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Results for “Electrorefining”

Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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155 records · Page 9
All sourcesNASA NTRSDOE OSTI
DOE OSTI2018

Analysis and modeling of the equilibrium behaviors of U and Pu in molten LiCl-KCl/Cd system at 500 °C

In the field of pyroprocessing of used nuclear fuels, liquid cadmium is applied in different ways to recover transuranics from molten salt, while effecting some degree of separations between transuranics and lanthanides. In this work, a model of the separation performance between uranium and plutonium across the phase boundary between molten salt and liquid cadmium is presented. The salt is the LiCl-KCl eutectic with solute concentrations of UCl 3 and PuCl 3 . The cadmium is the liquid cadmium phase field from the Cd-U-Pu ternary alloy system. Lastly, the model used to describe the behavior of uranium and plutonium in the salt/cadmium system is applied to describe the behavior of plutonium and lanthanide in the salt/cadmium system.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Uranium exchange kinetics in a molten LiCl-KCl/Cd system at 500 °C

A dissolution experiment is performed in order to examine the kinetics of uranium isotope equilibration in a molten salt and cadmium system. The salt contains a solute concentration of UCl 3 in LiCl-KCl eutectic and the cadmium contained dissolved uranium at 500 °C. A small piece of high enriched uranium is introduced to the cadmium, which perturbs the equilibrium with respect to the uranium enrichment of the salt and cadmium phases. The series of salt and cadmium samples trace how the enrichment of the two phases reach equilibrium (in the absence of a chemical driving force) over the course of approximately 24 h. Here, a model linking dissolution and isotope exchange kinetics is developed to track the evolution of uranium isotope composition. With the assumption of unaccounted salt mass incurred from the dated sample reporting practice, the model explains the trend of the measured uranium isotope composition well.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

History and status of spent fuel treatment at the INL Fuel Conditioning Facility

Experimental Breeder Reactor II (EBR-II) had an illustrious operational history from 1964 to 1994 as one of the first sodium-cooled fast reactors to use sodium-bonded metallic driver fuels and blankets. EBR-II was a testbed for fuel development and reactor design principles. Following shutdown, it was decided to use electrometallurgical methods to treat the spent fuel and blanket materials for final disposition. Here, this paper provides a brief history of U.S. liquid metal cooled reactor development with emphasis on EBR-II, and a technical description of the electrometallurgical methods.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Multi-species electrolyte migration kinetics with directed flow

Here, this paper investigates an electrochemical process designed to control mixing kinetics in liquid electrolytes. The process utilizes an electrochemical cell with anode and cathode chambers connected by a channel without a membrane. This configuration depletes multiple electroactive species in the cathode chamber while selectively replenishing them in the anode chamber. A key application is the removal of impurity species from the cathode chamber electrolyte and the selective replenishment of primary electroactive species in the anode chamber. Aqueous cell experiments were conducted to validate the theoretical model, and simulations were performed for molten salt electrolytes. This approach minimizes waste by reducing the need for additional electrolyte supplies and extends the utilization of anion species, contributing to environmentally sustainable electrochemical materials processing.

Electrochemical Process↗
DOE OSTI2025

Phosphate-Based Approaches for Dechlorination and Treatment of Salt Waste from Electrochemical Processing of Used Nuclear Fuel: A Perspective on Recent Work

Phosphate-based reagents are being considered by the U.S. Department of Energy (DOE) Office of Nuclear Energy to process halide salt-based nuclear wastes for stabilization prior to disposal. As evidenced by the Experimental Breeder Reactor-II (EBR-II) project, electrochemical processing (pyroprocessing) can be employed to recover uranium and other actinides for reintegration into the nuclear fuel cycle from metallic fuels. The resultant salt-based wastes generated from electrochemical processing of EBR-II fuel contains fission products within a LiCl–KCl eutectic salt that necessitate appropriate disposal. This paper provides an overview of recent efforts to support halide-based salt waste treatment for disposition, as well as a basis for comparison with other related efforts in salt waste treatment through salt partitioning initiatives. The U.S. DOE has selected a phosphate waste form reference material for further investigation and longer-term studies.

