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Forming shoe pounds edge of newly inserted plate workpiece. After many passes of shoe and advances of plate, thick land builds up at edge. Workpiece heated to enable metal to flow without strain hardening. Proposed upset-forming process replaces relatively expensive, time-consuming, and wasteful process in which integral weld lands created by machining metal away from plates everywhere except at lands.
We show very initial work on a specific bioelectrochemical system (BES), a bacteriologically driven 'fuel cell' (BFS), that is intended to process waste products, such as CO2 and brine. (1) Processing is the priority, not power generation (2) Really a Microbial Electrolysis Cell (MEC)
At nuclear facilities, accurate and precise measurements of quantitative activity values and isotopic distributions are needed for radioisotopes in waste items in order to meet the needs of safeguards, criticality controls, facility technical safety requirements, and waste disposal. At the Lawrence Livermore National Laboratory (LLNL), a Canberra WM2200 Segmented Gamma Scanner (SGS) (identified at LLNL as WS1331) was purchased to measure moderate to high levels of special nuclear materials (SNM) in Transuranic (TRU) waste drums generated at the Superblock (SBK) facility. As a result of changes in mission and in waste processing and packaging procedures, the majority of the waste items generated at SBK now only contain moderate to low levels of SNM. Modifications were needed to WS1331 to meet the measurement needs for the lower activity items. An additional Canberra LYNX Digital Signal Processor (DSP) was added to the nuclear instrumentation of WS1331 so optimized spectra could be acquired for both the quantitative activity and isotopic results. After the system was enhanced, measurements were collected to evaluate and assess the impact of the change on results. As a result, the measurement tests indicate that the additional DSP improved the performance of WS1331 for lower activity TRU waste drums.
Highlights: • Fractionation of solid wastes into readily and slowly biodegradable fractions. • Kinetic coefficients estimation from mono-digestion batch assays. • Validation of kinetic coefficients with a co-digestion continuous experiment. • Simulation of batch and continuous experiments with an ADM1-based model. - Abstract: A methodology to estimate disintegration and hydrolysis kinetic parameters of solid wastes and validate an ADM1-based anaerobic co-digestion model is presented. Kinetic parameters of the model were calibrated from batch reactor experiments treating individually fruit and vegetable wastes (among other residues) following a new protocol for batch tests. In addition, decoupled disintegration kinetics for readily and slowly biodegradable fractions of solid wastes was considered. Calibrated parameters from batch assays of individual substrates were used to validate the model for a semi-continuous co-digestion operation treating simultaneously 5 fruit and vegetable wastes. The semi-continuous experiment was carried out in a lab-scale CSTR reactor for 15 weeks at organic loading rate ranging between 2.0 and 4.7 g VS/L d. The model (built in Matlab/Simulink) fit to a large extent the experimental results in both batch and semi-continuous mode and served as a powerful tool to simulate the digestion or co-digestion of solid wastes.
The water from paper and pulp wastes obtained from a mill using the Kraft process is purified by precipitating lignins and lignin derivatives from the waste stream with quaternary ammonium compounds, removing other impurities by activated carbon produced from the cellulosic components of the water, and then separating the water from the precipitate and solids. The activated carbon also acts as an aid to the separation of the water and solids. If recovery of lignins is also desired, then the precipitate containing the lignins and quaternary ammonium compounds is dissolved in methanol. Upon acidification, the lignin is precipitated from the solution. The methanol and quaternary ammonium compound are recovered for reuse from the remainder.
