Immobilization Of High-level Waste Research Articles

Effects of high valency and polarizability ion (W6+) on the phase evolution of CaZr1–2xLn2xTi2–xWxO7 zirconolite-based solid solution (Ln = Nd, Sm, Gd, Ho, Yb)

Zirconolite-based structures have served as robust matrices for the incorporation of minor actinide-rich high-level waste (HLW). The co-doping effects are critical for exploring the phase evolution of zirconolite-based structures, thereby providing a guide for HLW immobilization. Nevertheless, few studies have studied the synergistic effect of high valent ion W6+ and Lanthanides (Ln, surrogates to minor actinides) co-doping. This study proposed a systematical study on the Ln-W (Ln = Nd, Sm, Gd, Ho, Yb) co-doping in zirconolite-derived structure using X-ray diffraction (XRD) and scanning electron microscope with energy dispersive X-ray spectroscopy (SEM-EDX). The results showed that near-single-phase zirconolite-2M was recorded for samples with low level doping of Ln-W (Nd–W, Sm–W and Gd–W with x = 0.05; Ho–W and Yb–W with x = 0.05–0.1), and its structure underwent a moderate expansion along the c-axis, as revealed by a Pawley refinement method. One key finding is that near-single-phase zirconolite-2M was recorded for samples with low level doping of Ln-W (Nd–W, Sm–W and Gd–W with x = 0.05; Ho–W and Yb–W with x = 0.05–0.1), and its structure underwent a moderate expansion along the c-axis, as revealed by a Pawley refinement method. Notably, in addition to a transformation from zirconolite-2M to pyrochlore, a pseudo-yellow phase CaWO4 was also noted with increasing co-doping contents. This work demonstrates that the high charge compensator W6+ and lanthanides can be simultaneously incorporated into zirconolite structure, and a different phase transformation was revealed compared to that in low valent element doping system, which gives a new insight into the optimization of HLW immobilization route.

A study of the mechanical and thermophysical properties of crystal matrices for the immobilization of high-level wastes

Objectives. The aim of the study was to confirm the compliance of the mechanical and thermophysical properties of titanate-zirconate mineral-like matrices intended for immobilization of the rare-earth-actinide fraction of high-level waste (HLW) with pyrochlore structures (Nd2ZrTiO7) and orthorhombic titanate of rare earth elements (Nd4Ti9O24+TiO2) with the Russian requirements for the final forms of radioactive waste sent for disposal. With regard to fractionated radioactive waste, this type of matrix is preferable when compared with conservative aluminophosphate and borosilicate glasses. This is due to larger capacity, and a better level of chemical, thermal, and radiation resistance.Methods. The synthesis of mineral-like matrices was carried out by remelting a granular precursor consisting of mineral-forming metal oxides and a solution imitating the rare earth-actinide fraction of HLW in an induction furnace with a cold crucible. The thermal diffusivity was determined by the laser flash method. The heat capacity of the matrix samples was measured by differential scanning calorimetry. Ultimate flexural and compressive strengths were determined using universal test machines. The elastic moduli (Young’s) were measured by the acoustic method. The temperature coefficients of linear expansion were determined using a high-temperature dilatometer.Results. The ultimate strength of the matrices (Nd2ZrTiO7) and (Nd4Ti9O24+TiO2) was found to be 150–179 and 20.6–57.8 MPa in compression and bending respectively. Young’s moduli vary from 3.7 ∙ 107 to 2.15 ∙ 108 kN/m2. With an increase in temperature from 50 to 500°C, the values of thermal conductivity have a pronounced tendency to decrease from 1.71 to 0.91 W/(m∙K). The temperature coefficients of linear expansion increase from 6.96 ∙ 10−6 to 1.01 ∙ 10−5 K−1 in the same temperature range.Conclusions. Comprehensive studies of titanate-zirconate mineral-like matrices show that their mechanical and thermal properties in certain cases significantly exceed the minimum requirements of regulatory documentation for the final forms of HLW.

