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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 1,135 records · Page 63
All sourcesNASA NTRSDOE OSTI
DOE OSTI2017

Proton transfer in microbial electrolysis cells

Proton transfer and electron transfer are of prime importance in the development of microbial electrochemical cells. While electron transfer is primarily controlled by biology, proton transfer is controlled by process engineering and cell design. To develop commercially feasible technologies around the concept of a bioelectrochemical cell, real feedstocks have to be explored and associated limitations have to be identified. Here in this study, the proton transfer rate was quantified for a microbial electrolysis cell (MEC) and its dependence on process parameters was investigated using a proton balance model. The reaction system consisted of a biomass-derived pyrolytic aqueous stream as a substrate producing hydrogen in a flow-through MEC. The proton transfer rate increased with anode flow rate and organic loading rate up to a maximum of 0.36 ± 0.01 moles per m 2 per h, equivalent to a hydrogen production rate of 9.08 L per L per day. Higher rates of hydrogen production, reaching 11.7 ± 0.2 L per L per day were achieved, when additional protons were provided via the cathode buffer. Electrochemical impedance spectroscopy shows that proton transfer was the dominant resistance in the production of hydrogen. The quantification of proton transfer rates for MECs with potential for biorefinery application and the demonstration of high hydrogen production rates approaching those required for commercial consideration indicate the strong potential of this technology for renewable hydrogen production. Understanding the transport phenomenon in bioelectrochemical cells is of great significance since these systems have potential for wide-ranging applications including energy production, bioremediation, chemical and nanomaterial synthesis, electro-fermentation, energy storage, desalination, and produced water treatment. Electron transfer in anode biofilms has been investigated extensively, but proton transfer studies are also important, since many cathodic half reactions require protons as the reactant. Determination of transport rates via proton balance was investigated in microbial electrolysis cells, which can be applied to other forms of microbial electrochemical systems. Lastly, these systems have a unique niche in the development of future biorefineries as a means of recovering energy from waste streams with potential for water recycle, making them an integral part of the water–energy nexus focus area.

59 BASIC BIOLOGICAL SCIENCES↗
NASA NTRS1987

Preliminary conceptual design and weight of a one-megawatt space-based laser power station utilizing a solar-pumped iodine lasant

A preliminary conceptual design of a space-based solar-pumped iodide laser emitting 1 MW of laser power for space-to-space power transmission is described. A near-parabolic solar collector focuses sunlight onto the t-C4F9I lasant within a transverse-flow optical cavity. Using waste heat, a thermal system supplies compressor and auxiliary power. The major system components are designed with weight estimates assigned. In particular, it is found that laser efficiency is not a dominant weight factor, the dominant factor being the laser transmission optics. The station mass is 92,000 kg, requiring approximately eight Shuttle flights to LEO, where an orbital transfer vehicle can transport it to the final altitude of 6378 km.

De Young, R. J.↗
NASA NTRS2011

Temporary Urine and Brine Stowage System (TUBSS) Development

International Space Station (ISS) crew liquid human waste is treated with chromic and sulfuric acids to maintain stability prior to processing to recover water. This pre-treated urine (PTU) and its processed by-product, brine, are highly toxic fluids that require special containment for on-orbit stowage. The temporary urine and brine stowage syste m (TUBSS) is an assembly used to store and transfer pre-treated urine (PTU) and brine for processing or disposal at a later date. This paper describes the development of the TUBSS, including design for two-fault tolerance and materials selection to maintain a soft, collapsible container. In addition, this paper will provide results of testing as well as lessons learned from the project, culminating in the successful launch of the hardware.

Dries, Kevin↗
DOE OSTI2019

Integrated Methodology for the Impact Assessment of an Uranium Mill Contaminated Site - 19031

