Supathorn Phongikaroon

Electrolytic Reduction of Spent Nuclear Oxide Fuel as Part of an Integral Process to Separate and Recover Actinides from Fission Products

Abstract Bench‐scale tests were performed to study an electrolytic reduction process that converts metal oxides in spent nuclear fuel to metal. Crushed spent oxide fuel was loaded into a permeable stainless steel basket and submerged in a molten salt electrolyte of LiCl–1 wt% Li2O at 650°C. An electrical current was passed through the fuel basket and a submerged platinum wire, effecting the reduction of metal oxides in the fuel and the formation of oxygen gas on the platinum wire surface. Salt and fuel samples were analyzed, and the extent of fission product separation and metal oxide reduction was determined.

Development of Computational Models for the Mark-IV Electrorefiner—Effect of Uranium, Plutonium, and Zirconium Dissolution at the Fuel Basket-Salt Interface

Abstract The electrochemical processing of spent metallic nuclear fuel has been demonstrated by and is currently in operation at the Idaho National Laboratory (INL). At the heart of this process is the Mark-IV electrorefiner (ER). This process involves the anodic dissolution of spent nuclear fuel into a molten salt electrolyte along with a simultaneous deposition of pure uranium on a solid cathode. This allows the fission products to be separated from the fuel and processed into an engineered waste form. A one-dimensional model of the Mark-IV ER has begun to be developed. The computations thus far have modeled the dissolution of the spent nuclear fuel at the anode taking into account uranium (U3+), plutonium (Pu3+), and zirconium (Zr4+). Uranium and plutonium are the two most important elements in the system, whereas zirconium is the most active of the noble metals. The model shows that plutonium is quickly exhausted from the anode, followed by dissolution of primarily uranium, along with small amounts of zirconium. The total anode potential as calculated by the model has been compared to experimental data sets provided by INL. The anode potential has been shown to match the experimental values quite well with root-mean-square (rms) values of 2.27 and 3.83% for two different data sets, where rms values closer to zero denote better fit.

A Computational Model of the Mark-IV Electrorefiner: Phase I―Fuel Basket/Salt Interface

Spent driver fuel from the Experimental Breeder Reactor-II (EBR-II) is currently being treated in the Mk-IV electrorefiner (ER) in the Fuel Conditioning Facility (FCF) at Idaho National Laboratory. The modeling approach to be presented here has been developed to help understand the effect of different parameters on the dynamics of this system. The first phase of this new modeling approach focuses on the fuel basket/salt interface involving the transport of various species found in the driver fuels (e.g. uranium and zirconium). This approach minimizes the guessed parameters to only one, the exchange current density (i0). U3+ and Zr4+ were the only species used for the current study. The result reveals that most of the total cell current is used for the oxidation of uranium, with little being used by zirconium. The dimensionless approach shows that the total potential is a strong function of i0 and a weak function of wt% of uranium in the salt system for initiation processes.

Diffusion Model for Electrolytic Reduction of Uranium Oxides in a Molten LiCl-Li2O Salt

Abstract In this study, a diffusion-based kinetic model essential for design and operational analysis of spent nuclear fuel reduction has been developed. The model considers the cathode side of the system to be rate limiting and deals with diffusion of lithium metal through the basket loaded with uranium oxide (UO2 or U3O8). Faraday’s law was implemented into the model to observe the electrochemical effect on the model. Solutions with different conditions are developed, and detailed results are presented. These solutions were compared against experimental bench scale data. At high operating current conditions (I > 0.8 A), the model fits the data well. The fitting resulted in estimated effective lithium diffusion coefficients for high and low void fraction UO2 crushed fuels of 8.5 × 10−4 cm2/s and 2.2 × 10−4 cm2/s, respectively. The effective diffusion coefficient for U3O8 is estimated to be 8.6 × 10−4 cm2/s. In some experiments, a porous magnesium oxide basket was used for containing the U3O8. It was estimated that the lithium diffusion coefficient through this magnesia basket is 3.3 × 10−5 cm2/s.

COMPUTATIONAL MODEL OF THE MARK-IV ELECTROREFINER: TWO-DIMENSIONAL POTENTIAL AND CURRENT DISTRIBUTIONS

Abstract A computational model of the Mark-IV electrorefiner is currently being developed as a joint project between Idaho National Laboratory, Korea Atomic Energy Research Institute, Seoul National University, and the University of Idaho. As part of this model, the two-dimensional potential and current distributions within the molten salt electrolyte are calculated for U3+, Zr4+, and Pu3+ along with the total distributions, using the partial differential equation solver of the commercial Matlab software. The electrical conductivity of the electrolyte solution is shown to depend primarily on the composition of the electrolyte and to average 205 mho/m with a standard deviation of 2.5 × 10–5% throughout the electrorefining process. These distributions show that the highest potential gradients (thus, the highest current) exist directly between the two anodes and cathode. The total, uranium, and plutonium potential gradients are shown to increase throughout the process, with a slight decrease in that of zirconium. The distributions also show small potential gradients and very little current flow in the region far from the operating electrodes.

