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At least 613 records · Page 34
All sourcesNASA NTRSDOE OSTI
DOE OSTI2019

Development and Demonstration of Monitoring Strategies and Technologies -Modern2020 Project - 19317

The Modern2020 project aims at providing the means for developing and implementing an effective and efficient repository operational monitoring program, taking into account the requirements of specific national programs. The work allows advanced national radioactive waste disposal programs to design monitoring systems suitable for deployment when repositories start operating in the next decade and supports less developed programs and other stakeholders by illustrating how the national context can be taken into account in designing dedicated monitoring programs tailored to their national needs. The work has been established to understand what should be monitored within the frame of the wider safety cases and to provide methodology on how monitoring information can be used to support decision-making and to plan for responding to monitoring results. Several developments on innovative repository monitoring techniques (wireless data transmission, alternative power supply sources, new sensors, geophysical methods) has been done from the proof of feasibility stage to the technology development and demonstration phase. Innovative technical solutions facilitate the integration and flexibility of required monitoring components to ease the final implementation and adaptation of the monitoring system. Full-scale in-situ demonstrations of innovative monitoring techniques were done in way to evaluate the performance of the state-of-the-art, the innovative techniques and their comparison with conventional ones. Finally, we will introduce the experiences of the local public stakeholders that have been involved in the Modern2020 project by focusing on the participation in practice of the Belgian, French, Finnish and Swedish local stakeholders, This paper provides an overview of the European Commission Horizon2020 Modern2020 project's main goals and progress to date. (authors)

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

Clay Generic Disposal System Model - Sensitivity Analysis for 32 PWR Assembly Canisters (+2 associated model files).

The Used Fuel Disposition Campaign (UFDC), as part of the DOE Office of Nuclear Energy’s (DOE-NE) Fuel Cycle Technology program (FCT) is investigating the disposal of high level radioactive waste (HLW) and spent nuclear fuela (SNF) in a variety of geologic media. The feasibility of disposing SNF and HLW in clay media has been investigated and has been shown to be promising [Ref. 1]. In addition the disposal of these wastes in clay media is being investigated in Belgium, France, and Switzerland. Thus, Argillaceous media is one of the environments being considered by UFDC. As identified by researchers at Sandia National Laboratory, potentially suitable formations that may exist in the U.S. include mudstone, clay, shale, and argillite formations [Ref. 1]. These formations encompass a broad range of material properties. In this report, reference to clay media is intended to cover the full range of material properties. This report presents the status of the development of a simulation model for evaluating the performance of generic clay media. The clay Generic Disposal System Model (GDSM) repository performance simulation tool has been developed with the flexibility to evaluate not only different properties, but different waste streams/forms and different repository designs and engineered barrier configurations/ materials that could be used to dispose of these wastes.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2025

Glass-Bonded Monazite Waste Forms for Lanthanide and Actinide Immobilization: From Theoretical Design to Scale-Up Production and Characterization

The development of nuclear waste forms for both existing and future nuclear wastes is critical to ensuring global environmental safety. This study focuses on waste management from molten salt reactors, where fuel exists in a salt form and could be processed in real time for the removal of neutron poisons such as xenon isotopes (e.g., 135 Xe) and rare earth elements (REEs, e.g., 149 Sm). To ensure safe, stable, and long-term disposal in geological repositories, REEs must be incorporated into a durable waste form. Iron-phosphate glasses are a promising candidate due to their low melting points, high chemical durability, and their ability to incorporate high concentrations of REEs. In this study, we successfully prepared iron-phosphate glass waste forms with high Nd loadings (up to 37 mass %) in batch sizes ranging from small (23 g) to large (1600 g). The resulting materials contained up to 75 mass % NdPO 4 , contributing to their mechanical resilience and exceptional chemical durability. These findings highlight the potential of iron-phosphate glasses as high-efficiency, chemically durable waste forms and demonstrate the successful transition from theoretical design to scaled-up production.

