Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Shimada, Michiya; Campbell, D. J.*; Mukhovatov, V.*; Fujiwara, Masami*; Kirneva, N.*; Lackner, K.*; Nagami, Masayuki; Pustovitov, V. D.*; Uckan, N.*; Wesley, J.*; et al.
Nuclear Fusion, 47(6), p.S1 - S17, 2007/06
Times Cited Count：744 Percentile：99.93(Physics, Fluids & Plasmas)The Progress in the ITER Physics Basis document is an update of the ITER Physics Basis (IPB), which was published in 1999. The IPB provided methodologies for projecting the performance of burning plasmas, developed largely through coordinated experimental, modeling and theoretical activities carried out on today's tokamaks (ITER Physics R&D). In the IPB, projections for ITER (1998 Design) were also presented. The IPB also pointed out some outstanding issues. These issues have been addressed by the International Tokamak Physics Activities (ITPA), which were initiated by the European Union, Japan, Russia and the U.S.A.. The new methodologies of projection and control developed through the ITPA are applied to ITER, which was redesigned under revised technical objectives, but will nonetheless meet the programmatic objective of providing an integrated demonstration of the scientific and technological feasibility of fusion energy.
Doyle, E. J.*; Houlberg, W. A.*; Kamada, Yutaka; Mukhovatov, V.*; Osborne, T. H.*; Polevoi, A.*; Bateman, G.*; Connor, J. W.*; Cordey, J. G.*; Fujita, Takaaki; et al.
Nuclear Fusion, 47(6), p.S18 - S127, 2007/06
no abstracts in English
Kamada, Yutaka; Leonard, A. W.*; Bateman, G.*; Becoulet, M.*; Chang, C. S.*; Eich, T.*; Evans, T. E.*; Groebner, R. J.*; Guzdar, P. N.*; Horton, L. D.*; et al.
Proceedings of 21st IAEA Fusion Energy Conference (FEC 2006) (CD-ROM), 8 Pages, 2007/03
no abstracts in English
Polevoi, A. R.*; Shimada, Michiya; Sugihara, Masayoshi; Igitkhanov, Y. L.*; Mukhovatov, V.*; Kukushkin, A. S.*; Medvedev, S. Y.*; Zvonkov, A. V.*; Ivanov, A. A.*
Nuclear Fusion, 45(11), p.1451 - 1456, 2005/11
Times Cited Count：33 Percentile：70.37(Physics, Fluids & Plasmas)Requirements for pellet injection parameters for plasma fuelling are assessed for ITER scenarios with enhanced particle confinement. A pellet injection throughput of 100 Pam
/s would be sufficient. The assessment is based on the integrated transport simulations including models of pedestal transport, reduction of helium transport and boundary conditions compatible with SOL/divertor simulations. The requirements for pellet injection for the inductive H-mode scenario (HH98(y,2) = 1) are reconsidered taking account of a possible reduction of the particle loss obtained in some experiments at low collisionalities. The assessment of fuelling requirements is carried out for the hybrid and steady state scenarios with enhanced confinement with HH98(y,2) 
1. A robustness of plasma performance to the variation of particle transport is demonstrated. A new type of steady state (SS) scenario is considered with neutral beam current drive (NBCD) and electron cyclotron current drive (ECCD) instead of lower hybrid current drive (LHCD).
Shimada, Michiya; Costley, A. E.*; Federici, G.*; Ioki, Kimihiro*; Kukushkin, A. S.*; Mukhovatov, V.*; Polevoi, A. R.*; Sugihara, Masayoshi
Journal of Nuclear Materials, 337-339, p.808 - 815, 2005/03
Times Cited Count：65 Percentile：96.35(Materials Science, Multidisciplinary)ITER is an experimental fusion reactor for investigation and demonstration of burning plasmas, characterised of its heating dominated by alpha-particle heating. ITER is a major step from present devices and an indispensable step for fusion reactor development. ITER's success largely depends on the control of plasma-wall interactions(PWI), with power and particle fluxes and time scales one or two orders of magnitude larger than in present devices. The strategy for control of PWI includes the semi-closed divertor, strong fuelling and pumping, disruption and ELM control, replaceable plasma-facing materials and stepwise operation.
Shimada, Michiya; Campbell, D.*; Stambaugh, R.*; Polevoi, A. R.*; Mukhovatov, V.*; Asakura, Nobuyuki; Costley, A. E.*; Donn
, A. J. H.*; Doyle, E. J.*; Federici, G.*; et al.
