Critical issues for the manufacture of the ITER TF coil winding pack

Koizumi, Norikiyo; Hemmi, Tsutomu; Matsui, Kunihiro; Nakajima, Hideo; Okuno, Kiyoshi; Kuno, Kazuo*; Nomoto, Kazuhiro*

Fusion Engineering and Design, 84(2-6), p.210 - 214, 2009/09

https://doi.org/10.1016/j.fusengdes.2008.12.105

ITER-TF coil, whose height and width are 14 m and 9 m, respectively, is scaled up by about 3 times from TF model coil (TFMC), which was developed in ITER EDA and successfully tested in EU. Although major technique of TF coil fabrication has been demonstrated in the TFMC development, new technical issues are initiated because of the scale-up. The remaining major issues are feasibilities of high accuracy automatic winding and optimization of welding of cover plates to fix the conductor in a radial plate, which is structure to enhance mechanical and electrical reliability. The authors therefore develop one of major parts of automatic winding machine, bending roller head, and successfully performed trial winding for 1/3 scale D-shaped winding. In addition, cover plate welding test was carried out using 1-m RP section and distortion of the radial plate is estimated to be in the requirement.

Design study of an ECRF antenna for JT-60SA

Kobayashi, Takayuki; Moriyama, Shinichi; Fujii, Tsuneyuki; Takahashi, Koji; Kajiwara, Ken; Sakamoto, Keishi

Fusion Engineering and Design, 84(2-6), p.1063 - 1067, 2009/06

https://doi.org/10.1016/j.fusengdes.2008.12.015

Design study of an ECRF antenna for JT-60SA is being carried out. An antenna concept which equips a linearly-driven flat mirror and a fixed-curved mirror was evaluated by using a numerical code. This antenna enables to alter millimeter-wave injection angle widely, only by a linear motion of the flat mirror. Therefore, the antenna eliminates a flexible tube for coolant supply and a link mechanism in vacuum vessel. In this work, it was clarified that the beam radius was expanded wider in poloidal direction compared with the expected one which was evaluated by the Gaussian optics. Then, the surface structures of the mirrors were modified in order to obtain the narrower beam radius on the resonance surface. It was found that the beam radius on the resonance surface became effectively narrower than that of the previous design by an increase in the curvature radius of the fixed-curved mirror or by a modification of the linearly-driven mirror as a diverging mirror.

Conceptual design of the quench protection circuits for the JT-60SA superconducting magnets

Gaio, E.*; Novello, L.*; Piovan, R.*; Shimada, Katsuhiro; Terakado, Tsunehisa; Kurihara, Kenichi; Matsukawa, Makoto

Fusion Engineering and Design, 84(2-6), p.804 - 809, 2009/06

https://doi.org/10.1016/j.fusengdes.2008.12.100

This paper deals with the conceptual design of the Quench Protection Circuits (QPC) of JT-60SA which have to provide a fast removal of the energy stored in the superconducting coils in case of quench. The core of the QPC units is constituted by a dc current breaker, which diverts the coil current into a resistor for a fast machine de-energization. In this paper, a hybrid solution, composed of a mechanical bypass switch paralleled to a static switch based on Integrated Gate Commutated Thyristor (IGCT) technology, has been chosen and worked out; a pyrobreaker is added in series to the hybrid switch as a backup protection. The resulting design allows benefiting from the fast breaking and very low maintenance requirement of the static switches, besides maintaining the advantage of the much lower power losses of the mechanical bypass in normal operation.

