237 results on '"Suguru Masuzaki"'
Search Results
2. Comparison of Hydrogen Isotope Retention in Divertor Tiles of JET with the ITER-Like Wall Following Campaigns in 2011–2012 and 2015–2016
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M. Oyaidzu, Moeko Nakata, Marek Rubel, Masayuki Tokitani, Yasuhisa Oya, Teppei Otsuka, Jari Likonen, Anna Widdowson, Jet Contributors, Fei Sun, Suguru Masuzaki, Kanetsuku Isobe, and Nobuyuki Asakura
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Nuclear and High Energy Physics ,Jet (fluid) ,Materials science ,plasma wall interactions ,020209 energy ,Mechanical Engineering ,Hydrogen isotope ,Divertor ,Nuclear engineering ,02 engineering and technology ,Plasma ,01 natural sciences ,010305 fluids & plasmas ,Chemical state ,Nuclear Energy and Engineering ,JET–ITER-like wal ,0103 physical sciences ,0202 electrical engineering, electronic engineering, information engineering ,General Materials Science ,Hydrogen isotope retention ,Civil and Structural Engineering - Abstract
Hydrogen isotope retention and chemical state for the tiles exposed to plasma in the JET–ITER-like wall (ILW) during two campaigns in 2011–2012 (first campaign, ILW-1) and 2015–2016 (third campaign, ILW-3) were studied and compared by means of X-ray photoelectron spectroscopy and thermal desorption spectroscopy. In both campaigns the upper part of the inner divertor tiles was the deposition-dominated area, while erosion was observed on the outer divertor tiles. Therefore, higher deuterium retention was found on the inner divertor tiles. The major D desorption peak for the inner divertor tiles from ILW-3 was located at the temperature range of 470°C to 520°C, which was higher than measured after ILW-1: 370°C to 430°C. The XPS analyses showed the formation of a BeO layer on the ILW-3 inner divertor tiles, while after ILW-1 the layers also contained a significant amount of carbon. Deuterium retention was reduced toward the outer divertor tiles. The differences could be related to the difference in the power level in the two campaigns.
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- 2020
3. Enhanced D retention in RAFM steel caused by D bubbles formed inside Cr-rich surface layer
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R.P. Doerner, Daisuke Nishijima, George Tynan, M. Tokitani, Mitsutaka Miyamoto, Suguru Masuzaki, and D. Nagata
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Nuclear and High Energy Physics ,GDOES ,Materials science ,Materials Science (miscellaneous) ,TK9001-9401 ,Nuclear Energy and Engineering ,RAFM steel ,Deuterium retention ,EDX ,TEM ,Nuclear engineering. Atomic power ,Bubbles ,Surface layer ,Composite material - Abstract
To identify the cause for an enhanced D retention in RAFM (reduced-activation ferritic/martensitic) steels with decreasing incident D fluence, ϕD, CLAM steel samples were exposed to D plasma at a low ϕD ∼ 3 × 1023 m−2 in the PISCES-A linear plasma device. At this low ϕD, a Cr-rich surface layer with a thickness of ∼ 10 nm still remains, confirmed with cross-sectional EDX (energy dispersive x-ray spectroscopy) elemental mapping. TEM (transmission electron microscopy) observations of a plasma-exposed surface reveal the formation of D bubbles inside the Cr-rich surface layer, while no D bubbles are seen in an unexposed surface. Furthermore, depth profiling using GDOES (glow discharge optical emission spectroscopy) detects D in the Cr-rich surface layer. These results demonstrate that the D retention at low ϕD is enhanced by D atoms trapped in bubbles created inside the Cr-rich surface layer.
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- 2021
4. Investigation on tritium retention and surface properties on the first wall in the large helical Device
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Suguru Masuzaki, Y. Torikai, Yasuhisa Oya, Nobuaki Yoshida, Teppei Otsuka, Gen Motojima, Masayuki Tokitani, and Miyuki Yajima
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inorganic chemicals ,Nuclear and High Energy Physics ,Materials science ,Materials Science (miscellaneous) ,chemistry.chemical_element ,Tritium ,Thermal desorption ,01 natural sciences ,010305 fluids & plasmas ,Large Helical Device ,0103 physical sciences ,Nuclear fusion ,Deposition (phase transition) ,Graphite ,010302 applied physics ,First wall ,Divertor ,TK9001-9401 ,Plasma ,Nuclear Energy and Engineering ,chemistry ,Deposition layer ,Nuclear engineering. Atomic power ,Atomic physics ,Carbon ,The Large Helical Device (LHD) - Abstract
In the Large Helical Device (LHD), the first deuterium plasma experiment was conducted in 2017. To investigate tritium migration in the LHD vacuum vessel, long-term material probes were installed on the first wall before the deuterium plasma experiment. After the experiment, the microstructure and amount of tritium remaining in each probe were analyzed. The results showed that a relatively large amount of tritium remained in the probes on the first wall, forming a thick deposition layer, rather than in the probes located in the erosion-dominant area. In the deposition layers on the probes, the dominant element is carbon, which can be generated on the divertor tiles made of graphite. The result of orbit calculation of the energetic tritons in the case of the standard magnetic configuration in the LHD showed that approximately 40% of the tritons generated by deuterium–deuterium fusion reactions were promptly lost mainly to the divertor. Thermalized tritons also flew to the divertor along with the background plasma. The divertor tiles, on which the tritons impinged, were eroded by the divertor plasma, and carbon atoms and tritiated hydrocarbon molecules were generated and deposited on the first wall. This can be the dominant mechanism of tritium retention in the first wall. Among the material probes located in the erosion-dominant area, the amount of tritium remaining in the probe on which the energetic tritons impinged was relatively large. The results of the tritium balance analysis show that the first wall is not the dominant reservoir of tritium in the LHD.
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- 2021
5. Line identification of boron and nitrogen emissions in EUV and VUV wavelength ranges in the impurity powder dropping experiments of LHD and its application to spectroscopic diagnostics
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E.P. Gilson, Alex Nagy, David Gates, Mamoru Shoji, Naoko Ashikawa, Robert Lunsford, Tetsutarou Oishi, Yasuko Kawamoto, F. Nespoli, Suguru Masuzaki, Zhen Sun, Chihiro Suzuki, Motoshi Goto, Shigeru Morita, and Tomohiro Morisaki
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Materials science ,business.industry ,chemistry.chemical_element ,Condensed Matter Physics ,Nitrogen ,Vacuum ultraviolet ,Wavelength ,Large Helical Device ,chemistry ,Impurity ,Optoelectronics ,Boron ,business ,Line (formation) - Published
- 2021
6. Deformation and fracture behavior of the W/ODS-Cu joint fabricated by the advanced brazing technique
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Masayuki Tokitani, Akio Sagara, Teruya Tanaka, Takeo Muroga, Hitoshi Tamura, Y. Hamaji, Suguru Masuzaki, Yutaka Hiraoka, and H. Noto
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Materials science ,Mechanical Engineering ,Divertor ,Fracture mechanics ,01 natural sciences ,010305 fluids & plasmas ,Nuclear Energy and Engineering ,Flexural strength ,0103 physical sciences ,Fracture (geology) ,Brazing ,General Materials Science ,Grain boundary ,Deformation (engineering) ,Composite material ,010306 general physics ,Joint (geology) ,Civil and Structural Engineering - Abstract
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.
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- 2019
7. New installation of in-vessel Non Evaporable Getter (NEG) pumps for the divertor pump in the LHD
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Gen Motojima, Mitsuhiro Yokota, T. Murase, Suguru Masuzaki, H. Sakurai, Fabrizio Siviero, M. Mura, T. Morisaki, Enrico Maccallini, A. Ferrara, H. Ogawa, and Mamoru Shoji
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Glow discharge ,Materials science ,Hydrogen ,Mechanical Engineering ,Divertor ,Nuclear engineering ,chemistry.chemical_element ,Non-Evaporable Getter ,Plasma radiation ,01 natural sciences ,010305 fluids & plasmas ,Large Helical Device ,Nuclear Energy and Engineering ,chemistry ,Power consumption ,Getter ,0103 physical sciences ,General Materials Science ,010306 general physics ,Civil and Structural Engineering - Abstract
A case of the application of Non Evaporable Getter (NEG) pumps to fusion devices is discussed. Compared with other pumping technologies used in fusion devices, the NEG pumps have advantages in terms of operation temperature at 100–200 °C (in which there is a high tolerance of the temperature increase due to a plasma radiation and/or an unexpected heat load from heating devices), low power consumption and compactness. Moreover, in the case of a power outage getters continue pumping and do not release hydrogen and its isotopes. The NEG pumps have been installed as an in-vessel pump into the divertor region in the Large Helical Device (LHD) for the first time. The pumping performance test shows that the effective pumping speed is 10 m3/s in hydrogen, which is close to the target value. In addition, the exposure of the NEG pumps to boronization and glow discharge treatments have been tested. While the boronization and glow discharge under noble gases do not show effect on the pumping performances of the NEG pumps, the establishment of an operational scenario including hydrogen and deuterium glow discharges consistent with other pumping facilities is still an aspect to be refined. However, the use of NEG pumps in LHD divertor is a milestone for the possibility of using this technology in future fusion devices.
