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DOE OSTI2023

Editorial: Sorption processes in nuclear waste management: data knowledge management and new methodologies for data acquisition/prediction

A fundamental approach to Nuclear Waste Repository research involves the collection of experimental data in a laboratory setting, development of empirical and/or mechanistic numerical models representing those observations, and application of these models (or Reduced Order Models) into reactive transport and performance assessment models as predictive tools for informing society of impacts and risks associated with nuclear waste repository scenarios (Stevens et al., 2020). Therefore, the assimilation and interpretation of experimental data must take advantage of both new data and the rich historical data available in the literature and apply novel modeling approaches to improve predictive tools, particularly from the standpoint of uncertainty quantification, for nuclear waste repository performance assessment (Zavarin et al., 2022).

sorption↗
DOE OSTI2020

NNSA-IAEC Science Area V Environmental ISR: Waste Management and Subsurface Science (Final Report FY18-FY20)

In August 2017, the U.S. National Nuclear Security Administration (NNSA) and the Israel Atomic Energy Commission (IAEC) launched a joint waste management program to evaluate the feasibility of siting a radioactive waste repository at intermediate depths in Israel. The bilateral collaboration involves three National Laboratories (LLNL, LANL, and Sandia), and the Nuclear Research Center, Negev, Geological Survey of Israel, and Ben Gurion University in Israel. This research is designed to assist with the evaluation of locations for an intermediate nuclear waste repository in Israel. This final report is a high-level summary of the results from the past three years by the team that worked on the WM-2: Radionuclide colloid-facilitated transport in fractured carbonate rock project for Science Area V, Environmental ISR: Subsurface Science and Waste Management. We have completed all of our tasks and met all of our milestones. MW-2 had three integrated tasks. Task 1 was laboratory experiments investigating colloid transport of radionuclides in natural fractured carbonate rocks from the Avdat formation, Israel. Task 2 was field experiments injecting radionuclide analogues pre-sorbed to clay colloids into fractured carbonate rock, and task 3 was numerical modeling of laboratory and field experimental results to assess the importance of colloids in the migration of relevant radionuclides. This task was incorporated into tasks 1 and 2. A 4 th task was initiated in FY20, to investigate the role of organics in facilitating radionuclide transport under the same conditions explored in task 1 due to the high organic content of the rocks being investigated for the immediate borehole locations. Preliminary results will be summarized here and the work will continue in FY21-23. The overall objective of this research is to evaluate the role of colloids (naturally occurring < 1 micron particles) in facilitating the transport of long-lived radionuclides from a nuclear waste repository situated in fractured carbonate rocks. Currently little data exists on radionuclide transport in carbonate rocks, but this is the main rock type available for siting a repository in Israel. Laboratory experiments were carried out in both Israel and the U.S., field experiments have taken place in Israel, and reactive transport modeling involved LLNL, LANL and Israel partner institutions. LLNL hosted a graduate student, Emily Tran, from Ben Gurion University in FY18 and FY19 and much of the work presented here was part of her PhD research.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

There is Nothing Here for You: Passive Institutional Controls and Preemptive Resource Exploitation - 16197

