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All sourcesNASA NTRSDOE OSTI
DOE OSTI2014

Alternative Electrochemical Salt Waste Forms, Summary of FY11-FY12 Results

The Fuel Cycle Research and Development Program, sponsored by the U.S. Department of Energy Office of Nuclear Energy, is currently investigating alternative waste forms for wastes generated from nuclear fuel processing. One such waste results from an electrochemical separations process, called the “Echem” process. The Echem process utilizes a molten KCl-LiCl salt to dissolve the fuel. This process results in a spent salt containing alkali, alkaline earth, lanthanide halides and small quantities of actinide halides, where the primary halide is chloride with a minor iodide fraction. Pacific Northwest National Laboratory (PNNL) is concurrently investigating two candidate waste forms for the Echem spent-salt: high-halide minerals (i.e., sodalite and cancrinite) and tellurite (TeO2)-based glasses. Both of these candidates showed promise in fiscal year (FY) 2009 and FY2010 with a simplified nonradioactive simulant of the Echem waste. Further testing was performed on these waste forms in FY2011 and FY2012 to assess the possibility of their use in a sustainable fuel cycle. This report summarizes the combined results from FY2011 and FY2012 efforts.

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DOE OSTI2019

Drying of Crystalline Silicotitanate (CST) Beds by Air Flow

The Direct Feed Low-Activity Waste (DFLAW) process has been proposed to support early production of immobilized low-activity waste (LAW) at the Hanford Site. The tank farm contractor Washington River Protection Solutions, LLC (WRPS) is developing the Tank-Side Cesium Removal (TSCR) system to accomplish the treatment objectives. The TSCR system is being designed and built by AVANTech, Inc. for WRPS and will use a filter to remove any entrained solids in the LAW and will use columns of crystalline silicotitanate (CST) ion exchange media to remove 137 Cs. A key step in preparing used columns of CST for transportation and interim storage is draining and drying of the CST beds. The purpose of this study is to document testing activities that quantify the drying of full-height beds of CST material in a laboratory column that is scaled to represent the drying process of the planned full-scale CST columns to be used in the TSCR system. The purpose also includes demonstrating that CST material within the scaled column, when dried with air injection, will pass the U.S. Environmental Protection Agency’s Paint Filter Liquids Test.

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DOE OSTI2021

Characterization of CST Post-Processing AP-105 Hanford Tank Waste

The primary goal of the Tank Side Cesium Removal (TSCR) system, under development by Washington River Protection Solutions, LLC (WRPS), is to remove entrained solids and 137 Cs from the Hanford tank waste supernate to expedite production of low-activity waste. Ion exchange (IX) testing of 10.9 L of waste from Hanford tank 241-AP-105 (AP-105), performed by Pacific Northwest National Laboratory, used a lead-lag-polish column format, with a bed volume of 10 mL per column, to decontaminate tank waste supernate using crystalline silicotitanate (CST) as the IX media. The AP-105 Cs IX processing test, discussed elsewhere, resulted in a shorter transition zone (i.e., steeper load curve) than those defined by wastes from tanks 241-AP-107 and 241-AW-102.1 The shorter transition zone was indicative of a matrix effect retarding Cs capacity. Therefore, aliquots of spent CST from the lead, lag, and polish columns were subjected to a digestion protocol to quantify analytes retained by the CST and extrapolate the impact on Cs capacity. The spent CST was digested using a combination of 5 M HNO 3 and H 2 O 2 with vigorous heating and stirring. Due to the radiation dose accompanying the 137 Cs on the CST columns, a secondary Cs separation by ammonium molybdophosphate embedded in polyacrylonitrile (AMP-PAN) was performed to separate the 137 Cs from the CST so the samples could be contact-handled for analysis outside of a shielded facility.

