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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 487 records · Page 27
All sourcesNASA NTRSDOE OSTI
DOE OSTI2018

Immobilization of organic radioactive and non-radioactive liquid waste in a composite matrix

A method for immobilizing liquid radioactive waste is provided, the method having the steps of mixing waste with polymer to form a non-liquid waste; contacting the non-liquid waste with a solidifying agent to create a mixture, heating the mixture to cause the polymer, waste, and filler to irreversibly bind in a solid phase, and compressing the solid phase into a monolith. The invention also provides a method for immobilizing liquid radioactive waste containing tritium, the method having the steps of mixing liquid waste with polymer to convert the liquid waste to a non-liquid waste, contacting the non-liquid waste with a solidifying agent to create a mixture, heating the mixture to form homogeneous, chemically stable solid phase, and compressing the chemically stable solid phase into a final waste form, wherein the polymer comprises approximately a 9:1 weight ratio mixture of styrene block co-polymers and cross linked co-polymers of acrylamides.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2023

Product Consistency Test and Vapor Hydration Test Comparisons of a Radioactive Hanford Waste Glass with its Non-Radioactive Simulant Glass

Product Consistency Tests (PCT) and Vapor Hydration Tests (VHT) were performed on a radioactive glass made from Hanford tank AP-107 waste (i.e., R-AP-107) and its non-radioactive equivalent glass made from simulated tank waste (i.e., S-AP-107). The radioactive and non-radioactive experiments were conducted in parallel with identical equipment, protocols, and staff to remove additional experimental uncertainties. PCT and VHT results are reported for all samples and are shown in Table S.1. Average normalized mass loss of B ( NL B ) and Na ( NL Na ) calculated from the three replicate test results for each glass with their respective standard deviations (SD) are presented for PCT results while individual VHT rates ( r VHT ) with their propagated errors ( σ r prop ) from individual thickness measurements on each sample are presented for the VHT results. Some VHT samples experienced more than 50% evaporation loss during the experiments. However, no visible alteration was detected on any VHT samples regardless on the amount of evaporation loss detected as shown in Table S.1.

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DOE OSTI2019

The Cigeo Pilot Phase - 19024

Within the incremental development of the Cigeo deep geological repository, the Pilot Phase (PP) concept emerged during the 2013 public debate on Cigeo. The PP is a period of time that covers the beginning of the project, after its construction license is granted. It mainly aims at acquiring and / or confirming in situ, in a representative environment and in the operating conditions of Cigeo, the data selected for the design and safety demonstration. In parallel, it will emphasize the gradual and cautious approach for starting a complex industrial installation such as Cigeo. The Pilot Phase which is scheduled for a total duration of at least a decade will include 'inactive' tests (without radioactive waste) and then radioactive waste package disposal operations. The PP should be considered as a phase during which additional knowledge will be obtained and demonstration that cannot be fully acquired within the Underground Research Laboratory will be completed. (authors)

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DOE OSTI2018

The Use of Macrophytes as Biosorbents in Radioactive Liquid Waste Treatment - 18039

An important stage of the management of radioactive liquid waste is the chemical treatment. In the last few years, bio-sorption has attracted attention as one alternative method to remove radionuclides in liquid LL/ILW due to its efficiency and low costs. There are many biomasses that can be applied for this purpose, as agricultural residues and Macrophytes. This study aims at studying Macrophytes (Pistia Stratiotes, Limnobium Laevigatum, Lemna sp and Azolla sp) ability to absorb uranium in liquid solutions. The bio-sorption capacity of uranium by the Macrophytes was experimentally determined. The experiments determine the metal uptake after the solution was in contact with each bio-sorbents by inductively coupled plasma optical emission spectrometry (ICP-OES). The results show that these materials are potentially applicable in the treatment of liquid radioactive waste. (authors)

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DOE OSTI2020

Considerations for Regulation and Practical Implementation of Analyses of Inadvertent Human Intrusion - 20169

