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Advances in High Throughput Screening of Biomass Recalcitrance (Poster)

Description: This was a poster displayed at the Symposium. Advances on previous high throughput screening of biomass recalcitrance methods have resulted in improved conversion and replicate precision. Changes in plate reactor metallurgy, improved preparation of control biomass, species-specific pretreatment conditions, and enzymatic hydrolysis parameters have reduced overall coefficients of variation to an average of 6% for sample replicates. These method changes have improved plate-to-plate variation of control biomass recalcitrance and improved confidence in sugar release differences between samples. With smaller errors plant researchers can have a higher degree of assurance more low recalcitrance candidates can be identified. Significant changes in plate reactor, control biomass preparation, pretreatment conditions and enzyme have significantly reduced sample and control replicate variability. Reactor plate metallurgy significantly impacts sugar release aluminum leaching into reaction during pretreatment degrades sugars and inhibits enzyme activity. Removal of starch and extractives significantly decreases control biomass variability. New enzyme formulations give more consistent and higher conversion levels, however required re-optimization for switchgrass. Pretreatment time and temperature (severity) should be adjusted to specific biomass types i.e. woody vs. herbaceous. Desalting of enzyme preps to remove low molecular weight stabilizers and improved conversion levels likely due to water activity impacts on enzyme structure and substrate interactions not attempted here due to need to continually desalt and validate precise enzyme concentration and activity.
Date: June 1, 2012
Creator: Turner, G. B.; Decker, S. R.; Tucker, M. P.; Law, C.; Doeppke, C.; Sykes, R. W. et al.
Partner: UNT Libraries Government Documents Department

OUT-OF-PILE PROPERTIES OF MIXED URANIUM-PLUTONIUM CARBIDES. Final Report

Description: Fabrication studies to produce high density solid solutions of 80% UC-- 20% PuC, with reproducible structure, composition, and density, were completed. Solid solution (U,Pu)C powder was produced by the oxide-carbon reaction, and the powders were consolidated by cold pressing and sintering. The studies were a continuation of work performed on Contract AT(30-1)-2899 with the USAEC. Two types of material were produced: 1. (U/sub 0.8/Pu/sub 0.2/)C/sub 0.95/, single,phase monocarbide pellets with average densities of 12.8 g/cm/sup 3/ (94% of theoretical), sintered at 1925 deg C; 2. (U/subn 0.95/ at 0.1 wt% Ni sintering aid, major monocarbide phase and minor amount of sesquicarbide phase pellets, with average densities of 13.1 g/cm/sup 3/ (96.5% of theoretical), sintered at 1550 deg C. The coefficient of thermal expansion was measured as 11.9 x 10/sup -6/ / deg C for (U/sub 0.8/Pu/sub 0.2/) x C/sub 0.95/ at 25 to 1400 deg C and 12.3 x 10/sup -6// deg C for (U/sub 0.8/Pu/sub 0 .2/)C/sub 0.95/ 0.1 wt% Ni at 25 to 1400 deg C. The presence of the nickel sintering aid did not make a significant difference. the values are similar to those of UC, compared at 950 deg C, the maximum temperature to which UC has been measured. The melting point of (U/sub 0.8/ Pu/sub 0.2/)C/sub 0.95/ was found to be 2480 P 20 deg C. The solidus is estimated to be about 2430 deg C. The vapor pressure data in the temperature range of 2100 to 2600 deg K show that the pressure of plutonium is about the same as uranium over (U/sub 0.8/Pu/sub 0.2/C/sub 0.95/. While the uranium data are about an order of magnitude higher than anticipated, based on data for U over UC (probably due to carbon diffusion across the cell), the thermal stability of the fuel in the temperature range ...
Date: December 1, 1963
Creator: Stahl, D.; Strasser, A.; Taylor, K. & Anderson, J.
Partner: UNT Libraries Government Documents Department

SAFETY CONSIDERATIONS IN AQUEOUS REPROCESSING PLANT OPERATIONS

Description: Safety precautions utilized for control and cominement of fissionable and radioactive materials in the various aqueous reprocessing operations performed at the Idaho Chemical Processing Plant are presented. Three primary nuclear safety controls, geometrical, mass limitation, and concentration control, are used. Operations are penformed according to standard operating procedures which are set up to prevent circumvention of the primary nuclear safety controls. The various processing operations with their particular safety features are discussed. The operations include receipt, handling, and storage of irradiated fuel elements, dissolution of the fuel elements in various reagents, separation of the unburned fissionable material from fission products and fuel element structural materials by solvent extraction, salvage or recycle operations of off- specifications product or waste solutions that exceed the dispossble fuel concentrationalimits, product packaging, storage and shipment, fission product recovery, and waste collection, handling and disposal. The originai plant design and later additions and modifications included built-in geometrical control wherever practical with allowances for possible neutron interaction between vessels. The standard operating procedures specificaily state mass limits and concentration controls required for certain operations which involve appreciable quantities of uranium. Administrative control insures compliance with the standard operating procedures. (auth)
Date: March 1, 1961
Creator: Morrison, W.G.
Partner: UNT Libraries Government Documents Department

