Salt Waste Processing Facility Research Articles

An Advanced Solvent for the Caustic-Side Solvent Extraction of Cesium from Nuclear Waste: Comparing Lipophilic Guanidines for Improved Hydrolytic Stability

ABSTRACT An advanced solvent has been developed for application in the extraction of cesium from alkaline nuclear waste. Four lipophilic guanidines have been compared to arrive at a more hydrolysis-resistant suppressor component of the solvent. Plans call for deployment of the Next-Generation Caustic Side-Solvent Extraction (NG-CSSX) process at industrial scale for separation of radioactive Cs-137 from the legacy salt waste stored at the US Department of Energy Savannah River Site (SRS). In the solvent used in NG-CSSX, an alkyl guanidine “suppressor” component facilitates efficient stripping of the loaded cesium from the solvent. However, as typical of guanidine compounds, the suppressor previously used in the NG-CSSX solvent, N,N‘,N’’-tri(3,7-dimethyloctyl)guanidine (TiDG), suffers from hydrolytic degradation under process conditions. Recently, the sterically hindered alkyl guanidine N,N’-dicyclohexyl-N’’-(10-nonadecyl)guanidine (DCNDG) has been found to offer 8 to 44 times greater hydrolysis resistance than TiDG. Here, the process performance characteristics of the NG-CSSX solvent incorporating DCNDG are determined in comparison with TiDG and two other guanidine suppressors. The solvent has been adjusted in density to accommodate use in the SRS Salt Waste Processing Facility and tested under aggressive bench-scale conditions intended to increase plant throughput. Experiments include the effect of ageing at normal and off-normal temperatures of the distribution of Cs+ through the NG-CSSX process, the fate and effects of guanidine degradation products, suppressor capacity, coalescence times for aqueous-solvent dispersions, tendency for emulsification, third-phase formation, and degree of protonation of the guanidine in stripping. While the chosen structures of the four compared guanidines mainly effect differences in stability, subtle differences in other system properties such as emulsion formation and suppressor capacity provide insight into the function of this important solvent component. In comparison with TiDG and two other guanidines, DCNDG provides greatly increased stability while not compromising the excellent functional properties of the NG-CSSX solvent.

Experiments on Cake Development in Crossflow Filtration for High Level Waste

Crossflow filtration is a key process step in many operating and planned waste treatment facilities to separate undissolved solids from supernatant slurries. This separation technology generally has the advantage of self-cleaning through the action of wall shear stress created by the flow of waste slurry through the filter tubes. However, the ability of filter wall self-cleaning depends on the slurry being filtered. Many of the alkaline radioactive wastes are extremely challenging to filtration, e.g., those containing compounds of aluminum and iron having particles whose particle size and morphology reduce cake permeability. Low filter flux can be a bottleneck in waste processing facilities such as the Salt Waste Processing Facility at the Savannah River Site and the Waste Treatment Plant at the Hanford Site. To date, increased rates are generally realized by either increasing the crossflow filter axial flow rate, limited by pump capacity, or by increasing filter surface area limited by space and increasing the required pump load. The Savannah River National Laboratory (SRNL) set up both dead-end and crossflow filter tests to better understand filter performance based on filter media structure, flow conditions, and filter cleaning. Using nonradioactive simulated wastes, both chemically and physically similar to the actual radioactive wastes, the authors performed several tests to demonstrate increases in filter performance. With the proper use of filter flow conditions, filter flow rates can be increased over rates currently realized today. This paper describes the selection of a challenging simulated waste and crossflow filter tests to demonstrate how performance can be improved by varied filter operation methods. Those methods were a slow startup to better develop the filter cake and scouring the filter wall. The results showed that for salt waste and metal oxide hydroxide sludges, the process of backpulsing is not necessary to maintain a good filter flux, and the process of periodically scouring the filter improves filter performance. The results also imply that initial filter operation is important to develop a filter cake that minimized pressure drop, the presence of a filter cake can lead to improved solids separation, and a well-developed cake with periodic scouring may allow a good filter flux to be maintained for long periods of time.

Case Studies from the United Kingdom and the United States of Stakeholder Decision Making on Radioactive Waste Management

Case studies of stakeholders in the United Kingdom and United States are presented. Recently, the United Kingdom nuclear industry has begun to consult stakeholders more widely. Methods of engagement within the British industry have varied. Discussions have been with the understanding that engagement with stakeholders will be “dialogue based” and “inform” final decisions. Engagement is being carried out at several levels: at the national level [via the Nuclear Decommissioning Authority’s (NDA) National Stakeholder Group (NSG)]; local site level (via Site Stakeholder Groups); and project level (usually via the best practicable environmental option process). Preliminary findings of a questionnaire issued to all members of the United Kingdom’s NDA NSG, and associated subgroups focused on stakeholder perceptions of the engagement process. In the United States, the Department of Energy’s Savannah River Site (SRS) Citizens Advisory Board, in Aiken, S.C., reviewed seismic designs for the salt waste processing facility at SRS. The final design, proposed by the Defense Nuclear Facilities Safety Board, provoked heated debate among all parties. Theory and future research are reviewed.

