The catastrophic nuclear reactor accident at Fukushima damaged public confidence in nuclear energy and a demand for new engineered safety features that could mitigate or prevent radiation releases to the environment in the future. We have developed a novel use of sacrificial material (SM) to prevent the molten corium from breaching containment during accidents as well as a validated, novel, high-fidelity modeling capability to design and optimize the proposed concept. Some new reactor designs employ a core catcher and a SM, such as ceramic or concrete, to slow the molten corium and avoid the breach of the containment. However, existing reactors cannot easily be modified to include these SMs but could be modified to allow injectable cooling materials (current designs are limited to water). The SM proposed in this Laboratory Development Research and Development (LDRD) project is based on granular carbonate minerals that can be used in existing light water reactor plants. This new SM will induce an endothermic reaction to quickly freeze the corium in place, with minimal hydrogen explosion and maximum radionuclide retention. Because corium spreading is a complex process strongly influenced by coupled chemical reactions (with underlying containment material and especially with the proposed SM), decay heat and phase change. No existing tool is available for modeling such a complex process. This LDRD project focused on two research areas: experiments to demonstrate the feasibility of the novel SM concept, and modeling activities to determine the potential applications of the concept to actual nuclear plants. We have demonstrated small-scale to large-scaled experiments using lead oxide (Pb0) as surrogate for molten corium, which showed that the reaction of the SM with molten Pb0 results in a fast solidification of the melt and the formation of open pore structures in the solidified Pb0 because of CO2 released from the carbonate decomposition.
Fluid inclusions are found within mineral crystals or along grain boundaries in many sedimentary rocks, notably in evaporite formations, and can migrate along a thermal or hydro-mechanical gradient. Shale and salt rocks have been considered potential host rocks for radioactive waste disposal, due to their low permeability. Previously stagnant inclusions may become mobilised by a perturbation of the in situ state by a geotechnical installation or the emplacement of heat-generating waste. The migration of fluid inclusions can thus have important impacts on the long-term performance of a geologic repository for high-level radioactive waste disposal. As a part of the international research project DECOVALEX-2019, two aspects of fluid inclusion migration in rock salt are currently investigated under different boundary conditions: a) altered hydro-mechanical conditions as a consequence of tunnel excavation or borehole drilling and b) coupled thermo-hydro-mechanical-chemical conditions during the heating period of the post-closure phase of a repository. To obtain a mechanistic understanding of underlying physical processes for fluid inclusion migration, a multi-scale modelling strategy has been developed. Microscale hydraulic and time-dependent mechanical conditions related to the creep behaviour of rock salt are constrained by considering the macroscale stress evolution of an underground excavation. An analysis using a coupled two-phase flow and elasto-plastic model with a consideration of permeability variation indicates that a pathway dilation along the halite grain boundary may increase the permeability by two orders of magnitude. The calculated high flow velocity may explain the fast pressure build-up observed in the field. In addition, a mathematical model for the migration and morphological evolution of a single fluid inclusion under a thermal gradient has been formulated. A first-order analysis of the model leads to a simple mathematical expression that is able to explain the key observations of thermally driven inclusion migration in salt. Finally, numerical methods such as a phase field method for solving a moving boundary problem of fluid inclusion migration have also been explored.
The work for Step 1 performed at Sandia National Laboratories and reported in Section 7 has been updated to incorporate new data and to conduct new simulations using a new larger base case domain. The new simulations also include statistical analysis for different fracture realizations. A sensitivity analysis was also conducted to the study of the effect of domain size. A much larger mesh was selected to minimize boundary effects. The DFN model was upscaled to the new base case domain and the much larger domain to generate relevant permeability and porosity fields for each case. The calculations updated for Step 2 are described in Section 12.1. New calculations have also been conducted to model the flooding of the CTD and the resulting pressure recovery. The modeling includes matching of pressure and chloride experimental data at the six observation locations in Well 12M133. The modeling was done for the 10 fracture realizations. The Step 2 recovery simulations are described in Section 12.2. The Step 2 work is summarized in Section 12.3.
Understanding the viscosity and friction of a fluid under nanoconfinement is the key to nanofluidics research. Existing work on nanochannel flow enhancement has been focused on simple systems with only one to two fluids considered such as water flow in carbon nanotubes, and large slip lengths have been found to be the main factor for the massive flow enhancement. In this study, we use molecular dynamics simulations to study the fluid flow of a ternary mixture of octane-carbon dioxide-water confined within two muscovite and kerogen surfaces. The results indicate that, in a muscovite slit, supercritical CO2 (scCO2) and H2O both enhance the flow of octane due to (i) a decrease in the friction of octane with the muscovite wall because of the formation of thin layers of H2O and scCO2 near the surfaces; and (ii) a reduction in the viscosity of octane in nanoconfinement. Water reduces octane viscosity by weakening the interaction of octane with the muscovite surface, while scCO2 reduces octane viscosity by weakening both octane-octane and octane-surface interactions. In a kerogen slit, water does not play any significant role in changing the friction or viscosity of octane. In contrast, scCO2 reduces both the friction and the viscosity of octane, and the enhancement of octane flow is mainly caused by the reduction of viscosity. Our results highlight the importance of multicomponent interactions in nanoscale fluid transport. The results presented here also have a direct implication in enhanced oil recovery in unconventional reservoirs.
