Scott B. Aase

DEVELOPMENT OF A SOLVENT EXTRACTION PROCESS FOR CESIUM REMOVAL FROM SRS TANK WASTE

An alkaline-side solvent extraction process was developed for cesium removal from Savannah River Site (SRS) tank waste. The process was invented at Oak Ridge National Laboratory and developed and tested at Argonne National Laboratory using singlestage and multistage tests in a laboratory-scale centrifugal contactor. The dispersion number, hydraulic performance, stage efficiency, and general operability of the process flowsheet were determined. Based on these tests, further solvent development work was done. The final solvent formulation appears to be an excellent candidate for removing cesium from SRS tank waste.

Actinide Separation of High-Level Waste Using Solvent Extractants on Magnetic Microparticles

Abstract Polymeric-coated ferromagnetic particles with an absorbed layer of octyl(phenyl)-N,N-diisobutylcarbamoylmethylphosphine oxide diluted by tributyl phosphate are being evaluated for application in the separation and the recovery of low concentrations of americium and plutonium from nuclear waste solutions. Due to their chemical nature, these extractants selectively complex americium and plutonium contaminants onto the particles, which can be recovered from the waste solution using a magnet. The effectiveness of the extractant-absorbed particles at removing transuranics (TRU) from simulated solutions and various nitric acid solutions was measured by gamma and liquid scintillation counting of plutonium and americium. The HNO3 concentration range was 0.01 to 6 M. The partition coefficients (K d) for various actinides at 2 M HNO3 were determined to be between 3000 and 30,000. These values are larger than those projected for TRU recovery by traditional liquid/liquid extraction. Results from transmission...

Optimizing the coating process of organic actinide extractants on magnetically assisted chemical separation particles

Abstract The coatings of ferromagnetic-charcoal-polymer microparticles (1–25 gm) with organic extractants specific for actinides were optimized for use in the magnetically assisted chemical separation (MACS) process. The organic extractants, octyl (phenyl)-N,N-diisobutylcarbamoylmethylphosphine oxide (CMPO) dissolved in tributyl phosphate (TBP), coated the particles when a carrier organic solvent was evaporated. Coated particles were heated in an oven overnight to drive off any remaining carrier solvent and fix the extractants on the particles. Partitioning coefficients for americium obtained with the coated particles routinely reached 3000–4000 ml g−1, approximately 10 times the separation efficiency observed with the conventional solvent extraction system using CMPO and TBP.

Experimental Verification of Caustic‐Side Solvent Extraction for Removal of Cesium from Tank Waste

Abstract A caustic‐side solvent extraction (CSSX) process was developed to remove Cs from Savannah River Site (SRS) high‐level waste. The CSSX process was verified in a series of flowsheet tests at Argonne National Laboratory (ANL) in a minicontactor (2‐cm centrifugal contactor) using simulant. The CSSX solvent, which was developed at Oak Ridge National Laboratory (ORNL), consists of a calixarene‐crown ether as the extractant, an alkyl aryl polyether as the modifier, trioctylamine as the suppressant, and Isopar®L as the diluent. For Cs removal from the SRS tank waste, the key process goals are that: (1) Cs is removed from the waste with a decontamination factor greater than 40,000 and (2) the recovered Cs is concentrated by a factor of 15 in dilute nitric acid. In the flowsheet verification tests, the objectives were to: (1) prove that these process goals could be met; (2) demonstrate that they could be maintained over a period of several days as the CSSX solvent is recycled; and (3) verify that the process goals could still be met after the solvent composition was adjusted. The change in composition eliminated the possibility that the calixarene‐crown ether could precipitate from the solvent. The process goals were met for each of the verification tests. The results of these tests, which are summarized here, show that the CSSX process is a very effective way to remove Cs from caustic‐side waste. #The submitted manuscript has been created by the University of Chicago as Operator of Argonne National Laboratory (“Argonne”) under Contract No. W‐31‐109‐ENG‐38 with the U.S. Department of Energy. The U.S. Government retains for itself, and others acting on its behalf, a paid‐up, nonexclusive, irrevocable worldwide license in said article to reproduce, prepare derivative works, distribute copies to the public, and perform publicly and display publicly, by or on behalf of the Government.