Realizing that challenges are facing US actinide science, separations, and waste form research and development activities including meeting future manpower needs, the U.S Department of Energy (DOE), Office of Environmental Management (EM), Office of Nuclear Energy (NE), and National Nuclear Security Administration (NNSA) jointly co-sponsored a workshop entitled 'Separations Technologies: 'Going from Where We Are to Where We Want to Be' on nuclear separations technologies in Bethesda, Maryland (July 2011) to identify common needs and potential requirements in separations technologies and opportunities for program partnerships. Recognizing the different mission requirements of the sponsoring programs, participants were able to identify areas with common uses and related applications. The two follow-on technical workshops held during the Fall 2012 addressed both the chemical aspects and the physical aspects of nuclear separations. Each workshop consisted of technical discussions regarding future work relevant to DOE goals. The large number of subject matter experts from the national laboratories, academia, and headquarters staff developed a long list of existing and new technology developments in the areas of aqueous separations, pyrochemical processing, waste forms, fuel technology etc... Each of these workshops identified the current state of knowledge or technology, the resources needed, technology development opportunities, and the objectives to reach the desired end-state(s). A consolidated report with its eight appendices will be issued at the end of December 2014 summarizing the outcomes of these workshops. Some examples of these findings will be presented in our contribution. It is clear that, in the current environment of decreasing budgets it is illogical for NE, EM and NNSA to develop its own home-grown technology. Therefore, there is a strong case to be made for leveraging limited resources in order to accomplish all objectives while significantly reducing development costs. (authors)
Thermal energy storage systems designed for energy conservation through the recovery, storage, and reuse of industrial process waste heat are reviewed. Consideration is given to systems developed for primary aluminum, cement, the food processing industry, paper and pulp, and primary iron and steel. Projected waste-heat recovery and energy savings are listed for each category.
The research program entitled 'Development of a Prototype Experiment for Treating CELSS (Controlled Ecological Life Support Systems) and PCELSS (Partially Controlled Ecological Life Support Systems) Wastes to Produce Nutrients for Plant Growth' consists of two phases: (1) the development of the neccessary facilities, chemical methodologies and models for meaningful experimentation, and (2) the application of what methods and devices are developed to the interfacing of waste oxidation with plant growth. Homogeneous samples of freeze-dried human feces and urine have been prepared to ensure comparability of test results between CELSS waste treatment research groups. A model of PCELSS food processing wastes has been developed, and an automated gas chromatographic system to analyze oxidizer effluents was designed and brought to operational status. Attention is given the component configuration of the wet oxidation system used by the studies.
Savannah River Remediation is evaluating changes to its current Defense Waste Processing Facility flowsheet to replace formic acid with glycolic acid in order to improve processing cycle times and decrease by approximately 100x the production of hydrogen, a potentially flammable gas. Higher throughput is needed in the Chemical Processing Cell since the installation of the bubblers into the melter has increased melt rate. Due to the significant maintenance required for the safety significant gas chromatographs and the potential for production of flammable quantities of hydrogen, eliminating the use of formic acid is highly desirable. Previous testing at the Savannah River National Laboratory has shown that replacing formic acid with glycolic acid allows the reduction and removal of mercury without significant catalytic hydrogen generation. Five back-to-back Sludge Receipt and Adjustment Tank (SRAT) cycles and four back-to-back Slurry Mix Evaporator (SME) cycles were successful in demonstrating the viability of the nitric/glycolic acid flowsheet. The testing was completed in FY13 to determine the impact of process heels (approximately 25% of the material is left behind after transfers). In addition, back-to-back experiments might identify longer-term processing problems. The testing was designed to be prototypic by including sludge simulant, Actinide Removal Product simulant, nitric acid, glycolic acid, and Strip Effluent simulant containing Next Generation Solvent in the SRAT processing and SRAT product simulant, decontamination frit slurry, and process frit slurry in the SME processing. A heel was produced in the first cycle and each subsequent cycle utilized the remaining heel from the previous cycle. Lower SRAT purges were utilized due to the low hydrogen generation. Design basis addition rates and boilup rates were used so the processing time was shorter than current processing rates.