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- 2019
8. Analysis of indefinite divertor footprint with proper orthogonal decomposition in hydrogen/deuterium plasmas in LHD
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Hirohiko Tanaka, Noriyasu Ohno, T. Morisaki, Masahiro Kobayashi, Suguru Masuzaki, Yasuhiro Suzuki, and Gakushi Kawamura
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Nuclear and High Energy Physics ,Materials science ,Hydrogen ,Materials Science (miscellaneous) ,Divertor ,chemistry.chemical_element ,Electron ,Plasma ,lcsh:TK9001-9401 ,Large Helical Device ,Nuclear Energy and Engineering ,chemistry ,Deuterium ,Physics::Plasma Physics ,Kinetic isotope effect ,lcsh:Nuclear engineering. Atomic power ,Atomic physics ,Pressure gradient - Abstract
Divertor particle flux signals on divertor plates were analyzed with the multivariable analysis technique, namely, proper orthogonal decomposition (POD), to characterize the indefinite divertor footprint in the Large Helical Device. To verify whether isotope effects exist in relation to the divertor plasmas, the POD outputs in hydrogen (H) and deuterium (D) plasmas were first compared by analyzing a number of discharges. It was found that the dominant components of the divertor particle flux profiles in H and D plasmas were similar, indicating no isotope effect. In addition, total reconstructed divertor particle fluxes in H plasmas were higher than those in D plasmas, suggesting an isotope effect. Furthermore, the footprint-profile modification with from the SOL-side to the private-side peak due to the edge electron pressure gradient in H and D plasmas were slightly different, relating to an isotope effect. Keywords: Divertor footprint, Isotope effect, Langmuir probe, Proper orthogonal decomposition, LHD
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- 2019
9. Spectroscopic studies on the enhanced radiation with high Z rare gas seeding for mitigating divertor heat loads in LHD plasmas
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Chihiro Suzuki, Masahiro Kobayashi, Tsuyoshi Akiyama, Suguru Masuzaki, Kiyofumi Mukai, Byron J. Peterson, and Izumi Murakami
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Nuclear and High Energy Physics ,Materials science ,Materials Science (miscellaneous) ,Divertor ,Krypton ,chemistry.chemical_element ,lcsh:TK9001-9401 ,Ion ,Neon ,Large Helical Device ,Nuclear Energy and Engineering ,chemistry ,Physics::Plasma Physics ,Extreme ultraviolet ,Emissivity ,lcsh:Nuclear engineering. Atomic power ,Emission spectrum ,Atomic physics - Abstract
We have measured extreme ultraviolet (EUV) and soft X-ray emission spectra of krypton (Kr) ions in the Large Helical Device (LHD) plasmas with Kr gas seeding intended for mitigating divertor heat loads. Several isolated lines or quasi-continuum bands from ion stages in the range of Kr5+–Kr25+ have been identified in a variety of wavelength ranges. The time trends of the line intensities indicate major contributions of ions with M-shell (n = 3) outermost electrons to the enhanced total radiation in the region closer to the core plasma. A line intensity ratio analysis for Kr25+ implies an unignorable contribution of the recombining component for the emissivity from this ion stage. A difference between the trends of neon (Ne) and Kr is also observed in a discharge with a combination of Kr and Ne seedings intended for more effective radiation enhancement. Keywords: Collisional-radiative model, Radiation enhancement, Impurity seeding, Large Helical Device, Spectroscopy, Krypton
- Published
- 2019
10. Impact of additional plasma heating on detached plasma formation in divertor simulation experiments using the GAMMA 10/PDX tandem mirror
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A. Terakado, Akiyoshi Hatayama, M. Fukumoto, Tsuyoshi Imai, Sotaro Yamashita, Makoto Ichimura, Ryutaro Minami, Junko Kohagura, Tsuyoshi Kariya, Yousuke Nakashima, Satoshi Togo, Kunpei Nojiri, M. Hirata, Mamoru Shoji, M. M. Islam, Keiji Sawada, Suguru Masuzaki, Tomoharu Numakura, Tsubasa Yoshimoto, T. Nakano, Md. Shahinul Islam, Nobuyuki Asakura, Naomichi Ezumi, Mizuki Sakamoto, Ryuya Ikezoe, Takaaki Iijima, and Masayuki Yoshikawa
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Nuclear and High Energy Physics ,Materials science ,Plasma parameters ,Materials Science (miscellaneous) ,Divertor ,Astrophysics::High Energy Astrophysical Phenomena ,Cyclotron ,Mathematics::General Topology ,Electron ,Plasma ,lcsh:TK9001-9401 ,Ion ,law.invention ,Nuclear Energy and Engineering ,Impurity ,law ,Physics::Plasma Physics ,Ionization ,lcsh:Nuclear engineering. Atomic power ,Atomic physics - Abstract
The transition of a detached to attached plasma experiment has been performed in GAMMA 10/PDX by applying an additional plasma heating pulse of electron cyclotron heating (ECH). In a plasma detachment experiment, a short pulse (25 ms) of ECH was applied at the east plug-cell to examine the effects of electron heating on the plasma parameters in the D-module. It was found that the ion flux increased significantly during ECH injection. In the ECH injection period, the ion flux increased with the increasing impurity injection, which indicates the impurity ion components are enhanced by ionization in the D-module due to the application of ECH. During ECH injection period, the increase of the ion flux near the corner of the target plate shows a clear dependence on the gas species. The 2D visible emission has been captured by the high-speed camera and a bright emission near the corner of the target has been observed. The spectroscopic measurement in the D-module shows that the impurity ion emission increases remarkably during ECH injection. These results indicate the detached plasma change to the attached state. Keywords: GAMMA 10/PDX, Divertor simulation experiment, Plasma detachment, Detached to attached transition
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- 2019
11. Demonstration of suppression of dust generation and partial reduction of the hydrogen retention by tungsten coated graphite divertor tiles in LHD
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Masayuki Tokitani, Suguru Masuzaki, and T. Murase
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010302 applied physics ,Nuclear and High Energy Physics ,Materials science ,Materials Science (miscellaneous) ,Divertor ,chemistry.chemical_element ,Plasma ,Tungsten ,01 natural sciences ,lcsh:TK9001-9401 ,010305 fluids & plasmas ,Large Helical Device ,Nuclear Energy and Engineering ,chemistry ,visual_art ,0103 physical sciences ,visual_art.visual_art_medium ,lcsh:Nuclear engineering. Atomic power ,Graphite ,Tile ,Composite material ,Layer (electronics) ,Carbon - Abstract
Three sets of tungsten coated graphite divertor tiles (VPS-W tiles) were installed in the closed helical divertor of the Large Helical Device (LHD) instead of the graphite divertor tiles in the 2012FY plasma campaign for examining the plasma wall interaction (PWI) with divertor plasma. The first wall panels and divertor tiles of the LHD consist of stainless steels (SUS316L) and graphite, respectively. The carbon based mixed-material deposition layer including a very small amount of Fe element has been formed on not only the divertor tiles but also the first walls near the divertor tiles through the PWI processes. Such a mixed layer often causes undesirable influences for maintaining a long pulse discharge in LHD, e.g., changing the particle recycling properties and dust generation. After the single plasma campaign of the 2012FY, we confirmed drastic suppression of the mixed-material deposition layer on the first wall panels just under the VPS-W tiles. On the other hand, carbon based mixed-material deposition layer was formed on the VPS-W tile surface, where the amount of the hydrogen retention was estimated to be over ∼4 × 1021 H/m2. If we would want further suppression of the hydrogen retention on the VPS-W tiles, the plasma facing components should be replaced to a full metal wall to avoid formation of the carbon co-deposition layer. Keywords: Tungsten divertor, Dust, Hydrogen retention
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- 2019
12. Core plasma confinement during detachment transition with RMP application in LHD
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Yoshiro Narushima, Masayuki Yokoyama, Ryosuke Seki, Takeshi Ido, Masahiro Kobayashi, Ichihiro Yamada, Tokihiko Tokuzawa, Kenji Tanaka, and Suguru Masuzaki
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Nuclear and High Energy Physics ,Electron density ,Materials science ,Materials Science (miscellaneous) ,Atmospheric-pressure plasma ,Plasma ,Radiation ,lcsh:TK9001-9401 ,01 natural sciences ,010305 fluids & plasmas ,Core (optical fiber) ,Nuclear Energy and Engineering ,Physics::Plasma Physics ,Phase (matter) ,0103 physical sciences ,Plasma parameter ,lcsh:Nuclear engineering. Atomic power ,Electron temperature ,Atomic physics ,010306 general physics - Abstract
The core plasma confinement during detachment phase is investigated in the discharges with application of resonant magnetic perturbation (RMP) field in LHD. The RMP application creates a remnant magnetic island in the edge stochastic layer, which largely changes the plasma parameter profiles including impurity radiation. The electron temperature and pressure profiles are flattened at the island, while the electron density is slightly peaked at the edge of the island. The estimated impurity radiation profile is enhanced and fixed around the magnetic island during the detached phase, where the discharge is stably sustained with controlled level of radiation. Without RMP, the radiation penetrates the confinement region, leading to radiation collapse. It is found that in the case of the RMP application the plasma stored energy increases discontinuously at the detachment transition. In spite of the reduced effective plasma volume caused by the edge magnetic island and by the enhanced radiation there, the central plasma pressure finally exceeds the case without RMP. This is caused by the pressure profile peaking at the central region in the case with RMP. These results indicate clear change of core plasma confinement during the detached phase with RMP.
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- 2018
13. Impurity transport simulation in the peripheral plasma in the large helical device with tungsten closed helical divertor
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Suguru Masuzaki, Gakushi Kawamura, Mamoru Shoji, and T. Morisaki
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Nuclear and High Energy Physics ,Materials science ,Materials Science (miscellaneous) ,Divertor ,chemistry.chemical_element ,Plasma ,Tungsten ,equipment and supplies ,01 natural sciences ,lcsh:TK9001-9401 ,010305 fluids & plasmas ,Ion ,Electric arc ,Large Helical Device ,Nuclear Energy and Engineering ,chemistry ,Impurity ,Sputtering ,0103 physical sciences ,lcsh:Nuclear engineering. Atomic power ,Atomic physics ,010306 general physics - Abstract
Long pulse plasma discharges in the Large Helical Device have been often interrupted by iron dust emission induced by electric arcing on the surface of the vacuum vessel. The iron ions in the peripheral plasma induced by the dust emission enhance physical sputtering and self-sputtering on the divertor plates, which can interrupt the long pulse plasma discharges by radiation collapse. The impurity transport simulation for a tungsten divertor configuration is performed using a three-dimensional edge plasma simulation code (EMC3-EIRENE) under the condition where the iron ions produced by the dust emission cause the sputtering on the tungsten divertor plates. The simulation shows that the sputtered tungsten significantly increases the radiation power by a factor of about 20 compared to that for the carbon divertor configuration in a high plasma temperature condition. The simulation reveals that plasma discharge operation with a high plasma density is desirable for the tungsten divertor. In this operational regime, the radiation power by tungsten ions is significantly reduced by the combined effect of the suppression of the sputtered tungsten and the control of the tungsten ion accumulation by the reduced thermal force in the peripheral plasma. Keywords: Tungsten, EMC3-EIRENE, Long pulse discharge, Sputtering, Dust
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- 2018
14. Heat loading behavior and thermomechanical analyses on plasma spray tungsten coated reduced-activation ferritic/martensitic steel
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Koichiro Ezato, M. Tokitani, T. Hotta, Suguru Masuzaki, Masato Akiba, Mikio Enoeda, Akira Kurumada, Katsumasa Nakamura, S. Suzuki, Kazutoshi Tokunaga, K. Araki, and Makoto Hasegawa
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Materials science ,Mechanical Engineering ,chemistry.chemical_element ,Atmospheric-pressure plasma ,engineering.material ,Tungsten ,01 natural sciences ,010305 fluids & plasmas ,Nuclear Energy and Engineering ,Coating ,Heat flux ,chemistry ,Martensite ,0103 physical sciences ,Thermal ,engineering ,Brazing ,General Materials Science ,Composite material ,010306 general physics ,Thermal spraying ,Civil and Structural Engineering - Abstract
Tungsten coating with a thickness of 1 mm on reduced-activation ferritic/martensitic steel (RAF/M) F82H (Fe-8Cr-2W), which is a leading structural material candidate for DEMO, have been produced by Atmospheric Plasma Spraying (APS) and Vacuum Plasma Spraying (VPS). Heat loading experiments on thermal response and fatigue using an electron beam have been carried out on W-coated F82H brazed on oxygen free high purity copper (OFHC) block with a cooling tube to evaluate their possibility as a plasma-facing armor in the fusion device. In addition, quantitative analyses about temperature profiles and thermal stress have been carried out using FEA. Thermal response experiments show that the temperatures increased monotonically with increasing heat flux. Surface temperature of the VPS-W/F82H/OFHC is always lower than that of the APS-W/F82H/OFHC. Surface modification, exfoliation and crack are not formed by thermal fatigue experiments up to 200 cycles at a heat flux of 3.2 MW/m2. Thermal response experiments under steady state condition have been successfully modeled by FEA. Interfacial strength of VPS-W and F82H is also discussed using the FEA results.