This work describes the development of mathematical models for radionuclides distribution and transport in the environment. The applicability of the models involves an integrated approach to assess the quantitative dispersion of the radionuclides in different environmental compartments when released from uranium tailings or from other low-activity mining wastes storages. For every compartment the important pathways of contamination and the radionuclides of major concern are identified and selected, based on their chemical, physical, radiological and toxic properties. Each compartment represents an environmental media: air, soil, water, vegetation and biota. Depending on the characteristics of the sources different types of possible releases are evaluated using phenomenological models. Models of environmental release, dispersion, transport and fate within each environmental compartment, as well as models of transfer between them, are applied to a specific site. The model outputs are the radionuclides activities for each one of the environmental compartments, where relevant exposure end-points are previously identified. The methodology was applied to a specific Uranium Mill Site, referred here as the 'Mill'. The site comprises approximately 3.2 km and includes two inactive mills, former ore stockpiles areas, a partially reclaimed tailings pond disposal area and a current tailings pond disposal area. The Mill site began operations in the late 1950's extracting uranium ore by alkaline leaching. Later on, a new facility using acid leaching for extracting uranium begun to operate in 1980. The mill operated first continuously and after intermittently until its closure, in March 2006. The data required for the model input was acquired from different sources: direct measurements from this particular site and available from multiple site investigation reports (in particular, environmental data for groundwater, soil, surface water, sediments and biota); other, were selected from published data. For the unknown data, assumptions were made. (authors)

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

X-Ray Fluorescence Quantitative and Qualitative Analysis Using Micro Glass Beads

The fused glass bead method is the accepted 'best practice' method of preparation of samples for X-ray fluorescence (XRF) analysis. It minimizes the effects of sample heterogeneity and simplifies calibration by dissolving the sample into a uniform glass matrix. Glass fusion XRF samples are typically prepared by grinding sample powder with Lithium Tetraborate / Lithium Bromide and fusing the mixture at 1000 deg. C. In most cases sample-to-flux ratios of between 1:1 to 1:10 are used (1:1 for determination of trace elements and 1:10 for major elements). Typical sample mass consumed is in the range of 1000 to 2500 mg. It would be advantageous to reduce this sample mass in cases where the sample material is rare, valuable or highly radioactive. This study examines the effects of using lower sample-to-flux ratios and smaller fusion bead size on the detection and quantification of major and trace elements by XRF analysis. We have employed a micro fused glass bead process, using sample-to-flux ratios of 1:10, 1:15 and 1:25, to determine the effects on measurement precision, accuracy and detection limit. Objectives: - Created micro fused glass beads of a series of standards and andesite sample with sample-to-flux ratios of 1:10, 1:15 and 1:25. - Perform an qualitative analysis of the different sample-to-flux ratio micro beads and compare the results with the 1:5.3 beads routinely used at SRNL. - Quantitate the major components in the different sample-to-flux ratio micro beads using a wavelength dispersive x-ray fluorescence instrument. Advantages: - Reduce sample amount needed. - Speed up waste tank characterization. - Reduce personnel exposure. - Potential sealed source. Fused micro glass beads with sample-to-flux ratios of 1:10, 1:15 and 1:25 were created using sample amounts of 123 mg, 82 mg and 52 mg respectively. It was possible to detect most of the major oxides and some trace elements in the fused micro glass beads with different sample-to-flux ratios. The quantitative analysis showed good linear correlation between the intensity of the analysis and the concentration of the elements. Good precision and accuracy results were obtained for the majority of the major oxides in the three fused micro glass beads with different sample-to-flux ratios. Path Forward: Enhance the precision and accuracy of the method by optimizing different sample preparation steps, such as weighing and sample transfer. Continue to increase the sample-to-flux ratio to determine the minimum amount of sample needed to perform a multielement analysis. Consider the use of synthetic standards to lower the detection limit of the analysis. Used the micro bead sample preparation method to analyze radioactive samples.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI1988

Nuclear Safety [Vol. 29, No. 1, January-March 1988]

