Publications

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This report summarizes the 2019 fiscal year (FY19) status of the borehole heater test in salt funded by the US Department of Energy Office of Nuclear Energy (DOE-NE) Spent Fuel and Waste Science & Technology (SFWST) campaign. This report satisfies SFWST level-three milestone report M3SF-19SN010303033. This report is an update of the April 2019 level-two milestone report M2SF-19SNO10303031 to reflect the nearly complete as-built status of the borehole heater test. This report discusses the fiscal year 2019 (FY19) design, implementation, and preliminary data interpretation plan for a set of borehole heater tests call the brine availability tests in salt (BATS), which is funded by the DOE Office of Nuclear Energy (DOE-NE) at the Waste Isolation Pilot Plant (WIPP), a DOE Office of Environmental Management (DOE-EM) site. The organization of BATS is outlined in Project Plan: Salt In-Situ Heater Test (SNL, 2018). An early design of the field test is laid out in Kuhlman et al. (2017), including extensive references to previous field tests, which illustrates aspects of the present test. The previous test plan by Stauffer et al. (2015) places BATS in the context of a multi-year testing strategy, which involves tests of multiple scales and processes, eventually culminating in a drift-scale disposal demonstration. This level-3 milestone report is an update of a level-2 milestone report from April 2019 by the same name. The update adds as-built details of the heater test, which at the time of writing (August 2019) is near complete implementation.

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This report discusses the fiscal year 2019 (FY19) design, implementation, and preliminary data interpretation plan for a set of borehole heater tests call the brine availability tests in salt (BATS), which is funded by the DOE Office of Nuclear Energy (DOE-NE) at the Waste Isolation Pilot Plant (WIPP). The organization of BATS is outlined in Project Plan: Salt In-Situ Heater Test. An early design of the field test is laid out in Kuhlman et al., including extensive references to previous field tests, which illustrates aspects of the present test. The previous test plan by Stauffer et al., places BATS in the context of a multi-year testing strategy, which involves tests of multiple scales and processes, possibly culminating in a drift-scale disposal demonstration.

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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.

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.

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In the published article (Xiong et al., 2017), there was an error for the reaction coefficient for the dissolution reaction of Ca2C10H12N2O8•7H2O(s) in the database used for the modeling. In the database for the modeling, the coefficient for water (i.e., 7H2O) was inadvertently omitted. Because of this omission, the results in Table 3 were affected. The authors wish to make the corrections to Table 3. The corrected values are tabulated in the revised Table 3. The corrected values reproduce the experimental data in MgCl2 solutions much better (see revised Fig. 5).

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Methane (CH4) and carbon dioxide (CO2), the two major components generated from kerogen maturation, are stored dominantly in nanometer-sized pores in shale matrix as (1) a compressed gas, (2) an adsorbed surface species and/or (3) a species dissolved in pore water (H2O). In addition, supercritical CO2 has been proposed as a fracturing fluid for simultaneous enhanced oil/gas recovery (EOR) and carbon sequestration. A mechanistic understanding of CH4-CO2-H2O interactions in shale nanopores is critical for designing effective operational processes. Using molecular simulations, we show that kerogen preferentially retains CO2 over CH4 and that the majority of CO2 either generated during kerogen maturation or injected in EOR will remain trapped in the kerogen matrix. The trapped CO2 may be released only if the reservoir pressure drops below the supercritical CO2 pressure. When water is present in the kerogen matrix, it may block CH4 release. However, the addition of CO2 may enhance CH4 release because CO2 can diffuse through water and exchange for adsorbed methane in the kerogen nanopores.

