Waste Form Performance Research Articles

Synergy between Ca2+ and high ionic field‐strength cations during the corrosion of alkali aluminoborosilicate glasses in hyper‐alkaline media

AbstractOne major factor impeding the design of nuclear waste glasses with enhanced waste loadings is our insufficient understanding of their composition–structure–durability relationships, specifically in the environments the waste form is expected to encounter in a geological repository. In particular, the high field‐strength cations (HFSCs) are an integral component of most waste streams. However, their impact on the long‐term performance of the glassy waste form remains mostly undeciphered. In this context, the present study aims to understand the impact of some HFSCs (i.e., Nb5+, Zr4+, Ti4+, and La3+) on the dissolution behavior of alkali/alkaline‐earth aluminoborosilicate‐based model nuclear waste glasses in hyper‐alkaline media. At pH = 13, the studied glasses dissolve through the dissolution–reprecipitation mechanism, with Ca precipitation being the most vital step to passivation. In Ca‐free glasses, although the HFSCs slow down the forward rate, they do not seem to impact the residual rate behavior of glasses. The presence of Ca2+, however, initiates the rapid precipitation of network polymerizing HFSCs (i.e., Nb5+, Zr4+, and Ti4+) into a Ca2+/HFSCs‐based passivating layer, thus suggesting a synergy between Ca2+ and HFSCs that contributes to the enhanced long‐term durability of the glasses. Such synergy is not strongly evident for La3+, but instead, a potential La/Si affinity is observed upon the formation of the alteration layer.

Chemical durability and corrosion-induced microstructure evolution of compositionally complex titanate pyrochlore waste forms with uranium incorporation

Chemical durability of compositionally-complex rare-earth titanate pyrochlore solid solutions with or without uranium incorporation was investigated by semi-dynamic leaching testing. A Ti-enriched amorphous passivation film formed on the surface due to the surface alteration-reorganization mechanism with preferential release of the weakly oxygen-bonded rare earth elements. Elemental release was found to be strongly correlated with chemical disorder, with a negligible correlation with entropy. The release rate of the uranium gradually decreased with increasing numbers of components, implying a possibility of designing and optimizing waste form performance with enhanced chemical durability by controlling composition complexity and chemical disorder for effective nuclear waste management.

Corrosion rates measured under potentiostatic conditions to represent performance of a cermet nuclear waste form

ABSTRACT Electrochemical tests were conducted to assess the corrosion performance of a developmental multiphase cermet nuclear waste form in different conditions. Corrosion currents were monitored as the surface equilibrated in an acidic brine solution during potentiostatic tests at different imposed potentials. Electrochemical impedance spectroscopy was used to quantify differences in the electrical properties of surfaces stabilised at different potentials. The passivating oxides that will stabilise the surface under disposal conditions are formed and maintained during testing. The measured corrosion currents were two orders of magnitude lower for passivated surfaces than for polished surfaces over a wide range of redox conditions. The electrochemical results were correlated with microscopic analyses of the corroded surfaces and physical models of the surface passivation and degradation processes are proposed. The measured corrosion rates are suitable for the long-term performance modeling of cermet nuclear waste forms.

Effect of Ti‐metal addition on hot‐isostatically pressed (HIPed) Synroc‐C

AbstractSynroc, a candidate nuclear wasteform and Synroc technology, a waste treatment solution utilizing hot‐isostatic pressing (HIPing) have significant potential for the immobilisation of challenging nuclear wastes from both current and innovative reactors and fuel cycles. Hot isostatic press (HIP) consolidation is undertaken within sealed metal HIP canisters, where metal buffers (e.g., Ti, Fe and Ni) can be incorporated to control the redox environment within the canister. This study, for the first time, reports the effect of varying Ti‐metal addition (0, 2, 4, and 8 wt.%) on phase formation, microstructural characteristics, and wasteform performance for HIP consolidated Synroc‐C containing 20 wt.% simulated PUREX type (PW‐4b) high level waste. Quantitative X‐ray diffraction analysis, scanning electron microscopy‐energy dispersive X‐ray spectroscopy (EDS) and transmission electron microscopy‐EDS analyses were undertaken for analytical investigations. The chemical durability of the samples was assessed using ASTM C1220‐21 standard test. Hot‐isostatically pressed (HIPed) samples with 0 and 8 wt.% Ti added for redox control produced unfavourable phase formation. However, the HIPed samples with Ti additions of 2 and 4 wt.% as a redox buffer showed the desired phase formation of Synroc‐C without any significant change to the partitioning of waste elements among the phases along with compatible durability results, when compared to previous literature for hot uniaxial pressing (HUPed) or sintered materials.

