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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.

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.

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.

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.

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.

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.

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.

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.

The present study investigated tensile properties of F82H steel after aging at 400–650 °C for 100,000 h (100 kh), and discusses the mechanisms for change in tensile properties due to the long-term aging. Tensile strength was degraded during the aging. The degradation of yield strength and ultimate tensile strength was 50 MPa or less at 550 °C and below in both room temperature and high temperature tests. Above 550 °C, yield strength and ultimate tensile strength decreased with increasing aging temperature. On the other hand, no degradation of ductility was detected after the aging. Most of the degradation of the tensile strength could be attributed to loss of solid solution hardening by W due to precipitation of Laves phase (Fe2W) at 550 °C. However, other mechanisms, such as coarsening of martensite structure and recovery of dislocations, should be taken into account to explain additional degradation above 550 °C.

The heat flux at the divertor in a fusion reactor is considered to have a peak of >10 MW/m 2 . In a design study of the helical reactor FFHR-d1, the feasibility of employing a copper alloy for divertor cooling pipes was investigated; however, radiation in the divertor area would quickly damage the copper alloy. The neutron load on the divertor can be reduced by a blanket arrangement; nevertheless, in the present divertor structure, irradiation damage of materials on the inboard side of the torus remains relatively high. If the divertor could be moved to an area receiving much less radiation, then the lifetimes of divertor materials should increase. In this paper, a novel divertor structure is introduced in which the coil-support structure is modified to create a region receiving relatively low amounts of radiation without changing the geometry of the helical or vertical field coils. Using this proposed design would increase the lifetime of the copper alloy in divertor components to more than an estimated six years. In addition, the divertor could be accessed from either the upper or lower sides of the device, simplifying maintenance.

NIFS launched the Fusion Engineering Research Project (FERP) in preparation for DEMO by starting the redesign of the LHD-type helical reactor FFHR-d1. In the first round, the main parameters were selected. The second round is preparing detailed three-dimensional (3D) design of the superconducting magnet support structures, and 3D neutronics analyses, where the diverter targets can be efficiently shielded from fast neutrons. A new Flinabe blanket mixed with metal powder was proposed. Fabrication of helical coils by connecting half-helical-pitch segments of 100 kA-class YBCO high-temperature superconductors is proposed as a promising method. Also in progress is improvement of the first round of the core plasma design, ignition start-up analyses, and fueling scenario. As a consequence, a multi-path strategy on FFHR-d1 has been introduced with versions of -d1A, -d1B, and -d1C, where design flexibility is expanded to include subignition with options FFHR-c1 for “before demo, compact, and component-test.”

Plasma operation control of the Large Helical Device (LHD)-type helical fusion reactor, FFHR-d1, was examined by using a quasi-one-dimensional particle balance calculation code. It was found that feedback control of the fusion power, which is proposed in the past study with a zero-dimensional model, was difficult because of the time delay between the pellet injection and response of the fusion power due to shallow pellet deposition. While the achievable parameter range of the fusion power at a steady state is limited by the pellet injection condition and magnetic field strength, smooth ignition access and steady-state sustainment can in principle be achieved via feedback control of the line-averaged electron density by manipulating the pellet fuelling rate and maintaining adequate control of the external heating power.

Hydrogen solubility in liquid Li–Na alloys (20, 50, and 90 at% Li) at 673, 773, and 873 K was investigated by adding and extracting hydrogen through the wall of a capsule made of pure iron. Hydrogen solubility in the Li90Na10 alloy was almost similar to that in pure Li while for Li50Na50 and Li20Na80, it was one and two orders of magnitude smaller, respectively. No obvious difference in temperature dependence was observed, similar to the case for pure Li. Li–Na alloy can thus be a potential liquid breeding material with almost all of the advantages of the Li blanket system and free of most of its drawbacks.

Neutron transport calculations with a three-dimensional model of the helical reactor FFHR-d1 have been performed for the accurate analysis of neutronics environment in the divertor areas. Based on the obtained neutron spectra, magnitudes of irradiation damage, contact dose rates and decay heat have been evaluated mainly for Fe, W and Cu. Since divertors can be placed behind radiation shields in helical reactors, magnitudes of damage and radioactivation at the outboard divertors are almost two orders lower than those at blanket first walls. Cu materials could be used as a cooling channel material of the outboard divertors. In contrast, the magnitudes at the inboard divertors are only one order lower compared with those at the first walls due to the limited space at the inboard side. Damage on Cu is evaluated to be ∼10 dpa after 6 years operation. Further efforts in divertor development and reactor design could suppress the magnitude of damage to less than half for adoption of Cu materials for the inboard divertors.