Recommendations from the International Commission for Radiological Protection (ICRP) and the International Atomic Energy Agency (IAEA) identify the need to consider potential inadvertent human intrusion into a near-surface radioactive waste disposal facility when developing limits for radionuclide concentrations that can be disposed. The recommendations also provide perspective on expectations for performance criteria and general assumptions for such analyses. Inadvertent human intrusion analyses are conducted to address the potential failure of institutional controls and loss of memory of a disposal facility. The analyses consider potential consequences of human actions that result in unknowingly bringing waste to the surface. Inadvertent human intrusion scenarios consider potential impacts to someone involved in an activity that actually disturbs the waste (e.g., drilling or excavation of a basement) or that exposes the person to emanations from the waste (e.g., radioactive gases) and also for the case where there may be longer term exposures to waste that has been brought to the surface. In practice, inadvertent human intrusion analyses have proven to be a critical consideration for the development of waste classification systems and for the determination of disposal requirements and limitations on wastes that can be disposed at shallow depths. The IAEA implemented a project on Human Intrusion in the context of Disposal of Radioactive waste (HIDRA) to further explore the regulatory expectations and practical implementation of analyses of potential inadvertent human intrusion. The paper begins with a discussion of international recommendations related to consideration of inadvertent human intrusion and then summarizes examples of some topical considerations related to addressing inadvertent human intrusion that are being discussed as part of the second phase of the HIDRA project. The topical considerations are planned to be addressed in a formal IAEA publication. All information in this paper simply reflects on-going project discussions and no formal recommendations are implied for these topical considerations at this time. (authors)

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DOE OSTI2017

Immobilization in Cement of Rice Husk used as Bio-sorbent of Radionuclides Present in Radioactive Liquid Waste - 17225

The use of new techniques for the treatment of radioactive liquid wastes has been studied in the last few years by many countries. In this context, the bio-sorption has been an attractive process. In Brazil the Radioactive Waste Management Laboratory of Nuclear and Energy Research Institute (IPEN-CNEN/SP), studied some bio-masses to be used as bio-sorbents to remove radionuclides of the radioactive liquid waste stored in this institute. The rice husk is a low cost and wide available byproduct of the agriculture and its bio-sorption capacity has been proved as feasible. Thus, it is necessary to evaluate if its immobilization in a cementitious matrix will result in a waste form that meets the requirements of Brazilian regulations for final disposal of low and intermediate level waste. This work aimed at evaluating the properties of the immobilization of rice husk in a cementitious matrix. Some properties of the waste form with incorporated rice husk were evaluated, as free water, workability, setting time and mechanical strength. The results indicated that cement matrices with 5 and 10% of biomass and water/cement ratio of 0.30 and 0.35 are viable for the production of waste forms that meets the Brazilian regulations. (authors)

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DOE OSTI2019

Development of Processing Technology for Radioactively Contaminated Water at Chernobyl Nuclear Power Plant - 19656

During the operation of the Chernobyl Nuclear Power Plant, distillation by evaporation was the main method for processing radioactively contaminated water. When the plant was decommissioned, the evaporation facility was mothballed and floor drain treatment was suspended in 2016. At present, the existing tanks with the capacity of 26,000 m{sup 3} are more than 60% full with radioactively contaminated floor drains, which requires alternative processing methods. The current process flow for radioactive waste management at State Specialized Enterprise 'Chernobyl NPP' includes processing of accumulated liquid radioactive wastes at a liquid radioactive wastes processing plant. Under the processing technology, liquid radioactive waste is immobilized at the plant by cementation and the resulting solid radioactive waste is then buried in a near-surface storage. Key factors for the selection of alternative radioactively contaminated water processing methods are the possibility to clear the processed water (no control monitoring), which requires the same degree of processing as for drinking water, and minimization of radioactive waste generation due to the lack of free space for solid waste disposal in the near-surface storage. Floor drains that require processing contain radionuclides such as {sup 137}Cs, {sup 90}Sr, {sup 60}Co and {sup 40}K. The current level of the total beta and gamma activity is about 4*105 Bq/kg. The following criteria are set for the Chernobyl NPP site for water to be disposed to the sewerage system: - Max. 37 Bq/kg for the total beta and gamma containing radionuclide activity; - Max. 0.7 Bq/kg for the total alpha containing radionuclide activity. METOIL produces inorganic sorbents for removing radionuclides from liquids and develops sorbent-based technology for a variety of applications related to liquid radioactive waste processing. In 2017-2018, experts from the Chernobyl NPP and METOIL ran several joint laboratory tests to evaluate the efficiency of inorganic sorbents made by METOIL for processing floor drains at the Chernobyl NPP. The test procedure included treatment of radioactively contaminated water in a two-stage sorption process: static sorption (sorbents are mixed in a liquid) and dynamic sorption (the liquid is filtered through a sorbent column). NIKET and MODIKS sorbents were provided for the tests. The results of the test measurements showed that the total alpha containing radionuclide activity is less than 0.026 Bq/kg and the total beta and gamma containing radionuclide activity is 9.2 Bq/kg. The physical and chemical parameters after processing also met the water disposal criteria. High sorption capacity of the sorbents used for the tests allows processing hundreds of water volumes with just one sorbent volume. At present, preparations are underway for full-scale tests at a pilot plant with a capacity of minimum 10 m{sup 3}/day. At a later stage, spent sorbents may be immobilized by inclusion into the cement matrix, which will substantially reduce the amount of radioactive waste to be buried. (authors)

