MODELING WASTE PACKAGE DEGRADATION AND WASTE FORM DISSOLUTION FOR GEOLOGIC REPOSITORY PERFORMANCE ASSESSMENT IN PFLOTRAN.
Abstract not provided.
Frederick, Jennifer M↗
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Results for “Waste Repository”
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Impact of High Performance Computing on the Detailed Simulation of Subsurface Flow and Contaminant Transport Using eSTOMP - 16341
The ability to model subsurface processes at the field-scale is essential to site closure and remedial action decisions. Field-scale models often require high-performance computing to meet the computational demands associated with complex systems, model calibration, and uncertainty quantification. eSTOMP, a parallel version of STOMP, is an analytical tool that meets the need for investigating coupled processes in the subsurface involving regional multi-fluid flow and biogeochemistry. The scalability of eSTOMP makes it ideal for simulating contaminant and geochemical reactive transport when long run times result from execution of a serial code. To demonstrate the impact of high-performance computation on subsurface analyses, the application of eSTOMP is presented for two sites. The first site involves uranium surface complexation within a highly dynamic, variably saturated flow field. The use of this model in conjunction with continued field monitoring is expected to provide a rigorous basis for developing operational strategies for field remediation and for defining defensible remediation endpoints. The second site involves the long-term dissolution of low-level radioactive glass emplaced within a near-surface repository. The waste form release rate is evaluated by modeling the basic physical and chemical processes that are known to control the waste form dissolution behavior, including pH and composition of the fluid contacting the glass. Simulations demonstrate that glass dissolution rates vary both in time and as a function of position within the disposal system, and that a 'leach rate' or radionuclide release rate parameter can only be estimated after identifying these impacts in a dynamic system. (authors)
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Non-destructive Material Characterization of Radioactive Waste Packages with QUANTOM{sup R} - 20130
During the last decades, the nuclear and non-nuclear industry has produced a considerable amount of low and intermediate level radioactive wastes. In many countries, it is foreseen to dispose of these waste packages safely in a final repository under strict waste acceptance requirements (e.g. the radiological and material characterization) defined by national licensing and supervisory authorities. The material characterization of waste packages can be performed on the basis of existing documentation or, if the documentation is insufficient, on further destructive or non-destructive analysis. Non-destructive methods are to be preferred to minimize radiation exposures of operating personnel as well as costs. We offer here an innovative non-destructive technology called QUANTOM{sup R} (Quantitative Analysis of Toxic and nontoxic Materials) based on prompt and delayed gamma neutron activation analysis (P and DGNAA). This technology is able to identify, quantify and thus verify the amount of hazardous and non-hazardous substances in 200-l radioactive drums, which is required for a final disposal characterization. (authors)
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DECOVALEX-2023: Task F2-Salt Final Report
The subject of Task F of DECOVALEX-2023 concerns performance assessment modelling of radioactive waste disposal in deep mined repositories. The primary objectives of Task F are to build confidence in the models, methods, and software used for performance assessment (PA) of deep geologic nuclear waste repositories, and/or to bring to the fore additional research and development needed to improve PA methodologies. In Task F2-(salt), these objectives have been accomplished through staged development and comparison of the models and methods used by participating teams in their PA frameworks. Coupled-process submodels and deterministic simulations of the entire PA model for a reference scenario for waste disposal in domal salt have been conducted. The task specification has been updated continuously since the initiation of the project to reflect the staged development of the conceptual repository model and performance metrics.