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DOE OSTI2017

Conversion of a wet waste feedstock to biocrude by hydrothermal processing in a continuous-flow reactor: grape pomace

Wet waste feedstocks present an apt opportunity for biomass conversion to fuels by hydrothermal processing. In this study, grape pomace slurries from two varieties, Montepulciano and cabernet sauvignon, have been converted into a biocrude by hydrothermal liquefaction (HTL) in a bench-scale, continuous-flow reactor system. Carbon conversion to gravity-separable biocrude product up to 56 % was accomplished at relatively low temperature (350 C) in a pressurized (sub-critical liquid water) environment (20 MPa) when using grape pomace feedstock slurry with a 16.8 wt% concentration of dry solids processed at a liquid hourly space velocity of 2.1 h-1. Direct oil recovery was achieved without the use of a solvent and biomass trace mineral components were removed by processing steps so that they did not cause processing difficulties. In addition, catalytic hydrothermal gasification (CHG) was effectively applied for HTL byproduct water cleanup using a Ru on C catalyst in a fixed bed producing a gas composed of methane and carbon dioxide from water soluble organics. Conversion of 99.8% of the chemical oxygen demand (COD) left in the aqueous phase was demonstrated. As a result, high conversion of grape pomace to liquid and gas fuel products was found with residual organic contamination in byproduct water reduced to <150 mg/kg COD.

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DOE OSTI2016

Dry Pretreatment of Used Nuclear Fuel for Tritium Removal: A Review

Virtually all the elements in the periodic chart are generated as a result of atomic fission in nuclear fuel. Several of the elements are volatile or can readily react with other chemicals to form volatile molecules that can be released during the processing of used nuclear fuel for the recovery of the valuable products. Among the volatile fission products, {sup 129}I, {sup 14}C, {sup 3}H, {sup 85}Kr, and {sup 79}Se constitute the main environmental concern for uncontrolled release. Storage of used fuel allows the short-lived isotopes to decay significantly (e.g., {sup 3}H decays in half every 12.32 years and {sup 85}Kr every 10.76 years). However, depending on regulatory limits they may remain an environmental concern up to about 50-100 years of fuel storage. In contrast, the concern for environmental release of the highly mobile {sup 129}I and {sup 14}C remains for geological times (half-lives of 15.7 million years for {sup 129}I and 5730 years for {sup 14}C). Processing of used commercial nuclear fuel requires significant head-end operations to remove the fuel from the protective coverings (e.g., assemblies). The current industrial process includes cropping the end caps of the fuel assemblies, chopping of the fuel pins and/or bundles into short segments, and prolonged leaching of the segments using concentrated nitric acid (HNO{sub 3}). The infrastructure, processes, and waste treatment systems in the head-end operations, including the associated off-gas system, are some of the larger and more costly components of the entire plant. The bulk of the volatile fission products are released to the off-gas during the leaching and dissolution. However, a small but significant amount will transfer to the following steps of processing such as solvent extraction and will show up in many different waste streams. The effective capture of {sup 3}H and iodine throughout the different processes and disposition of associated waste streams is costly and complicated. Up-front removal and capture is highly desirable. Currently, at the U.S. Department of Energy national laboratories and internationally, several dry pretreatment technologies are being studied and developed with the potential to simplify and increase the safety of subsequent storage or chemical recycling processes. These studies include (1) alternative decladding methods, (2) oxidation of exposed fuel following either conventional shearing or chemical decladding, and (3) the collection and stabilization of released volatile radionuclides. Several dry processes are being developed for the up-front removal of tritium and other fission products. They can be combined with different decladding alternatives and, in general, produce a fine powder that can be fed to aqueous or pyrochemical separations processes. (authors)

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DOE OSTI2020

Cesium Removal Performance Comparisons of Crystalline Silicotitanate Media Batches with Savannah River Site Waste Simulant

