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Journal Articles

Actuator and diagnostic requirements of the ITER plasma control system

Snipes, J. A.*; Beltran, D.*; Casper, T.*; Gribov, Y.*; Isayama, Akihiko; Lister, J.*; Simrock, S.*; Vayakis, G.*; Winter, A.*; Yang, Y.*; et al.

Fusion Engineering and Design, 87(12), p.1900 - 1906, 2012/12

 Times Cited Count:21 Percentile:86.21(Nuclear Science & Technology)

no abstracts in English

Journal Articles

Current ramps in tokamaks; From present experiments to ITER scenarios

Imbeaux, F.*; Citrin, J.*; Hobirk, J.*; Hogeweij, G. M. D.*; K$"o$chl, F.*; Leonov, V. M.*; Miyamoto, Seiji; Nakamura, Yukiharu*; Parail, V.*; Pereverzev, G. V.*; et al.

Nuclear Fusion, 51(8), p.083026_1 - 083026_11, 2011/08

 Times Cited Count:35 Percentile:84.28(Physics, Fluids & Plasmas)

Journal Articles

Current ramps in tokamaks; From present experiments to ITER scenarios

Imbeaux, F.*; Basiuk, V.*; Budny, R.*; Casper, T.*; Citrin, J.*; Fereira, J.*; Fukuyama, Atsushi*; Garcia, J.*; Gribov, Y. V.*; Hayashi, Nobuhiko; et al.

Proceedings of 23rd IAEA Fusion Energy Conference (FEC 2010) (CD-ROM), 8 Pages, 2011/03

Journal Articles

ECRH assisted plasma start-up with toroidally inclined launch; Multi-machine comparison and perspectives for ITER

Stober, J.*; Jackson, G. L.*; Ascasibar, E.*; Bae, Y.-S.*; Bucalossi, J.*; Cappa, A.*; Casper, T.*; Cho, M. H.*; Gribov, Y.*; Granucci, G.*; et al.

Proceedings of 23rd IAEA Fusion Energy Conference (FEC 2010) (CD-ROM), 8 Pages, 2010/10

Journal Articles

Current ramps in tokamaks; From present experiments to ITER scenarios

Imbeaux, F.*; Basiuk, V.*; Budny, R.*; Casper, T.*; Citrin, J.*; Fereira, J.*; Fukuyama, Atsushi*; Garcia, J.*; Gribov, Y. V.*; Hayashi, Nobuhiko; et al.

Proceedings of 23rd IAEA Fusion Energy Conference (FEC 2010) (CD-ROM), 8 Pages, 2010/10

In order to prepare adequate current ramp-up and ramp-down scenarios for ITER, present experiments from several tokamaks have been analyzed by means of integrated modeling in view of determining relevant heat transport models for these operation phases. The results of these studies are presented and projections to ITER current ramp-up and ramp-down scenarios are done, focusing on the baseline inductive scenario (main heating plateau current of 15 MA). Various transport models have been tested by means of integrated modeling against experimental data from ASDEX Upgrade, C-Mod, DIII-D, JET and Tore Supra, including both Ohmic plasmas and discharges with additional heating/current drive. With using the most successful models, projections to the ITER current ramp-up and ramp-down phases are carried out. Though significant differences between models appear on the electron temperature prediction, the final q-profiles reached in the simulation are rather close.

Journal Articles

Experimental studies of ITER demonstration discharges

Sips, A. C. C.*; Casper, T.*; Doyle, E. J.*; Giruzzi, G.*; Gribov, Y.*; Hobirk, J.*; Hogeweij, G. M. D.*; Horton, L. D.*; Hubbard, A. E.*; Hutchinson, I.*; et al.

Nuclear Fusion, 49(8), p.085015_1 - 085015_11, 2009/08

 Times Cited Count:48 Percentile:87.48(Physics, Fluids & Plasmas)