- Published
- 2018
15. Simulation of Impurity Transport and Deposition in the Closed Helical Divertor in the Large Helical Device
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S. Brezinsek, A. Kirschner, A. Eksaeva, Gakushi Kawamura, Mamoru Shoji, D. Borodin, Juri Romazanov, and Suguru Masuzaki
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Materials science ,peripheral plasma ,Divertor ,Condensed Matter Physics ,simulation ,7. Clean energy ,Molecular physics ,emc3-eirene ,Large Helical Device ,ero2.0 ,impurity transport ,plasma wall interaction ,13. Climate action ,Impurity ,Physics::Plasma Physics ,divertor ,ddc:530 ,large helical device ,Deposition (chemistry) - Abstract
Long pulse discharges in the Large Helical Device have often been interrupted by large amounts of dust particle emission from the divertor region caused by the exfoliation of carbon-rich mixed material deposition layers. The plasma wall interaction code ERO2.0 has provided the simulation results of the three-dimensional distribution of the carbon flux density in the divertor region which is quite reasonable with the observed distribution of the carbon-rich deposition layers. The code has also succeeded in reproducing the reduction of the carbon deposition layers on dome plates by changing the target plate configuration in the divertor region. The ERO2.0 simulations have also successfully explained dust particle emission from the inboard side near the equatorial plane for the new target plate configuration at the termination of a long pulse discharge. These simulation results prove that the ERO2.0 code is applicable to predicting the possible position from where the dust particles are released, and to designing an optimized divertor configuration for performing stable long pulse discharges with controlled dust particle emission.
- Published
- 2021
- Full Text
- View/download PDF
16. Tritium distribution analysis of Be limiter tiles from JET-ITER like wall campaigns using imaging plate technique and β-ray induced X-ray spectrometry
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Makoto Oyaizu, H. Kurotaki, Yuji Hatano, Haruto Nakamura, Masayuki Tokitani, Suguru Masuzaki, Masanori Hara, Yasuhisa Oya, S. Jachmich, S. E. Lee, Nobuyuki Asakura, K. Helariutta, Marek Rubel, D. Hamaguchi, Anna Widdowson, J. Likonen, Materials Physics, Tracers in Molecular Imaging (TRIM), and Department of Chemistry
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Materials science ,Tokamak ,Joint European Torus ,Analytical chemistry ,chemistry.chemical_element ,Tritium analysis ,Mass spectrometry ,01 natural sciences ,114 Physical sciences ,010305 fluids & plasmas ,law.invention ,law ,0103 physical sciences ,Limiter ,General Materials Science ,010306 general physics ,Civil and Structural Engineering ,Ion beam analysis ,ITER-like wall ,Mechanical Engineering ,X-ray ,Radiography ,Nuclear Energy and Engineering ,chemistry ,Deuterium ,Beryllium - Abstract
Tritium (T) distribution on the plasma-facing surfaces (PFSs) and inside castellation of Be limiter tiles from the JET tokamak with the ITER-like wall (ILW) was analyzed using imaging plate (IP) technique and β-ray induced X-ray spectrometry (BIXS). Regarding to PFSs, the outer poloidal limiter (OPL) showed significantly higher T concentrations than the inner wall guard limiter (IWGL) and upper dump plate (DP). The concentration of T on OPL was high at the central part. However, deuterium (D) and metallic impurities showed maximum concentration at the edges. This difference in distributions indicated different deposition and retention mechanisms between T and D. In contrast, deposition profiles of T concentrations on the castellated surfaces extended up to ∼ 5 mm into the gap, i.e. were similar to those of D and metallic impurities found by ion beam analysis.
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- 2020
17. Recent results from deuterium experiments on the large helical device and their contribution to fusion reactor development
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Masaki Osakabe, Hiromi Takahashi, Hiroshi Yamada, Kenji Tanaka, Tatsuya Kobayashi, Katsumi Ida, Satoshi Ohdachi, Jacobo Varela, Kunihiro Ogawa, Masahiro Kobayashi, Katsuyoshi Tsumori, Katsunori Ikeda, Suguru Masuzaki, Masahiro Tanaka, Motoki Nakata, Sadayoshi Murakami, Shigeru Inagaki, Kiyofumi Mukai, Mizuki Sakamoto, Kazunobu Nagasaki, Yasuhiro Suzuki, Mitsutaka Isobe, Tomohiro Morisaki, and the LHD Experiment Group
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Nuclear and High Energy Physics ,deuterium experiment ,Materials science ,Nuclear engineering ,energetic particle confinement ,Fusion power ,Condensed Matter Physics ,stellarator ,Large Helical Device ,Deuterium ,Physics::Plasma Physics ,LHD ,RMP induced H-mode ,tritium mass balance ,isotope effect - Abstract
In recent deuterium experiments on the large helical device (LHD), we succeeded in expanding the temperature domain to higher regions for both electron and ion temperatures. Suppression of the energetic particle driven resistive interchange mode (EIC) by a moderate electron temperature increase is a key technique to extend the high temperature domain of LHD plasmas. We found a clear isotope effect in the formation of the internal transport barrier in high temperature plasmas. A new technique to measure the hydrogen isotope fraction was developed in the LHD in order to investigate the behavior of the isotope mixing state. The technique revealed that the non-mixing and the mixing states of hydrogen isotopes can be realized in plasmas. In deuterium plasmas, we also succeeded in simultaneously realizing the formation of the edge transport barrier (ETB) and the divertor detachment. It is found that resonant magnetic perturbation plays an important role in the simultaneous formation of the ETB and the detachment. Contributions to fusion reactor development from the engineering point of view, i.e. negative-ion based neutral beam injector research and the mass balance study of tritium, are also discussed.
- Published
- 2022
18. Transition between Isotope-Mixing and Nonmixing States in Hydrogen-Deuterium Mixture Plasmas
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Mikirou Yoshinuma, Katsumi Ida, Kenji Tanaka, Gen Motojima, Motoki Nakata, T. Kobayashi, Suguru Masuzaki, Kotaro Yamasaki, Ryuichi Sakamoto, and Y. Fujiwara
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Electron density ,Materials science ,Hydrogen ,Isotope ,Turbulence ,digestive, oral, and skin physiology ,General Physics and Astronomy ,chemistry.chemical_element ,Plasma ,01 natural sciences ,Large Helical Device ,chemistry ,Deuterium ,Physics::Plasma Physics ,0103 physical sciences ,Physics::Atomic Physics ,Atomic physics ,Nuclear Experiment ,010306 general physics ,Beam (structure) - Abstract
The transition between isotope-mixing and nonmixing states in hydrogen-deuterium mixture plasmas is observed in the isotope (hydrogen and deuterium) mixture plasma in the Large Helical Device. In the nonmixing state, the isotope density ratio profile is nonuniform when the beam fueling isotope species differs from the recycling isotope species and the profile varies significantly depending on the ratio of the recycling isotope species, although the electron density profile shape is unchanged. The fast transition from nonmixing state to isotope-mixing state (nearly uniform profile of isotope ion density ratio) is observed associated with the change of electron density profile from peaked to hollow profile by the pellet injection near the plasma periphery. The transition from nonmixing to isotope-mixing state strongly correlates with the increase of turbulence measurements and the transition of turbulence state from TEM to ion temperature gradient is predicted by gyrokinetic simulation.
- Published
- 2020
19. In-vessel colorimetry of Wendelstein 7-X first wall components: variation of layer deposition distribution in OP1.2a and OP1.2b
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Chandra Prakash Dhard, Gen Motojima, Suguru Masuzaki, Dirk Naujoks, Y. Hayashi, S. Brezinsek, M. Krause, and W7-X Team, Max Planck Institute for Plasma Physics, Max Planck Society
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010302 applied physics ,Materials science ,0103 physical sciences ,Analytical chemistry ,Deposition (phase transition) ,Wendelstein 7-X ,Condensed Matter Physics ,01 natural sciences ,Layer (electronics) ,Mathematical Physics ,Atomic and Molecular Physics, and Optics ,Colorimetry (chemical method) ,010305 fluids & plasmas - Published
- 2020
- Full Text
- View/download PDF
20. Extended investigations of isotope effects on ECRH plasma in LHD
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Mamoru Shoji, Katsumi Ida, Ryosuke Seki, Gen Motojima, Ichihiro Yamada, Masanori Nunami, S. Satake, Hiroto Takahashi, Hisamichi Funaba, Masayuki Yokoyama, Tuomas Tala, Hiroshi Yamada, Clive Michael, Suguru Masuzaki, Shin Kubo, L. N. Vacheslavov, Y. Ohtani, Motoki Nakata, Masaki Osakabe, Tomohiro Morisaki, Tokihiko Tokuzawa, Kenji Tanaka, Toru Ii Tsujimura, Motoshi Goto, Ryo Yasuhara, Yasuo Yoshimura, Mikiro Yoshinuma, Felix Warmer, Y. Takemura, Tsuyoshi Akiyama, Hiroe Igami, Toshiki Kinoshita, Takashi Shimozuma, and LHD Experiment Group
- Subjects
Materials science ,Density gradient ,Electron ,Collisionality ,01 natural sciences ,Instability ,010305 fluids & plasmas ,law.invention ,Ion ,stellarator ,law ,Physics::Plasma Physics ,particle transport ,0103 physical sciences ,Diffusion (business) ,010306 general physics ,energy transport ,turbulence ,Plasma ,Condensed Matter Physics ,Nuclear Energy and Engineering ,Physics::Space Physics ,Atomic physics ,Stellarator ,isotope effect - Abstract
Isotope effects of ECRH plasma in LHD were investigated in detail. A clear difference of transport and turbulence characteristics in H and D plasmas was found in the core region, with normalized radius ρ < 0.8 in high collisionality regime. On the other hand, differences of transport and turbulence were relatively small in low collisionality regime. Power balance analysis and neoclassical calculation showed a reduction of the anomalous contribution to electron and ion transport in D plasma compared with H plasma in the high collisionality regime. In core region, density modulation experiments also showed more reduced particle diffusion in D plasma than in H plasma, in the high collisionality regime. Ion scale turbulence was clearly reduced at ρ < 0.8 in high collisionality regime in D plasma compared with H plasma. The gyrokinetic linear analyses showed that the dominant instability ρ = 0.5 and 0.8 were ion temperature gradient mode (ITG). The linear growth rate of ITG was reduced in D plasma than in H plasma in high collisionality regime. This is due to the lower normalized ITG and density gradient. More hollowed density profile in D plasma is likely to be the key control parameter. Present analyses suggest that anomalous process play a role to make hollower density profiles in D plasma rather than neoclassical process. Electron scale turbulence were also investigated from the measurements and linear gyrokinetic simulations.
