ALBERT

Description: 〈p〉Publication date: Available online 28 January 2020〈/p〉 〈p〉〈b〉Source:〈/b〉 Journal of Nuclear Materials〈/p〉 〈p〉Author(s): Shikai Wu, Yilei Shi, Guoyu Zhang, Song Zhang, Hongbin Liao, Xiaoyu Wang, Zhiyun Qin〈/p〉 〈div xml:lang="en"〉 〈h5〉Abstract〈/h5〉 〈div〉〈p〉In this paper, electronic beam butt welding of the reduced-activation ferrite/martenstic CLF-1 steel with a thickness of 32 mm was performed. Furthermore, the impact toughness, microstructure, phase compositions and impact fractures of the joints with different preheating, post welding heat treatment and heat inputs were investigated systematically and the influencing rules and mechanisms were analyzed. The experimental results shown that, with the optimization of the processes parameters, well-formed joints without any defects such as pores, incomplete fusion, and cracks was obtained by electronic beam welding, which also possessed favorable tensile and flexural performances. The impact absorbed energy of the joints was not remarkably improved with preheating or post welding heat treatment. As the heat input dropped from 2805 J mm〈sup〉−1〈/sup〉 to 1440 J mm〈sup〉−1〈/sup〉, the impact absorbed energy of the weld increased from 3.3–12.5 J–275 J after PWHT at 740 °C for 4 h. At a heat input of 1836 J mm〈sup〉−1〈/sup〉, the impact absorbed energy of the weld was comparable to the value of BM, and the specimens were not fractured after impact test. Preheating and PWHT only changed the size of martensite laths and carbides without eliminating the residual δ-ferrites in the weld. In addition, the impact fractures exhibited brittle fracture characteristics. As the heat input reduced, the content of δ-ferrites in the weld decreased significantly, the martensite laths in the weld zone after thermal treatment were refined, and Ta-rich MX carbides in the laths were precipitated. Therefore, the impact toughness of the joints increased significantly and the impact fractures exhibited brittle fracture characteristics.〈/p〉〈/div〉 〈/div〉

Description: 〈p〉Publication date: Available online 28 January 2020〈/p〉 〈p〉〈b〉Source:〈/b〉 Journal of Nuclear Materials〈/p〉 〈p〉Author(s): Peng Dou, Shaomin Jiang, Lanlan Qiu, Akihiko Kimura〈/p〉 〈div xml:lang="en"〉 〈h5〉Abstract〈/h5〉 〈div〉〈p〉FeCrAl oxide dispersion strengthened (ODS) steel is one of the most promising candidate cladding materials of generation IV nuclear reactors because of its excellent resistance to not only corrosion but also creep and irradiation due to the ultrahigh density nanometer-scale oxides. Effects of the contents of Al, Zr and Ti on the crystal and metal/oxide interface structures of the particles in Fe–15Cr–3.7Al–2W–0.1Ti–0.6Zr–0.35Y〈sub〉2〈/sub〉O〈sub〉3〈/sub〉 (3.7Al–0.1Ti–0.6Zr) and Fe–15Cr–3.8Al–2W–0.12Ti–0.5Zr–0.35Y〈sub〉2〈/sub〉O〈sub〉3〈/sub〉 (3.8Al–0.12Ti–0.5Zr) were studied by high resolution transmission electron microscopy (HRTEM). For 3.7Al–0.1Ti–0.6Zr and 3.8Al–0.12Ti–0.5Zr ODS steels, phase identification was accomplished on 175 and 213 particles, respectively, which have peak number fraction and, therefore, represent the oxides contributing most significantly to the macroscopic properties of the ODS steels. The proportions of various types of oxides were determined. For 3.7Al–0.1Ti–0.6Zr and 3.8Al–0.12Ti–0.5Zr ODS steels, the proportions of Y–Zr complex oxides are ∼87.4% and ∼54.4% while the number fractions of Y–Al complex oxides are ∼3.5% and ∼5.2%, respectively, indicating that 0.5–0.6 wt% Zr inhibits the formation of Y–Al complex oxides remarkably while prompts the significant occurrence of Y–Zr complex oxides. For 3.7Al–0.1Ti–0.6Zr and 3.8Al–0.12Ti–0.5Zr ODS steels, the proportions of Y–Ti complex oxides are ∼5.7% and ∼36.6%, respectively, indicating it is effective for increasing the proportion of Y–Ti complex oxides by increasing the content of Ti from 0.09 wt% to 0.12 wt% with the content of Zr decreased from 0.63 wt% to 0.49 wt%. The crystallographic orientation correlations of the oxides and matrix were determined. The formation and refinement mechanisms of oxides and, moreover, the reasons of the unusual irradiation tolerance and thermal stability of the ODS steel were discussed based on the results.〈/p〉〈/div〉 〈/div〉

Description: 〈p〉Publication date: Available online 27 January 2020〈/p〉 〈p〉〈b〉Source:〈/b〉 Journal of Nuclear Materials〈/p〉 〈p〉Author(s): M.H.H. Kolb, J.M. Heuser, R. Rolli, H.-C. Schneider, R. Knitter, M. Zmitko〈/p〉 〈div xml:lang="en"〉 〈h5〉Abstract〈/h5〉 〈div〉 〈p〉The HICU (High neutron fluence Irradiation of pebble staCks for fUsion) experiment was performed in the High Flux Reactor (HFR) in Petten, NL, in order to irradiate candidate tritium breeder materials in a fusion relevant environment. The presented work focuses on the post-irradiation examination of the irradiated lithium orthosilicate based breeder pebbles. The pebble samples showed three different contents of Li-6 and were irradiated at two different temperatures and in mechanically constrained and unconstrained state. In this particular publication, the influences of the irradiation conditions on the pebble morphology, microstructure, porosity, and mechanical strength are addressed.〈/p〉 〈p〉The results indicate that in general a high irradiation temperature seems to be advantageous for maintaining the mechanical strength of the irradiated pebbles. A higher mechanical strength and a significantly lower closer porosity is observed for samples that were irradiated at high temperatures in comparison to pebbles that were irradiated at low temperatures. The effects on the pebble properties with respect to the Li-6 content are small in contrast to effects of the irradiation temperature. With an increased Li-6 content, no deterioration of the material properties was observed, especially for samples irradiated at high temperatures.〈/p〉 〈/div〉 〈/div〉