Simulated Nuclear Waste Research Articles

Immobilization mechanism of Gd2Zr2O7 ceramic for nuclear waste treatment

Nuclear waste generated from the nuclear industry should be safely treated because of its harmful effects on the environment and humans. In this study, the mechanism of the as-prepared Gd2Zr2O7 for nuclear waste immobilization was investigated as an example in the simulation of waste immobilization under spark plasma sintering (SPS) conditions. The solid solubility limit of the waste in the sintered matrices was 45 mol% at 1725 °C and 1800 °C. The phase transformation from fluorite to pyrochlore could be detected owing to the increase in the ionic radius ratio when the waste was immobilized in the matrix structure. A high sintering temperature promotes transformation by increasing the orderliness of the matrix structure and destroying the crystalline structures to form an amorphous structure, results in a decrease in structural orderliness. Moreover, the leaching rates of Nd, at approximately 3.5 × 10−8 g·m−2·d−1 at 42 d, demonstrated high chemical stability in the matrices doping with 45 mol% simulated nuclear waste. A mechanism is proposed in this study for treating radioactive waste using Gd2Zr2O7 ceramic under SPS conditions.

Direct immobilization of simulated nuclear waste in preformed Gd2Zr2O7 pyrochlore via spark plasma sintering reaction

A rapid and efficient method is preferred for immobilizing nuclear wastes. In this study, Ce4+, a simulated tetravalent actinide (An4+) nuclear waste, was directly introduced into the Gd2Zr2O7 matrix (as-prepared) via spark plasma sintering (SPS) at different temperatures (1725, 1750, 1775, and 1800 °C) without a complicated lattice occupancy design. The solubility of CeO2 could be achieved to 18.8 wt%. Most of the sintered specimens exhibited the fluorite phase, but two samples sintered at 1775 °C exhibited the pyrochlore phase due to the Ce4+→Ce3+ transformation. Moreover, the samples sintered at various temperatures exhibited highly densified microstructures, in which Gd, Zr, and Ce were evenly distributed, indicating adequate immobilization results. By increasing the doping content to 60 mol%, the grain boundary of the sample sintered at 1800 °C became almost indistinguishable and exhibited slight vitrification, which was confirmed by transmission electron microscopy results. These results afforded a rapid and effective method for the immobilization of nuclear wastes in the as-prepared matrix without considering the lattice occupation.

Microstructural examination of interactions between chromia‐based refractory and nuclear glass in a melter

AbstractThis study shows the effects of nuclear waste glass production on Monofrax K‐3 refractory corrosion. A continuously fed research‐scale melter containing an Fe‐ and Ni‐rich simulated nuclear waste feed with borosilicate glass‐forming chemicals was cyclically melted at 1150°C and idled at 1050°C for a total of 11 weeks. Chemical maps using scanning electron microscopy show the interactions between the quenched melt and the refractory. Nanoscale X‐ray‐computed tomography was used for a three‐dimensional visualization of certain parts of the interface. Unreacted K‐3 consists of primarily corundum (Al,Cr)2O3 and spinel (Fe2+,Mg)(Al,Cr)2O4 interlocking crystalline phases. Some of the Cr from the refractory interacts with the Ni and Fe from the melt to form a reaction layer comprising (Ni,Fe2+)(Cr,Fe3+)2O4 spinel crystals. Simultaneously, melt components (Na,Si) infiltrate into the refractory. This interaction proceeds at the expense of the integrity of the refractory structure. Intact refractory grains (e.g., (Al,Cr)2O3) as well as the reaction layer itself can lose mechanical integrity and spall off into the melt, especially near the top of the melter. As the reaction layer can be a protective boundary for the refractory against further melt infiltration, a reduction in the reaction layer thickness allows an increase in refractory corrosion.

