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At least 199 records · Page 11

Coupled Thermo-Hydrological-Mechanical-Chemical Behavior of Anisotropic Granite for Geologic Disposal of High-Level Radioactive Waste: A Core-Scale Laboratory Investigation

The coupled thermo-hydrological-mechanical-chemical (THMC) behavior of rock within an Excavation Damaged Zone (EDZ) is critical for the safety and long-term performance of a geological repository for high-level radioactive wastes. While many laboratory experiments have been conducted to investigate EDZ rocks, the flow and deformation characteristics resulting from anisotropic rock textures and microcrack distribution under triaxial loading and elevated temperatures remain poorly understood. Particularly, cracks at various scales serve as fast paths for fluid flow and solute transport and present as focal points of mechanical weakness, which complicate the coupled THMC processes in anisotropic EDZ rocks and challenge modeling predictions. Here, in this study, a series of core-scale experiments was conducted on three granite samples under repository-relevant conditions. These rock samples were obtained from the Grimsel Underground Research Laboratory (URL), featured by anisotropic minerals (represented by bedding layers) and microcrack distributions and coarse cm-scale grain sizes. During the experiments, samples were subjected to an elevated temperature at 90 °C and different triaxial loading conditions either by radial (normal to bedding layers) or axial (parallel to bedding layers) compaction. Water was injected into the samples, and the rock permeability evolutions and effluent water chemistry were monitored closely. For intact samples, thermal expansion of minerals at 90 °C resulted in a large, 75% irreversible permeability reduction and rock strengthening under radial compaction, while thermal impact was limited to a 15% permeability reduction under axial compaction. In contrast, for a sample containing open cracks, the growth of fractures during the experiment resulted in an abrupt permeability increase and fast failure at 90 °C. The effluent water chemistry indicates much more considerable mineral dissolution from large shear sliding than that in rocks dominated by mechanical compaction. These results helped better understand the coupled THMC processes in anisotropic rocks containing cracks, evaluate the behaviors of EDZ rocks, and predict the long-term evolution of EDZ for the performance of the repositories.

Coupled THMC processes↗

Modeling glass degradation and release of radionuclides from vitrified waste for performance assessment simulations

The release of radionuclides initially encapsulated in a slowly degrading solid waste form and contained in an eventually corroding canister defines the source term for numerical simulations for the assessment of a geologic repository for high-level radioactive waste. While the details of waste degradation, canister corrosion, and dissolution and mobilization of the radionuclides in pore water include complex chemical reaction and transport processes that are coupled to the thermal, hydrological, microbiological, and mechanical conditions in the repository, the source-term model suitable for use in a numerical performance assessment model should be a defensible abstraction of these mechanisms. We developed a radiological source-term model and implemented it into a non-isothermal flow and transport simulator. While the proposed source-term model is applicable to various waste forms, canister systems, and disposal concepts, we specifically considered radionuclide releases from vitrified high-level waste placed in a cylindrical canister disposed in a deep vertical borehole repository. In this model, waste degradation is a function of temperature, and it can be adjusted to evaluate the influence of and propagate uncertainties in pH, passivation reactions, and chemical conditions as well as geometrical factors. The time-dependent, congruent release of safety-relevant radionuclides present in the decaying inventory is then calculated. Finally, the radionuclides are mobilized by diffusive and advective transport according to the thermo-hydraulic conditions prevailing in the near field of the repository, from where they migrate through the geosphere to the accessible environment. We examine the influence of the source-term model’s parameters on performance assessment calculations through sensitivity and uncertainty propagation analyses, identifying influential factors and confirming the upper bound of their impact. These considerations align with the overarching goal of repository design, which is to demonstrate that engineered and natural barriers can collectively delay radionuclide migration for timescales far exceeding human planning, thereby providing multiple, redundant barriers against environmental contamination.

iTOUGH2↗

Panel Session 116: Waste Management from Remediation of Legacy Sites or Unplanned Releases

