Effect of grain size on defect annealing in displacement-damaged tungsten
Not provided.
Nuclear Science & Technology↗
Engineering topics
Schwarz-Selinger, Thomas
Publications and source records attributed to Schwarz-Selinger, Thomas.
Deuterium retention in MeV ion-irradiated beryllium
Beryllium, as the material of choice for the ITER first wall, will be subject to high energy neutrons during the operation of the fusion device. Such irradiation will create displacement damage in the crystal lattice, which is known to influence hydrogen isotope retention, e.g. in iron and tungsten. Here in our study, we irradiated beryllium samples with high-energy oxygen ions up to various dpa levels (0.01 dpa, 0.1 dpa, and 1 dpa) and subsequently exposed the samples to deuterium plasma at 370 K and 573 K. The amount of retained deuterium was measured by thermal desorption spectroscopy and nuclear reaction analysis and was found to decrease by about 18% in the samples damaged up to 0.1 dpa compared to the undamaged ones. Further damaging up to 1 dpa did not show any additional effects on deuterium retention. Transmission electron microscopy was used to check for the existence of bubbles after plasma exposure.
36 MATERIALS SCIENCE↗
Data on erosion and hydrogen fuel retention in Beryllium plasma-facing materials
ITER will use beryllium as a plasma-facing material in the main chamber, covering a total surface area of about 620 m 2 . Given the importance of beryllium erosion and co-deposition for tritium retention in ITER, significant efforts have been made to understand the behaviour of beryllium under fusion-relevant conditions with high particle and heat loads. This paper provides a comprehensive report on the state of knowledge of beryllium behaviour under fusion-relevant conditions: the erosion mechanisms and their consequences, beryllium migration in JET, fuel retention and dust generation. The paper reviews basic laboratory studies, advanced computer simulations and experience from laboratory plasma experiments in linear simulators of plasma–wall interactions and in controlled fusion devices using beryllium plasma-facing components. A critical assessment of analytical methods and simulation codes used in beryllium studies is given. The overall objective is to review the existing set of data with a broad literature survey and to identify gaps and research needs to broaden the database for ITER.
Erosion–deposition↗
The influence of helium on deuterium retention in beryllium co-deposits
Tritium co-deposition with materials from the plasma-facing components is expected to be the main contributor to tritium retention in ITER. Since He will also be present in the plasma as fusion ash during the DT campaign, this study focuses on the effect of He on D retention in Be co-deposits. The PISCES-B linear plasma device was used to create co-deposited Be-D-He layers for various deposition temperatures (295 K - 475 K) and He concentrations in D - αHe plasma mixtures (0 ≤ α ≤ 0.1). Thermal desorption spectroscopy and nuclear reaction analysis were used to determine the D concentration in co-deposits. For the lowest He concentration (1%) an increase of D retention was observed for the deposition at room temperature, whereas for higher He concentrations a declining trend of D retention was found. Including 10% of He in the plasma is found to reduce D retention at deposition temperatures below 425 K and have a negligible effect at higher deposition temperatures.
36 MATERIALS SCIENCE↗