Nuclear Safety is a review journal that covers significant developments in the field of nuclear safety. Its scope includes the analysis and control of hazards associated with nuclear energy, operations involving fissionable materials, and the products of nuclear fission and their effects on the environment. Primary emphasis is on safety in reactor design, construction, and operation; however, the safety aspects of the entire fuel cycle, including fuel fabrication, spent-fuel processing, nuclear waste disposal, handling of radioisotopes, and environmental effects of these operations, are also treated. Table of Contents for this issue follows. THE CHERNOBYL ACCIDENT: 1 Core History and Nuclide Inventory of the Chernobyl Core at the Time of Accident by Gerald Kirchner and Cornelius C. Noack; GENERAL SAFETY CONSIDERATIONS: 6 Nonprescriptive Nuclear Safety Regulation: The Example of Loss of Offsite Power by M. W. Golay, V. P. Manno, and C. Vlahoplus, Jr., 20 Nuclear Power Safety Goals in Light of the Chernobyl Accident by C. Whipple and C. Starr; ACCIDENT ANALYSIS: 29 Containment Loads from Severe Accidents—U.S. Program by W. Kerr and M. K. Dey; PLANT SAFETY FEATURES: 36 Safety Characteristics of Modern High-Temperature Reactors: Focus on German Designs by W. Kröger, H. Nickel, and R. Schulten; ENVIRONMENTAL EFFECTS: 49 Radiation Hormesis and Nuclear Safety by C. C. Congdon; OPERATING EXPERIENCES: 58 Evaluation of Nonradiological Water Chemistry at Power Reactors by Harvey Zibulsky, James J. Kottan, Walter J. Pasciak, Mary Ann Castrogivanni, and Sujit Banerjee, 64 Reactor Shutdown Experience Compiled by J. W. Cletcher, 67 Operating U.S. Power Reactors Compiled by E. G. Silver; RECENT DEVELOPMENTS: 85 General Administrative Activities Compiled by E. G. Silver, 94 Reports, Standards, and Safety Guides by D. S. Queener, 100 Status of Power-Reactor Projects Undergoing Licensing Review Compiled by E. G. Silver, 103 Proposed Rule Changes as of Sept. 30, 1987; ANNOUNCEMENTS: 5 Northwestern University Short Course on Radiation Safety, 28 International Approach to Nuclear Safety (After Three-Mile Island and Chernobyl), 57 RPI Short Course on Modern Developments in Boiling Heat Transfer and Two-Phase Flow, 107 20th DOE/NRC Nuclear Air Cleaning Conference, 110 Third International Topical Meeting on Nuclear Power Plant Thermal Hydraulics and Operations, 108 The Authors, 111 Indexes to Nuclear Safety, Volume 28.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Integration of the Back End of the Nuclear Fuel Cycle

Management of spent nuclear fuel and high-level radioactive waste consists of three main phases – storage, transportation, and disposal – commonly referred to as the back end of the nuclear fuel cycle. Current practice for commercial spent nuclear fuel management in the United States (US) includes temporary storage of spent fuel in both pools and dry storage systems at operating or shutdown nuclear power plants. Storage pools are filling to their operational capacity, and management of the approximately 2,200 metric tons of spent fuel newly discharged each year requires transferring older and cooler spent fuel from pools into dry storage. Unless a repository becomes available that can accept spent fuel for permanent disposal, projections indicate that the US will have approximately 136,000 metric tons of spent fuel in dry storage systems by mid-century, when the last plants in the current reactor fleet are decommissioned. Current designs for dry storage systems rely on large multi-assembly canisters, the most common of which are so-called “dual-purpose canisters” (DPCs). DPCs are certified for both storage and transportation, but are not designed or licensed for permanent disposal. The large capacity (greater number of spent fuel assemblies) of these canisters can lead to higher canister temperatures, which can delay transportation and/or complicate disposal. This current management practice, in which the utilities continue loading an ever-increasing inventory of larger DPCs, does not emphasize integration among storage, transportation, and disposal. This lack of integration does not cause safety issues, but it does lead to a suboptimal system that increases costs, complicates storage and transportation operations, and limits options for permanent disposal. This paper describes strategies for improving integration of management practices in the US across the entire back end of the nuclear fuel cycle. The complex interactions between storage, transportation, and disposal make a single optimal solution unlikely. However, efforts to integrate various phases of nuclear waste management can have the greatest impact if they begin promptly and continue to evolve throughout the remaining life of the current fuel cycle. A key factor that influences the path forward for integration of nuclear waste management practices is the identification of the timing and location for a repository. The most cost-effective path forward would be to open a repository by mid-century with the capability to directly dispose of DPCs without repackaging the spent fuel into disposalready canisters. Options that involve repackaging of spent fuel from DPCs into disposalready canisters or that delay the repository opening significantly beyond mid-century could add 10s of billions of dollars to the total system life cycle cost.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Quantitative U=O bond activation in uranyl complexes via silyl radical transfer