The Department of Energy (DOE) Carlsbad Field Office (CBFO) is required by Title 40 CFR Part 191.14(c) and Title 40 CFR Part 194.43 to develop a Passive Institutional Controls (PICs) program to prevent the likelihood for future human intrusion into the Waste Isolation Pilot Plant (WIPP). The purpose of this paper is to explore how the exploitation of resources under nuclear waste repositories as a part of PICs could reduce the risk of future human intrusion. The WIPP, located in the northern Permian Basin near Carlsbad, New Mexico, is a geologic repository for the permanent isolation of defense-related transuranic waste. In 1995, a series of geophysical surveys by the New Mexico Bureau of Mines and Mineral Resources estimated the total hydrocarbon reserves underneath the WIPP, which included a 16-section area plus a one mile buffer. Since that time the Delaware Basin well database has been maintained to closely monitor hydrocarbon extraction activities. The database was used in compiling relevant oil and gas extraction data retrieved from The New Mexico Oil Conservation Division. The volume of hydrocarbons extracted from 1995 to the present suggests that if current extraction growth trends continue (as historic market and production data indicate they will), all known crude oil reserves under the WIPP, obtainable through current extraction technology, will be depleted within approximately 10 years, i.e., by 2026. This is well before the anticipated WIPP closure and start of the Active Institutional Controls (AICs) program called for in 40 CFR Part 194.41(b). In addition, gas resources, although more difficult to forecast, are not expected to last beyond 2137. Title 40 CFR 194.33 states, 'Inadvertent and intermittent intrusion by drilling for resources... is the most severe human intrusion scenario'. This human intrusion scenario is the feature, event, and process (FEP) that performance assessment (PA) has identified as having the most significant impact on radionuclide releases during the 10,000 year regulatory period for the WIPP. Title 40 CFR 194.43 requires a PICs program which contains '... [records] in the international archives that would likely be consulted by individuals in search of unexploited resources'. Due to political, social, and economic motivators for resource exploration and exploitation, the potential future absence of hydrocarbon resources under the WIPP is important in deterring possible human intrusion. To minimize the risk of future human intrusion, it may be necessary to consider actively removing as many of the exploitable hydrocarbon resources under and around a nuclear waste repository as is practicable before the end of the period of AICs (i.e., 100 years after closure of the repository). Additionally, communicating the absence of hydrocarbon resources under a nuclear repository should take a prominent role in any PICs program. Due to its high-cost nature, exploratory drilling of any kind currently involves cost/benefit analysis. If in the future the economic realities of resource exploration remain similar to today, and future oil companies know the resources for a particular area have already been heavily exploited, hypothetically, they would become less motivated to drill in that area. They would instead opt to explore an area with better prospects. It is important then that the remaining hydrocarbon reserves be clearly stated and prominently displayed. In essence we should communicate the message, loud and clear, 'There is nothing here for you'. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Engineered Barrier System Research Activities at LBNL (FY2020 Progress Report)

The design of a radioactive waste repository typically involves a multi-barrier system, including the natural barrier system (NBS), i.e., the host rock and its surrounding subsurface environment, and an engineered barrier system (EBS). The EBS is to be constructed from the man-made, engineered materials placed within a repository. The repository includes the waste form, waste canisters, buffer materials, backfill, and seals. The most common buffer material for EBS is compacted bentonite, which features low permeability and high retardation of radionuclides. Extensive studies concerning the behavior of bentonite backfill in crystalline and argillite/shale geologic environments for nuclear waste disposal have been conducted by means of laboratory experiments, numerical modeling, and large-scale in situ tests in Underground Research Laboratories (URLs) in Switzerland, France, Belgium, and Japan. This report includes the results of LBNL’s research activities conducted according to the objectives and scope of the work packages “SF-20LB01030802 Engineered Barrier System R&D - LBNL” and “SF-20LB01030806 Engineered Barrier System International Collaborations - LBNL” of the Spent Fuel and Waste Science and Technology (SFWST) (formerly called Used Fuel Disposal) Campaign of the Department of Energy’s (DOE) Office of Nuclear Energy. LBNL research studies included laboratory scale tests and modeling of the evolution of the EBS bentonite and associated coupled processes, and impacts of high temperature on parameters and processes relevant to the performance of crystalline and argillite repositories, including the evaluation of the technical basis for applying the maximum allowable temperature. The results of these studies are being addressing the technical elements necessary to evaluate the EBS design concepts. Emphasis is on the study of thermal, mechanical, and chemical processes that influence the performance of EBS, and the development of modeling capability for reliable assessment of these processes, and ultimately supporting the development of the GDSA model with detailed coupled THMC process models.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2024

Soft interface instability and gas flow channeling in low-permeability deformable media

Understanding gas percolation through a clay layer or a shale formation is of great importance for the development of a geologic repository for nuclear waste disposal, a subsurface system for gas storage, and an engineering approach for hydrocarbon extraction from unconventional reservoirs. Gas injection experiments have revealed complex dynamic behaviours of gas percolation through water saturated compacted bentonite, characterized by a high breakthrough pressure, rapid breakthrough, a pressure/stress decay after the breakthrough, a relatively high migration rate, high-frequency periodic/nonperiodic variations in flow rate, stepwise rate reductions during relaxation, and low gas saturation over the whole process, all indicating channelling nature of the processes. Using linear stability analyses, we show that this channelling can autonomously emerge from the instability of the deformable interface between the injected gas and the compacted bentonite matrix driven by local stress concentration, pore dilation, and hydrologic gradient. Channel patterns formed would possess a fractal geometry. We further show that, once a percolating channel is established, the gas injected would percolate through the channel in a chain of gas bubbles, also due to the interface instability, resulting in periodic/chaotic variations in gas flow rate. Our work provides a unified explanation for key features observed for gas percolation in low-permeability deformable media. The work also suggests a possibility of designing an engineered barrier system for a nuclear waste repository that can have controllable gas release while limit water transport.