The Tank Closure Cesium Removal (TCCR) system uses ion exchange columns filled with Crystalline Silicotitanate (CST) media to process radioactive waste solutions for the removal of Cs137. TCCR currently focuses on dissolving Savannah River Site (SRS) Tank 10H waste (primarily sodium salt cake solids) within the tank followed by at-tank ion exchange column treatment. Plans are underway to prepare and install a second TCCR unit at SRS. Capacity and particle size differences exist between archived (IE-911) and more recently-prepared CST media batches (9120- B and 9140-B). Side-by-side comparison testing was performed to evaluate the cesium removal performance of each batch to aid in selecting the preferred CST batch and media characteristics to load into the second TCCR unit. Batch contact equilibrium and flow-through column tests have been conducted with three CST batches using an SRS Average Simulant. Simulant batch contact equilibrium cesium loading results for the three CST batches (including two lots of one batch) are provided in Table ES-1. The 35 °C data indicates that the archived IE-911 batch has a higher cesium capacity than recently-prepared CST media and that the minor TCCR CST 9120-B media lot (2099000035) has similar cesium removal performance to the major TCCR lot (2099000034). Tests conducted at 25 ºC for the archived IE-911 CST batch indicated lower cesium removal performance with this simulant batch than was observed recently with a different SRS Average Simulant batch. A dilution factor (DF) is typically utilized when modeling engineered CST media cesium loading performance to account for mass contributions from the binder material. In cases where CST performance is lower than expected, this factor includes corrections for the binder and for low performance. The ZAM (Zheng, Anthony, and Miller) Isotherm Model DF values are provided in Table ES-1 for each simulant batch contact result. DF values near 0.5 were determined for the 9120-B and 9140-B CST batch contact tests while DF values near 0.6 (20% higher) were determined for the tests with IE-911. These dilution factors are lower than recently observed with a different SRS Average Simulant batch (9140-B DF = 0.68; IE-911 DF = 1.0). Based on these results, it appears that some component in the simulant solution used for equilibrium testing may have resulted in reduced cesium loading on the CST media.

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NASA NTRS2004

Improved Method of Purifying Carbon Nanotubes

An improved method of removing the residues of fabrication from carbon nanotubes has been invented. These residues comprise amorphous carbon and metal particles that are produced during the growth process. Prior methods of removing the residues include a variety of processes that involved the use of halogens, oxygen, or air in both thermal and plasma processes. Each of the prior methods entails one or more disadvantages, including non-selectivity (removal or damage of nanotubes in addition to removal of the residues), the need to dispose of toxic wastes, and/or processing times as long as 24 hours or more. In contrast, the process described here does not include the use of toxic chemicals, the generation of toxic wastes, causes little or no damage to the carbon nanotubes, and involves processing times of less than 1 hour. In the improved method, purification is accomplished by flowing water vapor through the reaction chamber at elevated temperatures and ambient pressures. The impurities are converted to gaseous waste products by the selective hydrogenation and hydroxylation by the water in a reaction chamber. This process could be performed either immediately after growth or in a post-growth purification process. The water used needs to be substantially free of oxygen and can be obtained by a repeated freeze-pump-thaw process. The presence of oxygen will non-selectively attach the carbon nanotubes in addition to the amorphous carbon.

Delzeit, Lance D.↗
DOE OSTI2012

Decontamination of Nuclear Liquid Wastes Status of CEA and AREVA R and D: Application to Fukushima Waste Waters - 12312

Liquid wastes decontamination processes are mainly based on two techniques: Bulk processes and the so called Cartridges processes. The first technique has been developed for the French nuclear fuel reprocessing industry since the 60's in Marcoule and La Hague. It is a proven and mature technology which has been successfully and quickly implemented by AREVA at Fukushima site for the processing of contaminated waters. The second technique, involving cartridges processes, offers new opportunities for the use of innovative adsorbents. The AREVA process developed for Fukushima and some results obtained on site will be presented as well as laboratory scale results obtained in CEA laboratories. Examples of new adsorbents development for liquid wastes decontamination are also given. A chemical process unit based on co-precipitation technique has been successfully and quickly implemented by AREVA at Fukushima site for the processing of contaminated waters. The asset of this technique is its ability to process large volumes in a continuous mode. Several chemical products can be used to address specific radioelements such as: Cs, Sr, Ru. Its drawback is the production of sludge (about 1% in volume of initial liquid volume). CEA developed strategies to model the co-precipitation phenomena in order to firstly minimize the quantity of added chemical reactants and secondly, minimize the size of co-precipitation units. We are on the way to design compact units that could be mobilized very quickly and efficiently in case of an accidental situation. Addressing the problem of sludge conditioning, cementation appears to be a very attractive solution. Fukushima accident has focused attention on optimizations that should be taken into account in future studies: - To better take account for non-typical aqueous matrixes like seawater; - To enlarge the spectrum of radioelements that can be efficiently processed and especially short lives radioelements that are usually less present in standard effluents resulting from nuclear activities; - To develop reversible solid adsorbents for cartridge-type applications in order to minimize wastes. (authors)