Overview of the main challenges for the engineering design of the test facilities system of IFMIF

Molla, J.*; Nakamura, Kazuyuki

Fusion Engineering and Design, 84(2-6), p.247 - 251, 2009/06

https://doi.org/10.1016/j.fusengdes.2008.12.071

IFMIF is considered as a key element in the achievement of fusion reactors. Experiments performed in IFMIF will produce valuable information on the behavior of materials under reactor conditions. The project entered in July 2007 into EVEDA phase. The engineering design will be completed by June 2013. The Test Facilities is one of the three main systems that compose IFMIF. Irradiation of samples, remote handling between different rooms, post-irradiation studies and all the required technologies to operate the Test Facilities will be part of this system. The intense neutron flux in the Test Cell, the heat load in the modules, the use of tritium for in-situ experiments and the presence of high radioactivity levels in the access Cell makes the design of the Test Cell system somewhat complicated. The main challenges for the design of the Test Cell and the required Validation of new technologies will be studied in this paper.

Development of 1 MeV accelerator and HV bushing at JAEA toward ITER neutral beam system

Umeda, Naotaka; Taniguchi, Masaki; Kashiwagi, Mieko; Dairaku, Masayuki; Hanada, Masaya; Tobari, Hiroyuki; Watanabe, Kazuhiro; Sakamoto, Keishi; Inoue, Takashi

Fusion Engineering and Design, 84(7-11), p.1875 - 1880, 2009/06

https://doi.org/10.1016/j.fusengdes.2008.12.018

The R&D of MeV class accelerator and high voltage bushing toward ITER neutral beam system are in progress at JAEA. To demonstrate the 1 MeV class negative ion beam acceleration, a MeV accelerator has been developed at MeV Test Facility. A H ion beam of 320 mA (current density: 140 A/m) was successfully accelerated up to 796 keV by fixing air leak in the ion source chamber caused by heat loading due to backstream positive ions. For development of the HV bushing, a half-size ( 800 mm) mockup single stage bushing was manufactured and tested, that demonstrated the dc voltage holding capability of up to 220 kV. Electrostatic analyses and design of the HV bushing are in progress, in which some design improvements are suggested to reduce electric field strength on the surface of ceramic insulator.

Approach to the lifetime assessment of the bayonet back plate for IFMIF target

Agostini, P.*; Ida, Mizuho; Miccich, G.*; Nakamura, Hiroo; Turroni, P.*

Fusion Engineering and Design, 84(2-6), p.364 - 368, 2009/06

https://doi.org/10.1016/j.fusengdes.2008.11.083

In the International Fusion Materials Irradiation Facility (IFMIF), a back-plate of liquid lithium (Li) target is most severely irradiated by high flux of neutrons. The conceptual configuration of the IFMIF target, based on the bayonet back plate, has been developed. The life time analysis of the back-plate has to be made carefully, in order to credibly estimate its expected replacement frequency. The various interconnections between the main damaging causes were discussed in order to evidence the most plausible reasons of the back-plate malfunctioning. The results were as follows. Under IFMIF Li condition with nitrogen content of 10 wppm or less, the erosion/corrosion rate of the back-plate made of the reduced activation ferritic steel is 0.003 mm/year or less. Even in case that the Bragg peak of heat deposition by deuterons approaches to the back-plate by 1 mm, temperature of the back-plate increases by only 12 degrees or less, and induced fatigue load is well low. Thermal stress due to the neutron irradiation is 93 MPa at the back-plate center, which is well below the plastic limit. The positive shift in the ductile brittle transition temperature (DBTT), the most effective factor, reduces the back-plate lifetime to three months or shorter under irradiation damage rate of 60 dpa/year. However, the lifetime can be recovered by the post irradiation annealing heat treatment.

Analysis of residual gas by high-resolution mass spectrometry during helium glow discharge cleaning in JT-60U

Hayashi, Takao; Kaminaga, Atsushi; Arai, Takashi; Sato, Masayasu

Fusion Engineering and Design, 84(2-6), p.908 - 910, 2009/06

https://doi.org/10.1016/j.fusengdes.2009.02.001

The residual gas analysis has been conducted by high-resolution mass spectrometry which can discriminate between D and He gas species during helium glow discharge cleaning (He-GDC) in JT-60U in order to investigate the effect of He-GDC. The residual gas analyzer was able to distinguish between D and He peaks during He-GDC. Since the He-GDC started, the partial pressure of D gas increases with time and reached its highest pressure (3.8 10 Pa), which is about ten times larger than that before the He-GDC (3.5 10 Pa). The total amount of D, which was released during the He-GDC (7 hours), was evaluated as 4 Pa m. The pressure of D (5.7 10 Pa) about 7 hours after the He-GDC (7 hours) is significantly lower than before the He-GDC, which indicates the He-GDC is effective to remove the deuterium from plasma facing components.

