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We are proposing to connect a supercritical CO2 gas turbine (SCOT) power generator to the Operational Recovery of Separated Hydrogen and Heat Inquiry-2 (Oroshhi-2) nuclear fusion primary mo...

In 2013–2014, the FliNaK/LiPb twin-loop system Operational Recovery Of Separated Hydrogen and Heat Inquiry-2 (Oroshhi-2) was constructed at National Institute for Fusion Science as a liquid blanket...

This report introduces a current status of the continuous tritium recovery test campaign of PbLi droplets in vacuum, using vacuum sieve tray (VST). The campaign aims to verify the viability of VST method at a prototype design level. The following verification are to be performed. 1) Verify the steady state extraction efficiency of tritium using a 0.6 mm nozzle. The target is greater than 80%. 2) Verify the mutual interference effects on efficiency degradation by multiple droplets by comparing 1, 4, 7, and 19 nozzles. 3) Verify the reliability of VST in continuous operation. At least 24 h of non-stop operation without efficiency degradation is the primary goal, which can be extended up to 48 h. A dedicated extraction setup is to be installed March 2019 into the Oroshhi-2 PbLi experimental loop at the National Institute for Fusion Science (NIFS) Toki-Japan, followed by a stand-alone test with a short flowing period at Kyoto University which is already finished. Deuterium (D2), instead of tritium, is used for the experiment. Extraction efficiency is calculated by comparing the D2 concentration before-drop and after-drop. The amount of released D2 is also measured by a quadrupole mass spectrometer (QMS) and verified by the concentration difference. The PbLi temperature is between 375 ℃ and 450 ℃, and the flow velocity at the nozzle is between 1.5 m s−1 and 4.5 m s−1. The final report is scheduled to be completed by the end of 2019.

Heat removal techniques of various liquid divertor concepts have been widely studied. In this study, thermal mixing characteristics of the liquid metal film-flow with locally heated on the surface under the vertical magnetic field were experimentally investigated by using various types of obstacles as a vortex generator. Especially, the efficiency of heat transport was investigated by the thermocouples installed on the bottom-wall. The velocity distributions were measured using the electric potential probe rake on the downstream of the channel. A delta-wing, a hemisphere or a cubic obstacle was used as a vortex generator installed at the center of the bottom-wall. The experiments were conducted for laminar flow region where Re = 487–1583 and in the range of N (=Ha2/Re) = 0–149 in the vertical magnetic field in the acrylic rectangular duct. GaInSn alloy was used as a working fluid. According to the comparison of heat flux distributions on the bottom-wall, the entire distributions moved towards the upstream with increasing of the strength of the vertical magnetic field for all cases. This tendency means the heat removal from the free-surface of liquid metal film-flow to the bottom-wall can be shortly achieved in case of the relatively high vertical magnetic field.

Design studies on the helical fusion reactor FFHR-c1 have progressed. The main goal of the FFHR-c1 is to demonstrate one-year steady-state sustainment of the fusion plasma with self-produced fuel and electricity. The major radius of the plasma, R, is ˜10 m and the magnetic field strength at the plasma center, B, is ˜8 T. The magnet coil arrangement is basically similar to that of the LHD. The fusion gain during the one-year operation will be ˜15 with the fusion output of ˜370 MW. Three new technologies of the high-temperature superconducting magnet coils, the cartridge-type liquid blanket, and the fusible metal pebble divertor are adopted. Including these three, we have defined 22 important issues. The strategy to efficiently address the 22 issues has been also discussed. As a result, we propose a step-by-step approach by developing the FFHR-a1 (R ˜ 2.5 m and B ˜ 4 T) and the FFHR-b1 (R ˜ 3.6 m and B ˜ 6 T), before the FFHR-c1. The FFHR-b1 is a HElical Volumetric Neutron Source (HEVNS) driven by the neutral beam injection heating, and the FFHR-a1 corresponds to the pilot HEVNS.

Magnetic interaction between a tokamak reactor and its reinforced-concrete building has been analyzed using the analytical method and ANSYS electromagnetic code. Although it is better to use the non-magnetized material for the foundation of the air core tokamak building, it is interesting to study whether the magnetic material can be used in order to save the total cost of a fusion reactor. This study is motivated from the present ITER under construction in France. Assuming the iron plate approximation instead of discrete rebars, we analyzed the effect of the magnetized material on the magnetic field null regime in the plasma break down phase, x point location in the divertor operation, the magnetic force applied to the floor, and the induction effect on the iron floor. Although the effect of the magnetic material on the plasma performance is found to be not so crucial, the mechanical soundness of the floor due to the magnetic attractive force is most concern for machine safety.

