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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&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.

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&D effort but promising certain advantages in return.

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.

The main feature of FFHR is force-free-like configuration of helical coils, which makes it possible to simplify the coil supporting structure and to use high magnetic field instead of high plasma beta. The other feature is the selection of molten-salt Flibe as a self-cooling tritium breeder from the main reason of safety. Collaboration works based on the LHD project have made great progress in the reactor studies by focusing on engineering aspects of the high magnetic field and Flibe system design. Encouraging positive results are shown on ignition access, mechanical stress in coils supporting structures, improvement in the blanket system including materials selection and tritium recovery. Critical issues on fundamental safety analysis and maintainability of reactor components are also discussed, and many subjects are pointed out as future works.

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.

The purpose of the ITER Blanket Testing Module is to test the operating and performance of candidate blanket concepts under a real fusion environment. To assure fuel self-sufficiency, the tritium breeding, recovery and processing have to be demonstrated. The tritium produced in the blanket has to be processed to a purity which can be used for refuelling. All these functions need to be accomplished so that the tritium system can be scaled to a commercial fusion power plant from a safety and reliability point of view. This paper summarizes the tritium processing steps, the size of the equipment, power requirements, space requirements, etc. for a self-cooled lithium blanket. This information is needed for the design and layout of the test blanket ancillary system and to assure that the ITER guidelines for remote handling of ancillary equipment can be met.

The ARIES-RS tokamak is a conceptual, D‐T-burning 1000 MWe power plant. As with earlier ARIES design studies, the final design of ARIES-RS was obtained in a self-consistent manner using the best available physics and engineering models. Detailed analyses of individual systems together with system interfaces and interactions were incorporated into the ARIES systems code in order to assure self-consistency and to optimize towards the lowest cost system. The ARIES-RS design operates with a reversed-shear plasma and employs a moderate aspect ratio (A4.0). The plasma current is relatively low (Ip11.32 MA) and bootstrap current fraction is high ( fBC 0.88). Consequently, the auxiliary power required for RF current drive is relatively low ( 80 MW). At the same time, the average

A concept to recover tritium from lithium, based on a cold trap, has been developed as part of the U.S. contribution to ITER. The cold trap process can only reduce the tritium concentration to about 400 appm, which is far above the ITER design goal of reducing the tritium concentration in lithium to about 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 °C for decomposition. The H and T then can be separated by a cryogenic distillation process.

The TITAN reversed-field-pinch (RFP) fusion-reactor study has two objectives: to determine the technical feasibility and key developmental issues for an RFP fusion reactor operating at high power density: and to determine the potential economic (cost of electricity), operational (maintenance and availability), safety and environmental features of high mass-power-density fusion-reactor systems. Mass power density (MPD) is defined as the ratio of net electric output to the mass of the fusion power core (FPC). The FPC includes the plasma chamber, first wall, blanket, shield, magnets, and related structure. Two different detailed designs TITAN-I and TITAN-II, have been produced to demonstrate the possibility of multiple engineering-design approaches to high-MPD reactors. TITAN-I is a self-cooled lithium design with a vanadium-alloy structure. TITAN-II is a self-cooled aqueous loop-in-pool design with 9-C ferritic steel as the structural material. Both designs use RFP plasmas operating with essentially the same parameters. Both conceptual reactors are based on the DT fuel cycle, have a net electric output of about 1000 MWe, are compact, and have a high MPD of 800 kWe per tonne of FPC. The inherent physical characteristics of the RFP confinement concept make possible compact fusion reactors with such a high MPD. The TITAN designs would meet the U.S. criteria for the near-surface disposal of radioactive waste (Class C, IOCFR61) and would achieve a high Level of Safety Assurance with respect to FPC damage by decay afterheat and radioactivity release caused by accidents. Very importantly, a “single-piece” FPC maintenance procedure has been worked out and appears feasible for both designs. Parametric system studies have been used to find cost-optimized designs. to determine the parametric design window associated with each approach, and to assess the sensitivity of the designs to a wide range of physics and engineering requirements and assumptions. The design window for such compact RFP reactors would include machines with neutron wall loadings in the range of 10–20 MW/m2 with a shallow minimum COE at about 18 MW/m2. Even though operation at the lower end of the this range of wall loading (10–12 MW/m2) is possible, and may be preferable, the TITAN study adopted the design point at the upper end (18 MW/m2) in order to quantify and assess the technical feasibility and physics limits for such high-MPD reactors. From this work, key physics and engineering issues central to achieving reactors with the features of TITAN-I and TITAN-II have emerged.