Handling Pyrophoric Reagents

Description: Pyrophoric reagents are extremely hazardous. Special handling techniques are required to prevent contact with air and the resulting fire. This document provides several methods for working with pyrophoric reagents outside of an inert atmosphere.
Date: August 14, 2009
Creator: Alnajjar, Mikhail S. & Haynie, Todd O.
Partner: UNT Libraries Government Documents Department

STUDY OF FACTORS INFLUENCING DUCTILITY OF IRON-ALUMINUM ALLOYS. Monthly Letter report No. 5

Description: Deformation mechnisms in Fe-Al single crystals were studied. and observations regarding deformation structures are discussed. In surface- preparation studies. a polishing technique was developed which requires only that the chemical bath temperatures be controlled and thnt the bath be agitated. Bath composition is given, and results obtained with Thermenol treated by this technique are included. (J.R.D.)
Date: October 1, 1960
Creator: Rauscher, G.P. Jr.; Perkins, F.C. & Nachman, J.F.
Partner: UNT Libraries Government Documents Department

EFFECTS OF TERNARY ADDITIONS OF ALUMINUM-35 w/o URANIUM ALLOYS

Description: The effects of a number of ternary additions on the constitution, casting, and fabricating characteristics and the physical properties of aluminum- 35 wt.% uranium were investigated. Initial investigations were concerned with the effects of 3 at.% ternary additions on the microstructure and press-forging characteristics of the alloy. It was found that additions of this magnitude often introduced extrinsic phases in the alloy. At the 3 wt% level, additions of germanium, silicon, tin, or zirconium inhibited the formation of UAl/sub 4/ and thereby increased the extent of the aluminum matrix in the alloy. It was also noted that these additions decreased the pressures required for extruding, and the tin addition also improved the homogeneity of cast shapes. Lead and palladium also improved the homogeneity of the cast material; however, neither of these was an effective inhibitor of UAl/sub 4/ and free lead was detected in the alloy to which lead had been added as the ternary. From these studies it appears that tin and zirconium are as effective as silicon in enhancing the fabricating characteristics of rior when evaluated on the bases of casting qualities and recycling characteristics. (auth)
Date: October 27, 1959
Creator: Daniel, N.E.; Foster, E.L. & Dickerson, R.F.
Partner: UNT Libraries Government Documents Department

EVALUATION OF AN ENGINEERING DEMONSTRATION OF THE MODIFIED ZIRFLEX AND NEUFLEX PROCESSES FOR THE PREPARATION OF SOLVENT EXTRACTION FEEDS FROM UNIRRADIATED ZIRCONIUM-BASE REACTOR FUELS

Description: In order to recover uranium from zirconium-base reactor fuels by solvent extraction, the metailic fuel and cladding must first be dissolved and a suitable feed solution prepared. Such preparations of solvent extraction feeds were successfully accomplished batchwise using both the Modified Zirflex and Neuflex processes employing an NH/sub 4/F -- oxidant mixture to dissolve the fuel elements, and the feed. (The d Zirflex feed, and H/sub 2/O for the Neuflex feed.) In the Modified Zirflex process, a dissolvent about 6 M in NH/sub 4/F with an excess of H/sub 2/O/sub 2/ to oxidize uranium to the more-soluble U(VI) valence state. The off-gas, after NH/sub 3/ removal, is an H/sub 2/-O/sub 2/ mixture of small volume, which is diluted with air to a safe concentration. Then nitric acid-aluminum nitrate is added to the dissolution product, yielding a solvent extraction feed from which uranium is recovered by using TBP-Amsco as the extractant. In the Neuflex process, the dissolvent is NH/sub 4/F--H/sub 2/O/sub 2/, with less than a stoichiometric amount of NH/sub 4/NO/sub 3/. Without NH/sub 4/NO/sub 3/, the scrubbed off-gas is principally hydrogen, on the hydrogen-rich side of the flammable range of H/sub 2/-O/sub 2/ mixtures, Only water is added to this dissolution product, yielding a neutral fluoride feed from which uranium is extractable by use of Dapex reagents. ln both processes the F: Zr charge ratio, initial surface condition, and maximum section thickness of the fuel element were the principa1 determinants of total dissolution time. The zirconium loading as determined by the free fluoride - zirconium solubility relationship limited the capacity of fuels containing less than 2% U, while the free-fluoride-to-uranium ratio of about 100 required for solution stability was the limiting factor with alloys containing higher percentages of uranium, Hydrogen peroxide concentration was not an important factor in solution stability; ...
Date: March 1, 1964
Creator: Kitts, F.G.
Partner: UNT Libraries Government Documents Department