- Published
- 2020
21. Full‐torus impurity transport simulation for optimizing plasma discharge operation using a multi‐species impurity powder dropper in the large helical device
- Author
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Gakushi Kawamura, Yoshihiko Uesugi, Roman Smirnov, Robert Lunsford, Erik P. Gilson, Suguru Masuzaki, Yasunori Tanaka, Naoko Ashikawa, and Mamoru Shoji
- Subjects
Imagination ,Thesaurus (information retrieval) ,EMC3-EIRENE ,Chemical substance ,Materials science ,media_common.quotation_subject ,impurity powder dropper ,Torus ,Plasma ,Condensed Matter Physics ,DUSTT ,Computational physics ,Search engine ,Large Helical Device ,ergodic layer ,Physics::Plasma Physics ,Impurity ,Condensed Matter::Superconductivity ,Condensed Matter::Strongly Correlated Electrons ,LHD ,media_common - Abstract
The transport of impurities supplied by a multi-species impurity powder dropper (IPD) in the large helical device (LHD) is investigated using a three-dimensional peripheral plasma fluid code (EMC3-EIRENE) coupled with a dust transport simulation code (DUSTT). The trajectories of impurity powder particles (Boron, Carbon, Iron, and Tungsten) dropped from the IPD and the impurity transport in the peripheral plasma are studied in a full-torus geometry. The simulation reveals an appropriate size of the impurity powder particles and an optimum operational range of the dust drop rates for investigating the impurity transport without inducing radiation collapse. The simulation also predicts a favourable plasma discharge condition for wall conditioning (boronization) using the IPD in order to deposit boron to high plasma flux and neutral particle density areas in the divertor region in the inboard side of the torus.
- Published
- 2019
22. Conductive and viscous sub-layers on forced convection and mechanism of critical heat flux during flow boiling of subcooled water in a circular tube at high liquid Reynolds number
- Author
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Koichi Hata, Qiusheng Liu, and Suguru Masuzaki
- Subjects
Fluid Flow and Transfer Processes ,Materials science ,Water flow ,Turbulence ,Critical heat flux ,020209 energy ,02 engineering and technology ,Mechanics ,Condensed Matter Physics ,Leidenfrost effect ,Forced convection ,Physics::Fluid Dynamics ,Subcooling ,020401 chemical engineering ,Boiling ,0202 electrical engineering, electronic engineering, information engineering ,0204 chemical engineering ,Nucleate boiling - Abstract
The turbulent heat transfer, the subcooled boiling heat transfer and the steady state CHF for a Pt-circular test tube of a 3 mm inner diameter and a 100 mm heated length are measured with a wide range of inlet subcooling and flow velocity at high liquid Reynolds number, i.e. Re-d = 3.01x10(4) to 1.43x10(5). The inner surface temperature of the Pt-circular test tube calculated by the steady one-dimensional heat conduction equation is compared with the values derived from authors' turbulent heat transfer correlation and with the numerical solutions of the RANS equations (Reynolds Averaged Navier-Stokes Simulation) of k-ε turbulence model for the flow velocities ranging from 4 to 21 m/s. The thicknesses of conductive sub-layer from non-boiling regime to CHF are measured by numerically analyzing the heat transfers with conductive sub-layer on forced convection and with thinner one dissipated by the evaporation on nucleate boiling. The thicknesses of viscous sub-layer on forced convection are estimated from the thicknesses of the conductive sub-layer and Prandtl numbers of the surface temperature on the heated surface. Furthermore, the thicknesses of conductive sub-layer at the CHF point are extrapolated from the measured values at various flow velocities. The experimental values of the CHF are also compared with authors' widely and precisely predictable correlations of critical heat flux during flow boiling of subcooled water and the corresponding theoretical values of the liquid sub-layer dry-out models suggested by other researchers, respectively. The authors' correlations and other researchers' theoretical values can represent the subcooled boiling CHFs obtained in this study within the ranges of -13.27 to 6.76% difference and - 32.51 to 13.16% one, respectively. A suggestion based on the experimental data as to what the dominant mechanism is for critical heat flux during flow boiling of subcooled water on a vertical circular tube is confirmed again at high liquid Reynolds number. The transitions to film boiling at the subcooled water flow boiling on the Pt test tube of d = 3 mm and L = 100 mm would occur due to the liquid sub-layer dry-out model at the steady-state CHF as well as those on the Pt test tube of d = 3 mm and L = 66.5 mm, but not due to the heterogeneous spontaneous nucleation and the hydro-dynamic instability.
- Published
- 2018
23. Deuterium retention behavior in simultaneously He+–D2+ implanted tungsten
- Author
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Suguru Masuzaki, Miyuki Yajima, Masanori Hara, Akihiro Togari, Yasuhisa Oya, Yuji Hatano, Keisuke Azuma, Masayuki Tokitani, Qilai Zhou, and Naoaki Yoshida
- Subjects
010302 applied physics ,Nuclear and High Energy Physics ,Materials science ,Astrophysics::High Energy Astrophysical Phenomena ,Materials Science (miscellaneous) ,Analytical chemistry ,chemistry.chemical_element ,Flux ,Trapping ,Tungsten ,equipment and supplies ,lcsh:TK9001-9401 ,01 natural sciences ,010305 fluids & plasmas ,Ion ,Nuclear Energy and Engineering ,chemistry ,Deuterium ,Vacancy defect ,0103 physical sciences ,lcsh:Nuclear engineering. Atomic power ,Irradiation ,Helium - Abstract
Poly-crystalline tungsten (W) samples were simultaneously irradiated with Helium (He) and Deuterium (D) ions using the triple-ion implantation device. He effect on D retention and transportation was studied using different combination of ion energies and He/D flux ratios in the simultaneous implantation. The experimental results show that D trapping at dislocation loops is significantly reduced in the case of 3 keV He+–3 keV D2+at He/D flux ratios over 0.6. D trapping by stronger trapping sites such as vacancies and vacancy clusters showed less dependence on the flux ratio. On the contrary, the D retention increases at each He/D flux ratio in the case of 3 keV He+–1 keV D2+compared to only D2+ implantation even the He/D flux ratio reaches a value of 1.0. TEM observations confirmed that dense dislocation loops are formed rather than He bubbles, which is responsible for the enhanced D retention in W. Keywords: Simultaneous implantation, D retention, Helium, Flux ratio, Transportation, Thermal desorption spectroscopy
- Published
- 2018
24. Conductive sub-layer of twisted-tape-induced swirl-flow heat transfer in vertical circular tubes with various twisted-tape inserts
- Author
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Katsuya Fukuda, Koichi Hata, and Suguru Masuzaki
- Subjects
Fluid Flow and Transfer Processes ,Pressure drop ,Materials science ,Turbulence ,business.industry ,020209 energy ,Drop (liquid) ,Thermodynamics ,02 engineering and technology ,Condensed Matter Physics ,01 natural sciences ,010305 fluids & plasmas ,Optics ,Heat flux ,0103 physical sciences ,Heat transfer ,0202 electrical engineering, electronic engineering, information engineering ,Twist ,business ,Reynolds-averaged Navier–Stokes equations ,Electrical conductor - Abstract
Twisted-tape-induced swirl-flow heat transfer due to exponentially increasing heat inputs with various exponential periods (Q = Q 0 exp(t/τ), τ = 6.04 to 23.07 s) and twisted-tape-induced pressure drop was systematically measured for various mass velocities (G = 4115 to 13,656 kg/m2 s), inlet liquid temperatures (T in = 285.88 to 299.09 K), and inlet pressures (P in = 847.45 to 943.29 kPa) using an experimental water loop flow. Measurements were made over a 59.2-mm effective length and three sections (upper, middle, and lower positions), within which four potential taps were spot-welded onto the outer surface of a 6-mm-inner-diameter, 69.6-mm-heated length, 0.4-mm-thickness platinum circular test tube. Type SUS304 twisted tapes with a width w = 5.6 mm, a thickness δ T = 0.6 mm, a total length l = 372 mm, and twist ratios y = 2.39 and 4.45 were employed in this study. The RANS equations (Reynolds Averaged Navier-Stokes Simulation) with a k–e turbulence model for a circular tube 6 mm in diameter and 636 mm in length were numerically solved for heating of water with a heated section 6 mm in diameter and 70 mm in length using the CFD code, under the same conditions as the experimental ones and considering the temperature dependence of the thermo-physical properties concerned. The theoretical values of surface heat flux q on the circular tubes with twisted tapes with twist ratios y of 2.39 and 4.45 were found to be almost in agreement with the corresponding experimental values of heat flux q, with deviations of less than 30% for the range of temperature difference between the average heater inner surface temperature and the liquid bulk mean temperature ΔT L [ = T s,av - T L , T L = (T in + T out )/2] considered in this study. The theoretical values of the local surface temperature T s , local average liquid temperature T f,av , and local liquid pressure drop ΔP x were found to be within almost 15% of the corresponding experimental ones. The thickness of the conductive sub-layer δ CSL and the nondimensional thickness of the conductive sub-layer y + CSL on the circular tubes with various twisted-tape inserts were determined on the basis of numerical solutions for the swirl velocities u sw ranging from 5.23 to 21.18 m/s. Correlations between the conductive sub-layer thickness δ CSL and the nondimensional thickness of the conductive sub-layer y + CSL for twisted-tape-induced swirl-flow heat transfer in a vertical circular tube were derived.
- Published
- 2018
25. Helium retention behavior in simultaneously He+-H2+ irradiated tungsten
- Author
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Miyuki Yajima, Suguru Masuzaki, Masanori Hara, Naoaki Yoshida, Qilai Zhou, Yuji Hatano, Akihiro Togari, Yasuhisa Oya, Masayuki Tokitani, and Keisuke Azuma
- Subjects
010302 applied physics ,Nuclear and High Energy Physics ,Materials science ,Hydrogen ,Thermal desorption spectroscopy ,Analytical chemistry ,chemistry.chemical_element ,Atmospheric temperature range ,Tungsten ,equipment and supplies ,01 natural sciences ,Fluence ,010305 fluids & plasmas ,Nuclear Energy and Engineering ,chemistry ,Desorption ,0103 physical sciences ,General Materials Science ,Irradiation ,Helium - Abstract
The purpose of this study is to elucidate helium (He) retention behavior in tungsten (W) under simultaneous He and hydrogen (H) irradiation. Polycrystalline-W was irradiated by He+ and H2+ simultaneously with the energy of 1.0 keV and 3.0 keV. He+ fluences were (0.5, 1.0, 10) × 1021 He+ m−2 and H2+ fluence was 1.0 × 1022 H+ m−2,respectively. After irradiation, He desorption behavior was investigated by high temperature thermal desorption spectroscopy (HT-TDS) in the temperature range of R.T.-1773 K. Micro-structure changes of W after irradiation were observed by TEM. It was found that simultaneous irradiation with different H2+ energy significantly changed He retention behavior. 1.0 keV H2+ suppressed the He bubble growth and no bubbles can be observed at room temperature. On the other hand, 3.0 keV H2+ facilitated the formation of He bubbles and increased the He retention due to the additional damage introduction by energetic H2+.
