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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.

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At least 343 records · Page 19
All sourcesNASA NTRSDOE OSTI
DOE OSTI2017

Legacy Waste Containers for Low and Intermediate Level Radioactive Waste in Germany - Actual Situation and Challenges - 17070

Low and intermediate level radioactive waste has been produced in Germany for several decades. While a part of the waste had been deposited in the Morleben repository and the Asse mine which have been operated for just a limited period of time, the major amount is conditioned and packaged in temporarily stored waste containers. Those containers are designated for final disposal in the KONRAD repository being currently under construction and expected to start operation in 2022. As a consequence, an assessment of the waste container documentation and safety is strongly needed to receive an approval for the Konrad repository. This requalification is challenging because those old containers have to meet the same safety requirements as new containers. (authors)

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DOE OSTI2018

The SD-20K, a large Type B(U)F Package for Waste Containers - 18181

The type B(U)F package SD-20K, currently being developed by DAHER Nuclear Technologies GmbH (DNT), will facilitate the transport of legacy waste from interim storage facilities to the German final repository Konrad. The package SD-20K could also be used in other cases where legacy packages are not fit for a final repository and/or for transport but could be easily conditioned into a suitable package. Since the 1990's German utilities, reactor operators and service providers conditioned waste according to the requirements of the preliminary acceptance criteria for the final repository Konrad. Some of this waste was conditioned in containers which were at that time licensed as type B packages but for which the license expired due to various reasons. Other waste was reclassified so that the used package type does no longer comply with the content. In recent times Konrad containers of type IP-2 are loaded with e. g. core components in compliance with the Konrad acceptance criteria but overloaded with respect to the transport regulations. DAHER Nuclear Technologies GmbH (DNT) in cooperation with Siemens AG investigated possible solutions in a general approach. The result of the evaluation was that the best solution for the transport problem was to develop a versatile type B(U)F package large enough to accommodate all containers and packages accepted for the final repository Konrad. Possible solutions like repacking (high costs and high dose exposure) or requalification (high costs and risk of failure to achieve the goal) were not considered to be generally applicable. The SD-20K packaging is currently designed to accommodate Konrad Containers type III, IV and VI; however, the dimensions of its cavity are sufficient to accommodate any container and package that meets the acceptance criteria for the final repository Konrad. The SD-20K packaging consists of an inner container which constitutes the containment, surrounded by a rather large shock absorbing structure which also contains the thermal protection system. The outer shape of the SD-20K resembles a 20' ISO container. A special feature of the SD-20K is the automatic unloading system to minimize dose exposure of the personnel in the final repository. After arrival in the repository the shock absorbing structure as well as the door of the containment system are opened automatically and the content is moved out of the SD-20K by a remote-controlled carrier system. (authors)

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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..

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DOE OSTI2019

GDSA Repository Systems Analysis FY19 Update

The Spent Fuel and Waste Science and Technology (SFWST) Campaign of the U.S. Department of Energy Office of Nuclear Energy, Office of Spent Fuel and Waste Disposition (SFWD), has been conducting research and development on generic deep geologic disposal systems (i.e., geologic repositories). This report describes specific activities in the second half of Fiscal Year (FY) 2019 associated with the Geologic Disposal Safety Assessment (GDSA) Repository Systems Analysis (RSA) work package within the SFWST Campaign. 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 host-rock environments, and to simulate and analyze these generic repository concepts and models using the GDSA Framework toolkit, and other tools as needed.

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DOE OSTI2019

GDSA Repository Systems Analysis Progress Report

The Spent Fuel and Waste Science and Technology (SFWST) Campaign of the U.S. Department of Energy Office of Nuclear Energy, Office of Spent Fuel and Waste Disposition (SFWD), has been conducting research and development on generic deep geologic disposal systems (i.e., geologic repositories). This report describes specific activities in fiscal year (FY) 2019 associated with FY19 Geologic Disposal Safety Assessment (GDSA) Repository Systems Analysis (RSA) work package within the SFWST Campaign. 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 host-rock environments, and to simulate and analyze these generic repository concepts and models using the GDSA Framework toolkit, and other tools as needed.

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DOE OSTI2020

GDSA Repository Systems Analysis Investigations (FY2020)

The Spent Fuel and Waste Science and Technology (SFWST) Campaign of the U.S. Department of Energy Office of Nuclear Energy, Office of Spent Fuel and Waste Disposition (SFWD), has been conducting research and development on generic deep geologic disposal systems (i.e., geologic repositories). This report describes specific activities in the Fiscal Year (FY) 2020 associated with the Geologic Disposal Safety Assessment (GDSA) Repository Systems Analysis (RSA) work package within the SFWST Campaign. 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 host-rock environments, and to simulate and analyze these generic repository concepts and models using the GDSA Framework toolkit, and other tools as needed.