FFHR-d1 is a conceptual design of the helical reactor being developed at the National Institute for Fusion Science. The maintenance of in-vessel components is very important for the fusion demo reactor. In addition, sufficient pathways are needed for the divertor exhaust. To implement these, the vacuum vessel, coil support structure, and cryostat require large apertures. However, the coil support structure has to be sufficiently rigid to remain within soundness and deformation limits. A design combining the structural components in the FFHR-d1A was developed from mechanical and thermal viewpoints. Consequently, components having a sufficiently large port area were provided. An investigation of the maintenance and exhaust schemes has been planned on the basis of this fundamental design.

The joint of the oxide dispersion strengthened copper alloy, ODS-Cu (ODS-Cu/ODS-Cu), was prepared. The applied ODS-Cu was GlidCop® (Cu-0.3 wt%Al2O3). The joint procedures were based on the brazing technique between GlidCop® and tungsten (W) with BNi-6 (Ni-11%P) filler material. This method was performed without compressive load and without any intermediate layer, and was named the advanced brazing technique. In this study, first, the joint GlidCop®/GlidCop® was prepared without any compressive load. The joint was subjected to the three-point bending test, the Vickers hardness test, microstructure observation and the leak tightness test. The yield strength of the joint was approximately one-half of that of the GlidCop® bulk. In addition, the joint was not leak tightness against fluids. Next, the joint was prepared with compressive load of ˜0.54 MPa which was applied in the direction perpendicular to the joint interface. The yield strength of the joint was almost as high as that of the GlidCop® bulk. Furthermore, the joint achieved leak tightness against fluids. The special feature of this joint method is that a limited volume of the GlidCop® surfaces near the bonding interface are melted during the bonding heat treatment phase at 960 °C due to the eutectic reaction of the Cu-P system. Then, after finalizing the solidification, a tight bonding layer are created without any cavities and cracks. This phenomenon seems to resemble a micro scale welding. The present brazing technique can promise fabrication of a high heat flux component with a complex shaped fluid flow path system with leak tightness.

We present the newly discovered control algorithm for the heating power during ignition access to the thermally unstable ignition regime in the FFHR helical reactor. We used the fact that the thermally unstable and stable ignition regimes on Plasma Operating Contour map (POPCON) have an inverse relationship with respect to the density and temperature. As the density limit restricts the operating point to the thermally stable regime, the postulated temperature limit restricts the operating point to the thermally unstable regime. As a result, the heating power for feedback control is expressed by the measured temperature and power balance terms such as alpha heating and bremsstrahlung loss powers. The external heating power thus obtained is applied during the low-density NBI penetration phase, and then switched off automatically after over-passing the saddle point by alpha-heating power, reaching high-density ignition regime. This control algorithm is robust to various perturbations of the confinement factor, impurity content and so on. Feedback control algorithms thus established for both fueling and heating makes the thermally unstable operation more realistic in FFHR.

The D 2 + fluence dependence on deuterium (D) retention was studied to clarify the D retention mechanism in tungsten. The additional D desorption stage was observed around 660 K in the TDS spectrum for a sample implanted with D 2 + up to the fluence of 10 23 D + m −2 , which desorption stage was not observed the D 2 + implanted sample with the fluence less than 10 22 D + m −2 . The TEM observation showed that the highly dense voids were formed in tungsten by D 2 + implantation with the fluence of 10 23 D + m −2 , considering that the D would be trapped by voids. To understand the D trapping by voids in C + implanted tungsten, C + –D 2 + sequential implantation experiments at various C + implantation temperatures were performed. It was found that the amount of D desorbed around 560 K was increased by increasing the C + implantation temperature. The formation of the voids was observed with increasing the C + implantation temperature by TEM, indicating that the increase of D desorption around 560 K was caused by the formation of voids. However, the desorption temperature of D trapped by voids in C + implanted sample was lower than that in D 2 + implanted one. TEM observation and XPS measurement indicated that this difference was caused by the increase of void size and/or the presence of implanted carbon.

A conceptual design for the helical DEMO reactor FFHR-d1 is under development at the National Institute for Fusion Science. In the large helical device (LHD)-type configuration of FFHR-d1, the space between the plasma boundary and the helical coil is limited. Many components are required to be installed at definite geometric positions within this space. The position of each component changes owing to variations in the surrounding conditions, such as temperature, electromagnetic forces, and support methods. This paper proposes a radial-build design concept; the application of this concept ensures constant geometric positions of the components and the gaps between them during normal operation and the construction/maintenance phase. On the other hand, large apertures are required for the coil support structure for the maintenance of blanket and divertor systems. This paper presents a structural design procedure for the coil support structure with large apertures and an LHD-type gravity support post. The analytic results show that the stress level of the structure is lower than the permissible limit.

