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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 307 records · Page 17
All sourcesNASA NTRSDOE OSTI
DOE OSTI2017

Representation of Global Climate Change in Performance Assessment Models for Disposal of Radioactive Waste - 17183

Evolving concepts of global climate change that include impacts of anthropogenic greenhouse gases require reassessment of conventional conservative approaches in performance assessment studies that assume cold and wet future glacial climates. Global paleo-climate data for the Quaternary (past 2 million years) show saw-tooth patterns (temperature/proxy temperature data versus time) of glacial and interglacial intervals accompanied by cyclical changes in the atmospheric content of CO{sub 2} (low contents during glacial cycles and higher contents during interglacial cycles). These cycles are correlated with changes in the earth's orbital configurations (Milankovitch cycles) but require nonlinear linkages with additional climate-forcing components to reproduce past glacial and interglacial transitions. The earth is currently in an interglacial period and the timing of the inception of the next glacial period is critically important for forecasting future climate states. Climate modeling studies coupling Milankovitch cycles with variable atmospheric CO{sub 2} content demonstrate conclusively that the earth is very unlikely to return to a glacial state when CO{sub 2} concentrations remain above pre-industrial levels (∼ 280 parts per million; current levels exceed 400 parts per million). A return to pre-industrial concentrations will likely require > 50,000 years and may require hundreds of thousands of years even if current emissions were to drop dramatically. The most likely scenario for the future global climate is persistence of the current interglacial period with variable but progressively increasing temperatures. Performance assessments need to revise site-specific evaluations of future climate assuming global warming within a prolonged interglacial period with increased probability of extreme weather events. Four radioactive waste disposal sites are illustrated that currently incorporate or will incorporate revised current climate concepts in modeling studies for performance assessments. (authors)

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DOE OSTI2017

Removal of Chalk River unidentified deposit (CRUD) radioactive waste by enhanced electrokinetic process

Decontamination techniques proposed and used to remove Chalk River unidentified deposit (CRUD) in radioactive waste management. In cases of huge volumes of metal or radionuclides contaminated by CRUD, removal of CRUD by mechanical or chemical decontamination is difficult. An advanced electrokinetic process combined with chemical decontamination was applied to remove CRUD and experimentally evaluated. We used oxalic acid for CRUD removal, and cobalt (Co) released from the CRUD was transferred to the cathode in an electrokinetic reactor. Our results indicate that the combined system is efficient for CRUD removal with enhanced, efficiency by use of the cation exchange membrane and zeolite.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2017

Development of an Integrated System for the Long-term Preservation of Records, Knowledge and Memory relating to Radioactive Waste Disposal Facilities - 17099

The Nuclear Energy Agency (NEA) has co-ordinated an international initiative to develop a common approach to the preservation of records, knowledge and long-term memory (RK and M), focused on radioactive waste disposals. The work programme has run in two phases, starting in 2011, with the current phase of the project due to be completed in 2018. Preparations for final reporting are now underway. The detailed implementation of an approach to RK and M preservation for any disposal facility will always remain a national decision. However, the NEA project has produced a 'menu' of tools and techniques that can be accessed and adapted to suit national needs. The common source of these menu components will help to ensure that the memory of the repository is kept alive, that messages to future populations can be clearly understood, and that evidence derived from the disposal environment can be properly interpreted. This paper describes the vision for implementation of the RK and M initiative within national programmes. Practical testing of some components of the menu of tools and techniques is currently underway in France, Sweden and USA, with encouraging results. Other disposal facilities, both under development or in planning, are invited to 'road test' the products and provide feedback. Refinement and maintenance of the tools remains important, and the NEA is committed to ensuring the continued accessibility and availability of the project outputs into the future. (authors)

