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Journal Articles

ITER test blanket module error field simulation experiments at DIII-D

Schaffer, M. J.*; Snipes, J. A.*; Gohil, P.*; de Vries, P.*; Evans, T. E.*; Fenstermacher, M. E.*; Gao, X.*; Garofalo, A. M.*; Gates, D. A.*; Greenfield, C. M.*; et al.

Nuclear Fusion, 51(10), p.103028_1 - 103028_11, 2011/10

 Times Cited Count:32 Percentile:15.51(Physics, Fluids & Plasmas)

Experiments at DIII-D investigated the effects of ferromagnetic error fields similar to those expected from proposed ITER Test Blanket Modules (TBMs). Studied were effects on: plasma rotation and locking; confinement; L-H transition; edge localized mode (ELM) suppression by resonant magnetic perturbations; ELMs and the H-mode pedestal; energetic particle losses; and more. The experiments used a 3-coil mock-up of 2 magnetized ITER TBMs in one ITER equatorial port. The experiments did not reveal any effect likely to preclude ITER operations with a TBM-like error field. The largest effect was slowed plasma toroidal rotation v across the entire radial profile by as much as $$Delta v/v_{0} sim 50%$$ via non-resonant braking. Changes to global $$Delta n/n_{0}$$, $$Delta v/v_{0}$$ and $$Delta H_{98}/H_{98,0}$$ were $$sim$$3 times smaller. These effects are stronger at higher $$beta$$ and lower $$v_{0}$$. Other effects were smaller.

Journal Articles

Six-party qualification program of FW fabrication methods for ITER blanket module procurement

Ioki, Kimihiro; Elio, F.*; Barabash, V.*; Chuyanov, V.*; Rozov, V.*; Wang, X.*; Chen, J.*; Wang, L.*; Lorenzetto, P.*; Peacock, A.*; et al.

Fusion Engineering and Design, 82(15-24), p.1774 - 1780, 2007/10

 Times Cited Count:11 Percentile:34.95(Nuclear Science & Technology)

In December 2005, the new procurement allocation plan of the ITER components among the seven Parties was prepared. The need to qualify for procurement of the specific components was especially introduced in the document. The main features and milestones of the qualification program are described in "Procurement Plan" for each specific component. Due to the complicated features of FW procurement, the procurement document has to be developed precisely. To guarantee high quality of 1700 FW panels produced by 6 different Parties, a qualification program is essential. The qualification mock-up is 80 mm wide, 240 mm long and 81 mm thick with 3 beryllium tiles 10 mm thick. Heat load tests will be performed on the qualification mock-ups in 2007 in EU and USA facilities. The maximum design heat load on the ITER FW is 0.5 MW/m $$^{2}$$ (steady state) $$times$$ 30,000 shots. Mechanical tests of joints are also required using standardized methods. Only Parties which have satisfied the acceptance criteria of the qualification tests can proceed to the procurement stage of the ITER FW. Semi-prototypes (roughly 1000 mm $$times$$ 200 mm) are also requested before the ITER FW manufacturing.

Journal Articles

Design progress of the ITER vacuum vessel sectors and port structures

Utin, Y.*; Ioki, Kimihiro; Alekseev, A.*; Bachmann, C.*; Cho, S. Y.*; Chuyanov, V.*; Jones, L.*; Kuzmin, E.*; Morimoto, Masaaki; Nakahira, Masataka; et al.

Fusion Engineering and Design, 82(15-24), p.2040 - 2046, 2007/10

 Times Cited Count:2 Percentile:79.6(Nuclear Science & Technology)

Recent progress of the ITER vacuum vessel (VV) design is presented. As the ITER construction phase approaches, the VV design has been improved and developed in more detail with the focus on better performance, improved manufacture and reduced cost. Based on achievements of manufacturing studies, design improvement of the typical VV sector (#1) has been nearly finalized. Design improvement of other sectors is in progress - in particular, of the VV sector #2 and #3 which interface with the ports for the neutral beam injection. For all sectors, the concept for the in-wall shielding has progressed and developed in more detail. The design progress of the VV sectors has been accompanied by the progress of the port structures. In particular, design of the NB Ports was advanced with the focus on the heat-flux components to handle the heat input of the neutral beams. Structural analyses have been performed to validate all design improvements.

Journal Articles

ITER limiters moveable during plasma discharge and optimization of ferromagnetic inserts to minimize toroidal field ripple

Ioki, Kimihiro; Chuyanov, V.*; Elio, F.*; Garkusha, D.*; Gribov, Y.*; Lamzin, E.*; Morimoto, Masaaki; Shimada, Michiya; Sugihara, Masayoshi; Terasawa, Atsumi; et al.

