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DOE OSTI2019

Managing Knowledge Over Time for Supporting Radioactive Waste Geological Disposal Project Development - 19028

Projects designed to define, build, operate and monitor radioactive waste disposal solutions require the transmission of knowledge over several decades and the preservation of memory over several centuries. This is particularly acute for the geological disposal project 'Cigeo' led by the French national radioactive waste management agency (Andra) as far as the requirements in terms of reversibility lead to the choice of an incremental design that allows to integrate learning resulting from continuous improvement of knowledge. For instance, for robustness purposes, Cigeo is designed on the basis of currently proven technologies, but its step-by-step construction in successive stages must make it possible to integrate improvements derived from the scientific and technical progress as well as from experience feedback from its exploitation. In the field of industrial activity, a significant proportion of the knowledge and skills implemented by employees of a company are not known or even identified by the human resources organization or the mid-management staff. This is mainly due to the fact that (i) the organized knowledge-sharing is often based on documentary tools but little on social modalities and (ii) the organization of knowledge is mostly cross-cutting to the managerial lines. To overcome this bias, a global and integrated approach to knowledge management is currently the focus of Andra's specific efforts. Cross-disciplinary groups and close intergenerational links positioned throughout the life of the projects aim specifically at continuous and participative knowledge sharing, capitalization and transmission. Then, devices aiming at systematic dissemination of and access to knowledge support experience feedback and provide unified collaboration, traceability and decision-support at all scales. (authors)

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DOE OSTI2019

Concept for Operation of the Belgian Geological Disposal Facility for Category B and C Wastes - 19247

For more than 40 years, ONDRAF/NIRAS (the federal agency responsible for managing radioactive waste and enriched fissile materials in Belgium) has been studying the concept of a geological disposal facility (GDF) in poorly indurated clays as an option for the long-term management of both category B waste (low-level and intermediate-level long-lived waste - LILW-LL) and category C waste (high-level waste - HLW and spent fuel). In the absence of a policy decision, the proposed Belgian reference option for disposal of such waste is a GDF in poorly indurated clays, located at a depth of between 200 m and 600 m. Within the framework of the first Safety and Feasibility Case (SFC-1) undertaken by ONDRAF/NIRAS, the organization is updating its former GDF layout, which dates back from 2003. The SFC-1 GDF concept is characterized by a central zone in which shafts, joined by a central connecting gallery, provide two surface connections. On opposing sides of the central gallery, four (two on each side) access galleries branch off to connect the disposal areas for B- and C-waste, respectively. Disposal dead end galleries, 400 m long, branch off perpendicular to the access galleries. Both waste areas will be operated during two separate time frames. The current underground layout has been significantly modified from the previous 2003 layout. The modifications are largely based on several R and D studies with regard to the future operations of the GDF carried out by BGE Technology GmbH in support to ONDRAF/NIRAS. This paper focuses on major operational processes such as waste transport and emplacement. In addition, the current conceptual design of the backfilling and closure concept is presented. Optimization studies on the further development of the operational process are also presented. (authors)

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DOE OSTI2019

Conceptual Design of the Belgian Geological Disposal Facility for Category B and C Wastes - 19341

For more than 40 years, ONDRAF/NIRAS (the federal agency responsible for managing radioactive waste and enriched fissile materials in Belgium) has been studying the concept of a geological disposal facility (GDF), within poorly indurated clays, as an option for the long-term management of both category B waste (low-level and intermediate-level long-lived waste - LILW-LL) and category C waste (high-level waste - HLW and spent fuel). The proposed Belgian reference option for disposal of such waste is a GDF within poorly indurated clays, located at a depth of between 200 m and 600 m. Within the framework of the first Safety and Feasibility Case (SFC-1) undertaken by ONDRAF/NIRAS, the former GDF layout, which dates back from 2003 has been updated. The first part of the paper describes the main drivers behind the evolution of the former 2003 GDF design towards the current reference SFC-1 design. It also details how these drivers, related to operational and constructional constraints, have affected the GDF layout, such as the disposal gallery length, the crossing shapes of the GDF galleries and the layout of access and waste shafts. The second part of the paper focuses on the engineering and construction challenges generated by GDF development within poorly indurated clays. This includes discussion of the impact of the specific mechanical characteristics such poorly indurated clays have on the construction methods used for the shafts, the rock support (concrete lining), the drift connections, the crossings and the global GDF design. These construction methods have been selected based on research conducted at the underground research laboratory HADES in Mol (Belgium). Finally, the main axes for further development of the GDF design within the SFC-1 framework is presented. (authors)

