Surface oxidation effect on deuterium permeation in reduced activation ferritic/martensitic steel F82H for DEMO application

Fuel loss and environmental contamination by tritium permeation through structural materials are critical issuesfor the establishment of a fusion DEMO reactor. In this study, the effectivity of a chromium oxide layer formedon reduced activation ferritic/martensitic steel F82H as a tritium permeation barrier and its stability undersimulated solid/liquid breeder blanket conditions have been investigated. A uniform 100-nm-thick chromiumoxide layer was formed by heat treatment at 710 °C for 5 min in 50% argon-50% hydrogen mixed gas with theflow rate of 200 standard cubic centimeter per minute. After exposure to simulated solid breeder blanket con-ditions, an iron oxide layer and a spinel-type iron-chromium oxide layer formed. In the case of a liquid breederblanket condition, the chromium oxide layer partly lost at 500 °C for 100 h. The chromium oxide-formed sampledecreased deuterium permeationflux by a factor of up to 150. The permeation reduction efficiency deterioratedafter exposure to a solid breeder blanket condition due to a change of the chromium oxide layer. However, thechromium oxide formation would play a role to reduce hydrogen isotope permeation even after reduction of theoxide layer

Design strategy and recent design activity on Japan\u27s DEMO

The Joint Special Design Team for Fusion DEMO was organized in 2015 to enhance Japan’s DEMO design activity and coordinate relevant research and development (R&D) toward DEMO. This paper presents the fundamental concept of DEMO and its key components with main arguments on DEMO design strategy. Superconducting magnet technology on toroidal field coils is based on the ITER scheme where a cable-in-conduit Nb3Sn conductor is inserted in the groove of a radial plate. Development of cryogenic steel with higher strength is a major challenge on the magnet. Divertor study has led to a baseline concept based on water-cooled single-null divertor assuming plasma detachment. Regarding breeding blanket, fundamental design study has been continued with focuses on tritium self-sufficiency, pressure tightness in case of in-box LOCA (loss of coolant accident) and material compatibility. An important finding on tritium permeation to the cooling water is also reported, indicating that the permeation to the cooling water is manageable with existing technology

Japan’s Efforts to Develop the Concept of JA DEMO During the Past Decade

This paper summarizes the evolution of Japanese DEMO design studies in a retrospective manner by highlighting efforts to resolve critical design issues on DEMO. Japan is currently working on the conceptual study of a steady-state DEMO (JA DEMO) with a major radius Rp of 8.5 m and fusion power Pfus of 1.5 to 2 GW based on water-cooled solid breeding blanket with pressurized water reactor water condition (290ºC to 325ºC, 15.5 MPa). Such a lower Pfus allows to find realistic design solutions for divertor heat removal. Recognizing that divertor heat removal is one of the most challenging issues on DEMO, the divertor design has been carried out in different approaches, including numerical divertor plasma simulation, magnetic configurations, heat sink design, etc. It is noteworthy that the latest divertor simulation led to a design window allowing divertor heat removal of the peak heat flux of <10 MW/m2. The breeding blanket (BB) design has been concentrated on simplification of the internal structure and pressure tightness of the BB casing against the in-box loss-of-coolant accident. Due to a large amount of radioactive waste generated in periodic replacement of in-vessel components, downsizing of waste-related facilities has come to be regarded as a significant design issue. A possible waste management for reducing temporary waste storage was proposed, and its impact on the plant layout was assessed