Spin resonance in the new-structure-type iron-based superconductor CaKFeAs

Iida, Kazuki*; Ishikado, Motoyuki*; Nagai, Yuki; Yoshida, Hiroyuki*; Christianson, A. D.*; Murai, Naoki; Kawashima, Kenji*; Yoshida, Yoshiyuki*; Eisaki, Hiroshi*; Iyo, Akira*

Journal of the Physical Society of Japan, 86(9), p.093703_1 - 093703_4, 2017/09

https://doi.org/10.7566/JPSJ.86.093703

Conceptual design of the SlimCS fusion DEMO reactor

Tobita, Kenji; Nishio, Satoshi*; Enoeda, Mikio; Nakamura, Hirofumi; Hayashi, Takumi; Asakura, Nobuyuki; Uto, Hiroyasu; Tanigawa, Hiroyasu; Nishitani, Takeo; Isono, Takaaki; et al.

JAEA-Research 2010-019, 194 Pages, 2010/08

JAEA-Research-2010-019-01.pdf:48.47MB
JAEA-Research-2010-019-02.pdf:19.4MB

This report describes the results of the conceptual design study of the SlimCS fusion DEMO reactor aiming at demonstrating fusion power production in a plant scale and allowing to assess the economic prospects of a fusion power plant. The design study has focused on a compact and low aspect ratio tokamak reactor concept with a reduced-sized central solenoid, which is novel compared with previous tokamak reactor concept such as SSTR (Steady State Tokamak Reactor). The reactor has the main parameters of a major radius of 5.5 m, aspect ratio of 2.6, elongation of 2.0, normalized beta of 4.3, fusion out put of 2.95 GW and average neutron wall load of 3 MW/m. This report covers various aspects of design study including systemic design, physics design, torus configuration, blanket, superconducting magnet, maintenance and building, which were carried out increase the engineering feasibility of the concept.

Compact DEMO, SlimCS; Design progress and issues

Tobita, Kenji; Nishio, Satoshi; Enoeda, Mikio; Kawashima, Hisato; Kurita, Genichi; Tanigawa, Hiroyasu; Nakamura, Hirofumi; Honda, Mitsuru; Saito, Ai*; Sato, Satoshi; et al.

Nuclear Fusion, 49(7), p.075029_1 - 075029_10, 2009/07

https://doi.org/10.1088/0029-5515/49/7/075029

Recent design study on SlimCS focused mainly on the torus configuration including blanket, divertor, materials and maintenance scheme. For vertical stability of elongated plasma and high beta access, a sector-wide conducting shell is arranged in between replaceable and permanent blanket. The reactor adopts pressurized-water-cooled solid breeding blanket. Compared with the previous advanced concept with supercritical water, the design options satisfying tritium self-sufficiency are relatively scarce. Considered divertor technology and materials, an allowable heat load to the divertor plate should be 8 MW/m or lower, which can be a critical constraint for determining a handling power of DEMO (a combination of alpha heating power and external input power for current drive).

Simulation study for divertor design to handle huge exhaust power in the SlimCS DEMO reactor

Kawashima, Hisato; Shimizu, Katsuhiro; Takizuka, Tomonori; Tobita, Kenji; Nishio, Satoshi; Sakurai, Shinji; Takenaga, Hidenobu

Nuclear Fusion, 49(6), p.065007_1 - 065007_7, 2009/06

https://doi.org/10.1088/0029-5515/49/6/065007

By the SOLDOR/NEUT2D simulation for divertor design study on DEMO reactor, SlimCS, we estimated a prospect of handling huge exhausted power. Assuming the exhausted power 500 MW and ion out flux 0.510 s into the scrape-off-layer, the peak heat load is estimated to be 70 MW/m on the outer target on the initial divertor design with introduction of moderate gas puff and Ar fraction. This value exceeds the allowable level 10 MW/m being an initial design target. By installing the "" in bottom of the outer divertor target, and using strong gas puffing or Ar impurity injection, the detached condition with high particle recycling and radiation loss conditions is formed, and the peak heat load is successfully reduced below 10 MW/m. It can also be demonstrated that the peak heat load is reduced exponentially with decrease of the exhaust power and reaches to 7 MW/m at = 300 MW for moderate gas puff flux and Ar fraction.

Momentum-resolved charge excitations in high- cuprates studied by resonant inelastic X-ray scattering

Ishii, Kenji; Hoesch, M.*; Inami, Toshiya; Kuzushita, Kaori*; Owada, Kenji; Tsubota, Masami; Murakami, Yoichi; Mizuki, Junichiro; Endo, Yasuo; Tsutsui, Kenji*; et al.

Journal of Physics and Chemistry of Solids, 69(12), p.3118 - 3124, 2008/12

https://doi.org/10.1016/j.jpcs.2008.06.041

Blanket-plasma interaction in tokamaks; Implication from JT-60U, JFT-2M and reactor studies

Kikuchi, Mitsuru; Nishio, Satoshi; Kurita, Genichi; Tsuzuki, Kazuhiro; Bakhtiari, M.*; Kawashima, Hisato; Takenaga, Hidenobu; Kusama, Yoshinori; Tobita, Kenji

Fusion Engineering and Design, 81(8-14), p.1589 - 1598, 2006/02

https://doi.org/10.1016/j.fusengdes.2005.09.048

Blanket-plasma interaction is important for plasma performance enhancement and reliability of first-wall/ blanket. Typical examples are harminization of wall stabilization and reduction of EM force during current quench, error field effect by ferritic steel, neutral-wall interaction under wall saturation, etc. JAERI reactor studies, JT-60U and JFT-2M results on this topics will be described.

