P. Vladimirov

Current Status of Beryllium Materials for Fusion Blanket Applications

Abstract Beryllium is a promising functional material for several breeder system concepts to be tested within the experimental fusion reactor ITER and, later, implemented in the first commercial demonstration fusion power plant DEMO. For these applications its resistance to neutron irradiation and the detrimental effects of radiogenic gases (helium and tritium) is crucial for fusion reactor safety, subsequent waste management and material recycling. A reliable prediction of beryllium behavior under fusion irradiation conditions requires both dedicated experiments and advanced modeling. Characterization of the reference and alternative beryllium pebble grades was performed in terms of their microstructure and tritium release properties. The results are discussed with respect to their application in fusion blanket systems. The outcomes from the HIDOBE-01 post irradiation experiment (PIE) are discussed to highlight several interesting features manifested by beryllium irradiation at fusion relevant temperatures. Titanium beryllide is presently developed as a possible substitute for beryllium pebbles as it shows better oxidation resistance, higher melting temperature and tritium release efficiency. Pebbles consisting predominantly of Be12Ti phase were successfully fabricated at Rokkasho, Japan. Recent advances in modeling provide new insights on the production of point defects and the behavior of helium and hydrogen impurities in beryllium, improving understanding of the mechanisms of primary damage production, hydrogen’s effect on the size and the shape of gas bubbles, and tritium removal from the pebbles. The relevance of the experimental and modeling results on irradiated beryllium for the design of a fusion demonstration reactor is evaluated, and recommendations for future R&D programs are proposed.

Overview of the HCPB Research Activities in EUROfusion

In the framework of the EUROfusion’s Power Plant Physics and Technology, the working package breeding blanket (BB) aims at investigating four different BB concepts for an EU demonstration fusion reactor (DEMO). One of these concepts is the helium-cooled pebble bed (HCPB) BB, which is based on the use of pebble beds of lithiated ternary compounds and Be or beryllides as tritium breeder and multiplier materials, respectively, EUROFER97 as structural steel and He as coolant. This paper aims at giving an overview of the EU HCPB BB Research and Development (R&D) being developed at KIT, in collaboration with Wigner-RCP, BUTE-INT, and CIEMAT. The paper gives an outline of the HCPB BB design evolution, state-of-the-art basic functionalities, requirements and performances, and the associated R&D activities in the areas of design, functional materials, manufacturing, and testing. In addition, attention is given also to the activities dedicated to the development of heat transfer augmentation techniques for the first wall and the corresponding testing. Due to their nature as design drivers, a brief overview in the R&D of key HCPB interfacing areas is given as well, namely, the tritium extraction and recovery system, the primary heat transfer and power conversion systems, and safety topics, as well as some specific activities regarding the integration of in-vessel systems through the BB. As concluding remarks, an outline of the standing challenges and future R&D plans is summarized.

ODS steel raw material local structure analysis using X-ray absorption spectroscopy

Oxide dispersion strengthened (ODS) steels are promising materials for fusion power reactors, concentrated solar power plants, jet engines, chemical reactors as well as for hydrogen production from thermolysis of water. In this study we used X-ray absorption spectroscopy at the Fe and Cr K-edges as a tool to get insight into the local structure of ferritic and austenitic ODS steels around Fe and Cr atoms and its transformation during mechanical alloying process. Using the analysis of X-ray absorption near edge structure (XANES) and extended X-ray absorption fine structure (EXAFS) we found that for austenitic samples a transformation of ferritic steel to austenitic steel is detectable after 10 hours of milling and proceeds till 40 hours of milling; only small amount of a-phase remains after 80 hours of milling. We found that the Cr K-edge EXAFS can be used to observe distortions inside the material and to get an impression on the formation of chromium clusters. In-situ EXAFS experiments offer a reliable method to investigate the ferritic to austenitic transformation.

Temperature-dependent EXAFS study of the local structure and lattice dynamics in cubic Y₂O₃.

The local structure and lattice dynamics in cubic Y2O3 were studied at the Y K-edge by X-ray absorption spectroscopy in the temperature range from 300 to 1273 K. The temperature dependence of the extended X-ray absorption fine structure was successfully interpreted using classical molecular dynamics and a novel reverse Monte Carlo method, coupled with the evolutionary algorithm. The obtained results allowed the temperature dependence of the yttria atomic structure to be followed up to ∼6 Å and to validate two force-field models.

Nanoscale characterization of 13.5% Cr oxide dispersion strengthened steels with various titanium concentrations

The influence of titanium alloying (Ti content of 0, 0.2, 0.3, and 0.4 wt %) on the nanostructure of yttrium oxide (Y2O3) dispersion strengthened steel with a composition Fe-13.5% Cr-2% W-0.3% Y2O3 is investigated. The spatial distribution of chemical elements is analyzed in the investigated volumes. The matrix composition and average size and concentration of nanoscale clusters are compared for different samples. It is shown that the average nanocluster size (∼3 nm) is almost unchanged with increasing Ti concentration, while the cluster concentration grows from ∼1 × 1023 m−3 (for Ti-free steel) to ∼1.5 × 1024 m−3 (for 0.4 wt % Ti alloy).