Reduction of plutonium(VI) in brine under subsurface conditions

The redox stability of PuO22+ was investigated in brine under subsurface conditions. In simulated brines, when no reducing agent was present, 0.1 mM concentrations of plutonium(VI) were stable as regards to reduction for over two years, which was the duration of the experiments performed. In these systems, the plutonyl existed as a carbonate or hydroxy-chloride species. The introduction of reducing agents (e.g. steel coupons, and aqueous Fe2+) typically present in a subsurface repository, however, led to the destabilization of the plutonium(VI) complexes and the subsequent reduction to Pu(IV) under most conditions investigated. X-ray Absorption Near-Edge Spectroscopy (XANES) confirmed that the final oxidation state in these systems was Pu(IV). This reduction lowered the overall steady state concentration of plutonium in the brine by 3−4 orders of magnitude. These results show the importance of considering repository constituents in evaluating subsurface actinide solubility/mobility and provide further evidence of the effectiveness of reduced iron species in the reduction and immobilization of higher-valent plutonium species.

EXAFS/XANES studies of plutonium-loaded sodalite/glass waste forms

A sodalite/glass ceramic waste form is being developed to immobilize highly radioactive nuclear wastes in chloride form, as part of an electrochemical cleanup process. Two types of simulated waste forms were studied: where the plutonium was alone in an LiCl/KCl matrix and where simulated fission-product elements were added representative of the electrometallurgical treatment process used to recover uranium from spent nuclear fuel also containing plutonium and a variety of fission products. Extended X-ray absorption fine structure spectroscopy (EXAFS) and X-ray absorption near-edge spectroscopy (XANES) studies were performed to determine the location, oxidation state, and particle size of the plutonium within these waste form samples. Plutonium was found to segregate as plutonium(IV) oxide with a crystallite size of at least 4.8 nm in the non-fission-element case and 1.3 nm with fission elements present. No plutonium was observed within the sodalite in the waste form made from the plutonium-loaded LiCl/KCl eutectic salt. Up to 35% of the plutonium in the waste form made from the plutonium-loaded simulated fission-product salt may be segregated with a heavy-element nearest neighbor other than plutonium or occluded internally within the sodalite lattice.

Proof-of-concept flowsheet tests for caustic-side solvent extraction of cesium from tank waste.

A caustic-side solvent extraction (CSSX) process to remove cesium from Savannah River Site (SRS) high-level waste was tested in a minicontactor (2-cm centrifugal contactor). In the first phase of this effort, the minicontactor stage efficiency was improved from 60% to greater than 80% to meet the SRS process requirements using a 32-stage CSSX flowsheet. Then, the CSSX flowsheet was demonstrated in a 32-stage unit, first without solvent recycle, then with it. In both cases, the key process goals were achieved: (1) the cesium was removed from the waste with decontamination factors greater than 40,000 and (2) the recovered cesium was concentrated by a factor of 15 in dilute nitric acid. Oak Ridge National Laboratory (ORNL) analysis of the recycled solvent showed no evidence of impurity buildup.

Radiogenic transmutation effects in a crystalline aluminosilicate ceramic : a TEM study.

We demonstrate the use of transmission electron microscopy (TEM) to study the effects of beta-decay of radioactive 137 Cs to 137 Ba in crystalline pollucite (CsAlSi 2 O 6 ). Most prior work on radiation effects in materials has focused on structural damage from alpha radiation. Beta radiation, on the other hand, causes little atomic displacement, but the decay transmutation, that is, the radioactive decay of a radioisotope to an isotope of another element, results in progeny with different the valence and ionic radius. Cesium-137, a fission product of uranium, is a major contaminant at U.S. Department of Energy production facilities. Pollucite is an aluminosilicate ceramic with potential use for long-term storage of 137 Cs. We focused on one of several available 137 Cs sources originally fabricated in the 1970s and 1980s. These sources were small, sealed, stainless steel capsules containing pollucite in which varying amounts of the natural Cs had been replaced by radioactive 137 Cs ( t 1/2 = 30.13 years). The sample chosen for TEM examination, aged for nearly 20 years, contained the most radiogenic barium and was expected to show the largest radiation effects. Bright field transmission images revealed a homogeneous crystalline matrix, with no evidence of distinct Ba phases or ex-solution phenomena resulting from the 137 Cs transmutation. Electron diffraction patterns obtained from several portions of the sample were consistent with literature values for pollucite. These data suggest that little substantial damage was done to the crystal structure of this sample, despite the transmutation of nearly 1.5% of the total cesium to barium over the elapsed 20 years. Although our observations are limited, to our knowledge these are the only available data in which transmutation effects have been isolated from other radiation damage phenomena.