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The manufacturing of tungsten (W) - graphite bonded divertor components for the Large Helical Device (LHD) has been investigated. The spark plasma sintering method was used to bond W and graphite with titanium (Ti) interlayer. Small specimens were fabricated to investigate the bonding strength and to diagnose the bonding interface. The granular structure was formed in the grooved area on the graphite surface. It was suggested that this granular structure had affected the bonding strength., source:https://doi.org/10.1585/pfr.18.1205003

A new cartridge-type blanket named the CARDISTRY-B is proposed for the helical fusion reactor FFHR-d1. This blanket is composed of the neutron shield and the tritium breeder using molten salt. Both of these are toroidally segmented every two degrees. At each toroidal angle, the segmented parts are divided further into several cartridges in order to make it possible to assemble these cartridges after completion of the superconducting magnet coils. The neutron shield is basically assembled by using mortise and tenon prepared on each of the cartridges and the lower port, instead of the wide area welding. After assembly, the plasma side of the neutron shield is welded to form a vacuum vessel. Another side of the neutron shield facing on the superconducting magnet coils is covered with the thermal shield, which was already attached before assembly. The tritium breeder cartridges can be replaced without cutting or welding of cooling pipes inside the vacuum vessel, where severe radiation dose is expected. Details of the CARDISTRY-B, including the results of motion analysis for all cartridges and estimation of the cartridge weight, are discussed., source:https://doi.org/10.1585/pfr.12.1405017, identifier:0000-0002-7901-6077

The stabilization effect of the Internal Flow Resistances (IFR) inserted into water jets has been investigated in order to apply the liquid metal divertor to the helical fusion reactor, FFHR. A high heat load exceeding 20 MW/m2 is one of the important issues for the divertor in the nuclear fusion reactor. A new concept of ergodic limiter/divertor called the REVOLVER-D has been proposed for FFHR. This divertor is expected to tolerate a high heat load larger than a few tens of MW/m2. Jets narrow after acceleration by gravity and transform to droplets due to the surface tension instability. Stabilization of jets is an important issue for realizing the REVOLVER-D. In this paper, stabilization of jet by various IFR has been examined. Here, the “sheath jet” is defined as a jet with an IFR inside. Experiments on sheath jet have been carried out and the stabilization effect of IFR has been confirmed. Numerical simulations on sheath jet using ANSYS have been carried out. It is shown that the velocity of the sheath jet reaches the terminal velocity, which is proportional to 0.36 ± 0.01 power of the flow rate and −0.26 ± 0.01 power of the outer circumference of the IFR., source:https://doi.org/10.1585/pfr.13.1405003, identifier:0000-0001-5079-6391

The cartridge-type helical blanket system called the CARDISTRY-B has been proposed for the helical fusion reactor FFHR-d1. The CARDISTRY-B is aimed at easy construction and maintenance. However, there remain many issues in the design. For example, these include a large number of cartridges, no tangential ports, low compatibility with the helical divertor, and others. To solve these issues, an improved design of the cartridge-type blanket is proposed and named the CARDISTRY-B2. This is composed of the tritium breeding blanket (BB) filled with the flowing molten salt and the neutron shielding blanket (SB) filled with the tungsten-carbide. The number of the BB (or, SB) cartridges in the 1/10 section of the full torus are reduced from 32 to 17 (or, from 44 to 17). Tangential ports for the neutral beam injection heating are newly equipped together with the multi-purpose blanket cartridge. Compatibility with the helical divertor is also improved. The BB cartridges are fixed on the fixing units set on the top of the SB. No other part of the BB is touching the SB except the fixing ribs. The SB cartridges surround the superconducting magnet coils and work as the cryostat inside the vacuum vessel. This makes the large bellows for the large maintenance ports unnecessary. Details of the CARDISTRY-B2 designed for a small helical fusion reactor FFHR-b1 are described in this paper., source:https://doi.org/10.1585/pfr.14.1405163, identifier:0000-0002-7901-6077