Reductive silylation of the uranyl dication with 1,4-bis(trimethylsilyl)dihydropyrazine, or “Mashima's reagent”, is detailed. The reagent simultaneously delivers silylium ions and electrons to multiple uranyl complexes, i.e. a pyridine dipyrrolide uranyl complex and a common uranyl-containing starting material, UO 2 Cl 2 (OPPh 3 ) 2 . This reaction results in quantitative activation of thermodynamically robust U=O bonds. Furthermore, the reductive functionalization of environmentally persistent actinyl species with Mashima's reagent is seen as a promising solution for nuclear waste remediation.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2016

Supplemental Environmental Baseline Survey for Proposed Land Use Permit Modification for Expansion of the Dynamic Explosive Test Site (DETS) 9940 Main Complex Parking Lot

The “subject property” is comprised of a parcel of land within the Kirtland Military Reservation, Bernalillo County, New Mexico, as shown on the map in Appendix B of this document. The land requirement for the parking lot addition to the 9940 Main Complex is approximately 2.7 acres. The scope of this Supplemental Environmental Baseline Survey (SEBS) is for the parking lot addition land transfer only. For details on the original 9940 Main Complex see Environmental Baseline Survey, Land Use Permit Request for the 9940 Complex PERM/0-KI-00-0001, August 21, 2003, and for details on the 9940 Complex Expansion see Environmental Baseline Survey, Proposed Land Use Permit Expansion for 9940 DETS Complex, June 24, 2009. The 2.7-acre parcel of land for the new parking lot, which is the subject of this EBS (also referred to as the “subject property”), is adjacent to the southwest boundary of the original 12.3- acre 9940 Main Complex. No testing is known to have taken place on the subject property site. The only activity known to have taken place was the burial of overhead utility lines in 2014. Adjacent to the subject property, the 9940 Main Complex was originally a 12.3-acre site used by the Department of Energy (DOE) under a land use permit from the United States Air Force (USAF). Historical use of the site, dating from 1964, included arming, fusing, and firing of explosives and testing of explosives systems components. In the late 1970s and early 1980s experiments at the 9940 Main Complex shifted toward reactor safety issues. From 1983 to 1988, fuel coolant interaction (FCI) experiments were conducted, as were experiments with conventional high explosives (HE). Today, the land is used for training of the Nuclear Emergency Response community and for research on energetic materials. In 2009, the original complex was expanded to include four additional 20-acre areas: 9940 Training South, 9940 Training East, T-Range 6, and Training West Landing Zone. The proposed use of the subject property is for the purpose of adding a parking lot to serve the increase in customer vehicles that is occurring as the 9940 Main Complex is more heavily utilized, and as the 2009 Expansion areas come online as operational training facilities. The subject property would be used only for parking, not for testing or training activities. The parking lot would have a gravel surface. Current and future work at the 9940 Main Complex involves arming, fuzing, and firing of explosives and the testing of explosive systems components in both terrestrial and aquatic settings. It also involves specialized training activities for a variety of first responder customers, both DOE and non-DOE agencies. The approach was to perform a document search, supplemented by a visual site inspection, to identify potential environmental contamination associated with the property. Factors evaluated included hazardous substances; petroleum products and derivatives; environmental restoration sites; areas of concern; storage tanks; oil/water separators; grease traps; wash racks; waste tanks; pesticides; military munitions/ordnance; medical or bio-hazardous waste; radioactive waste; solid/municipal waste; indoor air quality; groundwater; wastewater treatment, collection, and disposal/discharge; drinking water quality; utilities; asbestos; polychlorinated biphenyls (PCBs); radon; lead-based paint; cultural resources; floodplains; and natural/biological resources. Each of these factors is evaluated separately in Section 5, Findings for Subject Property. The property categorization for this subject property would be considered Category 1- “An area or real property where no storage, release, or disposal of hazardous substances or petroleum products or their derivatives has occurred into the environment or structures or disposed on the subject property (including no migration of these substances from adjacent properties).” There appears to be sufficient information to categorize the subject property and it appears that no further effort needs to be made to obtain additional information. There are no findings of an adverse nature on the subject property itself or from adjacent properties. It is recommended that the proposed transfer of the subject property from the USAF to DOE proceed.

