Benjamin D. Roach

N, N′-Dicyclohexyl-N″-Isotridecylguanidine as Suppressor for the Next Generation Caustic Side Solvent Extraction (NG-CSSX) Process

The purity, concentration, and source of the N,N′-dicyclohexyl-N″-isotridecylguanidine (DCiTG) suppressor (guanidine) used in the NG-CSSX process were found to influence solvent performance. As the starting isotridecanol used in the preparation of DCiTG is comprised of a mixture of branched-chain aliphatic alcohols, varying in composition with manufacturer, the resulting DCiTG itself is a mixture. Thus, it is necessary to address how the solvent performance will be affected by the different preparations of the DCiTG solvent component. In this study, four preparations of DCiTG from three sources were analyzed and evaluated for purity and performance, both in the absence and presence of a deliberately added anionic surfactant impurity.

Radiolytic Treatment of the Next-Generation Caustic-Side Solvent Extraction (NGS) Solvent and its Effect on the NGS Process

It is shown in this work that the solvent used in the Next Generation Caustic-Side Solvent Extraction (NGS) process can withstand a radiation dose well in excess of the dose it would receive in multiple years of treating legacy salt waste at the US Department of Energy Savannah River Site. The solvent was subjected to a maximum of 50 kGy of gamma radiation while in dynamic contact with each of the aqueous phases of the current NGS process, namely SRS−15 (a highly caustic waste simulant), sodium hydroxide scrub solution (0.025 M), and boric acid strip solution (0.01 M). Bench-top testing of irradiated solvent confirmed that irradiation has inconsequential impact on the extraction, scrubbing, and stripping performance of the solvent up to 13 times the estimated 0.73 kGy/y annual absorbed dose. Stripping performance is the most sensitive step to radiation, deteriorating more due to buildup of p-sec-butylphenol (SBP) and possibly other proton-ionizable products than to degradation of the guanidine suppressor, as shown by chemical analyses.

Thermal degradation of the solvent employed in the next-generation caustic-side solvent extraction process and its effect on the extraction, scrubbing, and stripping of cesium

As part of the ongoing development of the Next-Generation Caustic-Side Solvent Extraction (NG-CSSX) process, the thermal stability of its process solvent, the Next-Generation Caustic-Side Solvent (NGS) was investigated and shown to be adequate for industrial application. The solvent was thermally treated at 35°C over a period of 13 months whilst in dynamic contact with each of the aqueous phases of the current NG-CSSX process, namely SRS−15 (a highly caustic waste simulant), sodium hydroxide scrub solution (0.025 M), and boric acid strip solution (0.01 M). The effect of thermal treatment was evaluated by assessing batch extract/scrub/strip performance as a function of time, by monitoring the sodium extraction capacity of the solvent, and by analysis of the solvent using electrospray mass spectrometry. Current studies indicate that the NGS should be thermally robust without replenishment for a period of 7 months in the Modular Caustic-Side Solvent Extraction Unit (MCU), which has been treating waste using NG-CSSX since early 2014 at the Savannah River Site. The guanidine suppressor appears to be the solvent component most significantly impacted by thermal treatment of the solvent, showing significant degradation over extended operation.

Recommended Guanidine Suppressor for the Next-Generation Caustic-Side Solvent Extraction Process

The guanidine recommended for the Next-Generation Caustic-Side is N,N ,N -tris(3,7-dimethyloctyl)guanidine (TiDG). Systematic testing has shown that it is significantly more lipophilic than the previously recommended guanidine DCiTG, the active extractant in the commercial guanidine product LIX -79, while not otherwise changing the solvent performance. Previous testing indicated that the extent of partitioning of the DCiTG suppressor to the aqueous strip solution is significantly greater than expected, potentially leading to rapid depletion of the suppressor from the solvent and unwanted organic concentrations in process effluents. Five candidate guanidines were tested as potential replacements for DCiTG. The tests included batch extraction with simulated waste and flowsheet solutions, third-phase formation, emulsion formation, and partition ratios of the guanidine between the solvent and aqueous strip solution. Preliminary results of a thermal stability test of the TiDG solvent at one month duration indicated performance approximately equivalent to DCiTG. Two of the guanidines proved adequate in all respects, and the choice of TiDG was deemed slightly preferable vs the next best guanidine BiTABG.

Integration of a hyphenated HPIC-ICPMS protocol for the measurement of transplutonium isotopic mass distributions for 252Cf campaigns at Oak Ridge National Laboratory

Rapid measurement of transplutonium isotopic mass distributions during 252Cf production campaigns at Oak Ridge National Laboratory is a critical need. Mass measurements for the isotopes of plutonium, americium, curium, and californium are routinely requested to support the α-hydroxy-isobutyrate runs for the purification and recovery of the heavy curium target material and final 249Bk, 252Cf, and 254Es enriched isotope products. This paper presents the integration of an online high-pressure ion chromatography inductively coupled plasma mass spectrometry technique together with the protocol and chemistry that allows for rapid baseline separation and direct quantification of the transplutonium elemental concentrations and isotopic compositions.

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.