Search NASA⌕ Search

SEARCH · Search NASA

Results for “Repository”

Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 127 records · Page 7
All sourcesNASA NTRSDOE OSTI
DOE OSTI2012

Khazar Iodine Production Plant Site Remediation in Turkmenistan. NORM Contaminated Waste Repository Establishment - 12398

Radiation safety provisions for NORM contaminated areas are in use in a number of the former Soviet republics. Some of these areas were formed by absorbed radionuclides at the iodine and bromine extraction sites. As a rule, there are not any plant radiation monitoring systems nor appropriate services to ensure personnel, population and environmental radiation safety. The most hazardous sites are those which are situated in the Caspian Sea coastal zone. The bulk of the accumulated waste is represented by a loose mixture of sand and charcoal, which was basically used as the iodine extraction sorbent. The amounts of these wastes were estimated to be approximately 20,000 metric tons. The waste contamination is mainly composed of Ra-226 (U-238 decay series) and Ra-224, Ra-228 (Th-232 decay series). In 2009, the 'ECOMET-S', a Closed Joint-Stock Company from St. Petersburg, Russian Federation, was authorized by the Turkmenistan government to launch the rehabilitation project. The project includes D and D activities, contaminated areas remediation, collected wastes safe transportation to the repository and its disposal following repository closure. The work at the Khazar chemical plant started in September, 2010. Comprehensive radiological surveys to estimate the waste quantities were carried out in advance. In course of the rehabilitation work at the site of the Khazar chemical plant additional waste quantities (5,000 MT, 10,000 m{sup 3}) were discovered after the sludge was dumped and drained. Disposal volumes for this waste was not provided initially. The additional volume of the construction wastes was required in order to accommodate all the waste to be disposed. For the larger disposal volume the project design enterprise VNIPIET, offered to erect a second wall outside the existing one and this solution was adopted. As of May, 2011, 40,575 m{sup 3} of contaminated waste were collected and disposed safely. This volume represents 96.6% of the initial repository volume. Now work is underway to erect the second repository wall, which will allow housing of the additional 16,800 m{sup 3} The Khazar chemical plant territory restoration work is underway as well. (authors)

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

Nuclide Release Pathways in and Around a Trench Repository for Radioactive Waste - 16014

An effective model for a safety assessment of a conceptual trench repository system, in which low-and intermediate level radioactive waste (LILW) is disposed of has been developed. The repository system is assumed to be located in the surface area of a mountain hill near the seashore. An easy-to-use computer program based on this model, developed as a GoldSim template, has been extended based on the previous one in consideration of the rather possible complex pathways in and around the trench for its performance assessment, which is able to evaluate a nuclide release from the repository and farther transport into the geosphere and biosphere under various normal, disruptive natural, and manmade events, and scenarios that can occur after a failure of waste drums with associated uncertainty. To demonstrate the nuclide release and transport behavior through these various possible pathways in the near- and far-fields of the repository system under certain scenarios, some illustrative evaluations and a simple sensitivity investigation with C-14 for the concrete barriers of the trench were made. (authors)

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

Actinide Containment Safety Case in a TRU/HLW Repository in Salt - 16369

The safety case for the permanent isolation of actinides in a salt nuclear repository is discussed. The need for actinide and brine chemistry research is driven by the regulatory process that requires an accounting of low-probability 'worst case' scenarios to show that the release criteria are met under these unexpected conditions. An overview of the WIPP approach to establish the actinide source term, which is defined by the oxidation state distribution, solubility and tendency to form colloidal species, is presented for the key actinides. This approach is then discussed in the context of extending the salt repository concept to defense high level waste (HLW) and possibly spent fuel (SF) to identify key needs and gaps. It is an overall conclusion of this review that the WIPP serves as a good template for the HLW/SF case and the safety case can be readily extended to this repository concept with proper repository design and supporting experimental results. (authors)

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

Overview of Salt Repository Research and Development for Spent Nuclear Fuel and High-level Nuclear Waste in the United States - 20307

This paper summarizes the current US Department of Energy Office of Nuclear Energy's (DOE-NE) work towards developing and executing a research and development program that addresses both scientific and technical issues related to long-term disposal of spent nuclear fuel (SNF) and high-level waste (HLW) in a hypothetical bedded salt based geological repository. A primary goal of the program is to create a generic Geologic Disposal Safety Assessment (GDSA) Framework that can be used to help guide decisions on siting a possible future bedded salt repository. The generic GDSA work includes analysis of the impacts of heat generation caused by decay of short-lived radionuclides. We report progress in four primary areas. First, we discuss the development and recent modifications of a research and development road-map. Second, we briefly describe an experimental approach to better understand thermal processes in salt. Third, we highlight collaborations with the international research community that leverage salt-based repository science around the world. Finally, we discuss how our findings are being used to aid in the development of a generic safety assessment for a bedded salt repository containing SNF and HLW. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2021

