Plasma and Fusion Research
Online ISSN : 1880-6821
ISSN-L : 1880-6821
Volume 7
Showing 101-150 articles out of 159 articles from the selected issue
Regular Articles
  • Kazuya TAKAHATA, Sadatomo MORIUCHI, Kouki OOBA, Toshiyuki MITO, Shinsa ...
    Type: Regular Articles
    2012 Volume 7 Pages 2405008
    Published: February 17, 2012
    Released: March 09, 2012
    JOURNALS FREE ACCESS
    We present a fourteen-year data summary of the hydraulic characteristics of the large helical device (LHD) poloidal coils. The superconductors of the poloidal coils are cable-in-conduit conductors (CICC) cooled by circulated supercritical helium. The long-term operation of the LHD demonstrates that the initial hydraulic characteristics can be maintained without flow obstruction. Fine mesh filters installed at the inlet trapped impurities during cool-down of the coils, confirmed by monitoring the pressure drop of the filters. The filters have an important role in removing particles of impurities in the helium and maintaining the hydraulic characteristics of the coils.
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  • Yamato ASAKURA, Naoyuki SUZUKI
    Type: Regular Articles
    2012 Volume 7 Pages 2405009
    Published: February 17, 2012
    Released: March 09, 2012
    JOURNALS FREE ACCESS
    To understand the conditions of exhaust gas treatment at the transition point between the Large Helical Device (LHD) vacuum pumping system and the exhaust gas tritium recovery system, the gas flow rate and hydrogen concentration were measured. Simultaneous measurement of the exhaust gas flow rate and hydrogen concentration was made possible by applying two types of hydrogen monitors: a thermal conductivity sensor and a combustible gas sensor. The obtained results have led to remodeling of the LHD vacuum pumping system and an optimised plan of operation for the tritium recovery system.
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  • Yong SONG, Akio SAGARA, Takeo MUROGA, Qunying HUANG, Muyi NI, Yican WU
    2012 Volume 7 Pages 2405016
    Published: March 15, 2012
    Released: November 06, 2013
    JOURNALS FREE ACCESS
    In FFHR2 (LHD-type helical reactor) design, FLiBe has been selected as a self-cooling tritium breeder for low reactivity with oxygen and water and lower conductivity. Considering the fugacity of the tritium, particular care and adequate mitigation measures should be applied for the effectively extracting tritium from breeder and controlling the tritium release to the environment. In this paper, a tritium analysis model of the FLiBe blanket system was developed and the preliminary analysis on tritium permeation and extraction for FLiBe blanket system were done. The results of the analysis showed that it was reasonable to select W alloy as heat exchanger (HX) material, the proportion of FLiBe flow in tritium recover system (TRS) was 0.2, the efficiency of TRS was 0.85 and tritium permeation reduction factor (TPRF) was 20 in blanket etc.. In addition, further R&D efforts were required for FFHR2 tritium system to guarantee the tritium self-sufficient and safety, for example reasonable quality of tritium permeation barriers on blanket, requirement for the TRS and fabrication technology of the heat exchanger etc..
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  • Teruou TAKAYAMA, Atsushi KAMITANI, Ayumu SAITOH, Hiroaki NAKAMURA
    2012 Volume 7 Pages 2405017
    Published: March 01, 2012
    Released: November 06, 2013
    JOURNALS FREE ACCESS
    The accuracy and the resolution of two types of the contactless methods for measuring the critical current density in a high-temperature superconducting (HTS) film have been investigated numerically. To this end, a numerical code has been developed for analyzing the shielding current density in the film with a crack. The results of computations show that the accuracy of two contactless methods is degraded remarkably due to the crack. Specifically, in the permanent magnet method, the maximum repulsive force acting on the film decreases when the magnet approaches near the crack. It is found that, even if the crack size is small, the maximum repulsive force decreases. This means that the crack can be detected. In the inductive method, although the threshold current decreases because of the crack, its value does not necessarily decrease for the case with a small crack size. In fact, the accuracy is not degraded when the inner radius of the coil contains the crack of the film. For this reason, we conclude that the smallest possible inner radius is preferable to detect the crack.
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  • Takaaki WAJIMA, Kenzo MUNAKATA, Tatsuhiko UDA
    2012 Volume 7 Pages 2405021
    Published: March 15, 2012
    Released: November 06, 2013
    JOURNALS FREE ACCESS
    The deuterium-tritium (D-T) fusion reactor system is expected to provide the main source of electricity in the future. Large amounts of lithium will be required, dependent on the reactor design concept, and alternative resources should be found to provide lithium inventories for nuclear fusion plants. Seawater has recently become an attractive source of this element and the separation and recovery of lithium from seawater by co-precipitation, solvent extraction and adsorption have been investigated. Amongst these techniques, the adsorption method is suitable for recovery of lithium from seawater, because certain inorganic ion-exchange materials, especially spinel-type manganese oxides, show extremely high selectivity for the lithium ion. In this study, we prepared a lithium adsorbent (HMn2O4) by elution of spinel-type lithium di-manganese-tetra-oxide (LiMn2O4) and examined the kinetics of the adsorbent for lithium ions in seawater using a pseudo-second-order kinetic model. The intermediate, LiMn2O4, can be synthesized from LiOH·H2O and Mn3O4, from which the lithium adsorbent can subsequently be prepared via acid treatment., The adsorption kinetics become faster and the amount of lithium adsorbed on the adsorbent increases with increasing solution temperature. The thermodynamic values, ΔG0, ΔH0 and ΔS0, indicate that adsorption is an endothermic and spontaneous process.
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  • Masahiro TANAKA, Takuya GOTO, Yasuji KOZAKI, Akio SAGARA , the FFHR D ...
    2012 Volume 7 Pages 2405023
    Published: May 10, 2012
    Released: November 06, 2013
    JOURNALS FREE ACCESS
    Tritium particle balance in the FFHR DEMO reactor is investigated with consideration of the fueling efficiency by pellet injection system, retention loss in a vacuum vessel and permeation loss from the fuel processing system. In order to satisfy the fuel balance and the tritium safety management, it was necessary to suppress the tritium retention rate to be 10−5 and the DFs in the tritium cycle systems to above 107 with the tritium breeding ratio of 1.08. The processing throughput for the tritium processing system is estimated to be about 400 mol/h, which is almost same as the throughput of the fuel stream for the ITER. Therefore, the tritium processing system for vacuum exhaust gas for the DEMO will not be necessary to improve the system for the ITER further. On the other hands, the significant development of the tritium processing system for the effluent disposal and the waste materials from the safety aspect and the social acceptance will be required toward the DEMO reactor.