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DOE OSTI2017

Drying Model for Fukushima Fuel Debris - 17489

The FATE{sup TM} code is a well-established software tool for simulation of processing, shipping, and interim storage of damaged spent nuclear fuel (SNF) that was used for process design and safety analysis for the Hanford Spent Nuclear Fuel Program (SNFP). This paper describes new model developments that allow FATE to be used for simulation of drying of damaged LWR fuel and, in particular, for simulation of processing, shipping, and interim storage of Fukushima fuel debris. The model is capable of quantifying important tradeoffs in the debris chunk size, process conditions such as temperature and pressure, container size, etc. to optimize process design. The model is also capable of demonstrating proof-of-dryness and to demonstrate the link between observed quantities such as pressure and the amount of residual water. (authors)

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NASA NTRS1980

Applications of thermal energy storage to waste heat recovery in the food processing industry

A study to assess the potential for waste heat recovery in the food industry and to evaluate prospective waste heat recovery system concepts employing thermal energy storage was conducted. The study found that the recovery of waste heat in canning facilities can be performed in significant quantities using systems involving thermal energy storage that are both practical and economical. A demonstration project is proposed to determine actual waste heat recovery costs and benefits and to encourage system implementation by the food industry.

Wojnar, F.↗
NASA NTRS2000

Process and Equipment for Nitrogen Oxide Waste Conversion to Fertilizer

The present invention describes a process for converting vapor streams from sources containing at least one nitrogen-containing oxidizing agent therein to a liquid fertilizer composition comprising the steps of: (1) directing a vapor stream containing at least nitrogen-containing oxidizing agent to a first contact zone; (2) contacting said vapor stream with water to form nitrogen oxide(s) from said at least one nitrogen- containing oxidizing agent; (3) directing said acid(s) as a second stream to a second contact zone; (4) exposing said second stream to hydrogen peroxide which is present within said second contact zone in a relative amount of at least 0.1% by weight of said second stream within said second contact zone to convert at least some of any nitrogen oxide species or ions other than in the nitrite form present within said second stream to nitrate ion; (5) sampling said stream within said second contact zone to determine the relative amount of hydrogen peroxide within said second contact zone; (6) adding hydrogen peroxide to said second contact zone when a level on hydrogen peroxide less than 0.1% by weight in said second stream is determined by said sampling; (7) adding a solution comprising potassium hydroxide to said second stream to maintain a pH between 6.0 and 11.0 within said second stream within said second contact zone to form a solution of potassium nitrate; and (8) removing sais solution of potassium nitrate from said second contact zone.

Lueck, Dale E.↗
DOE OSTI2016

Utilization and Application of Kurion Mobile Processing System in Remediation of Radioactively Contaminated Water - 16370

Kurion{sup R} Mobile Processing Systems (KMPS) were deployed in Fukushima as first-of-a- kind at-tank isotope removal systems to remediate Sr from reverse osmosis concentrated salt water. The systems were successful in their efforts and achieved their decontamination and availability goals. The use of KMPS has been considered for other applications where a modular system is of benefit. The use of different Kurion Ion Specific Media (ISM) allows for precision removal of hazardous ions from a variety of waste streams. This paper describes a range of projects in which KMPS modularity will benefit the nuclear industry. The KMPS is a flexible system that can be deployed quickly and efficiently. One treatment system option being considered is the utilization of a KMPS for the removal of contamination from groundwater. Similar to its use in Japan, the KMPS can be sent to remote locations for the direct treatment of ground water. In particular, there are several sites across the world where uranium contamination is a problem. Uranium treatability testing was performed on collected groundwater samples from a radiologically contaminated site. The tests were used to help provide designs for KMPS modifications. These designs included the addition of chemical treatment module for the control of pH and other chemistry, the addition of an ultraviolet module for the destruction of biological material, among others. The tests determined the loading capacities of uranium capture resins (UCR) and media (UCM) for this specific waste stream. Based on uranium loading capacities of 37.3 mg/L (end of test C/C{sub 0} = 0.5) and 54.6 mg/L (extrapolated to C/C{sub 0} = 0.98), and the longevity of the resin (45,911 BV for C/C{sub 0} = 0.5, and 80,556 BV for C/C{sub 0} = 0.98) UCR-1 was selected as the media/resin to be proposed for a uranium treatment system. (authors)