Bentonite↗
DOE OSTI2018

Modeling the Long-Term Behavior of Cementitious Barriers - 18188

Cementitious materials are commonly included as backfill/barrier materials in geological repositories for radioactive waste. An understanding of the long-term behavior of cement-based barriers is therefore required to provide confidence in their safety functions within an environmental safety case. Geochemical modelling has been undertaken on several different radioactive waste repository concepts (and their natural/industrial analogues), to explore the nature of cement degradation over timescales relevant to postclosure safety (many tens of thousands of years to c. 1 Ma). In radionuclide transport calculations undertaken as part of performance assessments, it is typically considered that the properties of cementitious engineered barrier materials change as they degrade, potentially resulting in an increase in porosity hydraulic conductivity, and effective diffusion coefficients. However, reactive transport modelling (supported by some analogue data) suggests that, in some repository systems, a degree of clogging of the cement pore space could occur and that some advective flow paths (cracks in the barriers, fractures in the host rock) could become (to some extent) infilled by cement and rock alteration products. This could result in a reduction in water flow and contaminant transport in the vicinity of the cement-rock interface, but some uncertainties remain (especially process couplings and the nature of water flow in complex fracture networks that are becoming clogged). In addition, the armoring of fracture surfaces could result in decreased potential for diffusion of solutes into the rock matrix. It is envisaged that ongoing work will continue to improve reactive-transport model capabilities, such that the implications cement degradation has for contaminant transport can be predicted with greater confidence. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2014

Investigation of Coupled THMC Processes and Reactive Transport: FY14 Progress

The multi-barrier system in a nuclear waste repository typically consists of the natural barrier system (NBS), which includes the repository host rock and its surrounding subsurface environment, and the engineered barrier system (EBS). The EBS represents the man-made, engineered materials placed within a repository, including the waste form, waste canisters, buffer materials, backfill, and seals (OECD 2003).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Installation and operation of GeoMelt{sup R} in container vitrification in NNL Central Lab active rig hall at Sellafield - 15328

Cleanup and subsequent immobilization of wastes on the UK's NDA estate brings major challenges. With the exception of HLW where vitrification is utilized for immobilization, cement is typically used in the management and immobilization of non-heat generating wastes. However some legacy feeds such as those containing reactive metals can be incompatible with cementation technologies, and the application of engineered cement encapsulating product to raw wastes increases wasteform volume. Cementation has also been developed to be compatible with the repository concept, which provides it with a level of protection from the environment. Potential delays in repository implementation and as yet to be determined repository waste form acceptance criteria for some wastes may require the duration of above ground storage of immobilized wastes to be extended with potential effect on the integrity of the wasteform. Vitrification is a treatment option which destroys organics and hazardous inorganics such as asbestos, reduces wasteform volume, and produces a robust leach resistant wasteform. Kurion's GeoMelt{sup R} In-container Vitrification (ICV){sup TM} is a vitrification process wherein the treatment vessel can serve as an interim-storage, transport, and final long-term storage vessel - thus negating the need to pour glass into a separate container, effectively leading to a much simpler system than traditional waste melters. The inherent volume reduction from vitrification results in a dense product with self-shielding properties, and the GeoMelt treatment process has been demonstrated to homogenize elemental metals and radionuclide contaminants of concern, eliminating hot spots and mitigating criticality concerns. All of which opens up alternatives strategies for storage and disposal. GeoMelt has been proven in the commercial and US DOE waste treatment environment with over 26,000 tonnes of hazardous waste treated to date (half of this radioactive). Kurion and the UK's National Nuclear Laboratory (NNL) plan to demonstrate its application to Sellafield radioactive wastes. A small (engineering) scale demonstrator is being installed in the active rig hall at NNL's Central Laboratory at Sellafield UK with a twofold mission; 1) demonstration of the vitrification of ILW/LLW, and 2) processing of smaller volume orphan wastes that currently have no treatment path. The NNL facility will allow the handling of uranium containing materials which form a substantial component of ILW currently stored on the Sellafield site. Accurate DFs across the melter and off gas system are often difficult to obtain using inactive materials at the low concentrations present in radioactive wastes. Installation in the NNL facility will allow trace quantities of active species such as Cs-137 and Tc-99 to be processed as part of a simulant ILW stream and monitoring will allow accurate decontamination factors (DFs) to be established to support future vitrification facility design. In addition to underpinning melter operation and obtaining DFs across the system, the facility will allow the processing of radioactive waste arisings of low activity which may either be typical of larger waste volumes or may present as orphans without a disposition route. In either case the melter facility will be able to vitrify these wastes into a form that will be suitable for interim storage and disposal. Among the feeds envisaged for processing will be asbestos from power stations, contaminated soils, feeds containing uranium metals, ponds solids, sludge, and miscellaneous active laboratory wastes. The NNL Central Lab capability will be used to analyze the active product to support disposal requirements whether through letters of compliance for the Geological Disposal Facility (GDF) or waste acceptance criteria for the Low Level Waste Repository (LLWR). (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

