Immobilization Of High-level Nuclear Waste Research Articles

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.

Investigation of the leaching behavior of Na and Si in simulated HLW borosilicate glass obtained from the waste of a 1000 MWe class PWR reactor: using the response surface method.

The immobilization of high-level nuclear waste (HLW) in glass waste matrices provides the key safety function of slowing down radionuclide emissions from an underground disposal site. This study examines the leaching behavior of two major elements, Na and Si, in HLW borosilicate glass simulated from waste of a 1000 MWe class pressurized water reactor (PWR) using response surface methodology and Box-Behnken Design. The design of the experiment was carried out considering three independent variables: the pH of the solution, the contact time, and the leaching temperature, leading to 17 leaching runs performed using the static product consistency test (PCT). The results of statistical analysis (ANOVA: analysis of variance) indicated that the effects of the individual variables and the interactions between them were statistically significant, and the relative consistency of the data further confirmed the model's applicability. Data obtained from the PCT experiments revealed that the leaching behavior of Na and Si in the evaluated waste glass exhibited similar behavior to previously researched glasses for each condition tested.

Synthesis and characterization of sodium-aluminophosphate based glass-ceramics containing NZP phase for HLW immobilization

Stable sodium-aluminophosphate (SAP) based glass-ceramics with sodium zirconium phosphate (NZP) crystalline phase for high-level nuclear waste (HLW) immobilization were prepared by a melt-quenching process, and the effects of ZrO2 and B2O3 on their phase composition, structural features and properties were discussed in detail. The results show that the SAP based glass-ceramics obtained by the melt-quenching process show good aqueous durability and thermal stability. ZrO2 addition improves the formation of NZP phase but restrains the crystallization of AlPO4 phases. It is found that the correct ratios of ZrO2 and B2O3 allow only phase formation of NZP within the SAP glass, and the reasons were also analyzed. The conclusions demonstrate that the SAP based glass-ceramics with NZP phase as the crystalline phase are the potential matrices for specific HLW immobilization, and suggest that the melt-quenching process is an applicable way to prepare certain glass-ceramic waste forms.

Preparation and irradiation stability of A2B2O7 pyrochlore high-entropy ceramic for immobilization of high-level nuclear waste

The high-entropy pyrochlores have attracted increasing attention due to their ability to immobilize multiple nuclides simultaneously, compared with baseline single-component pyrochlores. In this work, nine kinds of multicomponent oxide ceramics are synthesized and eight of them exhibit a single pyrochlore phase. In addition, the ratio of rA to rB of high-entropy pyrochlore Gd2(Ti0.25Zr0.25Hf0.25Ce0.25)2O7 is 1.4499, this one observation suggests that for high-entropy pyrochlore, the rA/rB limit of pyrochlore phase is different from that of single-component pyrochlore (1.46 < rA/rB < 1.78). In order to evaluate the radiation resistance of high-entropy pyrochlores, four typical samples were irradiated by 800 keV Kr2+ ions beam. The GIXRD results show that all the high-entropy pyrochlores only undergo a phase transition from an ordered pyrochlore structure to a disordered fluorite structure (O-D transition) at the highest irradiation dose (∼ 40 dpa) in this study, instead of being amorphous. For (Sm0.5Gd0.5)2(Ti0.25Zr0.25Hf0.25Ce0.25)2O7 and (La0.2Nd0.2Sm0.2Eu0.2Gd0.2)2(Ti0.25Zr0.25Hf0.25Ce0.25)2O7, phase separation appears after irradiation. The results from the first principles calculations confirm that Gd2(Ti0.25Zr0.25Hf0.25Ce0.25)2O7 exhibits stronger radiation resistance than the corresponding single pyrochlores, which agrees with the irradiation experiment result. This study demonstrates that the high-entropy pyrochlores exhibit excellent irradiation stability and can be considered as the potential matrix material for immobilization of high-level nuclear waste.

Research Progress of Immobilization of High-Level Nuclear Waste by Using Phosphate Glass-Ceramics

Research Progress of Immobilization of High-Level Nuclear Waste by Using Phosphate Glass-Ceramics

Corrosion of ternary borosilicate glass in acidic solution studied in operando by fluid-cell Raman spectroscopy

Fluid-cell Raman spectroscopy is a space and time-resolving application allowing in operando studies of dynamic processes during solution–solid interactions. A currently heavily debated example is the corrosion mechanism of borosilicate glasses, which are the favoured material for the immobilization of high-level nuclear waste. With an upgraded fluid-cell lid design made entirely from the glass sample itself, we present the polymerization of the surface alteration layer over time in an initially acidic environment, including the differentiation between pore and surface-adsorbed water within it. Our results support an interface-coupled dissolution-precipitation model, which opposes traditional ion-exchange models for the corrosion mechanism. A sound description of the corrosion mechanism is essential for reliable numerical models to predict the corrosion rate of nuclear waste glasses during long-term storage in a geological repository.