High valency of charge compensator (Mo6+) to substitute Ti site in REE doped zirconolite (REE=Nd, Sm, Gd, Ho and Yb): Solid solubility, phase evolution and structural analysis

Zirconolite ceramics have been proven to be a promising matrix for high-level waste immobilization, especially for minor actinides. In this study, a series of CaZr1-2xREE2xTi2-xMoxO7 (REE = Nd, Sm, Gd, Ho, Yb) samples were prepared to investigate the formation of zirconolite REE as surrogates of minor actinides and Mo6+ as charge compensator. Synchrotron and in-house powder X-ray diffraction (XRD), X-ray absorption near edge spectroscopy (XANES), and scanning electron microscopy-energy dispersive X-ray spectroscopy (SEM-EDX) were used to study their solid solubility, phase evolution, and structural details. XRD results show that the formation of zirconolite-2M can be up to x = 0.15 in the CaZr1-2xNd2xTi2-xMoxO7 samples when co-existence of 2 M and 4 M was observed at x = 0.20 and 0.25. The “yellow phase” appears when x exceeds 0.4. The phase evolutions of CaZr1-2xSm2xTi2-xMoxO7, CaZr1-2xGd2xTi2-xMoxO7, CaZr1-2xHo2xTi2-xMoxO7 and CaZr1-2xYb2xTi2-xMoxO7 solid solutions are similar to each other: (1) zirconolite-2M in x = 0.05 and 0.10; (2) co-existence of 2 M and 4 M and then formation of 4 M phase from x = 0.15 to 0.25; (3) formation of “yellow phase” and pyrochlore when x ≥ 0.30. Interestingly, perovskite exits in all the ceramic samples, and its lattice parameters slightly change when increasing the dopants. Rietveld refinement results reveal that Sm/Ho mainly incorporates into 8f (Wyckoff position) Zr-sites while Mo replaces the 8f (Wyckoff position) Ti-sites. XANES results further confirm the occupation of Mo in TiO5 coordination environments. Furthermore, there is a considerable amount of Sm/Ho incorporated into 8f Ca-sites, demonstrating a different substitution mechanism from the preliminary design.

Nuclear Waste Management

Nuclear energy production generates nuclear waste. Nuclear and radioactive waste, especially high level waste (HLW) and intermediate level waste (ILW), require special long-term safe management solutions. One part of the solution involves recycling through reprocessing of spent nuclear fuel. Recycling reduces the volume of nuclear waste and allows for the reuse of some of its components. This can be achieved through methods such as adsorption, ion exchange, coagulation, flotation, filtration, chemical precipitation, reverse osmosis, and solvent extraction, like the PUREX process. Another possible future solution involves partitioning and transmutation technology, which can reduce the production of nuclear waste. The best long-term solution is the immobilisation of HLW and ILW in a solid matrix. Materials used for this purpose include glasses, cements, bitumen, geopolymers, concrete, and ceramics as a promising material. While cementation is still the most commonly used immobilisation method due to its low cost and simplicity, vitrification is a more permanent long-term solution. Deep geological disposal in combination with vitrification and a robust multi-barrier system is considered the most acceptable solution for safe nuclear waste isolation. This review provides insight into the mostly commonly used and promising immobilisation materials, as well as the most effective methods and technologies currently in use and under development for the management of HLW and ILW in nuclear waste management.

Evaluation of fracture resistance behavior of Alloy 690 material upto 1000°C

Alloy 690 is a nickel-based alloy with a nominal composition of 30% Cr and 10% Fe (wt%). It is denoted as Ni-30Cr-10Fe. This alloy is currently being investigated by various researchers for its use in nuclear industry as a possible replacement of alloy 600 steam generator tubes because of its excellent resistance to inter-granular stress corrosion cracking in water reactor environment. Alloy 690 is mainly used in vitrification melters for immobilization of high level waste in nuclear recycle plant. The temperature of water reactor environment is considerably lower compared to that of vitrification application, where temperature can reach an order of 1000 to 1100˚C. The mechanical and fracture properties of alloy 690 at the elevated temperature is important from the point of view of design and safety analysis of vitrification melter components and it is not available in literature. For this purpose, single edged notched tensile specimens have been fabricated from alloy 690 plates of 10 mm thickness. The fracture specimens have been tested in the temperature range of 25˚C to 1000˚C, which is the operating range of vitrification melters. The crack initiation toughness was obtained using modified blunting line equation in order to account for material strain hardening effect at a given temperature. From the fracture test results, it was observed that the initiation toughness is lowest at 700˚C. To understand this behavior, scanning electron microscopy was carried out on the fracture surface of the tested specimens. The fracture surface of the specimen tested at 700˚C showed presence of brittle phases in the microstructure, which are responsible for lower value of fracture toughness at this condition. The results of this work shall be useful for design and analysis of vitrification melter components.