In our previous work, the joint between oxide dispersion strengthened copper alloy (ODS-Cu), GlidCop® (Cu–0.3 wt%Al2O3) and tungsten (W) demonstrated superior fracture strength (∼200 MPa). This joint was fabricated by the direct brazing method between W and ODS-Cu using BNi-6 (Ni–11%P) filler material without any intermediate layer. This method was named as the improved or the advanced brazing technique. In the present study, deformation and fracture behavior of the joint after the three-point bending test was investigated. At first, it was found that the crack initiation points were dominantly in the W bulk, although it was not clear that the crack initiated from grain boundary or not. Secondly, the crack propagation proceeded mostly in the W bulk but tended to deflect towards the bonding layer. These results interpret the strength of the bonding interface is superior and the present bonding technique is applicable for severe environments such as a high heat flux divertor component on the fusion reactor. Based on the above physical and technological understanding, we successfully fabricated the large scale divertor mock-up which has twenty-eight plates of W with each size of 20 × 20 × 5 mm3.

The conceptual design of a helical fusion reactor was studied at the National Institute for Fusion Science in collaboration with other universities. Two types of the force free helical reactor (FFHR) are FFHR-d1 and FFHR-c1. FFHR-d1 is a self-ignition demonstration reactor that operates with a major radius of 15.6 m at a magnetic field intensity of 4.7 T. FFHR-c1 is a compact subignition reactor that aims to realize steady electrical self-sufficiency. Compared to FFHR-d1, FFHR-c1 has a magnetic field intensity of 7.3 T and a geometrical scale of 0.7. The location of the superconducting coils in both types of FFHR is based on that of the Large Helical Device (LHD). LHD has a major radius of 3.9 m. According to the design of LHD, the deformation must be within the required value to compensate for the accuracy of the magnetic field. According to this concept, the magnet support structure of LHD was fabricated using thick Type 316 stainless steel to impart sufficient rigidity. Thus, the stress of the magnet system of LHD is sufficiently below the permissible stress. In the case of FFHR, from the viewpoint of the reactor, a large access port is required for the maintenance of the in-vessel components. The mechanical design of the support structure is conceptualized by considering the basic thickness of the material and residual aperture space by referencing the mechanical analysis results. Details of the design concepts of LHD and FFHR-d1/FFHR-c1 as well as the results of mechanical analyses are introduced in this paper.

FFHR-d1A and c1 are the conceptual design of a helical fusion reactor. The positional relationship among superconducting coils, a pair of helical coils with two sets of vertical-field coils, are observed to be similar in both type of FFHR. Such a relation of coil configuration is based on the coil configuration of the Large Helical Device, which has been designed and constructed at the National Institute for Fusion Science. There is increasing demand to achieve an optimized coil configuration to anticipate improvements in plasma-confinement conditions. In this study, the structural design of FFHR based on the fundamental set of parameters of coil configuration is depicted, which satisfies the soundness of the structure. Further, the effects of the coil configuration parameters on the stress distributions are investigated. An effect of radius of curvature on a winding scheme of the helical coil is also discussed.

In this study, hydrogen permeation through Flinabe (equimolar LiF + NaF + BeF2 molten salt) including 0.1 wt% Ti particles is experimentally and analytically investigated. Ti particles mixed with Flinabe can suppress H2 permeation for a certain time when the upstream side is supplied with H2. The diffusion time lag after activation process is over 20 times larger than that before activation. H2 diffusion behavior in the Flinabe + Ti system is analyzed using an unsteady-state diffusion equation. The effective diffusivity in the Flinabe + Ti system is around 1/27 of that in the Flinabe system. It is confirmed that the activation of Ti particles is important for the overall permeation process in the Ti-Flinabe system.

