Barry B. Spencer

Extracting Acetic Acid from Acidic Solutions

Abstract In the UREX + process, acetic acid must be removed from the raffinate stream to avoid interference with the recovery and recycle of nitric acid solutions. Solvent extraction was selected to be the most promising approach to accomplish this cleanup. Acetic acid partitioning into pure diluents used in the UREX + process were found to be too low for an effective separation. Of the solvents tested, the most promising solvents for the extraction of acetic acid were found to be TBP in dodecane and TBP in FS-13.

Removing Acetic Acid from a UREX+ Waste Stream: A Review of Technologies

Abstract This study explores different technologies for removing acetic acid from a UREX+ waste stream. The waste stream contains both nitric and acetic acids, and the acetic acid must be removed from the waste stream to prevent potential problems in the downstream steps as well as affecting the recycle of nitric acid. The acetic acid is formed after the UREX step of the process as a result of hydrolytic degradation of acetohydroxamic acid used to suppress plutonium extraction. Of the available technologies, the two most attractive approaches are solvent extraction and distillation. In industry, solvent extraction is used for more dilute concentrations of acetic acid while distillation is used for concentrated acetic acid. If a liquid-liquid extraction is viable, this would be the best option with the addition of an extractant, like tributyl phosphate or tri-n-octyl amine, if needed. However, if acetic acid removal can be delayed until the end of the UREX+ process when the nitric acid may be concentrated for recycle, distillation may remain an option, though not necessarily a better option than solvent extraction.

Evaluation of Annular Centrifugal Contactors for the Extraction of Acetic Acid from Aqueous Streams

This study investigates the use of annular centrifugal contactors for the liquid-liquid extraction of acetic acid from an acidic aqueous phase into an organic phase consisting of 1.5 M Tributyl Phosphate in n-Dodecane. Initial break time tests were performed in order to investigate the mixing/separation viability of the organic/aqueous system, and after determining that centrifugal contactors could be used to perform the liquid-liquid extraction, hydraulic tests established the combination of rotation rate and throughput which should be used to ensure proper separation of the two outlet phases. Finally, extraction efficiency data was collected to examine the system conditions that provided the most efficient removal of acetic acid.

Water and Mean Ionic Activities of Aqueous HNO3 Solutions Calculated from an Extension of the Brunauer-Emmett-Teller (BET) Model

The objective for this research was to employ the extended adsorption isotherm (EAI) to develop a predictive relationship between the activity of H2O (aw) and the mean ionic activity (a±) of aqueous HNO3. The EAI model is a calculative approach to the estimation of solutions for activities of highly non-ideal chemical systems and is the collective efforts of Stokes and Robinson, Abraham, and Ally and Braunstein. From previous works of the HNO3-H2O system by Rains et al., a predictive model of relevant aqueous phase activities was developed that only needs knowledge of composition and temperature of the HNO3 solution.

Thermodynamics of Acetic Acid (aq) Calculated from the Modified Adsorption Isotherm Model for Aqueous Electrolytes

Abstract The modified adsorption isotherm model, incorporating the Stokes-Robinson modification of the Brunauer-Emmett-Teller (BET) adsorption isotherm, is used to calculate the activity of the organic acid and water, and the liquidus curve. This is the first known use of the modified adsorption isotherm for estimation of thermodynamic information for aqueous organic acid solutions and one of few works using the model for predicting information for aqueous acids in general. The method described here represents a simplified approach to estimating thermodynamic properties. The model offers some improvement over regular solution theory, but perhaps more importantly offers a method for adjusting thermodynamic properties in the presence of other components in the future. While other models such as UNIQUAC may prove to be more accurate, this model provides satisfactory results with more ease.

Effects of Sodium Hydroxide and a Chelating Agent on the Removal of Aluminum from Radioactive Sludge

Abstract: An essential step during the remediation of nuclear waste by the U.S. Department of Energy involves the separation of nonradioactive components such as aluminum from high-level waste sludges to minimize the ultimate volume to be stored in a nuclear waste repository. Plans for waste treatment at Hanford and the Savannah River Site include the use of 1 to 3m sodium hydroxide (NaOH) at an elevated temperature to leach the aluminum from the sludge. Triethanolamine (TEA) was added to caustic leaching solutions in an effort to improve the solubility of aluminum from authentic tank-waste sludge. High-level radioactive waste sludge with significant amounts of gibbsite and hard-to-dissolve boehmite phases was used in these tests. In concept, a chelating agent such as TEA can both improve the dissolution rate and increase the aluminum concentration in the liquid phase. However, TEA could also increase the solubility of other sludge components that are potentially problematic to downstream processing. Six tests were performed with leachate concentrations ranging from 0.1 to 3.0m NaOH, 0 to 3.0m TEA, and 0 to 2.9 m NaNO3. One test was performed using 3.0m NaOH at 80°C in order to simulate the baseline process, while the other tests were performed at 60°C. As expected, more aluminum entered the solution at 80 C than at 60°C when other test conditions were held constant. With caustic alone, equilibrium was achieved at both temperatures within 10 days. The addition of TEA significantly increased the concentration of aluminum in the leachate, and the aluminum concentration continued to increase even after 11 days of processing. The fraction of aluminum dissolved at 60°C increased from 35% using 3.0 in NaOH alone to 87% using a combination of 3.0 m NaOH and 3.0m TEA. The high-nitrate, low-hydroxide solutions did not significantly dissolve the aluminum because aluminate ion could not be produced. A small addition of TEA to these low-caustic solutions had no effect on aluminum removal. The use of TEA also increased the solubility of copper, nickel, and iron, which are only minor constituents. The TEA also had a significant effect on the solubility of the radionuclides 137Cs and 60Co. The significant presence of 137Cs in the leachates was expected with and without TEA. The high-nitrate leaches, which were the least effective of the leaching solutions, removed 69% of the 137Cs from the washed sludge, while a combination of 3.0m NaOH and 3.0m TEA removed 96%. Very little 60Co was removed from the sludge except with the use of the 3.0/h NaOH-3.0m TEA solution, which removed 53%. These results indicate that only TEA and 60Co need to be examined for potential chemical and radiological impacts, respectively, on downstream processes.

