Engineering Papers⌕ Search

SEARCH · Engineering Papers

Results for “RADIOACTIVE WASTES”

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 361 records · Page 20
All sourcesNASA NTRSDOE OSTI
DOE OSTI2017

Functions and Requirements of the Radioactive Waste Test Platform

The purpose of this document is to provide the function and requirements for the Radioactive Waste Test Platform (hereafter called “test platform”). Historically, testing has been performed at the bench scale with actual waste samples (Fiskum et al. 2006a, b, 2008, 2009; Lumetta et al. 2009; Edwards et al. 2009; Smith et al. 2000, 2001). These tests form the basis for waste treatment operations at the Hanford Site. However, since 2008, there has been limited capability to perform bench-scale tests with actual waste samples. Prior work has shown that simulants can provide useful information about the performance of unit operations; however, there have been examples in the past where simulants did not adequately capture the expected performance of the tank waste materials. Therefore, as the Hanford Site moves towards initiating treatment of tank waste, it is important to have a platform available that can evaluate process options utilizing tank waste samples. This platform can be utilized to evaluate processing options and to address troubleshooting for unexpected system performance. The initial operations for the test platform will be to contribute towards maturing the Direct Feed Low-Activity Waste (DFLAW) feed qualification program and to demonstrate an option for disposition of the melter offgas stream. This equipment will be the ideal platform to evaluate additional process options including filtration operating conditions associated with precipitation, methods to extend resin lifetime associated with multi-cycle degradation, and evaluation of other approaches to minimize the impact of recycle streams from the effluent management system (including Tc) on melter performance.

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

Savannah River Site High Level Radioactive Waste Tank Sample Thermolytic Hydrogen Generation - 20346

To support the revision of flammability calculations and controls for the Savannah River Site (SRS) High-Level Radioactive Waste (HLW) tanks, Savannah River National Laboratory (SRNL) conducted laboratory measurements with the goal of quantifying hydrogen produced in HLW by non-radiolytic chemical reactions (i.e., thermolysis). Testing was performed with HLW tank supernate samples that represented a cross section of types of waste at SRS, including tanks that store typical evaporator concentrate, dissolved salt-cake, dilute recycle stream from the Defense Waste Processing Facility (DWPF), evaporator concentrate of the recycle stream from DWPF, fresh waste from the canyon separations facility, and waste that has been through salt processing cesium removal. Non-radioactive simulant testing included the classes of organic compounds historically introduced into the HLW tanks that were recently shown to be the most active toward the thermolytic generation of hydrogen. The reaction rate equations developed for thermolytic hydrogen generation from each class of organic compound showed direct proportionality to organic compound concentration and hydroxide concentration. Simulant testing identified that highly concentrated waste with high hydroxide concentration had the highest rates of thermolytic hydrogen generation. Using the radioactive tank sample thermolysis measurements at high temperatures and the mechanistic salt dependence of the simulant models, a global model was developed for tank waste thermolytic hydrogen generation as a function of total organic carbon content of SRS HLW. Through the primary functionality of the global model (organic carbon and free hydroxide concentration), the relative reactivity of the organic carbon in the waste was quantified. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
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)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
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)

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

Verification of the Accountability Method as a Means to Classify Radioactive Wastes Processed Using THOR Fluidized Bed Steam Reforming at the Studsvik Processing Facility in Erwin, Tennessee, USA - 13087

Studsviks' Processing Facility Erwin (SPFE) has been treating Low-Level Radioactive Waste using its patented THOR process for over 13 years. Studsvik has been mixing and processing wastes of the same waste classification but different chemical and isotopic characteristics for the full extent of this period as a general matter of operations. Studsvik utilizes the accountability method to track the movement of radionuclides from acceptance of waste, through processing, and finally in the classification of waste for disposal. Recently the NRC has proposed to revise the 1995 Branch Technical Position on Concentration Averaging and Encapsulation (1995 BTP on CA) with additional clarification (draft BTP on CA). The draft BTP on CA has paved the way for large scale blending of higher activity and lower activity waste to produce a single waste for the purpose of classification. With the onset of blending in the waste treatment industry, there is concern from the public and state regulators as to the robustness of the accountability method and the ability of processors to prevent the inclusion of hot spots in waste. To address these concerns and verify the accountability method as applied by the SPFE, as well as the SPFE's ability to control waste package classification, testing of actual waste packages was performed. Testing consisted of a comprehensive dose rate survey of a container of processed waste. Separately, the waste package was modeled chemically and radiologically. Comparing the observed and theoretical data demonstrated that actual dose rates were lower than, but consistent with, modeled dose rates. Moreover, the distribution of radioactivity confirms that the SPFE can produce a radiologically homogeneous waste form. The results of the study demonstrate: 1) the accountability method as applied by the SPFE is valid and produces expected results; 2) the SPFE can produce a radiologically homogeneous waste; and 3) the SPFE can effectively control the waste package classification. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
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.