- Published
- 2018
26. Boiling incipience of subcooled water flowing in a narrow tube using wavelet analysis
- Author
-
Makoto Shibahara, Suguru Masuzaki, Katsuya Fukuda, Koichi Hata, and Qiusheng Liu
- Subjects
Materials science ,020209 energy ,Joule effect ,Flow (psychology) ,Energy Engineering and Power Technology ,02 engineering and technology ,Mechanics ,Industrial and Manufacturing Engineering ,Subcooling ,Superheating ,Boiling point ,Heat flux ,Boiling ,0202 electrical engineering, electronic engineering, information engineering ,Tube (fluid conveyance) - Abstract
Various incipient boiling phenomena for subcooled water flowing in a uniformly heated narrow tube were observed experimentally. The boiling signal was analyzed using the wavelet decomposition method. The boiling incipience of subcooled water in the narrow tube was recorded by a sound level meter at various flow velocities. A platinum tube was used as the experimental tube with an inner diameter of 1.0 mm. The length of the experimental tube was 23.2 mm. The tube was heated by the Joule effect using a direct current. The inlet temperature and flow velocities ranged 285–346 K and 2.5–14 m/s, respectively. The surface superheat ascended with an increase of the heat flux until the incipient boiling point was reached. The initial temperature overshoot did not appear as the outlet pressure increased. Since the existing correlations underestimated the incipient heat flux, a semi-empirical correlation of the boiling incipience was obtained based on the experimental data. The predicted value of the new correlation is in agreement with the experimental data within ±30%.
- Published
- 2018
27. Steady-state sustainment of divertor detachment with multi-species impurity seeding in LHD
- Author
-
Hikona Sakai, Byron J. Peterson, Yuki Hayashi, Toshiki Kinoshita, Kenji Tanaka, Hirohiko Tanaka, Tokihiko Tokuzawa, Suguru Masuzaki, Kiyofumi Mukai, Chihiro Suzuki, Masahiro Kobayashi, and T. Oishi
- Subjects
Nuclear and High Energy Physics ,Materials science ,Steady state (electronics) ,Divertor ,food and beverages ,Mechanics ,Condensed Matter Physics ,impurity seeding ,Impurity ,detachment ,Multi species ,divertor ,Seeding ,LHD - Abstract
Multi-species impurity seeding is an advanced operation scenario to mitigate the divertor heat load for the realization of future fusion reactors. In the large helical device, divertor detachment is successfully sustained using higher-Z (krypton, Kr) and lower-Z (neon, Ne) superimposed seeding. Emission from Kr impurities is drastically enhanced if it is followed by Ne seeding. Plasma radiation can be enhanced even at the upstream region in the edge plasma compared with Ne only seeded plasmas with suppression of impurity accumulation toward the central plasma. The characteristics of divertor heat load reduction and energy confinement are comparable between the Kr + Ne seeding and Ne only seeding under the same radiation fraction. However, while the detachment in Ne only seeding is transient, the detachment in Kr + Ne seeding is stable. It indicates that multi-species impurity seeding can be competitive for steady-state operation although further investigation is desired about the balance between divertor heat load reduction, impurity screening, and confinement degradation. The Kr emission enhancement is strongly affected by electron density and temperature at the last closed flux surface resulting in impurity penetration.
- Published
- 2021
28. The role of the graphite divertor tiles in helium retention on the LHD wall
- Author
-
Suguru Masuzaki, Yoshio Ueda, Masayuki Tokitani, Makoto Oya, Hirohiko Tanaka, and Gen Motojima
- Subjects
Nuclear and High Energy Physics ,Materials science ,Helium retention ,Thermal desorption spectroscopy ,Materials Science (miscellaneous) ,Analytical chemistry ,chemistry.chemical_element ,01 natural sciences ,Fluence ,010305 fluids & plasmas ,Large Helical Device ,Divertor ,0103 physical sciences ,Graphite ,Irradiation ,010306 general physics ,Helium ,Global particle balance ,lcsh:TK9001-9401 ,Nuclear Energy and Engineering ,chemistry ,Particle ,lcsh:Nuclear engineering. Atomic power ,LHD - Abstract
In this study, global particle balance of helium (He) long pulse discharge with graphite divertor in the Large Helical Device (LHD) and He retention in graphite from lab-scale experiments were compared in order to determine the role of the graphite divertor in He retention on the LHD wall. Global He particle balance analysis was conducted in long-pulse discharges in LHD with only turbo molecular pump. The analysis showed that static He retention was ∼2.9 × 1022 He and the ratio of retained He (= retention over fluence) was ∼0.45%. In the lab-scale study, He retention was measured by Thermal Desorption Spectroscopy (TDS) after ion irradiation or plasma exposure. The ratio of retained He were 0.1 ∼ 0.5% in all conditions, which was well consistent with LHD results. Therefore, it may be concluded that graphite divertor tiles have an important role in He absorption in the LHD wall at the initial phase of discharges.
- Published
- 2017
29. Influence of mixed material layer formation on hydrogen isotope and He retentions in W exposed to 2014 LHD experiment campaign
- Author
-
Hiroe Fujita, Masayuki Tokitani, Takumi Chikada, Suguru Masuzaki, Yuki Uemura, Keisuke Azuma, Miyuki Yajima, Naoaki Yoshida, Shodai Sakurada, Yasuhisa Oya, and Cui Hu
- Subjects
010302 applied physics ,Materials science ,Hydrogen ,Thermal desorption spectroscopy ,Scanning electron microscope ,Mechanical Engineering ,Analytical chemistry ,chemistry.chemical_element ,01 natural sciences ,010305 fluids & plasmas ,Nuclear Energy and Engineering ,X-ray photoelectron spectroscopy ,chemistry ,Transmission electron microscopy ,Desorption ,0103 physical sciences ,General Materials Science ,Atomic physics ,Layer (electronics) ,Deposition (law) ,Civil and Structural Engineering - Abstract
Influence of mixed material layer formation on hydrogen isotope retention in W exposed to 2014 Large Helical Device (LHD) experiment campaign was evaluated by Thermal Desorption Spectroscopy (TDS), X-ray Photoelectron Spectroscopy (XPS), Scanning Electron Microscope (SEM) and Transmission Electron Microscope (TEM). It was found that a lot of hydrogen isotopes were trapped by the carbon-dominated mixed-material layer deposited on the plasma facing materials. Most of He was also trapped in the carbon-dominated mixed-material layer and the corresponding desorption temperature was limited to be about 600 K, 900 K and 1200 K, respectively. However, the hydrogen retention behavior for erosion dominated area was clearly different from those for deposition dominated area and typical Plasma Wall Interaction (PWI) area, where He bubbles were introduced near the sample surface, leading to the introduction of various types of trapping sites in W.
- Published
- 2017
30. Development of the brazing technique of W and JLF-1 by Ni-P filler material
- Author
-
Y. Hamaji, Takeo Muroga, Suguru Masuzaki, Masayuki Tokitani, H. Noto, and T. Yamashita
- Subjects
Materials science ,Mechanical Engineering ,chemistry.chemical_element ,Tungsten ,01 natural sciences ,Copper ,Indentation hardness ,010305 fluids & plasmas ,Nuclear Energy and Engineering ,chemistry ,Residual stress ,Martensite ,0103 physical sciences ,Hardening (metallurgy) ,Brazing ,General Materials Science ,Composite material ,010306 general physics ,Joint (geology) ,Civil and Structural Engineering - Abstract
The reliable bonding between tungsten (W) and reduced activation ferritic/martensitic (RAFM) steel (JLF-1) was obtained by brazing with BNi-6 (Ni-11%P) filler material. In this work, a pure copper (Cu) interlayer was selected for the absorber of the residual stress, in which BNi-6 filler materials are inserted between W/Cu and Cu/JLF-1 interface. The stacked structure of the W/BNi-6/Cu/BNi-6/JLF-1 was subjected to the heat treatment procedure at 960℃. After the heat treatment procedure, fine joint structure without any crack and large size pore can be confirmed. In addition, several kinds of detailed analysis such as hardness testing and element distribution in the W/BNi-6/Cu/BNi-6/JLF-1 bonding structure, were performed. It was found that the Ni and P elements were preferentially located near each joint interface, and those elements caused a hardening effect of the bonding layer through some segregation mechanism. Also, the joint sample is subjected to the heat loading test by the electron beam device (ACT2) for evaluating the heat conducting characteristics of the joint. According to the temperature measurement during the heat loading, those hardening layers seemed to not cause the negative effects for degrading the joint properties.
- Published
- 2021
31. Quantitative evaluation of hydrogen retention of solid tin after exposure to hydrogen plasma
- Author
-
Haruka Suzuki, Suguru Masuzaki, Kota Tamura, Hirotaka Toyoda, and Junichi Miyazawa
- Subjects
Materials science ,Hydrogen ,Glow plasma ,Thermal desorption spectroscopy ,Mechanical Engineering ,Analytical chemistry ,chemistry.chemical_element ,Plasma ,01 natural sciences ,Ion fluence ,010305 fluids & plasmas ,Nuclear Energy and Engineering ,Plasma exposure ,chemistry ,0103 physical sciences ,General Materials Science ,010306 general physics ,Tin ,Civil and Structural Engineering - Abstract
In this study, the hydrogen retention properties of solid tin exposed to DC glow plasma were investigated using thermal desorption spectroscopy. The measurements were performed by varying the plasma exposure time from 0 (no exposure) to 40 min. The resulting retained hydrogen from the ion fluence of the plasma was quantified as ~10−3. The retained hydrogen increased with the exposure time.
- Published
- 2021
32. Divertor heat load distribution measurements with infrared thermography in the LHD helical divertor
- Author
-
Yuki Hayashi, Keisuke Mukai, Suguru Masuzaki, Makoto I. Kobayashi, and T. Murase
- Subjects
Materials science ,Mechanical Engineering ,Divertor ,Finite element analysis ,Mechanics ,Heat sink ,01 natural sciences ,IR thermography ,010305 fluids & plasmas ,symbols.namesake ,Nuclear Energy and Engineering ,Heat flux ,0103 physical sciences ,Thermography ,symbols ,Water cooling ,Langmuir probe ,Electron temperature ,General Materials Science ,Heat equation ,010306 general physics ,Civil and Structural Engineering - Abstract
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.