Corrosion testing of domestically produced baddeleyitecorundum and chromium-containing materials in borosilicate glass melts

Practical importance of corrosion and erosion resistance of refractories was justified for the design of furnaces with direct electric heating that are used for vitrification of high-level waste (HLW). The main promising materials are listed, and the algorithm of assessment of their resistance to the effects of borosilicate melts is given for the operating conditions of electric furnaces. Corrosion testing of refractories was carried out under static and dynamic conditions in a low-melting borosilicate glass melt and in a borosilicate glass melt containing simulated liquid HLW. Refractory materials were selected that have the highest resistance in a glass melt under conditions of HLW immobilization. The testing results will be used for choosing lining material in the design of removable and small-scale melters. Ill. 5. Ref. 11. Tab. 3.

Synthesis and feasibility studies of doping U at Ti site of Y2Ti2O7 as a radioactive waste immobilization matrix.

In pursuit of clean and green nuclear energy one of the major challenges is to effectively immobilize the nuclear waste. In this context A2B2O7 type pyrochlore owing to its structural flexibility, ability to accommodate ions at both A/B-sites and high radiation tolerance has demonstrated excellent capability to store highly radioactive actinide ions. To fill the major gap area of actinide doping at the B site we have taken up the challenge of doping uranium ions at the Ti site of Y2Ti2O7 type pyrochlore. An yttria titanate (Y2Ti2-xUxO7; x = 0.05, 0.075, 0.1, 0.2, and 0.3) based matrix with uranium doped at the Ti site was synthesized using a simple gel combustion route under an air atmosphere. Rietveld refined X-ray diffraction (XRD) demonstrated that Y2Ti2O7 can accommodate U up to 5 mol% in the Ti site without any phase separation, which was further confirmed using Raman spectroscopy. Y2Ti2O7 based matrices are found to be radiation stable up to 1000 kGy and at the same time they are moderately thermally stable and on a par with the values reported for pyrochlores. Uranium in Y2Ti2O7 stabilizes in +6 oxidation state in the form of uranyl ion distributed near and far off from titanium vacancies with distinct excited state lifetime. This work could provide a smart and strategic way for selecting a suitable advanced ceramic matrix for immobilization of high level waste with additional and important information on solubility limit, actinide speciation, radiation/thermal stability, actinide concentration, etc.

In-situ observation of Kr2+ irradiation induced amorphization on Ca10-xLax(PO4)6-x(SiO4)xF2 (x=1, 2 and 6)

Fluorapatite, with a formula of A10(XO4)6F2, is a potential immobilization waste form of high-level waste due to its variable composition and stable structure. In this study, Ca10-xLax(PO4)6-x(SiO4)xF2 (x=1, 2 and 6) compounds with different doping amounts were synthesized. Samples were in-situ irradiated by 800 keV Kr2+ ions in the temperature range of 293 K to 673 K, with a flux of ∼1.72 × 1012 Kr2+/(cm2·s). Each specimen was continuously irradiated until the Selected Area Electron Diffraction (SAED) pattern indicated that it was fully amorphous. The corresponding critical amorphous temperatures Tc for these apatites were 514 K, 612 K and 737 K, respectively. The incorporation of lanthanum and silicon significantly reduced the radiation tolerance of fluorapatites. In terms of structure, SiO44− with larger size is more difficult to migrate than PO43− during thermal spike annealing. Besides, the occupancy of La3+ was also analyzed using the Rietveld refinement method. It is found that the preference of La3+ for Ca2 sites reduces the stability of the apatite structure. The ionic and covalent properties of the samples are also discussed as an essential factor affecting radiation resistance. The covalent SiO44− is considered to have a greater impact on the irradiation performance than La3+, and this effect is enhanced with the increase of the doping amount. This study investigated the radiation-induced amorphization of lanthanum-silicon-doped fluorapatite, which is of great significance for the study of high-level waste immobilization materials with different loadings.