Various versions of deep borehole nuclear waste disposal have been proposed in the past in which effective sealing of a borehole after waste emplacement is generally required. In a high temperature disposal mode, the sealing function will be fulfilled by melting the ambient granitic rock with waste decay heat or an external heating source, creating a melt that will encapsulate waste containers or plug a portion of the borehole above a stack of the containers. However, there are certain drawbacks associated with natural materials, such as high melting temperatures, inefficient consolidation, slow crystallization kinetics, the resulting sealing materials generally being porous with low mechanical strength, insufficient adhesion to waste container surface, and lack of flexibility for engineering controls. In this study, we showed that natural granitic materials can be purposefully engineered through chemical modifications to enhance the sealing capability of the materials for deep borehole disposal. The present work systematically explores the effect of chemical modification and crystallinity (amorphous vs. crystalline) on the melting and crystallization processes of a granitic rock system. The approach can be applied to modify granites excavated from different geological sites. Several engineered granitic materials have been explored which possess significantly lower processing and densification temperatures than natural granites. Those new materials consolidate more efficiently by viscous flow and accelerated recrystallization without compromising their mechanical integrity and properties.
Montmorillonite with an empirical formula of Na0.2Ca0.1Al2Si4O10(OH)2(H2O)10 is a di-octahedral smectite. Montmorillonite-rich bentonite is a primary buffer candidate for high level nuclear waste (HLW) and used nuclear fuel to be disposed in mild environments. In such environments, temperatures are expected to be ≤ 90oC, the solutions are of low ionic strengths, and pH is close to neutral. Under the conditions outside the above parameters, the performance of montmorillonite-rich bentonite is deteriorated because of collapse of swelling particles as a result of illitization, and dissolution of the swelling clay minerals followed by precipitation of non-swelling minerals. It has been well known that tri-octahedral smectites such as saponite, with an ideal formula of Mg3(Si, Al)4O10(OH)2•4H2O for an Mg-end member (saponite-15A), are less susceptible to alteration under harsh conditions. Recently, Mg-bearing saponite has been favorably considered as a preferable engineered buffer material for the Swedish very deep holes (VDH) disposal concept in crystalline rock formations. In the VDH, HLW is disposed in deep holes at depth between 2,000 m and 4,000 m. At such deployment depths, the temperatures are expected to be between 100oC and 150oC, and the groundwater is of high ionic strength. The harsh chemical conditions of high pH are also introduced by the repository designs in which concretes and cements are used as plugs and buffers. In addition, harsh chemical conditions introduced by high ionic strength solutions are also present in repository designs in salt formations and sedimentary basins. For instance, the two brines associated with the salt formations for the Waste Isolation Pilot Plant (WIPP) in USA have ionic strengths of 5.82 mol•kg-1 (ERDA-6) and 8.26 mol•kg-1 (GWB). In the Asse site proposed for a geological repository in salt formations in Germany, the Q-brine has an ionic strength of ~13 mol•kg-1. In this work, we present our investigations regarding the stability of saponite under hydrothermal conditions in harsh environments.
Aqueous dissolution of silicate materials exhibits complex temporal evolution and rich pattern formations. Mechanistic understanding of this process is critical for the development of a predictive model for a long-term performance assessment of silicate glass as a waste form for high-level radioactive waste disposal. Here we provide a summary of a recently developed nonlinear dynamic model for silicate material degradation in an aqueous environment. This model is based on a simple self-organizational mechanism: dissolution of silica framework of a material is catalyzed by cations released from material degradation, which in turn accelerate the release of cations. This model provides a systematical prediction of the key features observed in silicate glass dissolution, including the occurrence of a sharp corrosion front, oscillatory dissolution, multiple stages of the alteration process, wavy dissolution fronts, growth rings, incoherent bandings of alteration products, and corrosion pitting. This work provides a new perspective for understanding silicate material degradation and evaluating the long-term performance of these materials as a waste form for radioactive waste disposal.
Uranyl ion, UO22+, and its aqueous complexes with organic and inorganic ligands, are the dominant species for transport of natural occurring uranium at the Earth surface environments. In the nuclear waste management, uranyl ion and its aqueous complexes are expected to be responsible for uranium mobilization in the disposal concepts where spent fuel is disposed in oxidized environments such as unsaturated zones relative to the underground water table. In the natural environments, oxalate, in fully deprotonated form, C2O42-, is ubiquitous, as oxalate is one of the most important degradation products of humic and fulvic acids. Oxalate is known to form aqueous complexes with uranyl ion to facilitate the transport of uranium. However, oxalate also forms solid phases with uranyl ion in certain environments, limiting the movement of uranium. Therefore, the knowledge of the stability constants of aqueous and solid uranyl oxalate complexes is important not only to the understanding of the mobility of uranium in natural environments, but also to the performance assessment of radionuclides in geological repositories for spent nuclear fuel. In this work, we present the stability constants for UO2C2O4(aq) and UO2(C2O4)22- at infinite dilution based on our evaluation of the literature data over a wide range of ionic strengths up to 9.5 mol•kg-1. We also obtain the solubility constants at infinite dilution for the following solid uranyl oxalates, UO2C2O4•3H2O and UO2C2O4•H2O, based on the solubility data in a wide range of ionic strengths up to 11 mol•kg-1. In our evaluation, we use the computer code EQ3/6 Version 8.0a. The model developed by us is expected to enable researchers to accurately assess the role of oxalate in mobilization/immobilization of uranium under various conditions including those in geological repositories.
