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Laboratory Studies for Separation of Solids from Synthoil Gross Product : Parts 1 and 2

Description: PART 1. A variety of surfactants and agglomerating agents have been added to coal-liquefaction product (from the SYNTHOIL process) to determine the effectiveness of these agents in decreasing the viscosity of the product or inducing agglomeration of suspended solids in the product (and consequently facilitating the removal of solids from the product). More than two dozen additives were tested; however, only two of the additives caused a small reduction in the viscosity of the coal-liquefaction product. PART 2. A variety of organic solvents have been added to coal-liquefaction product (from the SYNTHOIL process) to determine the effectiveness in promoting the precipitation of suspended solids in the product. High-temperature settling of the product in the absence of foreign solvent does not appear to be a promising mechanism for solids removal from this product. However, the promoter-solvent de-ashing scheme has been demonstrated to be a potentially attractive solids-liquid separation method, and kerosene has been found to be an effective promoter solvent for the SYNTHOIL gross product.
Date: November 1976
Creator: Huang, H. & Fischer, J.
Partner: UNT Libraries Government Documents Department

Sodium Waste Technology : a Summary Report

Description: The Sodium Waste Technology (SWT) Program was established to resolve long-standing issues regarding disposal of sodium-bearing waste and equipment. Comprehensive SWT research programs investigated a variety of approaches for either removing sodium from sodium-bearing items, or disposal of items containing sodium residuals. The most successful of these programs was the design, test, and the production operation of the Sodium Process Demonstration Facility at ANL-W. The technology used was a series of melt-drain-evaporate operations to remove nonradioactive sodium from sodium-bearing items and then converting the sodium to storable compounds.
Date: January 1987
Creator: Abrams, C. S. & Witbeck, L. C.
Partner: UNT Libraries Government Documents Department

Separation of Rubidium from Irradiated Aluminum-Encapsulated Uranium

Description: A procedure was developed for separating rubidium from irradiated aluminum encapsulated uranium. The separations procedure produces a final ultra-high purity rubidium chloride product for subsequent high performance mass spectrometric analysis. The procedure involves first removing most of the macro-components and fission products by strong base anion exchange using, first, concentrated HCl, then oxalic acid media and second, selectively separating rubidium from alkaline-earth ions and other alkali-metal ions, including cesium, using Bio-Rex-40 cation-exchange resin. The resultant rubidium chloride is then put through a final vacuum sublimation step. Ultra-pure reagents and specially clean glassware are used throughout the procedure to minimize contamination by naturally-occurring rubidium.
Date: January 1982
Creator: Horwitz, E. P.; Schmitz, F. J. & Rokop, D. J.
Partner: UNT Libraries Government Documents Department

Extraction and Recovery of Plutonium and Americium from Nitric Acid Waste Solutions by the TRUEX Process : Continuing Development Studies

Description: This report summarizes the work done to date on the application of the TRUEX solvent extraction process for removing and separately recovering plutonium and americium from a nitric acid waste solution containing these elements, uranium, and a complement of inert metal ions. This simulated waste stream is typical of a raffinate from a tributyl phosphate (TBP)-based solvent extraction process for removing uranium and plutonium from dissolved plutonium-containing metallurgical scrap. The TRUEX process solvent in these experiments was a solution of TBP and octyl(phenyl)-N,N-diisobutylcarbamoylmethylphosphine oxide (CMPO) dissolved in carbon tetrachloride. A flowsheet was designed on the basis of measured batch distribution ratios to reduce the TRU content of the solidified raffinate to less than or equal to 10 nCi/g and was tested in a countercurrent experiment performed in a 14-stage Argonne-model centrifugal contractor. The process solvent was recycled without cleanup. An unexpectedly high evaporative loss of CCl4 resulted in concentration of the active extractant, CMPO, to nearly 0.30M in the solvent. Results are consistent with this higher CMPO concentration. The raffinate contained only 2 nCi/g of TRU, but the higher CMPO concentration resulted in reduced effectiveness in the stripping of americium from the solvent. Conditions can be easily adjusted to give high yields and good separation of americium and plutonium. Experimental studies of the hydrolytic and gamma-radiolytic degradation of the TRUEX-CCl4 showed that solvent degradation would be (1) minimal for a year of processing this typical feed, which contained no fission products, and (2) could be explained almost entirely by hydrolytic and radiolytic damage to TBP. Even for gross amounts of solvent damage, scrubbing with aqueous sodium carbonate solution restored the original americium extraction and stripping capability of the solvent.
Date: September 1985
Creator: Leonard, R. A.; Vandegrift, G. F.; Kalina, D. G.; Fischer, D. F.; Bane, R. W.; Burris, L. et al.
Partner: UNT Libraries Government Documents Department

Recovery of Actinides from TBP-Na₂CO₃ Scrub-Waste Solutions: the ARALEX Process

Description: A flow-sheet for the recovery of actinides from TBP-Na2CO3scrub-waste solutions has been developed, based on batch extraction data, and tested, using laboratory-scale countercurrent extraction techniques. The process, called the ARALEX process, uses 2-ethyl-1-hexanol (2-EHOH) to extract the TBP degradation products (HDBP and H/sub 2/MBP) from acidified Na2CO3scrub waste leaving the actinides in the aqueous phase. Dibutyl and monobutyl phosphoric acids are attached to the 2-EHOH molecules through hydrogen bonds, which also diminish the ability of the HDBP and H2/MBP to complex actinides. Thus all actinides remain in the aqueous raffinate. Dilute sodium hydroxide solutions can be used to back-extract the dibutyl and monobutyl phosphoric acid esters as their sodium salts. The 2-EHOH can then be recycled. After extraction of the acidified carbonate waste with 2-EHOH, the actinides may be readily extracted from the raffinate with DHDECMP or, in the case of tetra- and hexavalent actinides, with TBP. The ARALEX process can also be applied to other actinide waste streams which contain appreciable concentrations of polar organic compounds (e.g., detergents) that interfere with conventional actinide ion exchange and liquid-liquid extraction procedures.
Date: August 1979
Creator: Horwitz, E. P.; Bloomquist, C. A. A.; Mason, G. W.; Leonard, R. A. & Ziegler, A. A.
Partner: UNT Libraries Government Documents Department

Compatibility of Technologies with Regulations in the Waste Management of H-3, I-129, C-14, and Kr-85: Part 1, Initial Information Base

Description: This report summarizes the information base that was collected and reviewed in preparation for carrying out an analysis of the compatibility with regulations of waste management technologies for disposal of Hydrogen-3, Iodine-129, Carbon-14, and Krypton-85. Based on the review of this literature, summaries are presented here of waste-form characteristics, packaging, transportation, and disposal methods. Also discussed are regulations that might apply to all operations involved in disposal of the four nuclides, including the processing of irradiated fuel in a fuel reprocessing plant, packaging, storage, transport, and final disposal. The compliance assessment derived from this information is reported in a separate document.
Date: August 1983
Creator: Trevorrow, L. E.; Vandegrift, G. F.; Kolba, V. M. & Steindler, M. J.
Partner: UNT Libraries Government Documents Department