As Permanent as Practicable: The reality of Permanent Markers - 16191

The regulatory requirements in the United States do not require any specific time frame for the duration of permanent marker systems to inform future generations of the presence of deep geologic nuclear waste repositories. The requirement is for markers to be 'as permanent as practicable'. For the Waste Isolation Pilot Plant in New Mexico, the U.S. Department of Energy chose a 10,000 year Passive Institutional Controls program goal. Passive Institutional Controls are to be designed prior to facility closure and implemented prior to the completion of the Active Institutional Controls program. Nuclear waste repositories across the world are actively developing marker and message systems, as required by their own regulations, for time frames either specified in those regulations or to be proposed to the regulator by the implementing organization. Time frames may be based upon the radioactive characteristics of the various isotopes being disposed of in the repository under consideration, or they may be based on some utility function that balances investment, practicability and risk. Scientifically, there are a number of time frames that make sense. However, an important determinant of effectiveness that has not been taken into consideration is the impact of the human connection, the receiver of the message. The major obstacle that the permanent marker systems need to overcome is the impact that humans will have over time. Historically, no monument has ever survived more than a few thousand years when humans come in contact with it. No matter what size of monument or topic the monument represents, humans have a history of destroying monuments beginning just a few generations after the monument's construction. The only way to ensure that a monument system can endure long time frames is to isolate it from humans. Eliminating contact with humans for whom the messages are meant defeats the very purpose of having an informative marker and message system. A more reasonable and historically supported time frame of 1,000 years would assure that no unreasonable cost would be incurred by current generations. Future generations could decide to continue maintaining the message, or not; and the likelihood of destruction of the markers by subsequent generations would be reduced since the knowledge of the markers and messages would have been transmitted within a more reasonable time frame over which language continuity, for example, would likely be manageable. Future preservation of the marker system and future human actions cannot be controlled, but the goal is to build a marker and message delivery system that reduces the likelihood of inadvertent future human disturbance of a deep geologic repository. This is the typical regulatory requirement on future information systems. Regulators who insist on an extremely long time frame for marker effectiveness should consider the historical realities of 'permanent' markers and monuments and propose a more reasonable time frame. Perhaps it would be more practicable to require 1,000 years for informing future generations and then allow them to determine what they wish to do to protect the future for another, perhaps similar, period of time. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

On the impact of temperatures up to 200°C in clay repositories with bentonite engineer barrier systems: A study with coupled thermal, hydrological, chemical, and mechanical modeling