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NASA NTRS1976

International Conference on Environmental Sensing and Assessment, Las Vegas, Nev., September 14-19, 1975, Proceedings. Volumes 1 & 2

The papers deal with the detection of hazardous environmental pollutants, the development of emission control plans, and the design of compliance monitoring systems. Topics include remote sensing techniques in environmental pollution monitoring, monitoring of atmospheric particulate matter, air pollution due to sulfur dioxide and other inorganic compounds, marine pollution, atmospheric aerosols, industrial pollution, and legal aspects of pollution monitoring. Other papers examine the toxic effects of heavy metals and halogenated hydrocarbons, pollution associated with waste-disposal processes, pesticide residues in soil and groundwater, evaluations of groundwater quality, and monitoring of nuclear wastes. The interaction of climate and pollution is also discussed along with global pollutant transport, environmental modeling, ambient environmental air quality, aircraft and ground-vehicle emissions, and pollution associated with energy extraction and utilization processes. Individual items are announced in this issue.

Source record↗
DOE OSTI2021

Refinement of Salt Dissolution Inhibitor Requirements (Interim Report)

At Savannah River Site, High-Level Waste is stored in below-grade carbon steel tanks. This waste in part consists of sludge, salt cake, and/or supernate. Preparation of this waste for future processing involves dissolution of the salt cake layer. The salt dissolution process can create conditions that leave the carbon steel tanks susceptible to localized corrosion. The salt to be dissolved contains high concentrations of nitrate, that once released, create an environment that may be conducive to pitting corrosion or stress corrosion cracking (SCC) of carbon steel. The salt dissolution process also liberates interstitial liquid trapped between the salt crystals. This liquid is initially high in nitrite and hydroxide concentration. High pH and greater ratios of nitrite to nitrate act as inhibitors to minimize corrosion of carbon steel in high nitrate environments. However, as dissolution proceeds, the concentration of nitrate increases while the hydroxide and nitrite concentration of the interstitial liquid depletes and becomes insufficient to prevent the onset of corrosion attack. While tank blending and the addition of inhibitors are used to ensure adequate concentrations of hydroxide and nitrite during normal operation, this is not desirable during salt dissolution as it can reduce process efficiency and increase the amount of waste that needs processed.

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DOE OSTI2019

Panel Session 51: Idaho Transuranic Waste Challenges, Solutions and Implications for the Future (3/5)

This panel focused on the retrieval, characterization, treatment, certification, and shipping of transuranic (TRU) waste at the Idaho National Laboratory (INL) Site, Idaho, USA. The INL began receiving TRU waste from offsite generators (primarily the Rocky Flats Plant in Colorado, USA) beginning in 1954. Until 1970, the INL Site disposed of barrels, boxes, and other debris in a shallow, unlined landfill. Beginning in 1970, the INL Site began storing TRU waste on an asphalt pad, which was then covered with several feet of soil. This practice continued until the late 1980's at which time a large air support building was erected over the waste. In the mid-to-late 1990's, the air support building was removed and a massive building was constructed over the entire dirt mound. Contractors retrieved the waste from the early 2000's through February 2017 to support an Idaho Settlement Agreement milestone that required the retrieval of 65,000 cubic meters of TRU waste and removal from Idaho. This panel focused on the Idaho Transuranic Waste Challenges, Solutions and Implications for the Future. The Idaho Transuranic Waste Program includes buried and stored waste retrieval and processing of contact- and remote-handled transuranic waste. Over the duration of the program, the Idaho site has conducted targeted retrieval of buried waste from the 1950's and 1960's; designed, permitted, built, operated, and completed the mission of a facility to process 65,000 cubic meters of waste; and shipped more than 57,000 cubic meters of waste to WIPP. Presentation proposed: Transuranic Waste Challenges and Solutions (Hoss Brown, Fluor Idaho, LLC (USA), John McCoy, Fluor Idaho, LLC (USA), James Malmo, US DOE (USA), Doug Pruitt, Department of Energy Idaho Field Office (USA), Kenneth Princen, US DOE (USA))