Selective Reduction of Active Metal Chlorides from Molten LiCl-KCl using Lithium Drawdown

In support of optimizing electrorefining technology for treating spent nuclear fuel, lithium drawdown has been investigated for separating actinides from molten salt electrolyte. Drawdown reaction selectivity is a major issue that requires investigation, since the goal is to remove actinides while leaving the fission products and other components in the salt. A series of lithium drawdown tests with surrogate fission product chlorides was run to obtain selectivity data with non-radioactive salts, develop a predictive model, and draw conclusions about the viability of using this process with actinide-loaded salt. Results of tests with CsCl, LaCl3, CeCl3, and NdCl3 are reported here. Equilibrium was typically achieved in less than 10 hours of contact between lithium metal and molten salt under well-stirred conditions. Maintaining low oxygen and water impurity concentrations (<10 ppm) in the atmosphere was observed to be critical to minimize side reactions and maintain stable salt compositions. An equilibrium model has been formulated and fit to the experimental data. Good fits to the data were achieved. Based on analysis and results obtained to date, it is concluded that clean separation between minor actinides and lanthanides will be difficult to achieve using lithium drawdown.

SALT-ZEOLITE ION-EXCHANGE EQUILIBRIUM STUDIES FOR A COMPLETE SET OF FISSION PRODUCTS IN MOLTEN LiCl-KCl

Abstract This paper presents results of experiments and modeling for ion exchange of LiCl-KCl-based molten salts with zeolite-A. The experiments examined the equilibrium distributions of various nuclear fuel fission products between the molten salt and zeolite phases. In addition to data that were collected in previous studies, new experiments were run using ternary salts (LiCl-KCl-YCl3, LiCl-KCl-LaCl3, and LiCl-KCl-PrCl3) and quaternary salts (LiCl-KCl-CsCl-NdCl3 and LiCl-KCl-CsCl-SrCl2). All contacting experiments were conducted at 500°C with a salt-zeolite contacting period of 24 h to allow for equilibrium to be reached. The developed equilibrium model assumes that there are ion-exchange and occlusion sites, both of which are in equilibrium with the molten salt phase. A systematic approach in estimating the total occlusion capacity of the zeolite-A was developed. The parameters of the model, including the total occlusion capacity of the zeolite-A, were determined from fitting the entire set of experimental data available between previous studies and the current one. Experiments involving ternary salts were used to estimate the parameters of the model, while those involving quaternary salts were used to validate the model.

Laser-Induced Breakdown Spectroscopy (LIBS) in a Novel Molten Salt Aerosol System

In the pyrochemical separation of used nuclear fuel (UNF), fission product, rare earth, and actinide chlorides accumulate in the molten salt electrolyte over time. Measuring this salt composition in near real-time is advantageous for operational efficiency, material accountability, and nuclear safeguards. Laser-induced breakdown spectroscopy (LIBS) has been proposed and demonstrated as a potential analytical approach for molten LiCl–KCl salts. However, all the studies conducted to date have used a static surface approach which can lead to issues with splashing, low repeatability, and poor sample homogeneity. In this initial study, a novel molten salt aerosol approach has been developed and explored to measure the composition of the salt via LIBS. The functionality of the system has been demonstrated as well as a basic optimization of the laser energy and nebulizer gas pressure used. Initial results have shown that this molten salt aerosol–LIBS system has a great potential as an analytical technique for measuring the molten salt electrolyte used in this UNF reprocessing technology.

SEPARATION OF CsCl FROM LiCl-CsCl MOLTEN SALT BY COLD FINGER MELT CRYSTALLIZATION

This study provides a fundamental understanding of a cold finger melt crystallization technique by exploring the heat and mass transfer processes of cold finger separation. A series of experiments were performed using a simplified LiCl-CsCl system by varying initial CsCl concentrations (1, 3, 5, and 7.5 wt%), cold finger cooling rates (7.4, 9.8, 12.3, and 14.9 L/min), and separation times (5, 10, 15, and 30 min). Results showed a potential recycling rate of 0.36 g/min with a purity of 0.33 wt% CsCl in LiCl. A CsCl concentrated drip formation was found to decrease crystal purity especially for smaller crystal formations. Dimensionless heat and mass transfer correlations showed that separation production is primarily influenced by convective transfer controlled by cooling gas flow rate, where correlations are more accurate for slower cooling gas flow rates.

Accelerator-driven subcritical fission in molten salt core: Closing the nuclear fuel cycle for green nuclear energy

A technology for accelerator-driven subcritical fission in a molten salt core (ADSMS) is being developed as a basis for the destruction of the transuranics in used nuclear fuel. The molten salt fuel is a eutectic mixture of NaCl and the chlorides of the transuranics and fission products. The core is driven by proton beams from a strong-focusing cyclotron stack. This approach uniquely provides an intrinsically safe means to drive a core fueled only with transuranics, thereby eliminating competing breeding terms.