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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 883 records · Page 49
All sourcesNASA NTRSDOE OSTI
DOE OSTI2018

2017 NE-4 Nuclear Technologies Research and Development (NTRD) Achievements Report

Fiscal year (FY) 2017 initiated new and reestablished presidential and departmental nuclear energy priorities for the U.S. Department of Energy (DOE), Office of Nuclear Energy (DOE-NE) that will: • Support the Administration’s priorities to achieve U.S. energy dominance and economic prosperity through U.S. technology leadership • Support DOE-NE mission priorities and goals for the existing nuclear fleet, advanced reactor pipeline and national fuel cycle infrastructure • Focus on early stage research, and effective cooperation with industry, academia, and other government agencies • Propose investments to maintain and improve capabilities and infrastructure to conduct leading-edge science, research, and technology development. DOE-NE’s mission priorities will focus on the existing nuclear fleet, advanced reactor pipeline, and fuel cycle infrastructure. Therefore, the DOE Nuclear Technology Research and Development (NE-4) mission aligned to (1) support the research and development (R&D) of innovative reactor technologies that may offer improved safety, functionality, and affordability and (2) conduct research to reduce long-term technical barriers for advanced nuclear energy systems, including space fission systems. Key activities to execute this mission include: • To achieve the NE-4 mission priorities, two major programs with eight R&D campaigns are engaged in impactful research, development, demonstration, and deployment: Fuel Cycle Technologies (FCT) Program • Fuel Cycle Options Campaign evaluates complex fuel cycle options and provides performance metrics on various systems and scenarios to improve the knowledge of interdependencies between technologies and different systems. • Advanced Fuels Campaign performs science-based R&D on accident tolerant fuels in addition to advanced light-water reactor (LWR) fuel and transmutation fuel. • Joint Fuel Cycle Studies (JFCS) is a collaboration with the Republic of Korea to assess the feasibility and nonproliferation of electrochemical recycling to manage used fuel. • Material Recovery and Waste Form Development (MRWFD) Campaign researches advanced fuel cycle material recovery and waste management capabilities to improve fuel cycle performance with less processing and waste generation. • Material Protection, Control, and Accountability Technologies (MPACT) Program develops tools and capabilities to secure the next generation nuclear materials management and safeguards for nuclear fuel cycles. Advanced Reactor Technologies (ART) Program • Fast Reactors Campaign targets benefits in either resource utilization or waste management and possesses favorable features for small reactor applications and plutonium management. • Gas-Cooled Reactors Campaign seeks to raise the technological readiness of high temperature gas-cooled reactors (HTGRs) and other reactor concepts utilizing coated particle fuel, graphite, or helium coolant by performing high risk, enabling research and development that commercial developers are unable or unwilling to undertake in the current energy market. • ART Energy Conversion (EC) Project provides solutions to convert heat from an advanced reactor to useful products that support commercial application of the reactor designs.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS1986

Controlled Ecological Life Support Systems (CELSS) physiochemical waste management systems evaluation

Parametric data for six waste management subsystems considered for use on the Space Station are compared, i.e.: (1) dry incineration; (2) wet oxidation; (3) supercritical water oxidation; (4) vapor compression distillation; (5) thermoelectric integrated membrane evaporation system; and (6) vapor phase catalytic ammonia removal. The parameters selected for comparison are on-orbit weight and volume, resupply and return to Earth logistics, power consumption, and heat rejection. Trades studies are performed on subsystem parameters derived from the most recent literature. The Boeing Engineering Trade Study (BETS), an environmental control and life support system (ECLSS) trade study computer program developed by Boeing Aerospace Company, is used to properly size the subsystems under study. The six waste treatment subsystems modeled in this program are sized to process the wastes for a 90-day Space Station mission with an 8-person crew, and an emergency supply period of 28 days. The resulting subsystem parameters are compared not only on an individual subsystem level but also as part of an integrated ECLSS.