One of the most important design variables for a geological nuclear waste repository is the temperature limit up to which the engineered barrier system (EBS) and the natural geologic environment can be exposed. Up to now, almost all design concepts that involve bentonite-backfilled emplacement tunnels have chosen a maximum allowable temperature of about 100°C. Such a choice is largely based on the consideration that in clay-based materials illitization and the associated mechanical changes in the bentonite (and perhaps the clay host rock) could affect the barrier attributes of the EBS. However, existing experimental and modeling studies on the occurrence of illitization and related performance impacts are not conclusive, in part because the relevant couplings between the thermal, hydrological, chemical, and mechanical (THMC) processes have not been fully represented in the models. This work presents a fully coupled THMC simulation of a nuclear waste repository in a clay formation with a bentonite-backfilled EBS for 1000 years. Two scenarios were simulated for comparison: a case in which the temperature in the bentonite near the waste canister can reach about 200°C and a case in which the temperature in the bentonite near the waste canister peaks at about 100°C. The model simulations demonstrate some degree of illitization in both the bentonite buffer and the surrounding clay formation. Other chemical alterations include the dissolution of K-feldspar and calcite, and precipitation of quartz, chlorite, and kaolinite. In general, illitization in the bentonite and the clay formation is enhanced at higher temperature. However, the quantity of illitization is affected by many chemical factors and therefore varies a great deal. The most important chemical factors are the concentration of K in the pore water as well as the abundance and dissolution rate of K-feldspar; less important ones are the concentration of sodium and the quartz precipitation rate. In our modeling scenarios, the calculated decrease in smectite volume fraction in bentonite ranges from 1 to 8% of the initial volume fraction of smectite in the 100°C scenario and 1–27% in the 200°C scenario. Chemical changes in the 200°C scenario could also lead to a reduction in swelling stress up to 15–18% whereas those in the 100°C scenario result in about 14–15% reduction in swelling stress for the base case scenario. Model results also show that the 200°C scenario results in a much higher total stress than the 100°C scenario, mostly due to thermal pressurization. While cautions should be taken regarding the model results due to some limitations in the models, the modeling work is illustrative in light of the relative importance of different processes occurring in EBS bentonite and clay formation at higher than 100°C conditions, and could be of greater use when site specific data are available.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Technical Approach for Retrieving Heat-Generating Waste from Repositories in Salt Formations - 16240

During the last decades, the management of SNF and HLW in Germany focused on deep geological disposal inside salt formations. The geological barrier of the rock salt is to safely isolate the waste over a period of one million years. In re-cent years, retrievability has moved into the focus of an ongoing public discus-sion about waste management strategies. The discussion is driven by socio-ethical, economical, safety and technical aspects. As DBE Technology GmbH has expertise in repository safety and technology, the development of technical concepts for retrieving SNF and HLW canisters was launched in an R and D project. This paper summarizes the latest results concerning the implementation of retrievability into existing emplacement concepts for rock salt in Germany and gives an overview of how retrieval could be realized safely and in a technically reliable manner. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

Review Guidelines for Biosphere Safety Assessment of Low-Level Waste Disposal Repository in Taiwan - 19606

Conducting reviews of radiological dose assessments is a crucial step toward the approval of long-term radiological safety requirements for low-level radioactive waste (LLW) disposal facilities. However, the concentration of radionuclides released from the near field into the biosphere and the radiation exposure levels to an individual can influence the performance acceptability of the LLW disposal facility and safety of the disposal environment. Therefore, all LLW must be managed and disposed of such that it minimizes risk to individuals and the environment. A dose of 0.25 mSv/year is the limit for individuals expected to receive the highest level of exposure to residual radioactivity. In Taiwan, a review for license applications has been proposed for LLW disposal facilities. These review guidelines are based on the Standard Review Plan (NUREG-1200) and are amended according to the environmental conditions in Taiwan. These guidelines can help reviewers to perform safety assessment reviews of applications for the construction and operation of LLW disposal facilities. The review guidelines for biosphere safety assessment include radiological dose assessment, description of waste characteristics, radionuclide transport characteristics, radiation dose under normal and unusual conditions, radionuclide release into the biosphere, and interpretation of evaluation results in various transfer mechanisms. These guidelines can also serve as references for regulatory departments conducting safety assessments of radiological doses in LLW disposal facilities in Taiwan. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

Building Sustainable and Efficient Markers to Bridge Ten Millennia: The Understanding Issues-18058