Structural change by phosphorus addition to borosilicate glass containing simulated waste components

Simulated nuclear waste glass samples containing phosphorus, which increase the solubility of molybdenum, were prepared and analyzed using synchrotron X-ray Absorption Fine Structure (XAFS) analysis for some constituent elements and Raman spectroscopic analysis of their complex structure. Changes in local structure and chemical state due to different phosphorus additions and waste loading rates were systematically studied. Consequently, no crystalline phase due to the molybdate compound was observed even at a maximum waste content of 30 wt% (corresponding to 1.87 mol% MoO3). Oxidation proceeded when the waste-loading rate was increased, whereas the reduction proceeded when phosphorus was added. In some cases, the effects of oxidation and reduction were offset. The local structure around specific elements can be classified as follows; Zn that is affected mainly by the waste-loading rate, Ce that is affected by both the waste-loading rate and phosphorus addition, and Zr element that is not affected by either of them. From the comparison between the analytical results of Mo and other elements, it was considered that the added phosphorus exists as a free PO4 structural unit and may deprive the alkali metal coordinated to the molybdate ion.

Direct single-step synthesis of phase pure zeolite Na–P1, hydroxy sodalite and analcime from coal fly ash and assessment of their Cs+ and Sr2+ removal efficiencies

Combustion by-products of coal from thermal power plants cause environmental pollution; efforts are made to develop techniques and green processes that could reduce the waste volume and convert the waste to useful materials. Synthesis of phase pure zeolites from coal fly ash and its use as adsorbent and ion exchangers to remove Cs+ and Sr2+ from simulated nuclear waste solutions is the focus of the current study. Zeolite Na–P1, hydroxy sodalite and analcime were synthesized from fly ash obtained from an Indian thermal power plant by using hydrothermal method. Zeolite synthesis conditions were optimized by studying the effect of concentration of NaOH solution, homogenization time, hydrothermal temperature and time. The zeolite phase formation, surface morphology, thermal stability studies were carried out by XRD, SEM, TG-DTA techniques; specific surface area was determined by BET analysis. Crystalline pristine zeolite Na–P1 was prepared from fly ash on reaction with 1 M NaOH at 150 °C. Similarly, hydroxy sodalite and analcime were also synthesized from the fly ash by hydrothermal reaction with 3 M NaOH at 150 °C and 5 M NaOH at 200 °C respectively. The removal efficiencies of synthesized zeolites were investigated with aqueous simulated nuclear waste solutions containing Cs+ and Sr2+. The adsorption capacities and cation exchange capacity (CEC) of zeolite Na–P1, hydroxy sodalite and analcime for Cs+ and Sr2+ ions were demonstrated. The results show the adsorption capacity of Na–P1 for Cs+ is 39.3 mg/g and for Sr2+ is 92.48 mg/g respectively and is a better adsorbent than analcime and hydroxy sodalite.

Conversion kinetics during melting of simulated nuclear waste glass feeds measured by dissolution of silica

We experimentally investigated the batch-to-glass conversion kinetics of two low-activity and two high-level nuclear waste melter feeds that represent widely different behavior during melting; tests comprised thermogravimetric analysis, the feed expansion test, evolved gas analysis, and x-ray diffraction. Our study indicates that the conversion processes, which include gas-evolving reactions, dissolution of crystalline phases, and primary foam expansion and collapse, are interconnected and influenced by the feed's thermal history. The fraction of dissolved silica was adopted as a measure of the extent of conversion inside the cold cap—the floating layer of feed material on the top surface of the melt in an electric melter. The interface between the cold cap and the free-flowing melt below was associated with a constant fraction of dissolved silica. The estimated viscosity of glass-forming melt at this interface lies within a narrow range for the feeds studied.