(Paper) TRIPWIRE: Multi-modal Distributed Sensing For Repository Verification

Underground geological repositories for storing used nuclear fuel are being planned and developed around the world. Research is being performed to understand technical aspects of sealing entombed materials and underground repositories to ensure the radiological and nuclear materials remain contained. To support safeguards verification for these facilities research at Idaho National Laboratory (INL) is underway to develop and demonstrate a multi-modal sensor system, TRIPWIRE, for containment verification in inaccessible radiological and nuclear waste repositories. The TRIPWIRE system will continuously monitor ionizing radiation and electromagnetic fields in the vicinity of emplaced nuclear materials buried in a repository, reporting on disturbances with a real-time alarm control station. The system will use long-length scintillating fiber bundles (SFBs) to perform area radiation monitoring; these will be coupled to kilometer-scale multimodal optical communication fibers – all light sensors and electronic components used with this system will be located above ground. Electromagnetic fields, and changes in local dielectric conditions caused by intrusion and soil movement, will be monitored using commercial grade, ported "leaky" coaxial cables (PCCs), with control electronics also located above ground. Tamper-indicating self-diagnostic assessments will be done using optical and electronic time domain reflectometry in the SFBs and PCCs, respectively. The result will be a long, kilometer-scale multi-modal SFB-PCC system. Simulation and modeling are being used to inform the work and a demonstration of the system's utility is planned at a nuclear storage facility in the future.

98 - NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL↗
DOE OSTI2022

Multiphysics modeling of a critical dual-purpose canister in a saturated geological repository

Dual purpose dry cask storage canisters for spent nuclear fuel are designed for storage and transportation, but are not licensed for permanent disposal in a geological repository. If dual purpose canisters were to be used to dispose of spent nuclear fuel in a geological repository, they would be expected to eventually breach and be flooded with groundwater, and it is shown that some fraction of these canisters will achieve criticality. In order to evaluate the consequences of canisters going critical in a repository, an initial capability has been developed for estimating the quasi-static power level of a critical canister using loosely coupled multiphysics simulations. The low power level in a critical canister enables coupling through precomputed physics proxies. This calculated power level is then used to compute the change in the critical canister’s isotopic inventory as a function of time. Three as-loaded canisters are evaluated and two were found to have power levels below 4 kW, with a modest effect on the radiological inventory over time. This effort also shows that although some DPCs will have extremely peaked power shapes, the relatively low power and long-time scales result in relatively homogenized thermohydraulic properties in the water within the DPC.

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 OSTI2019

Fault Stability Perturbation by Thermal Pressurization and Stress Transfer Around a Deep Geological Repository in a Clay Formation

The increase of temperature of a low-permeability, fluid-saturated media may trigger significant thermal pressurization, where the expansion of pore fluid cannot be accommodated by the thermal expansion of the pore space. With the aid of a coupled thermohydromechanical numerical simulator, we investigate the possible impact of thermal pressurization during the life of a deep geological repository (DGR) for high-level radioactive waste, characterized by the emplacement of radioactive material-filled canisters in a series of parallel tunnels, excavated in a low-permeability clay formation. We represent the fault as a planar structure embedded in a thermoporoelastic material and shear activation evaluated by a strain-softening Mohr-Coulomb failure criterion. The results show that stress changes caused by temperature and thermal pressurization of a rock mass around the emplacement tunnels may trigger a slip event on a fault plane in proximity of the geological disposal site: Rupture nucleates at depth, hundreds of meters below the DGR. Stress transfer plays a key role, while a direct hydraulic connection between the repository and the nucleation zone is not necessary in order to trigger rupture. A low stress ratio may favor the occurrence of slip up to a distance of 600 m of the fault from the outermost tunnel. Furthermore, these results highlight the need of investigating hydromechanical properties and local stress conditions at depth to characterize the geomechanical response of weak planes located in the surroundings of a high-level radioactive waste repository and to provide sufficient knowledge for the safe development of the DGR site.