Process Development of Zirconolite Ceramics for Pu Disposition: Use of a CuO Sintering Aid

Zirconolite-structured ceramics are candidate wasteform materials for the immobilisation of separated Pu. Due to the refractory properties of zirconolite and other titanates, removing residual porosity remains challenging in the final wasteform product when utilising a conventional solid state sintering route. Herein, we demonstrate that the addition of CuO as a sintering aid increases densification and promotes grain growth. Moreover, zirconolite phase formation was enhanced at lower process temperatures than typically required (≥1350 °C). CuO addition allowed an equivalent density to be reached using process temperatures of 250 °C lower than the undoped composition. At 150 °C lower than the undoped zirconolite, the addition of CuO resulted in a favourable microstructure and phase assemblage, as confirmed via X-ray diffraction and scanning electron microscopy. Secondary phases of CaTiO3 and Ca0.25Cu0.75TiO3 were observed at some processing temperatures, which may prove deleterious to wasteform performance. The use of a CuO sintering aid provides an avenue for the further development of the thermal processing of ceramic wasteform materials.

Immobilization of strontium and cesium by aluminosilicate ceramics derived from metakaolin geopolymer-zeolite A composites via 1100 °C heating treatment

In this work, K-based metakaolin geopolymer-zeolite A composites doped with strontium and cesium radionuclides have been subjected to 1100 °C to prepare aluminosilicate ceramic waste forms. Serial analysis of mineral characterization, physical properties and leaching resistance have been adopted to investigate the immobilization performance of aluminosilicate ceramic waste forms for strontium and cesium radionuclides. Compared to geopolymer-zeolite A composites, the obtained ceramic waste forms exhibit preferable mechanical property and leaching resistance (in deionized water, brine solutions and silicate solutions). The compressive strengths of aluminosilicate ceramic waste forms (Sr/Cs-loaded matrix) are both above 80 MPa, which are far higher than that of geopolymer-zeolite composites solidification matrices. Owing to the viscous sintering under 1100 °C heating treatment, even compact and dense ceramic waste forms have been successfully synthesized, accompanied by the main crystalline phase transformation in monoliths from zeolite A to leucite. Cs-leucite and Sr-feldspar also have been detected by XRD and TEM analyses, which are conducive to capturing strontium and cesium radionuclides in ceramic monoliths. All findings have proven that the ceramics waste forms derived from metakaolin geopolymer-zeolite A composites could be an ideal option for nuclear wastes immobilization.

Retention of strontium in high- & low-pH cementitious matrices – OPC vs. model systems

Strontium can be found in low/intermediate level radioactive waste streams, often solidified by Ordinary Portland Cement (OPC) matrices. Amorphous silica is added to improve wasteform performance, leading to low-pH systems. Our aim is to identify the phases controlling strontium immobilization in OPC matrices. The novelty of this work resides in combination between two approaches: 1) Comparing OPC systems to model systems based on alite, yielding matrices containing only major phases, emphasizing the relative role of minor phases; 2) Comparing high- & low-pH pastes. Pastes were characterized using calorimetry, XRD, DTG, SEM/EDS and porosimetry. Leaching experiments were conducted to study Sr retention and diffusivity. Strontium diffusivities in low-pH pastes are lower with respect to high-pH equivalents, despite their higher porosities, due to preferred Sr sorption to low Ca/Si C-S-H. OPC pastes show better Sr retention relative to model systems, suggesting that minor phases (like ettringite and AFm) improve Sr retention.

Toward Sustainable Cementitious Radioactive Waste Forms: Immobilization of Problematic Operational Wastes

Developing effective radioactive waste management practices is essential for ensuring the sustainability of the nuclear industry. The immobilization of radioactive wastes is one of the main activities conducted during the management of these wastes; it aims to produce a durable waste form that has sustainable performance over long periods of time. In this work, the challenges that face the design of durable cementitious waste forms are addressed for problematic operational wastes. In this respect, the problematic characteristics of evaporator concentrates, spent ion exchangers, and organic liquid wastes are overviewed, and the factors that affect the durability of their cementitious waste forms are identified. A summary of potential conventional and innovative cementitious matrices is presented by reviewing the cementation practices in national programs and recent research devoted to developing durable matrices. Finally, a guide to optimize the mix design of these waste forms was proposed that includes the selection of the testing procedure, factors that affect the waste form performance, and the optimization technique. This guide was presented with special focus on leaching tests, which are a means to test the stabilization performance of nuclear waste forms.

