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Based on the results of the IFMIF/EVEDA project in the broader approach (BA) activities, the conceptual design of the Advanced Fusion Neutron Source (A-FNS) to obtain irradiation data of DEMO blanket materials is under consideration in Japan. This paper describes the progress of the A-FNS conceptual design, which consists of a 40 MeV, 125 mA deuteron beam, a liquid lithium target system, and a test and post-irradiation test facilities. the main objectives of the A-FNS are the acquisition of RAFM irradiation data up to 2035 and the tritium recovery test on the blanket. In addition, it is expected to be used for various neutron researches. The technical design activities of the A-FNS are focused on improvement of the lithium target design, review of impurity criteria, remote handling, and irradiation application plan. [ABSTRACT FROM AUTHOR]

• Alleviation of impurities levels in Li for IFMIF-type neutron source are suggested. • No fatal problems are found by applying the alleviation of impurities levels. • Level of Be-7 in Li under process can be diminished by the alleviation. The alleviated limits of impurities of nitrogen and hydrogen isotopes in lithium are suggested to be 400wppm and 550appm, respectively for early realization of Advanced Fusion Neutron Source planned in Japan. The applicability of the suggestion is discussed by comparing to that of IFMIF. As the results, not less than 4-year buffer for R&D of nitrogen trap are brought in, which is thought as one of bottle neck. Additionally, the application of this suggestion would diminish the beryllium (a radioactive isotope will generate in A-FNS process) concentration in processing lithium, as well. On the other hand, it is indicated that there would be no fatal problems by application of the present suggestion. [ABSTRACT FROM AUTHOR]

Abstract: Depth profile and retention properties of hydrogen in the graphite armor tiles of JT-60 open-divertor used for HH discharge period were investigated by secondary ion mass spectroscopy and thermal desorption spectroscopy. Two tiles were investigated, one at the inner divertor area covered with the re-deposited layers of the largest thickness (70μm) and the other at the private flux region near the outer divertor striking point hardly covered with the re-deposited layers (only 1–2μm). In the former, hydrogen was retained uniformly in the carbon re-deposited layers with H/C of 0.03, which was significantly smaller than H/C observed in the re-deposited layers of other tokamaks. Such small hydrogen retention in the re-deposited layers is attributed to the temperature rise of the re-deposited layers during discharges due to their poor thermal contact to the substrate. The latter retained rather large amount of hydrogen only in the shallow surface layers and its depth profile decayed sharply to 1.3×10−4 in H/C of the bulk retention as expected from the very thin re-deposited layers and observed temperature rise above 900K. [Copyright &y& Elsevier]