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A digital process was developed to facilitate additive manufacturing for ceramic materials using digital light processing (DLP). A numerical model that predicts DLP sample properties can be generated from manufacturing inputs to forecast the effect of resin age on mechanical strength of the printed part based on data collected from experiments. Key parameters for printing the green bodies included determining the depth of cure, layer thickness, material composition, and solids loading. Thermogravimetric analyses were used to develop debinding and sintering curves. Debinding is used to remove the volatile organics comprising the photopolymer resin. Sintering is performed after debinding to increase density and mechanical strength of the printed parts. The sintered parts were then subjected to characterization and mechanical testing. The ensemble of data for various DLP-printed ceramic materials were added to a database. A design of experiments can be generated from the manufacturing process defined in the database with selected changeable parameters randomized over a range. Because the database is defined with an architecture to capture manufacturing processes, it can persist as a more generic platform for manufacturing digital twins. This can ease the development of future digital twins and can grow as a common repository for the insights gained from manufacturing research. Creating a digital twin of a DLP system for 3D printing parts enables manufacturers to simulate and assess the impact of resin age on printing parameters and part quality, facilitating optimization, predictive maintenance, and cost reduction. [ABSTRACT FROM AUTHOR]

A thermal neutron absorber material composed of Al3Hf particles in an aluminum matrix is under development for the Advanced Test Reactor. This metal matrix composite was fabricated via hot pressing of high-purity aluminum and micrometer-size Al3Hf powders at volume fractions of 20.0, 28.4, and 36.5%. Room temperature tensile and hardness testing of unirradiated specimens revealed a linear relationship between volume fraction and strength, while the tensile data showed a strong decrease in elongation between the 20 and 36.5% volume fraction materials. Tensile tests conducted at 200 °C on unirradiated material revealed similar trends. Evaluations were then conducted on specimens irradiated at 66 to 75 °C to four dose levels ranging from approximately 1 to 4 dpa. Tensile properties exhibited the typical increase in strength and decrease in ductility with dose that are common for metallic materials irradiated at ≤0.4Tm. Hardness also increased with neutron dose. The difference in strength between the three different volume fraction materials was roughly constant as the dose increased. Nanoindentation measurements of Al3Hf particles in the 28.4 vol% material showed the expected trend of increased hardness with irradiation dose. Transmission electron microscopy revealed oxygen at the interface between the Al3Hf particles and aluminum matrix in the irradiated material. Scanning electron microscopy of the exterior surface of tensile tested specimens revealed that deformation of the material occurs via plastic deformation of the Al matrix, cracking of the Al3Hf particles, and to a lesser extent, tearing of the matrix away from the particles. The fracture surface of an irradiated 28.4 vol% specimen showed failure by brittle fracture in the particles and ductile tearing of the aluminum matrix with no loss of cohesion between the particles and matrix. The coefficient of thermal expansion decreased upon irradiation, with a maximum change of −6.3% for the annealed irradiated 36.5 vol% specimen. [ABSTRACT FROM AUTHOR]

A predictive model of melt rate in waste glass vitrification operations is needed to inform melter operations during normal and off‐normal operations. This paper describes the development of a model of the cold cap (the reacting melter feed floating on molten glass in a glass melter) that couples heat transfer with the feed‐to‐glass conversion kinetics. The model was applied to four melter feeds designed for high‐level and low‐activity nuclear waste feeds using the material properties, either measured or estimated, to obtain temperature and conversion distribution within the cold cap. The cold cap model, when coupled with a computational fluid dynamics model of a Joule‐heated glass melter, allows the prediction of the glass production rate and power consumption. The results show reasonable agreement with the melting rates measured during pilot‐scale melter tests. [ABSTRACT FROM AUTHOR]

Microreactors, or very small, transportable or mobile nuclear reactors with a capacity of less than 20 MW(thermal), are being developed to provide heat and power for myriad applications in remote areas, military installations, emergency operations, humanitarian missions, and disaster relief zones. A wide variety of reactor types are under consideration, including sodium-cooled fast reactors, molten-salt reactors, very high-temperature gas reactors, and heat pipe reactors. One issue common to all microreactor designs is the need to remove heat from the core. The objective of this paper is to identify a spectrum of diverse approaches to thermal management that can be used develop advanced, high-performance heat removal systems to further enhance the expected performance of a 1- to 20-MW(thermal) nuclear reactor. The focus here is on concepts that can provide a passive means of heat removal and are new to nuclear reactors. Different types of passive heat removal strategies for microreactors are examined, including latent heat-transfer devices, such as various types of heat pipes, natural convection and conduction-radiation cooling, and other thermal devices, such as thermoelectrics and thermoacoustics, that can be used to provide power for auxiliary cooling. Many of these concepts have already been fielded in renewable energy systems. Concepts at different stages of technical maturity are outlined to present ideas that can push the boundaries of thermal management in present-day nuclear technology. Practical considerations relative to the integration of these concepts into nuclear systems are given. [ABSTRACT FROM AUTHOR]

New nuclear reactor designs that incorporate heat pipes are being investigated for possible near-term deployment in terrestrial applications. This study explores the use of screen-covered axially grooved sodium heat pipes and their applicability for providing heat removal for microreactors. A sodium working fluid is appropriate for microreactors operating in the 5 to 20 MW(thermal) range at approximately 650°C. HTPIPE, a legacy software code, was validated for the case of screen-covered grooves and used to perform steady-state analyses to determine the performance limits of a proposed heat pipe design. The performance limits of a sodium heat pipe with a screen-covered square grooved wick structure is compared to that of an equivalent heat pipe with an annular wick. In a horizontal orientation at an operating temperature of 650°C,the performance limits for the heat pipe with an annular wick configuration are 15% higher than for the screen-covered grooved wick. At operating temperatures below 777°C, the annular wick outperforms the screen-covered grooved wick, and at temperatures above 777°C, the screen-covered grooved wick outperforms the annular wick. However, the marginal performance gain at higher temperatures may not justify the use of heat pipes with a screen-covered grooved wick structure due to increased manufacturing costs. [ABSTRACT FROM AUTHOR]