A Flibe blanket has been proposed to be used in FFHR. Since Flibe has poor heat transfer performance, heat transfer promoter is required, and a sphere-packed pipe (SPP) has been proposed to enhance the heat transfer performance in the Flibe blanket. In this paper, the fluid flow and heat transfer characteristics in the SPP is evaluated numerically using a k–ɛ turbulent model for the flow field and an algebraic model for the thermal field. As a result, it was shown that bypass flows in the SPP play a significant role in heat transfer. Also it is thought that the turbulent energy can strongly affect heat transfer performance.

An experimental study was performed to investigate hydrogen transport under conditions of gas bubbling into the liquid breeders of Li, Pb–17Li and Flinak. Gas of H2 or high-purity Ar at 0.1 MPa was injected into the liquid breeders using an I-shape nozzle. Then, the H2 partial pressure in the cover gas under transition state was measured by a solid electrolyte sensor. Variations of the partial pressure in the exhaust gas were compared with a numerical model to describe hydrogen transfer through the bubble–liquid interface. The experimental rates of hydrogen transport in the liquid breeders were correlated in terms of a mass-transfer coefficient for the hydrogen transport under gas bubbling. The overall mass-transfer coefficient for H2 charging was compared with that for Ar purging. The coefficient for the recovery was about 10 times lower than that of the charging in the respective cases of the Pb–17Li and the Flinak.

A new design activity is under way for a helical type DEMO reactor FFHR-d1. The first stage of the activity involves the fundamental issues related to three-dimensional blanket design: (1) the minimum blanket space required for reactor parameter decisions, (2) the support method for the helical blanket system, and (3) the blanket module design. Investigations have been performed with neutronics and mechanical finite-element method calculations. Neutronics investigations indicate that a tungsten carbide radiation shield could reduce the minimum blanket space requirement by ∼30 cm at the inboard region of FFHR-d1 compared with the blanket space of ∼100 cm in the previous FFHR2 design. The investigations also showed that main shielding materials, ferritic steel and B4C, could be used separately in a two-layered shielding configuration. The ferritic steel layer of the radiation shield is considered suitable to support the helical blanket system instead of relying on a thin vacuum vessel of the helical reactor. A size of a blanket module for a replacement process and the preferable cooling channel direction under a magnetic field are also discussed.

Based on high-density and high-temperature plasma experiments in the large helical device (LHD), conceptual design studies of the LHD-type helical DEMO reactor FFHR-d1 have been conducted by integrating wide-ranged R&D activities on core plasmas and reactor technologies through cooperative researches under the fusion engineering research project, which has been launched newly in NIFS. Current activities for the FFHR-d1 in this project are presented on design window analyses with designs on core plasma, neutronics for liquid blankets, continuous helical magnets, pellet fueling, tritium systems and plasma heating devices.

A new concept of a fusion reactor system, MFE-IFE cooperative system, is proposed. This concept combines the merits of a small-size MFE reactor and a dry-wall IFE reactor and aims at sufficient amount of tritium production and electricity generation without advanced technology. Design window analysis shows a NIF-scale (5 m chamber radius) dry-wall laser fusion reactor with a ∼1 GWth fusion output and net tritium breeding ratio (TBR) of 1.74 can sustain an MFE power plant with a fusion power of 3 GWth and net TBR of 0.96. Although more detailed quantitative analyses are required, this concept can be a possible solution for a simultaneous achievement of tritium self-sufficiency and significant net electricity generation.

The behavior of hydrogen retention and the change of chemical states of boron film exposed to hydrogen plasma in LHD were investigated. The sample was prepared in LHD, and atomic concentrations for the boron film after hydrogen plasma exposure were changed from 75% for boron, 15% for carbon and 8% for oxygen to 53%, 18% and 22%, respectively. B C bond was a major chemical state of the boron film after hydrogen plasma exposure, although abundance of B B bond was the highest before the plasma exposure. Total hydrogen retention measured by TDS was evaluated to be 1.7 × 1020 H m−2, and the retentions of hydrogen as B H B, B H and B C H bonds were, respectively, 4.8 × 1019, 7.2 × 1019 and 5.2 × 1019 H m−2. It was concluded that the hydrogen retention could be estimated by taking account not only of chemical states of impurities, but also of hydrogen depth profile.

