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Smectite (e.g., Montmorillonite): phyllosilicate minerals found in bentonites. Bentonites have been considered as key backfill barrier materials in deep geological nuclear waste repository concepts. Swelling/shrinking of montmorillonite (MMT) occurs with increasing/decreasing relative humidity. Microscopically, how does the hydration/dehydration process occur?
To understand geological systems that are important to our national well-being -- such as nuclear waste repositories or hydraulically fractured rocks -- requires the ability to make quantitative projections of the evolution of rock-water systems forward in time. A fundamental goal of much of our geochemical research was focused on development of numerical models of reactive transport to simulate the evolution of natural water-rock systems. The geochemical community has provided such models and they are rapidly being improved with field-model cross-testing. This project was especially impactful in field-model testing for well-constrained weathering systems to provide new conceptual understanding of how these systems function.
This progress report (Level 4 Milestone Number M4SF-22LL010302062) summarizes research conducted at Lawrence Livermore National Laboratory (LLNL) within the Crystalline International Collaborations Activity Number SF-22LL01030206. The activity is focused on our long-term commitment of engaging our partners in international nuclear waste repository research. This includes participation in the Nuclear Energy Agency Thermochemical Database (NEA-TDB) Project (Cindy Atkins-Duffin, Mavrik Zavarin) and development of methodologies for integrating US and international thermodynamic databases for use in SFWST Generic Disposal System Assessment (GDSA) efforts.
Computational and mathematical models are essential to understanding complex systems and phenomena. However, when developing such models, limited knowledge and/or resources necessitates the use of simplifying assumptions. It is therefore crucial to quantify the impact of such simplifying assumptions on the reliability and accuracy of resulting model predictions. This work develops a first-of-its-kind approach to quantify the impact of physics modeling assumptions on predictions. Here, we leverage the emerging field of model-form uncertainty (MFU) representations, which are parameterized modifications to modeling assumptions, in combination with grouped Sobol’ indices to quantitatively measure an assumption’s importance. Specifically, we compute the grouped Sobol’ index for the MFU representation’s parameters as a single importance measure of the assumption for which the MFU representation characterizes uncertainty. To ensure this approach is robust to the subjective choice of how to parameterize a MFU representation, we establish bounds for the difference between sensitivity results for two different MFU representations based on differences in model prediction statistics. The capabilities associated with this approach are demonstrated on three exemplar problems: an upscaled subsurface contaminant transport problem, ablation modeling for hypersonic flight, and nuclear waste repository modeling. We found that our grouped approach is able to assess the impact of modeling assumptions on predictions and offers computational advantages over classical Sobol’ index computation while providing more interpretable results.
As the prospect of exceeding global temperature targets set forth in the Paris Agreement becomes more likely, methods of climate intervention are increasingly being explored. With this increased interest there is a need for an assessment process to understand the range of impacts across different scenarios against a set of performance goals in order to support policy decisions. The methodology and tools developed for Performance Assessment (PA) for nuclear waste repositories shares many similarities with the needs and requirements for a framework for climate intervention. Using PA, we outline and test an evaluation framework for climate intervention, called Performance Assessment for Climate Intervention (PACI) with a focus on Stratospheric Aerosol Injection (SAI). We define a set of key technical components for the example PACI framework which include identifying performance goals, the extent of the system, and identifying which features, events, and processes are relevant and impactful to calculating model output for the system given the performance goals. Having identified a set of performance goals, the performance of the system, including uncertainty, can then be evaluated against these goals. Using the Geoengineering Large Ensemble (GLENS) scenario, we develop a set of performance goals for monthly temperature, precipitation, drought index, soil water, solar flux, and surface runoff. The assessment assumes that targets may be framed in the context of risk-risk via a risk ratio, or the ratio of the risk of exceeding the performance goal for the SAI scenario against the risk of exceeding the performance goal for the emissions scenario. From regional responses, across multiple climate variables, it is then possible to assess which pathway carries lower risk relative to the goals. The assessment is not comprehensive but rather a demonstration of the evaluation of an SAI scenario. Future work is needed to develop a more complete assessment that would provide additional simulations to cover parametric and aleatory uncertainty and enable a deeper understanding of impacts, informed scenario selection, and allow further refinements to the approach.