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  • Atsushi KAMITANI, Teruou TAKAYAMA, Ayumu SAITOH, Hiroaki NAKAMURA
    2012 Volume 7 Pages 2405024
    Published: May 10, 2012
    Released: November 06, 2013
    JOURNALS FREE ACCESS
    A numerical method is proposed for analyzing the shielding current density in a high-temperature superconducting (HTS) film containing cracks/holes. If an HTS film contains cracks or holes, an integral form of Faraday's law is also imposed as the boundary condition. Since the integral form can be completely incorporated into the weak form, it is regarded as the natural boundary condition. Thus, the weak form has only to be solved with the essential boundary conditions. However, the resulting numerical solution does not satisfy the integral form exactly. In order to resolve this problem, the following method is proposed: virtual voltages be applied along the surfaces of cracks and holes so as to have Faraday's law numerically satisfied. By using the proposed method, the influence of a crack on the permanent magnet method is investigated numerically.
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  • Tsuyoshi YAGAI, Takataro HAMAJIMA
    2012 Volume 7 Pages 2405026
    Published: May 10, 2012
    Released: November 06, 2013
    JOURNALS FREE ACCESS
    The Cable-In-Conduit Conductor (CICC) is the most promising one for large scale fusion magnets. Now it has been adopted as conductors for ITER magnets. Although the conductor has good mechanical strength against large electromagnetic force, the performance is not so good because the Nb3Sn strands are fragile and the critical current density is sensitive to strain. Because the conductor is composed of hundreds or thousands of strands which are twisted and become tangled, the strands experience extra-bending during energizing magnets. It seems so difficult to analyze plastic deformation of the strands of whole conductor. Our approach to calculate it is unique in terms of using structural mechanics called “Beam Model” based on the measured strand traces inside the conduit. The calculated traces provide us the local curvatures of strands under electromagnetic force. This leads to the evaluate the conductor performance such as Ic degradation.
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  • Yoshiro TERAZAKI, Nagato YANAGI, Sai TOMIDA, Hiroki NOGUCHI, Kyohei NA ...
    2012 Volume 7 Pages 2405027
    Published: May 10, 2012
    Released: November 06, 2013
    JOURNALS FREE ACCESS
    Feasibility studies on applying high-temperature superconductors (HTS) to the LHD-type heliotron fusion energy reactor FFHR are being carried out. Because the HTS conductor has high cryogenic stability at elevated temperature operations (e.g. 20 K) and the refrigeration power has enough margins, it is considered that Joule heating dissipation generated at joints of conductors is acceptable to facilitate the segmented fabrication of the helical coils of FFHR. In this study, the joint resistance with 10-kA class YBCO conductors has been measured to evaluate the joule heating dissipation in the FFHR magnet. The experiment has been carried out by fabricating a soldered lap joint and a mechanical lap joint. The feasibility of segmented fabrication is examined by the measured results.
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  • Akihiko ISAYAMA, Takayuki KOBAYASHI, Kenji YOKOKURA, Mitsuru SHIMONO, ...
    2012 Volume 7 Pages 2405029
    Published: May 10, 2012
    Released: November 06, 2013
    JOURNALS FREE ACCESS
    Development of an electron cyclotron (EC) wave system was conducted in an effort to achieve the capability required in JT-60SA. Pulse duration at 1 MW output was extended to 31 s. Transmission line components with a diameter of 60.3 mm were installed in 2011 to reduce temperaturerise during a gyrotron oscillation. Development of a dual-frequency gyrotron was started to enable heating and current drive in the core region of the JT-60SA plasma for the toroidal field of 2.3 T. The EC wave frequency was chosen to be 138 GHz to meet the requirements of physics experiments and gyrotron design. Fabrication of the gyrotron began in 2011.
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  • George H. NEILSON, David A. GATES, Charles E. KESSEL, Jonathan E. MENA ...
    2012 Volume 7 Pages 2405035
    Published: May 10, 2012
    Released: November 06, 2013
    JOURNALS FREE ACCESS
    An assessment of Demo goals and of prerequisites for Demo readiness motivate an examination of a pilot plant: an intermediate facility designed to substantially narrow the technical gap to Demo in a next step. A pilot plant would: 1) test internal components and tritium breeding in a steady-state fusion environment, 2) prototype a maintainable design and maintenance scheme for a power plant, and 3) generate net electricity. Preconceptual designs based on the advanced tokamak (AT), spherical tokamak (ST), and compact stellarator (CS) have been developed in order to compare their relative merits as fusion systems. Any of them would take a large step toward Demo in key performance metrics, e.g. engineering gain QENG (≥1), neutron wall load (>1 MW/m2), tritium breeding ratio (>1), pulse length (106 - 107 s), blanket lifetime fluence (≥ 3MW-yr/m2), plant lifetime (6-20 MW-yr/m2), and availability (10-30%), but they differ in their associated risks.
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  • Masatsugu SAKAGUCHI, Hiroshi IDEI, Shin KUBO, Tetsuji SAITO, Takashi S ...
    2012 Volume 7 Pages 2405037
    Published: May 10, 2012
    Released: November 06, 2013
    JOURNALS FREE ACCESS
    Oversized corrugated waveguide transmissions are used to transmit high power millimeter wave for electron cyclotron heating in ITER. The HE11 mode purity of the waveguide components is a critical issue not only in the high power operation but also in the low power operation to evaluate the components. Conventional single Gaussian beam system causes edge diffraction and excites higher order modes at the waveguide aperture. A proposed HE11-mode exciter uses beam interference between two Gaussian beams interference and serves the high purity HE11 mode without the edge diffraction. The HE11-mode exciter consists of a scalar feed horn antenna, quasi-optical mirrors and a beam splitter (combiner). Phase matched mirrors correct the Gaussian-like beam excited from the horn antenna to pure Gaussian beam. Authors report the experimental study of the exciter.
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  • Ryuichi SANO, Byron J. PETERSON, Evgeny A. DRAPIKO, Dongcheol SEO, Yuj ...