amorphous materials↗
DOE OSTI2015

Encapsulation of ILW metals by melting technology - 15360

This paper outlines how encapsulation using melting technology can be performed and describe the positive effects in a handling and repository perspective. Based on a feasibility study of a conditioning method for reactor core components and other ILW metals with a high content of long-lived radionuclides, the paper gives an overview of the concept and how the method can be facilitated. The paper covers identification and analysis of alternative techniques that supports the aim from a repository perspective, which is to strengthen the defense in depth for long-lived waste to be disposed. Embedding of radioactive metallic ILW components can isolate the activated component from corrosion and leakage of radionuclides into the repository barrier systems. By implementing a system that delays the mobility of nuclides the leakage to the biosphere can be delayed and by then decreased significantly. Conditioning by embedding has potential to strengthen a repository concept significantly. It may in certain cases also reduce the requirements on the repository barriers. In addition to the above, embedding also serves as an efficient radiation protection, which simplifies post treatment handling of the waste. The concept can be applied on a wide spectrum of metallic components. Positive effects in a handling and repository perspective can be achieved if low level metal not subject to free release is used as embedding material, since conditioning of metallic ILW waste can be carried out at significant lower dose rates and without increasing the total amount to be disposed. The paper discuss why there are embedding alternatives for ILW metals that are considered to be both technically feasible and reasonable in cost both with regards to investment and operational cost. A cost-benefit analysis as well as a risk analysis is discussed in order to support the feasibility assessment of a facility for encapsulation of ILW metals by melting technology. (authors)

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

Fuel Cycle Study to Developing a Viable Long Term Nuclear Power Scenario for the UK

Within the UK a closed fuel cycle has been in operation for decades but with the introduction of a new build programme to create 16 GWe capacity using LWR technology and with the closure of the UK's reprocessing industry, an open fuel cycle will be implemented beyond 2020. However, it is expected that 16 GWe nuclear capacity will not be enough for the UK to curb greenhouse gas emissions through increased demand for low carbon electricity, and tackle energy security. As such further nuclear expansion scenarios are being considered between 40 and 75 GWe by 2050. This study evaluates a viable 40 GWe UK nuclear scenario, evaluating the extent to which key drivers can impact a future UK nuclear energy strategy, including; the ability to sustain a relatively high nuclear capacity, fresh uranium availability, accumulation of spent fuel at decommissioning nuclear reactor sites and capacity requirements of future GDF host sites. The thermal reactor open fuel cycle is the de-facto standard around the world for many reasons including the reliance on extremely mature technology, significant operational experience and most importantly a relatively low cost. However, the premise that a thermal reactor open fuel cycle is the correct one to adopt long term is dependent on the notion that large quantities of UO{sub 2} spent fuel can eventually be disposed of in a geological disposal facility and the future availability of fresh uranium. The total amount of available uranium is unknown and if extraction from seawater proves to be economic and practical then any future concerns regarding supply disappear. Spent fuel disposability on the other hand has never been performed on a commercial scale and there is some uncertainty when and whether this can be achieved. This paper will follow from previous work already presented in 2013 [REF] which examined UK fuel cycles involving the transition to a closed fast reactor fleet by the end of this century. The revised scenarios now assume fast reactors are required later in the scenario, in-line with other countries. In addition, the new study has used physics methods originally developed by SKB to assess the physical size of the repository required to dispose of all intermediate and high level waste whilst ensuring temperature limits imposed by the engineered barriers and surrounding geology are adhered to. The disposition of less Am-241 in the repository will no doubt lead to a reduction in GDF footprint. However, this may not be important to GDF design and economics. The purpose of this paper is to also critically review the importance of and put into perspective a smaller GDF footprint and lower fissile inventory. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Conception and Operation of a New Treatment Center for Radioactive Waste from High Energy Accelerators - 18293