We discussed with experiments the selective heating behavior of a mixture of molten salt and metal powder containing hydrogen in a single-mode cavity by 2.45 GHz microwave. The experimental device consists of two single-mode devices that intersected perpendicularly. In the common cavity at the two crossing single-mode devices, one microwave generates the maximum electric field of standing wave, and another microwave generates the maximum magnetic field of microwave. Thus, the sample located at the crossing can be selectively heated by the inputting electric field and/or the magnetic field. The frozen salt can be heated and melted by the electric field. Hydrogen was recovered from the metal powder in the molten salt by magnetic heating.

Sacrificial corrosion protection, a new use of Ti powder in molten salt blankets, is proposed. Redox reactions of mixed Ti powder into molten fluoride salts are predicted to suppress the TF corrosion of structural materials. Stainless Steel 430 (SS430) is immersed into FLiNaK containing HF at 823 K for 6 h. Before the immersion, the Ti powder either exposed to air or not exposed to air at all is mixed into FLiNaK. It has been found that either types of mixed Ti powder can suppress the HF corrosion of SS430.

The design of liquid metal cooling system for the liquid blanket concepts in magnetic fusion reactor requires the magnetohydrodynamics (MHD) friction loss coefficients for various pipes including pipes and bents to estimate the pressure drop of the whole blanket. The aim of this study is to reveal the MHD pressure drop of PbLi flow in a double-bended pipe. The experiments were conducted in the Oroshhi-2 loop having a super conducting magnet of 3T at NIFS in Japan. In this paper, the MHD pressure drops of the pipe under the magnetic field of 0.5T–3.0T, the ranges of Reynolds number (Re) of 2000–200,000 and Hartmann number (Ha) of 106–640 at the upstream, double-bended and downstream regions were measured. As the results, all the MHD friction loss coefficients in three regions were well correlated with the interaction parameter (Na = Ha2/Re) in the range of 0.1

Thermal and flow analyses on the cartridge-type molten salt blanket named the CARDISTRY-B (CARtridges Divided and InSerTed RadiallY – Blanket) for the compact helical fusion reactor FFHR-c1 have been carried out. As a result, it is shown that the maximum local temperature in the cartridges can be kept below 550 °C, which is an allowable temperature of the structure material, under the neutron wall loading condition of FFHR-c1. The CARDISTRY-B is composed of the tritium breeder (Breeding Blanket: BB) and the neutron shield (Shielding Blanket: SB). Both of them are toroidally segmented every two degrees. From a safety perspective, the molten salt FLiNaBe (melting point ∼305 °C) is chosen as the first option of the breeder material. In the case of FLiNaBe BB cartridge made of RAFM (Reduced Activation Ferritic/Martensitic steel), it is necessary to suppress the maximum temperature of BB cartridges to 550 °C or below. In this study, a finite volume method code considering beryllium (Be) pebbles inside a flow channel has been used to evaluate the cooling characteristics of a BB cartridge. By estimating the pressure drop through pebbles with the porous medium model, we have investigated how the temperature of BB cartridge change depends on the mass flow rate of FLiNaBe. The volumetric heating profile was modeled according to the neutronics calculation results obtained by the MCNP (Monte Carlo Neutron and Photon Transport) code.

The LHD-type helical fusion reactor FFHR has been studied to realize steady-state fusion power generation without a need for current drive and free from disruption. The conceptual design studies of FFHR are steadfastly progressing based on the presently ongoing experiments in the Large Helical Device (LHD). In order to enhance the attractive features of the base option of FFHR-d1A, which is similar to LHD, configuration optimization is being considered for FFHR-d1C. Slight modification of the helical coil trajectory gives an improved condition both for the plasma confinement and the MHD stability. In order to overcome the difficulty for construction and maintenance associated with the three-dimensional structure, innovative ideas are being explored for the superconducting magnet, divertor, and blanket. For the superconducting helical coils, the joint-winding method confirms a fast manufacturing process. The helical divertor is reexamined and practical feasibility is discussed. The maintenance method of the helical divertor and the helically-segmented breeder blanket is a serious issue and a plausible solution is proposed.