Proceedings of 20th IAEA Fusion Energy Conference (FEC 2004) (CD-ROM), 8 Pages, 2004/11
This paper summarises recent progress in the physics basis and its impact on the expected performance of ITER. Significant progress has been made in many outstanding issues and in the development of hybrid and steady state operation scenarios, leading to increased confidence of achieving ITER's goals. Experiments show that tailoring the current profile can improve confinement over the standard H-mode and allow an increase in beta up to the no-wall limit at safety factors 
4. Extrapolation to ITER suggests that at the reduced plasma current of 12MA, high Q 10 and long pulse (1000 s) operation is possible with benign ELMs. Analysis of disruption scenarios has been performed based on guidelines on current quench rates and halo currents, derived from the experimental database. With conservative assumptions, estimated electromagnetic forces on the in-vessel components are below the design target values, confirming the robustness of the ITER design against disruption forces.
Connor, J. W.*; Fukuda, Takeshi*; Garbet, X.*; Gormezano, C.*; Mukhovatov, V.*; Wakatani, Masahiro*; ITB Database Group; ITPA Topical Group on Transport and Internal Barrier Physics*
Nuclear Fusion, 44(4), p.R1 - R49, 2004/04
Times Cited Count：301 Percentile：76.74(Physics, Fluids & Plasmas)This paper first reviews the present state of theoretical and experimental knowledge regarding the formation and characteristics of ITBs in tokamaks. Specifically, the current status of theoretical modeling of ITBs is presented; then, an international ITB database based on experimental information extracted from some nine tokamaks is described and used to draw some general conclusions concerning the necessary conditions for ITBs to appear, comparing these with the theoretical models. The experimental situation regarding the steady-state, or at least quasi-steady-state, operation of tokamaks is reviewed and finally the issues and prospects for achieving such operational modes in ITER are discussed.
Shimada, Michiya; Mukhovatov, V.*; Federici, G.*; Gribov, Y.*; Kukushkin, A.*; Murakami, Yoshiki*; Polevoi, A. R.*; Pustovitov, V. D.*; Sengoku, Seio; Sugihara, Masayoshi
Nuclear Fusion, 44(2), p.350 - 356, 2004/02
Recent performance analysis has improved confidence in achieving Q 10 in inductive operation in ITER. Performance analysis based on empirical scaling shows the feasibility of achieving Q 10 in inductive operation with a sufficient margin. Theory-based core modeling indicates the need of high pedestal temperature (2-4 keV) to achieve Q 10, which is in the range of projection with pedestal scaling. The heat load of type-I ELM could be made tolerable by high density operation and further tilting the target plate (if necessary). Pellet injection from High-Field Side would be useful in enhancing Q and reducing ELM heat load. Steady state operation scenarios have been developed with modest requirement on confinement improvement and beta (HH98(y,2) 1.3 and betaN 2.6). Stabilisation of RWM, required in such regimes, is feasible with the present saddle coils and power supplies with double-wall structure taken into account.
Shimada, Michiya; Mukhovatov, V.*; Federici, G.*; Gribov, Y.*; Kukushkin, A. S.*; Murakami, Yoshiki*; Polevoi, A. R.*; Pustovitov, V. D.*; Sengoku, Seio; Sugihara, Masayoshi
Nuclear Fusion, 44(2), p.350 - 356, 2004/02
Times Cited Count：39 Percentile：75.67(Physics, Fluids & Plasmas)Performance analysis based on empirical scaling shows the feasibility of achieving Q 
10 in inductive operation. Analysis has also elucidated a possibility that ITER can potentially demonstrate Q 50, enabling studies of self-heated plasmas. Theory-based core modeling indicates the need of high pedestal temperature (3.2 - 5.3 keV) to achieve Q 10, which is in the range of projection with presently available pedestal scalings. Pellet injection from high-field side would be useful in enhancing Q and reducing ELM heat load in high plasma current operation. If the ELM heat load is not acceptable, it could be made tolerable by further tilting the target plate. Steady state operation scenarios at Q = 5 have been developed with modest requirement on confinement improvement and beta (HH98(y,2) 1.3 and betaN 2.6). Stabilisation of RWM, required in such regimes, is feasible with the present saddle coils and power supplies with double-wall structure taken into account.