ORNL experience and perspectives related to processing of thorium and 233U for nuclear fuel

Abstract Thorium-based nuclear fuel cycles have received renewed attention in both research and public circles since about the year 2000. Much of the attention has been focused on nuclear fission energy production that utilizes thorium as a fertile element for producing fissionable 233U for recycle in thermal reactors, fast reactors, or externally driven systems. Lesser attention has been paid to other fuel cycle operations that are necessary for implementation of a sustainable thorium-based fuel cycle such as reprocessing and fabrication of recycle fuels containing 233U. This paper first identifies recent literature that has resulted from the renewed interest in thorium-based fuel cycles. Next, differences in the radiation characteristics of nuclear materials associated with thorium-based and uranium-based fuels are discussed, and the generic implications of the differences to nuclear material processing are identified. Then, experience at Oak Ridge National Laboratory concerning processing of thorium and 233U is described in terms of the processing projects and campaigns undertaken and the facilities in which the processing was implemented. This experience then provides the basis for a generalized discussion of processing nuclear materials associated with thorium-based fuel cycles as compared to uranium-based fuel cycles. This comparative discussion focuses on key out-of-reactor fuel cycle operations: reprocessing of metal-clad oxide and graphite-matrix oxide used nuclear fuels (UNFs) including head-end processing (shearing and dissolution), solvent extraction, product conversion, fuel fabrication, and waste management. It is concluded that the recycle of thorium-based UNF constituents (233U and thorium) is more technically challenging than the recycle of uranium-based UNF constituents (plutonium and uranium) based on the radiation, chemical, and physical characteristics of nuclear materials in thorium-based fuel cycles as compared to uranium-based fuel cycles.

Performance Criteria for Capture and/or Immobilization Technologies - Milestone Report

The capture and subsequent immobilization of the four regulated volatile radionuclides (3H, 14C, 85Kr, and 129I) from the off-gas streams of a used nuclear fuel (UNF) reprocessing facility has been a topic of substantial research interest for the US Department of Energy and its international colleagues. Removal of some or all of these radionuclides (e.g., based upon fuel burnup, fuel type, cooling time) from the plant effluent streams prior to discharge to the environment is required to meet regulations set forth by the US Environmental Protection Agency. Upon removal, the radionuclide, as well as associated sorbents that cannot be cost-effectively regenerated, is destined for conversion to a waste form. Research in separation and capture methodologies has included a wide range of technologies, including liquid caustic scrubbing systems, solid adsorbents, and cryogenic distillation. The studies of waste forms have been correspondingly diverse. In considering the technologies available for future development and implementation of both sorbents and waste forms, it is necessary to identify benchmark measures of performance to evaluate objectively each sorbent system or waste form.

Complete Non-Radioactive Operability Tests for Cladding Hull Chlorination

Non-radioactive operability tests were made to test the metal chlorination reactor and condenser and their accessories using batch chlorinations of non-radioactive cladding samples and to identify optimum operating practices and components that need further modifications prior to installation of the equipment into the hot cell for tests on actual used nuclear fuel (UNF) cladding. The operability tests included (1) modifications to provide the desired heating and reactor temperature profile; and (2) three batch chlorination tests using, respectively, 100, 250, and 500 g of cladding. During the batch chlorinations, metal corrosion of the equipment was assessed, pressurization of the gas inlet was examined and the best method for maintaining solid salt product transfer through the condenser was determined. Also, additional accessing equipment for collection of residual ash and positioning of the unit within the hot cell were identified, designed, and are being fabricated.

Zirconium Recycle Test Equipment for Hot Cell Operations

The equipment components and assembly support work were modified for optimized, remote hot cell operations to complete this milestone. The modifications include installation of a charging door, Swagelok connector for the off-gas line between the reactor and condenser, and slide valve installation to permit attachment/replacement of the product salt collector bottle.