Estimation of tritium accumulation in tungsten and its application for ITER divertor

2014 Fiscal Year Final Research Report

• Project/Area Number: 24360380
• Research Category: Grant-in-Aid for Scientific Research (B)
• Allocation Type: Partial Multi-year Fund
• Section: 一般
• Research Field: Nuclear fusion studies
• Research Institution: University of Toyama
• Principal Investigator: TORIKAI Yuji 富山大学, 水素同位体科学研究センター, 准教授 (80313592)
• Co-Investigator(Kenkyū-buntansha): KURISHITA Hiroaki 東北大学, 金属材料研究所, 准教授 (50112298) ; ISOBE Kanetsugu 独立行政法人日本原子力研究開発機構, 核融合研究開発部門, 研究主幹 (00354613) ; OYAIZU Makoto 独立行政法人日本原子力研究開発機構, 核融合研究開発部門, 任期付研究員 (60516855)
• Project Period (FY): 2012-04-01 – 2015-03-31
• Keywords: プラズマ・壁相互作用 / タングステン / トリチウム蓄積 / ITER

Outline of Final Research Achievements

Tungsten(W) is currently contemplated as plasma facing material because of its advantageous thermo physical properties and rather low solubility of tritium. Tritium solubility of W estimated in this study is 3 order higher than that reported by literature. Traps or oxide films may affect the retention capability of W and lead significantly modified release properties. It became clear that there were capture sites that had different thermal stability and capture intensity in W after polishing, or oxide films that were grown on the surface of W and had barrier effects. Detailed investigation of the impact of possibly rather diverse traps produced either during manufacturing-or via radiation-induced processes and oxide films after annealing on the uptake and retention properties of hydrogen isotopes retained by W used in first wall components of fusion machines is therefore necessary in order to assess correctly and minimize the tritium inventory during various phases of operation.

Free Research Field

水素同位体学