Catherine H. Mattus

Dissolution and Separation of Aluminum and Aluminosilicates

Selection of an aluminum alloy for target cladding affects post-irradiation target dissolution and separations. Recent tests with aluminum alloy 6061 yielded greater than expected precipitation in the dissolver, forming up to 10 wt.% solids of aluminum hydroxides and aluminosilicates. Aluminosilicate dissolution presents challenges in a number of different areas, including metals extraction from minerals, flyash treatment, and separations from aluminum alloys. We present experimental work that attempts to maximize dissolution of aluminum metal in caustic, along with silicon, magnesium, and copper impurities, through control of temperature, the rate of reagent addition, and incubation time. Aluminum phase transformations have been identified as a function of time and temperature, using X-ray diffraction. Solutions have been analyzed using wet chemical methods and X-ray fluorescence. These data have been compared with published calculations of aluminum phase diagrams. Approaches are given to enhance the dissolution of aluminum and aluminosilicate phases in caustic solution.

Thermal And Chemical Stability Of Baseline And Improved Crystalline Silicotitanate

The U.S. Department of Energy (DOE) Savannah River Site (SRS) is evaluating technologies for removing radioactive cesium (137Cs) from the supernate solutions stored in the high-level waste tanks at the site. Crystalline silicotitanate (CST) sorbent (IONSIV® IE-911, UOP LLC, Des Plaines, IL), which is very effective at removing cesium from high-salt solutions, was one of three technologies that were tested. Small-scale batch and column tests conducted last year using samples of production batches of CST showed potential problems with CST clumping and loss of cesium capacity after extended contact with the simulant solutions. Similar tests using samples of a baseline and improved granular CST and the CST powder used to make both granular samples were performed this year to compare the performance of the improved CST. The column tests showed that the baseline CST generated more precipitates of sodium aluminosilicate than the improved CST. The sodium aluminosilicate formed bridges between the CST granules, causing clumps of CST to form in the column. Clumps were visible in the baseline CST column after 1 month of operation and in the improved CST column after 2 months. The cesium capacity of the CST samples from the column tests with recirculating simulant decreased slightly as the run time increased. Most of this decrease could be attributed to the weight of cancrinite (a sodium aluminosilicate) on the CST samples. Tests conducted last year using production batch samples of CST showed a more pronounced drop in cesium capacity under comparable conditions. A column test using the improved CST and once-through simulant showed few problems during 5 months of operation. The pressure drop through the column remained low; however, the CST in the column was clumped together when the final samples were taken after 5 months. The final sample taken from the top 1 cm of the column showed a 65% drop in cesium capacity compared with all the other samples from this column. This sample also contained the highest concentration of cancrinite, but the weight of cancrinite would only account for a small fraction of the drop in cesium capacity by simple dilution of the CST. The CST in the batch tests stored at elevated temperatures in average simulant formed clumps, but this occurred at a slower rate than that observed last year during comparable tests using production batch samples of CST. Storage at elevated temperatures caused a gradual decrease in cesium capacity as the storage time increased, with a loss in capacity of up to 20% after 5 to 6 months at 80°C. The results for the baseline and improved CST samples were essentially the same for these batch tests. #The submitted manuscript has been authored by a contractor of the U.S. Government under contract No. W-31-109-ENG-38. Accordingly, the U.S. Government retains a non-exclusive, royalty-free license to publish or reproduce the published form of this contribution, or allow others to do so, for U.S. Government purposes.

Aluminum Target Dissolution in Support of the Pu-238 Program

Selection of an aluminum alloy for target cladding affects post-irradiation target dissolution and separations. Recent tests with aluminum alloy 6061 yielded greater than expected precipitation in the caustic dissolution step, forming up to 10 wt.% solids of aluminum hydroxides and aluminosilicates. We present a study to maximize dissolution of aluminum metal alloy, along with silicon, magnesium, and copper impurities, through control of temperature, the rate of reagent addition, and incubation time. Aluminum phase transformations have been identified as a function of time and temperature, using X-ray diffraction. Solutions have been analyzed using wet chemical methods and X-ray fluorescence. These data have been compared with published calculations of aluminum phase diagrams. Temperature logging during the transients has been investigated as a means to generate kinetic and mass transport data on the dissolution process. Approaches are given to enhance the dissolution of aluminum and aluminosilicate phases in caustic solution.