Oleson, M.↗
DOE OSTI2017

Fabrication and Properties of Advanced Ceramic Technetium Waste Forms - 17311

Technetium-99 (Tc, t{sub 1/2} = 2.13 x 10{sup 5} years) is a challenge from a nuclear waste perspective and is one of the most abundant, long-lived radioisotopes found in used nuclear fuel (UNF). Within the Hanford Tank Waste Treatment and Immobilization Plant, technetium volatilises at typical glass melting temperature, is captured in the off-gas treatment system and recycled back into the feed to eventually increase Tc-loadings of the glass. In this research we are providing a technical solution for continuously depleting the Tc-content of the glass melter feed by precipitating Tc from the off-gas effluents and effectively immobilizing {sup 99}Tc as durable ceramic waste form. We have developed a wet-chemical coprecipitation-based synthesis route to fabricate gram-size Tc-based pyrochlores with the model composition Sm{sub 2}(Tc{sub 0.5-x}Ru{sub x}Ti{sub 0.5}){sub 2}O{sub 7}. These ceramic Tc waste forms exhibit better technetium retention and corrosion resistance in ASTM C1220-10 based testing compared with technetium-containing LAWE4-type borosilicate glass, combined with optimal waste loading. (authors)

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

Evaluating the Impacts of Reactive Gas (Carbon Dioxide and Oxygen) Ingress on Leaching from Cementitious Waste Forms - 17482

Cementitious materials are currently being used as the waste forms for low-activity waste disposal (Saltstone) at the Savannah River Site. In addition, cementitious materials have been selected as the waste form (Cast Stone) for secondary wastes and are being investigated for use for supplemental low-activity waste treatment (Cast Stone) at the Hanford Site. These materials are subjected to numerous aging and degradation phenomena in the disposal environment that potentially impact their long-term durability and performance. One important phenomenon is the ingress of reactive gases (i.e., carbon dioxide and oxygen) that can react with and change the chemistry and pore structure and thus mass transport properties of the waste forms. The reaction of carbon dioxide (i.e., carbonation) reduces the pH of the waste form pore solution impacting the solubility of constituents of potential concern (COPCs). In Cast Stone and Saltstone, the reaction of oxygen (i.e., oxidation) impacts the solubility of redox-sensitive COPCs, notably the technetium-99 (Tc-99) and chromium. The objective of this research is to experimentally characterize the carbonation and oxidation impacts on chemistry and leaching behavior in cementitious waste forms using experimental techniques including United States Environmental Protection Agency (U.S. EPA) Method 1313 and U.S. EPA Method 1315. Results of experimental characterization will be used to parameterize reactive mass transport models to project leaching under a range of field scenarios in future applications. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

Solidification of Low and Intermediate Level Waste - A Comparison between Different Methods - 19584

The treatment of radioactive materials results in products that have to be suitable for a period of interim storage as well as for disposal in a repository. Considering available options for treatment of radioactive waste, future disposal also has to be taken into account. Up to now, only cementation and encapsulation in geopolymer (only to limited extent) are relevant treatment methods for the solidification of German waste for disposal in the repository Grube Konrad. However, a synopsis of other solidification technologies is considered to be helpful in order to be prepared for future developments in the fields of radioactive waste treatment. Treatment methods addressed in this paper are: - Cementation; - Encapsulation into geopolymer matrix; - Vitrification; - SYNROC; - Geomelting. The criteria for evaluation of the treatment methods are as follows: - Requirements concerning waste properties; - Safety of treatment facilities; - Fulfillment of waste acceptance criteria for repository Grube Konrad; - Fulfillment of requirements for near surface disposal; - Suitability for long-term interim storage. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2024

Application of Process Chemical Modeling to Optimize Radioactive Waste Disposal at the Savannah River Site – 24242

The Liquid Waste Program (LWP) managed by Savannah River Mission Completion (SRMC) is responsible for the treatment and disposal of waste at the Savannah River Site (SRS). Radioactive waste at SRS is stored and processed at four key facilities – each with their respective functions to store, blend, grout, or vitrify waste. The tank farm, where waste is stored, consists primarily of legacy waste with new material incoming from the Accelerated Basin De-inventory program (ABD), which is managed by Savannah River Nuclear Solutions (SRNS). System planning is done by SRNS and SRMC to optimize ABD and LWP operations, respectively.