Divertor pumping was applied to plasma discharges for superior fuel particle control in the Large Helical Device (LHD). The LHD is equipped with two different pumping systems. One is the main pumping system, in which the pumping speed is 260 m3/s in hydrogen. The other pumping system is the divertor pumping system in which the pumping speed is 70 m3/s in hydrogen. Divertor pumping was applied to 40-second long pulse Electron Cyclotron Heating (ECH) discharges to assess the improvement in particle control provided by divertor pumping. The results show that without divertor pumping, the electron density was not controlled by gas puffing using the feedback signal of line-averaged electron density. Then, the plasma confinement deteriorated, finally leading to radiation collapse. On the other hand, with divertor pumping, the density was well-controlled by gas puffing using the feedback signal. The results indicate that divertor pumping is one of the key tools for controlling the particles in fusion plasmas., source:https://doi.org/10.1585/pfr.16.1202014, identifier:0000-0001-5522-3082

A feasibility study for measuring a deuterium-deuterium (D-D) fusion reaction radial profile by promptly lost D-D fusion born 3 MeV protons, whose energy Larmor radius is the same as the minor radius of CFQS, was performed. The Lorentz orbit code was utilized to estimate the predicted signals of collimated proton detectors using the D-D fusion radial profile calculated by the analytical Fokker-Planck code for steady-state plasma FIT3D-DD code. The inversion of the D-D fusion profile using the estimated signals was performed using a linear matrix solution library. The coarse agreement between input and inverted profiles shows the possibility of D-D fusion profile diagnostics by a 3 MeV proton in CFQS., source:http://www.jspf.or.jp/PFR/PFR_articles/pfr2022/pfr2022_17-2402012.html

The world's first quasi-axisymmetric stellarator CFQS (China First Quasi-axisymmetric Stellarator) will be constructed as the joint project of National Institute for Fusion Science (NIFS) in Japan and Southwest Jiaotong University (SWJTU) in China [1]. Physics design of CFQS plasma was completed [2, 3], and numerous efforts are now being made to finalize engineering design of CFQS [4]. The CFQS vacuum vessel (VV) will be made of thin plates of SUS316L with thickness of 6 mm. The thickness is determined by considering fabricability and manufacturing cost. In this report, analyses have been performed to confirm the structural reliability of VV and evaluate the influence of eddy current using finite element method software ANSYS/MechanicalTM and ANSYS/MaxwellTM., source:https://doi.org/10.1585/pfr.15.2405066

The world's first quasi-axisymmetric stellarator, CFQS, is now under construction. The CFQS will be dedicated to studies on the interaction between flow and turbulence, and confinement improvement by suppression of turbulence in connection with proof-of-principle experiment of quasi-axisymmetry. In order to conduct this experimental research, a heavy ion beam probe (HIBP) system is planned to be installed and utilized to measure the radial electric field and its fluctuation in a CFQS plasma. In this paper, an orbit calculation for a probe beam is performed to verify feasibility of the HIBP in the CFQS. The required beam energy, possible ion species, and the observable region in a CFQS plasma are investigated. The beam attenuation by a CFQS plasma is also estimated for different beam ion species. If we use 133Cs+ as a primary probe beam, the required beam energy is expected to be 30∼50 keV, which is relatively easy to handle. In this case the beam attenuation, evaluated by the ratio between the injected and detected beam currents, is 10−3∼10−2 in a CFQS plasma with a line-averaged electron density of <1.0 × 1019 m−3. For a higher density plasma, usage of 85Rb+ is better in terms of low-beam-attenuation, and a high signal-to-noise ratio. The HIBP in the CFQS will provide a great opportunity to study physics experimentally, related to the radial electric field, poloidal flow, and turbulence suppression., source:A. Shimizu et al 2022 JINST 17 C06004, source:https://doi.org/10.1088/1748-0221/17/06/C06004, identifier:0000-0003-3764-3184

Density control is crucial for maintaining stable confined plasma. Divertor pumping, where neutral particles are compressed and exhausted in the divertor region, was developed for this task for the Large Helical Device. In this study, neutral particle pressure, which is related to recycling, was systematically scanned in the magnetic configuration by changing the magnetic axis position. High neutral particle pressure and compression were obtained in the divertor for a high plasma electron density and the inner magnetic axis configuration. Density control using divertor pumping with gas puffing was applied to electron cyclotron heated plasma in the inner magnetic axis configuration, which provides high neutral particle compression and exhaust in the divertor. Stable plasma density and electron temperature were maintained with divertor pumping. A heat analysis shows that divertor pumping did not affect edge electron heat conductivity, but it led to low electron heat conductivity in the core caused by electron-internal-transport-barrier-like formation., source:G Motojima et al 2022 Phys. Scr. 97 035601, source:https://doi.org/10.1088/1402-4896/ac5269, identifier:0000-0001-5522-3082