Immobilization of a simulated HLW in phosphate based glasses/glass-ceramics by melt-quenching process

In this paper, monazite-iron phosphate based glass-ceramics for immobilization of a simulated high-level nuclear waste (HLW) derived from the composition of a nuclear waste in China were prepared through a melt-quenching process. The effect of HLW incorporation on phases, structural features, aqueous durability and properties of the as-prepared wasteforms were investigated in detail. The results show that the waste loading can be significantly elevated by forming monazite (Ce, La, Nd)PO4 containing glass-ceramic wasteforms through the melt-quenching process. The obtained glass/glass-ceramic wasteforms show good aqueous durability (∼10−4 g m−2 d−1) and thermal stability. Moreover, the formation of monazite (Ce, La, Nd)PO4 phase in the wasteforms is discussed. The conclusions suggest that it is feasible to synthesize (Ce, La, Nd)PO4-phosphate glass-ceramic wasteforms through melt-quenching process and monazite-iron phosphate based glass-ceramics are potential candidates for immobilization of the HLW.

Fabrication, chemical and thermal stability studies of crystalline ceramic wasteform based on oxyapatite phosphate host LaSr4(PO4)3O for high level nuclear waste immobilization

Reprocessed high-level nuclear waste (HLW) contains range of radioactive components. Crystalline oxyphosphate apatite ceramic of the formula LaSr4(PO4)3O [LSS] was investigated as a host for HLW immobilisation. The systematic study of solid solubility limit of individual rare earth ion substitution leads to the formulation of simulated wasteform of the formula La0.6Pr0.1Nd0.1Sm0.1Gd0.1Sr4(PO4)3O (WF1) with the waste loading of 17.95 wt% of rare-earth ions. Both parent and WF1 were synthesized by precipitation method. The thermal stress and groundwater inventory at the repository site can severely affect the wasteform performance, in addition to radiation and mechanical effects. Hence, the fabricated composition with high-level nuclear waste loading must be screened basically for chemical, thermal and radiation resistance. The present study investigated the thermal stability (by TGA), thermal expansion behaviour (by HT-XRD) and chemical durability (MCC-5) of the composition (WF1). The weight loss of WF1 being 2.2% and the average thermal expansion co-efficient (αavg) of 10.7 ± 1.2 × 10−6 K-1 in the temperature range (298–973 K) were comparatively lower than the parent phase, LaSr4(PO4)3O. The WF1 showed resistance to leaching of RE3+ and P5+ with only the leaching of Sr2+ ion whose leach rate was of the order 10-3-10-4 gm-2d-1.

Synthesis of pyrochlore-borosilicate glass-ceramics for immobilization of high-level nuclear waste

In this work, borosilicate based glass-ceramics with pyrochlore as crystalline phase for immobilization of high-level nuclear wastes (HLWs) were successfully synthesized by a one-step heat-treatment method. X-ray diffraction (XRD), scanning electron microscopy (SEM) and energy dispersive X-ray spectroscopy (EDXS) demonstrate that the obtained glass-ceramics show a regularly and uniformly distributed single pyrochlore (Ca,Na)(Nb,Ti)2Nd0.67O6F crystalline phase. Moreover, the glass-ceramics prepared show LRNa, LRB, LRAl, LRSi, LRNd, LRTi and LRNb of about 6.8 × 10−3, 3.7 × 10−4, 1.5 × 10−2, 2.2 × 10−3, 3.0 × 10−5, 5.1 × 10−5 and 5.5 × 10−6 g m−2 d−1 respectively after 28 leaching days, which are comparable to stable glass-ceramics for HLW immobilization. The results of this study are expected to provide an experimental reference for the engineering synthesis of glass-ceramics for the immobilization of certain HLWs.

The Effect of Heavy Ion Irradiation on the Forward Dissolution Rate of Borosilicate Glasses Studied in Situ and Real Time by Fluid-Cell Raman Spectroscopy.