Georgiou, Andreas↗
DOE OSTI2025

Technoeconomic Analysis of a Microwave-Assisted Novel Process That Converts Polypropylene Plastic Waste to Propylene

Propylene is an important petrochemical that is used in various industries, including the automobile and polymer sectors. Conventionally, propylene has been produced through fluid catalytic cracking, steam cracking, and propane dehydrogenation. In this project, a novel process to produce propylene from polypropylene plastic waste using a microwave reactor is introduced. Propylene production is simulated in Aspen Plus for both the conventional propane dehydrogenation process and the novel design, with heat integration applied to minimize utility demands and the conventional process serving as the base case for comparison. Finally, a tailored technoeconomic analysis is carried out for both simulated cases to estimate important economic indicators. Analysis of the results shows that the novel microwave-assisted process outperforms the conventional route for propylene production, achieving a single-pass conversion of almost 100%, compared with only 33.8% in the conventional plant. The Net Present Value of the novel plant is $\$1720.3$ MM, which is 3.5 times higher than the conventional process, while the Levelized Cost of Propylene is reduced by 40%. Capital and operating expenditure are also improved in the proposed scheme, with reductions of approximately $\$29$ MM and $\$222$ MM/yr, respectively. In addition to cost savings, this novel process also provides a convenient means to recycle waste polypropylene.

42 ENGINEERING↗
DOE OSTI2018

Design and Licensing of Irradiated Uranyl Nitrate Liquid Containers and Applicable Loading and Packaging System - 18481

Irradiated highly enriched uranyl nitrate liquid (HEUNL) solutions are typically residual materials resulting from the processing of uranium target materials in the production of radioisotopes. Traditionally, these residual materials have been processed and stored at the site of origin, due to the lack of an option for packaging, transportation and disposition. Modern alternatives to process, stabilize, package and dispose of these materials have been developed, but until recently there has not been a viable Type B package solution that could be used to ship these materials in sufficient quantities to justify the cost of an off-site disposition strategy. For this reason, most facilities that generate irradiated HEUNL materials have implemented on-site storage tanks or containers, resulting in the significant accumulation of HEUNL inventories worldwide. This paper provides an overview of the design and licensing activities for the development of sealed irradiated HEUNL transport containers to permit safe transportation of this material in the NAC Legal Weight Truck (NAC-LWT) cask. A description of the robust HEUNL container, licensing and fabrication is also presented. In addition, NAC has developed a specialized loading system to optimize the process to load, package and un-package HEUNL materials to minimize worker exposure during loading, and execute the packaging process efficiently for safe transport and retrieval. (authors)

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DOE OSTI2018

A Review of Flow Analysis Methods for Determination of Radionuclides in Nuclear Wastes and Nuclear Reactor Coolants

The safety and security nuclear power plant operations depends on the application of the most appropriate techniques and methods of chemical analysis, where modern flow analysis methods find increasing applications. Nevetheless, the current status of development of these methods is more limited than might be expected from their authentic advantages. The main aim of this contribution is to review automated flow analysis procedures developed with various detection methods for application in the nuclear energy industry. The flow analysis methods for determination of radionuclides that have been reported to date have been primarily focused mainly on environmental applications. The needs and benefits of the application of flow methods for the monitoring of nuclear wastes and for process analysis of the primary circuit coolants of light water nuclear reactors are discussed. Application of continuous flow methods (CFA) or injection methods (FIA, SIA) of flow analysis with radiometric detection shorterns the time of analysis and improves precision of determinations due to mechanization of time-consuming operations of sample processing. The use of mass spectrometry (MS) detection enables multicomponent analyses and determination of transuranic isotopes with much better limits of detection compared to radiometric detection.