58 GEOSCIENCES↗
DOE OSTI2024

Long-length Scintillating Fibers for Nuclear Geological Repositories

This research focuses on the usage of plastic scintillating fibers as ruggedized detectors in long-length sections coupled to long-length standard optical fiber segments for the purpose of radiation detection and monitoring. This technology addresses a need for monitoring radiation in difficult-to-access or inaccessible areas, such as monitoring: radioactive waste within geological repositories, nuclear material storage facilities, spent fuel dry storage areas, inaccessible areas within nuclear facilities including reactors and reprocessing facilities, decommissioning of nuclear facilities, and other similar areas. The fiber can be placed along the tunnels of a geological repository to monitor the radiation field along its length. This can detect an increase of radiation due to movement, leakage, or unexpected opening of the containers, as well as decrease in the radiation due to removal of material. These capabilities have been demonstrated at INL’s Radioactive Scrap Waste Facility by successfully detecting the removal of an interim storage containers from the storage area. The poster focuses on explaining the technology and its capability in monitoring deep geological repositories (DGRs), especially in the domain of safety and international safeguards. As more countries consider the development of DGRs, technologies such as the one being presented can help establish continuity of knowledge of the material, maintain public confidence, and ensure safe operation of the facility. Furthermore, the poster looks at the future steps of the technology and its potential role in the international safeguards regime.

98 - NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL↗
DOE OSTI2019

On the long-term migration of uranyl in bentonite barrier for high-level radioactive waste repositories: The effect of different host rocks

Repositories for high level radioactive waste hosted in argillite or granite with bentonite backfill are the two disposal concepts that been extensively studied. A general perception is that granite has very low sorption capacity and therefore is disadvantageous in retarding the migration of radionuclides in comparison with argillite. However, when there is bentonite backfill, does the sorption capacity of granite really matter for the migration of radionuclides? How different host rocks would affect the migration of U(VI) through the bentonite backfill of an engineered barrier system (EBS), and which properties of the host rocks cause those differences in migration, given that migration of radionuclides is a critical measure in assessing the performance of a repository? Here we present two coupled thermal, hydrological, and chemical models for the transport of U(VI), each with an identical setup except for the type of host rock (argillite vs granite rock). Comparisons between our models show that the different host rocks exert their influence on migration of U(VI) via regulating the chemical conditions in the bentonite, which affect the concentration of U(VI) at the source and the adsorption of U(VI) in the bentonite. The key chemical conditions are pH, Ca + , and HCO 3 - concentration in pore-water, which are strongly affected by soluble carbonate minerals. Our models also show the occurrence of illitization (dissolution of smectite and precipitation of illite) in the bentonite, which affects the migration of U(VI) through changing pH and the quantity of adsorbents. Although generalization of current model results for granite and argillite should be done carefully, the simulations highlight the importance of pore-water chemistry in host rock, as well as bentonite-host rock interactions, when assessing the performance of a repository. In comparison to properties such as thermal conductivity, permeability, and adsorption capacity, properties that have been studied intensively, the pore-water chemistry of host rocks deserve equal or even more attention because of its strong influence on radionuclide migration.

58 GEOSCIENCES↗
DOE OSTI2020

Modelling uranium dioxide corrosion under repository conditions: A pore-scale study of the chemical and thermal processes

This paper investigates the corrosion behaviors of spent uranium dioxide (UO 2 ) fuel when placed in geological repositories. We performed pore-scale reactive transport simulations of the UO 2 corrosion process in a defective fuel rod with different orientations of breach or fracture on its clad. It is found that the corrosion rate has strong dependency on the breach orientation. The highest corrosion rate of UO 2 is calculated when the angle between the groundwater flow direction and the clad fracture reaches 180°. UO 2 corrosion simulations with different flow rates are compared. The results show the higher flow rate accelerates the corrosion of UO 2 fuel. The effect of pH on the corrosion process is also determined. It is indicated that higher pH can limit the corrosion of UO 2 by reducing the reaction rate. The dependence of reaction rate and time on reactive surface area is explored. Spent fuel with lower surface area demonstrates longer lifetime under corrosion conditions. A numerical model coupling thermal conduction and chemical reactions is developed to assess the impact of temperature on the process of UO 2 corrosion. The results show that higher temperature leads to larger corrosion rates for UO 2 . Here, the predicted reaction rates are higher in comparison with the isothermal results. Hence, our results can help improve the fundamental understanding of UO 2 corrosion in geological repositories for long-term storage of spent nuclear fuels and provide guidance for the safe operations and selection of appropriate repositories.

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

Enthalpies of formation of polyhalite: A mineral relevant to salt repository

Polyhalite is an important coexisting mineral with halite in salt repositories for nuclear waste disposal, such as Waste Isolation Pilot Plant (WIPP) in Carlsbad, New Mexico. The thermal stability of this mineral is a key knowledge in evaluating the integrity of a salt repository in the long term, as water may release due to thermal decomposition of polyhalite. Previous studies on structural evolution of polyhalite at elevated temperatures laid the basis for detailed calorimetric measurements. Using high-temperature oxide-melt drop-solution calorimetry at 975 K with sodium molybdate as the solvent, we have determined the standard enthalpies of formation from constituent sulfates (ΔH° f,sul ), oxides (ΔH° f,ox ) and elements (ΔH° f,ele ) of a polyhalite sample with the composition of K 2 Ca 2 Mg(SO 4 ) 4 ·1.95H 2 O from the Salado formation at the WIPP site. The obtained results are: ΔH° f,sul = -152.5 ± 5.3 kJ/mol, ΔH° f,ox = -1926.1 ± 10.5 kJ/mol, and ΔH° f,ele = -6301.2 ± 9.9 kJ/mol. Furthermore, based on the estimated formation entropies of polyhalite, its standard Gibbs free energy of formation has been derived to be in the range of -5715.3 ± 9.9 kJ/mol to -5739.3 ± 9.9 kJ/mol. In conclusion, these determined thermodynamic properties provide fundamental parameters for modeling the stability behavior of polyhalite in salt repositories.

