A.V. Fedorov

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.

Helium implanted RAFM steels studied by positron beam Doppler Broadening and Thermal Desorption Spectroscopy

Reduced Activation Ferritic/Martensitic steels are being extensively studied because of their foreseen application in fusion and Generation IV fission reactors. To mimic neutron irradiation conditions, Eurofer97 samples were implanted with helium ions at energies of 500 keV and 2 MeV and doses of 5x1015-1016 He /cm2, creating atomic displacements in the range 0.07–0.08 dpa. The implantation induced defects were characterized by positron beam Doppler Broadening (DB) and Thermal Desorption Spectroscopy (TDS). The DB data could be fitted with one or two layers of material, depending on the He implantation energy. The S and W values obtained for the implanted regions suggest the presence of not only vacancy clusters but also positron traps of the type present in a sub-surface region found on the reference sample. The traps found in the implanted layers are expected to be HenVm clusters. For the 2 MeV, 1016 He/cm2 implanted sample, three temperature regions can be observed in the TDS data. Peaks below 450 K can be ascribed to He released from vacancies in the neighbourhood of the surface, the phase transition is found at 1180 K and the peak at 1350 K is likely caused by the migration of bubbles.

4.15 – Ceramic Breeder Materials

This article reviews the research on ceramic tritium breeding materials and the development of these materials, which aims at providing the tritium required for the deuterium–tritium fuel cycle in magnetic fusion reactors. In particular, the performance of various oxides of lithium is discussed. Currently, much of the ceramic breeder research and development is focused on lithium orthosilicate and metatitanate systems in the form of pebble beds. This chapter reviews design requirements, manufacturing routes, pebble and pebble-bed thermomechanics, tritium production and release properties, neutron-irradiation behavior, chemistry, and modeling. Ongoing work is summarized and the outlook for further work, including test programs in International Thermonuclear Experimental Reactor (ITER), is provided.

Dose effects in He implanted Eurofer97 steel

Reduced Activation Ferritic/Martensitic steels are being extensively studied because of their foreseen application in fusion reactors. To reproduce neutron irradiation conditions, Eurofer97 samples were implanted at room temperature with helium ions at energies of 500 keV and 2 MeV and doses of 1x1015-1017 He /cm2. The implantation induced defects were characterized by positron beam Doppler Broadening (DB). The samples were annealed in the range 300 – 1500 K, in 100 K steps. As the temperature increases, the annealing of vacancies and vacancy clusters is noticed and followed by the coalescence of HenVm clusters. At temperatures around 1200 K HeV pairs dissociate and bubbles are formed. Above 1300 K the helium release from bubbles is observed. The S-W graphs reveals that the samples have similar positron traps up to 1200 K. At 1200 K helium bubbles are noticed and the S,W pair shows a clearly distinct behaviour from the S,W values of vacancy type defects. As the temperature increases and helium is release, the S,W pairs shift towards the S,W’s of vacancies.

Ageing Behaviour of Neutron/Irradiated Eurofer97

Eurofer97 is a candidate structural material for nuclear fusion reactors. To better understand the ageing effects due to radiation during long-term use in real fusion conditions, Eurofer97 was neutron irradiated at the High Flux Reactor in The Netherlands. TEM images of post-irradiation Eurofer97 reveal a high density of irradiation-induced dislocation loops, fine precipitates and agglomeration of point defects. Microscopy results are related to post-irradiation tensile mechanical tests done at room temperature and at 300 ℃. An increase of yield and ultimate tensile strength combined with a decrease of total elongation is observed in both tests and correlated with the presence of radiation damage.