CERN has operated high-energy accelerators for more than 60 years of research in fundamental physics. The side-product of this activity is the generation of low- and intermediate-level radioactive waste, most of which is in solid form and free of surface contamination. Today, one operational year produces about 400 tons of radioactive waste which need to be disposed of towards the final repositories of the two CERN Host States, France and Switzerland. A new facility for the collection, interim storage, characterization and treatment of radioactive waste became operational at CERN in 2015. This facility was specifically conceived for an accelerator complex with a wide range of challenging waste items, which are often unique in their irradiation history as well as in their physical, chemical and radiological characteristics. The disposal rate was ramped up in 2016 with a unique treatment process tailored for metallic low-level waste. In 2017, two parallel processes were successfully tested for the elimination of radioactive cables and for the recycling of valuable metals. This paper gives an overview of the challenges in the design, commissioning and operation of the CERN radioactive waste treatment center. It outlines the methodology followed for the conception of the elimination processes, which can be of guidance for the development of similar facilities. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

Reactive transport in the complex heterogeneous alluvial aquifer of Fortymile Wash, Nevada

Yucca Mountain, Nevada, had been extensively investigated as a potential deep geologic repository for storing high-level nuclear wastes. Previous field investigations of stratified alluvial aquifer downstream of the site revealed that there is a hierarchy of sedimentary facies types. There is a corresponding log conductivity and reactive surface area subpopulations within each facies at each scale of sedimentary architecture. Here in this paper, we use a Lagrangian-based transport model in order to analyze radionuclide dispersion in the saturated alluvium of Fortymile Wash, Nevada. First, we validate the Lagrangian model using high-resolution flow and reactive transport simulations. Then, we used the validated model to investigate how each scale of sedimentary architecture may affect long-term radionuclide transport at Yucca Mountain. Results show that the reactive solute dispersion developed by the Lagrangian model matches the ensemble average of numerical simulations well. The link between the alluvium spatial variability and reactive solute dispersion at different spatiotemporal scales is demonstrated using the Lagrangian model. Finally, the longitudinal dispersivity of the reactive plume can be on the order of hundreds to thousands of meters, and it may not reach its asymptotic value even after 10,000 years of travel time and 2–3 km of travel distance.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2021

Modelling yield cap evolution in sandstone based on brittle creep experiments

The Zenifim Formation is being considered as a potential disposal formation for a deep borehole nuclear repository concept in Israel. Site selection and repository construction are intended to ensure that waste is separated from circulating groundwater, but long-term deformation of the wellbore could potentially create fluid flow pathways. To understand how time-dependent rock strength could affect wellbore stability, here we conducted creep tests under low to moderate confining pressures on retrieved core from the Zenifim formation. During creep, samples strain slowly as gradual damage accumulation progressively weakens the samples. Failure eventually occurred through the near-instantaneous formation of a shear fracture. Experimental results were used to calibrate a continuum damage poro-elastic model for sandstones. The calibrated damage-poro-elastic model successfully simulates different types of loading experiments including quasi-static and creep. The state of strain in experiments is close to yield during loading as the yield cap continuously evolves with damage accumulation. For creep tests, most damage occurs during triaxial loading. Minor damage accumulation occurs under constant load until the final stage of creep, where damage accelerates and promotes unstable fracturing.

58 GEOSCIENCES↗
DOE OSTI2024

Machine learning surrogates for surface complexation model of uranium sorption to oxides

Abstract The safety assessments of the geological storage of spent nuclear fuel require understanding the underground radionuclide mobility in case of a leakage from multi-barrier canisters. Uranium, the most common radionuclide in non-reprocessed spent nuclear fuels, is immobile in reduced form (U(IV) and highly mobile in an oxidized state (U(VI)). The latter form is considered one of the most dangerous environmental threats in the safety assessments of spent nuclear fuel repositories. The sorption of uranium to mineral surfaces surrounding the repository limits their mobility. We quantify uranium sorption using surface complexation models (SCMs). Unfortunately, numerical SCM solvers often encounter convergence problems due to the complex nature of convoluted equations and correlations between model parameters. This study explored two machine learning surrogates for the 2-pK Triple Layer Model of uranium retention by oxide surfaces if released as U(IV) in the oxidizing conditions: random forest regressor and deep neural networks. Our surrogate models, particularly DNN, accurately reproduce SCM model predictions at a fraction of the computational cost without any convergence issues. The safety assessment of spent fuel repositories, specifically the migration of leaked radioactive waste, will benefit from having ultrafast AI/ML surrogates for the computationally expensive sorption models that can be easily incorporated into larger-scale contaminant migration models. One such model is presented here.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2022

