Fusion tritons and plasma-facing components in a fusion reactor

Abstract

In future fusion reactors the long-term tritium retention will be a critical issue. In ITER, which will be the first device to demonstrate a burning plasma, the in-vessel tritium inventory limit will be 350 g. The primary retention mechanism of tritium is co-deposition with eroded first wall material. The measured tritium build-up rates at JET and TFTR tokamaks, both of which were operating with carbon walls at the time of their tritium campaigns, are too high for ITER. For these reasons there is today a growing interest in carbon-free fusion machines:ASDEX Upgrade in Garching, Germany, has replaced, step-by-step, all carbon fibre composite structures by tungsten-coated ones.The tritium inventory from the fuel tritium is typically found on top of the material surfaces, where it is deposited as amorphous hydrogen-rich carbon layers. This is because the fuel tritons are thermal, i.e., have low energy (E 100 eV) and, thus, a penetration depth of at most a few nanometers. However, tritium is also formed in deuterium-deuterium (D-D) reactions. The quantity of tritium formed inside the plasma will be insignificant compared to the amount of fuel tritium: the D-D fusion reaction rate is about 1/200 of the deuterium-tritium reaction rate. But due to its very high energy, contribution of the fusion-born tritium to the long-term tritium inventory and material damages can still be significant. This difference becomes evident by comparing the mean range and backscattered fraction of tritons in carbon, tungsten and beryllium, shown in tables 1 and 2. Energy 100 eV 1 MeV tungsten (2.8 ± 1.5) nm (4.3 ± 0.1) µm carbon (2.8 ± 1.6) nm (7.7 ± 0.1) µm beryllium (3.3 ± 1.6) nm (10.2 ± 0.1) µm 2) % (0.11 ± 0.01) % carbon (9.9 ± 0.1) % (0.000 ± 0.001) % beryllium (6.2 ± 0.1) % (0.000 ± 0.001) %

Overview of first Wendelstein 7-X high-performance operation

“…Further analysis of triton orbits and especially triton losses in W7-X with ASCOT could be beneficial in calculating triton losses to the W7-X wall. Determining triton loss patterns is important not only for heat load management but also for radiation safety: due to their high energy the tritons produced in DD can be deposited deeply into the device walls, which leads to tritium retention over time [27]. Studying triton orbits is also beneficial for designing future stellarator reactors, because the gyroradius r L of 1.01 MeV tritons in W7-X is similar to 3.6 MeV alpha particles in a foreseen HELIAS reactor.…”

Effects of complex symmetry-breakings on alpha particle power loads on first wall structures and equilibrium in ITER

“…The comparisons between the experimental data and the simulation were carried out in the present-day tokamaks, independently. The comparisons with the data of the triton surface deposition and the data of the neutral particle analyser were carried out in the ASDEX-U for ASCOT [22,23]. The comparisons with energetic deuteron confinement estimated by the neutron emission decay and the wall load estimated by the infrared TV camera measurement were carried out in the JT-60U [24], and the comparisons of the wall load estimated by the infrared TV camera measurement were carried out in the JFT-2M under the environment with ferritic inserts [12] for F3D OFMC.…”

ASCOT simulations of 14 MeV neutron rates in W7-X: effect of magnetic configuration

“…Further analysis of triton orbits and especially triton losses in W7-X with ASCOT could be beneficial in calculating triton losses to the W7-X wall. Determining triton loss patterns is important not only for heat load management but also for radiation safety: due to their high energy the tritons produced in DD can be deposited deeply into the device walls, which leads to tritium retention over time [30]. Studying triton orbits is also beneficial for designing future stellarator reactors, because the gyroradius r L of 1.01 MeV tritons in W7-X is similar to 3.6 MeV alpha particles in a foreseen HELIAS reactor.…”