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

Handling Radioactive Waste from the Proton Accelerator Facility at the Paul Scherrer Institut (PSI) - Always Surprising? - 13320

The Paul Scherrer Institut (PSI) is the largest national research centre in Switzerland. Its multidisciplinary research is dedicated to a wide field in natural science and technology as well as particle physics. In this context, PSI is operating, amongst others, a large proton accelerator facility since more than 30 years. In two cyclotrons, protons are accelerated to high speeds and then guided along roughly 100 m of beam line to three different target stations to produce secondary particles like mesons and neutrons for experiments and a separately beam line for UCN. The protons induce spallation processes in the target materials, and also at other beam loss points along the way, with emission of protons, neutrons, hydrogen, tritium, helium, heavier fragments and fission processes. In particular the produced neutrons, due to their large penetration depth, will then interact also with the surrounding materials. These interactions of radiation with matter lead to activation and partly to contamination of machine components and the surrounding infrastructures. Maintenance, operation and decommissioning of installations generate inevitably substantial amounts of radioactive operational and dismantling waste like targets, magnets, collimators, shielding (concrete, steel) and of course secondary waste. To achieve an optimal waste management strategy for interim storage or final disposal, radioactive waste has to be characterized, sorted and treated. This strategy is based on radiation protection demands, raw waste properties (size, material, etc.), and requirements to reduce the volume of waste, mainly for legal and economical reasons. In addition, the radiological limitations for transportation of the waste packages to a future disposal site have to be taken into account, as well as special regulatory demands. The characterization is a task of the waste producer. The conditioning processes and quality checks for radioactive waste packages are part of an accredited waste management process of PSI, especially of the Section Dismantling and Waste Management. Strictly proven and accepted methods needed to be developed and enhanced for safe treatment, transport, conditioning and storage. But in the field of waste from research activities, individual and new solutions have to be found in an increasingly growing administrative environment. Furthermore, a wide variety of components, with a really large inventory of radioactive nuclides, has to be handled. And there are always surprising challenges concerning the unusual materials or the nuclide inventory. In case of the operational and dismantling radioactive accelerator waste, the existing conditioning methods are in the process of a continuous enhancement - technically and administratively. The existing authorized specifications of conditioning processes have to be extended to optimize and fully describe the treatment of the inevitably occurring radioactive waste from the accelerator facility. Additional challenges are the changes with time concerning the legal and regulatory requirements - or do we have to consider it as business as usual? This paper gives an overview of the current practices in radioactive waste management and decommissioning of the existing operational accelerator waste. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
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 OSTI2013

Protecting Lake Ontario - Treating Wastewater from the Remediated Low-Level Radioactive Waste Management Facility - 13227

The Port Granby Project is part of the larger Port Hope Area Initiative, a community-based program for the development and implementation of a safe, local, long-term management solution for historic low level radioactive waste (LLRW) and marginally contaminated soils (MCS). The Port Granby Project involves the relocation and remediation of up to 0.45 million cubic metres of such waste from the current Port Granby Waste Management Facility located in the Municipality of Clarington, Ontario, adjacent to the shoreline of Lake Ontario. The waste material will be transferred to a new suitably engineered Long-Term Waste Management Facility (LTWMF) to be located inland approximately 700 m from the existing site. The development of the LTWMF will include construction and commissioning of a new Wastewater Treatment Plant (WWTP) designed to treat wastewater consisting of contaminated surface run off and leachate generated during the site remediation process at the Port Granby Waste Management Facility as well as long-term leachate generated at the new LTWMF. Numerous factors will influence the variable wastewater flow rates and influent loads to the new WWTP during remediation. The treatment processes will be comprised of equalization to minimize impacts from hydraulic peaks, fine screening, membrane bioreactor technology, and reverse osmosis. The residuals treatment will comprise of lime precipitation, thickening, dewatering, evaporation and drying. The distribution of the concentration of uranium and radium - 226 over the various process streams in the WWTP was estimated. This information was used to assess potential worker exposure to radioactivity in the various process areas. A mass balance approach was used to assess the distribution of uranium and radium - 226, by applying individual contaminant removal rates for each process element of the WTP, based on pilot scale results and experience-based assumptions. The mass balance calculations were repeated for various flow and load scenarios. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