Reduced diffusion and enhanced retention of multiple radionuclides from pore structure characterization of barrier materials for enhanced repository performance

Fluid flow and chemical transport in porous media are the macroscopic consequences of pore structure, which integrates geometry (e.g., pore size and surface area, pore-size distribution) and topology (e.g., pore connectivity). Low-permeability geological media whose pores are poorly interconnected will exhibit the characteristics of anomalous diffusion and sample size-dependent effective porosity, which will strongly impact long-term net diffusion and retention of radionuclides in geological repository settings involving different host rocks and barrier materials. A suite of innovative and complementary experimental approaches is utilized to study the microscopic pore structure and macroscopic fluid flow & chemical transport for a range of host rocks and barrier materials, in addition to standard clay minerals and reference rocks. With a particular focus on quantifying the presence and magnitude of “isolated” pores for a reduced effective porosity in low-permeability geomedia, the integrated methodologies for basic properties and pore structure characterization of these geomedia include X-ray diffraction, thin section petrography, grain size distribution, water immersion porosimetry after vacuum-pulling for full saturation, mercury intrusion porosimetry, nitrogen physisorption, scanning electron microscopy, X-ray computed tomography, and (ultra-)small angle neutron (X-ray) scattering. In addition, custom-designed gas diffusion, tracer recipe involving a range of anionic and cationic chemicals with subsequent analyses by laser ablation and inductively coupled plasma-mass spectrometry, along with batch sorption, column transport, and imbibition tests were conducted for coupled effects of pore structure and chemical retention/transport. From the perspectives of pore structure in conjunction with multiple and complementary approaches to examining a range of sample sizes under different observational scales, we find that the poor pore connectivity is prevalent in low-permeability media (mudstone and crystalline rock) that is related to geological processes (e.g., compaction, diagenesis and thermal maturation). For example, the deep and organic matter-rich mudstones have a much smaller effective porosity than the total porosity (as a result of poor pore connectivity) and associated diffusion coefficient, and the effective porosity & diffusion coefficients are also dependent upon the sample sizes used in the measurement. Similarly, most of the pore space in the shallow mudstone is also controlled by pore-throat diameters in the 5-50 nm range of intergranular pore types from its fine-grained nature, but with an overall good pore connectivity. However, the nm-sized pore space (physically pore-network architecture) and strong sorption capacities (chemical retention from clay minerals) of both shallow and deep mudstones lead to the synergistic retention of cationic radionuclides and their utilities as effective host rocks and barrier materials. Our unique approaches of studying how the micro-scale pore structure affect macro-scale fluid flow, diffusion & retention, and chemical transport produce improved mechanistic understanding, and realistic quantification, of diffusion and retention of typical radionuclides in a range of generic host rocks and barrier materials (clay/shale, salt, crystalline rock, and tuff), with the overall results leading to scientifically-based understanding of enhanced isolation (from both diffusion and retention) of radionuclides and improved confidence on the long-term performance of geological repository to store high-level radioactive wastes. In addition to the training of 25 undergraduates, graduates, and postdocs of UTA, the scientists (organizations) involved in performing this work (e.g., discussion, sample sharing, and operation of SANS and SAXS instruments) include Ed Matteo, Yifeng Wang, and Kristopher Kuhlman (Sandia National Laboratories), Jens Birkholzer, Liange Zheng, Tim Kneafsey, and Sharon Borglin (Lawrence Berkeley National Laboratory), Mavrik Zavarin (Lawrence Livermore National Laboratory), Yukio Tachi and Yuta Fukatsu (Japan Atomic Energy Agency), Mieke de Craen (Euridice, Belgium), Markus Bleuel (NIST), Wei-Ren Chen, Gergely Nagy, Changwoo Do, William Heller, Larry Anovitz, and Kenneth Littrell (ORNL), as well as Jan Illvsky, Ivan Kuzmenko, Ju-Sang Park and Jon Almers (ANL). Key deliverables include a total of 13 peer-reviewed journal articles (nine published and three under review), 23 presentations at scientific conferences (AAPG, AAPG Southwest Section, AGU, Asian Clay Conference, GSA, GSA South-Central Section, IHLRWM, InterPore, International Conference on Chemistry and Migration Behavior of Actinides and Fission Products in the Geosphere, International Conference on Coupled Processes in Fractured Geological Media: Observation, Modeling and Application), and academic institutions (UTA, New Mexico State University; University of Poitiers, France; University of Helsinki, Finland; Uppsala University, Sweden; Istanbul Technical University, Turkey) and other organizations (Andra, France; Posiva Oy, Finland).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