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DOE OSTI2020

KOMPASS: Compaction of crushed salt for the safe containment

In Germany, rock salt formations are a possible host rock taken into account for the safe disposal of heat-emitting radioactive waste. With respect to crushed salt will be used in the repository for backfilling of open cavitied (using dry material). With time, the crushed salt will be compacted by the convergence of the host rock and reaches porosities comparable with the rock salts. The compaction behaviour of crushed salt has been investigated within the last 40 years, however, its behaviour at low porosities and the resulting low permeabilities becomes relevant with the introduction of the approach of the containment providing rock zone. In the current state, the database and process understanding have some important gaps in knowledge referring the material behaviour, existing laboratory and numerical models, especially for the porosity range. The objective of this project was the development of methods and strategies for the reduction of deficits in the prediction of crushed salt compaction leading to an improvement of the prognosis quality. It includes the development of experimental methods for determining crushed salt properties in the range of low porosities, the enhancement of process understanding and the investigation and development of existing numerical models.

Crushed salt compaction↗
DOE OSTI2019

Approaches to multi-scale analyses of mechanically and thermally-driven migration of fluid inclusions in salt rocks

Fluid inclusions are found within mineral crystals or along grain boundaries in many sedimentary rocks, notably in evaporite formations, and can migrate along a thermal or hydro-mechanical gradient. Shale and salt rocks have been considered potential host rocks for radioactive waste disposal, due to their low permeability. Previously stagnant inclusions may become mobilised by a perturbation of the in situ state by a geotechnical installation or the emplacement of heat-generating waste. Here, the migration of fluid inclusions can thus have important impacts on the long-term performance of a geologic repository for high-level radioactive waste disposal. As a part of the international research project DECOVALEX-2019, two aspects of fluid inclusion migration in rock salt are currently investigated under different boundary conditions: a) altered hydro-mechanical conditions as a consequence of tunnel excavation or borehole drilling and b) coupled thermo-hydro-mechanical-chemical conditions during the heating period of the post-closure phase of a repository.

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DOE OSTI2018

Micromechanical processes in consolidated granular salt

Here, granular salt is likely to be used as backfill material and a seal system component within geologic salt formations serving as a repository for long-term isolation of nuclear waste. Pressure from closure of the surrounding salt formation will promote consolidation of granular salt, eventually resulting in properties comparable to native salt. Understanding dependence of consolidation processes on stress state, moisture availability, temperature, and time is important for demonstrating sealing functions and long-term repository performance. This study characterizes laboratory-consolidated granular salt by means of microstructural observations. Granular salt material from mining operations was obtained from the bedded Salado Formation hosting the Waste Isolation Pilot Plant and the Avery Island salt dome. Laboratory test conditions included hydrostatic consolidation of jacketed granular salt with varying conditions of confining isochoric stress to 38 MPa, temperature to 250 °C, moisture additions of 1% by weight, time duration, and vented and non-vented states. Resultant porosities ranged between 1% and 22%. Optical and scanning electron microscopic techniques were used to ascertain consolidation mechanisms. From these investigations, samples with 1% added moisture or unvented during consolidation, exhibit clear pressure solution processes with tightly cohered grain boundaries and occluded fluid pores. Samples with only natural moisture content consolidated by a combination of brittle, cataclastic, and crystal plastic deformation. Recrystallization at 250 °C irrespective of moisture conditions was also observed. The range and variability of conditions applied in this study, combined with the techniques used to display microstructural features, are unique, and provide insight into an important area of governing deformation mechanism(s) occurring within salt repository applications.

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DOE OSTI2020

Generic FEPs Catalogue and Salt Knowledge Archive

This report describes the development of a comprehensive catalogue of generic features, events, and processes (FEPs) that are potentially important for the post-closure performance of a repository for high-level radioactive waste (HLW) and spent nuclear fuel (SNF) in salt (halite) host rock. The FEPs and other supporting information have been entered into a “SaltFEP” Database. The generic salt repository FEPs include consideration of relevant FEPs from a number of U.S., Dutch, German, and international FEP lists and should be a suitable starting point for any repository program in salt host rock. The salt FEP catalogue and database employ a FEP classification matrix approach that is based on the concept that a FEP is typically a process or event acting upon or within a feature. The FEP matrix provides a two-dimensional structure consisting of a Features/Components axis that defines the “rows” and a Processes/Events axis that defines the “columns” of the matrix. The design of the FEP classification matrix is consistent with repository performance assessment – the Features/Components axis is organized vertically to generally correspond to the direction of potential radionuclide migration (from the waste to the biosphere) and the Processes/Events axis is designed to represent the common two-way couplings between thermal processes and other processes (such as thermal-mechanical or thermal-hydrologic processes). Related FEPs can be easily identified – related FEPs will typically be grouped in a single matrix cell or aligned along a common row (Feature/Component) or column (Process/Event). The online SaltFEP database can be downloaded from www.saltfep.org. It contains the FEP matrix, the FEPs, and the associated processes for each FEP. It provides a starting point to create and document site-specific individual FEPs. Furthermore, the FEP matrix is connected to the Salt Knowledge Archive (SKA), a database of about 20,000 references and documents representing the historical knowledge on radioactive disposal in salt. This work is the result of an ongoing collaboration between researchers in the U.S., the Netherlands, and Germany, and supports the NEA Salt Club Mandate. It builds upon prior work which is documented.