Fabrication of low-enriched experimental nuclear reactor fuel for the Advanced Gas Reactor (AGR) Fuel Development and Qualification program has concluded after several years of effort. Residual chemicals, uranium compounds, fuel intermediates and fuel compacts need to be dispositioned in an economically and environmentally sound manner that maximizes, within reason, the preservation and repurposing of valuable materials. Government Furnished Equipment (GFE), procured with AGR program funds, are also to be dispositioned as economically as reasonably possible while seeking to maintain the functionality of BWXT equipment and repurposing of stand-alone GFE. Any stand-alone equipment that cannot be repurposed will either be abandoned in place, with BWXT consent, or discarded to an appropriate waste repository.
Fabrication of low-enriched experimental nuclear reactor fuel for the Advanced Gas Reactor (AGR) Fuel Development and Qualification program has concluded after several years of effort. Residual chemicals, uranium compounds, fuel intermediates and fuel compacts need to be dispositioned in an economically and environmentally sound manner that maximizes, within reason, the preservation and repurposing of valuable materials. Government Furnished Equipment (GFE), procured with AGR program funds, are also to be dispositioned as economically as reasonably possible while seeking to maintain the functionality of BWX Technologies, Inc. (BWXT) equipment and repurposing of stand-alone GFE. Any stand-alone equipment that cannot be repurposed will either be abandoned in place, with BWXT consent, or discarded to an appropriate waste repository. Representatives for the United States Department of Energy may intervene in intended disposition of specific equipment or materials. Revision 3 on document
Fabrication of low-enriched experimental nuclear reactor fuel for the Advanced Gas Reactor (AGR) Fuel Development and Qualification program has concluded after several years of effort. Residual chemicals, uranium compounds, fuel intermediates and fuel compacts need to be dispositioned in an economically and environmentally sound manner that maximizes, within reason, the preservation and repurposing of valuable materials. Government Furnished Equipment (GFE), procured with AGR program funds, are also to be dispositioned as economically as reasonably possible while seeking to maintain the functionality of BWX Technologies, Inc. (BWXT) equipment and repurposing of stand-alone GFE. Any stand-alone equipment that cannot be repurposed will either be abandoned in place, with BWXT consent, or discarded to an appropriate waste repository. Representatives for the United States Department of Energy may intervene in intended disposition of specific equipment or materials.
Montmorillonite is an abundant clay mineral that has been identified as the primary component in engineered bentonite barrier systems in future nuclear waste repositories. Selective sorption onto planar and edge surfaces of clay particles make it a great candidate for the attentuation of uranium(VI). Although previous works have investigated uranium(VI) interactions with this clay, few have done so in the presence of calcite, a mineral impurity in bentonite that is understood to alter U(VI) speciation, sorption, and consequently its mobility. PHREEQC software is used to simulate the chemical solution speciation of U(VI) under various known repository conditions, in the absence and presence of calcite, and across a range of U(VI)concentrations (1.0E-7 to 1.0E 5 M), and partial pressures of CO 2 (zero to 2% CO2) (Fig. 1). In addition, the kinetic dissolution of calcite is simulated to gain a better understanding of how quickly chemical equilibrium can be reached in these systems. Lastly, surface complexation models will be developed to characterize the influence of calcite impurities on the formation of ternary U(VI)-Ca-carbonate solution species on U(VI) sorption behavior. Our results indiciate that the presence of Ca 2+ and CO 3 2- ions released from calcite leads to U(VI)-Ca-carbonate species with a negative or neutral charge even without the presence of CO 2 . Quantification of these species will allow for the development of more accurate U(VI)-montmorillonite sorption models. Calcite precipitation at alkaline pH further complicates speciation by creating a competation for Ca 2+ . This research is being performed using funding received from the DOE Office of Nuclear Energy’s NEUP.