FLiNaBe is a promising liquid blanket material for a fusion reactor. Since hydrogen solubility for FLiNaBe is low, one concern is the permeation of bred tritium. In order to increase effective solubility of hydrogen, the addition of Ti powder was proposed. In this study, the influence of Ti addition on the behavior of tritium generated inside FLiNaBe by neutron irradiation was investigated. The samples of FLiNaBe and FLiNaBe mixed with Ti of 5 wt% and 0.5 wt% were prepared and the solid-state samples were irradiated by thermal neutrons at Kyoto University Research Reactor. The neutron-irradiated samples were heated to 700 °C in an Ar gas flow. The released tritium was collected by a water bubbler and the temporal variation of TF and HT release was observed. Initially, the most of tritium was released as HT from both FLiNaBe with and without Ti but eventually, the most of released tritium was HT after long time heating from FLiNaBe with Ti. This result suggests that the addition of Ti can decrease the release of TF very much. The amount of tritium released from FLiNaBe with Ti was lower than that from FLiNaBe without Ti. This result suggests that the addition of Ti can suppress the release rate of tritium from FLiNaBe.

In the future fusion reactors, ferritic steel is a realistic candidate to be used as a structural material of the blanket, because of the low activation characteristics after neutron irradiation which can reduce the issue of radioactive waste disposal. However, since the ferritic/martensitic steel is a ferromagnetic material, distortion of magnetic field produced by the magnet coils is a concern on the plasma confinement due to island formation and partial changes of magnetic surfaces. The conceptual design studies on the LHD-type helical fusion reactor (FFHR) are progressing at National Institute for Fusion Science (NIFS) by employing the heliotron magnetic configuration with continuous helical coils. In order to include the distortion of magnetic field produced by the ferritic steels in the blanket, the numerical calculation code ANSYS based on the finite element method (FEM) is incorporated. First, the precise structure of the magnet coils, vacuum vessel and blanket is input to establish the 3D modelling for the finite element calculation. Second, the magnetic field by assuming nonmagnetic materials in the blanket is calculated as a reference. In this phase, the magnetic field-line tracing code is combined to obtain the magnetic surfaces and their associated parameters, such as the rotational transform and magnetic well. Third, the nonmagnetic material is changed to ferritic materials and the new magnetic fields are calculated. Comparison of magnetic field properties by changing the permeability of the structural material is investigated.

A conceptual design for a helical fusion reactor is currently being undertaken by the National Institute for Fusion Science, Japan. The coil support structure is designed from the perspective of the allowable stress of the material. A continuous helical coil winding with a low temperature superconductor and a water-cooled divertor made of tungsten and copper alloy have been considered for use in this reactor, and it is defined as the basic option. Several flexible design options have also been proposed; these options solve existing issues in the basic option and they are treated as the challenging options. They can be implemented by modifying the structural components of the basic option. The structural design modifications need to be made in order for the challenging options and the mechanical soundness were investigated.

Oxidation control of candidate duct materials for liquid cooled blanket systems (i.e. SS316, SS316L, Inconel 600 and Hastelloy X) was studied to form a single composition Cr 2 O 3 layer on the surfaces for (i) making a stable and robust underlayer for ceramic coating fabrication by the metal organic decomposition (MOD) method and (ii) suppression of hydrogen permeation by the Cr 2 O 3 layer itself. In heat treatments of the bare plates at 700 °C in air of atmospheric pressure and air of reduced pressure of 5 Pa, oxidized Fe and/or Ni diffused to the outermost surfaces of the oxide layers and a single composition layer of Cr 2 O 3 could not obtained on the surfaces. However, the Cr 2 O 3 surface layers could be formed by a heat treatment of chrome plated plates at 700 °C in air of atmospheric pressure. MOD coatings were successfully fabricated on the stable Cr 2 O 3 surface layers. The preliminary evaluation of deuterium permeation through the Cr 2 O 3 covered plates showed that the permeation reduction factors (PRFs) of the Cr 2 O 3 layers would be 1–2 orders.

The use of bare Reduced Activation Ferritic Martensitic (RAFM) steels has been proposed for the first wall in a fusion reactor. Thus, it is necessary to understand the erosion and surface morphology changes under direct and simultaneous hydrogen and helium implantation of RAFM steels under fusion-relevant condition. Experiments were performed at ion energies of ∼1 keV over the temperature range of 500–900 K, consistent with the anticipated operating conditions of the blanket region. The irradiation fluence was 1 × 10 24 m −2 and the He% in the ion beam was 0.5%. Mass loss was measured before and after irradiation to determine erosion. Surface morphology changes were observed by SEM/FIB. RBS was used to quantify the near surface W concentration. From SEM images, surface roughening was observed at T > 750 K and complex surface structures formed with peak-to-peak variation in the ∼μm range. Mass loss increased with increasing surface roughening but the loss for simultaneous hydrogen-helium case was systematically higher compared to hydrogen-only case. A corresponding lower surface W concentration was measured for the simultaneous case, indicating Helium has an important role in W surface enrichment process.