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DOE OSTI2019

Thermodynamic assessment of the hollandite high‐level radioactive waste form

Abstract Hollandite has been studied as a candidate ceramic waste form for the disposal of high‐level radioactive waste due to its inherent leach resistance and ability to immobilize alkaline‐earth metals such as Cs and Ba at defined lattice sites in the crystallographic structure. The chemical and structural complexity of hollandite‐type phases developed for high‐level waste immobilization limits the systematic experimental research that is required to understand phase development due to the large number of potential additives and compositional ranges that must be evaluated. Modeling the equilibrium behavior of the complex hollandite‐forming oxide waste system would aid in the design and processing of hollandite waste forms by predicting their thermodynamic stability. Thus, a BaO–Cs 2 O–TiO 2 –Cr 2 O 3 –Al 2 O 3 –Fe 2 O 3 –FeO–Ga 2 O 3 thermodynamic database was developed in this work according to the CALPHAD methodology. The compound energy formalism was used to model solid solution phases such as hollandite while the two‐sublattice partially ionic liquid model characterized the oxide melt. Results of model optimizations are presented and discussed including a 1473 K isothermal BaO–Cs 2 O–TiO 2 pseudo‐ternary diagram that extrapolates phase equilibrium behavior to regions not experimentally explored.

Utlak, Stephen A.↗
DOE OSTI2016

Nuclear Fuel Traces Definition in Storage Ponds of Research VVR-2 and OR Reactors in NRC 'Kurchatov Institute' - 16022

The Gas Plant complex is the experimental base of the Institute of Nuclear Reactors, which is part of the Kurchatov Institute. In 1954 the commissioning of the first Soviet water-cooled water-moderated research reactor VVR-2 on enriched uranium, and until 1983 the complex operated two research water-cooled water-moderated reactors 3 MW (VVR-2) and 300 kW (OR) capacity, which were dismantled in connection with the overall upgrades of the complex. The complex has three storage ponds in the reactor building. They are sub-surface vessels filled with water (the volume of water in each is about 6 m{sup 3}). In 2007-2013 the spent nuclear fuel from storages was removed for processing to 'Mayak'. Survey of Storage Ponds by Underwater Collimated Spectrometric System shows a considerable layer of slime on the bottom of ponds and traces of spent nuclear fuel in one of the storage pond. For determination qualitative and the quantitative composition of radionuclide we made complex α-, β-, γ - spectrometric research of water and bottom slimes from Gas Plant complex storage ponds. We found the spent nuclear fuel in water and bottom slime in all storage ponds. Specific activity of radionuclides in the bottom slime exceeded specific activity of radionuclides in the ponds water and was closed to levels of high radioactive waste. Analysis of the obtained data and data from earlier investigation of reactor MR storage ponds showed distinctions of specific activity of uranium and plutonium radionuclides. (authors)

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DOE OSTI2013

Experiment on the treatment of waste extraction solvent from the molybdenum-99 process

In the Mo-99 (Molybdenum-99) isotope extraction test process for radiopharmaceutical applications, organic solvent is used to extract Mo-99 from an irradiated UO{sub 2} dissolution. The extraction solvent was stored when the test work was stopped. A total of about 120 liters of waste solvent was stored at INER (Institute of Nuclear Energy Research, Taiwan). The extraction solvent consisted of 5% di-(2-ethylhexyl)-phosphoric acid (D2EHPA) and kerosene. The radionuclides found in the waste solvent include Cs-137, Am-241, Tc-99, and Sr-90, which give off gross alpha and beta radioactivity of 1898 and 471 Bq/ml, respectively. This study aims to remove radionuclides from the waste solvent using sodium carbonate and sodium hydroxide solutions in different concentrations. After mixing the waste solvent with the alkaline solution followed by settling, a third phase other than organic and aqueous phase appeared which is expected due to the saponification reaction. The experimental results showed that increasing the number of washing and the alkaline solution concentration could enhance the radionuclides removal rate. An optimal removal method was proposed using 2 M Na{sub 2}CO{sub 3} solution twice followed by 1 M NaOH solution one time for the third phase generated early in the mixing stages. The remaining gross alpha and beta radioactivity of the treated organic solvent was 2 and 3 Bq/ml, respectively. The treated solvent could be stabilized by ashing at 500 deg. C and then immobilized. The alkaline solution would be neutralized by hydrochloric or nitric acid and then treated using a variety of adsorbents or bone char via adsorption to remove nuclides to meet the wastewater discharge limitation. (authors)

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