    2012 Volume 7 Pages 2405039
    Published: May 10, 2012
    Released: November 06, 2013
    JOURNALS FREE ACCESS
    The IR imaging video bolometer (IRVB) provides the power distribution of plasma radiation. The radiation distribution is obtained from the temperature distribution on the bolometer foil. It is necessary to calibrate between the temperature distribution and the incident radiation power on the bolometer foil. This paper describes a new calibration technique for the foil which we have developed. The bolometer foil was irradiated with a He-Ne laser and the temperature distribution was measured by an IR camera while changing the irradiation position. The temperature distribution measured was analyzed by the comparison with the results calculated by FEM. We repeated this comparison while changing the parameters such as effective foil thickness and effective emissivity in the calculation until the calculated distribution converged to the measured one. The temperature distribution calculated by the FEM agreed well with the measured one, so the calibration between the radiation power and the temperature profile can be suitably conducted by this technique.
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  • Takayuki WATANABE, Tetsuro OHSHIMA, Toshiki TAKAHASHI, Naoyuki FUKUMOT ...
    2012 Volume 7 Pages 2405042
    Published: May 10, 2012
    Released: November 06, 2013
    JOURNALS FREE ACCESS
    The feasibility of compact torus (CT) neutralization fuel injection method is studied by a simulation model using particleand MHD hybrid techniques. The neutralization processis simulated by using rate-coefficients. The magnetic and electric fields are found to respond sluggishly to the neutralization process. Slow ions generated by charge-exchange have been added to the model, although the CT neutralization process was not significantly affected by this. Finally, the minimum length for the CT neutralizer is proposed as 2 m from the simulation run time 10 µs at a CT injection speed of 200 km/s.
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  • Daiji KATO, Hiroyuki A. SAKAUE, Izumi MURAKAMI, Teruya TANAKA, Takeo M ...
    2012 Volume 7 Pages 2405043
    Published: May 10, 2012
    Released: November 06, 2013
    JOURNALS FREE ACCESS
    Ion-beam induced luminescence of sintered Er2O3 samples irradiated by Ar+ ion-beams was measured in a visible range. In this experiment, three emission bands were observed at 500-520, 540-570, and 640-690 nm. The measured luminescence band at 640-690 nm is resolved into Lorentzian Stark component having a ∼1012 Hz width in frequency. Center wavelengths of the Stark components agree with those of intra-4f transitions of Er3+(4f11) ions situated at C2 symmetry cation sites in pure Er2O3. However, resonance broadening due to nearby Er3+ in the crystal accounts only 1% of the total width. Depopulation of the crystalline oxide in irradiated regions is inferred from decreasing intensity of the emission band at 640-690 nm during continuous irradiation. The emission bands at 500-520 and 540-570 nm still remained in heavily damaged samples look similar with those observed with a metallic Er target.
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  • Tatsuya OHINATA, Satoshi ITO , Hidetoshi HASHIZUME
    2012 Volume 7 Pages 2405045
    Published: May 10, 2012
    Released: November 06, 2013
    JOURNALS FREE ACCESS
    In this study, joint resistance in mechanical butt joint of a single GdBCO tape was evaluated by experiments and current distribution analyses to apply a remountable HTS magnet for a future fusion reactor. The results showed that thickness of the metal layers of GdBCO tape, cutting angle of the conductor and existence of soldered interface affect the joint resistance. According to discussion based on the results, an optimized joint structure could achieve joint resistance of 0.4 µΩ for single GdBCO tape, which correspondsto 5 nΩ for 100 kA class HTS conductor. The resistance has to be reduced almost a half value of the present result to be accepted from the view point of electric power for cooling.
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  • Fan XIA, Li PAN, Li ZHAO, Wei PAN, Xiao SONG, Xinyi LI, Chuan WANG, Ji ...
    2012 Volume 7 Pages 2405048
    Published: June 07, 2012
    Released: November 06, 2013
    JOURNALS FREE ACCESS
    In addition to the real time plasma magnetic and kinetic control, process control is also very important for the operation of a Tokamak. Process control aims at the normal control of the plant systems including information sharing, schema of the Tokamak operation, and alarm handling. A new tokamak named “HL-2M” will be built during the next four years and many conceptual designs for the operation of the new device might be validated on HL-2A, which is currently running at SWIP Chengdu, China. For those purposes, we select the EPICS as the basis for the process control system of HL-2A. Some user interface, data management, and development tools are adopted such as control system studio (CSS), MS SQL Server, Matlab, LabVIEW, VC++, and Java. In this study, we will describe the initial design of the whole HL-2A/HL-2M process control system and its related achievements in HL-2A.
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  • Hirotaka CHIKARAISHI, FFHR Design Group
    2012 Volume 7 Pages 2405051
    Published: June 07, 2012
    Released: November 06, 2013
    JOURNALS FREE ACCESS
    The force-free helical reactor (FFHR) is a helical-type fusion reactor whose design is being studied at the National Institute for Fusion Science. The FFHR will use three sets of superconducting coils to confine the plasma. It is not a fusion plasma experimental device, and the magnetic field configuration will be optimized for burning plasma. This paper introduces a conceptual design for a dc power system to excite the superconducting coils of the FFHR. In this design, the poloidal coils are divided into a main part, which generates a magnetic field for steady-state burning, and a control part, which is used in the ignition process to control the magnetic axis. The feasibility of this configuration was studied using the Large Helical Device coil parameters, and the coil voltages required to sweep the magnetic axis were calculated. It was confirmed that the axis sweep could be performed without a high output voltage from the main power supply. Finally, the power supply ratings for the FFHR were estimated from the stored magnetic energy.
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  • Mitsutaka ISOBE, Tsuyoshi AKIYAMA, Tokihiko TOKUZAWA, Teruya TANAKA, D ...
    2012 Volume 7 Pages 2405053
    Published: June 07, 2012
    Released: November 06, 2013
    JOURNALS FREE ACCESS
    The plasma diagnostics required for a heliotron-type DEMO reactor are discussed in terms of real-time burn control and safe operation of the machine. The minimum diagnostic set having the smallest footprint are essential in DEMO. Neutron transport calculation suggests that the diagnostic components used in existing experiments will deteriorate immediately in a DEMO reactor hall if they are not protected by a neutron shield. Neutron energy spectrometry is a promising diagnostic that is expected to play an important role in diagnosing DEMO plasmas, providing a fusion energy output, fuel ion temperature, ratio of deuteron density nD to triton density nT, and velocity distribution of confined α particles.
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  • Masahiko EMOTO, Masanobu YOSHIDA, Chihiro SUZUKI, Yasuhiro SUZUKI, Ka ...