FIT3D-DD code, which is a neutron emission and heating efficiency evaluation code based on a simple analytical solution of the Fokker-Planck equation, was applied to deuterium Chinese first quasi-axisymmetric stellarator (CFQS) plasmas heated by a deuterium neutral beam injector. A database of the neutron emission rate due to the beam-bulk D-D fusion reaction in the CFQS was created with the electron temperature and density, which were evaluated from the energy confinement time using ISS04 scaling. The neutron emission rate increases monotonically with the increase in the electron temperature and density. The neutron emission rate is expected to be up to ∼2.7 × 1012 [n/s] for an electron density of less than 4 × 1019 m−3., source:https://doi.org/10.1585/pfr.17.2403063, identifier:0000-0002-5364-805X

An innovative color measurement technique is employed in the Large Helical Device (LHD). This study provides a method for obtaining in broad spatial extent and in great detail the color information of the first wall relating to the thickness of the deposition layer. The RGB (Red, Green, and Blue) value, mainly of the stainless steel plates on the helically twisted coil, is measured by a color analyzer equipped with an integrated sphere light source. On the outer torus side, the colors of almost all stainless steel plates are close to carbon black, which suggests that deposition is dominant. On the inner torus side, all plates except for those neighboring the carbon divertor plates are almost white, as in the case of the stainless steel substrate of the first wall, which suggests that erosion is dominant. The relationship between the color and the distance from the stainless steel plates to the plasma is investigated., source:https://doi.org/10.1585/pfr.10.1202074, identifier:0000-0001-5522-3082

Long pulse discharges in the Large Helical Device (LHD) were interrupted by the emission of large amounts of carbon dusts released from a closed helical divertor in the previous experimental campaign (FY2013). In order to control the dust emission, the configuration of the closed divertor near lower and upper ports was modified in the last experimental campaign (FY2014). While the modification of the divertor configuration has successfully controlled the release of dusts, long pulse discharges were often terminated with abrupt increase in iron ion emission. The termination process of a long pulse plasma discharge was observed with a stereoscopic fast framing camera which can identify the three-dimensional positions of dusts. It proved that the dusts penetrated into the main plasma through the plasma periphery, which could induce the radiation collapse in the plasma discharge., source:https://doi.org/10.1585/pfr.11.2402056, identifier:0000-0003-0655-7347

The design activity on the CFQS quasi-axisymmetric stellarator which is conducted as a joint international project is ongoing by National Institute for Fusion Science (NIFS) in Japan and Southwest Jiaotong University (SWJTU) in China. The CFQS magnetic field generating coils consist of sixteen modular coils (MCs) in total with four different types, four poloidal field coils (PFCs), and twelve toroidal field coils (TFCs). In designing fusion devices, eddy currents and electromagnetic (EM) forces on the vacuum vessel (VV) have to be evaluated to verify the mechanical confidence and reliability. In this work, we consider the following three issues, (1) Influence of eddy current induced by external coil current change on the CFQS magnetic confinement, (2) Evaluation of EM force on VV by eddy current under a typical current of the MC and the PFC, and (3) Feasibility check of whether it is possible to heat the CFQS VV by using induction current., source:https://doi.org/10.1016/j.fusengdes.2020.111869

The Chinese First Quasi-axisymmetric Stellarator (CFQS) is now under design and construction. It will be the first quasi-axisymmetric (QA) configuration device to be operated in the world. The main parameters of the CFQS are as follows: the toroidal periodic number Np = 2, major radius R = 1.0 m, aspect ratio Ap = 4.0 and magnetic field strength Bt = 1.0 T. The low Ap makes it quite challenging to design a supporting structure because of the limited space and strong electromagnetic (EM) force. In this paper, a cage-like supporting structure is proposed for the CFQS modular coil (MC) system to sustain the EM force and the weight of entire device. A finite element analysis is carried out for ensuring the reliability of the supporting structure. The analysis results of the CFQS global model indicate that the cage-like supporting structure can basically satisfy the requirement., source:https://doi.org/10.1016/j.fusengdes.2020.112021, identifier:0000-0002-5601-7106