Past experiences have shown that there is often a disconnect between Indigenous societies and Western society. This disconnect can be attributed to different value systems between these two group and is not limited to the USA and Canada but also extends into Australia, New Zealand, and the Central Asia Republics (the 'five Stans'). The intent of this Panel Session to give voice to representatives from Indigenous communities, the challenges presented in remediation processes, and potential solutions. Standard risk assessment assumptions developed for Western societies often do not adequately address the cultural and spiritual values that may be unique to indigenous societies. This panel focused on the perspectives of Indigenous societies (e.g., Indian Tribes (US), First Nations (Canada)) related to risk assessment of environmental remediation challenges, including deep geological disposal of HLW and ILW, near-surface disposal of LLW and in-situ entombment of nuclear reactor components, and environmental remediation of abandoned uranium mines and historic wastes. Standard risk assessment assumptions often do not adequately address the cultural and spiritual values that may be unique to indigenous societies. Factoring those into a risk assessment that still meets regulatory requirements and policies such as reasonably anticipated land use at US EPA CERCLA sites can be challenging but has the potential to lead to a site or facility approach that is more accepted. Panelists with presentations: The U.S. Nuclear Regulatory Commission Regulatory Process - A Path for Indigenous Knowledge? (Francis Cameron); Local Community Perceptions and Remediation Projects (Michelle Roberts); Indigenous Perspectives on Risk Assessment (Bob Watts); Navajo Superfund Program: Fundamental Law and Traditional Ecological Knowledge (Dariel Yazzie); Risk Assessment: Two Indigenous Approaches (Whitney Fraser)

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Panel Session 120: Roundtable: Waste Management - Energy Facilities Contractor Operating Group (EFCOG WM)

This Roundtable focused on the WM EFCOG working group as they met to discuss a variety of issues of importance to US DOE waste management operations. The purpose of the WM EFCOG is to seek out and promote the best management and operating practices, cost effective technologies and disposal options for all waste streams generated at US DOE facilities whether destined for US DOE or commercial facilities. Each site representative provided a radioactive waste management lessons learned briefing. Panelists with presentations: Savannah River Site (SRS) Solid Waste Program Update (Kerri Crawford); Idaho Clean-up Project Status (Frederick Hughes); West Valley Demonstration Project Update - EFCOG Waste Management Working Group (John Rendall); Key Project Challenges and Successes (Kalli Shupe); EFCOG WMWG Challenging Waste Subgroup Update (Tammy Monday); CH TRU Difficult Waste Stream Processing Challenges (Daniel Pancake); Argonne Building 350 Clean-up Success Using Effective Work Control (Daniel Dilday); Problematic Waste Streams on the Oak Ridge Reservation (John Wrapp)

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Radioactive decay of $\mathrm {{}^{90}Sr}$ in cement: a non-equilibrium first-principles investigation

Cement is an inexpensive and relatively easily manageable material that is used as a last barrier for nuclear waste disposal. Under these conditions, the cement is in contact with low radiation doses, but there is a distinct possibility of being contaminated with radioactive products. Of particular concern is the medium lived half-life product 90 Sr (28.8 years) due to its ability to replace Ca. 90 Sr undergoes β-decay to 90 Y which, in turn, β-decays to stable 90 Zr. In this work, we discuss systematically the chain of non-equilibrium processes that result as a consequence of β-decay events in cement. We first use density functional-based methods to study the consequences of the sudden increase of the nuclear charge from Z to Z+1, a possible induced ionization and the perturbation of the surrounding electronic charge. Secondly, we use molecular dynamics simulations to study the recoil of the daughter nucleus. Finally, we discuss the damage caused by the ionization cascade produced during the propagation of the β-electron and the resulting chemical and structural perturbation.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

Initial Testing of Alkaline Earth Metal Ion Absorption on Crystalline Silicotitanate - 20440