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.

Behavior of Glass-like and Mineral-like Phosphate Compounds with an Immobilized Chloride Mixture in Hydrogen Peroxide Solutions

A new type of high-level waste (HLW) is generated during pyrochemical reprocessing of mixed nitride spent uranium–plutonium nuclear fuel. Such waste is a spent electrolyte, which is a mixture of chloride salts containing approximately 25.7 wt.% LiCl + 31.6 wt.% KCl + 4.1 wt.% CsCl + 5.1 wt.% BaCl2 + 3.8 wt.% SrCl2 + 29.7 wt.% LaCl3, and its immobilization in reliable matrices is an actual radiochemical problem. The structure and hydrolytic stability of sodium aluminoironphosphate (NAFP) glass and a low-temperature mineral-like magnesium potassium phosphate (MPP) matrix, which are promising for spent electrolyte immobilization in the presence of hydrogen peroxide solutions simulating natural water radiolysis products, were studied in this work. The structure of the samples was studied using the SEM-EDS method. It was shown that the initial samples of NAFP glass after leaching in hydrogen peroxide solutions are prone to precipitation of crystalline phases on the surface, which are mainly represented by a mixture of sodium–iron–aluminum pyrophosphates. It was established that the leaching rate of structure-forming components of NAFP and MPP matrices generally increase, but remain at a low level, meeting modern requirements for HLW immobilization. This confirms the effectiveness of the studied matrices for the industrial use of the spent electrolyte.

Effect of Nd2O3 dosages and melting temperatures on pyrochlore borosilicate glass-ceramics immobilization

In this study, borosilicate glass-ceramic waste forms containing pyrochlore crystalline phase (PBGC) were prepared by one-step heat-treated method. The effect of Nd2O3 dosages and melting temperatures on the crystalline phases and microstructure of PBGC were studied by using X-ray diffraction (XRD), scanning electron microscopy (SEM) and energy dispersive spectroscopy (EDS). Product Consistency Test (PCT) was used to investigate the aqueous durability of PBGC. The 14 wt% Nd2O3 doping in PBGC contributes to the formation of pyrochlore (Ca,Na)(Nb,Ti)2Nd0.67O6F. The optimal supercooling region of the glass-ceramic is from the melting temperature of 1325 °C to the initial cooling temperature of 1000 °C. The prepared glass-ceramic has a stable pyrochlore crystalline structure. In addition, the normalized leaching rates of major elements in PBGC are comparable to those of standard reference glass for immobilization of high-level wastes (HLWs). The results of this study provide useful guidance on the immobilization of certain HLWs in the industry using glass-ceramic materials.

Rapid solidification synthesis of novel (La,Y)2(Zr,Ti)2O7 pyrochlore-based glass-ceramics for the immobilization of high-level wastes

Pyrochlore-based glass-ceramics are attracting attention as a promising host material for immobilizing minor actinide in high-level wastes. However, the loading capacity of pyrochlore-based materials for actinide elements is to be improved, and only conventional synthetic approaches of powder sintering and glassy bulk crystallization were reported. Here, new (La,Y)2(Zr,Ti)2O7 pyrochlore-based glass-ceramics were subtly designed and successfully prepared through a rapid melting and solidification process. The solidification behavior of samples with different cooling rates was investigated, the dense and crack-free (La,Y)2(Zr,Ti)2O7 pyrochlore-based glass-ceramics were prepared at the cooling rate of 500 °C/s (within 1.5 s). The present amorphous La2O3-TiO2-Al2O3 phase demonstrates the outstanding thermal stability compared with the currently used borosilicate glass for high-level waste immobilization due to the lack of conventional glass-forming components. The results show the great feasibility of this novel synthesis route for glass-ceramics by direct solidification and its good potential for immobilizing high-level wastes.

