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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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At least 55 records · Page 3
All sourcesNASA NTRSDOE OSTI
DOE OSTI2013

Development of high temperature transport technology for LiCl-KCl eutectic salt in pyroprocessing

The development of high-temperature transport technologies for molten salt is a prerequisite and a key issue in the industrialization of pyro-reprocessing for advanced fuel cycle scenarios. The solution of a molten salt centrifugal pump was discarded because of the high corrosion power of a high temperature molten salt, so the suction pump solution was selected. An apparatus for salt transport experiments by suction was designed and tested using LiC-KCl eutectic salt. The experimental results of lab-scale molten salt transport by suction showed a 99.5% transport rate (ratio of transported salt to total salt) under a vacuum range of 100 mtorr - 10 torr at 500 Celsius degrees. The suction system has been integrated to the PRIDE (pyroprocessing integrated inactive demonstration) facility that is a demonstrator using non-irradiated materials (natural uranium and surrogate materials). The performance of the suction pump for the transport of molten salts has been confirmed.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Phase Equilibria in Systems Relevant to Pyroprocessing

Pyroprocessing has many advantages for the recovery and recycle of actinides, especially the transuranic elements, in used nuclear fuel. It has several unique features that distinguish it from other used fuel treatment technologies such as the ability to handle short-cooled fuel, a small process footprint, and production of metallic uranium and uranium - transuranic (U/TRU) products that can be directly fabricated into metallic fuel for advanced reactors. Recovery of the uranium and U/TRU products from the used fuel occurs in an electrorefining step. Uranium and transuranics in the used fuel (metal or reduced oxide) are dissolved at the anode in a molten chloride electrolyte, and electrodeposited at two different cathodes - one for uranium and another for the uranium - transuranic product. During this process, fission products less noble than uranium and transuranics (i.e., lanthanides, alkali and alkaline earth) anodically dissolve into the electrolyte, while only uranium and U/TRU are electrodeposited at the respective cathodes. Thus, there is a buildup of active metal and lanthanide fission products in the molten salt electrolyte. The goal of this work is to improve the fundamental understanding of the chemistry of actinides and transuranics in this complex, multicomponent solution. The fundamental thermodynamic properties of the multi-component molten salt electrolytes are complicated due to the formation of ternary and more complex compounds. As the actinides anodically dissolve at the anode and deposited at the cathode, a gradient in the concentration develops at the anode/electrolyte and the cathode/electrolyte interfaces. Ternary phase formation could impact the performance of the electrorefining system by influencing the chemical activity and thus the redox potentials of the actinide chlorides in the molten salt solution. Literature data on the LiCl-KCl-UCl{sub 3} system is limited, sometimes conflicting, and may reflect the use of older analytical instrumentation, which was less accurate than that currently available. In addition, there has not been any investigation of the low UCl{sub 3} concentration region of the system, which is the region relevant to electrorefining. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Development of a Discrete Event Simulation Model for Pyroprocessing Safeguardability

Advanced nuclear energy systems offer benefits for energy sustainability, independence, security, economic growth, and carbon emission reduction. Nuclear fuel fabricated for these systems utilize remotely-handled facilities for batch processing in hot cells. Because special nuclear material is in both different physical and chemical form than that of the contemporary fuel cycle, new challenges arise in terms of safeguardability. To that end, we are developing a system model for a commercial pyroprocessing facility that can also produce quantitative metrics to assess facility design. This paper presents the progress of a first build of a material throughput model using discrete event simulation for the fuel fabrication subsystem. The intent is to study the Type I error, or false alarm, as a metric for safeguardability. Initiating events that could cause false alarms arise from safeguards, safety, and security. False alarms can necessitate a facility shutdown until the anomaly is resolved by facility inspection. A high number would be costly for the facility operator both in terms of operation throughput goals and possible compensatory penalties. Therefore, different conceptual facility designs exhibit different frequencies for the false alarm per initiating event, and this can provide a measure of the safeguardability of each design. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

Investigation of Fission Product Transport into Zeolite-A for Pyroprocessing Waste Minimization

Methods to improve fission product salt sorption into zeolite-A have been investigated in an effort to reduce waste associated with the electrochemical treatment of spent nuclear fuel. It was demonstrated that individual fission product chloride salts were absorbed by zeolite-A in a solid-state process. As a result, recycling of LiCl-KCl appears feasible via adding a zone-freezing technique to the current treatment process. Ternary salt molten-state experiments showed the limiting kinetics of CsCl and SrCl2 sorption into the zeolite. CsCl sorption occurred rapidly relative to SrCl2 with no observed dependence on zeolite particle size, while SrCl2 sorption was highly dependent on particle size. The application of experimental data to a developed reaction-diffusion-based sorption model yielded diffusivities of 8.04 × 10-6 and 4.04 × 10-7 cm2 /s for CsCl and SrCl2, respectively. Additionally, the chemical reaction term in the developed model was found to be insignificant compared to the diffusion term.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Overview of Pyroprocessing