Markers of nuclear waste deposits that can be understood through ten millennia have been proposed for preventing from inadvertent human intrusion. It is postulated that currently designed waste repository markers are likely to fail, and an alternative system of markers is proposed based on a praxeological device, i.e., a language-learning by- action device. The key challenge is how should we represent radioactivity in waste repositories in such a way that it is understood by anyone who encounters such markers? Namely, for an adequate interpretation, two issues have to be addressed: to accommodate differing world views, and inevitable changes in languages. What is required is an integrated system that allows us to communicate a complex and abstract concept to all people at all times. That is the goal of our praxeological device, a sign that links a person's actions (praxis) in a particular context to a communication system about those actions and that particular context. That means anyone who experiences such a praxeological device will understand the actions that s/he is about to perform will help her/him to learn the iconic language s/he is reading. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2023

Surface Complexation/Ion Exchange Hybrid Model for Radionuclide Sorption to Clay Minerals (M4SF-23LL010301062)

This progress report (Level 4 Milestone Number M4SF-23LL010301062) summarizes research conducted at Lawrence Livermore National Laboratory (LLNL) within the Argillite International Collaborations Activity Number SF-23LL01030106. The activity is focused on our long-term commitment to engaging our partners in international nuclear waste repository research. The focus of this milestone is the establishment of international collaborations for surface complexation modeling and the associated impacts of unlocking larger, community-based datasets. More specifically, we are developing a database framework for Spent Fuel and Waste and Science Technology (SFWST) that is aligned with the Helmholtz Zentrum Dresden Rossendorf (HZDR) sorption database development group in support of the database needs of the SFWST program. In our FY22 effort, we described a detailed analysis of U(VI) sorption to quartz through both traditional surface complexation modeling and through a hybrid ML framework. In FY23, effort was placed on publication of these results and expansion of the LLNL surface complexation and ion exchange database (L-SCIE) in order to assess mineral-based radionuclide retardation under a wider variety of geochemical conditions (e.g., ionic strength, varying electrolyte compositions). Efforts were initiated to expand L-SCIE to include radionuclide surface complexation and ion exchange to clays that are relevant to subsurface geochemical processes occurring at nuclear waste repositories. In particular, a large source of sorption data for clays resides at the Paul Scherrer Institute (PSI) (work primarily by Bradbury and Baeyens) and we initiated discussions on how to retrieve those data and apply FAIR principles to those datasets. In addition to L-SCIE development, two hybrid models that incorporate AI/ML were investigated and compared to discern the most promising approaches for accurate and precise estimations of radionuclide retardation. Key considerations for future model development include (1) the ability to reduce computational burden on determining retardation coefficients for PA and (2) the ability to quantify and predict radionuclide-mineral partitioning at a more efficient, rapid pace due to automated workflows. Upon the careful consideration of the most effective modeling approaches, we are identifying ways to implement these approaches into PA. Ultimately, the data science-based workflows will provide a major incentive for other institutions to adopt a FAIR-formatted, interoperable database. LLNL will play a key role in disseminating sorption data and acting as good data stewards by updating the database in a consistent format and assessing the quality of the newly assimilated data in an organized fashion. To this end, all data and workflows are open access and made available on the LLNL Seaborg research website (https://seaborg.llnl.gov/resources/geochemical-databases-modeling-codes).

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2016

Quantification of Cation Sorption to Engineered Barrier Materials Under Extreme Conditions