- Published
- 2021
33. Damage and deuterium retention of re-solidified tungsten following vertical displacement event-like heat load
- Author
-
Arkadi Kreter, Yoshio Ueda, Ryuichi Sakamoto, Y. Hamaji, Marcin Rasinski, Makoto Oya, Akio Sagara, Masayuki Tokitani, Heun Tae Lee, Suguru Masuzaki, Sören Möller, and Kenzo Ibano
- Subjects
Nuclear and High Energy Physics ,High heat flux ,Materials science ,Materials Science (miscellaneous) ,chemistry.chemical_element ,Tungsten ,01 natural sciences ,010305 fluids & plasmas ,Divertor ,0103 physical sciences ,Vertical displacement ,Composite material ,010306 general physics ,Melting ,lcsh:TK9001-9401 ,NRA ,Crystallography ,Grain growth ,Retention ,Nuclear Energy and Engineering ,chemistry ,Heat flux ,Deuterium ,Cathode ray ,lcsh:Nuclear engineering. Atomic power ,ddc:333.7 ,Raster scan - Abstract
Surface morphology and hydrogen isotope retention of W specimen melted with vertical displacement event-like heat load and subsequent deuterium (D) plasma exposure were studied. Applied heat loads using electron beam without raster scanning were about 190 and 230 MW/m2 in heat flux and 0.08, 0.12 and 0.16s in duration. After the heat load application, specimens showed apparent melting spots with grain growth or dense micrometer scale convex structure. Cracks were observed only in the part with the convex structure. D retention in the melted part of specimens was not significantly larger than in the reference specimen despite large changes of surface characteristics. Keywords: Tungsten, Retention, Melting, Divertor, High heat flux, NRA
- Published
- 2017
34. Simulation of impurity transport in the peripheral plasma due to the emission of dust in long pulse discharges on the Large Helical Device
- Author
-
Mamoru Shoji, Yasunori Tanaka, Suguru Masuzaki, Gakushi Kawamura, Roman Smirnov, Yoshihiko Uesugi, and A. Pigarov
- Subjects
Nuclear and High Energy Physics ,education.field_of_study ,Long pulse ,Materials science ,Materials Science (miscellaneous) ,Divertor ,Carbon dust ,Peripheral plasma ,Plasma ,lcsh:TK9001-9401 ,complex mixtures ,01 natural sciences ,respiratory tract diseases ,010305 fluids & plasmas ,Large Helical Device ,Nuclear Energy and Engineering ,Impurity ,0103 physical sciences ,lcsh:Nuclear engineering. Atomic power ,Atomic physics ,010306 general physics ,education ,Helical coil - Abstract
Two different plasma termination processes by dust emission were observed in long pulse discharges in the Large Helical Device. One is a plasma termination caused by large amounts of carbon dust released from a lower divertor region. The other is termination caused by stainless steel (iron) dust emission from the surface of a helical coil can. The effect of the dust emission on the sustainment of the long pulse discharges are investigated using a three-dimensional edge plasma transport code (EMC3-EIRENE) coupled with a dust transport code (DUSTT). The simulation shows that the plasma is more influenced by the iron dust emission from the helical coil can than by the carbon dust emission from the divertor region. The simulation revealed that the plasma flow in divertor legs is quite effective for preventing dust from terminating the long pulse discharges. Keywords: Dust, EMC3-EIRENE, DUSTT, Long pulse discharge
- Published
- 2017
35. Effects of modified surfaces produced at plasma-facing surface on hydrogen release behavior in the LHD
- Author
-
N. Ashikawa, Yasuhisa Oya, Hiroshi Kasahara, Gen Motojima, D. Nagata, Mitsutaka Miyamoto, Yuji Yamauchi, Nobuaki Yoshida, M. Yajima, Noriyasu Ohno, Yuji Nobuta, M. Tokitani, and Suguru Masuzaki
- Subjects
Nuclear and High Energy Physics ,Materials science ,Hydrogen retention ,Desorption ,Redeposition ,Long-term sample ,LHD ,Hydrogen ,Materials Science (miscellaneous) ,Analytical chemistry ,chemistry.chemical_element ,Substrate (electronics) ,01 natural sciences ,010305 fluids & plasmas ,0103 physical sciences ,Boron ,010302 applied physics ,lcsh:TK9001-9401 ,Carbon film ,Nuclear Energy and Engineering ,chemistry ,lcsh:Nuclear engineering. Atomic power ,Carbon ,Deposition (chemistry) ,Layer (electronics) - Abstract
In the present study, an additional deuterium (D) ion irradiation was performed against long-term samples mounted on the helical coil can and in the outer private region in the LHD during the 17th experimental campaign. Based on the release behavior of the D and hydrogen (H) retained during the experimental campaign, the difference of release behavior at the top surface and in bulk of modified surfaces is discussed. Almost all samples on the helical coil can were erosion-dominant and some samples were covered with boron or carbon, while a very thick carbon films were formed in the outer private region. In the erosion-dominant area, the D desorbed at much lower temperatures compared to that of H retained during the LHD plasma operation. For the samples covered with boron, the D tended to desorb at lower temperatures compared to H. For the carbon deposition samples, the D desorbed at much higher temperatures compared to no deposition and boron-covered samples, which was very similar to that of H. The D retention capabilities at the top surface of carbon and boron films were 2–3 times higher than no deposition area. The results indicate that the retention and release behavior at the top surface of the modified layer can be different from that of bulk substrate material. Keywords: Hydrogen retention, Desorption, Redeposition, Long-term sample, LHD
- Published
- 2017
36. Initial growth phase of W-fuzz formation in ultra-long pulse helium discharge in LHD
- Author
-
Masayuki Tokitani, Nobuaki Yoshida, Yoshio Ueda, Takashi Mutoh, Shinji Nagata, Hiroshi Kasahara, Suguru Masuzaki, Yasuo Yoshimura, and Ryuichi Sakamoto
- Subjects
010302 applied physics ,Nuclear and High Energy Physics ,Ion beam analysis ,Materials science ,Nanostructure ,Materials Science (miscellaneous) ,Divertor ,chemistry.chemical_element ,Plasma ,Tungsten ,01 natural sciences ,Fluence ,lcsh:TK9001-9401 ,010305 fluids & plasmas ,Large Helical Device ,Nuclear Energy and Engineering ,chemistry ,0103 physical sciences ,lcsh:Nuclear engineering. Atomic power ,Atomic physics ,Helium - Abstract
In order to confirm the formation of a tungsten fiberform nanostructure (W-fuzz) by helium plasma exposure in the large-sized plasma confinement device, the bulk tungsten with the size of 80 × 20 × 1.5 mm3 was inserted into the divertor leg position in the Large Helical Device (LHD). Then, it was exposed to the divertor plasma during the ultra-long pulse helium discharges with 10,190s in total. The width of the divertor plasma, incident ion energy and total fluence were expected to be ∼2cm, 100–200eV and ∼5 ×1025 He/m2 (strike position), respectively. The surface temperature of the tungsten specimen was monitored by IR camera. The typical surface temperature of the divertor strike point was estimated to be around 1900∼2300K. After the exposure, an initial growth phase of tungsten fiberform nanostructure (W-fuzz) was able to be identified on the tungsten surface. The finest initial growth phase of the W-fuzz structure was able to be identified on the central region of the divertor strike point, where retention amount of helium was estimated to be ∼8 ×1021 He/m2. This study is the first simultaneous evaluation of the W-fuzz growth and quantification of the helium retention in the large-sized plasma confinement device. Keywords: W-fuzz, TEM observation, Ion beam analysis, LHD
- Published
- 2017
37. Conductive sublayer of turbulent heat transfer for heating of water in a circular tube
- Author
-
Katsuya Fukuda, Koichi Hata, and Suguru Masuzaki
- Subjects
Materials science ,Meteorology ,020209 energy ,Flow (psychology) ,Prandtl number ,Thermodynamics ,chemistry.chemical_element ,Liquid Temperature ,02 engineering and technology ,Computational fluid dynamics ,RANS Equation ,01 natural sciences ,010305 fluids & plasmas ,symbols.namesake ,0103 physical sciences ,0202 electrical engineering, electronic engineering, information engineering ,Transfer Coefficient ,Heat Flux Heat ,Electrical conductor ,Fluid Flow and Transfer Processes ,geography ,geography.geographical_feature_category ,Turbulence ,business.industry ,Turbulent Heat Transfer ,Condensed Matter Physics ,Inlet ,Exponential function ,chemistry ,symbols ,Platinum ,business - Abstract
The steady-state and transient turbulent heat transfer coefficients in circular platinum (Pt) test tubes (inner diameters: 3 and 6 mm; heated lengths: 66.5 and 100 mm and 69.6 mm, respectively) were systematically measured using an experimental water loop for a wide range of flow velocities, inlet liquid temperatures, Prandtl numbers, inlet pressures, and exponentially increasing heat inputs (Q 0 exp(t/τ), τ: exponential period). The Reynolds-averaged Navier–Stokes equations and the k–e turbulence model for unsteady turbulent heat transfer in circular test sections were numerically solved for heating of water with heated sections of diameter 3 and 6 mm and length 67 and 100 mm and 70 mm, respectively, by using computational fluid dynamics code under the same conditions as those in the experiment and with temperature-dependent thermophysical fluid properties. The thickness of the conductive sublayer, δ CSL,st and δ CSL [=(Δr) out /2], and the nondimensional thickness of the conductive sublayer, (y , + ) TEM [=(f F /2)0.5 ρ l u δ CSL,st /μ l ] and (y + ) TEM [=(f F /2)0.5 ρ l u δ CSL /μ l ], for steady-state and transient turbulent heat transfer at various heated length-to-inner diameter ratios, inlet liquid temperatures, and exponential periods were measured on the basis of the numerical solutions. The correlations of the thickness of the conductive sublayer, δ CSL,st , and nondimensional thickness of the conductive sublayer, (y , + ) TEM , for steady-state turbulent heat transfer and those of the thickness of the conductive sublayer, δ CSL , and nondimensional thickness of the conductive sublayer, (y + ) TEM , for transient turbulent heat transfer in a circular tube were derived.
- Published
- 2017
38. Conceptual design of a liquid metal limiter/divertor system for the FFHR-d1
- Author
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Hitoshi Tamura, Takuya Goto, Hirohiko Tanaka, Akio Sagara, Takeru Ohgo, Junichi Miyazawa, Juro Yagi, Teruya Tanaka, Suguru Masuzaki, Ryuichi Sakamoto, T. Murase, and Nagato Yanagi
- Subjects
Liquid metal ,Materials science ,Nuclear engineering ,Shields ,chemistry.chemical_element ,Solenoid ,remote maintenance ,01 natural sciences ,010305 fluids & plasmas ,law.invention ,blanket ,law ,Physics::Plasma Physics ,0103 physical sciences ,Limiter ,divertor ,General Materials Science ,010306 general physics ,liquid metal ,heliotron ,Civil and Structural Engineering ,Mechanical Engineering ,Divertor ,Fusion power ,Nuclear Energy and Engineering ,chemistry ,Vacuum pump ,fusion reactor ,Tin - Abstract
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.
- Published
- 2017
39. Design of a Closed Helical Divertor in LHD and the Prospect for Helical Fusion Reactors
- Author
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Lhd experimental groups, Masahiro Kobayashi, Mamoru Shoji, Akio Komori, Hiroshi Yamada, Suguru Masuzaki, and Akio Sagara
- Subjects
Nuclear and High Energy Physics ,Materials science ,020209 energy ,Mechanical Engineering ,Divertor ,Magnetic confinement fusion ,02 engineering and technology ,Plasma ,Mechanics ,Fusion power ,01 natural sciences ,010305 fluids & plasmas ,Magnetic field ,law.invention ,Large Helical Device ,Nuclear Energy and Engineering ,law ,Physics::Plasma Physics ,0103 physical sciences ,0202 electrical engineering, electronic engineering, information engineering ,General Materials Science ,Atomic physics ,Neutral particle ,Stellarator ,Civil and Structural Engineering - Abstract
A new closed helical divertor configuration for efficient particle control and reduction of the heat load on the divertor plates is proposed. The closed divertor configuration practically utilizes an ergodic layer and magnetic field line configuration on divertor legs in helical systems. For optimization of the design of the closed divertor, the distribution of the strike points is calculated in various magnetic configurations in the Large Helical Device (LHD). It suggests that the installation of the closed divertor components in the inboard side of the torus under an inward shift configuration (R ax =3.60m) is the best choice for achieving the above two purposes. This divertor configuration does not interfere with plasma heating and diagnostic systems installed in outer ports. The prospect of the closed divertor configuration to a helical fusion reactor is investigated using a three-dimensional neutral particle transport simulation code with a one-dimensional plasma fluid calculation on the divertor legs. The investigation shows efficient particle pumping from the inboard side and reduction of the heat load due to the combined effect of the optimized closed divertor geometry, ergodized divertor legs, and low electron temperature in the ergodic layer. It indicates a promising closed divertor configuration for helical fusion reactors.