This is a PPT presentation on the spent fuel treatment operations in the Fuel Conditioning Facility at MFC.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

Separation of CsCl from a Ternary CsCl-LiCl-KCl Salt via a Melt Crystallization Technique for Pyroprocessing Waste Minimization

A parametric study has been conducted to identify the effects of several parameters on the separation of CsCl from molten LiCl-KCl salt via a melt crystallization process. A reverse vertical Bridgman technique was used to grow the salt crystals. The investigated parameters were: (1) the advancement rate, (2) the crucible lid configuration, (3) the amount of salt mixture, (4) the initial composition of CsCl, and (5) the temperature difference between the high and low furnace zones. From each grown crystal, samples were taken axially and analyzed using inductively coupled plasma mass spectrometry (ICP-MS). Results show that CsCl concentrations at the top of the crystals were low and increased to a maximum at the bottom of the salt. Salt (LiCl-KCl) recycle percentages for the experiments ranged from 50% to 75% and the CsCl composition in the waste salt was low. To increase the recycle percentage and the concentration of CsCl in the waste form, the possibility of using multiple crystallization stages was explored to further optimize the process. Results show that multiple crystallization stages are practical and the optimal experimental conditions should be operated at 5.0 mm/hr rate with a lid configuration and temperature difference of 200 °C for a total of five crystallization stages. Under these conditions, up to 88% of the salt can be recycled.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

An Assessment of Applying Pyroprocessing Technology to Advanced Pebble-Type Fuels

With an expected increased demand for high-assay low enriched uranium (HALEU) to supply advanced reactors, the possibility of recovering actinides from various used nuclear fuels (UNF) is being reassessed. Pebble-type fuels such as tristructural isotropic (TRISO) fuel are intended to be directly disposed after use, but the HALEU remaining in each pebble is a resource with increasing value. An economically feasible recovery strategy would maximize the amount of HALEU recovered for reuse in new fuel without increasing the waste volume relative to the direct disposal of used TRISO fuel. We are evaluating the technical feasibility of recycling TRISO or pebble fuels to recover actinides for reuse as new fuel for advanced reactors by using pyrochemical methods.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

An Assessment of Applying Pyroprocessing Technology to Advanced Pebble-Type Fuels

With an expected increased demand for high-assay low enriched uranium (HALEU) to supply advanced reactors, the possibility of recovering actinides from various used nuclear fuels (UNF) is being re-assessed. Pebble-type fuels such as tristructural isotropic (TRISO) fuel are intended to be directly disposed after use, but the HALEU remaining in each pebble is a resource with increasing value. An economically feasible recovery strategy would maximize the amount of HALEU recovered for reuse in new fuel without increasing the waste volume relative to the direct disposal of used TRISO fuel. We are evaluating the technical feasibility of recycling TRISO or pebble fuels to recover actinides for reuse as new fuel for advanced reactors by using pyrochemical methods.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

Pyroprocessing of Oxidized Sodium-Bonded Fast Reactor Fuel -- an Experimental Study of Treatment Options for Degraded EBR-II Fuel

An experimental study was conducted to assess pyrochemical treatment options for degraded EBR-II fuel. As oxidized material, the degraded fuel would need to be converted back to metal to enable electrorefining within an existing electrometallurgical treatment process. A lithium-based electrolytic reduction process was studied to assess the efficacy of converting oxide materials to metal with a particular focus on the impact of zirconium oxide and sodium oxide on this process. Bench-scale electrolytic reduction experiments were performed in LiCl-Li2O at 650 °C with combinations of manganese oxide (used as a surrogate for uranium oxide), zirconium oxide, and sodium oxide. The experimental study illustrated how zirconium oxide and sodium oxide present different challenges to a lithium-based electrolytic reduction system for conversion of select metal oxides to metal.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2014

Actinide ion sensor for pyroprocess monitoring

An apparatus for real-time, in-situ monitoring of actinide ion concentrations which comprises a working electrode, a reference electrode, a container, a working electrolyte, a separator, a reference electrolyte, and a voltmeter. The container holds the working electrolyte. The voltmeter is electrically connected to the working electrode and the reference electrode and measures the voltage between those electrodes. The working electrode contacts the working electrolyte. The working electrolyte comprises an actinide ion of interest. The reference electrode contacts the reference electrolyte. The reference electrolyte is separated from the working electrolyte by the separator. The separator contacts both the working electrolyte and the reference electrolyte. The separator is ionically conductive to the actinide ion of interest. The reference electrolyte comprises a known concentration of the actinide ion of interest. The separator comprises a beta double prime alumina exchanged with the actinide ion of interest.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