Borosilicate glasses are the favored material for immobilization of high-level nuclear waste (HLW) from the reprocessing of spent fuel used in nuclear power plants. To assess the long-term stability of nuclear waste glasses, it is crucial to understand how self-irradiation affects the structural state of the glass and influences its dissolution behavior. In this study, we focus on the effect of heavy ion irradiation on the forward dissolution rate of a non-radioactive ternary borosilicate glass. To create extended radiation defects, the glass was subjected to heavy ion irradiation using 197Au ions that penetrated ~50 µm deep into the glass. The structural damage was characterized by Raman spectroscopy, revealing a significant depolymerization of the silicate and borate network in the irradiated glass and a reduction of the average boron coordination number. Real time, in situ fluid-cell Raman spectroscopic corrosion experiments were performed with the irradiated glass in a silica-undersaturated, 0.5 M NaHCO3 solution at temperatures between 80 and 85 °C (initial pH = 7.1). The time- and space-resolved in situ Raman data revealed a 3.7 ± 0.5 times increased forward dissolution rate for the irradiated glass compared to the non-irradiated glass, demonstrating a significant impact of irradiation-induced structural damage on the dissolution kinetics.

Real-time in situ observations of reaction and transport phenomena during silicate glass corrosion by fluid-cell Raman spectroscopy.

Borosilicate glass is an important material used in various industries due to its chemical durability, such as for the immobilization of high-level nuclear waste. However, it is susceptible to aqueous corrosion, recognizable by the formation of surface alteration layers (SALs). Here, we report in situ fluid-cell Raman spectroscopic experiments providing real-time insights into reaction and transport processes during the aqueous corrosion of a borosilicate glass. The formation of a several-micrometre-thick water-rich zone between the SAL and the glass, interpreted as an interface solution, is detected, as well as pH gradients at the glass surface and within the SAL. By replacing the solution with a deuterated solution, it is observed that water transport through the SAL is not rate-limiting. The data support an interface-coupled dissolution-reprecipitation process for SAL formation. Fluid-cell Raman spectroscopic experiments open up new avenues for studying solid-water reactions, with the ability to in situ trace specific sub-processes in real time by using stable isotopes.

Micron-sized spinel crystals in high level waste glass compositions: Determination of crystal size and crystal fraction

The compositions utilized for immobilization of high-level nuclear wastes (HLW) are controlled using glass property models to avoid the deleterious effects of crystallization in the high-level waste (HLW) vitrification melters. The type and size of the crystals that precipitate during melter operations (typically at 1150 °C) and idling (∼1000 °C) are significantly impacted by glass composition and thermal history. This study was conducted to measure the impact of melt composition and heat treatment temperature on crystal size and fraction. A matrix of 31 multi-component glasses canvasing the expected Hanford HLW compositional space was developed and the glasses fabricated and heat treated at 850, 900, and 950 °C. The crystal amounts, as determined by X-ray diffraction, varied from 0.2 to 41.0 wt%. Spinel concentrations ranged from 0.0 to 13.8 wt%. One glass of the matrix did not precipitate spinel and contained 0.2 wt% RuO2, which was assumed to be undissolved from the melting process. All compositions contained crystals in the as-quenched glass. All of the spinel based crystals present in the glasses were less than 10 μm in diameter, as determined by scanning electron microscopy with image analysis. Composition and temperature dependent models were generated using the resulting data and the best model fit was obtained by only considering spinel concentrations (R2 = 0.87). Two glasses were unable to be characterized because of an inability to process the glass under the conditions of this study. Those glasses were utilized to give insight into a potential multi-component constraint to aid in future statistical composition designs.

Treatment of Spent Nuclear Fuel for Separation, Immobilization and Disposal of Heat Generating High Level Fission Products, Caesium and Strontium

Separation of heat generating, high level fission product caesium and strontium from spent nuclear fuel boosts the capacity of waste repositories by as much as fifty times. For efficient use of already scarce repositories, separation of such fission products is mandatory. Separations of caesium and strontium using Chlorinated Cobalt Dicarbollide (CCD), PEG (Polyethylene Glycol), UNEX process and by Calixarenes or Fission Product Extraction Process (FPEX) were discussed. The UNEX method was then proposed as the most feasible method option. Following separation, nuclear waste immobilization techniques for such high-level fission product were discussed. The techniques included usage of concrete, glass and synthetic rock. Among them synthetic rock was identified as the most suitable one for immobilization of high-level nuclear waste. Finally, a safe disposal system with necessary required geology was shown for safe disposal of the waste.Journal of Chemical Engineering, Vol. 30, No. 1, 2017: 37-42

Effects of Ce+Gd on the structural features and thermal stability of iron-boron-phosphate glasses