Compound-tunable embedding potential method and its application to calcium niobate crystal CaNb2O6 with point defects containing tantalum and uranium

The compound-tunable embedding potential (CTEP) method developed for simulating the influence of environment on a fragment in the ionic-covalent crystal is presented in the form of a linear combination of particular short-range semilocal pseudopotentials for the atoms of nearest environment and the long-range Coulomb potentials from optimized fractional point charges centered on both nearest and more distant atoms of the environment. A pilot application of the CTEP method to calcium niobate crystal, ${\mathrm{CaNb}}_{2}{\mathrm{O}}_{6}$, is performed. A very good agreement of the electronic density and interatomic distances within the relaxed fragment with those of the original periodic crystal calculation is attained. Calcium niobate crystal can be considered as an idealized fersmite $({\mathrm{CaNb}}_{2\ensuremath{-}x}{\mathrm{Ta}}_{x}{\mathrm{O}}_{6},x\ensuremath{\approx}0.3)$ mineral when neglecting the contributions of the impurities with smaller molar fractions, and substitution $\mathrm{Nb}\ensuremath{\rightarrow}\mathrm{Ta}$ is considered here as a point Ta defect in ${\mathrm{CaNb}}_{2}{\mathrm{O}}_{6}$. Besides, uranium-containing point defects are also studied since the euxenite group minerals, to which fersmite belongs, are considered as prospective matrices for long-term immobilization of high-level waste. The chemical shifts of ${K}_{{\ensuremath{\alpha}}_{1,2}}$ and ${K}_{{\ensuremath{\beta}}_{1,2}}$ lines of x-ray emission (fluorescence) spectra in niobium are evaluated to analyze its chemical state in the crystal. Potential of CTEP for studying properties of point defects containing $f$ and heavy $d$ elements with relativistic effects, extended basis set, and broken crystal symmetry taken into account is discussed.

Probing electronic environment in gamma irradiated sodium borosilicate glass and simulated waste glass: a perturbed angular correlation spectroscopy study

Time Differential Perturbed Angular Correlation (TDPAC) has been used to investigate the local structural changes (if any) in Sodium borosilicate (NBS) Glass and Simulated Waste Oxide loaded sodium borosilicate glass (SNBS) used for immobilization of high-level liquid waste at Trombay, Mumbai. The glasses were doped with 181Hf tracer as TDPAC probe and irradiated with varying dose (1–50 MGy) of gamma radiation using 60Co (average γ energy 1.25 MeV) source. Hyperfine interaction parameters (quadrupole interaction frequency, its distribution and asymmetry of electric field gradient) were obtained for irradiated NBS and SNBS glasses from fitting of TDPAC data using DPAC software. The data was compared with the corresponding unirradiated glasses. Results do not show any micro-crystallinity upon irradiation in both the NBS and SNBS glass even at the highest gamma dose of 50 MGy even though small changes were observed in quadrupole interaction frequency at doses beyond 30 MGy. This suggests that the sodium borosilicate glass used for immobilization of high-level waste at Trombay is stable towards gamma radiation at least up to a gamma dose of 50 MGy.

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.

Rapid synthesis of Gd2Zr2O7 glass‐ceramics using spark plasma sintering

AbstractGlass‐ceramics are possible host matrix for high level waste immobilization. The Gd2Zr2O7 glass‐ceramic matrix was successfully synthesized using spark plasma sintering (SPS) method in 5 minutes. The phase transition with sintering temperature was studied using X‐ray diffraction, Raman and transmission electron microscopy. It revealed that samples kept a main defected fluorite phase as being sintered below 1800°C. Glass phase increased rapidly beyond 1850°C. The amorphous structure became the main body at 1900°C, with nanoscale crystal scattered in the bulk. With the increase of glass phase, the grain boundary became almost indistinguishable. The relationship between the final phase of Gd2Zr2O7 with its synthetic temperature range and corresponding technology was reviewed. Gd2Zr2O7 glass‐ceramics could be acquired by extending the sintering temperature beyond 1850°C using SPS method.