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DOE OSTI2022

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

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

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DOE OSTI2018

Rheology of simulated radioactive waste slurry and cold cap during vitrification

During the vitrification of radioactive waste in a Joule-heated melter, aqueous melter feed slurry forms a cold cap, a reacting and melting material, which floats on the surface of the molten glass. The rheological behavior of the feed affects cold cap formation and shape, and is vital for modeling the feed-to-melt conversion process. Here, we used slurry feed simulant and fast-dried slurry solids representing the cold cap to investigate the rheological behavior of the feed as it transforms into glass. Both low-temperature and high-temperature rheometry were performed and a new scheme was applied to estimate the feed viscosity. This study shows that the conversion advances in four sequential stages that form distinct regions in the cold cap: (i) a fast-spreading boiling slurry from which water evaporates, (ii) a porous solid region (viscosity > 10 8 Pa s) containing reacting solids and molten salts, (iii) a plastic region in which glass-forming melt connects the refractory solids (~10 8 to ~10 6 Pa s), and (iv) a viscous foam layer in which the viscosity drops from ~10 5 to ~10 1 Pa s. The implications for the mathematical modeling of the cold cap are discussed.

glass melting↗
NASA NTRS1973

Space transportation and destination considerations for extraterrestrial disposal of radioactive waste

A feasibility study is summarized of extraterrestrial (space) disposal of radioactive waste. The initial work on the evaluation and comparison of possible space destinations and launch vehicles is reported. Only current or planned space transportation systems were considered. The currently planned space shuttle was found to be more cost effective than current expendable launch vehicles, by about a factor of two. The space shuttle will require a third stage to perform the disposal missions. Depending on the particular mission this could be either a reusable space tug or an expendable stage such as a Centaur. Of the destinations considered, high earth orbits (between geostationary and lunar orbit altitudes), solar orbits (such as a 0.90 AU circular solar orbit) or a direct injection to solar system escape appear to be the best candidates. Both earth orbits and solar orbits have uncertainties regarding orbit stability and waste package integrity for times on the order of a million years.

Zimmerman, A. V.↗
DOE OSTI2018

Innovation in Contracting Strategy for a Master Services Agreement for Remediation of Low Level Radioactive Waste - 18381

The Port Hope Area Initiative (PHAI) is a federal environmental clean-up program. Its mandate is the remediation and local, long-term, safe management of approximately 1.7 million cubic metres of historic low-level radioactive waste in the adjacent communities of Port Hope and Port Granby in Southern Ontario, Canada. The PHAI is being undertaken by Canadian Nuclear Laboratories (CNL), on behalf of Atomic Energy of Canada Limited, a federal Crown corporation. It is currently one of the largest environmental remediation projects in Canada. The safety of workers, the public and the environment while delivering the PHAI on time and within budget is CNL's number one priority. The PHAI is being carried out as two projects - the Port Hope Project and the Port Granby Project. The Port Hope Project is the larger of the two and includes the cleanup of approximately 1.2 million cubic metres of historic waste from several locations, including the Port Hope Harbour, a closed municipal landfill and approximately 800 residential properties. This complex project requires contracting the services of qualified, skilled and experienced environmental remediation contractors to meet the diverse needs of the myriad tasks that comprise the remediation and restoration of all properties. This paper will highlight how the CNL contracting services team developed an innovative approach to a contracting strategy designed to deliver the safe execution of work and keep the project on time and within budget. The approach addresses the complexities of a scope of work that consists of multiple work packages to be executed both concurrently and in tandem, and that require highly specialized trades/skills, as well as high-level project management expertise to ensure seamless coordination among the various contractors. The author will discuss CNL's strategy to engage the market with a Request for Expression of Interest to identify potential qualified suppliers and gauge their level of interest, issue a Request for Prequalification (RFPQ) to determine eligible contractors and, ultimately, issue a Request for Proposal for a Master Services Agreement (MSA) that would competitively evaluate each contractor's direct experience, technical expertise, quality management programs, safety systems compliance and track record. Using CNL's experience as a case study, the paper will highlight how an innovative approach can positively affect project timelines by describing how CNL addressed the challenge associated with the limited time available to implement procurement processes and deliver a series of executed agreements while ensuring that work would begin on time and key project milestones would be met. The author will summarize how CNL's strategy put the project back on schedule, without compromising the process guideline/requirements, and ensured that a qualified and experienced set of contractors, committed to a strong safety culture, will be available to perform the work required when the cleanup in the Port Hope community begins in 2018. (authors)

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DOE OSTI2019

Removal of Historic Low-Level Radioactive Waste from the Highland Drive South and Alexander Street Ravines - 19155