The Tank Closure Cesium Removal process at the Savannah River Site (SRS) has processed aqueous tank waste using the inorganic ion exchange media IONSIV{sup TM} R9120-Ba (which is also known as Crystalline Silicotitanate (CST)). Salt-cake in Tank 10H at SRS was dissolved and processed through filters and ion exchange columns. The primary purpose of the process is to remove Cs-137 from the aqueous waste so that it can be disposed as low level waste. It is known that this inorganic media also absorbs strontium from solution, and that strontium competes with cesium ions for absorption sites. The strontium ion is actually more strongly absorbed than the cesium ion from typical tank waste. However, strontium is typically present in low concentrations so does not normally cause a significant impact. Strontium is present as both non-radioactive isotopes and the radioactive Sr-90 isotope; with the non-radioactive isotopes being much more abundant. Although the total strontium solubility is usually much lower than cesium, some tank waste compositions can have a high enough soluble strontium concentration to decrease the cesium absorption. Relatedly, some testing at SRS suggested that another alkaline earth metal, calcium, may also absorb onto CST and may decrease cesium absorption. Barium is also an important species in treatment of tank waste, but is also usually present at low concentrations. However, after the Cs-137 is absorbed onto CST, it emits a beta particle and converts to Ba-137m, which then decays to non-radioactive Ba-137 by emission of a gamma ray. If the Ba-137m were to desorb quickly, it could impact the dose rate in down-stream equipment. In order to understand the impact of these alkaline earth metals on CST, SRNL performed testing using simulants of SRS tank waste that contain soluble barium, strontium, and calcium. Testing examined both removal of the alkaline earth metals and their impact on removal of cesium. Testing involved first developing realistic waste simulant formulations and dissolving the alkaline earth metals to high enough concentrations to potentially impact the Cs absorption. Once the formulations were developed and prepared, computer modeling was used to calculate the expected Cs absorption behavior to determine if the alkaline earth metals impact the performance. Measurements of the alkaline earth metals absorption by the media is also important for disposition of spent media because of the added radionuclide inventory from the Sr-90. These initial tests are examining the general impact and will be used to determine if further testing or measurements are needed. (authors)

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Removal of iodine (I- and IO3-) from aqueous solutions using CoAl and NiAl layered double hydroxides

The treatment of radioactive iodine released from nuclear power plants and radiological waste disposal sites is of great concern due to its high mobility and toxicity. In particular, iodide (I-) and iodate (IO3-) are the major iodine species of concern under various pHs and groundwater conditions. Herein, CoAl and NiAl layered double hydroxides (LDHs) were synthesized by a hydrothermal method and investigated to identify the removal mechanisms and efficiencies of both I- and IO3-. Both CoAl and NiAl LDHs exhibited rapid iodine removal processes within 20 min, following the pseudo-second-order model via ion-exchange with parent NO3- anion in the LDHs. The CoAl LDH’s maximum sorption capacities for I- and IO3- were about 1.67 and 2.16 mmol g-1, respectively, whereas for the NiAl LDH, these were about 2.10 and 2.26 mmol g-1, and they followed the Langmuir isotherm model. Interestingly, both the CoAl and NiAl LDHs showed a preferential ion-exchange affinity for IO3- over I-, which was attributed to the structural similarity of the IO3- and NO3- as well as new formation of secondary Co(or Ni)(IO3)2·2H2O phases. In addition, a desorption study indicated that the selectivity order was SO42- = IO3- = OH- > HCO3- > Cl- > NO3- = I- and demonstrated the higher retention of the IO3- than I- anion. This study provides insights into promising iodine sorbents and the different removal mechanisms of I- and IO3- using CoAl and NiAl LDHs.

Kang, Jaehyuk↗

The Effect of DPC Fillers on FEPs Relevant to Disposal of SNF

The US Department of Energy (DOE) is investigating the use of different materials that could be used to fill the void space inside a dual-purpose canister (DPC) loaded with spent nuclear fuel (SNF) just before it is emplaced in a deep geologic repository. The purpose of adding filler material is to maintain subcritical conditions in the repository during the postclosure period, which can span up to 1,000,000 years. Several types of materials have been proposed, including metals, cements, particulates, and glass. Part of this investigation addresses how the presence of filler material inside a DPC will affect the performance of the repository with respect to the repository features; the consequences of events that may occur; and the multiple thermal, hydrologic, chemical, and mechanical processes that may occur in a deep geologic repository over long timescales. This report describes some of the filler materials that have been proposed and studied; identifies 11 features, 6 events, and 25 processes that may be affected by the presence of filler materials; and discusses the effects that may require consideration for each feature, event, or process. The results of this study can be used to direct appropriate research and to develop suitable models if the DOE decides to use fillers to maintain subcritical conditions in DPCs used to dispose of SNF.

