Publication: Tritium retention in W plasma-facing materials: Impact of the material structure and helium irradiation
Дата
2019
Авторы
Bernard, E.
Sakamoto, R.
Hodille, E.
Kreter, A.
Grisolia, C.
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© 2019 The Authors 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.
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Tritium retention in W plasma-facing materials: Impact of the material structure and helium irradiation / Bernard, E. [et al.] // Nuclear Materials and Energy. - 2019. - 19. - P. 403-410. - 10.1016/j.nme.2019.03.005
URI
https://www.doi.org/10.1016/j.nme.2019.03.005
https://www.scopus.com/record/display.uri?eid=2-s2.0-85063444261&origin=resultslist
http://gateway.webofknowledge.com/gateway/Gateway.cgi?GWVersion=2&SrcAuth=Alerting&SrcApp=Alerting&DestApp=WOS_CPL&DestLinkType=FullRecord&UT=WOS:000470746100064
https://openrepository.mephi.ru/handle/123456789/16724
https://www.scopus.com/record/display.uri?eid=2-s2.0-85063444261&origin=resultslist
http://gateway.webofknowledge.com/gateway/Gateway.cgi?GWVersion=2&SrcAuth=Alerting&SrcApp=Alerting&DestApp=WOS_CPL&DestLinkType=FullRecord&UT=WOS:000470746100064
https://openrepository.mephi.ru/handle/123456789/16724