4 Matching Results

Willis Library will be without power on Tuesday, August 20, 2019 from 5:00-7:00 AM CDT. All websites and web services will be down during this period.

Search Results

Advanced search parameters have been applied.

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

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

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

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