    2012 Volume 7 Pages 2405058
    Published: June 07, 2012
    Released: November 06, 2013
    JOURNALS FREE ACCESS
    More than 100 diagnostic devices are attached to the vacuum vessel of the Large Helical Device (LHD); they measure various aspects of the plasma physics. Because the shape of the LHD plasma is not symmetric, each diagnostic obtains the physical values in a different cross section. For example, the Thomson scattering system measures the electron temperature profile in the horizontally elongated cross section, and the laser interferometer measures the line-integrated electron density profile in the vertically elongated cross section. To analyze the data obtained by different diagnostics, their measurement positions must be mapped to a unified coordinate system, the flux coordinate system. Therefore, the authors have been building a database to map the physical coordinates to the flux coordinates. A system for mapping the electron temperature profile to the flux coordinates, TSMAP, has been developed using the database. The profiles calculated by TSMAP are fundamental data for analyzing the plasma physics during an experiment. Therefore, they must be obtained as soon as possible. However, the execution of TSMAP requires computational power, and the performance of a typical personal computer is not high enough to keep up with the 3-min plasma discharge cycle. To increase the performance, the authors use a parallel computing approach. Because the fitting calculation for each time is independent, the calculations for different times can be executed simultaneously. Using this approach, the authors increased the performance by 25 times, achieving a 25-s execution time.
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  • Shinya OGASAWARA, Shin KUBO, Masaki NISHIURA, Yoshinori TATEMATSU, Ter ...
    2012 Volume 7 Pages 2405061
    Published: June 07, 2012
    Released: November 06, 2013
    JOURNALS FREE ACCESS
    Frequency measurements at 74-80 GHz were conducted for the identification and suppression of spurious modes from the 77-GHz gyrotron, which was originally introduced as a power source for electron cyclotron resonance heating and is now also used as a probe beam for the collective Thomson scattering (CTS) diagnostic. The spurious modes are excited as the gyrotron output powe ris turned on and off by controlling the anode voltage. These modes are harmful for the CTS diagnostic, even though their power is less than approximately 50 dB below than that of the main mode. The measured frequency of one of the spurious modes is approximately 74.7 GHz. The cavity structure, starting current calculation, and mode competition calculation suggest that the spurious mode is the TE17,6 mode. The result is important for optimizing the gyrotron operation as a CTS probe beam and suppressing or minimizing the spurious mode excitation.
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  • Takuya KONO, Akinori ISHIKAWA, Seigo MISAKI, Atsushi SUNAHARA, Satoshi ...
    2012 Volume 7 Pages 2405065
    Published: July 26, 2012
    Released: November 06, 2013
    JOURNALS FREE ACCESS
    Plasma shielding is an important concept to study if the material damage could be suppressed with plasma layer properly prepared by absorbing the incoming plasma flux onto a divertor target in MFE or the first wall in IFE reactors. First experimental evidence of this effect is reported. Two plasma plumes (n ∼ 1012/cm3, Te ∼ 1 eV) are created with two laser beams. The laser ablation plasma plumes are created at laser energy density up to 10 J/cm2 and are crossed each other. 12∼59% of the incoming plasma particles are shielded with the collisions of the other plasma plume. By observing the material dependence of colliding effects, the effect to plasma shielding is discussed.
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  • Youji SOMEYA, Kenji TOBITA
    2012 Volume 7 Pages 2405066
    Published: June 07, 2012
    Released: December 25, 2013
    JOURNALS FREE ACCESS
    The decay heat of activated materials is important in safety assessment of fusion DEMO reactor against loss of coolant-flow accidents. Decay heat for reactor main components of the SlimCS DEMO reactor was studied with a one-dimensional code THIDA-2. The reactor main components consist of the inboard (IB) blanket module, outboard (OB) blanket module and divertor. For a reactor with a fusion output of 3.0 GW, the decay heat of IB blanket, OB blanket, divertor and radiation shield were estimated to be as high as 8.6 MW, 30.9 MW, 10.6 MW and 1.8 MW, respectively, immediately after the shutdown of operation. The total decay heat was as high as 52 MW immediately after the shutdown and 3.1 MW one month later. The blanket produces the largest portion of decay heat, about 76%.
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  • Takuya KONDO, Kozo YAMAZAKI, Tetsutarou OISHI, Hideki ARIMOTO, Tatsuo ...
    2012 Volume 7 Pages 2405067
    Published: July 26, 2012
    Released: December 25, 2013
    JOURNALS FREE ACCESS
    Because the D-3He reaction generates no neutrons and the D-D reaction can use abundant fuel resources, these reactions are expected to be used in advanced fuel fusion reactors. Economic considerations and engineering problems are important for realizing such reactors as commercial plants. Therefore, we estimate and compare the cost of electricity (COE) from D-T, D-3He, and catalyzed D-D (cat D-D) fusion reactors. D-3He and cat D-D reactors have a low neutron wall load. Therefore, the D-3He reactor has no wall replacement cost. In addition, no tritium breeding system is needed for the D-3He reactor, but 3He gas is rare. Because the reaction rates of the D-3He and D-D reactions are less, D-3He and D-D reactors require highly efficient confinement properties and operation at high ion temperatures. Furthermore, the power densities of D-3He and D-D reactors are smaller than that of the D-T reactor; thus, D-3He and D-D reactors require a large plasma volume. Assuming a high ion temperature (= 60 keV) and high normalized beta (= 7-8), the COE of a D-3He reactor is expected to be similar to that of a D-T reactor. In terms of cost, cat D-D is disadvantageous in comparison with D-3He and D-T reactors.
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  • Masatoshi KONDO, Teruya TANAKA, Takeo MUROGA, Hiroyuki TSUJIMURA, Yasu ...
    2012 Volume 7 Pages 2405069
    Published: July 26, 2012
    Released: December 25, 2013
    JOURNALS FREE ACCESS
    The electroplating of Er metal on the reduced activation ferritic martensitic steel, JLF-1 (Fe-9Cr-2W-0.1C), in a molten salt was studied. The specimen was immersed in the molten ErCl3 doped LiCl-KCl electrolyte. The electroplating was carried out by a constant potential electrolysis method and a pulsed current electrolysis method. It was found that the Er metal was deposited on the specimen surface due to the electrochemical reaction.
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  • Yoshiro MUNAKATA, Takashi KAWAGUCHI, Hiromasa TAKENO, Yasuyoshi YASAKA ...