radiochemical analysis, Sequential Injection and F↗
DOE OSTI2018

Tank 40 Sludge Batch 9 Waste Acceptance Product Specifications (WAPS) Solids and Chemical Analyses (2018)

A 3 L sample of sludge was collected from High Level Waste (HLW) Tank 40 (sample ID: HTF-40-18-9) at the Savannah River Site (SRS) and subsequently analyzed by the Savannah River National Laboratory (SRNL) in accordance with requirements for reporting the Waste Acceptance Product specification (WAPS) [1]. The collected sample was representative of Sludge Batch (SB) 9 that will be processed by the Defense Waste Processing Facility (DWPF) into a vitrified waste form.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2025

Advanced Computational Modeling of High-Level Waste Vitrification at the Hanford Site

The U.S. Department of Energy (DOE) has selected vitrification for stabilizing legacy tank waste at the Hanford site, where radioactive waste from plutonium production was historically stored in underground tanks. This waste will be separated into low-activity waste (LAW) and high-level waste (HLW) fractions and processed at the Waste Treatment and Immobilization Plant (WTP). At WTP, glass melters are used for the vitrification of radioactive tank waste, transforming it into a stable borosilicate glass form for safe long-term storage. The melter vessel is constructed from highly durable and heat-resistant materials, where the vitrification process occurs. The main regions that are modeled are the melt pool, plenum, cold cap, riser/discharge chamber, and surrounding structure with insulation layers. Forced convection induced by air bubblers at the base of the melter ensure uniform temperature distribution and provide heat to the cold cap layer. The cold cap is a region of reacting batch feed that floats on top of the molten glass and is where the batch-to-glass reactions occur. Joule heating provided by electrodes mounted along the vertical walls of the melter and immersed directly in the glass, generates the necessary heat for the net endothermic conversion processes that occur in the cold cap. The high temperatures, radioactivity, and opaque nature of the glass prevent direct observation inside the melters. Therefore, computational models are essential for providing insight into factors that affect melter throughput. Thermocouples in the plenum provide operators with plenum temperature measurements. Operational adjustments include bubbling rate, voltage supplied to the electrodes, feed adjustments, and glass removal rate. Different computational fluid dynamics (CFD) models have been developed, each serving a specific purpose. There are CFD models of different scale melters, as well as models that capture the two-phase flow interfaces of rising bubbles in the molten glass or models with a simplified molten glass region so that the surrounding structure and plenum can be feasibly incorporated. Pilot-scale melter models have been developed to serve as validation of the methods employed in the simulation of the full-scale WTP melters. Models incorporating resolved bubbling are used to develop momentum source terms to implement into a single phase, multi-region, steady-state flow model that is being validated by measured process parameters such as glass production rate, voltage, input power, plenum temperatures, etc. The resolved bubbling model uses the multiphase volume of fluid approach to model the system with a high-resolution interface capturing scheme to maintain sharp interfaces between the molten glass and the air phase. The suite of CFD models is continually being improved to incorporate more realistic physics and achieve faster turnaround time. For example, an incremental controller is implemented to automatically adjust electrode voltage within the simulation to a molten glass set point temperature of 1150°C. Newer models feature improved meshes to ensure conformal meshes between regions and eliminate unnecessary mesh refinement in areas that are not of interest (such as boundary layers in offgas ports). Instead of explicitly modeling the structural, refractory, and insulation layers of the melter, a thermal resistance approach is used with published correlations used for boundary conditions. The development of robust and efficient CFD models will be instrumental in enabling the WTP to successfully fulfill its mission of safely stabilizing legacy nuclear waste.

