TY - JOUR
T1 - Tritium retention in W plasma-facing materials
T2 - Impact of the material structure and helium irradiation
AU - Bernard, E.
AU - Sakamoto, R.
AU - Hodille, E.
AU - Kreter, A.
AU - Autissier, E.
AU - Barthe, M. F.
AU - Desgardin, P.
AU - Schwarz-Selinger, T.
AU - Burwitz, V.
AU - Feuillastre, S.
AU - Garcia-Argote, S.
AU - Pieters, G.
AU - Rousseau, B.
AU - Ialovega, M.
AU - Bisson, R.
AU - Ghiorghiu, F.
AU - Corr, C.
AU - Thompson, M.
AU - Doerner, R.
AU - Markelj, S.
AU - Yamada, H.
AU - Yoshida, N.
AU - Grisolia, C.
N1 - Publisher Copyright:
© 2019 The Authors
PY - 2019/5
Y1 - 2019/5
N2 - Plasma-facing materials for next generation fusion devices, like ITER and DEMO, will be submitted to intense fluxes of light elements, notably He and H isotopes (HI). Our study focuses on tritium (T) retention on a wide range of W samples: first, different types of W materials were investigated to distinguish the impact of the pristine original structure on the retention, from W-coated samples to ITER-grade pure W samples submitted to various annealing and manufacturing procedures, along with monocrystalline W for reference. Then, He and He-D irradiated W samples were studied to investigate the impact on He-damages such as nano-bubbles (exposures in LHD or PSI-2) on T retention. We exposed all the samples to tritium gas-loading using a gentle technique preventing any introduction of new damage in the material. Tritium desorption is measured by Liquid Scintillation counting (LSC) at ambient and high temperatures (800 °C). The remaining T inventory is then measured by sample full dissolution and LSC. Results on T inventory on He exposed samples highlighted that in all cases, tritium desorption as a gas (HT) increases significantly due to the formation of He damages. Up to 1.8 times more T can be trapped in the material through a competition of various mechanisms, but the major part of the inventory desorbs at room temperature, and so will most likely not take part to the long-term trapped inventory for safety and operational perspectives. Unfortunately, investigation of “as received” industrial W (used for the making of plasma-facing materials) highlighted a strong impact of the pre existing defects on T retention: up to 2.5 times more T is trapped in “as received W” compared to annealed and polish W, and desorbs only at 800 °C, meaning ideal W material studies may underestimate T inventory for tokamak relevant conditions.
AB - Plasma-facing materials for next generation fusion devices, like ITER and DEMO, will be submitted to intense fluxes of light elements, notably He and H isotopes (HI). Our study focuses on tritium (T) retention on a wide range of W samples: first, different types of W materials were investigated to distinguish the impact of the pristine original structure on the retention, from W-coated samples to ITER-grade pure W samples submitted to various annealing and manufacturing procedures, along with monocrystalline W for reference. Then, He and He-D irradiated W samples were studied to investigate the impact on He-damages such as nano-bubbles (exposures in LHD or PSI-2) on T retention. We exposed all the samples to tritium gas-loading using a gentle technique preventing any introduction of new damage in the material. Tritium desorption is measured by Liquid Scintillation counting (LSC) at ambient and high temperatures (800 °C). The remaining T inventory is then measured by sample full dissolution and LSC. Results on T inventory on He exposed samples highlighted that in all cases, tritium desorption as a gas (HT) increases significantly due to the formation of He damages. Up to 1.8 times more T can be trapped in the material through a competition of various mechanisms, but the major part of the inventory desorbs at room temperature, and so will most likely not take part to the long-term trapped inventory for safety and operational perspectives. Unfortunately, investigation of “as received” industrial W (used for the making of plasma-facing materials) highlighted a strong impact of the pre existing defects on T retention: up to 2.5 times more T is trapped in “as received W” compared to annealed and polish W, and desorbs only at 800 °C, meaning ideal W material studies may underestimate T inventory for tokamak relevant conditions.
KW - Helium
KW - Plasma-wall interactions
KW - Tritium inventory
KW - Tungsten
UR - https://www.scopus.com/pages/publications/85063444261
U2 - 10.1016/j.nme.2019.03.005
DO - 10.1016/j.nme.2019.03.005
M3 - Article
SN - 2352-1791
VL - 19
SP - 403
EP - 410
JO - Nuclear Materials and Energy
JF - Nuclear Materials and Energy
ER -