Control rods with low worth absorber materials may provide a mechanical means of making relatively small adjustments in core reactivity. Mo-Tb-Dy and Y-Tb-Dy alloys were developed to obtain appropriate nuclear performance for low worth absorber material. The two alloys were prepared by powder metallurgy technology and vacuum melting technology individually. To clarify the effects of Mo and Y diluents, Tb-Dy was also prepared to be compared. The microstructures were analyzed by X-ray diffraction (XRD), Scanning Electron Microscope (SEM) and Transmission Electron Microscopy (TEM). The experiment results showed that homogeneous microstructures were obtained. Out-pile properties, including mechanical properties, thermal conductivities, thermal expansion, corrosion resistance properties and ion irradiation properties were measured and analyzed. Y-Tb-Dy has similar properties with Tb-Dy. With temperature increasing, yield strengths of Tb-Dy and Y-Tb-Dy decreases largely while Mo-Tb-Dy decreases slightly. Thermal conductivities of Mo-Tb-Dy were four times more than Tb-Dy and Y-Tb-Dy. Mo element significantly increases thermal conductivity. Tb-Dy and Y-Tb-Dy showed severe corrosion and became powders in 280°C/10MPa de-ionized water while Mo-Tb-Dy had very slow corrosion rate. All three alloys were irradiated at 400∼700°C for 25 displacement per atom (dpa). No voids was observed for Tb-Dy and Y-Tb-Dy. Void diameter increases and its density decreases with temperature increasing for Mo-Tb-Dy. Maximum irradiation swelling rate with 0.5% was observed at 500°C. Irradiation swelling significantly decreased with increasing irradiation temperature.
- Nuclear Engineering Division
Preparation and Properties Investigation of Neutron Absorber Material Mo-Tb-Dy and Y-Tb-Dy Alloys
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Lu, J, Chen, L, Qiu, C, Yang, W, Zhang, F, & Yang, R. "Preparation and Properties Investigation of Neutron Absorber Material Mo-Tb-Dy and Y-Tb-Dy Alloys." Proceedings of the 2017 25th International Conference on Nuclear Engineering. Volume 3: Nuclear Fuel and Material, Reactor Physics and Transport Theory; Innovative Nuclear Power Plant Design and New Technology Application. Shanghai, China. July 2–6, 2017. V003T02A026. ASME. https://doi.org/10.1115/ICONE25-66747
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