Search NASA⌕ Search

SEARCH · Search NASA

Results for “Pyroprocessing”

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.

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 181 records · Page 10
All sourcesNASA NTRSDOE OSTI
DOE OSTI2012

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 needs to be investigated, since the goal is to remove actinides while leaving the fission products 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-loadd salt. Results of tests with CsCl, LaCl3, CeCl3, and NdCl3 are reported here. An equilibrium model has been formulated and fit to the experimental data. Excellent 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.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Water Sorption/Desorption Characteristics of Eutectic LiCl-KCl Salt-Occluded Zeolites

Molten salt consisting primarily of eutectic LiCl-KCl is currently being used in electrorefiners in the Fuel Conditioning Facility at Idaho National Laboratory. Options are currently being evaluated for storing this salt outside of the argon atmosphere hot cell. The hygroscopic nature of eutectic LiCl-KCl makes is susceptible to deliquescence in air followed by extreme corrosion of metallic cannisters. In this study, the effect of occluding the salt into a zeolite on water sorption/desorption was tested. Two zeolites were investigated: Na-Y and zeolite 4A. Na-Y was ineffective at occluding a high percentage of the salt at either 10 or 20wt% loading. Zeolite-4A was effective at occluding the salt with high efficiency at both loading levels. Weight gain in salt occluded zeolite-4A (SOZ) from water sorption at 20% relative humidity and 40°C was 17wt% for 10% SOZ and 10wt% for 20% SOZ. In both cases, neither deliquescence nor corrosion occurred over a period of 31 days. After hydration, most of the water could be driven off by heating the hydrated salt occluded zeolite to 530°C. However, some HCl forms during dehydration due to salt hydrolysis. Over a wide range of temperatures (320–700°C) and ramp rates (5, 10, and 20°C min -1 ), HCl formation was no more than 0.6% of the Cl - in the original salt.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

Potentiometric Sensor for Real-Time Remote Surveillance of Actinides in Molten Salts

A potentiometric sensor is being developed at the Idaho National Laboratory for real-time remote surveillance of actinides during electrorefining of spent nuclear fuel. During electrorefining, fuel in metallic form is oxidized at the anode while refined uranium metal is reduced at the cathode in a high temperature electrochemical cell containing LiCl-KCl-UCl3 electrolyte. Actinides present in the fuel chemically react with UCl3 and form stable metal chlorides that accumulate in the electrolyte. This sensor will be used for process control and safeguarding of activities in the electrorefiner by monitoring the concentrations of actinides in the electrolyte. The work presented focuses on developing a solid-state cation conducting ceramic sensor for detecting varying concentrations of trivalent actinide metal cations in eutectic LiCl-KCl molten salt. To understand the basic mechanisms for actinide sensor applications in molten salts, gadolinium was used as a surrogate for actinides. The ß?-Al2O3 was selected as the solid-state electrolyte for sensor fabrication based on cationic conductivity and other factors. In the present work Gd3+-ß?-Al2O3 was prepared by ion exchange reactions between trivalent Gd3+ from GdCl3 and K+-, Na+-, and Sr2+-ß?-Al2O3 precursors. Scanning electron microscopy (SEM) was used for characterization of Gd3+-ß?-Al2O3 samples. Microfocus X-ray Diffraction (µ-XRD) was used in conjunction with SEM energy dispersive X-ray spectroscopy (EDS) to identify phase content and elemental composition. The Gd3+-ß?-Al2O3 materials were tested for mechanical and chemical stability by exposing them to molten LiCl-KCl based salts. The effect of annealing on the exchanged material was studied to determine improvements in material integrity post ion exchange. The stability of the ß?-Al2O3 phase after annealing was verified by µ-XRD. Preliminary sensor tests with different assembly designs will also be presented.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