Mukhovatov, V.*; Shimada, Michiya; Chudnovskiy, A. N.*; Costley, A. E.*; Gribov, Y.*; Federici, G.*; Kardaun, O. J. F.*; Kukushkin, A. S.*; Polevoi, A. R.*; Pustovitov, V. D.*; et al.
Plasma Physics and Controlled Fusion, 45(12), p.235 - 252, 2003/12
Times Cited Count：55 Percentile：82.78(Physics, Fluids & Plasmas)ITER will be the first magnetic confinement device with burning DT plasma and fusion power of about 0.5 GW. During the past few years, new results have been obtained that substantiate the confidence in achieving Q 10 in ITER with inductive H-mode operation. These include achievement of a good H-mode confinement near the Greenwald density at high triangularity of the plasma cross section; improvements in theory-based confinement projections for the core plasma; improvement in helium ash removal due to the elastic collisions of He atoms with D/T ions in the divertor predicted by modelling; demonstration of feedback control of NTMs and resultant improvement in the achievable beta-values; better understanding of ELM physics and development of ELM mitigation techniques; and demonstration of mitigation of plasma disruptions. ITER will have a flexibility to operate also in steady state and intermediate (hybrid) regimes. The paper concentrates on inductively driven plasma performance and discusses requirements for steady-state operation in ITER.
Mukhovatov, V.*; Shimomura, Yasuo; Polevoi, A. R.*; Shimada, Michiya; Sugihara, Masayoshi; Bateman, G.*; Cordey, J. G.*; Kardaun, O. J. F.*; Pereverzev, G. V.*; Voitsekhovich, I.*; et al.
Nuclear Fusion, 43(9), p.942 - 948, 2003/09
Times Cited Count：43 Percentile：76.69(Physics, Fluids & Plasmas)The values of Q = (fusion power)/(auxiliary heating power) predicted for ITER by three different methods are compared. The first method utilises an empirical confinement time scaling and prescribed radial profiles of transport coefficients, the second approach extrapolates from especially designed ITER similarity experiments, and the third approach is based on partly theory-based transport models. The energy confinement time given by the ITERH-98(y,2) scaling for an inductive scenario with plasma current of 15 MA and plasma density 15% below the Greenwald density is 3.7 s with one estimated technical standard deviation of 14%. This translates in the first approach into an interval for Q of [6-15] at the auxiliary heating power Paux = 40 MW and [6-30] at the minimum heating power satisfying a good confinement ELMy H-mode. Predictions of similarity experiments from JET and of theory-based transport models overlap with the prediction using the empirical confinement-time scaling within its estimated margin of uncertainty.
Sugihara, Masayoshi; Mukhovatov, V.*; Polevoi, A.*; Shimada, Michiya
Plasma Physics and Controlled Fusion, 45(9), p.L55 - L62, 2003/09
Times Cited Count：40 Percentile：74.98(Physics, Fluids & Plasmas)Improvement of the scaling of H-mode edge pedestal pressure for a Type-I ELM regime based on a simple analytical formula is achieved by introducing shaping factors into the scaling. The physics basis of these shaping factors is an enhancement of the critical pressure gradient by the magnetic well. With increasing the magnetic well, a high toroidal mode number ideal ballooning mode and peeling mode decouple and critical gradient is determined by intermediate mode number, which results in significant enhancement of the critical pressure gradient. The improved scaling can reproduce the pedestal pressure reasonably well for ASDEX-U, JET, DIII-D and JT-60U Type-I ELmy data archived in the ITER pedestal database.
Chankin, A. V.; Mukhovatov, V. S.*; Fujita, Takaaki; Miura, Yukitoshi
Europhysics Conference Abstracts (CD-ROM), 26B, 4 Pages, 2002/00
no abstracts in English
Pustovitov, V. D.*; Mikhailovskii, A. B.*; Kobayashi, Noriyuki*; Konovalov, S. V.*; Mukhovatov, V. S.*; Zvonkov, A. V.*
Proceedings of IAEA 18th Fusion Energy Conference (CD-ROM), 5 Pages, 2001/00
no abstracts in English
Matsumoto, Hiroshi; Boucher, D.*; Mukhovatov, V.*
26th European Physical Society Conference on Controlled Fusion and Plasma Physics (CD-ROM), 4 Pages, 1999/00
no abstracts in English
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.
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.
