Joseph F. Birdwell

Robustness of the CSSX Process to Feed Variation: Efficient Cesium Removal from the High Potassium Wastes at Hanford

This contribution finds the Caustic‐Side Solvent Extraction (CSSX) process to be effective for the removal of cesium from the Hanford tank‐waste supernatant solutions. The Hanford waste types are more challenging than those at the Savannah River Site (SRS) in that they contain significantly higher levels of potassium, the chief competing ion in the extraction of cesium. By use of a computerized CSSX thermodynamic model, it was calculated that the higher levels of potassium depress the cesium distribution ratio (D Cs), as validated to within ±11% by the measurement of D Cs values on various Hanford waste‐simulant compositions. A simple analog model equation that can be readily applied in a spreadsheet for estimating the D Cs values for the varying waste compositions was developed and shown to yield nearly identical estimates as the computerized CSSX model. It is concluded from the batch distribution experiments, the physical‐property measurements, the equilibrium modeling, the flowsheet calculations, and the contactor sizing that the CSSX process as currently formulated for cesium removal from alkaline salt waste at the SRS is capable of treating similar Hanford tank feeds, albeit with more stages. For the most challenging Hanford waste composition tested, 31 stages would be required to provide a cesium decontamination factor (DF) of 5000 and a concentration factor (CF) of 2. Commercial contacting equipment with rotor diameters of 10 in. for extraction and 5 in. for stripping should have the capacity to meet throughput requirements, but testing will be required to confirm that the needed efficiency and hydraulic performance are actually obtainable. Markedly improved flowsheet performance was calculated based on experimental distribution ratios determined for an improved solvent formulation employing the more soluble cesium extractant BEHBCalixC6 used with alternative scrub and strip solutions, respectively 0.1 M NaOH and 0.010 M boric acid. The improved solvent and flowsheet can meet minimum requirements (DF = 5000 and CF = 2) with 15 stages or more ambitious goals (DF = 40,000 and CF = 15) with 19 stages. Thus, a modular CSSX application for the Hanford waste seems readily obtainable with further short‐term development.

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.

Coupling a transient solvent extraction module with the separations and safeguards performance model.

A past difficulty in safeguards design for reprocessing plants is that no code existed for analysis and evaluation of the design. A number of codes have been developed in the past, but many are dated, and no single code is able to cover all aspects of materials accountancy, process monitoring, and diversion scenario analysis. The purpose of this work was to integrate a transient solvent extraction simulation module developed at Oak Ridge National Laboratory, with the SSPM Separations and Safeguards Performance Model, developed at Sandia National Laboratory, as a first step toward creating a more versatile design and evaluation tool. The SSPM was designed for materials accountancy and process monitoring analyses, but previous versions of the code have included limited detail on the chemical processes, including chemical separations. The transient solvent extraction model is based on the ORNL SEPHIS code approach to consider solute build up in a bank of contactors in the PUREX process. Combined, these capabilities yield a much more robust transient separations and safeguards model for evaluating safeguards system design. This coupling and the initial results are presented. In addition, some observations toward further enhancement of separations and safeguards modeling based on this effort are provided, including: items to be addressed in integrating legacy codes, additional improvements needed for a fully functional solvent extraction module, and recommendations for future integration of other chemical process modules.

URANIUM EXTRACTION SELECTIVITIES OF DIBUTYL CARBITOL AND TRIBUTYL PHOSPHATE IN THE SYSTEM UO2(NO3)2-HNO3H2O-Al(NO3)3-SOLVENT

ABSTRACT Historically, both the BUTEX (dibutyl carbitol-based) and PUREX (tributyl phosphate-based) processes have been used for uranium recovery and purification. Currently, BUTEX- and PUREX-type extraction processes are being used in series for recovery of isotopically enriched uranium at the Oak Ridge Y-12 Plant. The use of two solvents is predicated on the differing selectivities of each with regard to the contaminant elements present in the uranium source stream. As part of efforts to streamline plant operations in response to decreasing throughput requirements, the Y-12 Development Division is evaluating options for converting the existing two-solvent operation to a single-solvent process. At the request of the Y-12 Development Division, the Robotics and Process Division at the Oak Ridge National Laboratory has undertaken evaluation of solvents for use in a single-solvent recovery process. Initial efforts have been directed toward development of a single-solvent, dibutyl carbitol-or tributyl phospha...

FY 2007 LDRD Director's R&D Progress SummaryProposal Title: Developing a Science Base for Fuel Reprocessing Separations in the Global Nuclear Energy Program

This work is aimed at developing an experimentally validated computational capability for understanding the complex processes governing the performance of solvent extraction devices used for separations in nuclear fuel reprocessing. These applications pose a grand challenge due to the combination of complicating factors in a three-dimensional, turbulent, reactive, multicomponent, multiphase/interface fluid flow system. The currently limited process simulation and scale-up capabilities provides uncertainty in the ability to select and design the separations technology for the demonstration plan of the Global Nuclear Energy Partnership (GNEP) program. We anticipate the development of science-based models for technology development and design. This project will position ORNL to address the emerging opportunity by creating an expandable process model validated experimentally. This project has three major thrusts, namely, a prototype experimental station, a continuum modeling and simulation effort, and molecular modeling and kinetics support. Excellent progress has been made in corresponding activities in this first year in: (1) defining, assembling, and operating a relevant prototype system for model validation; (2) establishing a mathematical model for fluid flow and transport; (3) deploying sub-scale molecular modeling.