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ABSTRACT: Recent results of R&D on plasma facing components for fusion
experimental reactors such as the ITER/FER (International Thermonuclear
Experimental Reactor/Fusion Experimental Reactor) are presented. Plasma
facing components will be exposed to severe heat loads. Thermal cycling
tests for first wall mock-ups and divertor mock-ups have been carried
out under specified heat flux conditions of ITER/FER in high heat flux
test facilities at JAERI. Tests at a stationary heat flux of 0.2 MW/m
<sup>2</sup> for first wall mock-ups, exposed in the normal heat flux
part, have confirmed their durabilities and reliabilities against the
cyclic heat loads. In experiments at a stationary heat flux of 0.6 MW/m
<sup>2</sup>, with exposure also in the high heat flux part, only the
radiatively cooled first wall mock-up was tested. A ceramic sleeve
covered with pure titanium between the armor and the cooling structure
was damaged before reaching a stationary heat flux of 0.6 MW/m<sup>2
</sup>. The authors have also confirmed the integrity and the durability
of the bonds in the thermal cycling test of brazed
carbon-fiber-composite/copper divertor mock-ups at a simulated
stationary surface heat flux of 10 MW/m<sup>2</sup> for 1000
cycles
Fusion Engineering, 1991. Proceedings., 14th IEEE/NPSS Symposium on; 11/1991
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M. Kuriyama,
M. Akiba,
N. Akino,
M. Dairaku,
N. Ebisawa,
K. Hiruta,
M. Kawai,
S. Kitamura,
K. Kikuchi,
M. Komata, [......],
T. Ohga,
Y. Ohara,
H. Oohara,
Y. Okumura,
M. Seki,
K. Shibanuma,
T. Sugawara, S. Tanaka,
H. Usami,
K. Usui
[show abstract]
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ABSTRACT: The JT-60 JAERI (Japan Atomic Energy Research Institute)
Tokamak-60 neutral beam injector (NBI) has been successfully operated
for three years, delivering a neutral beam power of 20 MW at a beam
energy of 75 keV. The helium-beam injection into JT-60 was performed to
investigate exhaustion of fusion alpha particles from the plasma. The
helium-ion beam current extracted from ion sources was 40-45 A at an
energy of 31 keV. The injection power was about 0.4 MW. In the
helium-beam operation, the helium gas in the beamline chamber was pumped
out by a condensed SF<sub>6</sub> gas cryosorption pump. The JT-60
neutral beam injector will be modified to inject a 120-keV deuterium
beam into the JT-60 upgrade, which will start to operate at the
beginning of 1991
Fusion Engineering, 1989. Proceedings., IEEE Thirteenth Symposium on; 11/1989
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Y. Ohara,
M. Akiba,
M. Araki,
M. Hanada,
T. Inoue,
H. Kojima,
M. Kuriyama,
M. Matsuoka,
M. Mizuno,
Y. Matsuda,
Y. Okumura,
M. Seki, S. Tanaka,
K. Watanabe
[show abstract]
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ABSTRACT: Recent results from the negative-ion source experiments and design
works of negative-ion-based NBI (neutral beam injection) systems for the
JAERI (Japan Energy Research Institute) tokamak (JT-60U) and the FER
(Fusion Experimental Reactor) are presented together with a description
of the long-range R&D program. A 50-keV, 7.8-A negative-hydrogen-ion
beam has been produced successfully for a duration of 100 ms using a
cesium-supplied volume source which has 253 extraction apertures of 11.3
mm in diameter. The maximum ion-beam current density is 30 mA/cm<sup>2
</sup> at the source pressure of 1.3 Pa. On the basis of recent
promising results from the negative-ion sources, injection of a 500-keV,
10-MW deuterium beam into the JT-60U plasmas in 1994 to heat the core
plasma and to drive the plasma current is proposed
Fusion Engineering, 1989. Proceedings., IEEE Thirteenth Symposium on; 11/1989
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M. Akiba,
M. Araki,
M. Dairaku,
K. Fukaya,
T. Horie,
K. Iida,
H. Ise,
M. Mizuno,
M. Ogawa,
Y. Ohara,
Y. Okumura,
M. Seki,
H. Takatsu, S. Tanaka,
K. Watanabe,
K. Yokoyama
[show abstract]
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ABSTRACT: Recent R&D results on high-heat flux components are presented,
including construction of a new test stand. The test stand can extract
an electron beam of 4.1 at 100 keV. E-folding divergence of the beam is
1.7 mrad, and the latest beam performance is also described. At the
original test stand, which can produce hydrogen-ion beams of 50 A at 100
keV for 10 s, high- Z divertor armors, were tested. Tungsten
plates brazed on copper blocks have been proven to have enough
durability against heat flux under 10 MW/m<sup>2</sup>.
