[en] A design window concept is developed for a He-cooled fusion reactor blanket and divertor design. This concept allows study of a parameter regime under which a possible design exists with different design requirements, such as allowable pumping fraction. The concept identifies not only the required parameter regime, but also investigates the robustness of the design, i.e., the validity of the design with change of design parameters and requirements. Some recent directions of helium cooled design for ITER and for divertor can also be explained by this design window concept

[en] Liquid breeding materials have received wide attention in the U.S. fusion reactor design studies over the past few years. There are major uncertainties of the tritium recovery system design over the constraints of tritium inventory, tritium containment, and recovery system costs. The key problems are identified, and the data required to resolve these issues are discussed. (author)

[en] The ARIES-ST (Spherical Tokamak) is to investigate the attractiveness of a low-aspect-device as the confinement concept for a fusion power plant. The key driven force of the ST design is caused by the center column conductor. The design selected is a water-cooled Cu normal conductor. This selection has a major impact on the blanket design and selection, tritium breeding and over-all power balance. The blanket selected is a dual coolant concept, partially decided by the characteristics of the center conductor. The final blanket design is modified from the dual coolant concept, which developed under the EC DEMO program. The reason for this selection and the design issues are summarized in this paper

[en] The technical issues, development problems, and required experiments and facilities for both solid and liquid breeder blankets have been investigated. The results have been used to develop a technical framework for a test plan that identifies the role, timing, and characteristics of major experiments and facilities. A major feature of this framework is the utilization of non-fusion facilities over the next 15 years, followed by testing in fusion devices beyond about the year 2000. Basic, separate-effect and multiple-interaction experiments in non-fusion facilities will provide property data, explore phenomena and provide input to theory and analytic modeling. Experiments in fusion facilities can proceed in two phases: 1) concept verification and 2) component reliability growth. Integrated testing imposes certain requirements on fusion testing device parameters; these requirements have been quantified. (author)

[en] Thermonuclear fusion power stations based on the deuterium-tritium reaction require breeding blankets to produce the tritium (T) fuel consumed in the plasma. This paper resumes the state-of-the-art of the T related technology from the initial T production in the lithium-bearing breeder material to a T stream which is ready for re-injection into the plasma. The remaining development issues are outlined and conventional techniques are confronted with advanced methods requiring more R and D effort but promising certain advantages in return

[en] A concept to recover tritium from lithium, based on a cold trap, has been developed as a part of the US contribution to ITER. The cold trap process can only reduce tritium concentration to about 400 appm, which is far above the ITER design goal of reducing tritium concentration in lithium to ∼1 appm. To achieve this lower goal, protium is added to the lithium to a concentration higher than the saturation concentration of the hydrogen isotope at the cold trap temperature. Thus, LiH and LiT will precipitate out together at the cold trap. The tritium from the cold trap can be recovered by heating the Li(H+T) to 600 degrees C for decomposition. The H and T then can be separated by cryogenic distillation process

[en] One of the critical issues of Flibe/V-alloy blanket with REDOX control by Be is a large tritium inventory in V-alloy structures. Among the possible solutions to this issue would be to control REDOX not by Be but by addition of MoF₆ or WF₆ enhancing the reaction from T₂ to TF. The present study investigated feasibility of this procedure by thermodynamic and neutronics calculations. Using the blanket dimensions of Force Free Helical Reactor (FFHR), tritium inventory in V-alloy structure and Flibe were estimated based on the calculated equilibrium partial pressures of T₂ and TF in various cases of REDOX control by MoF₆ or WF₆. Also carried out were neutronics examinations for the impact of Mo or W doping in the blanket. The results showed that the tritium inventory in the blanket area would be less than 100g at the TF level of 0.1 and 1 ppm in Flibe with addition of WF₆ and MoF₆, respectively. WF₆ doping is far more advantageous than MoF₆ doping for low activation purposes. (author)

[en] Blanket design is in progress in helical-type compact reactor FFHR-2. A localized blanket concept is proposed by selecting molten-salt Flibe as a self-cooling tritium breeder from the main reason of safety: low tritium solubility, low reactivity with air and water, low pressure operation, and low MHD resistance which is compatible with the high magnetic field design in force-free helical reactor (FFHR). Numerical results are presented on nuclear analyses using the MCNP-4B code, on thermal and stress analyses using the ABAQUS code, and heat exchange efficiency from Flibe to He. R and D programs on Flibe engineering are also in progress in material dipping-tests and in construction of molten salt loop. Preliminary results in these experiments are also presented

[en] Japan-US cooperation program TITAN (Tritium, Irradiation and Thermofluid for America and Nippon) started in April 2007 as 6-year project. This is the summary report at the midterm of the project. Historical overview of the Japan-US cooperation programs and direction of the TITAN project in its second half are presented in addition to the technical highlights. Blankets are component systems whose principal functions are extraction of heat and tritium. Thus it is crucial to clarify the potentiality for controlling heat and tritium flow throughout the first wall, blanket and out-of-vessel recovery systems. The TITAN project continues the JUPITER-II activity but extends its scope including the first wall and the recovery systems with the title of 'Tritium and thermofluid control for magnetic and inertial confinement systems'. The objective of the program is to clarify the mechanisms of tritium and heat transfer throughout the first-wall, the blanket and the heat/tritium recovery systems under specific conditions to fusion such as irradiation, high heat flux, circulation and high magnetic fields. Based on integrated models, the breeding, transfer, inventory of tritium and heat extraction properties will be evaluated for some representative liquid breeder blankets and the necessary database will be obtained for focused research in the future.

[en] The ARIES-RS fusion power plant design study is based on reversed-shear (RS) physics with a Li/V (lithium breeder and vanadium structure) blanket. The reversed-shear discharge has been documented in many large tokamak experiments. The plasma in the RS mode has a high beta, low current, and low current drive requirement. Therefore, it is an attractive physics regime for a fusion power plant. The blanket system based on Li/V has high temperature operating capability, good tritium breeding, excellent high heat flux removal capability, long structural life time, low activation, low after heat and good safety characteristics. For these reasons, the ARIES-RS reactor study selected Li/V as the reference blanket. The combination of attractive physics and attractive blanket engineering is expected to result in a superior power plant design. This paper summarizes the power core design of the ARIES-RS power plant study. (orig.)