    2012 Volume 7 Pages 2405071
    Published: July 26, 2012
    Released: December 25, 2013
    JOURNALS FREE ACCESS
    In an advanced fusion, fusion-produced charged particles must be separated from each other for efficient energy conversion to electricity. The CuspDEC performs this function of separation and direct energy conversion. Analysis of working characteristics of CuspDEC on plasma density is an important subject. This paper summarizes and discusses experimental and theoretical works for high density plasma by using a small scale experimental device employing a slanted cusp magnetic field. When the incident plasma is low-density, good separation of the charged particles can be accomplished and this is explained by the theory based on a single particle motion. In high density plasma, however, this theory cannot be always applied due to space charge effects. In the experiment, as gradient of the field line increases, separation capability of the charged particles becomes higher. As plasma density becomes higher, however, separation capability becomes lower. This can be qualitatively explained by using calculations of the modified Störmer potential including space charge potential.
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  • Akifumi IWAMOTO
    2012 Volume 7 Pages 2405073
    Published: July 26, 2012
    Released: December 25, 2013
    JOURNALS FREE ACCESS
    The surface orientation dependence of heat transfer characteristics in liquid helium was discussed based on previous studies. Judging from their discussions and experimental data, the critical heat fluxes of our measurements come from the upper limit of the heat flux in the regime of continuous vapor columns and patches. To compensate the surface orientation dependence, we modified the gravitational force term in a theoretical equation for the critical heat flux with a horizontal surface. Then, the evaluations by the modified equation were compared with our experimental results. Film boiling heat transfer coefficient with the variation of surface orientation was also discussed based on two-phase boundary layer treatment of free convection film boiling. It was confirmed that our experiments were consistent with the theory.
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  • Makoto OKADA, Yuki EDAO, Hiroaki OKITSU, Satoshi FUKADA
    2012 Volume 7 Pages 2405074
    Published: July 26, 2012
    Released: December 25, 2013
    JOURNALS FREE ACCESS
    Clarification of tritium transfer in blanket is an important issue for realizing fusion reactor. We perform an experiment of simultaneous H and D permeation through Li17Pb83 by means of an unsteady permeation method in order to clarify the interactions and isotope effects between H and D atoms. The experimentis conducted under the condition where the Sieverts' law holds. As a result, it is found that (i) H and D atoms permeate independently regardless of the H/D component ratio in the upstream gas and (ii) diffusion process is the rate-determining step in the overall permeation process. The diffusivity of H is around 1.4 times larger than that of D. The solubility of H is close to that of D. We estimate two vibration modes at an absorption site and a saddle point of H in Li17Pb83 based on the ratio of isotope effect. It is considered that the zero-point vibration energy of H at the absorption site in Li17Pb83 is around 0.173 eV and that at the saddle point is around 0.235 eV. The ratio of isotope effect is almost in proportion to the square root of mass ratio of D to H. The diffusivity of T can be estimated as 1/1.7 times of H.
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  • Satoshi SHIGEHARU, Yusuke HATACHI, Tetsushi HIROMOTO, Yuki EDAO, Satos ...
    2012 Volume 7 Pages 2405080
    Published: July 26, 2012
    Released: December 25, 2013
    JOURNALS FREE ACCESS
    Materials to compose a fusion reactor chamber are subjected to neutron irradiation. International Fusion Materials Irradiated Facility (IFMIF) will be used to analyze how neutron damage affects material durability. The material is irradiated by high-intensity neutron beam, which is generated by the D-Li stripping reaction. Tritium (T) generated by by-pass reaction needs to be recovered from liquid Li for safety. Recovery of tritium by an yttrium (Y) hot trap is one of the ways to recover T from liquid Li in IFMIF. Therefore, it is necessary to analyze the behavior of T in the liquid Li. In this study, the authors measured H2 absorption rates under stirred conditions of liquid Li and analyzed the effect of hydrogen absorption rate of Y by elevating temperature from 250C to 400C to establish the way to recover T from liquid Li with the Y hot trap in IFMIF. Judging from the comparison, we considered the rate-determining step is H diffusion in Y. Mass transfer coefficients at each temperature were determined by fitting the results between calculation and experiment.
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  • Kozo YAMAZAKI, Tetsutarou OISHI, Hideki ARIMOTO
    2012 Volume 7 Pages 2405082
    Published: July 26, 2012
    Released: December 25, 2013
    JOURNALS FREE ACCESS
    For burning plasma simulation and reactor system analysis on steady-state high beta fusion reactors, TOTAL physics code and PEC engineering code have been developed. From TOTAL analysis, it is clarified that by choosing appropriate external current drive profile, high bootstrap-current fraction is achieved in steady-state. From PEC analysis, it is found that the current drive efficiency should be raised for cost of electricity (COE) and CO2 reductions in rather low-beta reactors. Newly derived scaling formulas on COE and life-cycle CO2 emission rate might contribute to the future reactor design projection.
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  • Takuya GOTO, Junichi MIYAZAWA, Hitoshi TAMURA, Teruya TANAKA, Shinji H ...
    2012 Volume 7 Pages 2405084
    Published: July 26, 2012
    Released: December 25, 2013
    JOURNALS FREE ACCESS
    Conceptual design activity for the LHD-type helical DEMO reactor FFHR-d1 has been conducted at the National Institute for Fusion Science under the Fusion Engineering Research Project since FY2010. In the first step of the conceptual design process, design window analysis was conducted using the system design code HELIOSCOPE by the “Design Integration Task Group”. On the basis of a parametric scan with the core plasma design based on the DPE (Direct Profile Extrapolation) method, a design point having a major radius of 15.6 m and averaged magnetic field strength at the helical coil winding center of 4.7 T was selected as a candidate. The validity of the design was confirmed through the analysis by the related task groups (in-vessel component, blanket, and superconducting magnet).
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  • Tatsuhiko UDA, Masahiro TANAKA, Shizuhiko DEJI, Jianqing WANG, Osamu F ...
    2012 Volume 7 Pages 2405085
    Published: July 26, 2012
    Released: December 25, 2013
    JOURNALS FREE ACCESS
    The environmental electromagnetic fields were measured around a magnetic confinement fusion test facility namely Large Helical Device (LHD) which is equipped with large superconducting magnet coils system and high-power plasma heating systems of Neutral Beam Injection, Electron Cyclotron resonance Heating and Ion Cyclotron Range of Frequencies (ICRF) heating. The leakage of the static magnetic field from the LHD was less than 1.2 mT, and it varied according to the coil operation. The extremely low frequency electromagnetic field was measured around power supply units for the coil system, and the magnetic field of higher than the guideline level of the International Commission on Nonionizing Radiation Protection (ICNIRP) was predicted. Leakage of high frequency electromagnetic field from the ICRF was observed in bursts according to plasma shots. The measured values were less than the occupational guideline levels. Although the electromagnetic fields were less than the regulation levels, more monitoring survey is necessary from the view point of occupational safety.