Sorption Modeling of Strontium, Plutonium, Uranium, and Neptunium Adsorption on Monosodium Titanate

:We examined the ability of various equilibrium isotherms to replicate the available data for the adsorption of strontium (Sr), plutonium (Pu), uranium (U), and neptunium (Np) on monosodium titanate (MST) during the treatment of simulated and actual Savannah River Site high-level waste. The data come from numerous experimental studies conducted between 1999 and 2002. The analysis considered 29 isotherm models from the literature. As part of this study, we developed a general method for selecting the best isotherm equation. The selection criteria for rating the isotherm equations considered the relative error in predicting the experimental data, the complexity of the mathematical expressions, the thermodynamic validity of the expressions, and statistical significance for the expressions.The Fowler Guggenheim-Jovanovic Freundlich (FG-JF), the Fowler Guggenheim-Langmuir Freundlich (FG-LF) and the Dubinin-Astashov (DA) isotherms each reliably predicted the actinide and Sr adsorption on MST. The first two models describe the adsorption process by single layer formation and lateral interactions between adsorbed sorbates, while the DA model assumes volume filling of micropores (by osmotic pressure difference). These two mechanisms include mutually exclusive assumptions. However, we cannot determine which model best represents the various adsorption mechanisms on MST. Based on our analysis, the DA model predicted the data well. The DA model assumes that an initial sorption layer forms after which networking begins in the pore spaces, filling the volume by a second mechanism. If this mechanism occurs in MST, as the experimental data suggest, then we expect all the empty and closed spaces of MST to contain actinides and Sr when saturated. Prior microstructure analyses determined that the MST surface is best described as heterogeneous (i.e., a semicrystalline outer layer on an amorphous core) or composite material for adsorption. Therefore, we expect the empty spaces (of nanometer size) between the crystalline units in the fibrous material to provide sorption area for the actinides and Sr. Additional conclusions from this study follow.Since each of the three models work reliably, we recommend use of the computationally simplest model as the primary tool until future work can differentiate between the two mechanisms. The DA model possesses a simpler mathematical form with fewer parameters and operations.The experimental data for actual and simulated wastes generally showed consistent agreement. However, the data sets do include considerable variance from a number of causes including the following:• The Pu sorption data appear most consistent (e.g., between actual and simulated waste) and most easily predicted. Since Pu removal efficiency proves most important for the process design efforts, this consistency of the data proves especially beneficial.• Extremely high mass loadings of U on MST result in multilayer sorption behavior and divergence from classical single monolayer isotherm forms. Prior X-ray studies demonstrate that U begins to network, or form dimers, which agrees with this interpretation. This U behavior also shows a complex interaction, and a direct correlation, with sorption data for the other radionuclides. We believe these data suggest nucléation (e.g., precipitation) of the actinides in the micropore space for both Np and Pu. For Sr, the high U loadings appear to inhibit the sorption of Sr.• Nearly all the solutions contained U as the radionuclide with the highest mass concentration. These data show the widest variance.• The composite data set indicates a notable variance in sorption for different batches of MST. The sorption of Sr with different batches of MST shows the largest variance among the four radionuclides for different batches of MST. This variance remains a relatively unexplored aspect of the process design.• Similarly, the experimental data included a wide variety of solution compositions. As such, the mathematical expressions implicitly account for variances in solution chemistry typical of that anticipated within the Salt Waste Processing Facility and Actinide Removal Process. The reader must consider the ranges of these concentrations when applying the expressions.• Increasing temperature decreases Sr, Pu, and, to a lesser extent, U sorption on MST. The opposite effect occurs with Np. This temperature variance further suggests a nucléation behavior for Np.• Nearly all the data used in developing the sorption models came from experiments using solutions with all the principle radionuclides of interest present simultaneously. We modeled the data without invoking competition between the actinides and Sr despite the large concentrations of both U and Np. Since the model does not explicitly invoke competition, the optimized parameters implicitly carry the impact of interaction within the concentration ranges of the original data. Hence, extrapolation of the models to concentrations markedly outside those ranges may result in poorer predictive ability.