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In this study, creep-fatigue (CF) tests on F82H at 823 K were conducted according to the standard specimen and requirement of ASTM. Tension holding (TH) showed higher lifetime than compression holding (CH). Compared with 9Cr-1Mo-V steels, CF lifetime degradation was not observed at the holding time less than 0.05 h in case of CH and 0.17 h in case of TH. CF damages were summarized from the CF test results by Campbell method. The obtained damages distributed around CF damage envelope of Grade 91 in ASME NH. CF damage of cylindrical water-cooled ceramic breeder (WCCB) test blanket module (TBM) container at the zenith of hemisphere body under the ITER thermo-mechanical load was calculated. The calculated damage located within of the damage envelope of Grade 91. It was indicated that the present design of cylindrical WCCB TBM might resist CF interaction under the ITER thermo-mechanical load conditions.

Reduced activation ferritic/martensitic (RAFM) steel, e.g., F82H, is the leading candidate structural material for fusion blanket. Of many blanket concepts, the water-cooled ceramic breeder blanket is an attractive concept because of its compactness and its compatibility with the technologies which using high-temperature water as coolant in conventional light water reactor. For tritium breeding, it is necessary to manage the corrosion of F82H by controlling not chemicals but hydrogen and oxygen in the cooling water. Understanding the effects of oxygen and hydrogen on corrosion property was therefore required for the blanket design. This work aims to investigate the corrosion behavior of F82H in high-temperature water with hydrogen added. Based on the results of the corrosion test, it is speculated that hydrogen addition expected no considerable change of the flow-accelerated corrosion (FAC). In contrast, it was found that the number of cycles to fracture was decreased in the high-temperature water in comparison to the fatigue test at room temperature in air.

Reduced activation ferritic/martensitic (RAFM) steel, e.g., F82H, is the leading candidate structural material for fusion blanket. Of many blanket concepts, the water-cooled ceramic breeder blanket is an attractive concept because of its compactness and its compatibility with the technologies which using high-temperature water as coolant in conventional light water reactor. For tritium breeding, it is necessary to manage the corrosion of F82H by controlling not chemicals but hydrogen and oxygen in the cooling water. Understanding the effects of oxygen and hydrogen on corrosion property was therefore required for the blanket design. This work aims to investigate the corrosion behavior of F82H in high-temperature water with hydrogen added. Based on the results of the corrosion test, it is speculated that hydrogen addition expected no considerable change of the flow-accelerated corrosion (FAC). In contrast, it was found that the number of cycles to fracture was decreased in the high-temperature water in comparison to the fatigue test at room temperature in air.

A water-cooled ceramic breeder blanket system is the Japanese promising design, and the candidate structuralmaterial is a reduced activation ferritic/martensitic steel, F82H. On the other hands, the material of the pipes ofcooling system is planned to an austenitic stainless steel, SUS316 L. The cooling water will commonlyflow bothF82H cooling channels and SUS316 L cooling pipes. Appropriate water chemistry condition for each of them isdifferent, but it is necessary tofind out a reasonable and common water chemistry condition for both of them. Inthe fusion blanket, since the cooling water will be irradiated by neutrons and brought about the radiolysis ofwater, the dissolved oxygen concentration of the cooling water is considered to change by the radiolysis ofwater. This research provides the simultaneous effects of dissolved oxygen and applied stress on surface mor-phology of F82H and SUS316 L by the pressurized water immersion test in dissolved oxygen concentration up to100 ppb.

Reduced activation ferritic/martensitic (RAFM) steels are assumed as structural material of blanket design activities. Determining the design limit is essential for the RAFM steels to qualify their use in fusion blankets. This work investigates fatigue properties of a RAFM steel, F82H at elevated temperature up to 823 K in the air and vacuum condition. The tests were carried out with axial strain controlled condition using cylindrical specimen, and the results were summarized using the Manson’s modified universal slope method, which can describe fatigue lifetime as a function of tensile properties. Although fatigue lifetime of F82H decreased with temperature, the fatigue lifetime up to 723 K fell into factor of 2 of the lifetime at ambient temperature. The lifetime at vacuum condition was successfully described using universal slope method. On the other hand, the lifetime in the air was 1/3 or shorter than that in vacuum condition. It is notable that a heat of F82H prepared for IEA round robin tests demonstrated 1/5 of shorter lifetime than latest F82H. The difference is discussed with inclusion morphologies affected by steel manufacturing process and so on. This report has been prepared as an account of work assigned to the Japanese Implementing Agency within the ‘‘Broader Approach Agreement’’ between the Government of Japan and the European Atomic Energy Community.

In this study, creep-fatigue (CF) tests on F82H at 823 K were conducted according to the standard specimen and requirement of ASTM. Tension holding (TH) showed higher lifetime than compression holding (CH). Compared with 9Cr-1Mo-V steels, CF lifetime degradation was not observed at the holding time less than 0.05 h in case of CH and 0.17 h in case of TH. CF damages were summarized from the CF test results by Campbell method. The obtained damages distributed around CF damage envelope of Grade 91 in ASME NH. CF damage of cylindrical water-cooled ceramic breeder (WCCB) test blanket module (TBM) container at the zenith of hemisphere body under the ITER thermo-mechanical load was calculated. The calculated damage located within of the damage envelope of Grade 91. It was indicated that the present design of cylindrical WCCB TBM might resist CF interaction under the ITER thermo-mechanical load conditions.

This work investigates mechanical properties of weld joints of a reduced activation ferritic/martensitic steel, F82H and effects of post weld heat treatment on the welds. Vickers hardness, tensile and Charpy impact tests were conducted on F82H weld joints prepared using tungsten-inert-gas and electron beam after various heat treatments. Although narrow groove tungsten-inert-gas welding reduced volume of weld bead, significant embrittlement was observed in a heat affected zone transformed due to heat input. Post weld heat treatment above 993 K successfully moderated the brittle transformed region. The hardness of the brittle region strongly depends on the heat treatment temperature. Meanwhile, strength of base metal was slightly reduced by the treatment at temperature ranging from 993 to 1053 K. Moreover, softening due to double welding was observed only in the weld metal, but negligible in base metal.

Japan Atomic Energy Agency (JAEA) is performing the development of a Water Cooled Ceramic Breeder (WCCB) Test Blanket Module (TBM) as one of the most important steps toward DEMO blanket. Regarding the blanket module fabrication technology development using F82H, the fabrication of a real scale mockup of the back wall of TBM was completed. In the design activity of the TBM, electromagnetic analysis under plasma disruption events and thermo-mechanical analysis under steady state and transient state of tokamak operation have been performed and showed bright prospect toward design justification. Regarding the development of advanced breeder and multiplier pebbles for DEMO blanket, fabrication technology development of Li rich Li2TiO3 pebble and BeTi pebble was performed. Regarding the research activity on the evaluation of tritium generation performance, the evaluation of tritium production and recovery test using D-T neutron in the Fusion Neutronics Source (FNS) facility has been performed. This paper overviews the recent achievements of the development of the WCCB Blanket in JAEA.