In this study, solubility measurements regarding boracite [Mg3B7O13Cl(cr)] and aksaite [MgB6O7(OH)6·2H2O(cr)] from the direction of supersaturation were conducted at 22.5 ± 0.5 °C. The equilibrium constant (log10K0) for boracite in terms of the following reaction, Mg3B7O13Clcr+15H2Ol⇌3Mg2++7BOH4 −+Cl−+2H+ is determined as −29.49 ± 0.39 (2σ) in this study. The equilibrium constant for aksaite according to the following reaction, MgB6O7OH6·2H2Ocr+9H2Ol⇌Mg2++6BOH4 −+4H+ is determined as −44.41 ± 0.41 (2σ) in this work. This work recommends a set of thermodynamic properties for aksaite at 25 °C and 1 bar as follows: ΔHf 0 = −6063.70 ± 4.85 kJ·mol−1, ΔGf 0 = −5492.55 ± 2.32 kJ·mol−1, and S0 = 344.62 ± 1.85 J·mol−1·K−1. Among them, ΔGf 0 is derived from the equilibrium constant for aksaite determined by this study; ΔHf 0 is from the literature, determined by calorimetry; and S0 is computed in the present work from ΔGf 0 and ΔHf 0. This investigation also recommends a set of thermodynamic properties for boracite at 25 °C and 1 bar as follows: ΔHf 0 = −6575.02 ± 2.25 kJ·mol−1, ΔGf 0 = −6178.35 ± 2.25 kJ·mol−1, and S0 = 253.6 ± 0.5 J·mol−1·K−1. Among them, ΔGf 0 is derived from the equilibrium constant for boracite determined by this study; S0 is from the literature, determined by calorimetry; and ΔHf 0 is computed in this work from ΔGf 0 and S0. The thermodynamic properties determined in this study can find applications in many fields. For instance, in the field of material science, boracite has many useful properties including ferroelectric and ferroelastic properties. The equilibrium constant of boracite determined in this work will provide guidance for economic synthesis of boracite in an aqueous medium. Similarly, in the field of nuclear waste management, iodide boracite [Mg3B7O13I(cr)] is proposed as a waste form for radioactive 129I. Therefore, the solubility constant for chloride boracite [Mg3B7O13Cl(cr)] will provide the guidance for the performance of iodide boracite in geological repositories. Boracite/aksaite themselves in geological repositories in salt formations may be solubility-controlling phase(s) for borate. Consequently, solubility constants of boracite and aksaite will enable researchers to predict borate concentrations in equilibrium with boracite/aksaite in salt formations.

In this work, a solubility study on brucite [Mg(OH)2(cr)] in Na2SO4 solutions ranging from 0.01 to 1.8 mol·kg−1, with 0.001 mol·kg−1 borate, has been conducted at 22.5 °C. Based on the solubility data, the Pitzer interaction parameters for MgB(OH)4 + − SO4 2− and MgB(OH)4 + − Na+ along with the formation constant for MgSO4(aq) are evaluated using the Pitzer model. The formation constant (log10β10 = 2.38 ± 0.08) for MgSO4(aq) at 25 °C and infinite dilution obtained in this study is in excellent agreement with the literature values. The experimental data on the solubility of gypsum (CaSO4·2H2O), at 25 °C, in aqueous solutions of MgSO4 with ionic strengths up to ~ 11 mol·kg−1 were analyzed using models with and without considering the MgSO4(aq) species. The model incorporating MgSO4(aq) fits better to the experimental data than the model without MgSO4(aq), especially in the ionic strength range beyond ~ 4 mol·kg−1, demonstrating the need for incorporation of MgSO4(aq) into the model to improve the accuracy.