- Published
- 2017
40. Development of H, D, T Simultaneous TDS Measurement System and H, D, T Retention Behavior for DT Gas Exposed Tungsten Installed in LHD Plasma Campaign
- Author
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Shodai Sakurada, Yuki Uemura, Takumi Chikada, Hiroe Fujita, Yasuhisa Oya, Masayuki Tokitani, Yuji Hatano, Cui Hu, Miyuki Yajima, Kenta Yuyama, and Suguru Masuzaki
- Subjects
010302 applied physics ,Nuclear and High Energy Physics ,Materials science ,Hydrogen ,Thermal desorption spectroscopy ,Mechanical Engineering ,Analytical chemistry ,chemistry.chemical_element ,Tungsten ,Mass spectrometry ,01 natural sciences ,Ion source ,010305 fluids & plasmas ,Large Helical Device ,Nuclear Energy and Engineering ,chemistry ,Desorption ,0103 physical sciences ,Ionization chamber ,General Materials Science ,Civil and Structural Engineering - Abstract
All the hydrogen isotope (H, D, T) simultaneous TDS (Thermal desorption spectroscopy) measurement system (HI-TDS system) was newly designed to evaluate all hydrogen isotope desorption behavior in materials. The present HI-TDS system was operated under Ar purge gas and the H and D desorptions were observed by a quadruple mass spectrometer equipped with an enclosed ion source, although T desorption was evaluated by an ionization chamber or proportional counters. Most of the same TDS spectra for D and T were derived by optimizing the heating rate of 0.5 K s−1 with Ar flow rate of 13.3 sccm.Using this HI-TDS system, D and T desorption behaviors for implanted or DT gas exposed tungsten samples installed in LHD (Large Helical Device) at NIFS (National Institute for Fusion Science) was evaluated. It was found that major hydrogen desorption stages consisted of two temperature regions, namely 700 K and 900 K, which was consistent with the previous hydrogen plasma campaign and most of hydrogen would be trap...
- Published
- 2017
41. Temperature impact on the micro structure of tungsten exposed to He irradiation in LHD
- Author
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Naoaki Yoshida, E. Bernard, Hiroshi Yamada, Ryuichi Sakamoto, Hiromi Hayashi, Masayuki Tokitani, and Suguru Masuzaki
- Subjects
010302 applied physics ,Coalescence (physics) ,Nuclear and High Energy Physics ,Materials science ,Annealing (metallurgy) ,chemistry.chemical_element ,Tungsten ,Atmospheric temperature range ,01 natural sciences ,Crystallographic defect ,010305 fluids & plasmas ,Crystallography ,Nuclear Energy and Engineering ,chemistry ,0103 physical sciences ,General Materials Science ,Irradiation ,Surface layer ,Composite material ,Dislocation - Abstract
A new temperature controlled material probe was designed for the exposure of tungsten samples to helium plasma in the LHD. Samples were exposed to estimated fluences of ∼1023 m−2 and temperatures ranging from 65 to 600 °C. Transmission Electron Microscopy analysis allowed the study of the impact of He irradiation under high temperatures on tungsten micro structure for the first time in real-plasma exposure conditions. Both dislocation loops and bubbles appeared from low to medium temperatures and saw an impressive increase of size (factor 4 to 6) most probably by coalescence as the temperature reaches 600 °C, with 500 °C appearing as a threshold for bubble growth. Annealing of the samples up to 800 C highlighted the stability of the dislocation damages formed by helium irradiation at high surface temperature, as bubbles and dislocation loops seem to conserve their characteristics. Additional studies on cross-sections showed that bubbles were formed much deeper (70–100 nm) than the heavily damaged surface layer (10–20 nm), raising concern about the impact on the material mechanical properties conservation and potential additional trapping of hydrogen isotopes.
- Published
- 2017
42. Influence of carbon-dominated deposition layer on He retention and desorption in tungsten
- Author
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Naoaki Yoshida, Shodai Sakurada, Takumi Chikada, Masayuki Tokitani, Keisuke Azuma, Yuji Uemura, Miyuki Yajima, Cui Hu, Yasuhisa Oya, Hiroe Fujita, and Suguru Masuzaki
- Subjects
010302 applied physics ,Materials science ,Thermal desorption spectroscopy ,Mechanical Engineering ,Layer by layer ,Analytical chemistry ,chemistry.chemical_element ,Tungsten ,01 natural sciences ,010305 fluids & plasmas ,Nuclear Energy and Engineering ,chemistry ,Impurity ,Desorption ,0103 physical sciences ,General Materials Science ,Layer (electronics) ,Helium ,Deposition (law) ,Civil and Structural Engineering - Abstract
Pure tungsten (W) samples were respectively installed on the two positions of typical plasma wall interaction area (PI) and erosion dominated position (ER) on the first wall in the Large Helical Device (LHD) at National Institute for Fusion Science (NIFS), Japan. After the experiment campaign in 2014, these samples were picked up and thermal desorption spectroscopy (TDS) were applied to evaluate their desorption behavior of He which was implanted by He discharge. It was found that a carbon-dominated mixed-material layer with impurities, such as Fe, Cr, Mo, O and N, etc., was deposited layer by layer during the plasma exposure on the PI sample. On the other hand, W-C mixed layer was formed on the ER sample. The results showed that a large amount of He was trapped in the samples on both PI and ER, and the total He retention for ER is about twice as large as that of PI. He was trapped in various types of trapping sites in the ER sample and their desorption peaks were located at temperatures of about 425 K, 755 K, 1130 K and 1630 K. For PI sample, most of He was trapped in the carbon-dominated mixed-material layer and the corresponding desorption temperature was limited to be about 600 K, 900 K and 1200 K. The additional 3.0 keV helium ion (He+) implantation was performed for several samples to investigate the He retention characteristics in these samples and it was found that no additional desorption stage was found. These results suggested that the He discharge history and its deposition on the W plasma facing wall would affect He desorption behavior of W.
- Published
- 2016
43. The isotope effect on impurities and bulk ion particle transport in the Large Helical Device
- Author
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Gen Motojima, Hiroto Takahashi, Motoki Nakata, Masayuki Yokoyama, K. Yamazaki, T. Oishi, Ryuichi Sakamoto, M. Emoto, Ryosuke Seki, Naoki Tamura, Suguru Masuzaki, R. Mackenbach, Kenichi Nagaoka, Izumi Murakami, Kiyofumi Mukai, Masaki Osakabe, Mikirou Yoshinuma, Junwei Chen, Chihiro Suzuki, S. Morita, Katsumi Ida, Haruhisa Nakano, Sadayoshi Murakami, Tomohiro Morisaki, Shuji Kamio, Masanori Nunami, Y. Fujiwara, Motoshi Goto, T. Kobayashi, K. Fuji, Hiroshi Yamada, Science and Technology of Nuclear Fusion, and Turbulence in Fusion Plasmas
- Subjects
Nuclear and High Energy Physics ,Materials science ,Hydrogen ,Analytical chemistry ,chemistry.chemical_element ,Plasma ,Condensed Matter Physics ,7. Clean energy ,01 natural sciences ,010305 fluids & plasmas ,Ion ,Large Helical Device ,impurity transport ,Deuterium ,chemistry ,13. Climate action ,Impurity ,Physics::Plasma Physics ,0103 physical sciences ,Kinetic isotope effect ,ion particle transport ,large helical device ,010306 general physics ,Ion transporter ,isotope effect - Abstract
The isotope effect on impurities and bulk ion particle transport is investigated by using the deuterium, hydrogen, and isotope mixture plasma in the Large Helical Device (LHD). A clear isotope effect is observed in the impurity transport but not the bulk ion transport. The isotope effects on impurity transport and ion heat transport are observed as a primary and a secondary effect, respectively, in the plasma with an internal transport barrier (ITB). In the LHD, an ion ITB is always transient because the impurity hole triggered by the increase of ion temperature gradient causes the enhancement of ion heat transport and gradually terminates the ion ITB. The formation of an impurity hole becomes slower in the deuterium (D) plasma than the hydrogen (H) plasma. This primary isotope effect on impurity transport contributes the longer sustainment of the ion ITB state because the low ion thermal diffusivity can be sustained as long as the normalized carbon impurity gradient R/L n,c, where L c =-(∇n c/n c) -1 is above the critical value (∼-5). Therefore, the longer sustainment of the ITB state in the deuterium plasma is considered to be a secondary isotope effect due to the mitigation of the impurity hole. The radial profile of H and D ion density is measured using bulk charge exchange spectroscopy inside the isotope mixture plasma. The decay time of H ion density after the H-pellet injection and the decay time of D ion density after D-pellet injection are almost identical, which demonstrates that there is no significant isotope effect on ion particle transport.
- Published
- 2019
44. Synergistic effect of nitrogen and hydrogen seeding gases on plasma detachment in the GAMMA 10/PDX tandem mirror
- Author
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Md. Shahinul Islam, Sotaro Yamashita, Yousuke Nakashima, Mizuki Sakamoto, Makoto Ichimura, Ryuya Ikezoe, Naomichi Ezumi, Kazuya Ichimura, Toru Iijima, R. Perillo, Akiyoshi Hatayama, Yosuke Kinoshita, Ryutaro Minami, Noriyasu Ohno, Hirohiko Tanaka, S. Sawada, Satoshi Togo, Shinichiro Kado, A. Terakado, Masayuki Yoshikawa, Tomohiro Mikami, Mafumi Hirata, S. Jang, M. M. Islam, Junko Kohagura, Toshiki Hara, Akira Tonegawa, Tsuyoshi Kariya, Suguru Masuzaki, Tomoharu Numakura, Isao Katanuma, Tsuyoshi Imai, Tsubasa Yoshimoto, Kunpei Nojiri, T. Nakano, and Science and Technology of Nuclear Fusion
- Subjects
Nuclear and High Energy Physics ,Electron density ,Tokamak ,Materials science ,Density gradient ,Hydrogen ,Divertor ,chemistry.chemical_element ,plasma detachment ,Plasma ,Condensed Matter Physics ,01 natural sciences ,Molecular physics ,nitrogen ,010305 fluids & plasmas ,Ion ,law.invention ,chemistry ,Impurity ,law ,hydrogen ,0103 physical sciences ,divertor ,N-MAR ,010306 general physics - Abstract
We have investigated the synergistic effect of a combination of various impurity gases and hydrogen gas on plasma detachment of high temperature plasma, equivalent to scrape-off layer plasma of tokamaks in the GAMMA 10/PDX end region, utilizing an open magnetic field configuration. A small puff of an impurity gas (N2, Ne, Ar, Kr, Xe) in combination with a puff of H2 gas is examined to evaluate their synergistic effect on the formation of detached plasma; the following results are obtained. (i) A combination of N2 and H2 puffs showed a clear decrease of electron density and ion flux; (ii) N2 and H2 puffs form a strong density gradient along the axial direction; and (iii) other noble impurity gases showed an insufficient synergistic effect. The new results indicate the possibility of achieving a reliable divertor operation scheme and the importance of a deeper understanding of the H2 and N2 assisted recombination process.