The objective of this research is to examine mechanisms and thermodynamics of actinide sorption to engineered barrier materials (iron (oxyhydr)oxides and bentonite clay) for nuclear waste repositories under high temperature and high ionic strength conditions using a suite of macroscopic and microscopic techniques which will be coupled with interfacial reaction models. Gaining a mechanistic understanding of interfacial processes governing the sorption/sequestration of actinides at mineral-water interfaces is fundamental for the accurate prediction of actinide behavior in waste repositories. Although macroscale sorption data and various spectroscopic techniques have provided valuable information regarding speciation of actinides at solid-water interfaces, significant knowledge gaps still exist with respect to sorption mechanisms and the ability to quantify sorption, particularly at high temperatures and ionic strengths. This objective is addressed through three major tasks: (1) influence of oxidation state on actinide sorption to iron oxides and clay minerals at elevated temperatures and ionic strengths; (2) calorimetric titrations of actinide-mineral suspensions; (3) evaluation of bentonite performance under repository conditions. The results of the work will include a qualitative conceptual model and a quantitative thermodynamic speciation model describing actinide partitioning to minerals and sediments, which is based upon a mechanistic understanding of specific sorption processes as determined from both micro-scale and macroscale experimental techniques. The speciation model will be a thermodynamic aqueous and surface complexation model of actinide interactions with mineral surfaces that is self-consistent with macroscopic batch sorption data, calorimetric and potentiometric titrations, X-ray absorption Spectroscopy (XAS, mainly Extended X-ray Absorption Fine Structure (EXAFS)), and electron microscopy analyses. The novelty of the proposed work lies largely in the unique system conditions which will be examined (i.e. elevated temperature and ionic strength) and the manner in which the surface complexation model will be developed in terms of specific surface species identified using XAS. These experiments will thus provide a fundamental understanding of the chemical and physical processes occurring at the solid-solution interface under expected repository conditions. Additionally, the focus on thermodynamic treatment of actinide ion interactions with minerals as proposed will provide information on the driving forces involved and contribute to the overall understanding of the high affinity many actinide ions have for oxide surfaces. The utility of this model will be demonstrated in this work through a series of advective and diffusive flow experiments.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2014

Evaluation of Used Fuel Disposition in Clay-Bearing Rock

Radioactive waste disposal in shale/argillite rock formations has been widely considered given its desirable isolation properties (low permeability), geochemically reduced conditions, anomalous groundwater pressures, and widespread geologic occurrence. Clay/shale rock formations are characterized by their high content of clay minerals such as smectites and illites where diffusive transport and chemisorption phenomena predominate. These, in addition to low permeability, are key attributes of shale to impede radionuclide mobility. Shale host-media has been comprehensively studied in international nuclear waste repository programs as part of underground research laboratories (URLs) programs in Switzerland, France, Belgium, and Japan. These investigations, in some cases a decade or more long, have produced a large but fundamental body of information spanning from site characterization data (geological, hydrogeological, geochemical, geomechanical) to controlled experiments on the engineered barrier system (EBS) (barrier clay and seals materials). Evaluation of nuclear waste disposal in shale formations in the USA was conducted in the late 70’s and mid 80’s. Most of these studies evaluated the potential for shale to host a nuclear waste repository but not at the programmatic level of URLs in international repository programs.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

The OECD-NEA RK and M Initiative (2011-2018) - a 7-year step in a 10,000-year journey: What next? - 19432

The OECD Nuclear Energy Agency (NEA) has coordinated an international initiative to develop understanding and guidelines related to the long-term preservation of records, knowledge and memory (RK and M) for radioactive waste repositories. After seven years of work, it has come to a close in 2018. Currently, the 'Final Report of the RK and M Initiative' is in the final process of publication [the status/sentence will be updated in early 2019]. It includes a historical review, sets out the RK and M fundamentals, presents a menu of approaches and mechanisms of RK and M preservation, and finally provides suggestions on how to develop a systemic strategy aimed at embedding radioactive waste repositories in society over extended periods of time. This paper highlights the 'multi-dimensionality' of RK and M preservation: Multiple time frames need to be considered, multiple transmission modes explored, and multiple actors brought together. The latter seems to be the primary aspect to be considered in the discussion about future (international) cooperation in the RK and M preservation field. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Applicability of Wireless Power Transfer for Monitoring Technology of Radioactive Waste Geological Disposal - 16156

To assure the engineered barrier of a high-level radioactive waste repository functions correctly, long-term monitoring inside the backfilled deposition holes and tunnels is planned. Cables are used to power monitoring equipment and send signals however complete isolation will not be possible. Previously developed monitoring systems send information wirelessly and used batteries. Due to the limited battery life, these systems cannot be used for long-term monitoring. In this paper a wireless power transfer method based on magnetic resonant coupling is presented. The necessary transmitter and receiver coils have been developed. The performance of the individual coils has been tested. The performance of the system of transmitter and receiver with both air and bentonite as medium was tested. The results showed that the system with bentonite performed similarly well as the system with air. Wireless power transfer based on magnetic resonant coupling is therefore a feasible method for high-level radioactive waste repository monitoring applications which could in principle work for a period of 100 year or longer. (authors)

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