the advantages of higher specific activity, negligible the content of impurities and does not require immobilization of radioactive waste in comparison with a reactor and a cyclotron.

Bi0-Reduced Graphene Oxide Composites for the Enhanced Capture and Cold Immobilization of Off-Gas Radioactive Iodine.

Radioactive waste management is critical for maintaining the sustainability of nuclear fuel cycles. In this study, we propose a novel bismuth-based reduced graphene oxide (Bi0-rGO) composite for the immobilization of off-gas radioactive iodine. This material synthesized via a solvothermal route exhibited a low surface area (2.96 m2/g) combined with a maximum iodine sorption capacity of 1228 ± 25 mg/g at 200 °C. The iodine sorbent was mixed with Bi2O3 powder and distilled water to fabricate waste matrices, which were cold-sintered at 300 °C under a uniaxial pressure of 500 MPa for 20 min to achieve a relative density of ∼98% and Vickers hardness of 1.3 ± 0.1 GPa. The utilized methodology reduced the iodine leaching rate by approximately 3 orders of magnitude through the formation of a chemically durable iodine-bearing waste form (BiOI). This study demonstrates the high potential of Bi0-rGO as an innovative solution for the immobilization of radioactive waste at relatively low temperatures.

Waste acceptance criteria testings of a phosphate-based geopolymer waste form to immobilize radioactive borate waste

Immobilization of radioactive borate waste (BW) is an urgent issue in South Korea because of the lack of appropriate waste forms. This study aimed to investigate whether phosphate-based geopolymers could satisfy the waste acceptance criteria established by Korean government. Geopolymers with 40 wt% BW did not exceed the criterion of compressive strength (3.445 MPa) after the water immersion and thermal cycling tests, whereas geopolymers containing 30 wt% BW still showed higher than 3.445 MPa after waste acceptance criteria tests. In particular, the compressive strength increased and the total porosity significantly decreased after the thermal cycling and gamma irradiation tests. The leachability index of boron calculated according to the ANSI/ANS 16.1 test was found to be higher than the criterion value (>6.0), and the controlling leaching mechanism was diffusion. The characteristic analyses indicated that water-soluble borate phases were leached from the geopolymer during the water immersion and leaching tests. However, the structurally incorporated 4-coordinated boron was also present as a sparingly soluble phase. Consequently, phosphate-based geopolymers can be used as stable binder materials to immobilize radioactive BW.

Synthesis and study of zirconium-containing sodium-cesium aluminoborosilicate matrix materials

During the development of new composition versions for basic systems immobilizing high-level radioactive waste, the investigation of physico-chemical peculiarities has been carried out for various glass-ceramic materials, synthesized by the quench method from melts of the Na2O – Cs2O – CaO (SrO, BaO) – Al2O3 – B2O3 – SiO2 system with high zirconium content. Factors that determine solubility and zirconium retention have been studied, as well as homogeneity of the synthesized materials. The influence of zirconium upon the phase composition of materials, formation of their structure, crystallization processes, density values, and thermal properties have been considered. The obtained data point at possibilities in application of the chosen glass-forming system with the investigated combination of components in the field of synthesis of matrix materials for immobilization of high-level radioactive waste with high zirconium content.

Production and Properties of Magnesium Potassium Phosphate Cements Containing Ash and Metallurgical Slag Additives for Radioactive Waste Immobilization

Magnesium potassium phosphate cements (MKPCs) are currently considered as the matrix materials for immobilization of radioactive waste of low and medium activity, as an alternative to Portland cements. At the same time, MKPC compound has the following advantages: fast hardening, high early strength, low shrinkage and high chemical resistance. Magnesium potassium phosphate cement is prepared at room temperature by an acid-base reaction between calcined magnesium oxide (MgO), potassium dihydrogen phosphate (KH2PO4) and water to form magnesium potassium phosphate hexahydrate MgKPO4 ∙ 6H2O known as K-struvite. To obtain MKPC, starting components of high purity and, accordingly, high cost are required. From an industrial point of view, this leads to an excessive price of the final product, and therefore, the use of inexpensive industrial waste as cement fillers is economically cost effective.
 In the present paper, the effect of fly ash and blast-furnace slag additives on the microstructure, compressive strength, and chemical resistance of MKPC samples was studied. The results indicate that particles of both ash and blast-furnace slag are involved in the reaction of MKPC obtaining with the formation of predominantly crystalline K-struvite. According to X-ray diffraction analysis, the content of K-struvite in samples with additives of both fly ash and blast-furnace slag is almost the same and amounts to 58–60 %. With the same addition of filler, samples with the addition of blast-furnace slag have a denser structure, which is formed due to the increased reactivity of blast-furnace slag particles. Moreover, MKPC samples with the addition of blast-furnace slag exhibit higher strength and better chemical resistance to leaching compared to MKPC samples with fly ash additions. The obtained results demonstrate the prospects of using industrial waste additives in the production of MKPC compounds characterized by high mechanical strength and chemical resistance.