This study evaluates convective and boiling heat transfer characteristics of a water impinging jet flow in porous media in order to remove the heat flux of 10 MW/m2 imposed to fusion divertors. The...

Dissimilar-metals joints between vanadium alloy NIFS-HEAT-2 (NH2) and nickel alloy Hastelloy X (HX) were developed by electron beam welding (EBW). If without filler, the joint fractured several minutes after the welding due to hard and brittle intermetallics formed in the weld metal (WM). Pure Ni filler with 0.2–1.0 mm in thickness decreased the content of intermetallics and eliminate hardening in the WM. However, there is always a hardening interlayer estimated as Ni2V and σ intermetallics, existed with thickness of 50 µm between NH2 base metal (BM) and WM. The hardening cannot be eliminated by annealing at high solution temperature of 1373 K. Aging at 723–973 K for 100 h further increased the hardening not only in the hardening interlayer but also in the WM. Pure Cu filler was also investigated. For the joint with 0.5 mm thick Cu filler, there are still hardening interlayer and hardening areas in the WM due to Ni2V and σ intermetallics. However, by increasing the Cu filler to 1 mm thick, the har...

H2 permeation behavior in a new molten salt of Flibe mixed with 0.5 wt% Ti powder is experimentally and analytically investigated. Ti powder included in Flibe can suppress H2 permeation for...

Surface morphology and hydrogen isotope retention of W specimen melted with vertical displacement event-like heat load and subsequent deuterium (D) plasma exposure were studied. Applied heat loads using electron beam without raster scanning were about 190 and 230 MW/m2 in heat flux and 0.08, 0.12 and 0.16s in duration. After the heat load application, specimens showed apparent melting spots with grain growth or dense micrometer scale convex structure. Cracks were observed only in the part with the convex structure. D retention in the melted part of specimens was not significantly larger than in the reference specimen despite large changes of surface characteristics. Keywords: Tungsten, Retention, Melting, Divertor, High heat flux, NRA

A systematic study of the temperature effect on the microstructure and the deuterium retention property in beryllium co-deposition layers has been carried out using W samples exposed to D + Be and D + Be + He mixture plasmas in the linear divertor plasma simulator PISCES-B. A deposition layer consisting of small grains of ∼10 nm with original hexagonal close-packed structure of beryllium was formed on a sample exposed to D + Be mixture plasmas at low temperature of 373 K. He seeding to the mixture plasmas was found to cause amorphization of the layer. In contrast, columnar structure consisting of Be 2 C due to crystal growth appeared at high temperature exposure cases of > 773 K both with and without He seeding. The formation of these deposition layers brought about a significant D retention. Assuming all retained D atoms uniformly distribute within the deposition layers, the D/Be ratios are estimated to be about 0.05 for the low temperature exposure case of 573 K, and to be about 0.01 even for the high temperature at 773 K. In addition, post-irradiation of 3 keV-D 2 + for Be deposited sample demonstrated that the hydrogen isotope retention behavior can be modified drastically, once the Be deposition layer is formed.

A new liquid metal divertor system named the REVOLVER-D (Reactor-oriented Effectively VOLumetric VERtical Divertor) is proposed for the helical fusion reactor FFHR-d1. The REVOLVER-D is composed of molten tin shower jets stabilized by internal flow resistances of wire/tape/chain. These shower jets are inserted into the ergodic layer surrounding the main plasma. Tin is selected as the liquid metal because of its low melting point, low vapor pressure, low material cost, and high safety. The liquid metal pumps, cryopumps, and turbo molecular pumps are installed in the central vacuum vessel connected to the main vacuum vessel via 10 inner ports equipped with maze neutron shields. Central solenoid coils made of high-temperature superconductors are installed inside the central vacuum vessel to shield the pumps from the strong magnetic field. The REVOLVER-D has a good possibility to satisfy important characteristics required for the divertor system in a fusion reactor, that is, high heat load tolerance, high maintainability, sufficient vacuum pump speed, high level of safety, and a small amount of radioactive wastes.