Measurements and Analysis of Oxygen Bubble Distributions in LiCl-KCl Molten Salt

Transparent system experimental studies have been performed to provide measurement and analysis of oxygen bubble distributions and mass transfer coefficients at different sparging rates ranging from 0.05 to 0.20 L/min in LiCl-KCl molten salt at 500 degrees C using a high-speed digital camera and an oxygen sensor. The results reveal that bubble sizes and rise velocities increased with an increase in oxygen sparging rate. The bubbles observed were ellipsoidal in shape, and an equivalent diameter based on the ellipsoid volume was calculated. The average equivalent bubble diameters at 500 degrees C and these oxygen sparging rates range from 2.63 to 4.07 mm. Results show that the bubble equivalent diameters at each respective sparging rate are normally distributed. A Fanning friction factor correlation was produced to predict a bubble’s rise velocity based on its equivalent diameter. The oxygen mass transfer coefficients for four sparging rates were calculated using the oxygenation model. These calculated values were within the order of magnitude of 10-2 cm/sec and followed a decreasing trend corresponding to an increasing bubble size and sparging rate. The diffusivities were calculated based on two different types of mechanisms, one based on physics of the bubbles and the other on systematic properties. The results reveal that diffusivity values calculated from bubble physics are 1.65 to 8.40 x 10-5 cm2/sec, which are within the range suggested by literature for gases in liquids of a similar viscosity.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

Management of Salt Waste from Electrochemical Processing of Used Nuclear Fuel

Electrochemical processing of used nuclear fuel involves operation of one or more cells containing molten salt electrolyte. Processing of the fuel results in contamination of the salt via accumulation of fission products and transuranic (TRU) actinides. Upon reaching contamination limits, the salt must be removed and either disposed or treated to remove the contaminants and recycled back to the process. During development of the Experimental Breeder Reactor-II spent fuel treatment process, waste salt from the electrorefiner was to be stabilized in a ceramic waste form and disposed of in a high-level waste repository. With the cancellation of the Yucca Mountain high-level waste repository, other options are now being considered. One approach that involves direct disposal of the salt in a geologic salt formation has been evaluated. While waste forms such as the ceramic provide near-term resistance to corrosion, they may not be necessary to ensure adequate performance of the repository. To improve the feasibility of direct disposal, recycling a substantial fraction of the useful salt back to the process equipment could minimize the volume of the waste. Experiments have been run in which a cold finger is used for this purpose to crystallize LiCl from LiCl/CsCl. If it is found to be unsuitable for transportation, the salt waste could also be immobilized in zeolite without conversion to the ceramic waste form.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

Analysis of Cadmium in Undissolved Anode Materials of Mark-IV Electrorefiner

The Mark-IV electrorefiner (Mk-IV ER) contains an electrolyte/molten cadmium system for refining uranium electrochemically. Typically, the anode of the Mk-IV ER consists of the chopped sodium-bonded metallic driver fuels, which have been primarily U-10Zr binary fuels. Chemical analysis of the residual anode materials after electrorefining indicates that a small amount of cadmium is removed from the Mk-IV ER along with the undissolved anode materials. Investigation of chemical analysis data indicates that the amount of cadmium in the undissolved anode materials is strongly correlated with the anode rotation speeds and the residence time of the anode in the Mk-IV ER. Discussions are given to explain the prescribed correlation.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Experience of Graphite Crucibles during the Pyrometallurgical Processing of Uranium

Throughout the history of processing molten uranium, various materials have been investigated and utilized for the containment of uranium at temperatures greater than 1135oC. Of these, graphite is the most prevalent material due to its ease of fabrication, expense, compatibility with induction systems, and robust characteristics. Graphite in contact with molten uranium will react to form uranium carbide species, thus a coating or mold is needed to prevent reaction of the uranium and the subsequent adherence of the solidified metal. During the pyrometallurgical treatment of used nuclear fuel for uranium recovery at the Idaho National Laboratory, two operational steps utilize graphite crucibles for containment of molten uranium. The cathode processor step separates molten salt from the uranium while the casting step blends uranium for de-enrichment and allows molten sampling of the uranium. Zirconia and yttria based materials on graphite are employed for the cathode processor and casting steps, respectively. For the cathode processor, zirconia coatings were applied initially to graphite crucibles but more recently a zirconia castable material has been molded into the graphite. Although not perfect, zirconia has been found to be compatible with both molten salt and uranium whereas yttria will react with molten salts if carried over to the casting operation. With development, the uranium losses associated with the reaction of uranium and zirconia have been reduced to approximately 1 wt. % of the total uranium throughput. Losses from yttria coating reaction during casting are typically an order of magnitude less than zirconia. The real benefit of these materials is their cost effectiveness given the multiple number of times the graphite crucibles can be reused and/or recoated. Nearly thirty uses have been recorded for graphite crucibles at both the cathode processor and casting operations. The amount of impurity pickup to the uranium products from the graphite crucibles has been minimal (<300 ppm) in terms of carbon, oxygen, zirconium, and yttrium content. Based on the promising results with the zirconia castable at the cathode processor, zirconia material has been cast into a graphite casting crucible and will be tested for comparison in the absence of salt.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

The MPACT 2020 Milestone: Safeguards and Security by Design of Future Nuclear Fuel Cycle Facilities.