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DOE OSTI2012

Development of an Advanced Performance Assessment Modeling Capability for Geologic Disposal of Nuclear Waste: Methodology and Requirements

This report describes the planning and initial development of an advanced disposal system PA modeling capability to facilitate the science-based evaluation of disposal system performance for a range of fuel cycle alternatives in a variety of geologic media and generic disposal system concepts. The advanced modeling capability will provide a PA model framework that facilitates PA model development, execution, and evaluation within a formal PA methodology. The PA model framework will provide a formalized structure that enables (a) representation and implementation of a range of generic geologic disposal system options, (b) representation of subsystem processes and couplings at varying levels of complexity in an integrated disposal system model, (c) flexible, modular representation of multi-physics processes, including the use of legacy codes, (d) evaluation of system- and subsystem-level performance, (e) uncertainty and sensitivity analyses to isolate key subsystem processes and components, (f) data and configuration management functions, and (g) implementation in HPC environments.

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DOE OSTI2026

Lessons Learned from Site Selection Efforts for Deep Geological Disposal

This report discusses international experiences from site selection of deep geological repositories (DGR) for radioactive waste disposal. Based on experiences from three countries with recent or ongoing site selections, the report describes different site selection approaches and summarizes lessons learned. The German and Swiss programs are two slightly different examples of one end of the spectrum of potential siting approaches: Starting with the entire country, a technically based downselection is performed to select the repository site. The Canadian program is an example of the other end of the spectrum: Here, the program started with an extensive engagement program with Canadians to collaboratively develop a socially acceptable and technically sound national approach for the long-term management of used nuclear fuel. This was followed by the launch of a collaboratively developed consent-based siting process designed to seek an informed and willing host with suitable sites, considering both technical and community well-being site requirements.

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DOE OSTI2024

Scenario development for safety assessment in deep geologic disposal of high-level radioactive waste and spent nuclear fuel: A review

Radiation and radioactive substances result in the production of radioactive wastes which require safe management and disposal to avoid risks to human health and the environment. To ensure permanent safe disposal, the performance of a deep geological repository for radioactive waste is assessed against internationally agreed risk-based standards. Assessing postclosure safety of the future system's evolution includes screening of features, events, and processes (FEPs) relevant to the situation, their subsequent development into scenarios, and finally the development and execution of safety assessment (SA) models. Global FEP catalogs describe important natural and man-made repository system features and identify events and processes that may affect these features into the future. By combining FEPs, many of which are uncertain, different possible future system evolution scenarios are derived. Repository licensing should consider both the reference or “base” evolution as well as alternative futures that may lead to radiation release, pollution, or exposures. Scenarios are used to derive and consider both base and alternative evolutions, often through production of scenario-specific SA models and the recombination of their results into an assessment of the risk of harm. Furthermore, while the FEP-based scenario development process outlined here has evolved somewhat since its development in the 1980s, the fundamental ideas remain unchanged. A spectrum of common approaches is given here (e.g., bottom–up vs. top–down scenario development, probabilistic vs. bounding handling of uncertainty), related to how individual numerical models for possible futures are converted into a determination as to whether the system is safe (i.e., how aleatoric uncertainty and scenarios are integrated through bounding or Monte Carlo approaches).