A new method named direct profile extrapolation (DPE) has been developed to estimate the radial profiles of temperature and density in a fusion reactor. This method directly extrapolates the radial profiles observed in present experiments to the fusion reactor condition assuming gyro-Bohm type parameter dependence. The magnetohydrodynamic equilibrium that fits the experimental profile data is used to determine the plasma volume. Four enhancement factors for the magnetic field strength, the density, the plasma beta, and the energy confinement are assumed. Then, the plasma size is determined so as to fulfill the power balance in the reactor plasma. The plasma performance can be measured by an index, C exp , introduced in the DPE method. The minimum magnetic stored energy of the fusion reactor to achieve self-ignition is shown to be proportional to the cube of C exp and inversely proportional to the square of magnetic field strength. Using this method, the design window of a self-ignited fusion reactor that can be extrapolated from recent experimental results in the Large Helical Device (LHD) is considered. Also discussed is how large an enhancement is needed for the LHD experiment to ensure the helical reactor design of FFHR2m2.

Behaviors of hydrogen isotope retention and damages in tungsten and SS-316 with simultaneous C+–D2+ implantation were compared to those with only D2+ implantation using X-ray photoelectron spectroscopy (XPS), Thermal desorption spectroscopy (TDS), glow discharge-optical emission spectroscopy (GD-OES) and Transmission electron microscopy (TEM). The total D retention for SS-316 with only D2+ implantation was about 45% as large as that for tungsten. The D retention for simultaneous C+–D2+ implanted tungsten and SS-316 clearly increased as a factor of 1.7, which is almost the same among these samples. The density of dislocation loops was enhanced by the simultaneous C+–D2+ implantation, indicating the D trapping site would be produced by C+ implantation. As for the D desorption temperature, small shift toward lower temperature side was found for SS-316 compared to tungsten, indicating the D trapping energy by dislocation loops and grain boundary for SS-316 is lower than that for tungsten.

The electrical insulating coating on the blanket components such as ducts and walls is an attractive concept for reducing the Magneto Hydrodynamic (MHD) pressure drop. Erbium oxide (Er 2 O 3 ) is a promising candidate coating because of its high stability in liquid lithium and high electrical resistivity according to the results of Er 2 O 3 bulk and Physical Vapor Deposition (PVD) thin film. We have investigated the Metal Organic Chemical Vapor Deposition (MOCVD) process for the large area and complicatedly shaped Er 2 O 3 coating. The Er 2 O 3 insulator coating formation on the various metal disk substrates was successfully carried out. The crystallinity of the Er 2 O 3 coating on the metal substrate increased with the decrease in the surface roughness of the metal substrate and, thus, the crystallinity of the coating can be improved by reducing the roughness of the substrate surface. Furthermore, the Er 2 O 3 coating into the interior surface of the honing SUS pipe, whose interior surface was polished by rotating grinding and brush, was formed stably through the MOCVD process.

W coating by vacuum plasma spray process and Cr coating by chromizing process were performed on fusion low activation materials, JLF-1 ferritic steel and NIFS-HEAT-2 vanadium alloy. The present study discusses feasibility of the coatings as anti-corrosion coating against fluoridation in Flibe for fusion low activation materials. Coatings were characterized by microstructural analysis and examination on chemical stability by corrosion tests. The corrosion tests were conducted with H2O–47% HF solution at RT and He–1% HF–0.06 H2O gas mixture at 823 K to simulate fluoridation and oxidation in Flibe. The coatings presented suppression of fluoride formation compared with JLF-1 or NIFS-HEAT-2, however weight loss due to WF6 formation was induced, and much Cr2O3 was formed.

The transport of hydrogen (H 2 ) in Flinak was investigated using hydrogen sensor made of proton conductive solid electrolyte. The test in Flinak was performed after the validity of the sensor was checked in the test in Ar–H 2 gas mixture with known hydrogen partial pressure. The sensor showed stable output in the Flinak at homogeneous temperature condition. The electromotive force (EMF) of the sensor agreed with theoretically estimated EMF. The sensor however showed the EMF much different from the theoretically estimated EMF when there was temperature difference between the free surface and the immersed sensor position. The temperature dependence of the Henry's law which controlled the hydrogen transport in Flinak was investigated from the sensor EMF. It was indicated that the Henry's constant for H 2 in Flinak had the negative temperature dependence, which agrees with the previous study.

The first review is presented for the recent stellarator designs of FFHR, HSR, ARIES-CS, which are based on the experimental device projects of LHD, W-7X, NCSX, respectively. The main advantageous features of those designs are common on (1) current-free plasmas, which leads to steady, high Q and disruption-free reactors and (2) high density operations, which largely reduces the diverter heat load. Design optimization and engineering feasibility are highlighted including maintainability, fabricability and showing future R&D key issues with prospect towards DEMO.