Key parts of the ITER scenarios are determined by the capability of the proposed poloidal field (PF) coil set. They include the plasma breakdown at low loop voltage, the current rise phase, the performance during the flat top (FT) phase and a ramp down of the plasma. The ITER discharge evolution has been verified in dedicated experiments. New data are obtained from C-Mod, ASDEX Upgrade, DIII-D, JT-60U and JET. Results show that breakdown for $$E$$$$_{axis}$$ $$<$$ 0.23-0.33 V m$$^{-1}$$ is possible unassisted (ohmic) for large devices like JET and attainable in devices with a capability of using ECRH assist. For the current ramp up, good control of the plasma inductance is obtained using a full bore plasma shape with early X-point formation. This allows optimization of the flux usage from the PF set. Additional heating keeps $$l$$$$_{i}$$(3) $$<$$ 0.85 during the ramp up to $$q$$$$_{95}$$ = 3. A rise phase with an H-mode transition is capable of achieving $$l$$$$_{i}$$(3) $$<$$ 0.7 at the start of the FT. Operation of the H-mode reference scenario at $$q$$$$_{95}$$ $$sim$$ 3 and the hybrid scenario at $$q$$$$_{95}$$ = 4-4.5 during the FT phase is documented, providing data for the $$l$$$$_{i}$$(3) evolution after the H-mode transition and the $$l$$$$_{i}$$(3) evolution after a back-transition to L-mode. During the ITER ramp down it is important to remain diverted and to reduce the elongation. The inductance could be kept $$leq$$ 1.2 during the first half of the current decay, using a slow $$I$$$$_{p}$$ ramp down, but still consuming flux from the transformer. Alternatively, the discharges can be kept in H-mode during most of the ramp down, requiring significant amounts of additional heating.

JAEA Reports

Studies on representative disruption scenarios, associated electromagnetic and heat loads and operation window in ITER

Fujieda, Hirobumi; Sugihara, Masayoshi*; Shimada, Michiya; Gribov, Y.*; Ioki, Kimihiro*; Kawano, Yasunori; Khayrutdinov, R.*; Lukash, V.*; Omori, Junji; Neyatani, Yuzuru

JAEA-Research 2007-052, 115 Pages, 2007/07

JAEA-Research-2007-052.pdf:3.58MB

Impacts of plasma disruptions on ITER have been investigated to confirm the robustness of the design of the machine to the potential consequential loads. The loads include both electro-magnetic (EM) and heat on the in-vessel components and the vacuum vessel. Several representative disruption scenarios are specified. Disruption simulations with the DINA code and EM load analyses with a 3D finite element method code are performed for these scenarios. Some margins are confirmed in the EM load. Heat load on the first wall due to the vertical movement and the thermal quench (TQ) is calculated with a 2D heat conduction code. For vertical displacement event, beryllium ($$Be$$) wall will not melt during the vertical movement, prior to the TQ. Significant melting is anticipated for the upper $$Be$$ wall and tungsten baffle due to the TQ after the vertical movement. However, its impact could be mitigated by implementing a reliable detection system of the vertical movement and a mitigation system.

Journal Articles

Progress in the ITER physics basis, 1; Overview and summary

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:624 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.

Journal Articles

Progress in the ITER physics basis, 3; MHD stability, operational limits and disruptions

Hender, T. C.*; Wesley, J. C.*; Bialek, J.*; Bondeson, A.*; Boozer, A. H.*; Buttery, R. J.*; Garofalo, A.*; Goodman, T. P.*; Granetz, R. S.*; Gribov, Y.*; et al.

Nuclear Fusion, 47(6), p.S128 - S202, 2007/06

 Times Cited Count:759 Percentile:98.25(Physics, Fluids & Plasmas)

no abstracts in English

Journal Articles

Progress in the ITER physics basis, 8; Plasma operation and control

Gribov, Y.*; Humphreys, D. A.*; Kajiwara, Ken*; Lazarus, E. A.*; Lister, J. B.*; Ozeki, Takahisa; Portone, A.*; Shimada, Michiya*; Sips, A. C. C.*; Wesley, J. C.*

Nuclear Fusion, 47(6), p.S385 - S403, 2007/06

 Times Cited Count:114 Percentile:97.32(Physics, Fluids & Plasmas)

This chapter describes the progress achieved in these areas in the tokamak experiments since the ITER Physics Basis (1999 Nucl. Fusion 39 2577) was written and the results of assessment of ITER to provide the plasma initiation and basic control. This chapter considers only plasma initiation and plasma basic control. The experiments on plasma initiation performed in DIII-D and JT-60U, as well as the theoretical studies performed for ITER, have demonstrated that, ITER can produce plasma initiation in a low toroidal electric field of 0.3V/m, if it is assisted by about 2MW of ECRF heating. The plasma basic control is described, which includes control of the plasma current, position and shape - the plasma magnetic control, as well as control of other plasma global parameters or their profiles - the plasma performance control.

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.

Journal Articles

Disruption scenarios, their mitigation and operation window in ITER

Shimada, Michiya; Sugihara, Masayoshi; Fujieda, Hirobumi*; Gribov, Y.*; Ioki, Kimihiro*; Kawano, Yasunori; Khayrutdinov, R.*; Lukash, V.*; Omori, Junji

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

Several representative disruption scenarios are specified and disruption simulations are performed with the DINA code and EM load analyses with the 3D FEM code for these scenarios based on newly derived physics guidelines. Although some margin is confirmed in the EM loads due to induced eddy and halo currents on the in-vessel components for all of the representative scenarios, but the margin is not large. The heat load on various parts of the first wall due to vertical movements and thermal quenches is calculated. The beryllium wall will not melt during vertical movement. Melting is anticipated at the thermal quench during a VDE, though its impact could be reduced substantially by implementing a reliable detection and mitigation system, e.g., massive gas injection. With unmitigated disruptions, the loss of beryllium layer is expected to be within 30 $$mu$$m/event out of 10 mm thick beryllium first wall.