Solvent extraction is used to separate metals or other complexes into two different immiscible liquids and is an essential component of the PUREX Process. Improvements to the solvent extraction process can directly contribute to an organization’s ability to ensure the purity and recovery of special nuclear materials from spent nuclear fuel. Such improvements can benefit nuclear reprocessing efforts, increase fuel reutilization, and limit the concentration of actinides in nuclear waste repositories.
Spatially heterogeneous uncertainties are prevalent in geophysical modeling applications, such as probabilistic post-closure performance assessment (PA) of deep geologic repositories for nuclear waste. Such uncertainties are often highly influential to model outputs, so it is desirable to identify the most important mechanisms by which they influence model predictions. However, these uncertainties are often unparameterized in the sense that there is no set of parameters that can be specified to yield a particular realization of the uncertainty. Additionally, the uncertainty is not intrinsically endowed with a parameterization that captures a realization’s mechanistic influence on model outputs. Therefore, in this work we present a novel methodology to develop and assess a set of proxy variables that aim to represent this influence. We show how they can be computed, downselected, and incorporated into the construction of statistical surrogate models mapping model inputs to outputs. We present our methodology in the context of a motivating application problem in deep geologic repository PA and discuss the challenges in capturing the effects of these spatial heterogeneities in uncertainty analyses.
This study examines methods that can help maximize confidence in maintaining Continuity of Knowledge (CoK) on plutonium-bearing wastes, from a final safeguards-verification measurement through emplacement underground. The study identifies Containment and Surveillance (C/S) measures that can be applied during packaging of plutonium wastes at the Savannah River Site (SRS) in South Carolina, USA, through shipment to, and receipt and disposal at the Waste Isolation Pilot Plant (WIPP) near Carlsbad, New Mexico, USA. Results of this study could apply to countries with a Comprehensive Safeguards Agreement (CSA) that plan to dispose in a geological repository plutonium or other non-fuel nuclear materials that are under international safeguards.
The Nuclear Waste Management Organization (NWMO) has conceptualized a unique method of long-term storage for spent nuclear waste. The concept encompasses all aspects of the end-of-life process from the point the spent fuel is no longer in-service; including handling, shipping, containing, storing, monitoring, and finally decommissioning. The equipment developed for each process is unique and designed specifically for each operation. Since these methods have not been implemented before, it is required to prove each process and the related equipment through simulations and eventually practical tests to evaluate performance and to confirm effective and safe operation. The spent fuel will be stored in an underground repository several hundred meters below the surface using autonomous machinery. Considering the significant investment and labour involved in creating full-scale tests underground, it would be impractical at this project stage to attempt tests of the newly developed machinery and methods in other than a surface setting. Therefore, our scope was to provide a customizable mock-up structure to simulate the conditions and constraints an underground repository would present. (authors)
Advances in simulation of coupled heat and mass transfer in granular salt using the Finite Element Heat and Mass Transfer Code (FEHM) is presented in this paper. The modeling focuses on reproducing the results of an in-situ subsurface evaporation experiment beginning during the summer of 2015 carried out in the underground at the Waste Isolation Pilot Plant (WIPP). Run-of-mine (granular) salt was placed in a metal pan, and sample mass, as well as ambient temperature and humidity, were measured continuously on a data logger through May 2016. Changes in moisture content of the salt are interpreted from the change in mass, and these data provide insight into evaporation and water retention characteristics from salt in the WIPP underground. To allow simulation of the experimental data, FEHM was modified to include a variable humidity boundary condition. The new humidity boundary condition can adjust streams of inflowing air to reflect fractions of air and water at prescribed temperature and pressure conditions that in turn represent relative humidity. Attributes for the flow such as temperature, air mass fraction, and humidity can be specified for inflowing air. The in-situ experiment provided data for validation of FEHM code development to accurately compute water-vapor pressure lowering due