Guillermo D. Del Cul

Preparation of spherical, dense uranium fuel kernels with carbon

The internal gelation process and a suitable broth formulation with an uranium concentration of 1.3 M was used to produce air-dried uranium trioxide dihydrate (UO3·2H2O) and carbon microspheres with crush strengths greater than 600 g per microsphere. The addition of carbon lowered the slow-pour densities of the air-dried microspheres by a minimum of 9% if all other conditions were held constant. The crush strengths of the air-dried microspheres with and without carbon remained very good. These microspheres were not prone to leach when they were washed with ammonium hydroxide, and they did not have the tendency to crack during subsequent heat treatments. For the UO3·2H2O microspheres with and without carbon, dehydration occurred at the same rate. The dehydration was accompanied by spontaneous reduction of the urania to UO2.67. In the same temperature range, hydrogen and carbon can be used to further reduce the urania to uranium dioxide. Therefore, the loss of carbon during calcination appears to be unavoidable. The current recommendation on calcinations is to use a temperature of 600 °C or higher to minimize the loss of carbon. Dense and strong uranium fuel kernels with carbon were produced in argon at 1680 °C.

Vibrational spectrum of strychnine: detection at the nanogram level using a raman microscope

The complete vibrational spectra of strychnine is reported. Strychnine, a compound with medicinal value at the proper dosage, is also a potent toxin at certain concentration levels. Down to 5 ng of strychnine were detected from its spectrum using a microscope attached to a Raman spectrophotometer. The complete Raman and Fourier transform IR spectra were also recorded for positive identification. The prominent vibrational band wavenumbers useful as a reference source for the identification of strychnine in an emergency situation are discussed. The Raman spectra of a bulk sample and of a strychnine microcrystal are given to facilitate positive identification. The potential application of the vibrational analysis of strychnine for forensic diagnostics is discussed. Copyright

The Influence of Lewis Acid/Base Chemistry on the Removal of Gallium by Volatility from Weapons-Grade Plutonium Dissolved in Molten Chlorides

Abstract It has been proposed that GaCl3 can be removed by direct volatilization from a Pu-Ga alloy that is dissolved in a molten chloride salt. Although pure GaCl3 is quite volatile (boiling point: 201°C), the behavior of GaCl3 dissolved in chloride salts is quite different because of solution effects and is critically dependent upon the composition of the solvent salt (i.e., its Lewis acid/base character). In this technical note, the behavior of gallium in prototypical Lewis acid and Lewis base salts is contrasted. It is found that gallium volatility is suppressed in basic melts and promoted in acidic melts. These results have an important influence on the potential for simple gallium removal in molten salt systems.

Evaluation of Annular Centrifugal Contactors for the Extraction of Acetic Acid from Aqueous Streams

This study investigates the use of annular centrifugal contactors for the liquid-liquid extraction of acetic acid from an acidic aqueous phase into an organic phase consisting of 1.5 M Tributyl Phosphate in n-Dodecane. Initial break time tests were performed in order to investigate the mixing/separation viability of the organic/aqueous system, and after determining that centrifugal contactors could be used to perform the liquid-liquid extraction, hydraulic tests established the combination of rotation rate and throughput which should be used to ensure proper separation of the two outlet phases. Finally, extraction efficiency data was collected to examine the system conditions that provided the most efficient removal of acetic acid.

Water Content of Organic Solvents and their Relationship to Extraction of Nitric and Acetic Acid from UREX+ Streams

This study investigates the equilibrium absorption of water in various solvents and solvent-mixtures being considered for the counter-current solvent extraction of acetic acid from improved Uranium Extraction (UREX+) process solutions. It then seeks to determine if there is any correlation between the equilibrium water content of these solvents and their equilibrium extraction of 0.25 M nitric and 0.025 M acetic acid. The UREX+ process is a proliferation resistant version of the Plutonium Uranium Extraction (PUREX) process. The solvents studied were n-Dodecane (nDD), 1,2 Dichloroethane (DCE), and Phenyltrifluoromethyl Sulfone (FS-13), and mixtures of these solvents with Tributyl Phosphate (TBP). After studying both pure water and acidified aqueous systems, it seems the water absorption mechanism is independent of the diluent used and remains constant with the addition of the 0.25 M nitric and 0.025 M acetic acid.

Development of the process for the recovery and conversion of 233UF6 chemisorbed in NaF traps from the molten salt reactor remediation project

Abstract The Molten Salt Reactor Experiment (MSRE) site at Oak Ridge National Laboratory is being cleaned up and remediated. The removal of ~37 kg of fissile 233U is the main activity. Of that inventory, ~23 kg has already been removed as UF6 from the piping system and chemisorbed in 25 NaF traps. This material is in temporary storage while it awaits conversion to a stable oxide. The planned recovery of ~11 kg of uranium from the fuel salt will generate another 15 to 19 NaF traps. The remaining 2 to 3 kg of uranium are present in activated charcoal beds, which are also scheduled to be removed from the reactor site. Since all of these materials (NaF traps and the uranium-laden charcoal) are not suitable for long-term storage, they will be converted to a uranium oxide (U3O8), which is suitable for long-term storage. The conversion of the MSRE material into an oxide presents unique problems, such as criticality concerns, a large radiation field caused by the daughters of 232U (an impurity isotope in the 233U), and the possible spread of the high-radiation field from the release of 220Rn gas. To overcome these problems, a novel process was conceived and developed. This process was specially tailored for providing remote operations inside a hot cell while maintaining full containment at all times to avoid the spread of contamination. This process satisfies criticality concerns, maximizes the recovery of uranium, minimizes any radiation exposure to operators, and keeps waste disposal to a minimum.