The Enhanced Sealing Project, Monitoring a Full-Scale Shaft Seal at Canada's Underground Research Laboratory (2009-2015) - 16070

A full-scale shaft seal was designed and installed in the 5-m-diameter main shaft at Canada's Underground Research Laboratory (URL) in 2009. Located where the main shaft intersects a water-bearing fracture zone at a depth of approximately 273 m, a 6-m-thick clay component was sandwiched between two 3-m-thick concrete components. The clay component was 40:60 mixture of bentonite and quartz/feldspar sand and spans approximately 1 m beyond the maximum identified vertical extent of the fracture zone. Its purpose is to limit the mixing of saline groundwater from the deeper region with the fresher, near-surface groundwater. Its construction was part of Canada's Nuclear Legacy Liabilities Program (NLLP). The 1. phase of the Enhanced Sealing Project (ESP), a jointly-funded Research, Development and Demonstration (RD and D) project was started in 2009 to monitor the saturation of the shaft seal until the end of 2013. The 2. phase, extended monitoring of the ESP is supported by Posiva (Finland), Andra (France) and CNL (Canada) and this will run until the end of 2016. The slow evolution of the seal and its surroundings and complete removal of the surface facilities at the URL site in 2014, has meant that the ESP monitoring has evolved to a stand-alone, lower frequency data collection program. Information gained from the ESP monitoring is applicable to support the development of sealing technology to safely close a deep geological repository for the permanent disposal of radioactive wastes. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2022

Dehalogenation reactions between halide salts and phosphate compounds

Reactions between phosphoric acid [H 3 PO 4 ] or ammonium hydrogen phosphates [i.e., NH 4 H 2 PO 4 , (NH 4 ) 2 HPO 4 ] and halide salts can be used to dehalogenate (remove halides from) salt-based waste streams, where the process of removing halides yields products that have more efficient disposal pathways for repository storage. In this context, the term efficiency is defined as higher waste loadings and simplified immobilization processes with potential for recycle of certain salt components (e.g., 37 Cl as H 37 Cl or NH 4 37 Cl). The main streams identified for these processes are nuclear wastes generated during electrochemical reprocessing of used nuclear fuel as well as used halide salts from molten salt reactor operation. The potential byproducts of these reactions are fairly consistent across the range of halide species (i.e., F, Cl, Br, I) where the most common are hydrogen halides [e.g., HCl (g) ] or ammonium halides (e.g., NH 4 Cl). However, trihalide compounds (e.g., NCl 3 ), nitrogen triiodide ammine adducts [NI 3 ·(NH 3 ) x ], and ammonium triiodide (NH 4 I 3 ) are also possible. Several of these byproducts (i.e., NCl 3 , NBr 3 , NI 3 , and NH 4 I 3 ) are shock-sensitive contact explosives so their production in these processes must be tracked and carefully controlled, which includes methods of immediate neutralization upon production such as direct transport to a caustic scrubber for dissolution. Several benefits arise from utilizing H 3 PO 4 as the phosphate additive during dehalogenation reactions for making iron phosphate waste forms including more oxidized iron (higher Fe 3+ :Fe 2+ ratios), higher chemical durabilities, and the avoidance of trihalides, but the byproducts are hydrogen halides, which are corrosive and require special handling.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2015