The lead-lithium eutectic alloy (PbLi) is a promising coolant for nuclear fusion reactors. Although PbLi flow and heat transfer in a nuclear fusion reactor condition needs to be comprehended, no ve...

A new closed helical divertor configuration for efficient particle control and reduction of the heat load on the divertor plates is proposed. The closed divertor configuration practically utilizes an ergodic layer and magnetic field line configuration on divertor legs in helical systems. For optimization of the design of the closed divertor, the distribution of the strike points is calculated in various magnetic configurations in the Large Helical Device (LHD). It suggests that the installation of the closed divertor components in the inboard side of the torus under an inward shift configuration (R ax =3.60m) is the best choice for achieving the above two purposes. This divertor configuration does not interfere with plasma heating and diagnostic systems installed in outer ports. The prospect of the closed divertor configuration to a helical fusion reactor is investigated using a three-dimensional neutral particle transport simulation code with a one-dimensional plasma fluid calculation on the divertor legs. The investigation shows efficient particle pumping from the inboard side and reduction of the heat load due to the combined effect of the optimized closed divertor geometry, ergodized divertor legs, and low electron temperature in the ergodic layer. It indicates a promising closed divertor configuration for helical fusion reactors.

Accelerator-driven neutron sources have been developed worldwide for boron neutron capture therapy (BNCT) instead of nuclear reactors. Nagoya University is currently developing a compact DC accelerator driven BNCT system using a sealed lithium (Li) target. The target receives a high heat load by a proton beam, and the heat then requires efficient removal. This study developed a high-efficiency heat removal technique that uses V-shaped staggered ribs on the heat removal side of the flow channels. We performed experiments using an electron beam to confirm the technique performance. The results showed double the heat removal performance compared to the smooth surface channel. The value expected to sufficiently remove 6.6 MW/m2 of heat load provided by the proton beam at the Nagoya University neutron source for BNCT.

In order to achieve high power generation efficiency, facility compactness, high safety, and high coexistence with fuel tritium in the 3 G Wth class FFHR(Force Free Helical Reactor) power plant, optimization, performance and operation scenario of the power generation system of the power plant model using the axial flow type uniaxial supercritical CO2 gas turbine power generation system were examined. As a result, the following conclusions were obtained. (1) The most efficient and compact axial flow uniaxial design supercritical CO2 gas turbine power generation system suitable for 3 G WthFFHR power plant has a maximum heat capacity of 1.5GWth×2, and the gas turbine conversion rate at that time was evaluated as 46.9 %(= 1,267MWe/2,700MWth). Therefore, split connection with two power generation systems is optimal for a 3 G Wth class FFHR power plant. Here, the operating speed of the turbo equipment was set to 3600 rpm in order to adapt this plant model to the power supply system in Japan. (2) Simultaneous supply of 818MWe/300 s electricity and 300MWth/30days heat is required at startup of the FFHR. (3) The in-house power during steady operation under the self-ignition condition was reduced to about 31.7MWe, and the amount of power generated at the transmission end was evaluated as 1,234MWe.

Optical reflectivity measurement is proposed as a convenient diagnostics method of surface modifications in plasma facing materials (PFMs) and its applicability to a plasma confinement device is evaluated. The optical reflectivity for mirror polished stainless steel (SUS316L) exposed to the Large Helical Device (LHD) plasma was examined with a spectroscopic ellipsometer. In-situ measurement of the reflectivity change under the glow discharge was also performed using a super continuum white laser. The detectable change of the reflectivity was observed and the level of the degradation depended on the location, exposure period and employing wavelength. Microstructure observation of the mirror surface revealed that the behavior of the reflectivity change correlated closely with the ion-induced damages and the depositions thickness in the vicinity of the sample surface. Thus, the optical reflectivity measurement is considered to be a possible method for convenient in-situ diagnostics for PFMs.

The high-temperature superconducting (HTS) option is employed for the conceptual design of the LHD-type helical fusion reactor FFHR-d1. The 100-kA-class STARS (Stacked Tapes Assembled in Rigid Structure) conductor is used for the magnet system including the continuously wound helical coils. Protection of the magnet system in case of a quench is a crucial issue and the hot-spot temperature during an emergency discharge is estimated based on the zero-dimensional and one-dimensional analyses. The number of division of the coil winding package is examined to limit the voltage generation. For cooling the HTS magnet, helium gas flow is considered and its feasibility is examined by simple analysis as a first step.