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DOE OSTI2024

Modelling of failure and fracture development of the Callovo-Oxfordian claystone during an in-situ heating experiment associated with geological disposal of high-level radioactive waste

Not Available

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DOE OSTI2021

An Alternative Conceptual Model for the Spent Nuclear Fuel–Water Interaction in Deep Geologic Disposal Conditions

For the long-term safety assessment of direct disposal of spent nuclear fuel in deep geologic repositories, knowledge on the radionuclide release rate from the UO 2 matrix is essential. This work provides a conceptual model to explain the results of leaching experiments involving used nuclear fuel or simulant materials in confirmed reducing conditions. Key elements of this model are: direct effect of radiation from radiolytic species (including defects and excited states) in the solid and in the first water layers in contact with its surface; and excess H 2 may be produced due to processes occurring at the surface of the spent fuel and in confined water volumes, which may also play a role in keeping the spent fuel surface in a reduced state. The implication is that the fractional radionuclide release rate used in most long-term safety assessments (10 -7 year -1 ) is over estimated because it assumes that there is net UO 2 oxidation caused by radiolysis, in contrast with the alternative conceptual model presented here. Furthermore, conventional water radiolysis models and radiation chemical yields published in the literature are not directly applicable to a heterogeneous system such as the spent fuel–water interface. Suggestions are provided for future work to develop more reliable models for the long-term safety assessment of spent nuclear fuel disposal.

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DOE OSTI2015

Effects of pH on deposition rate of supersaturated silicic acid around geological disposal system - 15245

For the migration assessment of radionuclides around geological repository, it is important to estimate the dynamic behavior of silicic acid under a high alkalinity condition due to the cementitious materials for the construction of the repository. This study examined the pH dependencies of the deposition behavior of supersaturated silicic acid in the co-presence of solid phase. The parameters for the deposition experiments were as follow: the pH was set to 8, 9 and 10, the initial saturated concentration was set to 2-8 mM, the amount of amorphous silica (350 m{sup 2}/g) as solid phase was set to 0.1-1.0 g. The experimental results showed that the initial supersaturated concentrations strongly affected the formation of colloidal silicic acid, increasing the deposition amount of silicic acid and decreasing the soluble silicic acid with time. In other words, the polymerization and the deposition of silicic acid simultaneously arose in these experiments. Furthermore, this study evaluated the apparent deposition rates of silicic acid by assuming the first-order reaction equation considering the specific surface area. As a result, the deposition rate constants, k, were in a narrow range of 10{sup -11} to 10{sup -10} m/s under all experimental conditions in this study. This means that the deposition rate constants do not depend both on pH-value (in the range from 8 to 10) and the initial concentration of supersaturated silicic acid. That is, the estimated deposition rate constants are useful in estimation of the deposition behavior of supersaturated silicic acid, such as the clogging in fractures (flow-paths) around the repository, even if the pH-value of groundwater gradually or dramatically changes. (authors)

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DOE OSTI2013

Hydrologic Nuclide Transport Models in Cyder, A Geologic Disposal Software Library - 13328

Component level and system level abstraction of detailed computational geologic repository models have resulted in four rapid computational models of hydrologic radionuclide transport at varying levels of detail. Those models are described, as is their implementation in Cyder, a software library of interchangeable radionuclide transport models appropriate for representing natural and engineered barrier components of generic geology repository concepts. A proof of principle demonstration was also conducted in which these models were used to represent the natural and engineered barrier components of a repository concept in a reducing, homogenous, generic geology. This base case demonstrates integration of the Cyder open source library with the Cyclus computational fuel cycle systems analysis platform to facilitate calculation of repository performance metrics with respect to fuel cycle choices. (authors)

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DOE OSTI2013

Considerations of human intrusion in U.S. programs for deep geologic disposal of radioactive waste