Computational scheme for burnup credit applied to long term waste disposal - 14448

The application of burnup credit (BUC) to criticality safety evaluations (CSE) of geological repositories for long term disposal of spent nuclear fuel (SNF) is under consideration in Switzerland. Therefore, in collaboration with the national waste management organization, PSI is conducting research towards the development of a BUC methodology tailored for Swiss specific applications. As part of this research, a computational sequence from the in-core burnup phase, through the interim storage, until final disposal is being established. The tools selected for this methodology, referred to as BUCSS-R (Burnup Credit System for Swiss reactors and Repositories), are CASMO for the in-core fuel depletion, the decay module of SERPENT for the decay calculations, and MCNP for the criticality calculations. In connection to BUCSS-R, a computational interface to validated reference core models of the Swiss reactors is being developed. The purpose is to apply reconstructed depletion histories based on real core operation conditions with CASMO assembly calculations, to transfer the pin wise nuclide compositions at discharge to the subsequent SERPENT decay calculations, and afterwards to the final MCNP criticality calculations. An objective of this paper is to present the elements of the BUCSS-R methodology, and to show a first application to so-called bounding analyses of a preliminary Swiss disposal canister design loaded with fuel assembly types operated in a Swiss PWR reactor. The rationale for these bounding analyses is that before starting the integration of methodologies for bias and uncertainty quantifications, the span of the k-eff variation between fresh and burnt fuel loadings should first be quantified in a realistic manner. The obtained results are reported in this paper together with studies performed in order to evaluate the effects from burnup levels and nuclides credit at flooded conditions on the k-eff values. (authors)

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

Performance assessment modeling of a generic SNF/HLW repository in salt with coupled thermal-hydrologic effects - 15423

This paper describes advances in performance assessment modeling of deep geologic repositories facilitated by a massively parallel, high-performance computing (HPC) environment. Our new Generic Disposal System Analysis (GDSA) Framework utilizes the massively parallel PFLOTRAN multi-physics code to simulate repository performance in the presence of coupled thermal-hydrologic-chemical processes, linked to DAKOTA, an HPC uncertainty sampling and propagation code that provides nested parallelism for multi-realization performance assessment and sensitivity analysis. This enhanced performance assessment (PA) modeling capability is demonstrated with deterministic and probabilistic simulations of a generic repository for SNF and HLW in bedded salt host rock, by comparing repository performance between a case with heat-generating waste ('thermal' case) and a case without heat generation ('isothermal' case). The simulation results provide preliminary insights into the effect of multi-physics processes and thermal-hydrologic coupling on the long-term behavior of a reference-case salt repository. (authors)

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

Unsaturated alluvium disposal modelling with improved geological realism

This report describes specific activities in the Fiscal Year (FY) 2023 associated with the Geologic Disposal Safety Assessment (GDSA) Repository Systems Analysis (RSA) work package funded by the Spent Fuel and Waste Science and Technology (SFWST) Campaign of the U.S. Department of Energy Office of Nuclear Energy (DOE-NE), Office of Spent Fuel and Waste Disposition (SFWD). The overall objective of the GDSA RSA work package is to develop generic deep geologic repository concepts and system performance assessment (PA) models in several potential host-rock environments, and to simulate and analyze these generic repository concepts and models using the GDSA Framework toolkit and other tools as needed. The specific GDSA goal addressed in this report is reference case development, simulation, and analysis for the unsaturated alluvium (UZ), one of the four potential host-rocks considered by the GDSA. Further, we aim to exercise the simulation tools and methodologies under development by GDSA for PA modelling.