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  • Masao MATSUYAMA, Jyunpei SAIKAWA, Shinsuke ABE, Kiyohiko NISHIMURA, Na ...
    2012 Volume 7 Pages 2405091
    Published: July 26, 2012
    Released: December 25, 2013
    JOURNALS FREE ACCESS
    Tritium retention of samples exposed to plasmas in the Large Helical Device (LHD) during each campaign in 12th, 13th and 14th cycles has been studied. Small sample plates made of stainless steel type 316L were fixed in advance at four different walls in LHD: location of a sample plate was 1.5U, 5.5U, 6.5L and 9.5L. After plasma exposure in each cycle, these samples were exposed to tritium gas at a temperature of 300 or 623 K. Retention behavior of tritium in surface layers of each sample was mainly examined using β-ray-induced X-ray spectrometry (BIXS) and X-ray photoelectron spectroscopy (XPS). The energy spectra observed by BIXS and XPS showed the depositions of boron, carbon, titanium, chromium, iron, nickel and molybdenum with oxygen. Tritium retention of the samples exposed to plasma increased than that of a bare SS316L sample, although it was largely different in the location of a sample. When the samples were exposed to tritium gas at 300 K, the order of magnitude of tritium retention was as follows: 9.5L≫5.5U>6.5L>1.5U for 12th cycle, 6.5L>9.5>1.5U>5.5U for 13th cycle, and 6.5L>1.5U∼5.5U>9.5L for 14th cycle.
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  • Akihito TSUCHIYA, Tomoaki HINO, Yuji YAMAUCHI, Yuji NOBUTA, Masato AKI ...
    2012 Volume 7 Pages 2405097
    Published: July 26, 2012
    Released: December 25, 2013
    JOURNALS FREE ACCESS
    Lithium titanate, Li2TiO3, was irradiated with deuterium ions with energy of 1.7 keV and then exposed to helium or helium-hydrogen mixed gas at different temperature, in order to evaluate the effect of exposure gas on removing deuterium from the pebbles. The amount of residual deuterium in the irradiated pebbles after the gas exposure was measured by thermal desorption spectroscopy. The mixing of hydrogen gas into helium gas enhanced the removal amount of deuterium. In other words, the amount of residual deuterium after the helium-hydrogen mixed gas exposure at low temperature was lower than that after the helium gas exposure. The ion irradiation followed by heating for the pebbles was repeated, and the residual amount of deuterium was measured. The residual amount increased with the number of irradiation/heating cycles. The result suggests that the tritium inventory in the blanket in a fusion reactor changes after the operation.
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  • Mikio SAIGUSA, Shuhei SUGAWARA, Kohei ATSUMI, Yasuhisa ODA, Tomoki YAM ...
    2012 Volume 7 Pages 2405099
    Published: July 26, 2012
    Released: December 25, 2013
    JOURNALS FREE ACCESS
    For improving a stabilizing efficiency of neoclassical tearing modes, the new type diplexer as a fast switching device of high power millimeter wave is proposed for an electron cyclotron current driving system. The principle is a ring resonator-type switch consisting of a ring corrugated circular waveguide, a pair of mitre-bends, and a pair of half mirrors. A mock-up diplexer was designed, fabricated, and tested in low power. The switching operation of the mock-up diplexer with slotted metal half mirrors was verified at the frequency bands of 137 GHz and 170 GHz.
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  • Hiroyasu UTOH, Kenji TOBITA, Youji SOMEYA , Makoto NAKAMURA
    2012 Volume 7 Pages 2405109
    Published: July 26, 2012
    Released: December 25, 2013
    JOURNALS FREE ACCESS
    For optimization of coolant conditions in DEMO blanket design, a two-dimensional (2D) nuclear-thermal-coupled analysis code, DOHEAT, has been modified. A striking feature of DOHEAT is to have a user-friendly interface that enables users to create an appropriate analysis model for different blanket concepts without much diffculty. In the modified DOHEAT, the coolant condition calculation module was added into the 2D thermal analysis module, and the temperature profile in the blanket was provided based on the nuclear heating rate profile and coolant temperature. In addition, numerical solution of simultaneous linear equation is changed from successive over relaxation (SOR) method to bi-conjugate gradient stabilized (Bi-CGStab) method for calculation speed-up. By improving DOHEAT, a series of blanket analysis including not only neutronics and thermal analysis but also coolant condition can be done. The modified DOHEAT allows to calculate the temperature change of the coolant along the cooling tube, and to evaluate the accurate temperature distribution of blanket.
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  • Tetsutarou OISHI, Kozo YAMAZAKI, Hideki ARIMOTO, Kanae BAN, Takuya KON ...
    2012 Volume 7 Pages 2405115
    Published: September 13, 2012
    Released: December 25, 2013
    JOURNALS FREE ACCESS
    Cost evaluation analysis of the tokamak-type demonstration reactor DEMO using the PEC (physics-engineering-cost) system code is underway to establish a cost evaluation model for the DEMO reactor design. As a reference case, a DEMO reactor with reference to the SSTR (steady state tokamak reactor) was designed using PEC code. The calculated total capital cost was in the same order of that proposed previously in cost evaluation studies for the SSTR. Design parameter scanning analysis and multi regression analysis illustrated the effect of parameters on the total capital cost. The capital cost was predicted to be inside the range of several thousands of M$s in this study.
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  • Hyuck Jong KIM, Jun Ho YEOM, Hyung Chan KIM, Myeun KWON
    2012 Volume 7 Pages 2405118
    Published: July 26, 2012
    Released: December 25, 2013
    JOURNALS FREE ACCESS
    The fusion DEMO program of Korea has been conceptualized to realize magnetic fusion energy with the tokamak concept at the end of 2030s or early 2040s. In this program, to expedite the development of a fusion DEMO plant, cross-cutting based on the commonalities between the fusion DEMO plant and existing systems. Among the existing systems, the current and generation IV nuclear power plants will have many areas of commonalities with the fusion DEMO plant including regulatory requirements and licensing processes, codes and standards, design methods and computational codes for thermo hydraulic analysis, and safety analysis methods. Theses commonalities will be used for discovering a pathway of resolving the nested logic dilemma incurred by the inherent first-of-a-kind nature of the fusion DEMO plant. This paper presents the result of an exploratory study on the subject cross-cutting.