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

Iodobutane Deep Bed Adsorption Test Report

The Department of Energy (DOE) Nuclear Technology Research and Development (NTRD) Program Material Recovery and Waste Form Development (MRWFD) Campaign Off-gas Sigma Team has supported research and development on iodine control and iodine waste forms for the past several years. As a part of this on-going research, two deep-bed iodobutane adsorption tests have been performed with higher NO x levels. High (>1,000) iodine decontamination factors (DFs, the ratios of the inlet iodine concentrations divided by the outlet iodine concentrations, when the gas flowrate is constant) were achieved with both the silver aerogel and silver zeolite sorbents. The iodine adsorption mass transfer zone (the depth of the sorbent bed in which adsorption is occurring at any specific time) for these tests was estimated at 4 inches for both sorbents under the test conditions. Silver utilization for silver aerogel sorbent reached 41% (up to 12 g adsorbed iodine per 100 g silver aerogel sorbent), while silver utilization reached 43% (up to 6.8 g adsorbed iodine per 100 g silver zeolite sorbent). These values are generally consistent with results for prior methyl iodide adsorption tests. The results also indicate that, during iodine adsorption process, the organic iodide molecule decomposes so that the iodine can be adsorbed. Even when sorbent breakthrough is reached, the organic iodide molecule continues to decompose; and the unadsorbed iodine that passes through bed is no longer in the form of the original organic iodide; but in the form of iodine compounds that are soluble in NaOH solutions, such as HI or I 2 . This research and development has included iodine adsorption tests using a laboratory-scale test system containing segmented fixed beds of iodine adsorbents in series. Non-radioactive synthetic gas mixtures were blended from air, nitrogen, NO, NO 2 , water, and the target iodine compounds to represent dissolver off-gas (DOG) streams from used nuclear fuel aqueous reprocessing.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

Thermodynamic analysis on heavy metals partitioning impacted by moisture during the MSW incineration

Highlights: Black-Right-Pointing-Pointer Partitioning of HMs affected by moisture was investigated by thermodynamic analysis. Black-Right-Pointing-Pointer Increase in moisture and in temperature was opposite impact on HMs contribution. Black-Right-Pointing-Pointer The extent of temperature decreased by increase in moisture determines the impact. - Abstract: A thermodynamic calculation was carried out to predict the behavior and speciation of heavy metals (HMs), Pb, Zn, Cu, and Cd, during municipal solid waste (MSW) incineration with the different moisture levels. The calculation was based on the minimization of the total Gibbs free energy of the multi-components and multi-phases closed system reaching chemical equilibrium. The calculation also indicated the reaction directions and tendencies of HMs components. The impacts of chlorine additives (No PVC, 1%PVC, and 5%PVC) and moisture on the behavior of HMs were investigated at different temperature levels in the system (750 Degree-Sign C, 950 Degree-Sign C, and 1150 Degree-Sign C). Furthermore, because the incineration temperature falls down with the increase in moisture in waste, the co-influence of moisture and temperature in combusting MSW on the HMs was also studied with the given chlorine (as 1%PVC + 0.5%NaCl). The results showed that in the non-chlorine system, the impact of the moisture on Pb, Zn, and Cu was not significant, and the ratio of compound transformation was less than 10%, except the Cd compounds at 950 Degree-Sign C and 1150 Degree-Sign C. In the system with low chlorine (as 1%PVC) at constant temperature, the chlorides of HMs (Cd, Pb, Zn, and Cu) transferred to oxides, and when the content of chlorine rose up (as 5%PVC), the ratio of the chlorides of HMs (Cd, Pb, Zn, and Cu) transferring to oxides fell down noticeably. When the moisture varied together with the temperature, the Zn and Cu compounds transferred from chlorides to oxides with increase in moisture as well as decrease in temperature. At the temperature of 700-1000 Degree-Sign C, the impact of temperature on Pb and Cd was little and the moisture was the main factor; while at the temperature of 1000-1200 Degree-Sign C, the impact of increase in moisture and decrease in temperature on Pb and Cd was almost equal and reversed.