Competitive TcO4-, IO3-, and CrO42- Incorporation into Ettringite.

Ettringite is a naturally occurring mineral found in cementitious matrices that is known for its ability to incorporate environmentally mobile oxyanion contaminants. To better assess this immobilization mechanism for contaminants within cementitious waste forms intended for nuclear waste storage, this work explores how mixed oxyanion contaminants compete for ettringite incorporation and influence the evolving mineralogy. Ettringite was precipitated in the presence of TcO4-, IO3-, and/or CrO42-, known contaminants of concern to nuclear waste treatment, over pre-determined precipitation periods. Solution analyses quantified contaminant removal, and the collected solid was characterized using bulk and microprobe X-ray diffraction coupled with pair distribution function and microprobe X-ray fluorescence analyses. Results suggest that ≥96% IO3- is removed from solution, regardless of ettringite precipitation time or the presence of TcO4- or CrO42-. However, TcO4- removal remained <20%, was not significantly improved with longer ettringite precipitation times, and decreased to zero in the presence of IO3-. When IO3- is co-mingled with CrO42-, calcite and gypsum are formed as secondary mineral phases, which allows for oxyanion partitioning, e.g., IO3- incorporation into ettringite, and CrO42- incorporation into calcite. Results from this work exemplify the importance of competitive immobilization when assessing waste form performance and environmental risk of contaminant release.

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.

Geochemical modelling of the effect of waste degradation processes on the long-term performance of waste forms

The near field of the planned deep geological repository for low- and intermediate-level radioactive waste (L/ILW) in Switzerland will consist of different waste materials with different reactivities that are conditioned in cementitious matrices. Geochemical modelling was applied to predict the temporal evolution of the chemical conditions in a cement-stabilised model waste form by simulating the degradation of the waste materials and the alteration of the cementitious matrix. The model waste form contains large amounts of metallic waste and a low amount of organic waste. The geochemical modelling considered the use of either siliceous or calcareous aggregates for fabrication of the cementitious matrix used to condition the waste form. The results show that the type of aggregates has a major effect on the temporal evolution of the chemical conditions in the waste form. The use of calcareous aggregate instead of the commonly used siliceous aggregate is expected to maintain highly alkaline conditions of the waste form over the period of concern for the L/ILW repository. The latter conditions prevent accelerated iron corrosion and, related to that, accelerated H2 production from occurring due to the high pH. The evolution of the waste form is decisively controlled by the availability of water. In the absence of water ingress from the near field, the degradation processes cease as free water in the waste package is exhausted. In the case of saturation achieved by water ingress with time, the degradation processes continue over the entire period of concern for the L/ILW repository. The present study aims to illustrate the potential of geochemical modelling for predicting the temporal evolution of the chemical conditions of a cement-stabilised waste form.

Self-accelerated corrosion of nuclear waste forms at material interfaces.

The US plan for high-level nuclear waste includes the immobilization of long-lived radionuclides in glass or ceramic waste forms in stainless-steel canisters for disposal in deep geological repositories. Here we report that, under simulated repository conditions, corrosion could be significantly accelerated at the interfaces of different barrier materials, which has not been considered in the current safety and performance assessment models. Severe localized corrosion was found at the interfaces between stainless steel and a model nuclear waste glass and between stainless steel and a ceramic waste form. The accelerated corrosion can be attributed to changes of solution chemistry and local acidity/alkalinity within a confined space, which significantly alter the corrosion of both the waste-form materials and the metallic canisters. The corrosion that is accelerated by the interface interaction between dissimilar materials could profoundly impact the service life of the nuclear waste packages, which, therefore, should be carefully considered when evaluating the performance of waste forms and their packages. Moreover, compatible barriers should be selected to further optimize the performance of the geological repository system.

Nepheline crystallization behavior in simulated high-level waste glasses

The U.S. Department of Energy immobilizes legacy nuclear high-level waste (HLW) in borosilicate glass. HLW glass with elevated Na2O and Al2O3 concentrations have a demonstrated propensity for precipitating nepheline (NaAlSiO4) during cooling, which adversely affects the waste form performance. Currently, nepheline is suppressed by controlling the HLW glass composition, but this can limit the waste loading and result in longer mission life and cost. Correlation between composition and thermal history experienced during processing with time-temperature relationships that enable nepheline crystal growth would facilitate a more realistic prediction of nepheline crystallization in HLW glass and allow for a wider range of compositions to be accessed. In this work, crystallization behavior in simulated HLW glass was investigated to identify key attributes that lead to nepheline formation, with results indicating that the rate of diffusion for crystal growth may be controlled by glass structure connectivity.