Synthesis of nano-crystalline zeolite-A and zeolite-X from Indian coal fly ash, its characterization and performance evaluation for the removal of Cs+ and Sr2+ from simulated nuclear waste

Phase pure zeolite-A and zeolite-X were synthesized using coal fly ash (CFA) obtained from Indian thermal power plants by employing alkali fusion method followed by hydrothermal technique. The fusion of fly ash with Na2CO3 was accomplished by heating at 800 °C/2 h by maintaining fly ash to Na2CO3 ratio at 1.2. The fused mass was found to be nepheline (Na4Al4Si4O16); and on subsequent treatment of the fused mass with 3 M NaOH under hydrothermal condition transformed to zeolite-A (Na12Al12Si12O48.27H2O) and zeolite-X (Na88Al88Si104O384.194H2O). The effluent solution from zeolite-A synthesis was utilized to prepare cancrinite. The zeolites were characterized by XRD, FTIR, TG-DTA, SEM and surface area of the powders were measured by BET technique. The specific surface area of the zeolite-A and zeolite-X were found to be 58.29 ± 0.20 and 164.34 ± 5.4 m2g−1 respectively. The TG-DTA studies showed the conversion of nano-crystalline to micro-crystalline zeolites with loss of adsorbed water. The ion exchange capacities of these nano-crystalline zeolites were evaluated by using simulated nuclear waste solutions containing Cs+or Sr2+ ions. The adsorption capacity of zeolite-A was found to be 95.74 mg/g and 54.12 mg/g respectively for Sr2+ and Cs+ions. Similarly, zeolite-X shows the adsorption capacity of 93.14 mg/g and 53.14 mg/g respectively for Sr2+ and Cs+ ions.

Immobilize CeO2 as simulated nuclear waste in natural magmatic granite: maximum solid solubility

Immobilize CeO2 as simulated nuclear waste in natural magmatic granite: maximum solid solubility

Thermal evolution effects on the properties of converting Cs-polluted soil into pollucite-base glass-ceramics for radioactive cesium immobilization

The Fukushima nuclear accident in Japan on March 11, 2011, produced large amounts of Cs-polluted soil which must be controlled to prevent the spread of hazardous Cs into the environment. In this paper, the effects of heat treatment on the structure and properties of Cs-containing glass-ceramics, as a simulated nuclear waste form, were systematically investigated. Cesium atoms are chemically bonded in the pollucite structure, and the amorphous phase further encapsulates the pollucite crystals in the glass-ceramics, thus providing an extra protective layer for the immobilized Cs. XRD analysis, Raman and FT-IR studies on the glass-ceramics synergistically indicated that the optimum crystallization temperature for pollucite is around 1000 °C. The pollucite is predominantly the main crystalline phase with a narrow crystal size distribution between 0.5 and 2 μm. Standard leaching test results show that the leaching rate of Cs was very low (3.0 × 10−3 g/(m2·d)). The study offers a practical method for immobilizing Cs in pollucite-base glass-ceramics. Moreover, the experimentally obtained data may provide some important references for converting Cs-polluted soil into pollucite-base glass-ceramics waste form.

Plasma spectroscopy + chemometrics: An ideal approach for the spectrochemical analysis of iron phosphate glass samples

AbstractThe penultimate aim of all analytical techniques is to provide best quantitative information with high sensitivity and accuracy. Such techniques then can be ideal to estimate the concentration of different species in a sample or to measure the surface concentration and so on. The preliminary step involved in quantitative analysis is evaluation of the system response for a given amount of the analyte under investigation. Different methodologies are proposed and executed for the above mentioned purpose. In this work, we have demonstrated the performance of chemometric methods to study the atomic emissions from simulated nuclear waste glasses. Five samples of iron‐phosphate glasses were synthesized by doping Cr, Sr, and Ti at various concentrations. The spectra recorded using laser‐induced breakdown spectroscopy (LIBS) system were analyzed using principal component analysis (PCA), partial least square‐discriminant analysis (PLS‐DA), and support vector machines (SVM) for the classification. As compared with the PLS‐DA, SVM provided better results for classification with accuracy of 100% on both calibration and validation sets, respectively. The quantitative analysis of glass samples is carried out using partial least square regression (PLSR) and support vector regression (SVR). The root mean squared error of prediction (RMSEP) is found to be 0.16, 0.20, and 0.08 (wt%) for Cr, Sr, and Ti using PLSR. The detailed investigation elucidates the advantages of chemometrics in handling the complex LIBS spectrum of glass samples and their significance in process control of nuclear waste, nuclear forensics, and so on.