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DOE OSTI2021

A novel defoamer for processing nuclear waste: Testing and performance

Abstract Legacy radioactive waste from nuclear weapons material processing is the biggest environmental challenge in the state of South Carolina, and one of the biggest in the United States. Although substantial progress has been made in processing and vitrifying high‐level radioactive waste at the Savannah River site, approximately 35 million gallons remains to be treated and dispositioned. In this article, we show the development of a new defoamer for use in the processing of high‐level radioactive waste. The application of the new defoamer is not only more effective at controlling foam but will shorten the batch processing time leading to safer and more efficient processing. This has the potential to help shorten the overall site mission lifetime, saving the federal government hundreds of millions of dollars. The novel aspect of the new defoamer is its effectiveness and stability in the harsh conditions needed for processing high‐level radioactive waste: high temperatures, pH 4–13, strong oxidizing and reducing agents, and numerous metal catalysts. The defoamer will replace an existing antifoam agent that decomposes to form multiple flammable gases. Because the existing antifoam agent decomposes quickly, processing changes increased processing time. Statement of novelty We demonstrate that a new defoamer has been developed for use in processing high‐level radioactive waste. This novel defoamer is effective and stable in the harsh processing conditions with no detectable flammable by‐products. Because the existing antifoam agent decomposes quickly, processing changes, including lowering acid addition rates and evaporation rates, increased processing time. The new defoamer is not only more effective at controlling foam but will shorten the batch processing time and lead to safer and more efficient processing by eliminating the production of flammable gases. Statement of industrial relevance Foaming is of great concern in many industrial processes involving three‐phase gas/liquid/fine‐solids systems, like in water evaporation, mineral floatation, air sparging in situ remediation techniques, and in those found in the paper industry.

Lambert, Dan P.↗
NASA NTRS1973

Feasibility of Space Disposal of Radioactive Nuclear Waste. 1: Executive Summary

This NASA study, performed at the request of the AEC, concludes that transporting radioactive waste (primarily long-lived isotopes) into space is feasible. Tentative solutions are presented for technical problems involving safe packaging. Launch systems (existing and planned), trajectories, potential hazards, and various destinations were evaluated. Solar system escape is possible and would have the advantage of ultimate removal of the radioactive waste from man's environment. Transportation costs would be low (comparable to less than a 5 percent increase in the cost of electricity) even though more than 100 space shuttle launches per year would be required by the year 2000.

