REMOVAL OF CHLORIDE FROM ACIDIC SOLUTIONS USING NO2

Chloride (Cl{sup -}) salt processing in strong acids is used to recycle plutonium (Pu) from pyrochemical residues. The Savannah River National Laboratory (SRNL) is studying the potential application of nitrogen dioxide (NO{sub 2}) gas to effectively convert dissolved pyrochemical salt solutions to chloride-free solutions and improve recovery operations. An NO{sub 2} sparge has been shown to effectively remove Cl{sup -} from solutions containing 6-8 M acid (H{sup +}) and up to 5 M Cl{sup -}. Chloride removal occurs as a result of the competition of at least two reactions, one which is acid-dependent. Below 4 M H+, NO2 reacts with Cl- to produce nitrosyl chloride (ClNO). Between 6 M and 8 M H{sup +}, the reaction of hydrochloric acid (HCl) with nitric acid (HNO{sub 3}), facilitated by the presence of NO{sub 2}, strongly affects the rate of Cl{sup -} removal. The effect of heating the acidic Cl{sup -} salt solution without pre-heating the NO{sub 2} gas has minimal effect on Cl{sup -} removal rates when the contact times between NO{sub 2} and the salt solution are on the order of seconds

PURIFICATION OF URANIUM FROM URANIUM/MOLYBDENUM ALLOY

The Savannah River Site will recycle a nuclear fuel comprised of 90% uranium-10% molybdenum by weight. The process flowsheet calls for dissolution of the material in nitric acid to a uranium concentration of 15-20 g/L without the formation of precipitates. The dissolution will be followed by separation of uranium from molybdenum using solvent extraction with 7.5% tributylphosphate in n-paraffin. Testing with the fuel validated dissolution and solubility data reported in the literature. Batch distribution coefficient measurements were performed for the extraction, strip and wash stages with particular focus on the distribution of molybdenum

Flammability Analysis for Actinide Oxides Packaged in 9975 Shipping Containers

Packaging options are evaluated for compliance with safety requirements for shipment of mixed actinide oxides packaged in a 9975 Primary Containment Vessel (PCV). Radiolytic gas generation rates, PCV internal gas pressures, and shipping windows (times to reach unacceptable gas compositions or pressures after closure of the PCV) are calculated for shipment of a 9975 PCV containing a plastic bottle filled with plutonium and uranium oxides with a selected isotopic composition. G-values for radiolytic hydrogen generation from adsorbed moisture are estimated from the results of gas generation tests for plutonium oxide and uranium oxide doped with curium-244. The radiolytic generation of hydrogen from the plastic bottle is calculated using a geometric model for alpha particle deposition in the bottle wall. The temperature of the PCV during shipment is estimated from the results of finite element heat transfer analyses

Benchmarking of Improved DPAC Transient Deflagration Analysis Code

The transient deflagration code DPAC (Deflagration Pressure Analysis Code) has been upgraded for use in modeling hydrogen deflagration transients. The upgraded code is benchmarked using data from vented hydrogen deflagration tests conducted at the HYDRO-SC Test Facility at the University of Pisa. DPAC originally was written to calculate peak deflagration pressures for deflagrations in radioactive waste storage tanks and process facilities at the Savannah River Site. Upgrades include the addition of a laminar flame speed correlation for hydrogen deflagrations and a mechanistic model for turbulent flame propagation, incorporation of inertial effects during venting, and inclusion of the effect of water vapor condensation on vessel walls. In addition, DPAC has been coupled with CEA, a NASA combustion chemistry code. The deflagration tests are modeled as end-to-end deflagrations. The improved DPAC code successfully predicts both the peak pressures during the deflagration tests and the times at which the pressure peaks