to dissolved salt and capillary pressure for pore water stored granular WIPP salt. Using this new boundary condition capabilities, a numerical representation of the pan experiment was created. The numerical domain is a 2-dimensional square grid of 30 by 30 cm axes. Salt and air properties are assigned based on measured values. A low flow rate of air is specified across the pile of salt to represent drift ventilation airflow. Humidity and temperature of the flowing air are fixed to measured values, and the simulated pile of salt is allowed to take up or release water based only on gradients in humidity and water-vapor pressure. Mass changes of the modeled salt pan were compared to measured mass changes in the bench-scale experiment. Directionality of the modeled changes tracks well with measured values, with good correlation between peak and valley inflection points. This shows that in situations with airflow where the relative humidity boundary can apply, FEHM can produce accurate values if humidity, air flow, and initial conditions are well established. Finally, this work supports phased-field-thermal testing in salt planned for the underground at WIPP that is expected to begin in early 2017. (authors)
Concepts for the final disposal of high level radioactive waste in deep geological formations must consider the heat pulse produced from radioactive decay during operations and in the post-closure phase as an important factor for the integrity of barrier systems. Several safety cases in different countries limit the maximum temperature at the boundary between the technical barrier/canister surface to the host rock/buffer material to 100 C. Increases in maximum temperature have very diverse effects with respect to the type, intensity and range of the heat pulse. These depend, among other factors, on i) the amount of heat emitted, ii) the type of host rock including site-specific characteristics, iii) the safety and repository concept and further site-specific boundary conditions. The German Commission on the Disposal of High-level Radioactive Waste (Commission) recommended, for precautionary reasons, that the temperature limit at the outer surface of waste containers should not exceed 100 C. Higher temperatures could only be considered if the maximum physically possible temperatures in the respective host rock have been reliably set as a result of research findings. (authors)
When disposing of radioactive waste with negligible heat generation, the activity of the radionuclides is limited. However, the waste including radioactive material comprises a certain proportion of non-radioactive material. The risk of hazardous contamination of the groundwater by non-radioactive substances must be taken into account. It must be ascertained that the non-radioactive material cannot affect near-surface groundwater, which is a resource for drinking water supply. To achieve this, limit values on the amount of hazardous substances in the total waste volume shall be determined. In 2002 a regulatory approval was granted to the 'Konrad' final storage facility for radioactive waste with negligible heat generation according to the German Water Resources Act. This approval is embedded in the license (plan-approval decision) based on the German Atomic Energy Act. The approval is based on the model of a possible discharge of contaminants from the final storage facility. In a long-term simulation spanning 300,000 years, contact of the waste with the groundwater in the surrounding rock formations over the postoperational phase was conservatively assumed. The maximum amount of allowable disposal of substances - such that no concern of affecting the near-surface groundwater arises - can be derived from the solubility of the respective substances. Implementing the permit according to the Water Resources Act requires: - classification of the waste through delivery declaration obligations - monitoring the substance inspection in the waste to assess its suitability for the disposal, and - stock keeping procedures for the disposal waste by the operator. While the repository is operated by the Federal Office for Radiation Protection (BfS) (http://www.bfs.de./EN/home/home{sub n}ode;), reporting to the Federal Environment Ministry, the statutory water permit is issued and applied by the State of Lower Saxony as responsible state authority. A waste package/container list supplements a comprehensive list of substances. From this - depending on the respectively conditioned waste - material vectors can be derived. For this purpose, BfS provides a database with a substances and container list to the waste owners. In the meantime, first experiences have been collected in applying the approval. More than 540 waste containers with a total mass of almost 6,000 Mg have been controlled and are ready for disposal. In our presentation we will show the recent development in supervising the preparation for the disposal of mixed waste. (authors)