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Abstract: This paper summarizes the proposals and findings of the IFMIF Specification Working Group established to update the users requirements and top level specifications for the facility. Special attention is given to the different roadmaps of fusion pathway towards power plants, of materials R&D and of facilities and their interactions. The materials development and validation activities on structural materials, blanket functional materials and non-metallic materials are analyzed and specific objectives and requirements to be implemented in IFMIF are proposed. Emphasis is made in additional potential validation activities that can be developed in IFMIF for ITER TBM qualification as well as for DEMO-oriented mock-up testing. [Copyright &y& Elsevier]

Abstract: Reduced activation ferritic/martensitic steels (RAFMs) are recognized as the primary candidate structural materials for fusion blanket systems. The RAFM F82H was developed in Japan with emphasis on high-temperature properties and weldability. Extensive irradiation studies have conducted on F82H, and it has the most extensive available database of irradiated and unirradiated properties of all RAFMs. The objective of this paper is to review the R&D status of F82H and to identify the key technical issues for the fabrication of an ITER test blanket module (TBM) suggested from the recent research achievements in Japan. This work clarified that the primary issues with F82H involve welding techniques and the mechanical properties of weld joints. This is the result of the distinctive nature of the joint caused by the phase transformation that occurs in the weld joint during cooling, and its impact on the design of a TBM will be discussed. [Copyright &y& Elsevier]

Abstract: This paper describes the latest design of liquid lithium (Li) target system in International Fusion Materials Irradiation Facility (IFMIF). IFMIF is an accelerator-driven intense neutron source for fusion reactor materials testing. Two 40MeV deuteron beams with a total current of 250mA strikes a liquid Li target circulating in a Li loop. The Li target is to provide a stable Li jet at a speed of 10–20m/s to handle an averaged heat flux of 1GW/m2. A concaved Li flow is applied to avoid Li boiling. A cold trap and two kinds of hot trap are applied to control impurities (T, 7Be, C, O, N) below permissible levels. A back wall made of RAFM is located in the most severe region of neutron irradiation (50dpa/year). System availability requires more than 95% during plant lifetime. EVEDA tasks including Li loop are now being performed under the Broader Approach. [Copyright &y& Elsevier]

Abstract: The possibility of reducing the concentration of 14C produced in irradiation of reduced activation ferritic steels (RAFS) intended as structural materials in first wall and blanket of fusion power plants was investigated. During RAFS irradiation 14C is mainly produced from the more abundant of the two isotopes of nitrogen present in the steel, namely 14N. The method proposed consists in increasing the enrichment of the other isotope (15N). One-dimension transport calculations show that for a typical DEMO blanket configuration an 15N enrichment from the natural value (0.37%) to 95% was sufficient to keep the end of life 14C concentration in the RAFS below the limit (3.7×107 Bq/kg) fixed by the Japanese Nuclear Safety Commission for qualifying it as a low level material (LLM) which can be disposed by shallow land burial. [Copyright &y& Elsevier]

Abstract: The reduced activation ferritic steel (RAFS) has been selected as structural material for a variety of blanket systems for ITER test blanket modules (TBM). In the evaluation of integrated performance of ferritic steels as structural components of blanket systems, there are unique issues as well as common issues for each blanket system. One of the unique issues for each system is the compatibility of ferritic steels with the coolant materials. The corrosion rate of ferritic steels in hot water, super critical pressurized water (SCPW), humid air, Pb–17Li, lithium and Flibe at various temperatures is reviewed in this work. Efforts to improve corrosion resistance have been made, taking the alloy design into account. A dispersion of yttria was effective to improve corrosion resistance of a RAFS. The compatibilities of RAFSs with hot water, Pb–17Li, lithium and Flibe are considered to be good enough for the TBM applications. The liquid metal embrittlement (LME) is considered to be a critical issue for the utilization of RAFSs for the lithium systems. Several issues towards DEMO and beyond are shown from the compatibility point of view. [Copyright &y& Elsevier]

Abstract: Small specimen test technology (SSTT) has been developed to investigate mechanical properties of nuclear materials. SSTT has been driven by limited availability of effective irradiation volumes in test reactors and accelerator-based neutron and charged particle sources. In this study, new bend test machines have been developed to obtain fracture behaviors of F82H steel for very small bend specimens of pre-cracked t/2-1/3CVN (Charpy V-notch) with 20mm length and deformation and fracture mini bend specimen (DFMB) with 9mm length and disk compact tension of 0.18DCT (disk compact tension) type, and fracture behaviors were examined at 20°C. The effect of specimen size on ductile–brittle transition temperature (DBTT) of F82H steel was examined by using 1/2t-CVN, 1/3CVN and t/2-1/3CVN, and it was revealed that DBTT of t/2-1/3CVN and 1/3CVN was lower than that of t/2-CVN. DBTT behaviors due to helium and displacement damage in F82H-std irradiated at about 120°C by 50 or 100MeV He ions to 0.03dpa were also measured by small punch tests. [Copyright &y& Elsevier]