Addendum to the Calcined Waste Storage at the Idaho Nuclear Technology Center

Description: This report is an addendum to the report Calcined Waste Storage at the Idaho Nuclear Technology and Engineering Center, INEEL/EXT-98-00455 Rev. 1, June 2003. The original report provided a summary description of the Calcined Solids Storage Facilities (CSSFs). It also contained dozens of pages of detailed data tables documenting the volume and composition (chemical content and radionuclide activity) of the calcine stored in the CSSFs and the liquid waste from which the calcine was derived. This addendum report compiles the calcine composition data from the original report. It presents the compiled data in a graphical format with units (weight percent, curies per cubic meter, and nanocuries per gram) that are commonly used in regulatory and waste acceptance criteria documents. The compiled data are easier to use and understand when comparing the composition of the calcine with potential regulatory or waste acceptance criteria. This addendum report also provides detailed explanations for the large variability in the calcine composition among the CSSFs. The calcine composition varies as a result of reprocessing different types of fuel that had different cladding materials. Different chemicals were used to dissolve the various types of fuel, extract the uranium, and calcine the resulting waste. This resulted in calcine with variable compositions. This addendum report also identifies a few trace chemicals and radionuclides for which the accuracy of the amounts estimated to be in the calcine could be improved by making adjustments to the assumptions and methods used in making the estimates.
Date: May 1, 2004
Creator: Staiger, M. D.; Swenson, Michael & Thomas, T. R.
Partner: UNT Libraries Government Documents Department

Alternative Feedstocks Program Technical and Economic Assessment: Thermal/Chemical and Bioprocessing Components

Description: This resource document on biomass to chemicals opportunities describes the development of a technical and market rationale for incorporating renewable feedstocks into the chemical industry in both a qualitative and quantitative sense. The term "renewable feedstock?s" can be defined to include a huge number of materials such as agricultural crops rich in starch, lignocellulosic materials (biomass), or biomass material recovered from a variety of processing wastes.
Date: July 1, 1993
Creator: Bozell, J. J. & Landucci, R.
Partner: UNT Libraries Government Documents Department

Fission Product Traps for Use in High-Temperature Gas-Cooled Graphite Reactors

Description: A proposal is given of an approach to a fission-product trapping system which appears feasible on the basis of thermodynamic and other data available. Reactor and trapping conditions are outlined. The half-lives, fission yields, and volatility of the fission products of interest are described. To provide the most effective retention at elevated temperatures, two types of reagents are required: a highly electropositive metal that will not melt or appreciably vaporize and which will form stable non-volatile compounds with non-metallic or near non-metallic fission products; and a reagent to provide a highly electronegative element to form stable, non-volatile compounds with metallic fission products. Thermodynamic properties are included for compounds formed by reactions between the fission products and the trapping reagents. (B.O.G.)
Date: March 13, 1958
Creator: Zumwalt, L. R.
Partner: UNT Libraries Government Documents Department

Sequential Separation of Some Actinide Elements by Anion Exchange

Description: Methods are presented by which trace amounts of several actinide elements are separated. Use is made of the large differences in distribution coefficients, so careful chromatographic techniques are not necessary. Small columns are used, allowing the desired constituent to be obtained in 10 ml or less. One method is used to separate americium, plutonium, and neptunium. These elements are sequentially eluted from columns of Dowex 1 resin in that order by 8 M HNO/sub 3/, 0.02 M ferrous sulfamate in 4.5 M HNO/sub 3/, and 0.001 M ceric sulfate in 0.25 M HNO/sub 3/. Another method is used to separate americium, thorium, plutonium, and neptunium sequentially in that order by S M HNO/sub 3/, 12 M HCl, 12 M HCL--0.1 M NH/sub 4/I, and 4 M HCL by elutriation. Protactinium and uranium follow the americium in both methods. The methods presented are characterized by a low degree of cross contsmnination. Yields are greater than 95 percent. (auth)
Date: June 1, 1959
Creator: Roberts, F. P. & Brauer, F. P.
Partner: UNT Libraries Government Documents Department