In Port Hope, the Highland Drive South Ravine (HDSR) is the discharge area for groundwater originating from the closed Highland Drive Landfill site. The presence of low-level radioactive waste (LLRW) in the HDSR is the result of a contaminated groundwater plume flowing from the landfill site. The contaminant plume contains elevated concentrations of uranium and arsenic and extends from the landfill south to discharge groundwater to shallow soils in a wet discharge area within the ravine. The groundwater discharges to Hunt's Ponds and Highland Drive South Creek, which are immediately downstream of the wet discharge area and will be included in the remediation design. Remediation and environmental management plans for HDSR have been developed within the framework of the Port Hope Project. Recent subsurface investigations have identified LLRW at greater depths than previously anticipated. Other investigations of surrounding properties have identified LLRW that is primarily from filling activities in backyards. The steep topography of the site and the presence of the creek and seepage face present remediation challenges. Also in Port Hope, the LLRW in the Alexander Street Ravine Area is the result of end dumping, filling of private backyards and stack deposition from the former refinery. The original site extent was focused on the deeper contamination due to end dumping and included a portion of Alexander Creek. However, recent investigations have confirmed other fill areas and the extent of stack deposition that are generally contiguous with the end dumping area. Restricted access to the site will result in all remediation in this area taking place via a single access point along the narrow existing Hayward Street. While both of these sites are ravines in Port Hope containing LLRW, the origins of contamination and remediation approaches are very different. This paper will compare these two challenging remediation sites. (authors)

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DOE OSTI2014

The Performance of Underground Radioactive Waste Storage Tanks at the Savannah River Site: A 60-Year Historical Perspective

The Savannah River Site produced weapons-grade materials for nearly 35 years between 1953 and 1988. The legacy of this production is nearly 37 million gallons of radioactive waste. Since the 1950s, the liquid waste has been stored in large, underground carbon steel waste tanks. During the past 20 years, the site has begun to process the waste so that it may be stored in vitrified and grout forms, which are more suitable for long-term storage. Over the history of the site, some tanks have experienced leakage of the waste to the secondary containment. This article is a review of the instances of leakage and corrosion degradation that the tanks and associated equipment have experienced since the first tanks were built. Furthermore, the activities that the site has taken to mitigate the degradation and manage the service life of the tank for its anticipated lifetime are reviewed.

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

Safe Packaging of Chemically Reactive Radioactive Waste - Addressing the Data Needs - 17487

Industry-standard thermal hazard screening is an effective, cost efficient approach to quickly obtain the required data typically utilized for safe scale-up of chemical processes and to accommodate changes to process recipes. Such thermal hazard screening is directly relevant to the packaging, transport, and storage of radioactive waste that is or can become chemically reactive. For such waste streams it is vital to identify safe temperature and pressure conditions and quantify adiabatic heat and gas generation rates in order to safely accommodate (or preclude) thermal instability within the waste package or storage facility. This paper illustrates widely-used thermal hazard screening bench-scale techniques that lend themselves to quickly identifying reactive hazards while providing directly scalable data for package/storage facility design. Example data are presented with discussion of how the data are analyzed for application to safe packaging and storage. (authors)

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DOE OSTI2022

Analysis of Glycolate in Radioactive Waste by Ion Chromatography (IC) and Proton Nuclear Magnetic Resonance (H NMR)

Here, in preparation for implementing the Nitric-Glycolic (NG) acid flowsheet for the Savannah River Site (SRS) Liquid Waste System (LWS), analytical methods for determining glycolate at low concentration, below 20 mg/L in radioactive samples, were developed to support system management and safety. To accommodate the wide range of LWS matrix conditions, two alternative methods were developed, refined, and demonstrated for glycolate analysis in radioactive waste samples: ion chromatography (IC) and a proton nuclear magnetic resonance (H NMR). Investigators validated IC and H NMR methods for glycolate analysis, defined the range of applicability, and demonstrated key supporting analytical protocols. The deployed IC method is applicable in low to moderate ionic strength samples and requires sample pretreatment using a Dionex OnGuard II H + cartridge. The deployed H NMR method is more labor intensive but provides options for a broader range of matrices. Based on the results, high quality glycolate analysis of the Defense Waste Processing Facility (DWPF) condensate in Tank 22 is feasible by IC down to approximately 12 mg/L. Using H NMR, glycolate may be determined to 8 mg/L or lower depending on the run time with the potential for broader applicability of the method to higher ionic strength conditions in other tanks of the SRS LWS.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2015