Electrochemical reduction behavior of vitrified nuclear waste simulant in molten CaCl2

The electrochemical reduction of vitrified nuclear waste simulant glass produced by melting oxides of 33 different elements, including Si and four types of long-lived fission product elements (Cs, Zr, Pd, and Se), was investigated in molten CaCl2 at 1103 K. Before the experiment, potential−pO2− diagrams were constructed from the thermodynamic data to predict the behaviors of the constituent elements during the electrochemical reduction. In the first experiment, small crucible electrodes containing approximately 100 mg of the simulant were electrolyzed at 0.6–1.1 V vs. Ca2+/Ca. X-ray diffraction analysis confirmed that SiO2 was reduced to Si. In addition, the formation of XB6 (X = Ca, Si, or rare earth elements (REEs)) was confirmed. In the second experiment, a large crucible electrode containing approximately 10 g of the simulant was galvanostatically reduced at −2 A for 5 h. The distribution ratios of the constituent elements to the solid product and to the molten salt were calculated from the result of composition analysis using inductively coupled plasma atomic emission spectrometry and mass spectrometry. In particular, four types of long-lived fission product elements (Cs, Zr, Pd, and Se) were investigated. Consequently, alkali and alkaline earth elements, including Cs, were dissolved into the molten salt. Si mostly remained in the solid product as elemental silicon. Most of the other elements, including Zr and Pd, were also found in the solid product. Se and Zn were indicated to partially evaporate depending on the duration of electrolysis.

Partitioning of rare earths in multiphase nuclear waste glass‐ceramics

AbstractAlumino‐borosilicate glass‐ceramics (GCs) containing alkali (Na), alkaline earth (Ca), rare earth (RE) (one of La … Lu, Y), and transition metals (Zr and Mo) were formulated to test the effects of composition and RE cation size on the crystallization of simulated nuclear waste GCs. Eight‐oxide glasses were formulated in the peraluminous ([Na2Oexcess]=[Na2O]‐[Al2O3]‐[ZrO2]<0) region with 10 mol.% RE2O3 and 15 mol.% B2O3. Glasses were melted and quenched, reheated, and slow‐cooled to promote crystallization, and then characterized using quantitative X‐ray diffraction and electron probe microanalysis. Transition metal phases [powellite (Ca,Na,RE)MoO4 and zirconia (Zr,RE)O2 or zircon (Zr,RE)SiO4] always formed, while borosilicate (RE3BSi2O10) or silicate [oxyapatite (RE,Ca,Na)10Si6O26 or keiviite (RE2Si2O7)] plus borate (REBO3) phases also formed, depending on the size of the RE cation and hence its preferred coordination number. All crystalline phases were found to contain RE elements, and often phases incorporated other cations (Zr, Ca, B) which were not nominally part of their pure stoichiometry. These results suggested that the glass formulation and RE size strongly influenced the RE environment in the glass and hence the crystal phases formed. The formation of RE borosilicate, versus separate RE borate and RE silicate phases, also suggested original differences in the glass structure.

Densifications and mechanical properties of single-phase Gd2Zr2O7 ceramic waste forms with improved TRPO waste load