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Conditioning and Requalification of Legacy Waste Packages - 18648
In most German interim storage facilities, hundreds of waste packages are waiting for re-conditioning and qualification to comply with the waste acceptance criteria (WAC) for the Konrad repository. Especially the waste classification as well as some characteristics and the description of the material composition in its current state do not comply with the Konrad requirements since the conditioning of these waste packages took place up to 40 years ago. The German name for these waste packages is 'Altgebinde', hereafter referred to as 'legacy waste packages'. The present article will provide an overview of the measures for the qualification of these legacy waste packages to fulfill the Konrad requirements. The following measures, when necessary, are carried out for the qualification of legacy waste packages: - Collection and evaluation of all documents related to the primary conditioning process, e. g. the used equipment, documented waste and packaging properties, conditioning and radiology, if available (hereafter 'old documents'). - Preparation of a conceptual report and a process control quality plan based on existing data, with appropriate measures to qualify the waste packages. Assessment and approval of the process control quality plan by the competent authority. - Characterization of the waste and packaging properties by visual inspection or tomography, recording or updating of dose rate, contamination and mass, verification of product stability, for example by gas analysis of the interior atmosphere of the container. - Characterization of the nuclide inventory by gammas scans, sample taking to analyze nuclides which can not be measured by means of in-situ gamma spectrometry, and for material declaration. - Qualification or post-conditioning of the legacy waste packages if necessary, e. g. by drying the waste packages and packaging them in containers approved for disposal. - Compilation of all results of the above mentioned tests and steps to create a waste package documentation, including the calculations for the nuclide inventory and the description of the material composition of the waste packages. - Receiving approval by the competent authority after evaluation of the final documentation. The aim behind this structured approach is to prevent additional, unnecessary measures and treatment of the waste. Consequently, the additional characterization is preferably carried out on the basis of the available documents. (authors)
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Documentation of Radioactive Waste Packages for Disposal in the Konrad Repository - 16253
For the Konrad repository, the plan approval procedure has been completed in 2002. In 2007 all complaints filed against the repository were dismissed, so that there is a legal and un-appealable license decision available. The waste acceptance requirements were revised taking into account additional requirements from the plan approval decision. Today, there are already several hundred waste packages produced that meet all requirements for final disposal in Konrad. The acquisition, verification and quality assurance of the data relevant for disposal is performed in parallel to the waste treatment and packaging. This constitutes the development of database applications, on the one hand to capture and handle specific data for the waste packages to another to manage the disposal procedures. The development of a database application for recording disposal-relevant data of waste packages has already been completed with regard to the radiological requirements. Additional aspects based on the requirements of the plan approval decision, e.g. water law relevant aspects, have to be taken into account. Thus, an upgrade relating to the recording of groundwater debilitating substances will be carried out in a second step, which is expected to be completed in 2016. In parallel, the development of a database application is carried out to coordinate the waste package deliveries to the repository and to document their disposal. (authors)
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International Collaboration Activities on Engineered Barrier Systems
The Used Fuel Disposition Campaign (UFDC) within the DOE Fuel Cycle Technologies (FCT) program has been engaging in international collaborations between repository R&D programs for high-level waste (HLW) disposal to leverage on gathered knowledge and laboratory/field data of near- and far-field processes from experiments at underground research laboratories (URL). Heater test experiments at URLs provide a unique opportunity to mimetically study the thermal effects of heat-generating nuclear waste in subsurface repository environments. Various configurations of these experiments have been carried out at various URLs according to the disposal design concepts of the hosting country repository program. The FEBEX (Full-scale Engineered Barrier Experiment in Crystalline Host Rock) project is a large-scale heater test experiment originated by the Spanish radioactive waste management agency (Empresa Nacional de Residuos Radiactivos S.A. – ENRESA) at the Grimsel Test Site (GTS) URL in Switzerland. The project was subsequently managed by CIEMAT. FEBEX-DP is a concerted effort of various international partners working on the evaluation of sensor data and characterization of samples obtained during the course of this field test and subsequent dismantling. The main purpose of these field-scale experiments is to evaluate feasibility for creation of an engineered barrier system (EBS) with a horizontal configuration according to the Spanish concept of deep geological disposal of high-level radioactive waste in crystalline rock. Another key aspect of this project is to improve the knowledge of coupled processes such as thermal-hydro-mechanical (THM) and thermal-hydro-chemical (THC) operating in the near-field environment. The focus of these is on model development and validation of predictions through model implementation in computational tools to simulate coupled THM and THC processes.