12 - MGMT OF RADIOACTIVE AND NON-RADIOACTIVE WASTE↗
DOE OSTI2025

Advanced Processing of Coal and Coal Waste to Produce Graphite for Fast-Charging Lithium-Ion Battery Anode

The University of North Dakota (UND) Energy & Environmental Research Center (EERC), in collaboration with the UND Center for Process Engineering Research (CPER), conducted a project to validate two technologies capable of converting North Dakota lignite and lignite coal waste to high-quality graphite for fast-charging lithium-ion battery (LIB) anode. The project was conducted over about 3 years from April 7, 2022, to July 6, 2025. The two technological paths pursued in this project include path A – direct conversion of coal or coal waste to graphite by the upgraded carbon ores to products (UCOP) process being developed at the EERC and path B – lignite-derived coal tar pitch (CTP) conversion to graphite (CTP2G) process being developed at CPER. The results from this project validate the two technological approaches and are expected to be an integral part of a portfolio of emerging technologies for making high-quality graphite not only from North Dakota lignite, but from all ranks of U.S. domestic coal and coal waste resources. The quality of the graphite produced by these technologies is high enough for various applications, including batteries for the fast-growing electric vehicle industry, energy storage applications, electric arc furnace electrodes for steel production, and graphene production, among others. Although the two technologies can produce high-quality graphite, they are fundamentally different in that the UCOP technology provides a direct path to transform coal to graphite, while the CTP2G technology needs to go through a CTP intermediate and a coking process for the intermediate, which requires a special facility to accomplish. For application in the industry, the UCOP process is designed to be more flexible, with feedstock to include potentially any carbonaceous material such as all coal ranks and biochar, while the CTP2G process is designed to utilize CTP as the starting precursor. The key project accomplishments include the following: • Successful preparation of high-quality synthetic graphite from North Dakota lignite coal/coal wastes and lignite-derived CTP. • Patent application has been filed for the UCOP process and an internal invention disclosure has been filed for the CTP2G process. • The produced graphite performs better than a commercial battery-grade sample in LIB coin cells, especially fast-charging capability, stability, and long-duration cycling. • Coin-type Li-ion half-cells with CTP2G graphite showed excellent performance, with >370 mAh/g capacity, >90% initial coulombic efficiency, and 93%/67% retention at 1C/2C rate, which outperforms commercial graphite in charging speed, stability, and cycling. • Results of fabricated 18650 cells were consistent with the observations in coin cells. • Preliminary techno-economic analysis (TEA) estimates for the UCOP technology indicate a manufacturing cost of about $\$$39/kg based on 50-metric ton/year capacity. • Preliminary TEA estimates for the CTP2G technology indicate a market price of about $\$$7107/ton ($\$$7/kg) based on 22,000-ton/year production capacity.

01 COAL, LIGNITE, AND PEAT↗
NASA NTRS1987

USBI Booster Production Company's Hazardous Waste Management Program at the Kennedy Space Center, FL

In response to the hazardous-waste generating processes associated with the launch of the Space Shuttle, a hazardous waste management plan has been developed. It includes waste recycling, product substitution, waste treatment, and waste minimization at the source. Waste material resulting from the preparation of the nonmotor segments of the solid rocket boosters include waste paints (primer, topcoats), waste solvents (methylene chloride, freon, acetone, toluene), waste inorganic compounds (aluminum anodizing compound, fixer), and others. Ways in which these materials are contended with at the Kennedy Space Center are discussed.

Venuto, Charles↗
NASA NTRS2011

Microbial Characterization of Solid-Wastes Treated with Heat Melt Compaction Technology