Sedimentary activity (erosion, transportation and deposition) is a major landscape-shaping process on many icy worlds of the outer solar system for which an understanding of this process is central to their geologic characterizations. Several recent studies, lead by the authors, have identified the effects of sedimentary activity on icy satellite landforms and recognized the role of mass wasting and volatile loss and redistribution in the evolution of these features. We apply state of the art, physics based, landform evolution modeling to icy satellite sedimentary landforms in order to fully quantitatively characterize how their morphologies change with time, and what their present appearances imply for the initial abundance and distribution of loose material, volatiles, and refractories in original pristine landforms. We will present results of our landform evolution modeling to the development of the landscapes of the Saturnian moons of Hyperion and Helene. Our current conclusions for Hyperion are that its unique appearance can be explained in part by the loss to space of ballistic ejecta during impact events. In order to create the smooth surfaces and the reticulate, honeycomb pattern of narrow divides between old craters, appreciable subsequent modification of crater morphology must occur through mass-wasting processes accompanied by sublimation, probably facilitated by the loss of CO2 as a component of the relief-supporting matrix of the bedrock. This mass wasting effectively destroys small craters, at least in part accounting for the paucity of sub-kilometer craters on Hyperion. Helene’s unusual morphology consists of broad depressions (modified large craters) and a generally smooth surface patterned with streaks and grooves. The streaks appear to be oriented down-gradient, as are the grooves. This pattern suggests intensive mass-wasting as a dominant process. Our initial modeling of this surface suggests a Bingham-like rheological behavior for the loose down-slope moving material. Interestingly, as a Bingham flow, the models indicate that aperiodic "intermittent" behavior to be present, suggesting that periods of quiet steady landform evolution are punctuated short periods of active surface readjustment.
Optional grouted waste forms were formulated for waste streams generated during the production of 99Mo to be compliant with low-level radioactive waste regulations. The amounts and dose rates of the various waste form materials that would be generated annually were estimated and used to determine the effects of various waste processing options, such as the of number irradiation cycles between uranium recovery operations, different combinations of waste streams, and removal of Pu, Cs, and Sr from waste streams for separate disposition (which is not evaluated in this report). These calculations indicate that Class C-compliant grouted waste forms can be produced for all waste streams. More frequent uranium recovery results in the generation of more chemical waste, but this is balanced by the fact that waste forms for those waste streams can accommodate higher waste loadings, such that similar amounts of grouted waste forms are required regardless of the recovery schedule. Similar amounts of grouted waste form are likewise needed for the individual and combined waste streams. Removing Pu, Cs, and Sr from waste streams lowers the waste form dose significantly at times beyond about 1 year after irradiation, which may benefit handling and transport. Although these calculations should be revised after experimentally optimizing the grout formulations and waste loadings, they provide initial guidance for process development.
In recent years there has been growing interest in developing recycling technologies for composites manufacturing scrap, process waste and end-of-life parts. The focus of this work was to establish processing routes and mechanical property bounds for glass-polypropylene (PP-GF) scrap from the production of parts for truck trailers, automobiles, and rail cars. This study considered PP-GF scrap and demonstrated extrusion-compression molding (ECM) as a viable route for the closed-loop manufacture of composite parts. The results were promising in terms of the strength and modulus retention of the PP-GF recyclate. The tensile strength and modulus was the highest for 50% and 66% recycled content, compared with 100% and 83% recycle content. The flexural strength and modulus of the 100% and 83% recycled compositions was higher than the 66% and 50% recycled content, respectively. The impact energy absorption of the PP-GF recyclate at at all fiber loadings was superior in absorbing energy compared with the incumbent (benchmark) plywood. This work is useful to designers seeking to incorporate recycled materials in their products.