Carbon-fiber-carbon (CFC) composites were tested at the new
electron-beam test stand and an electron-beam welding machine. Under
disruption-simulation conditions, evaporation weight loss of CFC was
lower than that of isotropic graphite
Fusion Engineering, 1989. Proceedings., IEEE Thirteenth Symposium on; 11/1989
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[show abstract]
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ABSTRACT: A design of a high-voltage DC power supply for future neutral beam
injectors (NBI) is presented together with the latest R&D results
for a 100-kV, 5-A, 5-kHz inverter-type power supply. A negative-ion
based NBI is used for heating and current drive in the JT-60U (JAERI
Tokamak-60 upgrade) and the FER (Fusion Experimental Reactor). For a
high-voltage DC power supply of 500-1000 kV, an inverter-type DC power
supply has been adopted. It consists of low-voltage converters,
inverters, step-up transformers, and high-voltage rectifiers. Ion
sources are protected from electrical breakdowns by stopping the
inverters. As the first step, the same system was used in the 100-kV,
5-A acceleration power supply of the JAERI (Japan Atomic Energy Research
Institute) electron-beam irradiation stand. The power supply is based on
5-kHz inverters. After the dummy load test, the power supply operated in
conjunction with a plasma electron gun, and a 100-keV, 4-A electron beam
has already been extracted
Fusion Engineering, 1989. Proceedings., IEEE Thirteenth Symposium on; 11/1989
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E. Tada,
K. Shibanuma,
M. Sugihara,
K. Yoshida,
H. Tsuji,
Y. Shimomura,
S. Matsuda,
T. Abe,
N. Fujisawa,
M. Hasegawa, [......],
K. Sato,
S. Seki,
Y. Seki,
Y. Shinya,
H. Takatsu, S. Tanaka,
T. Takizuka,
T. Tsunematsu,
S. Yamamoto,
H. Yoshida
[show abstract]
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ABSTRACT: The Fusion Experimental Reactor (FER) is a D-T-burning tokamak
machine currently being designed. It is expected to provide physical
information and technical experiences that will be sufficient to proceed
towards the DEMO Fusion Reactor which will demonstrate electric power
generation by fusion energy. An efficient ash exhaust, a hybrid current
drive operation, the use of a 3% ripple field, the technological
achievements in R&D of the magnets, and the negative-ion beam system
are expected to allow the FER to achieve its cost-effectiveness
Fusion Engineering, 1989. Proceedings., IEEE Thirteenth Symposium on; 11/1989
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K. Shibanuma,
S. Matsuda,
H. Tsuji,
H. Kimura,
Y. Ohara,
Y. Seki,
E. Tada,
H. Takatsu, S. Tanaka,
H. Yoshida,
K. Yoshida
[show abstract]
[hide abstract]
ABSTRACT: The time has come for fusion experimental reactor programmes (ITER/FER) to enter the engineering design phase, where the large scale models which are capable of extrapolation to the construction of reactors will be developed as the core of the main activities. A brief review of the present status of the required R&D for experimental reactors, and the technological realization perspectives, are described.
Fusion Engineering and Design.