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  • Kanae BAN, Kozo YAMAZAKI, Hideki ARIMOTO, Tetsutarou OISHI, Tatsuo SHO ...
    2012 Volume 7 Pages 2405119
    Published: September 13, 2012
    Released: December 25, 2013
    JOURNALS FREE ACCESS
    We analyzed several types of fusion reactors, tokamak (TR), spherical tokamak (ST), helical (HR), and inertial fusion reactor (IR) using physics, engineering and cost (PEC) code, which evaluates economic and lifecycle energy amount quantitatively. We compared the cost of electricity (COE) and the energy payback ratio (EPR) of each fusion reactors with those of fission power plant. Especially, we focus on the EPR of TR with several blanket and shield designs having scarce materials such as silicon carbide (SiC), vanadium alloy (V), and ferritic steel (FS). As the result, we found that the EPR of TR with SiC/LiPb blanket/shield model is the lowest. The COEs and the input energy of TR (βN = 4.0) and IR are lower than those of ST and HR. The COE of fusion reactor is two times higher than that of fission power plants. However the EPR of fusion reactor is as high as that of fission reactor.
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  • Shin KAJITA, Takaki HATAE, Takeshi SAKUMA, Shuichi TAKAMURA, Noriyasu ...
    2012 Volume 7 Pages 2405121
    Published: September 13, 2012
    Released: December 25, 2013
    JOURNALS FREE ACCESS
    Some discharge phenomena seriously damaged the secondary mirror for Thomson scattering diagnostics, which was located outside the vacuum vessel. In this paper, the surface damages recorded on the mirror are observed in detail with an optical microscope. Many fine trails were found on the surface. The trails could be categorized into two different types with respect to the trail width. The mechanisms to lead the damages were discussed based on the observation. This study issues warning on the components to be installed in future fusion devices both inside and outside the vacuum vessel.
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  • Valentyn TSISAR, Olga YELISEYEVA, Takeo MUROGA, Takuya NAGASAKA
    2012 Volume 7 Pages 2405123
    Published: September 13, 2012
    Released: December 25, 2013
    JOURNALS FREE ACCESS
    Effect of Li purity with respect to N on the features of in-situ formation of Er2O3 oxide coating on the surface of V-4Ti-4Cr alloy was investigated. Samples of V-4Ti-4Cr alloy (NIFS HEAT 2) were pre-charged by oxygen in Ar-7%O2 at 700C. The hardened zone with needle shaped Ti-O net structure was formed in the near-surface layers of V-alloy after oxidation. Oxygen pre-charged samples were exposed to N-containing Li (CN[Li] ≤0.2 wt%) at 700C and Li pre-cleaned by getter (Zr) at 650C for 100 h. Both melts were doped with active Er impurity. It was shown that in-situ formation of E2O3 oxide coating on the surface of V-alloys depends strongly on the purity level of Li with regard to the N. In N-containing Li the coating was not formed on the surface of V-alloy, while whole oxygen pre-stored in V-matrix was dissolved by Li. On the contrary, in Li pre-cleaned with respect to N by Zr getter, the insulating oxide film was formed.
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  • Yoshimitsu HISHINUMA, Satoshi MURAKAMI, Kenji MATSUDA, Tsutomu TANAKA, ...
    2012 Volume 7 Pages 2405127
    Published: September 13, 2012
    Released: December 25, 2013
    JOURNALS FREE ACCESS
    Erbium oxide (Er2O3) was shown to be a high potential candidate for tritium permeation barrier and electrical insulator coating for advanced breeding blanket systems such as liquid Li, Li-Pb or molten-salt blankets. Recently, we succeeded to form Er2O3 coating layer on large interior surface area of metal pipe using Metal Organic Chemical Vapor Deposition (MOCVD) process. In this paper, we investigated the microstructure of Er2O3 coating layer on stainless steel 316 (SUS 316) plate before and after heat treatments with hydrogen or argon gases. From the results of TEM observations, we confirmed that Er2O3 coating layer with 700 nm thickness was formed on the SUS 316 plate and this layer was identified to poly-crystal phase because the diffraction fleck which was arranged like a ring was observed in the selected electron diffraction pattern. No macroscopic defects such as crack and peeling in Er2O3 coating layer were observed before and after thermal cycling test. The change of microstructure of the Er2O3 coating layer on before and after heat cycling test was reported.
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  • Teruya TANAKA, Akio SAGARA, Takuya GOTO, Nagato YANAGI, Suguru MASUZAK ...
    2012 Volume 7 Pages 2405132
    Published: October 15, 2012
    Released: January 10, 2014
    JOURNALS FREE ACCESS
    Radiation shielding and tritium breeding performances of the helical DEMO reactor FFHR-d1 have been investigated for the present proposed reactor component configuration. Since the core plasma position shifts to inboard side of the torus, the total thickness of the inboard breeding blanket and radiation shield of FFHR-d1 is limited to ∼70 cm. To simulate the geometric features of the helical reactor, three-dimensional neutronics calculation model consisted of ∼4,000 cells have been prepared for the neutron and gamma-ray transport calculations using the MCNP code. It has been confirmed that the present radiation shield configuration with WC (tungsten carbide) and FS (ferritic steel) + B4C layers would provide sufficient shielding performance for the helical coils. The tritium breeding ratio (TBR) of 1.08 has been obtained with a Flibe+Be/FS breeding blanket for the present component configuration of FFHR-d1.
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  • Yuji NOBUTA, Kenji YOKOYAMA, Jun KANAZAWA, Yuji YAMAUCHI, Tomoaki HINO ...
    2012 Volume 7 Pages 2405134
    Published: October 15, 2012
    Released: January 10, 2014
    JOURNALS FREE ACCESS
    Co-deposition of deuterium with carbon in an opening on a plasma-facing surface, a so-called ‘gap', was simulated by using a deuterium arc discharge with carbon electrodes. The carbon deposition distribution and deuterium retention/desorption behavior of the carbon film were investigated. The amount of deposited carbon decreased exponentially with an increase of the distance from the gap entrance and more rapidly decreased with an increase in discharge gas pressure. The deuterium concentration in the carbon film increased with discharge gas pressure. At a high discharge gas pressure of 36 Pa, the atomic ratio of D/C in the carbon film reached as high as 0.9. Deuterium retained in the film desorbed mainly in the forms of D2, HD, CD4 and C2D4. The desorption behavior of retained deuterium depended on D/C. In a film with a high D/C ratio, desorption of D2 started at lower temperatures. The amount of desorbed hydrocarbons (CD4 and C2D4) increased with D/C. Carbon film with high D/C tended to contain a polymer-like structure, which could be related to the desorption behavior of the retained deuterium.