Study on Mixing Phenomena in T-Pipe Junction; Temperature Measurement Test in Pipe by Liquid Crystal Sheet

Tanaka, Masaaki; Kawashima, Shigeyo*; Igarashi, Minoru; Hayashi, Kenji; Tobita, Akira; Kamide, Hideki

JNC TN9400 2003-117, 65 Pages, 2004/03

JNC-TN9400-2003-117.pdf:3.49MB

Temperature fluctuation due to mixing of hot and cold fluids gives thermal fatigue to the structure (thermal striping phenomena).Investigation of this phenomenon is significant for the safety of a fast breeder reactor, which uses liquid metal as a coolant. In Japan Nuclear Cycle Development Institute, experiments and numerical analyses have been carried out to understand this phenomenon and also to construct the evaluation rule, which can be applied to the design. A water experiment of fluid mixing in T-pipe with long cycle fluctuation (WATLON),which notices thermal striping phenomena in the T-pipe junction, is performed to investigate the key factor of mixing phenomena by reason of long cycle fluctuation observed in a plant. By the former visualization test, it was showed that the flow pattern of branch pipe jet could be classified into (A) impinging jet, (B) deflecting jet (C) re-attachment jet and (D) wall jet according to the inflow condition. It was confirmed that the each jet pattern could be predicted by the momentum ratio of the each piping fluid. In this study, a thermo-chromic liquid crystal sheet was put on the inner wall in the main pipe, and temperature field on the wall surface was visualized. We established a new method to convert the color image data to temperature data. And measurement uncertainty of this method was evaluated + and - about 2.0 [deg-C], using by the typical picture in the temperature calibration test. From the temperature fluctuation visualization test by liquid crystal sheet, the cold spot was formed in just downstream region from the outlet of the branch pipe in the cases of the wall jet and impinging jet. Since this cold spot moved in time, high value of temperature fluctuation intensity was shown around the cold spot. And the validity of this method was shown from the comparison of the thermocouple data installed in a wall surface with the temperature conversion result.

Effects of complex magnetic ripple on fast ions in JFT-2M ferritic insert experiments

Shinohara, Koji; Kawashima, Hisato; Tsuzuki, Kazuhiro; Urata, Kazuhiro*; Sato, Masayasu; Ogawa, Hiroaki; Kamiya, Kensaku; Sasao, Hajime; Kimura, Haruyuki; Kasai, Satoshi; et al.

Nuclear Fusion, 43(7), p.586 - 593, 2003/07

https://doi.org/10.1088/0029-5515/43/7/312

In JFT-2M, the ferritic steel was installed inside the vacuum vessel as the third step of tne Advanced Material Tokamak Experiment (AMTEX) programme. The magnetic field structure has become the complex ripple strucuture such as non-periodic ripple structure in the toroidal direction and high Fourier-number ripple because of the existence of the port and the limitation of the periodic installation. Under such a complex ripple structure, we have performed the experiment to understand its effect on fast ions. To actively change the ripple structure, we have installed the additional ferritic steel plates (FPs). We also have compared the experiment result with the newly updated OFMC code which does not need the toroidal 16-folded symmetry. The experiment results were almost consistent with the OFMC calcution with the complex magnetic field and the complex first wall.

Progress in design study of low aspect ratio DEMO, "SlimCS"

Tobita, Kenji; Nishio, Satoshi; Yamada, Masao*; Kakudate, Satoshi; Nakamura, Hirofumi; Hayashi, Takumi; Enoeda, Mikio; Tsuru, Daigo; Kawashima, Hisato; Kurita, Genichi; et al.

no journal, , 

Maintenance is one of critical issues that determines the torus configuration of a fusion reactor. For high availability of operation, in-vessel components should be replaceable easily. On the other hand, the in-vessel components must be supported robustly enough to withstand the electromagnetic forces acting on a disruption. In addition, the fraction of the structural material of blanket should be reduced to keep the breeding region high as possible. In order to meet these requirements, JAEA has attempted to find a sector transport maintenance scheme based on the existing technologies in industries. Possible ideas on transportation of the sector, cask and support structure of toroidal field coils for the sector transport maintenance will be presented.

Study of optimized fuelling scenario and particle balance in a steady-state fusion reactor

Takenaga, Hidenobu; Kawashima, Hisato; Tobita, Kenji; Ichige, Hisashi; Miyo, Yasuhiko; Nishiyama, Tomokazu

no journal, , 

A fuelling scenario in a fusion reactor has been investigated, where tritium is fuelled in the main plasma and deuterium is fuelled in both the main plasma and the edge plasma. The tritium fuelling in the main plasma minimizes the tritium fuelling rate necessary for sustaining the high tritium density in the main plasma. The deuterium fuelling in the main plasma sustains the high deuterium density in the main plasma and the deuterium fuelling in the edge plasma enhances the deuterium recycling level for reducing the divertor temperature. Based on this scenario, particle balance was quantitatively investigated using the SlimCS design parameters at 2.95 GW fusion output. The fuelling rates in the main plasma were evaluated to be 2.510/s for tritium and 1.410/s for deuterium when the confinement times for the particles fuelled in the main and edge plasmas were assumed to be 2 s and 2 ms, respectively, and the divertor pumping rate was assumed to be 3% of the particle flux to the divertor plates. For enhancement of the recycling level, the additional deuterium fuelling in the edge plasma of 3.610/s was required in this case. Dependence of fuelling rates on particle confinement times and acceptable tritium retention will be reported together with development of new fuelling systems.