The research purpose of the project was to determine the fate of microorganisms in space-generated solid wastes after processing by a Heat Melt Compactor (HMC), which is a candidate solid waste treatment technology. Five HMC product disks were generated at Ames Research Center (ARC), Waste Management Systems element. The feed for two was simulated space-generated trash and feed for three was Volume F compartment wet waste returned on STS 130. Conventional microbiological methods were used to detect and enumerate microorganisms in HMC disks and in surface swab samples of HMC hardware before and after operation. Also, biological indicator test strips were added to the STS trash prior to compaction to test if HMC processing conditions, 150 C for approx 3 hr and dehydration, were sufficient to eliminate the test bacteria on the strips. During sample acquisition at KSC, the HMC disk surfaces were sanitized with 70% alcohol to prevent contamination of disk interiors. Results from microbiological assays indicated that numbers of microbes were greatly reduced but not eliminated by the 70% alcohol. Ten 1.25 cm diameter cores were aseptically cut from each disk to sample the disk interior. The core material was run through the microbial characterization analyses after dispersal in sterile diluent. Low counts of viable bacteria (5 to 50 per core) were found but total direct counts were 6 to 8 orders of magnitude greater. These results indicate that the HMC operating conditions might not be sufficient for complete waste sterilization, but the vast majority of microbes present in the wastes were dead or non-cultivable after HMC treatment. The results obtained from analyses of the commercial spore test strips that had been added fo the wastes prior to HMC operation further indicated that the HMC was sterilizing the wastes. Nearly all strips were recovered from the HMC disks and all of these were negative for spore growth when run through the manufacturer's protocol. The 10(exp 6) or so spores impregnated into the strips were no longer viable. Control test strips, i.e., not exposed to the HMC conditions, were all strongly positive. All isolates from the cultivable counts were identified, leading to one concern: several were identified as Staphylococcus aureus, a human pathogen. The project reported here provides microbial characterization support to the Waste Management Systems element of the Life Support and Habitation Systems program.

Strayer, Richard F.↗
DOE OSTI2018

Evaluation of Immobilizing Secondary Waste from a Proposed Treatment Process for Hanford WTP LAW Melter Condensate

The off-gas system in the Low Activity Waste (LAW) vitrification facility at the Hanford Waste Treatment and Immobilization Plant (WTP) will generate an aqueous condensate recycle stream. Two unit operations generate this condensate, a Submerged Bed Scrubber and a Wet Electrostatic Precipitator (LAW SBSWESP condensate). The expected disposition of this stream during baseline operations is to send it to the WTP Pretreatment Facility, where it will be blended with LAW, concentrated by evaporation and recycled to the LAW vitrification facility again. The primary reason to recycle this stream is to immobilize the semivolatile 99 Tc isotope into the glass. During the Direct Feed Low Activity Waste (DFLAW) portion of the mission, the condensate is sent to the Effluent Management Facility (EMF) where it is evaporated then recycled to the LAW facility where the evaporated condensate is combined with the LAW feed from the pretreatment system.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2022

Process-microstructure relationship of laser processed thermoelectric material Bi2Te3

Additive manufacturing allows fabrication of custom-shaped thermoelectric materials while minimizing waste, reducing processing steps, and maximizing integration compared to conventional methods. Establishing the process-structure-property relationship of laser additive manufactured thermoelectric materials facilitates enhanced process control and thermoelectric performance. This research focuses on laser processing of bismuth telluride (Bi 2 Te 3 ), a well-established thermoelectric material for low temperature applications. Single melt tracks under various parameters (laser power, scan speed and number of scans) were processed on Bi 2 Te 3 powder compacts. A detailed analysis of the transition in the melting mode, grain growth, balling formation, and elemental composition is provided. Rapid melting and solidification of Bi 2 Te 3 resulted in fine-grained microstructure with preferential grain growth along the direction of the temperature gradient. Experimental results were corroborated with simulations for melt pool dimensions as well as grain morphology transitions resulting from the relationship between temperature gradient and solidification rate. Samples processed at 25 W, 350 mm/s with 5 scans resulted in minimized balling and porosity, along with columnar grains having a high density of dislocations.