Transport of tritium (T) and heat is calculated to estimate the performance of liquid Li17Pb83 (Li–Pb) or Li2BeF4 (Flibe) as a T-breeder in a fusion reactor blanket. T is bred in such a way of low leak to facilities outside and continuous recovery by a removal system, and heat is transferred through structural walls to He coolant efficiently. In this paper, T permeation in a blanket composed of structural materials and liquid Li–Pb or Flibe is calculated based on data of previous experiment. The effects of T recovery ratio by the outside removal apparatus and fluid-film T diffusion resistance in the liquid blanket on overall T permeation rates are analytically clarified. Design of a liquid blanket with low T leak and high T recovery is discussed here. In addition, possibility in that microbubbles may be generated at interfaces between a liquid blanket and a structural wall is investigated.

Deuterium ion irradiations with an ion with energy of 1.7 keV were conducted for boron–titanium (B–Ti) film prepared by electron beam evaporation and hot pressed titanium-boride, TiB 2 . The amount of retained deuterium was measured for these materials using a technique of thermal desorption spectroscopy. The amount of deuterium retained in TiB 2 was comparable with that in B–Ti. Desorption peaks of deuterium in B–Ti were 470 K and 620 K, corresponding to a desorption in the low temperature regime observed in boron (B) and a desorption in titanium (Ti), respectively. The desorption peaks in TiB 2 were 620 K and 750 K, which correspond to the desorption in Ti and that in the high temperature regime in B, respectively. The desorption temperature in B–Ti was approximately 100 K lower than that in TiB 2 . This difference is discussed based upon chemical bindings and amorphous/crystal structures of B–Ti and TiB 2 . Irradiation of helium ion with energy of 5 keV was conducted for B–Ti after the deuterium ion irradiation. The amount of retained deuterium decreased and the desorption temperature shifted to the lower temperature regime, as the helium ion fluence increased. The shift to the low temperature regime is due to the enhancement of amorphous structure of B in B–Ti.

To assess the technical potential of the FFHR operation style, dedicated hydrogen production from electrolysis by the electric power of 1 GW is investigated in this report. Gaseous hydrogen of 700 tons per day can be produced. The steam of 6354 tons per day at 150 °C is necessary in this case. In FFHR, about 450 MW of thermal power is delivered via the scrape-off layer plasma to the divertors with double-null structure. The divertor may be one of the potential heat sources to produce the steam for the electrolysis. Hydrogen has to be packaged by compression or liquefaction, transported by trailer or pipeline, stored, and transferred to the end users. The four different styles of plant outputs were investigated: (A) 1 GW of power generation, (B) pressurized hydrogen gas of 625 tons per day, (C) liquid hydrogen of 574 tons per day, and (D) 0.824 GW of electricity plus 100 tons per day of liquid hydrogen.

Static corrosion tests were performed in molten salts, LiF–BeF2 (Flibe) and LiF–NaF–KF (Flinak), at 500 °C and 600 °C for 1000 h. The purpose is to investigate the corrosion characteristics of reduced activation ferritic steels, JLF-1 (8.92Cr–2W) in the fluids. The concentration of hydrogen fluoride (HF) in the fluids was measured by slurry pH titration method before and after the exposure. The HF concentration determined the fluoridation potential. The corrosion was mainly caused by dissolution of Fe and Cr into the fluids due to fluoridation and/or electrochemical corrosion. Carbon on the surface might be dissolved into the fluids due to the corrosion, and this resulted to the decrease of carbide on the surface. The corrosion depth of the JLF-1 specimen, which was obtained from the weight losses, was 0.637 μm in Flibe at 600 °C and 6.73 μm in Flinak at 600 °C.

Progress of design studies on a heliotron reactor is reported. Recent experimental achievements in large helical device (LHD) are building up the physical basement. Integrated designs on LHD-type energy reactors have improved the base design of force free helical reactor and revealed economical feasibility. Large helical device is exploring a new horizon of net current free helical plasmas. Inherent characteristics demonstrate reactor relevant performances individually; steady state (1 h), high beta (5%), high density (1 × 10 21 m −3 ), etc. In particular, discovery of internal diffusion barrier provides a novel attractive scenario of super high density operation. It greatly facilitates physical requirements as well as mitigates engineering demands. New methods have been proposed to access and operate ignited plasmas; a long rise-up over 300 s to reduce the heating power to 30 MW and a new proportional-integration-derivative (PID) control of the fueling to handle the thermally unstable plasma at high-density operations.