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Abstract In the framework of the European "HORIZON 2020" innovation and research programme, the EUROfusion Consortium develops a design of a fusion power demonstrator (DEMO). One of the key components in the fusion reactor is the Breeding Blanket (BB) surrounding the plasma, ensuring tritium self-sufficiency, heat removal for conversion into electricity, and neutron shielding. CEA-Saclay, with the support of Wigner-RCP and Centrum výzkumu Řež, is in charge of the development of one of the four BB concepts investigated in Europe for DEMO: the Helium Cooled Lithium Lead (HCLL) BB. The rationales of the HCLL are the use of Eurofer as structural material, eutectic liquid lithium-lead (PbLi) as tritium breeder and neutron multiplier, and helium gas as coolant. This paper shows the basic description of the DEMO HCLL BB concept and its design evolution during the past years, from a design based on the ITER Test Blanket Module (TBM) concept to a more advanced design called "Advanced-Plus" concept. This new HCLL BB concept that has been designed in order to improve Tritium Breeding Ratio (TBR) and shielding performances is presented. This new reference HCLL BB design has been analyzed and show very promising nuclear performances. Nevertheless, the "Optimized Conservative" concept, based on ITER TBM, is still considered as a robust back-up solution since structural improvements are still necessary on the "Advanced-Plus" concept. Moreover, a new Back Supporting Structure (BSS) is presented in this paper, designed to support the BB modules, with the aim to reduce pressure drops and thermal stresses. [ABSTRACT FROM AUTHOR]

The paper describes the organization of the Test Blanket Module (TBM) program, its overall objective and schedule and the status of the technical activities within the ITER Organization-Central Team (IO-CT). The latter include the design integration of the Test Blanket Systems (TBSs) into the nuclear buildings, ensuring all interfaces with other ITER systems, the design of the common TBS components such as the TBM Frames, the Dummy TBMs, and the TBS maintenance tools and equipment in the TBM Port Cell as well as in the Hot Cell building, the design of the TBS connection pipes and the definition of the required maintenance operations and associated R&D. The paper also discusses the major challenges that the TBM Program will be facing in ITER such as the potential impact of the TBMs ferritic/martensitic structures on plasma operations, the approaches to tritium and contamination confinement, the required mitigation and recovery actions in case of accidents, and the assessment of the reliability aspects that could have an impact on ITER availability. [ABSTRACT FROM AUTHOR]

Abstract Korea has own programs toward DEMO and fusion reactors that will require further improved DEMO blanket and energy utilization systems. Primary goals of DEMO blanket development in Korea are to develop and verify the integrated blanket design tools; to develop blanket materials, cooling and tritium fuel cycle technologies; and to develop and evaluate fabrication and joining technologies. The concept of helium-cooled ceramic reflector (HCCR) blanket is adopted to be tested in ITER as Test Blanket Module (TBM). Currently, the design and R&D activities are mainly performed through the ITER TBM program in Korea. It is expected to demonstrate the major objectives of the breeding blanket: extraction of heat from burning plasma, tritium self-sufficiency and its integrity within the whole systems considering safety features. Although TBM program will provide very essential data and experience, there will be a considerable technical gap from DEMO blanket. This paper presents the technical gap in the main technical categories of the breeding blanket based on the experience of the development of HCCR TBM and breeding blanket. It is found that still about 60–70% of the DEMO blanket technology can be achieved through the HCCR TBM depending on the technology maturity level, and some ways to DEMO blanket from HCCR TBM are proposed. [ABSTRACT FROM AUTHOR]

The Lead-Lithium Ceramic Breeder Test Blanket Module (LLCB-TBM) is the Indian TBM representing the DEMO breeding blanket, to be installed in ITER radial port no-2 for testing in ITER. The conceptual design of LLCB TBM systems and their engineering design are under progress. The neutronic analysis and engineering design of LLCB TBM set (TBM + TBM Shield) is undergoing optimization. The LLCB TBS auxiliary systems; Helium cooling systems, Lead-Lithium Cooling System and Tritium Extraction Systems design are conceptualized and detail design is under progress. The system arrangements in port cell area, Tokamak Cooling Water System (TCWS) vault and Tritium building along with interface requirements have been worked out within the allocated space. LLCB TBS related R&D activities are under progress at Institute for Plasma Research (IPR), Gandhinagar in association with Bhabha Atomic Research Centre (BARC), Mumbai and Indira Gandhi Centre for Atomic Research, Kalpakkam. The major research and development areas are liquidmetal technologies, lithium ceramic pebbles, lead-lithium eutectic alloy, India specific Reduced Activation Ferritic Martensitic Steels (IN-RAFMS) development and fabrication technologies development. This paper will highlight the current LLCB TBM set and axillary systems design and status of R&D activities in various areas. [ABSTRACT FROM AUTHOR]

This paper describes the main acceptance criteria and required acceptance tests for the components of the six Test Blanket Systems to be installed and operated in ITER. It summarizes the guide-line toward the establishment of detailed test plans for the TBS, starting from the end-product at the ITER Members factories, and to generally define the type of tests that have to be performed on the ITER site after shipment, and/or prior to the systems final commissioning phase. [ABSTRACT FROM AUTHOR]