Oztan, Cagri↗
DOE OSTI2018

A Composition-Based Approach for Predicting Hanford Tank Waste Liquid Viscosity

Liquid viscosity is an important Hanford waste physical property that governs, in part, the performance of key Hanford flowsheet unit operations and processes, including waste filtration, ion exchange, pumping and transport, and mixing. Evaluation and optimization of these processes rely on liquid viscosity correlations subject to substantial error (at least ±60%) that can lead to overly conservative, and sometimes physically unachievable, limits for validation and operation of these unit operations. The current liquid viscosity correlations used for the Hanford flowsheet attempt to capture the effects of specific waste chemistry (namely, the speciation of various salts) on liquid viscosity using liquid density alone. It is postulated that the ±60% or greater error associated with these “density-based” liquid viscosity predictions is driven in part by the inability of waste density to capture the full impact of waste chemistry on liquid viscosity, and that improved prediction accuracy could be achieved by implementing a composition-based approach to evaluating liquid viscosity for Hanford wastes. Recent publications, namely those of Lencka et al. (1998), Hu (2004), and Laliberté (2007a), provide correlations and comprehensive, consolidated databases of single-salt solution viscosity that enable composition-based predictions for Hanford-type liquids (i.e., concentrated multi-component aqueous electrolyte solutions). In the current study, the viscosities of Hanford wastes and waste simulants were predicted using a composition-based model developed by Laliberté (2007a) (hereafter referred to as the “Laliberté viscosity model”) to evaluate its accuracy relative to the current density-based models used to predict waste viscosity for engineering applications at the Hanford Site. A viscosity database that captured the composition and viscosity of wastes and waste simulants was developed to support this evaluation. To facilitate application of the Laliberté viscosity model, waste and waste simulant compositions were converted to an equivalent salt recipe composed of salts with known viscosity coefficients using a number of simplifying assumptions that have the potential to impact the fidelity of the model prediction. Despite these simplifications, the Laliberté viscosity model was able to predict viscosity more accurately than the density-based correlations currently available for Hanford flowsheet use, such as that proposed in RPP-RPT- 51652. The use of composition-based viscosity models, like that proposed by Laliberté (2007a), shows promise for leading to improved prediction of liquid waste viscosity. As implemented in this study, application of the Laliberté viscosity model could lead to a decrease in the uncertainty associated with liquid viscosity prediction by as much as a factor of two, allowing for reduction in unit operation design conservatism and for the more realistic selection of test simulants for flowsheet process evaluation and optimization, all of which could potentially reduce flowsheet design margins, lower costs and reduce duration of the Hanford mission, and reduce Hanford flowsheet risks.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

Uranium Ion Exchange Resin Performance - A Case Study at 200 West Pump and Treat - 18103

The 200 West Pump and Treat Facility (200W P and T), located on the U.S. Department of Energy's Hanford site, is a 2,500 gallon per minute (GPM) plant designed to capture and treat an eight square kilometer groundwater plume. The groundwater contamination is largely a result of process liquids disposed from operations involving the separation of nuclear materials and storage of spent nuclear fuel. This document will report on the performance of a full scale uranium ion exchange system used to treat contaminated groundwater and other waste streams. The uranium ion exchange system at 200W P and T was designed as a three vessel operation with a lead, lag and polish vessel, which contain 300 cubic feet of resin in each vessel. The ion exchange system treats a portion of the total plant flow with a maximum capacity of 400 GPM. Prior to start-up, one key issue was the potential to treat feed streams with a wide range of uranium concentrations (22 to 150,000 microgram per liter [μg/L]). The resin change-out frequency was a primary concern. A literature review was performed, resulting in resin capacity and change out frequency that varied by two orders of magnitude (1.2 milligram per gram [mg/g] to 104 mg/g capacity). Since startup in September 2015, the ion exchange system has experienced a dynamic feed stream as wells were brought online, which changed the contaminant concentrations. The plant also treats leachate liquid waste from on-site low-level waste site, known as the Environmental Restoration Disposal Facility (ERDF), which contains uranium as well as other competing anions, such as sulfate. The uranium breakthrough is closely monitored throughout the resin beds with weekly sampling events at the influent, effluent, and intermediate ports in the first vessel. In the influent to the ion exchange system, the uranium concentration fluctuated from 24 μg/L up to 910 μg/L. The information on influent and effluent was used to determine the change out frequency and the capacity of the resin for uranium under these operating conditions. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