Segmented fabrication of high-temperature superconducting (HTS) coils has been proposed for the helical fusion reactor FFHR-d1; the joint-winding and “remountable” (demountable) helical coils concept are the primary and advanced options, respectively. This paper introduces recent research activities on a bridge-type mechanical lap joint of HTS Stacked Tapes Assembled in Rigid Structure (STARS) conductors, particularly to be used for joint winding of the helical coils, and also discusses some technical issues to develop a reliable joint. First, the required joint performance is discussed from the viewpoints of operation cost and mechanical behavior, taking the present design of joint winding of the helical coils for the FFHR-d1 into consideration. Then, mechanical and electrical characteristics of a prototype STARS conductor joint, which achieved a joint resistance of 1.8 nΩ at a current of 100 kA, are analyzed based on small- and largescale joint tests. It is suggested that the joint performance of the prototype joint can be improved more, according to tests of small-scale conductor joints. Finally, the study introduces a fabrication procedure for a reliable joint. We need to develop nondestructive testing for a joint at room temperature. For this purpose, we also perform an X-ray computed tomography inspection as a first step. The inspection can roughly judge whether a fabricated joint is acceptable.

In order to establish a liquid blanket using a Li-containing molten fluoride salt, the corrosion rate and mechanism of pure Fe and JLF-1 steel in FLiNaK (LiF-NaF- KF = 46.5-11.5-42.0) was investigated by one of the electrochemical methods, the potential step method. Simulating the TF breeding in the blanket, HF gas was injected into FLiNaK at 773 K. The partial pressure of HF gas (pHF) was varied from 8.59 to 2.46 × 103 Pa. The current-potential curves of pure Fe and JLF-1 steel were measured in an electrochemical cell and fit by the Butler-Volmer equation with limited current density. The corrosion current density (icor) of pure Fe increased with pHF as i c o r = 0.6 p H F . The corrosion rates of pure Fe and JLF-1 steel were close. In order to keep the uniform corrosion rate of JLF-1 steel below 0.1 mm/year in the blanket, HF concentration in FLiNaK is required less than 1.2 × 10−4 molHF/m3. The cathodic current density (iLc) was proportional to pHF as i L c = - 0.11 p H F , and, therefore, the rate-limiting process of HF corrosion of pure Fe is predicted to be HF or H+ diffusion in FLiNaK.

Dissimilar-metals bonding between V–4Cr–4Ti and Hastelloy X alloy was investigated by explosive welding, which is essential to connect coolant pipes of Flibe blanket and external components such as tritium extractor and heat exchanger in fusion reactor. The nano-indentation hardness and corresponding microstructure at the interface are evaluated. After the explosive welding, a characterization of the typical wavy interface indicates good weldability. Discontinuous vortex regions appear at the interface and reveal obvious hardening up to 806 HV. Microstructural analysis indicates that the interlayer consists of Ni37.1V21Cr20.5Fe15.4Mo4.4Ti1.0 Co0.6 with a fcc lattice structure and a high density of dislocations, and no intermetallic forms. Based on the compositions, the formation of solid solution was predicted by the entropy and enthalpy of the multi-component alloy system, despite the phase diagrams of binary alloys of the constituent elements indicating the formation of intermetallics. Study of hardening mechanism at the interlayer indicates that obvious hardening is mainly attributed to solid-solution hardening and dislocation hardening.

A novel fabrication method for a divertor heat removal component with tungsten (W) armour and copper alloy heat sink was newly developed and named as the “Advanced Multi-Step Brazing (AMSB). The basic principle of multi-step brazing is to apply the “advanced brazing technique” repeatedly during the manufacturing process of the single divertor heat removal component. The advanced brazing technique was originally developed by our previous work for joining between oxide dispersion strengthened copper alloy (ODS-Cu) and W with BNi-6 (Ni–11%P) filler material. The applied ODS-Cu was GlidCop® (Cu-0.3wt%Al2O3). One of the possible examples of the AMSB fabrication process can be considered as follows. First, an appropriate cooling flow path channel is processed into a GlidCop® heat sink. Then, the flow path channel is sealed in a leak tightness condition with a lid made by GlidCop®. The leak tightness joint between GlidCop® (GlidCop®/GlidCop®) can be realized by application of the “advanced brazing technique.” Second, in order to facilitate the weldability to connect the other cooling pipe system, the sleeves made by stainless steel (SUS) are jointed on the interface edge of the flow path channel of the GlidCop® heat sink with a leak tightness condition by the “advanced brazing technique.” Finally, W armour is jointed on the GlidCop® heat sink also by the “advanced brazing technique.” In this study, the mechanical strength of the SUS/GlidCop® joint was confirmed by the three-point bending test. Then, AMSB divertor mock-up with leak tightness condition was successfully produced.

