Brian R. Westphal

On the Development of a Distillation Process for the Electrometallurgical Treatment of Irradiated Spent Nuclear Fuel

As part of the spent fuel treatment program at the Idaho National Laboratory, a vacuum distillation process is being employed for the recovery of actinide products following an electrorefining process. Separation of the actinide products from a molten salt electrolyte and cadmium is achieved by a batch operation called cathode processing. A cathode processor has been designed and developed to efficiently remove the process chemicals and consolidate the actinide products for further processing. This paper describes the fundamentals of cathode processing, the evolution of the equipment design, the operation and efficiency of the equipment, and recent developments at the cathode processor. In addition, challenges encountered during the processing of irradiated spent nuclear fuel in the cathode processor will be discussed.

Engineering-Scale Liquid Cadmium Cathode Experiments

Abstract Recovery of uranium and transuranic (TRU) actinides from spent nuclear fuel by an electrorefining process was investigated as part of the U.S. Department of Energy Advanced Fuel Cycle Initiative. Experiments were performed in a shielded hot cell at the Materials and Fuels Complex at Idaho National Laboratory. The goal of these experiments was to collect, by an electrochemical process, kilogram quantities of uranium and plutonium into what is called a liquid cadmium cathode (LCC). For each experiment, a steel basket loaded with chopped spent nuclear fuel from the Experimental Breeder Reactor II acted as the anode in the electrorefiner. The cathode was a beryllium oxide crucible containing ~26 kg of cadmium metal (the LCC). In the three experiments performed to date, between 1 and 2 kg of heavy metal was collected in the LCC after passing an integrated current between 1.80 and 2.16 MC (500 and 600 A h) from the anode to the cathode. Sample analysis of the processed LCC ingots measured detectable amounts of TRUs and rare earth elements.

Separation and Recovery of Uranium and Group Actinide Products From Irradiated Fast Reactor MOX Fuel via Electrolytic Reduction and Electrorefining

A series of bench-scale tests was conducted with irradiated fast reactor MOX fuel to separate and recover refined uranium and group actinide products via electrolytic reduction and electrorefining. The fuel was declad, crushed, immersed in a pool of molten LiCl −1 wt% Li2O at 650°C, and electrolyzed to convert the mixed oxide fuel to metal. The reduced fuel was then electrorefined in LiCl-KCl-UCl3 at 500°C, yielding a refined uranium metal product. Additional electrorefining experiments were performed in which actinides (that is, uranium, neptunium, plutonium, and americium) were recovered as a group metal product.

Selective Trapping of Volatile Fission Products with an Off-Gas Treatment System

Abstract An off-gas treatment system is being developed for the collection of volatile fission products during a head-end processing step. The head-end processing step employs high temperatures to oxidize UO2 to U3O8 resulting in the separation of fuel from cladding and the removal of volatile fission products. Three volatile fission products have been targeted for trapping on distinct zones of the off-gas system and within those zones, on individual filters. A description of the filter media and a basis for its selection will be given along with the collection mechanisms and design considerations. Results from testing with the off-gas treatment system will also be presented.

INTEGRATED EFFICIENCY TEST FOR PYROCHEMICAL FUEL CYCLES

Abstract An engineering-scale pyroprocessing integrated efficiency test was conducted with sodium-bonded, spent Experimental Breeder Reactor II drive fuel elements. The major pieces of equipment used to conduct the test were the element chopper, Mk-IV electrorefiner, cathode processor, and casting furnace. Four batches of the spent fuel (containing 50.4-kg heavy metal) were processed under a set of fixed operating parameters. The primary goal of the test was to demonstrate the actinide dissolution and recovery efficiencies typical of the fixed operating parameters that have been developed for this equipment based on over a decade’s worth of processing experience. The total mass balance for the test was 101.28% (slightly more output than input). The uranium mass balance for the test was 100.13%. The test results indicate that 99.3 wt% of uranium in the feed was electrochemically dissolved and 98.4 wt% of the uranium was collected as metal ingots. The complexity of zirconium behavior during electrorefining was confirmed by the test results. More than 85 wt% of the zirconium was electrochemically dissolved during the later stages of the electrorefining process. However, only 33.7 wt% of the zirconium was collected as metal in the ingots. The balance of the zirconium is believed to reside in the cadmium pool. The test also identified that the dross streams from the cathode processor and casting furnace account for ˜2.4 wt% of the uranium relative to the feed.

Separation Characteristics of Manganese as a Surrogate for Americium during the Distillation Operations of Pyroprocessing

The loss of americium metal during the pyroprocessing of used nuclear fuel has long been a concern due to its high vapor pressure relative to the operating conditions of the process. Of the two high temperature vacuum operations (distillation and casting) performed during pyroprocessing, the distillation operation would incur significantly more americium losses by several orders of magnitude. The distillation operation is required for the removal of cadmium from the transuranic products. Thus, a series of tests were initiated to investigate the evaporative characteristics of manganese as a surrogate for americium during distillation operations. The results for the separation of manganese are presented from the test program and compared against modeled data. Based on the modeling of manganese, similar data were calculated for americium evaporation during a typical liquid cathode operation in the cathode processor. It is anticipated that less than 0.15 wt.% of the americium will be lost during the distillation operation of liquid cathodes during the pyroprocessing of used nuclear fuel.

Second Generation Experimental Equipment Design to Support Voloxidation Testing at INL

Voloxidation is a potential head-end process used prior to aqueous or pyrochemical spent-oxide-fuel treatment. The spent oxide fuel is heated to an elevated temperature in oxygen or air to promote separation of the fuel from the cladding as well as volatize the fission products. The Idaho National Laboratory (INL) and the Korea Atomic Energy Research Institute (KAERI) have been collaborating on voloxidation research through a joint International Nuclear Energy Research Initiative (I-NERI). A new furnace and off-gas trapping system (OTS) with enhanced capability was necessary to perform further testing. The design criteria for the OTS were jointly agreed upon by INL and KAERI. First, the equipment must accommodate the use of spent nuclear fuel and be capable of operating in the Hot Fuel Examination Facility (HFEF) at the INL. This primarily means the furnace and OTS must be remotely operational and maintainable. The system requires special filters and distinctive temperature zones so that the fission products can be uniquely captured. The OTS must be sealed to maximize the amount of fission products captured. Finally, to accommodate the largest range of operating conditions, the OTS must be capable of handling high temperatures and various oxidizing environments. The constructed system utilizes a vertical split-tube furnace with four independently controlled zones. One zone is capable of reaching 1200°C to promote the release of volatile fission products. The three additional zones that capture fission products can be controlled to operate between 100-1100°C. A detailed description of the OTS will be presented as well as some initial background information on high temperature seal options.