Regulations in the United States that govern the permanent disposal of spent nuclear fuel and high-level radioactive waste in deep geologic repositories require the explicit consideration of hypothetical future human intrusions that disrupt the waste. Specific regulatory requirements regarding the consideration of human intrusion differ in the two sets of regulations currently in effect in the United States; one defined by the Environmental Protection Agency's 40 Code of Federal Regulations part 197, applied only to the formerly proposed geologic repository at Yucca Mountain, Nevada, and the other defined by the Environmental Protection Agency s 40 Code of Federal Regulations part 191, applied to the Waste Isolation Pilot Plant in New Mexico and potentially applicable to any repository for spent nuclear fuel and high-level radioactive waste in the United States other than the proposed repository at Yucca Mountain. This report reviews the regulatory requirements relevant to human intrusion and the approaches taken by the Department of Energy to demonstrating compliance with those requirements.

Swift, Peter N.↗

DOE OSTI2013

A Safety Case Approach for Deep Geologic Disposal of DOE HLW and DOE SNF in Bedded Salt - 13350

The primary objective of this study is to investigate the feasibility and utility of developing a defensible safety case for disposal of United States Department of Energy (U.S. DOE) high-level waste (HLW) and DOE spent nuclear fuel (SNF) in a conceptual deep geologic repository that is assumed to be located in a bedded salt formation of the Delaware Basin [1]. A safety case is a formal compilation of evidence, analyses, and arguments that substantiate and demonstrate the safety of a proposed or conceptual repository. We conclude that a strong initial safety case for potential licensing can be readily compiled by capitalizing on the extensive technical basis that exists from prior work on the Waste Isolation Pilot Plant (WIPP), other U.S. repository development programs, and the work published through international efforts in salt repository programs such as in Germany. The potential benefits of developing a safety case include leveraging previous investments in WIPP to reduce future new repository costs, enhancing the ability to effectively plan for a repository and its licensing, and possibly expediting a schedule for a repository. A safety case will provide the necessary structure for organizing and synthesizing existing salt repository science and identifying any issues and gaps pertaining to safe disposal of DOE HLW and DOE SNF in bedded salt. The safety case synthesis will help DOE to plan its future R and D activities for investigating salt disposal using a risk-informed approach that prioritizes test activities that include laboratory, field, and underground investigations. It should be emphasized that the DOE has not made any decisions regarding the disposition of DOE HLW and DOE SNF. Furthermore, the safety case discussed herein is not intended to either site a repository in the Delaware Basin or preclude siting in other media at other locations. Rather, this study simply presents an approach for accelerated development of a safety case for a potential DOE HLW and DOE SNF repository using the currently available technical basis for bedded salt. This approach includes a summary of the regulatory environment relevant to disposal of DOE HLW and DOE SNF in a deep geologic repository, the key elements of a safety case, the evolution of the safety case through the successive phases of repository development and licensing, and the existing technical basis that could be used to substantiate the safety of a geologic repository if it were to be sited in the Delaware Basin. We also discuss the potential role of an underground research laboratory (URL). (authors)

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DOE OSTI2013

Performance assessment for the geological disposal of Deep Burn spent fuel using TTBX

The behavior of Deep Burn Modular High Temperature Reactor Spent Fuel (DBSF) is investigated in the Yucca Mountain geological repository (YMR) with respect to the annual dose (Sv/yr) delivered to the Reasonably Maximally Exposed Individual (RMEI) from the transport of radionuclides released from the graphite waste matrix. Transport calculations are performed with a novel computer code, TTBX which is capable of modeling transport pathways that pass through heterogeneous geological formations. TTBX is a multi-region extension of the existing single region TTB transport code. Overall the peak annual dose received by the RMEI is seen to be four orders of magnitude lower than the regulatory threshold for exposure, even under pessimistic scenarios. A number of factors contribute to the favorable performance of DBSF. A reduction of one order of magnitude in the peak annual dose received by the RMEI is observed for every order of magnitude increase in the waste matrix lifetime, highlighting the importance of the waste matrix durability and suggesting graphite's utility as a potential waste matrix for the disposal of high-level waste. Furthermore, we see that by incorporating a higher fidelity far-field model the peak annual dose calculated to be received by the RMEI is reduced by two orders of magnitude. By accounting for the heterogeneities of the far field we have simultaneously removed unnecessary conservatisms and improved the fidelity of the transport model. (authors)