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RH-TRU Waste Streams: HFDA Crucibles Create New Complex Waste Stream

This project focuses on the removal of the Hot Fuel Dissolution Apparatus (HFDA) crucibles from the Hot Fuel Examination Facility’s (HFEF) hot cells to create space for upcoming projects. The project team has created a crucible and pyro-salt non-debris waste stream that can be disposed of at Waste Isolation Pilot Plant (WIPP). Chemical Compatibility Examinations (CCEs) must be performed on candidate waste by reviewing the types of chemicals used in the experiments, information found in INL Process Knowledge Summary Reports (PKSRs), and Environmental Protection Agency guides on chemical compatibility. The CCEs identify which chemicals can be safely combined to ensure adverse reactions do not occur in the waste packages. The CCEs are then used to update the PKSR, Basis of Knowledge, and Acceptable Knowledge documents and demonstrate compliance with the WIPP Waste Acceptance Criteria. Once documents are finalized, they are reviewed and approved by the WIPP Central Characterization Project (CCP) to ensure proper waste stream characterization. After the documents are approved, the crucibles can be packaged and shipped to the Idaho Nuclear Technical and Engineering Center for storage awaiting certification. CCP will certify the waste using the approved documents and a series of confirmatory testing before being shipped and disposed in WIPP’s deep geological repository. Currently, the CCE is in progress and the PKSRs are being updated.

Basis of Knowledge↗

Sludge Processing Options for early HLW Treatment at Hanford

The U.S. Department of Energy’s (DOE) Hanford Site has 177 underground storage tanks that contain wastes from past nuclear fuel reprocessing and waste-management operations. Over 20% of this waste is in the form of an insoluble sludge that will require solids concentration and washing prior to vitrification for long-term disposal. An assessment of potential flowsheet operations to support feed preparation activities prior to high level waste (HLW) vitrification has been conducted to better evaluate pretreatment processing options. Settling studies assessing the baseline approach of a settle-decant method were explored as well as a crossflow filtration system to be used alternatively for concentrating and washing HLW sludge. Significant variations in behavior of settling rates and sludge characteristics give reason to evaluate alternative pretreatment options for the HLW. Non-radioactive sludge containing iron oxide, boehmite, and gibbsite were evaluated via gravity settling and crossflow filtration to determine the behavior of these compounds in various tank waste matrices. Understanding the predictive capabilities of HLW solids settling as well as sludge concentration via crossflow filtration can help provide technical guidance during flowsheet planning.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Disposal Concepts for a High-Temperature Repository in Shale.

Disposal of large, heat-generating waste packages containing the equivalent of 21 pressurized water reactor (PWR) assemblies or more is among the disposal concepts under investigation for a future repository for spent nuclear fuel (SNF) in the United States. Without a long (>200 years) surface storage period, disposal of 21-PWR or larger waste packages (especially if they contain high-burnup fuel) would result in in-drift and near-field temperatures considerably higher than considered in previous generic reference cases that assume either 4-PWR or 12-PWR waste packages (Jové Colón et al. 2014; Mariner et al. 2015; 2017). Sevougian et al. (2019c) identified high-temperature process understanding as a key research and development (R&D) area for the Spent Fuel and Waste Science and Technology (SFWST) Campaign. A two-day workshop in February 2020 brought together campaign scientists with expertise in geology, geochemistry, geomechanics, engineered barriers, waste forms, and corrosion processes to begin integrated development of a high-temperature reference case for disposal of SNF in a mined repository in a shale host rock. Building on the progress made in the workshop, the study team further explored the concepts and processes needed to form the basis for a high-temperature shale repository reference case. The results are described in this report and summarized..

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Cigeo Design Options, Risks and Opportunities Management: A Continuous Project Management Goal - 20015

Andra is currently completing the detailed engineering phase of the stepwise design development of Cigeo, the French HLW and IL-LLW Deep Geological Repository. If licensed in the coming years, Cigeo's progressive construction, industrial operation and closure will last over more than a century. Such a duration, associated with the requirements in terms of reversibility, lead to the choice of an incremental design that allows, for next operator generations, to integrate improvements derived from scientific and technical progress as well as from experience feedback from its operation. This incremental design approach, associated with a permanent risks and opportunities management process, will continue after licensing of Cigeo as a tool for the governance of the reversibility. (authors)

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GDSA PFLOTRAN Development (FY2021)