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Practical Considerations for the Implementation of a World Information Library (WIL) - 17112
The U.S. Department of Energy's (DOE) Waste Isolation Pilot Plant (WIPP) is a deep geologic repository for the permanent disposal of defense-generated transuranic waste. Title 40 Code of Federal Regulations (CFR) Part 194.43, Passive institutional controls, requires that measures be employed, '...to preserve knowledge about the location, design, and contents of the [WIPP] disposal system'. These measures include the, 'Placement of records in the archives and land record systems of local, State, and Federal governments, as well as international archives, that would likely be consulted by individuals in search of unexploited resources'. One possibility for allowing individuals to quickly and easily access information on the location, design, and contents of a disposal system and its relation to unexploited resources is the development of an electronic geographic information system (GIS) that would be used to interpret spatial data. In 2016, R. Patterson proposed such a system, known as the 'World Information Library (WIL)'. Information in this database system would be designed to be publicly accessible through a simple web-based user interface that would focus on spatial databases with categories such as; waste repositories, resources, historical sites, wildlife, infrastructure, medicine, agriculture, and technology. The conceptual WIL would be a compilation of searchable libraries and databases. A category in the WIL would consist of spatial databases created for the purpose of preserving knowledge and providing a means of communicating location information in a GIS-based Land Use Information System (LUIS) user interface. The successful implementation of the WIL would require the robust use of project controls, an effective Quality Assurance program, and the development of international standards in collaboration with industry experts. The purpose of this paper is to identify and discuss known practices and industry standards that could be utilized for the implementation of a WIL as was introduced by Patterson et al. in 'A Conceptual World Information Library (WIL) and Land Use Information System (LUIS)'. (authors)
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Summary of Information Available and Recommendations for a Doe Complex-Wide Waste Glass Database
The purpose of this document is to summarize the existing repositories of nuclear waste glass property-composition data produced for DOE. A framework is described for consolidation of the data into a single database. Recommended attributes of the consolidated database are provided, including quality assurance, the user interface, and means for updates and maintenance. Future work will utilize the framework described herein to produce the complex-wide database and establish the user interface.
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Qualification of Numerical and Semi-Analytical Modeling Codes for Thermal Design of HLW Repository - 18455
Disposal in a geological repository is a viable option for safely isolating high-level radioactive waste. Thermal processes induced by decay heat from the radioactive waste may impact the integrity of the engineered barrier system, and may lead to fluid flow and transport; and the associated migration of radionuclides. This study looks at an analysis of thermal conduction for the emplacement of high-level waste in a repository located in bedded salt. The simulations used analytical and numerical methods to study conductive heat transport in a repository. This study concentrated on heat flow in a central waste package with contributions from adjacent waste packages and drifts. Of particular importance to this work is comparison of predictions of peak temperature at the drift wall and at the waste package surface using different solution methods. Simulation results of analytical and numerical modeling methods are compared. The results show that predictions of the codes were comparable under identical initial and boundary conditions. (authors)
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Coupled Processes Modeling in Rock Salt and Crushed Salt Including Halite Solubility Constraints: Application to Disposal of Heat-Generating Nuclear Waste
This work presents numerical modeling of coupled thermal, hydraulic and mechanical processes in rock salt and crushed salt considering halite solubility constraints. The TOUGH-FLAC simulator is used, with a recently enhanced Equation-Of-State module that includes the thermodynamic properties of aqueous fluids of variable salinity. Laboratory and field scale tests performed on rock salt and crushed salt under temperature gradients are modeled first to evaluate the capabilities of the simulator to reproduce important features, such as porosity changes induced by halite dissolution/precipitation, and brine and heat migration. Since the results are quite satisfactory, the simulator is used to predict the long-term response of a generic salt repository for heat-generating nuclear waste. To evaluate the impacts of halite solubility on the predictions, two simulations that respectively consider or neglect solubility constraints are performed. In the scenario studied, the results are not significantly affected by dissolution/precipitation, and only some differences are observed due to changes in porosity, but the dominating processes remain the same. Finally, with the new provisions, TOUGH-FLAC is more complete in terms of processes occurring around a heat-releasing nuclear waste package and can therefore provide more accurate predictions of the long-term performance of a nuclear waste repository in salt formations.