Source record↗
DOE OSTI2016

WOW Technology's Innovative Radioactive Liquid Waste Treatment - 16128

WOW presents its revolutionary technology and equipment for liquid radioactive waste treatment: outperforming ultimate water decontamination and purification process, enhanced sludge concentration, no secondary waste nor consumables, fully automated, remote controlled and self-decontaminating device. The WOW's technology is based upon a never before observed discovery of fluid dynamics science: the possibility of performing a molecular separation between solute and suspended elements and the solvent. The combination of such a molecular separation process with a standard vacuum evaporation improves the abatement performances by thousands of times, with respect to those of the state of the art vacuum evaporators. In addition to this, no secondary waste is produced during the process, as no filters, membranes, resins or additives are used. WOW equipment, automated and remote controlled, self decontaminates after use and can be designed and constructed either tailored to the application needs or with a modular approach for enhanced transportability and application flexibility. After the preliminary verification by CNR, the Italian National Research Center, WOW Technology's decontamination device was tested c/o LENA, the Laboratory of Applied Nuclear Energy of the University of Pavia, Italy with a simulated solution 6000 times more contaminated than the nuclear reactor's cooling water of Fukushima-Daiichi NPP. In addition to that, WOW's technology was also used in a real case at the Radiochemistry laboratory of the Pavia's University Chemistry department. Both the above mentioned contaminated fluids have been successfully decontaminated without production of additional or secondary waste. WOW's technology has already performed on industrial scale c/o the Nuclear Repository of S.S.M. in Saluggia, Italy: 45000 liters of acid radioactive solution have been successfully decontaminated to a Decontamination Factor (DF) of 335000 for Cs-137 by one single evaporation step and without using any additional tool to control steam. [Performances independently certified by: National Physics Laboratory (UK) and Pavia University Radiochemistry Laboratory, Chemistry Department (Italy)]. The high decontamination factors and the high volume reduction of the sludge make this technology suitable for the ultimate treatment of highly contaminated liquid waste such as the Fukushima-Daiichi nuclear reactor's cooling water. The WOW's technology addresses as well to interventions in both, operating and decommissioning nuclear power plants and any other nuclear and radioactive facility whose produced and/or stored radioactive liquid solutions, including acid liquid solutions, can be successfully treated, the waste volume reduced and no secondary waste produced. (authors)

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DOE OSTI2016

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

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

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DOE OSTI2019

Effects of bentonite heating on U(VI) adsorption

Engineered barrier systems designed to isolate high-level radioactive waste utilize bentonite, a montmorillonite rich material, to restrict contaminant transport due to its low hydraulic conductivity and high adsorption capacity. High temperatures (100–200 °C) near waste canisters resulting from radioactive decay may alter the clay's ability to adsorb contaminants. In this study, we examine U(VI) adsorption onto two different bentonite samples subjected to (1) 18 years of in situ heating during an underground experiment and (2) short-term (7-week), high temperature (300 °C) heating in the laboratory. Results show that U(VI) adsorption was lower for field-heated bentonite located closest to the heater, which experienced temperatures of approximately 95 °C, compared to a control sample, which experienced temperatures of approximately 20 °C, over a range of aqueous chemical conditions. On average, K d values for U(VI) adsorption were 31% lower for 95 °C heated samples. By contrast, U(VI) adsorption onto intermediate-heated bentonite (50 °C) was indistinguishable from the adsorption onto the cold-zone (20 °C) sample. U(VI) adsorption onto lab-heated bentonite was also lower than onto control bentonite over the pH range 4.5–8.0, with an average decrease in K d values of 50% after heating. Lower U(VI) adsorption to field-heated bentonite persisted after bentonite was purified to isolate the clay fraction. This allows us to rule out changes in pore-water chemistry or accessory mineral composition as causes of the lower adsorption. No evidence of montmorillonite illitization was observed in the heated samples. While some of the lower U(VI) adsorption in the lab-heated bentonite can be explained by changes in aqueous U(VI) speciation, we propose that lower U(VI) adsorption to field-heated bentonite may be primarily due to changes in the montmorillonite edge structure. The observed changes in U(VI) adsorption to bentonite after heating have implications for U(VI) diffusive transport through engineered barriers and must be considered when designing radioactive waste disposal repositories.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

System Model Calculations for the Active Trenches of the Low-Level Burial Grounds, Hanford Site, Washington

Two lined trenches (Trench 31 and 34) located within the 200 West Low-Level Burial Ground (LLBG) area of the U.S. Department of Energy’s (DOE) Hanford Site are designated for permanent disposal of low-level radioactive waste (LLW) and mixed low-level radioactive waste (MLLW). In accordance with DOE O 435.1, Radioactive Waste Management, radioactive waste shall be managed and disposed in a manner that is protective of worker, public health and safety, and the environment. DOE O 435.1 requires a site-specific radiological performance assessment (PA) that includes calculations of potential releases and subsequent doses to members of the public for a period of 1,000 years after closure of a low-level waste (LLW) disposal facility. The purpose of this environmental calculation file (ECF) is to document the various dose calculations performed to support DOE’s LLBG PA. The dose assessments are performed to evaluate the potential exposure of an all-pathway representative person to radionuclide contaminants of potential concern (COPCs) that may be released from the LLBG Trenches 31 and 34 to the point of calculation (POC) located at the outer edge of a 100 m buffer zone surrounding the trench boundary. This ECF uses inputs from other ECFs and their associated model package reports (MPRs) supporting the LLBG PA. The five major objectives of this ECF are to present the results of the following: (1) Deterministic dose assessments for both groundwater and atmospheric exposure scenarios; (2) Radon-222 (Rn-222) flux emanating at the surface from the disposed wastes; (3) Uncertainty analyses of the groundwater pathway annual dose due to uncertainty in input parameters; (4) Sensitivity analyses of the input parameters that could potentially impact the fate and transport of the contaminants and doses; (5) Doses arising from inadvertent intruder exposure scenarios.