electrorefiner↗
DOE OSTI2020

Electrochemical Separation of Alkaline-Earth Elements from Molten Salts Using Liquid Metal Electrodes

Closing the nuclear fuel cycle requires recycling used nuclear fuel. Additional waste is generated during recycling due to fission products accumulating in processing salts (LiCl-KCl). Reducing waste generated during recycling entails recovering alkaline-earth fission products (Ba 2+ /Sr 2+ ) from molten chlorides with a minimal loss of bulk electrolyte constituents (Li + /K + ). Electrochemical co-deposition of Ba 2+ /Li + and Sr 2+ /Li + into liquid metal (Bi, Sb, Sn, Pb) and alloy (Bi-Sb) electrodes was investigated in LiCl-KCl-(BaCl 2 , SrCl 2 ) electrolytes at 500 °C and 650 °C. For the pure Bi (500 °C) and Sb (650 °C) electrodes, the greatest percentage of charge was used to deposit Ba and Sr. Effective recovery of Ba/Sr by liquid Bi and Sb electrodes is supported via experimentally determined activity values of Ba/Sr in Bi and Sb. Alloying Sb with Bi increased Ba recovery but decreased Sr recovery, as compared to recovery using the liquid Bi electrode. Here, the results suggest that alkaline-earth fission products can be recovered from molten chlorides by liquid metal electrodes via electrochemical separation, thereby providing a methodology to reduce the generation of nuclear waste from nuclear fuel recycling.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Discrete-event Simulation Process Model for the Pyrochemical Processing of Plutonium at Los Alamos National Laboratory

The pyrochemical metal production operations that occur in the Plutonium Facility at Los Alamos National Laboratory perform plutonium purification with the aim to provide plutonium metal for a variety of defense- and non-defense missions within the National Nuclear Security Administration. The demands and constraints associated with the pyrochemical processing of plutonium are complex, making decision analyses challenging for program managers who require plutonium production for their mission applications. The construction of a discrete-event simulation process model is proposed to measure and report the process capacity, material throughput, equipment requirements, and dose accumulation for operators of the pyrochemical metal production operations. The process model, constructed in the ExtendSim™ software, will represent the cause-and-effect relationships between the pyrochemical processing environment and the process constraints, including criticality limitations, material control and accountability measures, chemical analysis requirements, and equipment availability. An accurate representation of the pyrochemical metal production process capacity through simulation modeling will be helpful to program managers in their efforts to forecast plutonium availability for mission applications. Furthermore, the proposed process model will be vital for future analyses that will measure the interactions between the pyrochemical metal production operations and the aqueous reprocessing operations and their ability to minimize transuranic waste disposal.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Evaluating High-Halide Waste Form Options for Salt-Based Nuclear Waste Simulants

In this study, waste forms were being explored for an electrochemical salt simulant, , referred to as ERV3, which is a high-LiCl/KCl salt containing simulated fission products (i.e., Sr, Cs, Nd) and Na to represent bond sodium from Experimental Breeder Reactor-II metallic fast reactor fuel. The goal was to find glassy systems that could be used to immobilize the salt in a single-step process. The envisionment of this process would be find a frit glass that could be added to the salt waste, heat treated, poured into waste canisters, and then stored for disposal. To perform this study, a literature review was conducted, the most promising seven systems were fabricated without the salt, and then mixed with salt simulant and heat treated under different processes. The criteria that were used to screen potential compositions included demonstrated alkali incorporation, could be melted at reasonably low temperature ( T ≤ 1000°C), and if the compositions had some demonstrated data for waste-form-related properties, that was a benefit. High marks were given for compositions that showed amorphous nature after slow cooling of samples containing ERV3.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