In this paper, effect of Ce + Gd incorporation on the structural features and properties of iron borophosphate (IBP) glasses was investigated by X-ray diffraction (XRD), Fourier transform infrared spectroscopy (FTIR), Raman spectroscopy and differential scanning calorimetric (DSC) measurements. The results show that the solubility limit of Ce + Gd in the IBP glass with the composition of 36Fe2O3-10B2O3-54P2O5 is about 8 mol%. The structural networks of the Ce + Gd containing IBP glasses consist predominantly of orthophosphate, Q0, groups and pyrophosphate, Q1, groups, along with [BO4] units and a small amount of metaphosphate, Q2, groups. The Ce + Gd incorporation increases the density and Tg but decreases thermal stability of the IBP glass. However, the thermal stability of the Ce + Gd doped IBP glasses increases with Ce + Gd until the contents of Ce + Gd nearly reach the limit solubility. Moreover, the incorporation of Ce + Gd offers depolymerizing effect on the glass networks, which increases the fraction of Q0 and Q1 groups, and decreases the fraction of Q2 groups. The conclusions provide researchers with useful information to understand the roles of Ce + Gd in iron-boron-phosphate based glasses for high-level nuclear waste immobilization.

Materials Science of High-Level Nuclear Waste Immobilization

AbstractWith the increasing demand for the development of nuclear power comes the responsibility to address the issue of waste, including the technical challenges of immobilizing high-level nuclear wastes in stable solid forms for interim storage or disposition in geologic repositories. The immobilization of high-level nuclear wastes has been an active area of research and development for over 50 years. Borosilicate glasses and complex ceramic composites have been developed to meet many technical challenges and current needs, although regulatory issues, which vary widely from country to country, have yet to be resolved. Cooperative international programs to develop advanced proliferation-resistant nuclear technologies to close the nuclear fuel cycle and increase the efficiency of nuclear energy production might create new separation waste streams that could demand new concepts and materials for nuclear waste immobilization. This article reviews the current state-of-the-art understanding regarding the materials science of glasses and ceramics for the immobilization of highlevel nuclear waste and excess nuclear materials and discusses approaches to address new waste streams.

Metamict fergusonite-(Y) in a spessartine-bearing granitic pegmatite from Adamello, Italy

A granitic pegmatite associated with the Monte Bruffione granodiorite in the Tertiary Adamello Massif, Italy, consists primarily of albite, potassium feldspar, muscovite, quartz, and spessartine (graphic intergrowths with quartz). As accessory minerals, the pegmatite contains magnetite, pyrophanite, monazite-(Ce), uraninite, xenotime-(Y), zircon, and fergusonite-(Y). This yttrium niobate mineral contains inclusions of Th-rich uraninite, and is itself rich in UO 2 (average 7.1 ± 1.1 wt.%, n = 34) and ThO 2 (average 3.5 ± 1.0 wt.%). Many fergusonite crystals display growth zoning, characterized by a general increase towards the rim in the contents of Y and rare earth elements at the expense of U and Th. Irregular or patchy zoning as well as sector zoning are also observed in some of the crystals. Due to the alpha-decay of the U and Th fergusonite is metamict, as documented by transmission electron microscopy (TEM) and micro-Raman spectroscopy. Nevertheless, the mineral could be identified as fergusonite-(Y) on the basis of a canonical discriminant analysis of its chemical composition and on the basis of the close similarity of its Raman spectrum with that of a reference β-fergusonite. The crystalline-to-metamict transformation was associated with macroscopic swelling, as indicated by microfractures that are arranged radially around fergusonite inclusions in pyrophanite. The TEM data revealed that the amorphous fergusonite contains U-rich nanocrystals (5–15 nm across), which in most cases are distributed randomly and which probably nucleated after metamictization. The TEM investigations further revealed the ubiquitous presence of nano-sized (typically 5–25 nm across), nearly circular features, which exhibit a low diffraction contrast and which we interpret as nanopores. We propose that these nanopores represent former bubbles of radiogenic helium. The data presented here allowed us to determine that the critical amorphization dose of fergusonite is ≤ 0.97 × 10 16 alphas/mg, i.e., lower than that of other actinide-rich oxide minerals (e.g., pyrochlore). The presence of the U-rich nanocrystals indicates that fergusonite is able to retain actinides even when it is entirely metamict and even if it is surrounded by microfractures, which represent potential fluid pathways. This result suggests that fergusonite could be a phase suitable to be included in ceramics designed for the immobilization of high-level nuclear waste.