- Published
- 2019
45. Influence of thermal shocks on the He induced surface morphology on tungsten
- Author
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Y. Hamaji, Ryuichi Sakamoto, Akio Sagara, Suguru Masuzaki, Hitoshi Tamura, Arkadi Kreter, and Masayuki Tokitani
- Subjects
010302 applied physics ,Nuclear and High Energy Physics ,Materials science ,Materials Science (miscellaneous) ,Pulse duration ,chemistry.chemical_element ,Flux ,Plasma ,Tungsten ,lcsh:TK9001-9401 ,01 natural sciences ,Fluence ,Molecular physics ,010305 fluids & plasmas ,Ion ,Nuclear Energy and Engineering ,Heat flux ,chemistry ,0103 physical sciences ,lcsh:Nuclear engineering. Atomic power ,Irradiation ,ddc:624 - Abstract
In this study, the effect of ELM-like thermal shocks on He induced surface morphology were investigated. A W sample was exposed to pure He plasma. He ion incident energy, flux and fluence were 80 eV, 1 × 1022 /m2/s and 3 × 1025 /m2, respectively. Irradiation temperature were approximately 470 and 1100 K. Then, thermal shocks were applied on the sample using focused electron beam. The peak heat flux, pulse duration and base temperature were 500 MW/m2, 500 μs and R.T., respectively. After the thermal shocks, He induced morphologies such as holes and periodic undulations were flatten completely at the region exposed to the highest heat flux. At the peripheral regions of the electron beam spot, the hole density increase or partial flattening of morphologies were observed. These results suggested that in order to anticipate surface morphology with He irradiation and ELMs, the peak temperature should play a more important role than base temperature. Keywords: Tungsten, Divertor, He induced surface morphology, Thermal shocks
- Published
- 2019
46. Erosion and deposition investigations on Wendelstein 7-X first wall components for the first operation phase in divertor configuration
- Author
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Thomas Sunn Pedersen, Suguru Masuzaki, V. V. Burwitz, Dag Hathiramani, G. Ehrke, Dirk Naujoks, M. Rack, Cristian Ruset, D. Höschen, Thomas Schwarz-Selinger, M. Mayer, Miyuki Yajima, Gen Motojima, Ralf König, Chandra Prakash Dhard, Rudolf Neu, M. Balden, Olaf Neubauer, Marcin Rasinski, Christian Linsmeier, S. Brezinsek, M. Krause, Cong Li, J. Oelmann, J. W. Coenen, Masayuki Tokitani, and W7-X Team, Max Planck Institute for Plasma Physics, Max Planck Society
- Subjects
Toroidal and poloidal ,Materials science ,Mechanical Engineering ,Nuclear engineering ,Divertor ,Plasma ,01 natural sciences ,010305 fluids & plasmas ,law.invention ,Magnetic field ,Nuclear Energy and Engineering ,law ,0103 physical sciences ,Erosion ,Deposition (phase transition) ,General Materials Science ,ddc:530 ,Wendelstein 7-X ,010306 general physics ,Stellarator ,Civil and Structural Engineering - Abstract
In the stellarator Wendelstein 7-X with its twisted 3D magnetic field geometry, studies of material migration with respect to first wall components become very important in view of the envisioned long-pulse operation. A variety of erosion/deposition samples were installed on the plasma-facing components exposed at three different nominal heat load levels between 0.1 and 10 MW/m2. After the first successful operation phase in divertor configuration, all the probes at higher and lower load levels were removed, whereas at the intermediate load levels, 352 out of 30 000 screws have been exchanged at selected locations along the toroidal and poloidal directions. The exchanged probes have been analyzed by various measurement techniques. At the higher load levels where the probes were installed within the divertor, heavy erosion has been observed presumably at the strike line positions. Both, erosion and deposition phenomena have been found on the screw heads. The optical reflection measurement profile of the whole plasma vessel show the deposition patterns at similar locations in all the five modules. At the low load level, the Si-wafer probes are under investigation.
- Published
- 2019
47. Plasma-wall interaction on the divertor tiles of JET ITER-like wall from the viewpoint of micro/nanoscopic observations
- Author
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Alexander Lukin, Stefan Matejcik, Ryuichi Sakamoto, Soare Sorin, Francesco Romanelli, Kalle Heinola, Masayuki TOKITANI, Anna Widdowson, Bohdan Bieg, Suguru Masuzaki, Vladislav Plyusnin, José Vicente, Alberto Loarte, Axel Jardin, Rajnikant Makwana, CHIARA MARCHETTO, Marco Wischmeier, Yuji Hatano, William Tang, Choong-Seock Chang, Manuel Garcia-munoz, Department of Physics, and Materials Physics
- Subjects
010302 applied physics ,Jet (fluid) ,Materials science ,Tokamak ,EROSION ,Mechanical Engineering ,Divertor ,JET-ILW ,Plasma ,Mechanics ,01 natural sciences ,114 Physical sciences ,010305 fluids & plasmas ,law.invention ,Fuel inventory ,Nuclear Energy and Engineering ,law ,0103 physical sciences ,General Materials Science ,Nanoscopic scale ,Erosion-deposition ,Civil and Structural Engineering - Abstract
Micro/nanoscopic observations on the surface of the divertor tiles used in the first campaign (2011-2012) of the JET tokamak with ITER-like Wall (JET ILW) have been carried out by means of several material analysis techniques. Previous results from the inner divertor were reported for a single poloidal section of the tile numbers 1, 3 and 4, i.e., upper, vertical and horizontal targets, respectively. The formation of the thick stratified mixed-material deposition layer on tiles 1 and 4, and erosion on tile 3 were identified. This study is mostly focused on the outer divertor: tiles 6, 7 and 8. In contrast to the inner tile, remarkable surface modifications have not been observed on the vertical target (tiles 7 and 8) where sputtering erosion and impurity deposition would have been almost balanced. Only a specific part of tile 6 (horizontal target) located near the exhaust channel was covered with a stratified ("geological-like") mixed-material deposition layer which mainly included Be and Ni with the thickness of similar to 2 mu m. Special feature of this mixed layer was that a certain amount of nitrogen (N) was clearly detected in the layer. Since the concentration of N varied with the depth position, it could be depended on the amount of that gas puffed for plasma edge cooling during the JET experimental campaign. In addition to the outer divertor tiles, a very interesting feature of the local erosion and deposition effects is reported in this paper.
- Published
- 2018
48. EUV and VUV Spectra of NeIII-NeX Line Emissions Observed in the Impurity Gas-Puffing Experiments of the Large Helical Device
- Author
-
Gakushi Kawamura, Masahiro Kobayashi, Shigeru Morita, Yuki Hayashi, Yasuko Kawamoto, Motoshi Goto, Tetsutarou Oishi, Suguru Masuzaki, Kiyofumi Mukai, and Chihiro Suzuki
- Subjects
Materials science ,Extreme ultraviolet lithography ,extreme ultraviolet ,divertor detachment ,Condensed Matter Physics ,impurity seeding ,Spectral line ,Large Helical Device ,Impurity ,vacuum ultraviolet ,Atomic physics ,magnetically confined fusion ,Line (formation) ,plasma spectroscopy - Abstract
Extreme ultraviolet (EUV) and vacuum ultraviolet (VUV) wavelength spectra including line emissions released from neon (Ne) ions ranging from low to high charge states observed simultaneously in a single discharge are summarized for contribution to compile a fundamental spectral dataset for the Ne-seeded divertor heat load reduction experiments in Large Helical Device (LHD). NeIX and NeX lines were observed in the EUV wavelength range of 10∼50 Å and NeIII-NeVIII lines were observed in the VUV wavelength range of 400∼1000 Å. The temporal evolutions of the line intensities exhibited different behaviors between the edge emissions of NeIII-NeVIII with the ionization potential, Ei, of 63∼239 eV and the core emission of NeX with Ei of 1362 eV. NeIX with Ei of 1196 eV exhibited a marginal behavior between the edge emission and the core emission.
- Published
- 2021
49. Application of Divertor Pumping to Long-Pulse Discharge for Particle Control in LHD
- Author
-
Yasuhiko Takeiri, T. Murase, Tomohiro Morisaki, Ryuichi Sakamoto, Gen Motojima, Masahiro Kobayashi, Suguru Masuzaki, and Yasuyuki Tsuchibushi
- Subjects
Materials science ,Long pulse ,Divertor pump ,Nuclear engineering ,Divertor ,density control ,LHD ,Condensed Matter Physics ,Particle control - Abstract
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.
- Published
- 2021
50. Boron transport simulation using the ERO2.0 code for real-time wall conditioning in the large helical device
- Author
-
Gakushi Kawamura, Mamoru Shoji, S. Brezinsek, A. Kirschner, A. Eksaeva, D. Borodin, Suguru Masuzaki, and Juri Romazanov
- Subjects
Nuclear and High Energy Physics ,EMC3-EIRENE ,Materials science ,Toroidal and poloidal ,Materials Science (miscellaneous) ,ERO2.0 ,chemistry.chemical_element ,Large Helical Device ,Physics::Plasma Physics ,Impurity ,Condensed Matter::Superconductivity ,Boron ,Plasma density ,Divertor ,Plasma ,Mechanics ,DUSTT ,lcsh:TK9001-9401 ,Boron transport ,Nuclear Energy and Engineering ,chemistry ,Impurity powder dropper ,lcsh:Nuclear engineering. Atomic power ,LHD ,Boronization ,ddc:624 - Abstract
The three-dimensional Monte-Carlo impurity transport and plasma surface interaction code ERO2.0 is applied to a full-torus model for the Large Helical Device (LHD). In order to find an optimum experimental condition for effective real-time wall conditioning (boronization) using an Impurity Powder Dropper (IPD), the toroidal and poloidal distribution of the boron flux density on the divertor components and the vacuum vessel are surveyed in various experimental conditions. The source profile of the neutral boron atoms originated from boron powders supplied from the IPD is calculated using the DUSTT code in background plasmas provided by the EMC3-EIRENE code. The simulations using ERO2.0 predict that higher plasma density operation is inappropriate for the effective wall conditioning because of the toroidally localized boron flux density in a closed helical divertor region. The ERO2.0 simulations have successfully revealed an optimum experimental condition for the wall conditioning with the toroidally uniform boron flux density in the closed helical divertor region.
- Published
- 2020
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