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DOE OSTI2020

Brine Availability Test in Salt: Thermal-Hydrological-Mechanical-Chemical Simulations of a Heated Borehole in Salt - 20239

The Brine Availability Test in Salt (BATS) is a collaboration between Los Alamos, Sandia, and Lawrence Berkeley National Laboratories that is being conducted to reduce the uncertainties associated with disposal of spent nuclear fuel (SNF) and high-level radioactive waste (HLW) in a hypothetical bedded salt repository. Bedded salt formations may be an acceptable host rock for disposal of nuclear waste due to salt's high thermal conductivity, extremely low permeability, and self-healing capability. In this paper, we report on the first round of the BATS project, a 'shakedown test' known as Phase 1s, with an emphasis on the supporting numerical modeling. The experiment was conducted at the Waste Isolation Pilot Plant (WIPP) using previously drilled horizontal boreholes. The test included a heater borehole in which temperature, pressure, and moisture was monitored, and two nearby temperature monitoring boreholes. Numerical simulations that reflect the 36-year-long history of activity at the WIPP site are necessary to develop the appropriate initial conditions. Testing with different heater designs reveals that a 750 W radiative heater within a borehole can achieve the 120 deg. C borehole wall temperature goal. (authors)

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DOE OSTI2018

Leaching of Degraded Preservative Matrices, Based on Sodium Aluminophosphate Glasses, for High-Level Wastes

The chemical stability of vitrified high-level nuclear industry wastes in contact with water was studied experimentally in application to underground disposal of the wastes. The experiments were performed on samples of sodium aluminophosphate glass used in industrial solidification of wastes. Stable isotopes of strontium, cesium, uranium, and rare-earth elements as geochemical analogs of actinides were introduced into the glass to simulate radionuclides. It was observed that water vapor affects the degradation of the glass when vitrified wastes are placed in temporary repositories. It is shown that as a result of the degradation the rate of leaching of the glass in water increases significantly. In addition, a substantial portion of radionuclide simulators in colloidal form goes into the water medium.

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DOE OSTI2018

Operational Safety and Post-Closure Safety: Investigation of their Interactions by a FEP Analysis - 18040

The consideration of operational safety for a repository for high-level waste is a challenge, since it has to integrate the dimension of long-term safety. The operator has to demonstrate on the one hand the safety of the workers in the mine and of men and environment during operation. On the other hand a proper operation of the mine has to guarantee that the state of the facility at the time of closure will be compatible with the requirements for long-term safety. To achieve this, a pre-closure safety analysis should include (a) a general description of the structures, systems, components, equipment, and process activities during the operational phase, (b) identification and a systematic analysis of naturally occurring and human-induced hazards and (c) identification and documentation of the strong relationships between operational safety and post-closure safety. In a German R and D project, a FEP catalogue for the operational phase was developed based on the national disposal concepts for clay and salt. The FEP catalogue contains analyses of the processes and events that affect the components (features) of the repository system in its entirety. The FEP catalogue was applied to identify hazards for the operational phase and to document the impacts of these FEP (including the identified hazards) on post-closure safety. (authors)