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DOE OSTI2019

Panel Session 105: US DOE LLW On-Site Disposal: LLW Disposal Facility Federal Review Group Review Process

This panel focuses on risk-informed provision for waste management of US DOE radioactive wastes. Detailed information on the high-level radioactive waste interpretation Federal Register Notice will be discussed to better educate the audience in understanding that interpretation. The Co-Chairs and panelists each introduced themselves. Presentations proposed include: 1 - Overview of Public Comments on the Proposed High-Level Radioactive Waste Interpretation (Theresa Kliczewski, US DOE EM (USA)); 2 - ECA Presentation on Implementation Approaches and Expectations for Radwaste Management; DOE Order 435.1 (Kara Colton, Energy Communities Alliance (USA)); 3 - 2019 Grassroots Efforts to Clarify HLW Definition (Rick McLeod, Savannah River Site Community Reuse Organization - SRSCRO (USA))

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DOE OSTI2020

Low Level Radon Concentration and Radon Daughter Dose Monitoring in an Arid-zone Waste Store - 20081

Radioactive waste containing predominantly soil and rubble from the clean-up of a former uranium and thorium minerals processing research facility is currently stored at the Woomera Test Range in South Australia awaiting final disposal. Prior to commencing any manipulation of the nearly 10,000 waste drums (205 L or 55 gallon each), radon concentration monitoring and radon daughter dosimetry has been undertaken to (i) gain understanding of the temporal and spatial variation of radon and its daughters within the storage facility, which could provide insights into the radiological source term and (ii) confirm that radiological risks to workers are below the guideline values. A time series of nearly 12 months of detailed observations shows low level radon concentrations with annual average values that are slightly above the average radon concentrations in Australian homes. Peak values in summer are about three times higher than the annual average based on the CRM measurements. Both radon concentrations and annual effective dose are well below the regulatory limits. (authors)

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DOE OSTI2016

Hot Isostatic Pressing of Radioactive Nuclear Waste: The Calcine at INL - 16348

Hot Isostatic Pressing (HIP) is a technology that has been around for 60+ years. By using high temperature and high gas pressure, dry metal and ceramic powders can be consolidated and a volume decrease can be achieved. This paper presents the efforts using the HIPing process at the Idaho National Laboratory to treat the calcine radioactive waste. Once loaded into collapsible canisters, the HIPing would be used to treat the waste before final disposal. The resulting volume reduction was shown to be 20-70% and the cost ratio vs vitrification is 1:1.74. (authors)

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

Disposition Options for Sodium Cooled Fast Reactor (A White Paper)

The sodium-cooled fast reactor (SFR) design concept is one of the six classes of nuclear reactors in the GenIV initiative. SFRs are uranium or plutonium-fueled reactors operating in the fast neutron spectrum using liquid sodium as the coolant. SFRs can be designed as a breeder reactor or actinide-burning reactor in addition to operating the thorium fuel cycle. While having different fuel designs, the anticipated waste streams, and the necessary management strategies for spent nuclear fuel (SNF) and radioactive wastes from SFRs are very similar. This includes the SNF, activated sodium coolant, in-core stainless-steel components, piping, resins and filters, solidified liquid waste, contaminated equipment, and other radioactive wastes. Modern SFR designs are based on a long and rich operating history of several liquid-metal-cooled fast reactors with sodium coolant. Several of these reactors have been shut down, the fuel has been placed in safe storage, and they have undergone some degree of decommissioning. As such, there is significant experience in the management of the SNF and radioactive wastes associated with operating these reactors. This white paper will identify the definitions and regulations that apply to the safe and secure management, storage, and disposal of radioactive waste and identify the key radioactive waste streams from SFRs. Idaho National Laboratory has significant experience in the management of the SNF from the SFR predecessors. This experience should form the basis for the management and disposition efforts of the radioactive waste from any new SFR-type small modular reactor or microreactor intended for deployment at the Idaho National Laboratory Site

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