Design and performance testing of a novel far field Gamma system to assay radioactive waste in 400 and 200 litre drums - 15339

A Far Field Gamma ray Measurement system has been developed for measuring and assaying both 400 and 200 litre waste drums with a wide range of activity at nuclear power stations. These include radioactive waste drums with surface dose-rates of less than 2 mSv/h (200 mrem/h) and, at the higher range of activity, 400 litre drums with surface dose-rates of up to 50 mSv/h (5 rem/h). The assay system incorporates an electro-mechanically cooled high purity Germanium profile detector and advanced digital spectroscopic analysis electronics. A novel optimised conical-trapezoidal collimator has been incorporated to reduce the effect of the radioactive background in addition to the conventional detector lead shield with tin and copper graded lining to reduce the effect of lead X-rays. Two features have been incorporated to reduce detector dead time and maintain detector resolution when measuring high activity drums. The first is the inclusion of two automated tungsten filters of different thickness as an integral part of the detector collimator. The filters are used in conjunction with an automated rail system on which the detector platform is mounted. The detector can be positioned at different distances from the drum surface in order to reduce detector dead time. The deployment of the filters and the adjustment of the drum - detector position are automatically controlled based on a user defined dose-rate limit table. The data is provided by two Geiger-Muller dose-rate detectors, which are used to measure the drum surface dose-rate. This data is subsequently used to adjust the detector position and in cases of higher dose-rate deploy one or other of the tungsten filters. The drum rotation platform incorporates a load cell to determine drum weight and hence the drum density. This information is used to calculate an attenuation correction based on the assumption of uniform drum density and uniform distribution of activity within the drum. Where the waste drums have a more complicated and known regular internal structure, such as a small radioactive region surrounded by an annulus of shielding material, the analysis algorithms incorporated into the far field geometry and attenuation correction spectroscopic data analysis code are able to make a more accurate determination of drum activity than would be possible assuming a completely uniform drum. The detector energy and efficiency calibration is achieved using a point source with multiple gamma ray energy peaks. The response of the high purity Germanium detector is also modelled using the Monte Carlo Neutron Photon code and the model is benchmarked using the measured point source calibration data. The performance of the instrument has been determined for a range of uniform density matrix drums containing a set of Eu-152 line or rod sources located in re-entrant tubes positioned on an equal volume basis. When rotated, the test drums with volume distributed line sources simulate waste drums of different density each with a uniformly distributed radioactive source. The measurements have been validated using Monte Carlo simulations. (authors)

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DOE OSTI2025

Processes in Salt Repositories for Radioactive Waste Disposal

This document summarizes the key processes (thermal, hydrological, mechanical, and chemical; THMC) impacting the features of a deep geological repository for radioactive waste in salt. Some processes are natural and on-going whether the repository is there or not, and other processes are driven by the perturbation associated with the repository. The features considered here include both engineered and natural components of the repository system. The engineered barrier system (EBS) in a salt repository is quite different from those implemented for a repository in clay or crystalline rocks, because it is comprised mostly of granular salt and salt-compatible cements, rather than bentonite. When compared to other rocks (i.e., silicates), salt has unique properties that make it an excellent potential host rock. Openings and fractures in salt creep closed readily. Salt has high thermal conductivity, which can reduce peak temperatures. Additionally, far away from the excavations the porosity of salt is unconnected, which leads to essentially zero advective or diffusive transport. The small amount of hypersaline brine occurring in salt minimizes microbial activity, reduces colloid-assisted transport, and eliminates in-package criticality (i.e., chloride is a neutron poison). At the end of the report, we present a brief outline for a potential salt repository, including considerations avoided in previous repository disposal concepts. We propose considering higher-temperature processes in future disposal concepts, rather than trying to minimize the thermal perturbation of the repository. Since hot salt is drier, a dry repository would limit corrosion, gas generation, and solute transport. Openings and fractures creep shut faster in hot salt. Therefore, higher temperatures could be seen as beneficial, rather than something to minimize, through increased spacing between waste packages (increasing repository costs).

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