In the Force Free Helical Reactor FFHR-d1 (R∼15.7 m, a∼2.5 m, Bo ∼4.5 T, 〈β〉∼5 % and the fusion power of 3 GW) [1, 2], it is demonstrated that the thermally unstable operation is better to achieve the higher density and lower temperature plasma (n(0)∼ 9 × 1020 m−3 and T(0)∼7 keV) than the impurity injection method. One drawback of the thermally unstable operation is a possibility of the thermal runaway when fueling systems have failures. However, we have found that the inherently safe function exists owing to the plasma outward shift during the thermal runaway. Thus we continue to pursue this high-density operation scenario for FFHR-d1. Preliminary estimation of the alpha energy loss fraction of ∼3 % is obtained for a broad density profile in the high-density ignition regime.

In this study, the effect of ELM-like thermal shocks on He induced surface morphology were investigated. A W sample was exposed to pure He plasma. He ion incident energy, flux and fluence were 80 eV, 1 × 1022 /m2/s and 3 × 1025 /m2, respectively. Irradiation temperature were approximately 470 and 1100 K. Then, thermal shocks were applied on the sample using focused electron beam. The peak heat flux, pulse duration and base temperature were 500 MW/m2, 500 μs and R.T., respectively. After the thermal shocks, He induced morphologies such as holes and periodic undulations were flatten completely at the region exposed to the highest heat flux. At the peripheral regions of the electron beam spot, the hole density increase or partial flattening of morphologies were observed. These results suggested that in order to anticipate surface morphology with He irradiation and ELMs, the peak temperature should play a more important role than base temperature. Keywords: Tungsten, Divertor, He induced surface morphology, Thermal shocks

MgB 2 is one of the “low activation superconductors”, because the half-life of the induced radio-activity on the MgB 2 is much shorter than that of Nb-based superconductors such as NbTi, Nb 3 Sn and Nb 3 Al. MgB 2 wire would become an alternative material for Nb-Ti wire for the Poloidal Field and correction feedback coils, if the critical current density ( J c ) around the magnetic field of 4 T could be improved. We found that addition of Cu using the compound Mg 2 Cu was an effective method to enhance J c and succeed to fabricate 100 m length of MgB 2 wire with Cu addition. J c of the 100 m long MgB 2 with Cu addition was comparable to that of a NbTi wire, and the small solenoid coil using Cu added MgB 2 wire reached a central magnetic field of 2.62 T. We found that MgB 2 wire is an alternative material to commercial NbTi wires under the hard neutron irradiation in a fusion reactor. Furthermore, we also fabricated a Cu added MgB 2 wire using isotope pure boron-11 ( 11 B) powder as the boron source to reduce the radioactivity. The critical temperature ( T c ) of the MgB 2 wire using 11 B powder was found to be 37 K, and T c was found to be changed by different boron materials.

Performance and applicability of hydride shielding materials are investigated in the helical reactor FFHR-d1 design. Performance of ZrH2 and TiH2 in fast neutron shielding are close to that of WC, ...

The hydrogen solubility in a FLiNaK molten salt mixed with Ti powder was investigated. A hydrogen-soluble metal powder mixed with a molten salt can increase the effective hydrogen solubility of the molten salt, which is currently a major disadvantage of molten salts. A significant increase in hydrogen solubility was observed, even with a mass fraction of Ti powder of only 0.1 wt%. The increase of hydrogen solubility was so large that a vanadium alloy would be compatible with the Ti powder/molten salt mixture, unlike typical molten salts that result in an unacceptably large tritium inventory in the vanadium alloy. In addition, contamination of the Ti powder by oxidation suppressed the hydrogen uptake and release capability. Controlling the purity of the molten salt was found to be one of the key issues for the metal powder mixture concept.