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DOE OSTI2020

Preliminary Analysis of Postclosure DPC Criticality Consequences

One of the objectives of the United States (U.S.) Department of Energy's (DOE) Office of Nuclear Energy's Spent Fuel and Waste Science and Technology Campaign is to better understand the technical basis, risks, and uncertainty associated with the safe and secure disposition of spent nuclear fuel (SNF) and high-level radioactive waste. Commercial nuclear power generation in the U.S. has resulted in thousands of metric tons of SNF, the disposal of which is the responsibility of the DOE (Nuclear Waste Policy Act 1982). Any repository licensed to dispose the SNF must meet requirements regarding the longterm performance of that repository. For an evaluation of the long-term performance of the repository, one of the events that may need to be considered is the SNF achieving a critical configuration. Of particular interest is the potential behavior of SNF in dual-purpose canisters (DPCs), which are currently being used to store and transport SNF but were not designed for permanent geologic disposal. A two-phase study has been initiated to begin examining the potential consequences, with respect to longterm repository performance, of criticality events that might occur during the postclosure period in a hypothetical repository containing DPCs. Phase I, a scoping phase, consisted of developing an approach intended to be a starting point for the development of the modeling tools and techniques that may eventually be required either to exclude criticality from or to include criticality in a performance assessment (PA) as appropriate; Phase I is documented in Price et al. (2019). The Phase I approach guided the analyses and simulations done in Phase II to further the development of these modeling tools and techniques as well as the overall knowledge base. The purpose of this report is to document the results of the analyses conducted during Phase II. The remainder of Section 1 presents the background, objective, and scope of this report, as well as the relevant key assumptions used in the Phase II analyses and simulations. Subsequent sections discuss the analyses that were conducted (Section 2), the results of those analyses (Section 3), and the summary and conclusions (Section 4). This report fulfills the Spent Fuel and Waste Science and Technology Campaign deliverable M2SF-20SN010305061.

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DOE OSTI2021

GDSA Repository Systems Analysis Investigations in FY2021

The Spent Fuel and Waste Science and Technology (SFWST) Campaign of the U.S. Department of Energy Office of Nuclear Energy, Office of Spent Fuel and Waste Disposition (SFWD), has been conducting research and development on generic deep geologic disposal systems (i.e., geologic repositories). This report describes specific activities in the Fiscal Year (FY) 2021 associated with the Geologic Disposal Safety Assessment (GDSA) Repository Systems Analysis (RSA) work package within the SFWST Campaign. 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 host-rock environments, and to simulate and analyze these generic repository concepts and models using the GDSA Framework toolkit, and other tools as needed.

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DOE OSTI2022

GDSA Repository Systems Analysis Investigations in FY2022

The Spent Fuel and Waste Science and Technology (SFWST) Campaign of the U.S. Department of Energy Office of Nuclear Energy, Office of Spent Fuel and Waste Disposition (SFWD), has been conducting research and development on generic deep geologic disposal systems (i.e., geologic repositories). This report describes specific activities in the Fiscal Year (FY) 2022 associated with the Geologic Disposal Safety Assessment (GDSA) Repository Systems Analysis (RSA) work package within the SFWST Campaign. 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 host-rock environments, and to simulate and analyze these generic repository concepts and models using the GDSA Framework toolkit, and other tools as needed.

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DOE OSTI2023

Advective gas flow in bentonite: Development and comparison of enhanced multi-phase numerical approaches

Understanding the impact of repository gas, generated from degradation of waste and its interaction with the host rock, is essential when assessing the performance and safety function of long-term disposal systems for radioactive waste. Numerical models based on conventional multi-phase flow theory have historically been applied to predict the outcome and impact of gas flow on different repository components. However, they remain unable to describe the full complexity of the physical processes observed in water-saturated experiments (e.g., creation of dilatant pathways) and thus, the development of novel representations for their description is required when assessing fully saturated clay-based systems. This was the primary focus of Task A within the international cooperative project DECOVALEX-2019 (D-2019) and refinement of these approaches is the primary focus of this study (Task B in the current phase of DECOVALEX-2023). This paper summarises development of enhanced numerical representations of key processes and compares the performance of each model against high-quality laboratory test data. Experimental data reveals that gas percolation in water-saturated compacted bentonite is characterised by four key features: (i) a quiescence phase, followed by (ii) the gas breakthrough, which leads to a (iii) peak value, which is then followed by (iv) a negative decay. Three models based on the multiphase flow theory have been developed. These models can provide good initial values and reasonable responses for gas breakthrough (although some of them still predict a too-smooth response). Peak gas pressure values are in general reasonably well captured, although maximum radial stress differences are observed at 48 mm from the base of the sample. Here, numerical peak values of 12.8 MPa are predicted, whereas experimental values are about 11 MPa. These models are also capable of providing a reasonable representation of the negative pressure decay following peak pressure. However, other key specific features (such as the timing of gas breakthrough) still require a better representation. The model simulations and their comparison with experimental data show that these models need to be further improved with respect to model parameter calibration, the numerical representation of spatial heterogeneities in material properties and flow localisation, and the upscaling of the related physical processes and parameters. To further understand gas flow localisation, a new conceptual model has been developed, which shows that discrete channels can possibly be induced through the instability of gas-bentonite interface during gas injection, thus providing a new perspective for modeling gas percolation in low-permeability deformable media.

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