A nitrate–citrate combustion method combined with microwave sintering was firstly employed for the rapid fabrication of the Gd2Zr2O7 matrix immobilizing various amounts of simulated nuclear wastes. Phase evolutions, microstructure changes, element distributions, densification processes and mechanical properties of the as-prepared (1 – x)Gd2Zr2O7·xTRPO (0.0 ≤ x ≤ 0.6) at various temperatures were investigated. Compared to the reported studies, we have increased the immobilization amount of simulated TRPO waste within a crystal structure from 45 to 60 wt%. Td and Tg (the threshold temperatures to trigger accelerated densification and grain growth, respectively) were employed to divide the densification process into three stages. the mechanical properties and the densification stages of the (1 – x)Gd2Zr2O7·xTRPO (0.0 ≤ x ≤ 0.6) ceramics sintered under microwave sintering at various temperatures were finally determined. Fine-grained (1 – x)Gd2Zr2O7·xTRPO (0.0 ≤ x ≤ 0.6) ceramic waste forms with average grain size less than 200 nm and relative density higher than 90% can be obtained by microwave sintering at sintering temperature less than 1400 °C.

Performance of Some Ion-Exchange Resins in Removal of 241Am(III), 152+154Eu(III), 99Mo(VI), 137Cs(I), and 60Co(II) from Simulated Nuclear Acidic Solutions

Separation of 241Am(III), 152+154Eu(III), and 60Co(II) radionuclides from nitric acid solutions by the batch method using KU-2, TRU, and Amberlite® IR-120(H) resins was studied. The separation factors (SFs) of radionuclides are affected by the HNO3 concentration and resin type. In 3 M HNO3, the SF of 152+154Eu from 60Co and 241Am with the KU-2 resin was 628 and 26.48, respectively. In 0.01 M HNO3, the SF of 241Am from 60Co with the TRU resin reached 83. All the resins used showed good radiation resistance and stability up to 100 kGy. Simulated nuclear liquid waste containing minor actinides (241Am), lanthanides (152+154Eu), fission products (137Cs and 99Mo), and activation products (60Co) was successfully treated using column chromatography with the KU-2 resin. The radionuclides were stripped using 1.5 M aqueous HNO3 with 70–82% degree of stripping in one stage. Thus, the KU-2 and TRU resins can be used for treatment of the high level corrosion products containing trivalent long-lived minor actinides (5d elements, e.g., Am and Cm), lanthanides (4f elements) and/or divalent heavy metals (3d elements), associated with a nuclear power plant.

Effective decontamination of 99TcO4−/ReO4− from Hanford low-activity waste by functionalized graphene oxide–chitosan sponges

Treatment and recycling of radioactive waste are major challenges for the nuclear energy sector. Pertechnetate anions, 99TcO4−, occur in nuclear waste and are formed mainly during nuclear fission. 99TcO4− is a very hazardous contaminant due to its long half-life and high solubility and mobility in water. The effective capture of 99Tc contamination at legacy nuclear sites remains a serious and unsolved environmental issue. In this work, 99TcO4−/ReO4− was adsorbed by compressible graphene–chitosan sponges with grafted tertiary amine groups. Moreover, the hydrophobic branched chains strongly reduced the interference of competitive anions during adsorption. Results show that 99TcO4−/ReO4− can be effectively removed even under highly acidic conditions and in the presence of competing anions. More than 98% of 99TcO4−/ReO4− was removed using simulated nuclear wastes.

In situ characterization of foam morphology during melting of simulated waste glass using x-ray computed tomography

During batch-to-glass conversion, a glass-forming melt connects, creating a foam layer between the batch and the glass melt. Due to its transience and opacity, investigation of this foam layer presents a formidable challenge. In this work, we use in situ x-ray computed tomography to characterize the foam morphology that evolves during batch-to-glass conversion of a simulated nuclear waste glass. Rapid 1-min scans with 38 μm voxels were performed to capture the foam structure during heating. Geometric volume, total porosity, and bubble size distribution are reported. Using evolved gas analysis and combining the temperature-dependent melt viscosity with x-ray data, we describe the evolution of foam structure during the foam growth and subsequent collapse.