Journal Articles

Analysis of the direction of plasma vertical movement during major disruptions in ITER

Lukash, V.*; Sugihara, Masayoshi; Gribov, Y.*; Fujieda, Hirobumi*

Plasma Physics and Controlled Fusion, 47(12), p.2077 - 2086, 2005/12

 Times Cited Count:8 Percentile:26.36(Physics, Fluids & Plasmas)

Vertical directions of plasma movement after the thermal quench (TQ) of major disruptions in ITER are investigated using the predictive mode of the DINA code. Three dominant parameters in determining the direction of plasma movement are identified; (1) the rate of plasma current quench, (2) change of the internal plasma inductance li associated with the TQ and (3) the initial vertical position of plasma column before the TQ. It is shown that the reference ITER plasma moves upward after the TQ, if the current quench rate is higher than 200kA/ms and the drop of li does not exceed 0.2 for the present reference initial vertical position (55.5 cm above the center of machine).

Journal Articles

Progress in physics basis and its impact on ITER

Shimada, Michiya; Campbell, D.*; Stambaugh, R.*; Polevoi, A. R.*; Mukhovatov, V.*; Asakura, Nobuyuki; Costley, A. E.*; Donn$'e$, 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 $$sim$$ 4. Extrapolation to ITER suggests that at the reduced plasma current of $$sim$$ 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.

Journal Articles

Performance of ITER as a burning plasma experiment

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.

Journal Articles

Performance of ITER as a burning plasma experiment

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:78.24(Physics, Fluids & Plasmas)

Performance analysis based on empirical scaling shows the feasibility of achieving Q $$geq$$ 10 in inductive operation. Analysis has also elucidated a possibility that ITER can potentially demonstrate Q $$sim$$ 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 $$geq$$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) $$geq$$ 1.3 and betaN $$geq$$ 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.

Journal Articles

Current profile behavior during ramping-up phase in high bootstrap current tokamak plasmas

Nakamura, Yukiharu; Tsutsui, Hiroaki*; Takei, Nahoko*; Shirai, Hiroshi; Sugihara, Masayoshi; Gribov, Y.*; Ozeki, Takahisa; Tobita, Kenji; Iio, Shunji*; Jardin, S. C.*

Journal of Plasma and Fusion Research SERIES, Vol.6, p.196 - 198, 2004/00

no abstracts in English

Journal Articles

Overview of physics basis for ITER

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:46 Percentile:80.66(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.

Journal Articles

Examinations on plasma behaviour during disruptions on existing tokamaks and extrapolation to ITER

Sugihara, Masayoshi; Lukash, V.*; Kawano, Yasunori; Yoshino, Ryuji; Gribov, Y.*; Khayrutdinov, R.*; Miki, Nobuharu*; Omori, Junji*; Shimada, Michiya

Proceedings of 30th EPS Conference on Controlled Fusion and Plasma Physics (CD-ROM), 4 Pages, 2003/07

We examine plasma behaviours during plasma disruptions in detail in JT-60U and other tokamaks to derive appropriate physics guidelines for the behaviours. Their interpretations and their extrapolations to ITER are incorporated into the DINA code, which solves plasma transport and 2D free boundary plasma equilibrium simultaneously with circuit equations for the vacuum vessel and the PF coils. Sensitivity of the plasma behaviours and their impact on the EM force during disruptions due to the range of variation and uncertainty of the experimental data are examined.

Journal Articles

Wave form of current quench during disruptions in Tokamaks

Sugihara, Masayoshi; Lukash, V.*; Kawano, Yasunori; Yoshino, Ryuji; Gribov, Y.*; Khayrutdinov, R.*; Miki, Nobuharu*; Omori, Junji*; Shimada, Michiya

Purazuma, Kaku Yugo Gakkai-Shi, 79(7), p.706 - 712, 2003/07

The time dependence of the current decay during the current quench phase of disruptions, which can significantly influence the electro-magnetic force on the in-vessel components due to the induced eddy currents, is investigated using data obtained in JT-60U experiments in order to derive a relevant physics guideline for the predictive simulations of disruptions in ITER. It is shown that an exponential decay can fit the time dependence of current quench for discharges with large quench rate (fast current quench). On the other hand, for discharges with smaller quench rate (slow current quench), a linear decay can fit the time dependence of current quench better than exponential.

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