The TITAN reversed-field-pinch (RFP) fusion-reactor study has two objectives: to determine the technical feasibility and key developmental issues for an RFP fusion reactor operating at high power density; and to determine the potential economic operational, safety, and environmental features of high mass-power-density (MPD) fusion-reactor systems. Parametric system studies have been used to find cost-optimized designs. The design window for compact RFP reactors includes the range of 10–20 MW/m2. The reactors are physically small, and a potential benefit of this “compactness” is improved economics. The TITAN study adopted 18 MW/m2 in order to assess the technical feasibility and physics limits for such high-MPD reactors. The TITAN-I design is a lithium self-cooled design with a vanadium-alloy (V-3Ti-1Si) structural material. The magnetic field topology of the RFP is favorable for liquid-metal cooling. The first wall and blanket consist of single pass poloidal-flow loops aligned with the dominant poloidal magnetic field. A unique feature of the TITAN-I design is the use of the integrated-blanket-coil (IBC) concept. The lithium coolant in the blanket circuit is also used as the electrical conductor of the toroidal-field and divertor coils. A “single-piece” FPC maintenance procedure is used, in which the first wall and blanket are removed and replaced by vertical lift of the components as a single unit. This unique approach permits the complete FPC to be made of a few factory-fabricated pieces, assembled on site into a single torus, and tested to full operational conditions before installation in the reactor vault. A low-activation, low-afterheat vanadium alloy is used as the structural material throughout the FPC in order to minimize the peak temperature during accidents and to permit near-surface disposal of waste. The safety analysis indicates that the liquid-metal-cooled TITAN-I design can be classified as passively safe, without reliance on any active safety systems. The results from the TITAN study support the technical feasibility, economic incentive, and operational attractiveness of compact, high-MPD RFP reactors. Many critical issues remain to be resolved, however. The physics of confinement scaling, plasma transport and the role of the conducting shell are already major efforts in RFP research. However, the TITAN study points to three other major issues. First, operating high-power-density fusion reactors with intensely radiating plasmas is crucial. Second, the physics of toroidal-field divertors in RFPs must be examined. Third current drive by magnetic-helicity injection must be verified. The key engineering issues for the TITAN I FPC have also been defined. Future research and development will be required to meet the physics and technology requirements that are necessary for the realization of the significant potential economic and operational benefits that are possible with TITAN-like RFP reactors.

The requirements for vacuum pumping and fuel processing for deuterium/helium (D/3He) burns in the physics operating phase for the International Thermonuclear Experimental Reactor (ITER) were assessed. These burns are expected to have low fusion power (100 MW), short burn times (≤ 30 s), limited operation (2000 shots), and a fractional burn ∼0.3%. For the physics phase, the fuel processing system will include several units to separate deuterium and helium (activated charcoal bed, SAES getter and a Pd/Ag diffuser), as well as an isotopic separation system to separate 3He and 4He. The needed vacuum system's cryorption surface area may be as large as 10 m2 if the burn time is ∼ 200 s, the fractional burn is 100 MW.

The requirements of tritium technology are centered in three main areas, i.e., (1) fuel processing, (2) breeder tritium extraction, and (3) tritium containment. The gaseous tritium stream from the breeder tritium extraction system is significantly different from the plasma exhaust stream and, therefore, may have an important impact on the operation of the fuel processing system. For some blankets, such an aqueous solution blanket, the blanket tritium stream may dominate the fuel processing system in terms of component size and power consumption. The importance of the blanket interface to a fuel processing experiment, such as TSTA, has been identified. The initial work to define the blanket processing system, which is proposed to be added as part of TSTA, will be discussed here.