The dissolution of simulant UK Ca/Zn-modified nuclear waste glass: Insight into Stage III behavior

ABSTRACTThe dissolution of the United Kingdom’s vitrified high-level-waste simulant, CaZn MW28, was investigated following the Product Consistency Test-B protocol for 112 d at 90 °C and in ultra-high-quality water. Residual rate dissolution (stage II) and rate resumption (stage III), after 28 d, was observed. Thermodynamic modelling suggested that solutions were saturated with respect to Mg- and Zn-bearing phases, and the presence of Mg- and Zn-smectite clays was tentatively observed. The formation of these phases was concurrent with a significant increase in the dissolution rate, similar to Stage III behavior seen in other nuclear waste simulant glass materials, indicating that the addition of Mg and Zn to high-level-waste glass (7.3 wt. % combined) significantly influences the dissolution rate.

Preparation and Study of Corrosion Behavior of Some Simulated Radioactive Waste Glasses

To safe environment from radioactive waste it is important to fix them as radioactive waste glasses. The corrosion behavior of radioactive waste glasses in water is significantly important. Radionuclides return to the biosphere by means of leaching from waste form into ground water. Finally the ground water containing the radionuclide are transported to the surface. In this study, the preparation, characterization and leaching behavior of some borosilicate (BS) and lead iron phosphate (LIP) of different chemical composition doped with simulated nuclear waste oxide were investigated. We measured the pH found to be in the range from 6.78 up to 7.79 of the leachate solution at normal temperature and at varying time intervals. Leaching study of these glasses were conducted with the help of Soxhlet distillation apparatus with distilled water upto 24 hours and for BS9 - BS12 upto 100 hours duration. Weight losses were are measured with respect to time of leaching. Leach rate of some borosilicate glass samples loaded with uranium are calculated from surface area measurements. The results are reported in the range 1.34x10-4 g.m-2.hr-1 and 6.26 x 10-4 g.m-2.hr-1 respectively at 90°C.

Nonlinear logistic regression mixture experiment modeling for binary data using dimensionally reduced components

AbstractThis article motivates, presents, and illustrates an approach using nonlinear logistic regression (NLR) for modeling binary response data from a mixture experiment when the components can be partitioned into groups used to form dimensionally reduced components (DRCs). A DRC is formed from a linear combination of the components in a group having similar roles and/or effects of the same sign, where the linear combinations over all groups are normalized so that the DRC proportions sum to one. Linear combinations of a particular form provide for quantifying the effects of the remaining components in a group relative to a chosen component. This reason, plus dimensional reduction, are the primary motivations for the proposed DRC mixture experiment modeling approach. NLR is required because models expressed in terms of the DRCs are nonlinear in the parameters that specify the linear combinations. A method for obtaining nonparametric tolerance limits on the probability of a “success” for the binary response variable using a bootstrap approach is also presented. Finally, the article shows how DRCs provide for visualizing data and modeling results that otherwise would be impossible. A real database on the presence or absence of nepheline crystals in simulated nuclear waste glass is used to illustrate the NLR modeling and nonparametric tolerance limit approaches. The methodology is general and can be applied to other applications.

Titanium-doped iron phosphate based glass ceramic waste forms containing 50 wt% simulated nuclear waste

50 wt% of the simulated nuclear waste derived from the high level waste (HLW) in China is immobilized by using titanium-doped iron phosphate based glass ceramics. The effect of TiO2 content on the crystalline phases, structure and aqueous dissolution rate of the glass ceramic waste forms was investigated in detail. The results show that the crystalline phases of the obtained waste forms are monazite (Ce, La)PO4 accompanying with a small concentration of zirconium pyrophosphate ZrP2O7. The crystallization of ZrP2O7 phase is suppressed by adding TiO2 in the phosphate baseline glasses. Meanwhile, smaller grain size of the products is obtained. Moreover, the aqueous dissolution rates of the glass ceramic waste forms keep in the order of magnitude of 10−2 g m−2 d−1 after 14 leaching days, indicating their good chemical durability which are comparable to widely used borosilicate glasses and iron phosphate glasses/waste forms. The conclusions suggest that phosphate-based glass ceramics may potentially be an applicable matrix for the disposal of HLWs which are rich in actinide elements.