Status of engineering design of liquid lithium target in IFMIF-EVEDA

Nakamura, Hiroo; Agostini, P.*; Ara, Kuniaki; Fukada, Satoshi*; Furuya, Kazuyuki*; Garin, P.*; Gessii, A.*; Giusti, D.*; Groeschel, F.*; Horiike, Hiroshi*; et al.

Fusion Engineering and Design, 84(2-6), p.252 - 258, 2009/06

https://doi.org/10.1016/j.fusengdes.2009.02.019

Upgrading the NIFS superconductor test facility for JT-60SA cable-in-conduit conductors

Obana, Tetsuhiro*; Takahata, Kazuya*; Hamaguchi, Shinji*; Yanagi, Nagato*; Mito, Toshiyuki*; Imagawa, Shinsaku*; Kizu, Kaname; Tsuchiya, Katsuhiko; Hoshi, Ryo; Yoshida, Kiyoshi

Fusion Engineering and Design, 84(7-11), p.1442 - 1445, 2009/06

https://doi.org/10.1016/j.fusengdes.2008.12.085

The superconductor test facility in National Institute for Fusion Science (NIFS) has been upgraded in order to test cable-in-conduit (CIC) conductors for the JT-60SA equilibrium field (EF) coil. In the test facility, supercritical helium (SHe) lines were newly assembled with transfer tubes and a heat exchanger. The CIC conductor was covered with a thermal insulation vessel filled with gas helium at atmospheric pressure. The temperature of the conductor was varied using a film heater attached to an inlet pipe. The Ic and Tcs measurements of the prototype CIC conductor have successfully been carried out with the upgraded test facility. In the measurements, the conductor temperature was precisely controlled.

Problems of lead nuclear data in fusion blanket design

Kondo, Keitaro; Murata, Isao*; Klix, A.*; Seidel, K.*; Freiesleben, H.*

Fusion Engineering and Design, 84(7-11), p.1076 - 1086, 2009/06

https://doi.org/10.1016/j.fusengdes.2009.02.032

Several participants of the International Thermonuclear Experimental Reactor (ITER), such as Japan and EU, intend to introduce a Test Blanket Module (TBM) using a liquid lithium lead eutectic, which is used for the neutron multiplier and the tritium breeder. Recently a preliminary experiment in which a LiAlPb assembly was irradiated with 14 MeV neutrons was conducted at Technische Universitt Dresden. We found out that the neutron flux inside the assembly calculated with JENDL-3.3 underestimates an experimental value in the 10-16 MeV region by around 30% and that in the 0.5-5 MeV region by around 15%, while the calculated flux with JEFF-3.1 overestimates the measurement in the 5-10 MeV region by around 20%. In order to reveal a reason of the discrepancy, problems of the nuclear data libraries for lead were investigated. As a result, the following problems of the evaluated libraries were pointed out: The cross sections of the elastic scattering in JENDL-3.3 for lead isotopes are too small and cause a significant underestimation of the neutron flux above 10 MeV, which appeared in the analysis of the above experiment. Inelastic scattering data for Pb in JENDL-3.3 reproduce previous experimental double-differential cross section data most well. However, those for the other lead isotopes have some problems and cause a large underestimation of the neutron flux from 0.5 to 5 MeV. The reason of the overestimation in the energy region of 5-10 MeV with JEFF-3.1 is still unclear.