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

An innovative implementation of LCA within the EIA procedure: Lessons learned from two Wastewater Treatment Plant case studies

Life Cycle Assessment (LCA) has been identified in the literature as a promising tool to increase the performance of environmental assessments at different steps in the Environmental Impact Assessment (EIA) procedure. However, few publications have proposed a methodology for an extensive integration, and none have compared the results with existing EIA conclusions without LCA. This paper proposes a comprehensive operational methodology for implementing an LCA within an EIA. Based on a literature review, we identified four EIA steps that could theoretically benefit from LCA implementation, i.e., (a) the environmental comparison of alternatives, (b) the identification of key impacts, (c) the impact assessment, and (d) the impact of mitigation measures. For each of these steps, an LCA was implemented with specific goal and scope definitions that resulted in a specific set of indicators. This approach has been implemented in two contrasting Wastewater Treatment Plant (WWTP) projects and compared to existing EIA studies. The results showed that the two procedures, i.e., EIAs with or without inputs from LCA, led to different conclusions. The environmental assessments of alternatives and mitigation measures were not carried out in the original studies and showed that other less polluting technologies could have been chosen. Regarding the scoping step, the selected environmental concerns were essentially different. Global impacts such as climate change or natural resource depletion were not taken into account in the original EIA studies. Impacts other than those occurring on the project site (off-site impacts) were not assessed, either. All these impacts can be significant compared to those initially considered. On the other hand, unlike current LCA applications, EIAs usually address natural and technological risks and neighbourhood disturbances such as noises or odours, which are very important for the public acceptability of projects. Regarding the impact assessment, even if the conclusions of the EIAs with or without LCA were partially common for local on-site impacts, LCA gives crucial additional information on global and off-site impacts and highlights the processes responsible for them. Finally, for all EIA steps investigated, interest in LCA was demonstrated for both WWTP case studies. The feasibility in terms of skills, time and cost of such implementation has also been assessed. - Highlights: • An innovative methodology for a first-stage implementation of LCA in EIA is proposed. • Its applicability is demonstrated on two Wastewater Treatment Plant case studies. • The conclusions for the four EIA steps investigated differ with or without LCA. • LCA provides valuable additional information on 1) global and 2) off-site impacts. • LCA identifies pollution transfers towards a life cycle perspective.

54 ENVIRONMENTAL SCIENCES↗
NASA NTRS1992

US Department of Energy's Efforts in Intelligent Processing Equipment

The Department of Energy (DOE) uses intelligent processing equipment (IPE) technologies to conduct research and development and manufacturing for energy and nuclear weapons programs. This paper highlights several significant IPE efforts underway in DOE. IPE technologies are essential to the accomplishment of DOE's missions, because of the need for small lot production, precision, and accuracy in manufacturing, hazardous waste management, and protection of the environment and the safety and health of the workforce and public. Applications of IPE technologies include environmental remediation and waste handling, advanced manufacturing, and automation of tasks carried out in hazardous areas. DOE laboratories have several key programs that integrate robotics, sensor, and control technologies. These programs embody a considerable technical capability that also may be used to enhance U.S. industrial competitiveness. DOE encourages closer cooperation with U.S. industrial partners based on mutual benefits. This paper briefly describes technology transfer mechanisms available for industrial involvement.

Peavy, Richard D.↗
DOE OSTI2015

Application of the FW-CADIS variance reduction method to calculate a precise N-flux distribution for the FRJ-2 research reactor - 15603

In order to avoid unnecessary radiation exposure and to minimize the radioactive waste and respectively cost by the decommissioning of nuclear facilities, the radiological characteristics and the activity atlases are required. Therefore, a significant step is the accurate determination of the neutron fluence distribution resulting from the operation of a reactor. For this purpose, computer codes based on the Monte-Carlo method are employed because of the flexibility in the modelling of the complex structures and physical processes. In the scope of the research project founded by the German Federal Ministry of Education and Research (BMBF), and implemented by the Institute of Nuclear Engineering and Technology Transfer (NET) -RWTH Aachen, sophisticated simulation tools are developed, which have the capability to cover the range of demands for activity and dose rate distribution using the example of the German research reactor FRJ-2. Due to the complex geometry of FRJ-2, which consists of different structures and configuration, the MCNP5 Monte Carlo Code is the optimal method for a precise neutron transport simulation. This work presents a method to achieve a precise N-Flux distribution for the FRJ-2 research reactor using hybrid deterministic Monte Carlo method. (authors)