The Spent Fuel & Waste Science and Technology (SFWST) Campaign of the U.S. Department of Energy (DOE) Office of Nuclear Energy (NE), Office of Spent Fuel & Waste Disposition (SFWD) is conducting research and development (R&D) on geologic disposal of spent nuclear fuel (SNF) and high-level nuclear waste (HLW). A high priority for SFWST disposal R&D is to develop a disposal system modeling and analysis capability for evaluating disposal system performance for nuclear waste in geologic media. This report describes fiscal year (FY) 2021 advances of the PFLOTRAN Development group of the SFWST Campaign. The mission of this group is to develop a geologic disposal system modeling capability for nuclear waste that can be used to probabilistically assess the performance of generic disposal concepts. In FY 2021, development proceeded along three main thrusts: software infrastructure, code performance, and process model advancement. Software infrastructure improvements included implementing an Agile software development framework and making improvements to the QA Test Suite. Code performance improvements included development of advanced linear and nonlinear solvers as well as design of flexible smoothing algorithms for capillary pressure functions. Process modeling advancements included the addition of flexible thermal conductivity function definitions and refinement of multi-continuum reactive transport to support Sandia’s participation in DECOVALEX

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Image Classification Using Convolutional Neural Networks to Automate Visual Inspection of CCO Containers (Rev. 1)

The Savannah River Site is automating the receipt and inspection of Criticality Control Overpack drums to aid in nuclear waste disposal. The Advanced Engineering group is automating this receipt and inspection process to assist the site’s goals. Part of this process includes visually inspecting CCO drums to ensure that no defects or security risks are present. This research focuses on developing a machine learning model to automatically classify drums as passing or failing inspection. The machine learning model was implemented using the open-source Tensorflow library and uses a Convolutional Neural Network architecture to extract features and differentiate between images. The model was able to achieve an average of 85% accuracy on a dataset of 288 drums. This project’s goal was to prove the validity of computer vision in an automation process and provide a starting place for continued research.

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Efficient Multi-Scale, Thermo-Hydro-Mechanical Numerical Model for Simulation of Long-Term Stability of Rooms for Nuclear Waste Disposal

This project demonstrated the feasibility of developing a numerical modeling tool that provides an efficient and unique approach to analyze the geomechanical performance of a deep geological repository (DGR) for high-level nuclear waste. The numerical modeling tool is physics-based and models the thermo-hydro-mechanical (THM) processes during the required time (up to one million years) for a DGR. The unique capability of the developed code is explicit representation of fracturing and a discrete fracture network (DFN) in the model that includes the entire repository layout for simulated time of one million years. Therefore, it can potentially address one of the challenges of analysis of a DGR, which is efficient numerical simulation of a multi-scale, physics-based models.

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Stabilization/Solidification of Tritiated Water - 20135

Tritiated water was stabilized/solidified by using calcium oxide and a polymer. First, tritiated water was changed into the powder form of calcium hydroxide by addition of tritiated water into calcium oxide. Then the produced calcium hydroxide was solidified in waste form by using a polymer binder. The specimens of waste form prepared by surrogate calcium hydroxide were subjected to compressive strength testing and water immersion testing. In addition, the specimens of waste form prepared by tritiated calcium hydroxide were subjected to leach testing. The mean compressive strength of the specimens containing 40 wt % of the surrogate calcium hydroxide was about 98.5 MPa (14,300 psi). After water immersion testing, there was no distinct change in the appearance of the specimens. The mean compressive strength of the specimens after water immersion was 79.1 MPa (11,470 psi). The mean diffusion coefficient of tritium for the specimens containing 40 wt% of tritiated calcium hydroxide was 5.4 x 10-10 cm{sup 2}/s. These findings indicate that the waste form has enough mechanical integrity and durability for final disposal and releases tritium very slowly. (authors)

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International Collaboration Activities in Geologic Disposal Research: FY2021 Progress

This report describes the FY21 status of international collaboration regarding geologic disposal research in the Spent Fuel and Waste Disposition (SFWD) Campaign. Since 2012, in an effort coordinated by Lawrence Berkeley National Laboratory, SFWD has advanced active collaboration with several international geologic disposal programs across the world. Such collaboration has allowed the SFWD Campaign to benefit from a deep knowledge base in regards to alternative repository environments developed over decades and has provided a framework for active peer-to-peer research participation in international groups which conduct, analyze, and model performance-relevant processes. Via international collaboration, the SFWD Campaign also benefits from substantial international investments in research facilities (such as underground research laboratory testing and modeling) and achieves cost savings via joint funding of expensive field experiments. To date, SFWD’s International Disposal R&D Program has established formal collaboration agreements with multiple international initiatives and various international partners, and national lab scientists associated with SFWD have conducted a large number of specific collaborative R&D activities that align well with its R&D priorities.

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