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

Implementation of FAIR principles in the IPCC: the WGI AR6 Atlas repository

The Sixth Assessment Report (AR6) of the Intergovernmental Panel on Climate Change (IPCC) has adopted the FAIR Guiding Principles. We present the Atlas chapter of Working Group I (WGI) as a test case. We describe the application of the FAIR principles in the Atlas, the challenges faced during its implementation, and those that remain for the future. We introduce the open source repository resulting from this process, including coding (e.g., annotated Jupyter notebooks), data provenance, and some aggregated datasets used in some figures in the Atlas chapter and its interactive companion (the Interactive Atlas), open to scrutiny by the scientific community and the general public. We describe the informal pilot review conducted on this repository to gather recommendations that led to significant improvements. Finally, a working example illustrates the re-use of the repository resources to produce customized regional information, extending the Interactive Atlas products and running the code interactively in a web browser using Jupyter notebooks.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2016

Damage-plasticity model of the host rock in a nuclear waste repository

The paper describes damage-plasticity model for the modelling of the host rock environment of a nuclear waste repository. Radioactive Waste Repository Authority in Czech Republic assumes the repository to be in a granite rock mass which exhibit anisotropic behaviour where the strength in tension is lower than in compression. In order to describe this phenomenon, the damage-plasticity model is formulated with the help of the Drucker-Prager yield criterion which can be set to capture the compression behaviour while the tensile stress states is described with the help of scalar isotropic damage model. The concept of damage-plasticity model was implemented in the SIFEL finite element code and consequently, the code was used for the simulation of the Äspö Pillar Stability Experiment (APSE) which was performed in order to determine yielding strength under various conditions in similar granite rocks as in Czech Republic. The results from the performed analysis are presented and discussed in the paper.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2013

Dose rate profile surrounding a repository

The focus of this research was to analyze the dose rate profile around a waste repository using Monte Carlo techniques. Dose rates at various heights and distances were analyzed outside of the waste repository using MCNPX [1]. The heights measured extended the height of the building and the distances varied between 0 and 22 m away from the waste repository. The simulation data were fitted by smooth analytical functions for different height levels and distances, such that vertical and horizontal dose rates as functions of source-detector distance were achieved. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2012

Report on Modeling Coupled Processes in the Near Field of a Clay Repository

Clay/shale has been considered as potential host rock for geological disposal of high-level radioactive waste throughout the world. Coupled thermal, hydrological, mechanical, and chemical (THMC) processes have a significant impact on the long-term safety of a clay repository. For example, the excavation damaged zone (EDZ) near repository tunnels can modify local permeability (resulting from induced fractures), potentially leading to less confinement capability. This report documents results from three R&D activities: (1) implementation and validation of constitutive relationships, (2) development of a discrete fracture network (DFN) model for investigating coupled processes in the EDZ, and (3) development of a THM model for the FE tests at Mont Terri, Switzerland, for the purpose of model validation. The overall objective of these activities is to provide an improved understanding of EDZ evolution in clay repositories and the associated coupled processes, and to develop advanced relevant modeling capabilities.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2020

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..

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2021

Draft Experimental Plan for Commercial SNF Degradation in Repository Environments with a Focus on Fracture Matrix Degradation

This report documents work performed supporting US Department of Energy (DOE) Nuclear Energy Spent Fuel and Waste Disposition, Spent Fuel and Waste Science and Technology, under work breakdown structure element 1.08.01.03.04, “Inventory and Waste Form Characteristics and Performance.” In particular, this report fulfills the M3 milestone, M3SF-21OR010309072 "Draft Experimental Plan for Commercial SNF Degradation in Repository Environments” as described in work package SF-21OR01030907 "SNF Degradation Testing-ORNL".This experimental plan is a collaboration between ORNL, SNL, ANL, and PNNL. It is intended to support source term model development and spent fuel degradation testing in a range of generic repository conditions. Potential testing and experiments to support validation of the Fuel Matrix Dissolution Model (FMDM) and the overall development/validation of repository source term modeling are discussed.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