Proceedings of 21st IAEA Fusion Energy Conference (FEC 2006) (CD-ROM), 8 Pages, 2007/03

Two important design updates have been made in the ITER VV and in-vessel components recently. One is the introduction of limiters moveable during a plasma discharge, and the other is optimization of the ferromagnetic insert configuration to minimize the toroidal field ripple. In the new limiter concept, the limiters are retracted by 8 cm during the plasma flat top phase in the divertor configuration. This concept gives important advantages: (1) the particle and heat loads due to disruptions, ELMs and blobs on the limiters will be mitigated approximately by a factor 1.5 or more; (2) the gap between the plasma and the ICRH antenna can be reduced to improve the coupling of the ICRH power. The ferromagnetic inserts have previously not been planned to be installed in the outboard midplane region between equatorial ports due to irregularity of tangential ports for NB injection. The result is a relatively large ripple (1 %) in a limited region of the plasma, which nevertheless seems acceptable from the plasma performance viewpoint. However, toroidal field flux lines fluctuate 10 mm due to the large ripple in the FW region. To avoid problems due to the large TF flux line fluctuation, additional ferromagnetic inserts are now planned to be installed in the equatorial port region.

Oral presentation

An Assessment of ITER scenarios under varying assumptions of NBI and LHCD capability

Oikawa, Toshihiro; Polevoi, A. R.*; Mukhovatov, V.*; Shimada, Michiya; Bonoli, P.*; Campbell, D.*; Chuyanov, V.*

no journal, , 

In the ITER design review, reducing the NBI energy is proposed for increasing the plasma rotation. NBI capability is assessed for various design assumptions. In the D-T operation, reliable operation would be obtained well above the H-L transition boundary. In the hydrogen operation, the beam energy has to be less than 500keV due to the NBI shinethrough, and the H-mode operation regime is narrow. The NBI central heating at high density necessary for the ITER mission is difficult at 500keV. NB current drive decreases by 20% at 750keV, which makes it difficult to prospect the steady state scenario. The rotation increases by 13% at 750keV. Based on recent experiments, however, MHD instabilities can be suppressed at a rotation velocity available with 1MeV NBIs. A steady state scenario using LHCD is explored with a transport code employing a LHCD code. With a reasonable HH=1.4, 93% of the plasma current is no-inductively driven, and hence the pulse duration is extended to the machine limit.

Oral presentation

Physics assessment of the NBI capability in ITER plasmas

Oikawa, Toshihiro; Polevoi, A. R.*; Mukhovatov, V.*; Sakamoto, Yoshiteru; Kamada, Yutaka; Shimada, Michiya*; Campbell, D. J.*; Chuyanov, V.*; Schunke, B.*; Tanga, A.*; et al.

no journal, , 

In the ITER design review, reducing the NBI energy is proposed for increasing the plasma rotation in terms of suppressing MHD instabilities. NBI capability has been assessed for various design possibilities. In D-T operation, the planned auxiliary heating power makes possible reliable operation well above the H-L transition boundary. In hydrogen operation the beam energy should be limited to 500keV due to the NBI shinethrough. However, reduction of the NBI central heating at high density necessary for the ITER mission is significant at 500keV, confirming the need for higher energy in DT operation. Although the rotation increases by 13% at 750keV relative to 1MeV at constant power, NB current drive decreases by 20%, which would be problematic for the development of steady-state scenarios. Therefore it is concluded that the beam energy should be kept 1MeV in DT operation. The beam energy variation in a pulse enables the plasma beta control for avoiding the stability boundary.

Oral presentation

Results of ITER test blanket module mock-up experiments on DIII-D

Snipes, J. A.*; Schaffer, M. J.*; Gohil, P.*; de Vries, P.*; Fenstermacher, M. E.*; Evans, T. E.*; Gao, X. M.*; Garofalo, A.*; Gates, D. A.*; Greenfield, C. M.*; et al.

no journal, , 

A series of experiments was performed on DIII-D to mock-up the field that will be induced in a pair of ferromagnetic Test Blanket Modules (TBMs) in ITER to determine the effects of such error fields on plasma operation and performance. A set of coils producing both poloidal and toroidal fields was placed inside a re-entrant horizontal port close to the plasma. The coils produce a localized ripple due to the toroidal field (TF) + TBM up to 5.7%, which is more than four times that expected from a pair of representative 1.3 ton TBMs in ITER. The experiments show that the reduction in the toroidal rotation is sensitive to the ripple. On the other hand, the confinement is reduced by up to 15-18% for local ripple $$ge$$ 3% but is hardly affected at 1.7% local ripple.

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