The Materials Protection, Accounting, and Control Technologies (MPACT) campaign, within the U.S. Department of Energy Office of Nuclear Energy, has developed a Virtual Facility Distributed Test Bed for safeguards and security design for future nuclear fuel cycle facilities. The purpose of the Virtual Test Bed is to bring together experimental and modeling capabilities across the U.S. national laboratory and university complex to provide a one-stop-shop for advanced Safeguards and Security by Design (SSBD). Experimental testing alone of safeguards and security technologies would be cost prohibitive, but testbeds and laboratory processing facilities with safeguards measurement opportunities, coupled with modeling and simulation, provide the ability to generate modern, efficient safeguards and security systems for new facilities. This Virtual Test Bed concept has been demonstrated using a generic electrochemical reprocessing facility as an example, but the concept can be extended to other facilities. While much of the recent work in the MPACT program has focused on electrochemical safeguards and security technologies, the laboratory capabilities have been applied to other facilities in the past (including aqueous reprocessing, fuel fabrication, and molten salt reactors as examples). This paper provides an overview of the Virtual Test Bed concept, a description of the design process, and a baseline safeguards and security design for the example facility. Parallel papers in this issue go into more detail on the various technologies, experimental testing, modeling capabilities, and performance testing.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗
DOE OSTI2020

Safeguards Modeling for Advanced Nuclear Facility Design.

Future nuclear fuel cycle facilities will see a significant benefit from considering materials accountancy requirements early in the design process. The Material Protection, Accounting, and Control Technologies (MPACT) working group is demonstrating Safeguards and Security by Design (SSBD) for a notional electrochemical reprocessing facility as part of a 2020 Milestone. The idea behind SSBD is to consider regulatory requirements early in the design process to provide more optimized systems and avoid costly retrofits later in the design process. Safeguards modeling, using single analyst tools, allows the designer to efficiently consider materials accountancy approaches that meet regulatory requirements. However, safeguards modeling also allows the facility designer to go beyond current regulations and work toward accountancy designs with rapid response and lower thresholds for detection of anomalies. This type of modeling enables new safeguards approaches and may inform future regulatory changes. The Separation and Safeguards Performance Model (SSPM) has been used for materials accountancy system design and analysis. This paper steps through the process of designing a Material Control and Accountancy (MC&A) system, presents the baseline system design for an electrochemical reprocessing facility, and provides performance metrics from the modeling analysis. The most critical measurements in the electrochemical facility are the spent fuel input, electrorefiner salt, and U/TRU product output measurements. Finally, material loss scenario analysis found that measurement uncertainties (relative standard deviations) for Pu would need to be at 1% (random and systematic error components) or better in order to meet domestic detection goals or as high as 3% in order to meet international detection goals, based on a 100 metric ton per year plant size.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗
DOE OSTI2022

Nuclear Qualms Intern Poster

A 2022 Reuters poll found that 45% of Americans support nuclear power, 33% oppose it, and 20% are unsure, with opposers primarily concerned about meltdowns and nuclear waste. Establishing trust with the public has historically been a challenge for the nuclear energy development community. Technical jargon, specialized concepts and popular fictional portrayals are among the barriers to reassuring the public of modern reactor designs’ safety. Using ten letters sent in 1993 to Congress by interest groups and other archival sources, we demonstrate how the public’s misguided nuclear qualms contributed to Congress cancelling the promising Integral Fast Reactor (IFR) project at Argonne National Laboratory West’s (later INL) EBR II reactor in 1994. The decision halted the completion of a groundbreaking technology that cost the world 28 years and counting in lost time to validate and commercialize IFR technology.

99 GENERAL AND MISCELLANEOUS↗