Reaction rates and prediction of thermal instability during aluminum alloy 6061 dissolution

ABSTRACT Chemical kinetics of dissolution of aluminum alloy 6061 was investigated for the processing of Pu-238 for deep space missions. The rate of dissolution was measured by the heat release and appeared to be controlled by the rate of release of Al(OH)4− from the metal surface. Rates of reaction were measured from 273 to 365 K, giving an activation energy of 72 ± 13 kJ⋅(mol Al)−1 and a pre-exponential factor of 5 ± 3 × 109 dm3mol−1min−1. Minor alloying elements did not appear to affect the reaction kinetics. The average heat of dissolution was −360 ± 70 kJ⋅(mol NaAlO2)−1. When extrapolated to an infinitely dilute solution of aluminum, kJ⋅(mol NaAlO2)−1.

Hydroxylamine Nitrate Decomposition under Non-radiological Conditions

Hydroxylamine nitrate (HAN) is used to reduce Pu(IV) to Pu(III) in the separation of plutonium from uranium. HAN becomes unstable under certain conditions and has been known to explode, causing injury to humans including death. Hence, it is necessary to deactivate HAN once the reduction of plutonium is finished. This report reviews what is known about the chemistry of HAN and various methods to achieve a safe decomposition. However, there are areas where more information is needed to make a decision about the handling of HAN in reprocessing of nuclear fuel. Experiments have demonstrated a number of non-radiolytic ways to safely decompose HAN, including heating in HNO3, photolytic oxidation in the presence of H2O2, and the addition of a metal such as Fe(III) that will oxidize the HAN.

Phosphate Ions: Does Exposure Lead to Degradation of Cementitious Materials?

An assessment of the potential effects of phosphate ions on cementitious materials was made through a review of the literature, contacts with concrete research personnel, and conduct of a “bench-scale” laboratory investigation [1]. The objectives of this limited study were to: (1) review the potential for degradation of cementitious materials due to exposure to high concentrations of phosphate ions; (2) provide an improved understanding of any significant factors that may lead to a requirement to establish exposure limits for concrete structures exposed to soils or ground waters containing high levels of phosphate ions; (3) recommend, as appropriate, whether a limitation on phosphate ion concentration in soils or ground water is required to avoid degradation of concrete structures; and (4) provide a “primer” on factors that can affect the durability of concrete materials and structures in nuclear power plants. Results of a literature review, contacts with industry personnel, and a laboratory investigation indicate that no harmful interactions occur between phosphate ions and cememtitious materials unless phosphates are present in form of phosphoric acid. Relative to the “primer,” a separate NUREG report has been prepared that provides a review of pertinent factors that can affect the durability of nuclear power plant reinforced concrete structures.

Liquid Hydrofluoric Acid Sorption Using Solid Media - Part 1

The conversion of the uranium hexafluoride (UF{sub 6}) which is removed from the Molten Salt Reactor Experiment (MSRE), into a stable oxide for long-term storage will produce a significant amount of slightly contaminated, concentrated aqueous hydrofluoric acid (HF). Sin&the handling of this HF is complicated and dangerous, it was decided to transform it into a stable solid fluoride (e.g., CaF{sub 2}, AlF{sub 3}, and MgF{sub 2}). Tests have been performed to identify the best media to use for trapping the HF. These tests are described in this report. The first series of tests evaluated 37 trapping materials using a 6 wt % solution of HF. The solution was pumped through a 3.8-cm-diam column at a slow rate, and samples were taken in 100-mL batches until it was determined that the media could no longer neutralize the solution. Each bed volume of media was evaluated for its retention of fluoride and for its plugging problems. Mixtures of calcium hydroxide and blast furnace slag (BFS) with high Surface areas (18-30 mesh) performed the best. A mixture of 80 wt % calcium hydroxide and 20 wt % BFS was capable of loading 0.134 g HF per cubic centimeter (cm{sup 3}) of media. Other media that performed well were (a) mixtures of calcium hydroxide and portland cement and (b) pure calcium hydroxide. The second series of tests evaluated media using a 33 wt % HF solution. The best performing media from the first series and some new ones were tested. A 2.54-cm-diam, clear, polyvinyl chloride pipe was used as the column, and solution was introduced to different types or sizes of media using slugs from a pipette or constant flow of {approx}10.7 mL/min from a metering pump. The transparent PVC allowed for observation of acid-media interaction and provided a glimpse into how the media and cartridge were performing in this highly corrosive environment. Results from the second series of tests showed that many of the best performing media from the first series of tests would not do well under the more concentrated solutions of HF. Plugging and vigorous reactions were common in the second series, and calcium hydroxide-based media was ruled out due to its disintegration at any size (1.25-cm diam to 30 mesh). The best performing media was mid-sized (4-18 mesh) soda and lime (soda lime). This media not only stood up well in the HF solution, but it also had great neutralization capability, effectively neutralizing up to {approx}0.5 g HF/cm{sup 3} of media. It is expected that a cartridge of this sorbent will be capable of handling approximately seven batches of HF from the uranium conversion.