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

Glycolate Analysis in Tank 22: Developing and Testing Analytical Methods for the Savannah River Site Liquid Waste System

Researchers developed and tested a range of techniques to support a low mg/L Limit of Quantitation (LOQ) of glycolate in radioactive Tank 22 waste solution by Ion Chromatography (IC) and by proton NMR. For the IC method, Dionex OnGuard II cartridges were tested as a means of removing alkali earth and transition metals that can interfere during ion chromatography applications especially when analyzing for low-concentration, chelating analytes. Additionally, concentrations of nitrate in the raw Tank 22 sample (5,000 to 10,000 mg/L) were managed by using reasonable levels of sample dilution. The resulting IC performance quality was enhanced by improving the baseline, increasing sensitivity [Limit of Quantitation (LOQ) of 12 mg/L and Limit of Detection (LOD) of 4.0 mg/L], and resolving analytes into well-defined Gaussian peaks when using the Dionex OnGuard II H⁺ cartridges to remove matrix interferences. High concentrations of nitrate limited the performance of the IC method. To achieve an acceptable baseline, samples with high nitrate content require more dilution, resulting in higher detection limits. Tests on samples with higher concentrations of nitrate (Tank 30 and 32 supernate with approximately 150000 mg/L nitrate) suggested that IC detection limits for glycolate in these samples would be > 500 mg/L. Thus, an LOQ of 12 mg/L is not feasible by this IC method on evaporator feed samples. An alternative method of glycolate analysis, using proton nuclear magnetic resonance (H NMR), was developed by the research team. In initial tests, the H NMR technique provided reasonable quantitation of glycolate in Tank 22 conditions by direct observation of the liquid. The H NMR method may provide improved detection limits for solutions with higher nitrate concentrations. We recommend further development of this analysis for high nitrate LWS samples such as evaporator feed and evaporator drop tank content. A method for pretreatment of samples using crystalline silicotitanate (CST) was developed and tested. The objective of the pretreatment was to facilitate analysis of samples with higher levels of radioactivity. The pretreatment did not influence the glycolate concentration in solution, and the pretreatment is projected to reduce Cs 137 activity in a sample by a factor of 16,200. The double strike CST pretreatment was demonstrated and would allow milliliters of higher activity samples to be transferred from the Shielded Cells and handled in a containment unit for glycolate analysis. The various studies validated IC and H NMR methods for glycolate analysis, defined the range of applicability, and demonstrated key supporting analytical protocols. Based on the results, high quality glycolate analysis of Tank 22 is feasible down to approximately 12 mg/L, with the potential for broader applicability of the methods to other conditions in the Savannah River Site Liquid Waste System (LWS).

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Gas Retention and Release from Nuclear Legacy Waste - 16449

The retention of gas within corroded magnox sludge waste at Sellafield, UK and secondary reprocessing waste at Hanford, USA has significant economic and safety implications for decommissioning various nuclear legacy buildings, including the magnox swarf storage silos and first generation magnox storage pond. A series of laboratory scale gas retention tests within magnesium hydroxide soft sediments have revealed a 4-51 Pa yield stress range where consolidated beds could retain sufficient gas to become buoyant with respect to a water supernatant. Density inversion could lead to a Rayleigh-Taylor style instability which could result in an upward transfer of radioactive material from the consolidated bed. The applicable yield stress range suggests that such rollover events may occur in weaker sediment than previously hypothesized, based on current understanding of the fluidization and stable channel mechanisms for gas release from weak and very strong sediments respectively. X-ray computed tomography images of gas retained by 7 Pa yield stress soft sediment reveal both a stable foam layer at the top of the bed and regions dense with micro-bubbles which could provide pathways for gas transport through the bed. Extension of these pathways, hidden below the surface of the sediment, to the container walls and the foam layer could represent a novel mechanism for gas release from intermediate strength sediments of <100 Pa yield stress. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2011