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KOMPASS-II: Compaction of Crushed salt for Safe Containment – Phase 2
Long-term stable sealing elements are a basic component in the safety concept for a possible repository for heat-emitting radioactive waste in rock salt. The sealing elements will be part of the closure concept for drifts and shafts. They will be made from a welldefinied crushed salt in employ a specific manufacturing process. The use of crushed salt as geotechnical barrier as required by the German Site Selection Act from 2017 /STA 17/ represents a paradigm change in the safety function of crushed salt, since this material was formerly only considered as stabilizing backfill for the host rock. The demonstration of the long-term stability and impermeability of crushed salt is crucial for its use as a geotechnical barrier. The KOMPASS-II project, is a follow-up of the KOMPASS-I project and continues the work with focus on improving the understanding of the thermal-hydraulic-mechanical (THM) coupled processes in crushed salt compaction with the objective to enhance the scientific competence for using crushed salt for the long-term isolation of high-level nuclear waste within rock salt repositories. The project strives for an adequate characterization of the compaction process and the essential influencing parameters, as well as a robust and reliable long-term prognosis using validated constitutive models. For this purpose, experimental studies on long-term compaction tests are combined with microstructural investigations and numerical modeling. The long-term compaction tests in this project focused on the effect of mean stress, deviatoric stress and temperature on the compaction behavior of crushed salt. A laboratory benchmark was performed identifying a variability in compaction behavior. Microstructural investigations were executed with the objective to characterize the influence of pre-compaction procedure, humidity content and grain size/grain size distribution on the overall compaction process of crushed salt with respect to the deformation mechanisms. The created database was used for benchmark calculations aiming for improvement and optimization of a large number of constitutive models available for crushed salt. The models were calibrated, and the improvement process was made visible applying the virtual demonstrator.
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Thermal Analysis of Disposal of High-Level Nuclear Waste in a Generic Bedded Salt repository using the Semi-Analytical Method.
An example case is presented for testing analytical thermal models. The example case represents thermal analysis of a generic repository in bedded salt at 500 m depth. The analysis is part of the study reported in Matteo et al. (2016). Ambient average ground surface temperature of 15°C, and a natural geothermal gradient of 25°C/km, were assumed to calculate temperature at the near field. For generic salt repository concept crushed salt backfill is assumed. For the semi-analytical analysis crushed salt thermal conductivity of 0.57 W/m-K was used. With time the crushed salt is expected to consolidate into intact salt. In this study a backfill thermal conductivity of 3.2 W/m-K (same as intact) is used for sensitivity analysis. Decay heat data for SRS glass is given in Table 1. The rest of the parameter values are shown below. Results of peak temperatures at the waste package surface are given in Table 2.