In Situ Radiation Damage Studies of Ca3Zr2FeAlSiO12 and Ca3Hf2FeAlSiO12

AbstractGarnets, A3B2C3O12, are considered to be potential host phases for the immobilization of high-level nuclear waste as they can accommodate a number of elements of interest, including Zr, Ti and Fe. The naturally occurring garnet, kimzeyite, Ca3(Zr,Ti)2(Si,Al,Fe)3O12, can contain ˜30wt% Zr. An understanding of the radiation tolerance of these materials is crucial to their potential use in nuclear waste immobilization. In this study two synthetic analogues of kimzeyite of composition Ca3Zr2FeAlSiO12 and Ca3Hf2FeAlSiO12 were monitored in situ during irradiation with 1.0 MeV Kr ions using the intermediate voltage electron microscope-Tandem User Facility (IVEM) at Argonne National Laboratory. The structure of these materials was previously determined by neutron diffraction and 57Fe Mössbauer spectroscopy. Ca3Zr2FeAlSiO12 and Ca3Hf2FeAlSiO12 have very similar structural properties with cubic Ia3d symmetry, the only significant difference being the presence of Zr and Hf, respectively, on the 6 coordinated B sites.

TEM Study of Short-Range-Order in Zirconolite Induced by High Temperature Ion Irradiation

Abstract Zirconolite (CaZrTi207) is an important phase proposed for high level nuclear waste immobilization. Zirconolite was irradiated by 1 MeV Kr+ at various temperatures. At room temperature, zirconolite became amorphous after a dose of 7x1014 ions/cm2.1 Amorphization dose increased with temperature due to thermal annealing. The critical temperature, above which amorphization does not occur, was estimated to be 654 K. During the low temperature irradiation (&amp;lt;654 K), concurrent with amorphization, zirconolite transformed from a monoclinic structure to the cubic pyrochlore structure and then to the fluorite substructure. The structural change is due to the disordering between cations and between oxygen and oxygen vacancies. After an irradiation at 673 K to a dose of 3.6x1015 ions/cm, the zirconolite samples remained crystalline. The diffraction pattern consists of strong maxima from the fluorite structure and diffuse maxima surrounding the Bragg positions of the pyrochlore superlattice (FIG. 1). Diffuse scattering patterns have been reported in other phases, and were generally attributed to the shortrange- order (SRO) domains.

High-temperature melter tests for vitrificationof bnfl high-level nuclear wastes

AbstractVitrification has become the international method of choice for the immobilization of highlevel nuclear wastes (HLW) from both commercial and defense fuel reprocessing activities. However, there are clear economic incentives for "next-generation" melters that might achieve increased waste loadings and higher throughput rates. In addition, BNFL is investigating potential immobilization processes for projected high-zirconium waste streams that would be generated from advanced nuclear fuel reprocessing processes. One approach towards addressing both of these challenges is waste vitrification at temperatures above the limits that are imposed on all currently operating HLW melters (around 1100 -1200°C). A unique high-temperature joule-heated melter that has been operating at the Vitreous State Laboratory for several years was used to conduct a matrix of tests designed to address these issues. The tests used simulants of two reprocessing waste streams: (i) A blend of "Magnox" and "Oxide" wastes, similar to those currently being vitrified at BNFL's Sellafield site; and (ii) The projected high-zirconium wastes. Each type of waste was investigated both as a wet slurry feed using chemical additives and as a dry calcined waste feed mixed with glass frit. Three target glass compositions were designed for these tests in order to permit processing over a wide range of temperatures. Data were collected on processing rates, off-gas composition, glass product composition and product quality, as well as on general processing behavior. In all cases, glass production rates increased rapidly with processing temperature, typically doubling for every 100°C, and decreased with increasing water content of the feed. The glass production rates per unit melt surface area were significantly higher than those typically observed for hot-wall induction melters operating under comparable conditions. Interestingly, off-gas emissions were much greater for the slurry feeds than for the comparable solid feeds. The resulting glass products were characterized for melt viscosity and electrical conductivity, leachability by both soxhlet and PCT procedures, glass transition temperature, and the amounts and types of any secondary phases that were present.

Radiation effects in crystalline ceramics for the immobilization of high-level nuclear waste and plutonium

This review provides a comprehensive evaluation of the state-of-knowledge of radiation effects in crystalline ceramics that may be used for the immobilization of high-level nuclear waste and plutonium. The current understanding of radiation damage processes, defect generation, microstructure development, theoretical methods, and experimental methods are reviewed. Fundamental scientific and technological issues that offer opportunities for research are identified. The most important issue is the need for an understanding of the radiation-induced structural changes at the atomic, microscopic, and macroscopic levels, and the effect of these changes on the release rates of radionuclides during corrosion.