We evaluated the two dimensional (2D) distribution of the divertor heat flux in LHD. The Infrared (IR) thermography was performed to measure the surface temperature at a divertor plate. The 3D heat conduction equation was solved using the finite element method (FEM) by taking into account the practical divertor geometry including the heat sink behind the graphite tile and the cooling system. The FEM analysis successfully reconstructs heat load distribution from the measured temperature pattern. Significant difference was found between the temperature and the heat load patterns. The FEM analysis shows that the highest heat deposition is observed at the strike line with 5–10 MWm−2 of ∼ 10 mm width in NBI heated discharge. In addition to the strike line there is also found the lower heat deposition region of ∼ 1 MW m−2 with wide channel width ∼ 30 mm. The detailed heat transport analysis inside the divertor components shows that the heat transport process is different between the strike line and the other region due to the heat channel width and to the divertor component structure. The comparison between the heat flux obtained by the thermography and that by the Langmuir probes shows reasonable agreement, except that the peak value of the heat load is higher in the IR thermography than in the Langmuir probe. By relaxing the assumption that electron temperature, Te, equal to be ion temperature, Ti, in the sheath model for the Langmuir probe analysis, the agreement becomes better for the case with Ti > Te., source:https://doi.org/10.1016/j.fusengdes.2021.112235, identifier:0000-0001-6090-5010

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., source:https://doi.org/10.1016/j.fusengdes.2017.07.003, identifier:0000-0002-7901-6077

The maintenance scheme in the compact helical fusion reactor, FFHR-c1, equipped with the liquid metal divertor, REVOLVER-D (Reactor-oriented Effectively VOLumetric VERtical Divertor) and the cartridge-type molten salt blankets, CARDISTRY-B (CARtridges Divided and InSerTed RadiallY − Blanket), is investigated. The magnetic configuration of the compact variant FFHR-c1 is similar to that of the Large Helical Device (LHD), while the device size is 2.8 times enlarged from LHD and a strong magnetic field strength of ∼8 T at the plasma center is adopted. The maintenance of the REVOLVER-D is simpler than that of the helical divertor with a complicated structure as seen in the LHD. In the REVOLVER-D, showers of molten tin are injected into the ergodic layer at 10 inner ports. To circulate the molten tin, 10 sets of the shower system including a liquid metal pump, ducts, a showerhead, a pool, and a heat exchanger, are installed. These can be replaced with simple up/down motions. The CARDISTRY-B consists of 320 tritium breeding blanket cartridges, which are toroidally segmented every two degrees. These cartridges are maintained by using heavy manipulators with a simple 4DOF motion at a constant toroidal angle., source:Junichi Miyazawa, Takuya Goto, Hitoshi Tamura, Teruya Tanaka, Nagato Yanagi, Tananori Murase, Ryuichi Sakamoto, Suguru Masuzaki, Takeru Ohgo, Akio Sagara, Maintainability of the helical reactor FFHR-c1 equipped with the liquid metal divertor and cartridge-type blankets, Fusion Engineering and Design, Volume 136, Part B, 2018, Pages 1278-1285, source:https://doi.org/10.1016/j.fusengdes.2018.04.118, identifier:0000-0002-7901-6077

A feasibility study for installing fast-ion loss diagnostics (FILDs) is initiated to understand energetic ion loss process in a quasi-axisymmetric stellarator CFQS. The calculations of beam ion orbit, the guiding center orbit code including the slowing down by bulk plasma in the Boozer coordinates and the collisionless Lorentz orbit code in the Cartesian coordinates, are performed to estimate the toroidal/poloidal distribution of beam ion losses on a vacuum vessel. We observed that loss points on the vacuum vessel are mainly located at the upper side of torus according to gradient magnetic field drift direction. Based on the loss distribution on the vacuum vessel, a possible location of FILD for CFQS is investigated. Although the energy of lost beam ions is ∼10 keV, relatively high-energy (> 20 keV) components can be measured on the upper side of the torus in the low-density case. Furthermore, the radial or vertical dependence of beam ion loss flux at candidate movable scintillator-type FILD positions is obtained. A considerable number of particles reach the scintillator-type FILD position. The energy and pitch angle distribution of beam ions reaching the FILD position shows that scintillator-type FILDs can measure co-going transit beam ions with energy of ∼25 keV and pitch angle of ∼140 degrees and transition beam ions with energy of 10–15 keV and pitch angle of ∼80–100 degrees., source:Citation K. Ogawa et al 2021 JINST 16 C09029, source:https://doi.org/10.1088/1748-0221/16/09/C09029