Selective sequestration of sulfate (SO42-) in the form of barite (BaSO4) from alkaline solutions of high ionic strength containing carbonate at high concentrations resembling legacy radioactive waste generated during plutonium cold war era production is problematic due to the preferential formation of BaCO3. Incorporation of sulfate into the insoluble and thermally stable BaSO4 phase can potentially benefit radioactive waste processing in reducing operational challenges and suppressing volatilization of other waste components such as technetium-99. To enhance selectivity of SO42- sequestration, a series of Ba(II)-Sn(IV) materials was prepared using simple hydrothermal synthesis from different Sn(II) and Sn(IV) precursors. Structural characterization indicated that all obtained products predominantly contained BaSn(OH)6 and Ba2SnO2(OH)4(H2O)10 crystalline phases which were disrupted upon exposure to SO42- due to formation of BaSO4. Performance of the Ba(II)-Sn(IV) materials was tested using complex alkaline solutions simulating radioactive waste containing 0.094 M SO42- and 0.5 M CO32- among other constituents. About 54 – 66% of SO42- was converted to BaSO4 when a quantity of Ba(II)-Sn(IV) material containing approximately a stoichiometric amount of Ba2+ relative to SO42- was used. In comparison, previous studies indicate negligible BaSO4 formation under similar conditions when a simple Ba2+ salt is used. This improvement is attributed to the selective replacement of the Ba(II) stannate by SO42-. Thermal stability of the sulfate-loaded product material up to 1100 °C was demonstrated. Obtained materials offer a convenient and economical option for the selective sequestration or removal of SO42- from carbonate containing solutions.
From its inception in the early 1950s through the end of the Cold War in the early 1990s, the Savannah River Site (SRS) produced nuclear materials for national defense in five reactors. Additionally, irradiated reactor fuel and target tubes were dissolved in nitric acid to recover plutonium and uranium using the PUREX (Plutonium Uranium Reduction EXtraction) process. Liquid waste from these chemical separations processes was then stored onsite in 51 underground tanks. Eight waste storage tanks have been operationally closed (i.e. cleaned and grouted) and the remaining tanks hold a mixture of liquids, insoluble solids, and precipitated salts (SRMC-LWP-2022-00001), the latter generated by evaporating water from the liquid waste. Waste is currently being retrieved from tanks and separated into 1) high-radioactivity, low-volume, and 2) low-radioactivity, high-volume components, principally through the Salt Waste Processing Facility (SWPF) (SRMC-LWP-2023-00001). The former waste stream is vitrified in the Defense Waste Processing Facility (DWPF), stored onsite, and destined for offsite disposal in a deep geologic repository. The latter stream is mixed with dry cementitious materials in the Saltstone Production Facility (SPF) and the wet slurry placed in onsite Saltstone Disposal Units (SDUs) within the Saltstone Disposal Facility (SDF), where it hardens into a cement waste form termed saltstone. A low-infiltration surface cover system will be placed over the SDF at closure, where SDUs will then be in the subsurface post-closure (SRR-CWDA-2019-00001).
Expensive tooling often constraints the use of composites in the design and development of automotive parts. While there is significant confidence and knowledge in sheet and bulk metals, composite processes are less understood in mass production environment. The processes used to produce composites and resulting properties are influenced by fiber length attrition, resin to fiber ratio, process waste etc. Tool designs are determined very early in the engineering process. It is cost prohibitive to build additional tools, in the event it becomes obvious a better processing method and material would be beneficial, the original decision is not easily changed. In the present work we recognize the bottleneck of tooling costs and provide an approach of multi-process tooling. The innovation of this work is the design and demonstration of a single tool for different processes namely injection, injection-compression and extrusion-compression. The materials used in this study were long and short fiber thermoplastics (LFTs and SFTs). The resulting structure-property relationships have been reported for the materials and processing methods with a battery tray (BT) tool.