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  • Yuya MIYOSHI, Yuichi OGAWA, Makoto NAKAMURA
    2012 Volume 7 Pages 2405135
    Published: October 15, 2012
    Released: January 10, 2014
    JOURNALS FREE ACCESS
    For the fusion reactors or experimental devices, one will be required to control several plasma parameters, like the fusion power, the heat flux, the neutron flux, the beta-value and so on. To control these parameters, many diagnostics and actuators are needed, but the diagnostics and actuators available in DEMO/commercial reactors are limited because of the high heat or neutron flux. For these reasons, to realize the fusion reactors, the construction of the reactor control logic is required. We are developing the burn control logic in the core plasma with a 1.5D transport code, and discussing on the relationship between control parameters and actuators. To demonstrate the feasibility of the core plasma control, we have demonstrated the simultaneous control of the fusion power and the safety factor profile with the gas-puff and NBI.
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  • Takuya NAGASAKA, Teruya TANAKA, Akio SAGARA, Teruo MUROGA, Masatoshi K ...
    2012 Volume 7 Pages 2405141
    Published: November 22, 2012
    Released: January 10, 2014
    JOURNALS FREE ACCESS
    Hydrogen recovery unit is developed for the molten salt loop Orosh2i-1. Pure Ni was selected as hydrogen permeation material due to its industrial maturity of fabrication technology and good compatibility with molten salt in fusion reactor condition. No significant degradation of hydrogen permeability of the pure Ni during the fabrication process. Advanced hydrogen permeation materials, such as Pd, V, Nb, and Ta, maintaining higher hydrogen permeability are also discussed to develop more compact hydrogen recovery systems.
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  • Mamoru SHOJI, Suguru MASUZAKI, Tomohiro MORISAKI, Masahiro KOBAYASHI ...
    2012 Volume 7 Pages 2405145
    Published: October 15, 2012
    Released: January 10, 2014
    JOURNALS FREE ACCESS
    A vacuum pumping system is installed in a Closed Helical Divertor (CHD) in the Large Helical Device (LHD) at the National Institute for Fusion Science for active control of the peripheral plasma density and impurity suppression in the core plasma. In the CHD configuration, the distance between the pumping system and the divertor plates (heat and particle source) is very short (only ∼0.1 m). One of the major issues in designing the pumping system is the reduction of heat load by radiation and thermal conduction due to the neutral particles being released from the heated divertor plates while keeping a high pumping efficiency. Here the heat load and the pumping efficiency are analyzed using a neutral particle transport simulation and a finite element method based software for multi-physics analysis. We propose a new design for a pumping system with an expanded area of the inlet of the water-cooled blinds and a bottom slit beneath the pumping system. This increases the pumping efficiency by approximately 60% over that of our previous design. It also predicts that the increase in heat load on the pumping system for the new design would be reasonably suppressed by a buffer plate with high emissivity on the surface of the vacuum vessel on the inboard side of the torus.
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  • Kazuya ICHIMURA, Yousuke NAKASHIMA, Katsuhiro HOSOI, Hisato TAKEDA, Ta ...
    2012 Volume 7 Pages 2405147
    Published: November 22, 2012
    Released: January 10, 2014
    JOURNALS FREE ACCESS
    The end-loss ion flux in GAMMA 10 is measured with a view to use it for a divertor simulation experiment or other studies that require high-performance plasma flux. First, the basic parameters of the end-loss ion flux, such as its energy and current density, were measured in typical plasma shots in GAMMA 10. A diagnostic device, the end loss ion energy analyzer (ELIEA), was used to the measure these parameters. An investigation of the relationship between the parameters of the end-loss ion flux and the plasma parameters in the central-cell revealed linear-like relationships between these parameters. We also analyzed the effects of plasma heating and fueling by using devices installed in GAMMA 10 (ion cyclotron resonance frequency (ICRF), electron cyclotron resonance heating (ECRH) and supersonic molecular beam injection (SMBI)) in order to generate more intense ion flux. The results showed that the energy distribution of the ion flux is more closely resembles a double component Maxwellian than a simple Maxwellian. Plasma heating schemes such as ECRH and ICRF are found to be effective for the generation of a more intense ion flux.
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  • Hisato TAKEDA, Yousuke NAKASHIMA, Katsuhiro HOSOI, Kazuya ICHIMURA, Te ...
    2012 Volume 7 Pages 2405151
    Published: November 22, 2012
    Released: January 10, 2014
    JOURNALS FREE ACCESS
    In GAMMA 10 which is a tandem mirror machine in Plasma Research Center at University of Tsukuba, divertor simulation studies have been planned and started. This paper describes the recent results of divertor simulation experiment in the GAMMA 10 end-mirror cell. As a part of these experiments, characteristics of end-loss plasma have been investigated by using probes and calorimeters. The diagnostics are installed at two positions, which are separated by 40 cm in the axial direction. In the results, ion temperature of the end-loss flux in GAMMA 10 is much higher than that of other divertor simulators. It is found that heat-flux density can be controlled within the range from 0.4 to 0.8 MW/m2 by changing the ICRF power. In addition, heat flux density has strong dependence on diamagnetism in the central-cell which is time integrated in the plasma duration. While, particle-flux density is proportional to electron line-density in the central-cell. Particle and heat fluxes measured at axially different positions agree well with the calculation results for considering the influence of particle reflection phenomena.
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  • Kazuo OGURA, Hiroshi IIDUKA, Kiyoyuki YAMBE
    2012 Volume 7 Pages 2406022
    Published: March 15, 2012
    Released: January 10, 2014
    JOURNALS FREE ACCESS
    We present a numerical and experimental study of the dispersion characteristics of cylindrical surface waves on metallic cylinders with rectangular corrugations. In actual devices, reflections at both ends quantize the electromagnetic modes into resonant axial modes. A cavity method based on axial mode measurements is applied to study the properties of the cylindrical surface waves. The resonances are relatively sharp near the upper cutoff, but at low frequencies far from the cutoff, the resonances broaden to resemble those of a Sommerfeld wave.
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