Evaluation of AP-Farm Simulant Composition for Rotary Microfilter Testing

This document identifies the feed composition of a Hanford AP tank farm simulant for rotary microfiltration testing. The composition is based on an Hanford Tank Waste Operations Simulator (HTWOS) model run in combination with Tank Waste Information Network (TWINS) data and mineralogical studies of actual waste solids. The feed simulant is intended to be used in test runs at SRNL. The simulant will be prepared in two parts: (1) A supernate, composed of water-soluble salts and (2) The undissolved (actually, undissolvable) solids. Test slurries with distinct solids concentrations (e.g., 0.5, 5 and 10 wt%) are then prepared as needed. The base for the composition of supernate and solids is the modeled feed sequence for a deployment scenario of the Supplemental Pretreatment units within AP-farm. These units comprise a filtration part, the RMF, and a Cesium-removal part, a Small Column Ion Exchange. The primary use of this simulant is for filtration testing - however, in case that it is also used for ion-exchange tests, the amount of Cs-137 that would need to be added is available in Table 1 and Attachment 3. A modified model run (MMR-049) of the Hanford Tank Waste Operations Simulator (HTWOS) system plan 6 case 3 was performed to identify the feed sequence. Case 3 assumed supplemental treatment besides the low activity waste (LAW) melter with supplemental pretreatment supporting the pretreatment facility. The MMR did not cap the duration of supplemental pretreatment to 15 months, but rather used it throughout the entire treatment mission as an add-on option to the pretreatment facility at the Waste Treatment and Immobilization Plant (WTP). Tank 241-AP-105 (AP-105) was chosen as the feed tank to the filtration unit. Other parameters included a fixed minimum of 0.5 wt% solids in the feed and a maximum Na-concentration of 5M in the supernate. The solids rejection from the filtration unit was set to 99.99% and the maximum allowed amount of solids within tank AP-105 was set to 10 wt%. A comprehensive description of the run and the full suite of results were issued as SVF-2364-00. The list of individual feed events including the amounts of liquid and solids transferred for the first five years is added as Attachment 2; the chemical composition of the supernate feed comprises Attachment 3. For the simulant composition, only the first five years of proposed feed delivery were taken into account. The main outcome of MMR-049 was that for the first five years, the feed would come mostly from AP-farms. Multiple delivery campaigns to AP-105 are included in this average feed, while minimizing the amount of contributing tanks to the solids in the feed mix.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2021

CO 2 utilization in built environment via the P CO 2 swing carbonation of alkaline solid wastes with different mineralogy

Carbon mineralization to solid carbonates is one of the reaction pathways that can not only utilize captured CO 2 but also potentially store it in the long term. In this study, the dissolution and carbonation behaviors of alkaline solid wastes (i.e., waste concrete) was investigated. Concrete is one of the main contributors to a large carbon emission in the built environment. Thus, the upcycling of waste concrete via CO 2 utilization has multifaceted environmental benefits including CO 2 emission reduction, waste management and reduced mining. Unlike natural silicate minerals such as olivine and serpentine, alkaline solid wastes including waste concrete are highly reactive, and thus, their dissolution and carbonation behaviors vary significantly. Here, both conventional acid (e.g., hydrochloric acid) and less studied carbonic acid (i.e., CO 2 saturated water) solvent systems were explored to extract Ca from concrete. Non-stoichiometric dissolution behaviors between Ca and Si were confirmed under far-from-equilibrium conditions (0.1 wt% slurry density), and the re-precipitation of the extracted Si was observed at near-equilibrium conditions (5 wt% slurry density), when the Ca extraction was performed at a controlled pH of 3. These experiments, with a wide range of slurry densities, provided valuable insight into Si re-precipitation phenomena and its effect on the mass transfer limitation during concrete dissolution. Next, the use of the partial pressure of CO 2 for the pH swing carbon mineralization process was investigated for concrete, and the results were compared to those of Mg-bearing silicate minerals. In the P CO 2 swing process, the extraction of Ca was significantly limited by the precipitation of the carbonate phase (i.e., calcite), since CO 2 bubbling could not provide a low enough pH condition for concrete–water–CO 2 systems. Furthermore, this study showed that the two-step carbon mineralization via P CO 2 swing, that has been developed for Mg-bearing silicate minerals, may not be viable for highly reactive Ca-bearing silicate materials (e.g., concrete). The precipitated calcium carbonate (PCC) derived from waste concrete via a pH swing process showed very promising results with a high CO 2 utilization potential as an upcycled construction material.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