In this memorandum, the chemical and radionuclide contaminant results from the Fourth Quarter Calendar Year 2015 (CY15) sample of Tank 50H salt solution are presented in tabulated form. The Fourth Quarter CY15 Tank 50H samples were obtained on October 29, 2015 and received at Savannah River National Laboratory (SRNL) on October 30, 2015. The information from this characterization will be used by Defense Waste Processing Facility (DWPF) & Saltstone Facility Engineering for the transfer of aqueous waste from Tank 50H to the Salt Feed Tank in the Saltstone Production Facility, where the waste will be treated and disposed of in the Saltstone Disposal Facility. This memorandum compares results, where applicable, to Saltstone Waste Acceptance Criteria (WAC) limits and targets. Data pertaining to the regulatory limits for Resource Conservation and Recovery Act (RCRA) metals will be documented at a later time per the Task Technical and Quality Assurance Plan (TTQAP) for the Tank 50H saltstone task. The chemical and radionuclide contaminant results from the characterization of the Fourth Quarter Calendar Year 2015 (CY15) sampling of Tank 50H were requested by SRR personnel and details of the testing are presented in the SRNL Task Technical and Quality Assurance Plan.
In this memorandum, the chemical and radionuclide contaminant results from the Third Quarter Calendar Year 2016 (CY16) sample of Tank 50H salt solution are presented in tabulated form. The Third Quarter CY16 Tank 50H samples (a 200 mL sample obtained 6” below the surface (HTF-5-16-63) and a 1 L sample obtained 66” from the tank bottom (HTF-50-16-64)) were obtained on July 14, 2016 and received at Savannah River National Laboratory (SRNL) on the same day. Prior to obtaining the samples from Tank 50H, a single pump was run at least 4.4 hours, and the samples were pulled immediately after pump shut down. The information from this characterization will be used by Defense Waste Processing Facility (DWPF) & Saltstone Facility Engineering for the transfer of aqueous waste from Tank 50H to the Saltstone Production Facility, where the waste will be treated and disposed of in the Saltstone Disposal Facility. This memorandum compares results, where applicable, to Saltstone Waste Acceptance Criteria (WAC) limits and targets. Data pertaining to the regulatory limits for Resource Conservation and Recovery Act (RCRA) metals will be documented at a later time per the Task Technical and Quality Assurance Plan (TTQAP) for the Tank 50H saltstone task. The chemical and radionuclide contaminant results from the characterization of the Third Quarter CY16 sampling of Tank 50H were requested by Savannah River Remediation (SRR) personnel and details of the testing are presented in the SRNL TTQAP.
Millions of gallons of radioactive waste are stored in underground storage tanks at the Hanford Site in Washington State. This waste will be vitrified at the Waste Treatment and Immobilization Plant that is under construction. Vessels in the pretreatment portion of the plant are being configured for processing waste slurries with challenging physical and rheological properties that range from Newtonian slurries to non-Newtonian sludge. Pulse jet mixing technology has been selected for mobilizing and mixing this waste. In the pulse jet mixing process, slurry is expelled from pulse tube nozzles directed towards the vessel floor. The expelled fluid forms a radial jet that erodes the settled layer of solids. The pulse tubes are configured in a ring or multiple rings and operate concurrently. The expelled fluid and mobilized solids traverse toward the center of the tank. At the tank center the jets from pulse tubes in the ring collide and lift solids upward in a central plume. At the end of the pulse, when the desired fluid volume is expelled from the pulse tube, the applied pressure switches to suction and the pulse tube is refilled. This cycle is used to mobilize and mix the tank contents. An initial step of the process is the erosion of solids from the vessel floor by the radial jets that form on the vessel flow beneath each pulse tube. Experiments have been conducted using simulants to evaluate the ability of the pulse jet mixing system radial jets to combine to develop the central upwell and lift solids into the vessel. These experiments have been conducted at three scales using a range of granular simulants over a range of concentrations. The vessels have elliptical, spherical, or flanged and dished bottoms. Process parameters evaluated include the velocity of fluid expelled from the pulse tube, the duration of the pulse and the duty cycle, the ratio of pulse duration to cycle time. Videos taken from beneath the vessel show the growth of the cleared area from each pulse tube as a function of time. All solids are lifted from the vessel bottom when the system is operating at the critical suspension velocity. The focus of this paper is to compare and contrast erosion patterns developed from different simulants and pulse tube configurations. The cases are evaluated to determine how changes in process parameters affects the PJM ability to mobilize solids from the vessel floor.