Analysis of Krypton-85 Legacy Waste Forms: Part II

During a removal of legacy materials from one hot cell at the Idaho National Laboratory in 2010, five metal capsules and some loose zeolite material were identified as krypton (Kr) immobilization test specimens produced in the late 1970s under the Airborne Waste Management Program (AWMP). This AWMP research and development effort examined the encapsulation of 85Kr within a collapsed zeolite structure for use as a potential waste form for long-term storage. The recovered capsules appeared to have been placed to the side and remained untouched after the AWMP was halted in the mid-1980s. These reclaimed capsules and loose material presented a unique opportunity to study a potential 85Kr waste form after three half-lives had elapsed. The first phase of this study included two parts: The first was to assess the physical condition of the capsule walls, and the second was to examine the Kr-containing material within the capsules. The first part of this study was previously reported and noted that substantial corrosion was observed throughout each capsule wall of the two previously breached capsules that were examined. One of these capsules had been hot isostatic pressed (HIPed) and one was not HIPed. The second part of the study examined the materials contained in the two previously breached capsules. There appears to be a relatively uniform distribution of Kr and rubidium throughout the pellets examined. The chemical composition of the pellets appears to be consistent with 5A molecular sieves. The material contained within the HIPed capsules showed ~1 at. % lead (Pb). The origin of the Pb is currently indeterminate. X-ray diffraction analysis shows a significant shift from the 5A structure, most likely due to the Kr encapsulation/sintering process that occurred when the samples were made. Calculations based on the energy dispersive spectrometry elemental analysis show a residual Kr level within the pellets that is within a factor of 2 of the reported Kr capacities for this type of processed material. This provides a clear indication that a significant fraction of the Kr initially encapsulated in the material remained within the waste form even following a significant breach of the capsule wall. As a result, it would appear that this Kr immobilization method, even in non-HIPed form, is very promising as a waste form for long-term storage. The successful analysis of these two breached capsules forms a solid basis for the future analysis of the remaining unbreached capsules, which offers the opportunity to provide an even more complete understanding of the long-term Kr retention performance of this promising waste form.

Effects of heavy-ion irradiation on Gd2Zr2O7 bearing simulated TRPO waste

Considering the urgent demand of nuclear waste disposal, a comprehensive study on the irradiation performance of nuclear waste forms is extremely necessary, especially on those containing multiple-nuclides with multiple-valence. The irradiation effects of Gd2Zr2O7 pyrochlore bearing simulated trialkyl phosphine oxides (TRPO) waste (consisting of ZrO2, MoO3, RuO2 and PdO) were studied in this work. Gd10.685Mo4.734Zr28.924Ru1.000Pd1.745O91.825 and Gd8.883Mo9.901Zr33.760Ru2.089Pd3.644O118.368 were irradiated by 1.5 MeV Xe20+ to fluences from 1 × 1012 to 1 × 1015 ions/cm2 at room temperature. The results showed an irradiation induced disordering with slight micro-swelling. In addition, the ion irradiation tolerance of the solid solutions decreased as the TRPO waste content increased. The decrease and shift of Raman peaks indicated the structural disordering and inner stress after irradiation. The micromorphology and element distribution of the irradiated surface remained almost unchanged. The structural evolution from pyrochlore to fluorite and even amorphous structure were disclosed in this study.

Advanced Experimental Technique for Radiation Damage Effects in Nuclear Waste Forms: Neutron Total Scattering Analysis