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Radionuclides and heavy metals easily sorb onto colloids. This phenomenon can have a beneficial impact on environmental clean-up activities if one is trying to scavenge hazardous elements from soil for example. On the other hand, it can have a negative impact in cases where one is trying to immobilize these hazardous elements and keep them isolated from the public. Such is the case in the field of radioactive waste disposal. Colloids in the aqueous phase in a radioactive waste repository could facilitate transport of contaminants including radioactive nuclides. Salt formations have been recommended for nuclear waste isolation since the 1950's by the U.S. National Academy of Science. In this capacity, salt formations are ideal for isolation of radioactive waste. However, salt formations contain brine (the aqueous phase), and colloids could possibly be present. If present in the brines associated with salt formations, colloids are highly relevant to the isolation safety concept for radioactive waste. The Waste Isolation Pilot Plant (WIPP) in southeast New Mexico is a premier example where a salt formation is being used as the primary isolation barrier for radioactive waste. WIPP is a U.S. Department of Energy geological repository for the permanent disposal of defenserelated transuranic (TRU) waste. In addition to the geological barrier that the bedded salt formation provides, an engineered barrier of MgO added to the disposal rooms is used in WIPP. Industrial-grade MgO, consisting mainly of the mineral periclase, is in fact the only engineered barrier certified by the U.S. Environmental Protection Agency (EPA) for emplacement in the WIPP. Of interest, an Mg(OH)2-based engineered barrier consisting mainly of the mineral brucite is to be employed in the Asse repository in Germany. The Asse repository is located in a domal salt formation and is another example of using salt formations for disposal of radioactive waste. Should colloids be present in salt formations, they would facilitate transport of contaminants including actinides. In the case of colloids derived from emplaced MgO, it is the hydration and carbonation products that are of interest. These colloids could possibly form under conditions relevant in particular to the WIPP. In this chapter, we report a systematic experimental study performed at Sandia National Laboratories in Carlsbad, New Mexico, related to the WIPP engineered barrier, MgO. The aim of this work is to confirm the presence or absence of mineral fragment colloids related to MgO in high ionic strength solutions (brines). The results from such a study provides information about the stability of colloids in high ionic strength solutions in general, not just for the WIPP. We evaluated the possible formation of mineral fragment colloids using two approaches. The first approach is an analysis of long-term MgO hydration and carbonation experiments performed at Sandia National Laboratories (SNL) as a function of equivalent pore sizes. The MgO hydration products include Mg(OH)2 (brucite) and Mg3 Cl(OH)5•4H2O (phase 5), and the carbonation product includes Mg5(CO3)4(OH)2•4H2O (hydromagnesite). All these phases contain magnesium. Therefore, if mineral fragment colloids of these hydration and carbonation products were formed in the SNL experiments mentioned above, magnesium concentrations in the filtrate from the experiments would show a dependence on ultrafiltration. In other words, there would be a decrease in magnesium concentrations as a function of ultrafiltration with decreasing molecular weight (MW) cut-offs for the filtration. Therefore, we performed ultrafiltration on solution samples from the SNL hydration and carbonation experiments as a function of equivalent pore size. We filtered solutions using a series of MW cut-off filters at 100 kD, 50 kD, 30 kD and 10 kD. Our results demonstrate that the magnesium concentrations remain constant with decreasing MW cutoffs, implying the absence of mineral fragment colloids. The second approach uses spiked Cs+ to indicate the possible presence of mineral fragment colloids. Cs+ is easily absorbed by colloids. Therefore, we added Cs+ to a subset of SNL MgO hydration and carbonation experiments. Again, we filtered the solutions with a series of MW cut-off filters at 100 kD, 50 kD, 30 kD and 10 kD. This time we measured the concentrations of Cs. The concentrations of Cs do not change as a function of MW cut-offs, indicating the absence of colloids from MgO hydration and carbonation products. Therefore, both approaches demonstrate the absence of mineral fragment colloids from MgO hydration and carbonation products. Based on our experimental results, we acknowledge that mineral fragment colloids were not formed in the SNL MgO hydration and carbonation experiments, and we further conclude that high ionic strength solutions associated with salt formations prevent the formation of mineral fragment colloids. This is due to the fact that the high ionic strength solutions associated with salt formations have high concentrations of both monovalent and divalent metal ions that are orders of magnitude higher than the critical coagulation concentrations for mineral fragment colloids. The absence of mineral fragment colloids in high ionic strength solutions implies that contributions from mineral fragment colloids to the total mobile source term of radionuclides in a salt repository are minimal.

In this work, a Pitzer model is developed for the K+(Na+)-Am(OH)4−-Cl−-OH− system based on Am(OH)3(s) solubility data in highly alkaline KOH solutions. Under highly alkaline conditions, the solubility reaction of Am(OH)3(s) is expressed as: Solubilities of Am(OH)3(s) based on the above reaction are modeled as a function of KOH concentrations. The stability constant for Am(OH)4− is evaluated using Am(OH)3(s) solubility data in KOH solutions up to 12 mol•kg-1 taken from the literature. The Pitzer interaction parameters related to Al(OH)4- are used as analogs for the interaction parameters involving Am(OH)4- to obtain the stability constant for Am(OH)4-. The for the reaction is -11.34 ± 0.15 (2σ).

In this paper, the experimental results from long-term solubility experiments on micro crystalline neodymium hydroxide, Nd(OH)₃(micro cr), in high ionic strength solutions at 298.15 K under well-constrained conditions are presented. The starting material was synthesized according to a well-established method in the literature. In contrast with the previous studies in which hydrogen ion concentrations in experiments were adjusted with addition of either an acid or a base, the hydrogen ion concentrations in our experiments are controlled by the dissolution of Nd(OH)₃(micro cr), avoiding the possibility of phase change.

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