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

Framework for The Analysis of US Permanent Geologic Disposal Options for Radioactive Waste

The U.S. study of permanent geologic disposal options of spent nuclear fuel (SNF) and high-level radioactive waste (HLW) provided a technical basis for informing policy decisions regarding strategies for the management and disposal of radioactive waste requiring geologic isolation through the evaluation of potential impacts of waste forms on the feasibility and performance of representative generic concepts for geologic disposal. The goal of the study was to help inform relevant policy questions including: Is a "one-size-fits-all" repository a good strategic option for disposal? Do different waste types and forms perform differently enough in different disposal concepts that they warrant different treatment? Do some disposal concepts perform significantly better with or without specific waste types or forms?

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

Prediction of the thermal-hydraulic-mechanical response of a geological repository at large scale and sensitivity analyses

To better understand the Thermal-Hydraulic-Mechanical (THM) response of the rock caused by the high-level radioactive waste (HLW) released heat in the Callovo-Oxfordian formation (COx) in the near-field and far-field areas, a series of coupled THM modelling is performed. This study presents a case study based on the French geological repository for HLW (Cigéo project) to better assess modelling of deep geological repositories (DGR) within the DECOVALEX-2019 framework. In this study, the proposed coupled THM model, implemented in the platform of COMSOL, is validated by comparing the modelled THM results for a point heat source in an infinite rock mass with an analytical solution. It is also validated by comparing its thermal component for the Base Case with accurate results calculated using another numerical method. To shorten the calculation time, the DGR is modelled using six cuboid blocks representing 162 HLW cells and only six central HLW cells of interest modelled in detail. This simplification is validated. This model can easily be used to model not only the near-field THM response of the rock but also the far-field THM response. The influence of the boundary conditions applied on the gallery wall, HLW cell walls, and external surfaces of the model and the influence of different vertical dimensions of the geometry are studied. The sensitivity analyses of the THM parameters of COx on the THM response at different locations are performed. Finally, the influence of using a 2-dimensional (2D) model to represent a 3-dimensional (3D) repository is also investigated.

58 GEOSCIENCES↗
DOE OSTI2016

U.S. Senate Bill S.854-IS - A Maladjusted Politicized Maze for Consent-Based Siting of New HLW-Repositories - 16019

At the end of 2015, the USA's HLW-disposal program had been on hold since 2010 pending the enactment of enabling legislation for one or both of: 1) The Yucca Mountain HLW repository; and/or 2) The law proposed in Senate Bill S.854-IS 'to establish a new organization to manage nuclear waste, provide a consensual process for siting nuclear waste facilities, ensure adequate funding for managing nuclear waste, and for other purposes'. But, S.854-IS does not acknowledge the existence of the Yucca Mountain HLW repository, which, if re-enabled, would affect both when another HLW repository would be needed and its disposal capacity. Furthermore, it neither timely nor meaningfully integrates waste generators, facility-host(s) and other directly affected parties with the new 'implementing' organizations deemed to be more susceptible to political influence and interferences than those plaguing the Yucca Mountain HLW repository program since 1987 that resulted in a multitude of contentions, lawsuits, delays, and increased costs. For these and other reasons very compelling and conclusive to the author, the legal status of the YM HLW repository should be 'definitive' before a law based upon S.854-IS is enacted. Instead, the near-term focus should be on: 1. Defining 'consent-based' quantitatively. 2. Promptly siting and opening at least one large, expandable, 'consolidated' HLW-storage facility allowing the government to begin taking title to commercially-generated HLW; most-expeditiously accomplished by amending existing legislation. 3. Updating S.854-IS based on lessons learned about siting, developing, and licensing nuclear-waste-storage and -disposal facilities in the USA and abroad the past 30+ years. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