ABSTRACTFor the past 30 years, the development of durable materials for radionuclide immobilization has been driven by efforts to dispose of wastes generated by the nuclear fuel cycle [National Research Council, ‘Waste Forms Technology and Performance: Final Report’, the National Academies Press, Washington D.C., 2011]. Many materials have been developed, but there still exist large gaps in the knowledge of fundamental modes of waste form degradation in repository environments. An important aspect of waste form science is the behavior of the materials under intense irradiation from decaying actinides and fission products. This irradiation induces a wide range of defects and disorder, the details of which depend on the specific waste form material. At the present time, it is not fully explained how radiation effects will influence the performance of nuclear waste forms and their long-term retention of fission products and actinides under operational conditions. The complex defect behavior and radiation damage must be understood over a range of length scales, from the initial atomic-scale defect structure to the long-range observable material modification. This is particularly challenging and requires advanced characterization techniques. This contribution describes how pair distribution function (PDF) analysis obtained from neutron total scattering experiments can be applied in the research field of waste form science to uniquely characterize radiation effects in a wide range of materials, including crystalline complex oxides and waste glasses. Neutron scattering strength does not have an explicit Z-dependence; this allows access to many low-Z elements, such as oxygen, that cannot be accurately studied with X-rays. In many cases, this can permit a detailed analysis of both cation (often high-Z) and anion (often low-Z) defect behavior. In contrast to traditional crystallography, which relies on long-range order, PDF analysis probes the local defect structure, including changes in site occupation, coordination, and bond distance. This is particularly important when characterizing aperiodic waste glasses with no long-range order at all. In contrast to X-ray characterization which requires very little sample mass (∼0.1 mg), neutron characterization (even at state-of-the-art spallation facilities) requires relatively large sample mass (∼50 - 100 mg). Obtaining this quantity is challenging for studies of irradiated materials, but by tailoring our experimental approach to use high-energy ions (GeV) with very high penetration depth, we are able to produce the required mass.

CEM V based special cementitious materials investigated by means of SANS method. Preliminary results

The management of the radioactive waste assume the conditioning in a cement matrix as an embedding, stable, disposal material. Cement matrix is the first and most important engineering barrier against the migration in the environment of the radionuclides contained in the waste packages. Knowing how the microstructure develops is therefore desirable in order to assess the compatibility of radioactive streams with cement and predict waste form performance during storage and disposal. For conditioning wastes containing radioactive aluminum new formulas of low basicity cements, using coatings as a barrier between the metal and the conditioning environment or introducing a corrosion inhibitor in the matrix system are required. Preliminary microstructure investigation of such improved CEM V based cement matrix is reported.

Studies of the radioactive waste confinement matrix using neutron scattering methods

The management of the radioactive waste implies the conditioning in a cement matrix as an embedding, stable, disposal material. Cement matrix is the first and most important engineering barrier against the migration in the environment of the radionuclides contained in the packages. Chemically, after mixing with water, cement produces a reacting matrix with a porous microstructure. Knowing how the microstructure develops is therefore desirable in order to assess the compatibility of radioactive streams with cement and predict waste form performance during storage and disposal. This article highlights the work performed in SANS investigation of a new type of cements to be used for conditioning of aluminium waste coming from reactor decommissioning activities. It is well known that the expansion of corrosion products formed on aluminium can lead to cracking of the waste form.

Mineral assemblage transformation of a metakaolin-based waste form after geopolymer encapsulation

Mitigation of hazardous and radioactive waste can be improved through conversion of existing waste to a more chemically stable and physically robust waste form. One option for waste conversion is the fluidized bed steam reforming (FBSR) process. The resulting FBSR granular material was encapsulated in a geopolymer matrix referred to here as Geo-7. This provides mechanical strength for ease in transport and disposal. However, it is necessary to understand the phase assemblage evolution as a result of geopolymer encapsulation. In this study, we examine the mineral assemblages formed during the synthesis of the multiphase ceramic waste form. The FBSR granular samples were created from waste simulant that was chemically adjusted to resemble Hanford tank waste. Another set of samples was created using Savannah River Site Tank 50 waste simulant in order to mimic a blend of waste collected from 68 Hanford tank. Waste form performance tests were conducted using the product consistency test (PCT), the Toxicity Characteristic Leaching Procedure (TCLP), and the single-pass flow-through (SPFT) test. X-ray diffraction analyses revealed the structure of a previously unreported NAS phase and indicate that monolith creation may lead to a reduction in crystallinity as compared to the primary FBSR granular product.

Cast Stone Formulation for Nuclear Waste Immobilization at Higher Sodium Concentrations

Low activity radioactive waste at U.S. Department of Energy sites can be immobilized for permanent disposal using cementitious waste forms. This study evaluated waste forms produced with simulated wastes at concentrations up to twice that of currently operating processes. The simulated materials were evaluated for their fresh properties, which determine processability, and cured properties, which determine waste form performance. The results show potential for greatly reducing the volume of material. Fresh properties were sufficient to allow for processing via current practices. Cured properties such as compressive strength meet disposal requirements. Leachability indices provide an indication of expected long‐term performance.