Effect of the Synthesis Atmosphere on the Process of Formation of Pyrochlore Matrices for HLW Immobilization

The synthesis of ceramic matrices, based on zirconium-enriched pyrochlore for the immobilization of high-level waste (HLW), by means of self-propagating high-temperature synthesis (SHS) under the pressure of argon and compressed air in a SHS-reactor was investigated. The dependences of the phase composition and characteristics of the crystalline phases of the material on the gas pressure were determined. It is shown that all rare-earth elements (REE) (Ce4+, La3+) from the model mixture of the HLW are incorporated into the structure of the ceramic. It is established based on evidence derived from elemental analysis that lanthanides are concentrated in the perovskite and pyrochlore structures.

Hydrolytic and thermal stability of magnesium potassium phosphate compound for immobilization of high level waste

The samples of the magnesium potassium phosphate (MPP) compound have been synthesized during solidification of high level waste (HLW) surrogate. The compound consists of a main phosphate phases Mg1.1Na0.35K0.45PO4 × (4–6)H2O and MgCs0.5Na0.2K0.3PO4 × (5–6)H2O. Differential leaching rates of 239Pu, 152Eu and 90Sr from the MPP compound after heat treatment (450 °C) are 7.8 × 10−9; 1.7 × 10−7 and 8.8 × 10−6 g cm−2 day−1, respectively. The coefficient of thermal expansion of MPP compound—(11.6 ± 0.3) × 10−6 °C−1; coefficient of thermal conductivity averages 0.5 W m−1 K−1. The properties of MPP compound meet the regulatory requirements for solidified HLW.

A carbon composite based on pyrolyzed diphthalocyanines for immobilization of high-level waste from nuclear industry

A new procedure for immobilization of high-level waste from spent nuclear fuel (SNF) of nuclear power plants (NPP) is suggested. The HLW immobilization process consists in the radionuclide transfer into the form of acetates, followed by synthesis of their diphthalocyanines and their subsequent pyrolysis in an inert medium. Real raffinate solutions from reprocessing of SNF of Novovoronezh NPP were used as HLW. The efficiency of the immobilization of radionuclides (mainly REE and minor actinides) was found to be higher than 99%. The thermal stability, chemical durability (leaching with aqueous solutions with pH from 2 to 9), and radiation resistance of the samples were evaluated using α-, β-, and γ-ray spectrometry. The evolution of the structure of the pyrolyzed diphthalocyanines in the temperature range 800–1600°C was studied by small-angle neutron scattering, X-ray diffraction analysis, and atomic force microscopy. The results obtained have confirmed the hypothesis that the pyrolysis of diphthalocyanines yields a branched network of carbon atom nanoclustes of size from 5 to ~150 nm, ensuring efficient retention of radionuclides in the host material. The host material is comparable in the thermal stability, i.e., in the degree of radionuclide retention at 1000–1200°С, with borosilicate or phosphate glasses, which are the most widely used host materials for radioactive waste immobilization, and considerably surpasses them in the resistance to leaching and radiation.

Oxidation state and local environment of iron and hydrolytic stability of multicomponent aluminum-iron-phosphate glasses for immobilization of high-level waste

The nuclear gamma resonance (Mossbauer effect) is used to determine the oxidation state and the coordination environment of iron ions in model glasses that simulate vitrified high-level waste. A major fraction of iron is present in the samples with relatively low contents of transition metal oxides as Fe(III) in the octahedral oxygen environment. When the content of transition metal oxides is high, iron is distributed between a glassy phase (10–15%) and a crystalline phase (85–90%). Iron in the first phase exists predominantly as Fe(III) in the octahedral environment, while in the second phase it is present as Fe(II) and Fe(III) ions as well in a form of octahedrally coordinated and participating ions in a “fast” electron exchange proceeding as mFe3+ + nFe2+ ↔ mFe2+ + nFe3+. The leach rate of Cs ions from boron-free glasses amounts to ~1 × 10–5 g/(cm2 day) and lower, and the value for Fe is three orders of magnitude lower. Upon annealing, both the rate and the degree of leaching increase by several times, while the boron-containing glasses exhibit lower hydrolytic durability. Both the rate and the degree of leaching of iron change insignificantly after annealing.