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Postclosure assessment modelling of a deep geologic repository for low and intermediate level radioactive waste in Ontario, Canada - 15145
In 2011, Ontario Power Generation (OPG) submitted an Environmental Impact Statement and a Preliminary Safety Report (PSR) to the federal authorities in Canada to support an application for a site preparation and construction licence for a Deep Geologic Repository (DGR) for low and intermediate level radioactive waste (L and ILW). The DGR, which is proposed to be located close to the existing Western Waste Management Facility (WWMF) at the Bruce nuclear site, would receive L and ILW arising from the operation and refurbishment of OPG's nuclear reactors; the wastes are currently stored safely at the WWMF. The PSR is supported by postclosure safety assessment modelling, including a total-systems level model implemented in the AMBER compartment modelling software that draws directly on detailed groundwater and coupled groundwater and gas modelling using the FRA3DVS-OPG and TOUGH2 codes. A full suite of calculation cases is managed within a single AMBER model, which aided efficient management of the calculation files and quality assurance. Assessment of expected/normal evolution cases results in maximum calculated effective doses to humans that are more than five orders of magnitude below associated criteria. Assessment of conservative disruptive and 'what if' scenarios results in maximum calculated doses that are higher than those for the normal evolution cases, but which, when the low likelihood of occurrence is taken into account, remain significantly below the associated risk criterion. The assessment demonstrates that deep geologic repositories can provide a safe disposal option and provides an example of evaluating the safety of projects on long time scales. It shows how results from detailed groundwater and gas codes can be incorporated into assessment-level codes such as AMBER, thereby providing a single systems model that can be used to efficiently evaluate a wide range of scenarios and associated calculation cases. (authors)
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Measurement and modelling of reactive transport in geological barriers for nuclear waste containment
Compacted clays are considered as excellent candidates for barriers to radionuclide transport in future repositories for nuclear waste due to their very low hydraulic permeability. Diffusion is the dominant transport mechanism, controlled by a nano-scale pore system. Assessment of the clays' long-term containment function requires adequate modelling of such pore systems and their evolution. Existing characterisation techniques do not provide complete pore space information for effective modelling, such as pore and throat size distributions and connectivity. Special network models for reactive transport are proposed here using the complimentary character of the pore space and the solid phase. Here, this balances the insufficient characterisation information and provides the means for future mechanical–physical–chemical coupling. The anisotropy and heterogeneity of clays is represented using different length parameters and percentage of pores in different directions. Resulting networks are described as mathematical graphs with efficient discrete calculus formulation of transport. Opalinus Clay (OPA) is chosen as an example. Experimental data for the tritiated water (HTO) and U(VI) diffusion through OPA are presented. Calculated diffusion coefficients of HTO and uranium species are within the ranges of the experimentally determined data in different clay directions. This verifies the proposed pore network model and validates that uranium complexes are diffusing as neutral species in OPA. In the case of U(VI) diffusion the method is extended to account for sorption and convection. Finally, rather than changing pore radii by coarse grained mathematical formula, physical sorption is simulated in each pore, which is more accurate and realistic.
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International Approaches for Nuclear Waste Disposal in Geological Formations: Report on Fifth Worldwide Review
An important issue for present and future generations is the final disposal of spent nuclear fuel. Over the past over forty years, the development of technologies to isolate both spent nuclear fuel (SNF) and other high-level nuclear waste (HLW) generated at nuclear power plants and from production of defense materials, and low- and intermediate-level nuclear waste (LILW) in underground rock and sediments has been found to be a challenging undertaking. Finding an appropriate solution for the disposal of nuclear waste is an important issue for protection of the environment and public health, and it is a prerequisite for the future of nuclear power. The purpose of a deep geological repository for nuclear waste is to provide to future generations, protection against any harmful release of radioactive material, even after the memory of the repository may have been lost, and regardless of the technical knowledge of future generations. The results of a wide variety of investigations on the development of technology for radioactive waste isolation from 19 countries were published in the First Worldwide Review in 1991 (Witherspoon, 1991). The results of investigations from 26 countries were published in the Second Worldwide Review in 1996 (Witherspoon, 1996). The results from 32 countries were summarized in the Third Worldwide Review in 2001 (Witherspoon and Bodvarsson, 2001). The last compilation had results from 24 countries assembled in the Fourth Worldwide Review (WWR) on radioactive waste isolation (Witherspoon and Bodvarsson, 2006). Since publication of the last report in 2006, radioactive waste disposal approaches have continued to evolve, and there have been major developments in a number of national geological disposal programs. Significant experience has been obtained both in preparing and reviewing cases for the operational and long-term safety of proposed and operating repositories. Disposal of radioactive waste is a complex issue, not only because of the nature of the waste, but also because of the detailed regulatory structure for dealing with radioactive waste, the variety of stakeholders involved, and (in some cases) the number of regulatory entities involved.
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Long-term modeling of the thermal–hydraulic–mechanical response of a generic salt repository for heat-generating nuclear waste
Not Available
Blanco Martín, Laura↗