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DOE OSTI2012

Towards a defensible safety case for deep geologic disposal of DOE HLW and DOE SNF in bedded salt

Mackinnon, Robert James↗

DOE OSTI2022

Hydro-mechanical behavior of heated bentonite buffer for geologic disposal of high-level radioactive waste: A bench-scale X-ray computed tomography investigation

The behavior of heated bentonite buffer is critical for the security and long-term performance of a geological repository for high-level radioactive waste (HLW). While laboratory column experiments have been conducted to investigate compacted bentonite and coupled THMC (thermal-hydro-mechanical and chemical) processes for a moderate temperature range of up to 100 °C, data for a higher temperature range are limited. Understanding bentonite behavior and coupled THMC processes under higher temperatures (e.g., up to 200 °C) could allow for a more economic repository design and would expand the data and knowledge base for more reliable modeling. In this study, a bench-scale experiment was conducted in a compacted bentonite column experiencing both heating up to 200 °C in the center and hydration from a sand-clay boundary surrounding the column. During the experiment run for 1.5 years, frequent X-ray computed tomography (CT) scanning of bentonite provided insights into the spatiotemporal evolution of (1) hydration/dehydration, (2) clay swelling/shrinkage, (3) displacement, and (4) mineral precipitation. After the experiment, a comprehensive post-dismantling characterization of bentonite samples was conducted. Results showed that the bentonite hydration was axi-symmetrical despite the initial heterogeneity due to packing, confirming the ability of bentonite to seal fast flow/transport paths. Further, compared to a non-heated control experiment, the heated column showed greater CT density variations along the radial distance, indicating that homogenization of bentonite might be more difficult if a temperature gradient is maintained in the repository. Precipitation of an anhydrite layer occurred in the inner hot zone, pointing to potential concerns about salt precipitation causing canister corrosion. Ultimately, the experiments provided a high-resolution window into the strongly dynamic and coupled behavior of bentonite exposed to heating, hydration and swelling, which will be valuable for improving modeling of coupled processes, especially for the early state of a HLW repository.

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DOE OSTI2025

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↗

DOE OSTI2017

Experimental determination of solubilities of di-calcium ethylenediaminetetraacetic acid hydrate [Ca2C10H12N2O8·7H2O(s)] in NaCl and MgCl2 solutions to high ionic strengths and its Pitzer model: Applications to geological disposal of nuclear waste and other low temperature environments

In this study, solubility measurements on di-calcium ethylenediaminetetraacetic acid [Ca 2 C 10 H 12 N 2 O 8 (s), abbreviated as Ca 2 EDTA(s)] as a function of ionic strength are conducted in NaCl solutions up to I = 5.0 mol•kg –1 and in MgCl 2 solutions up to I = 7.5 mol•kg –1 , at room temperature (22.5 ± 0.5oC).

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DOE OSTI2018

Corrigendum to “Experimental Determination of Solubilities of Di-calcium Ethylenediaminetetraacetic Acid Hydrate [Ca 2 C 10 H 12 N 2 O 8 •7H 2 O(s)] in NaCl and MgCl 2 Solutions to High Ionic Strengths and Its Pitzer Model: Applications to Geological Disposal of Nuclear Waste and Other Low Temperature Environments”

In the published version (Xiong et al., 2017), there was an error for the reaction coefficient for the dissolution reaction of Ca 2 C 10 H 12 N 2 O 8 •7H 2 O(s) in the database used for the modeling. In the database for the modeling, the coefficient for water (i.e., 7H 2 O) was inadvertently omitted.

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