Publications (96)187.7 Total impact
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ABSTRACT: The effects of a radial electric field (EF) on the losses of alpha particles and neutral beam injection (NBI) fast ions in typical ITER operation scenarios for both error fields due to test blanket modules (TBMs) and toroidal field (TF) ripple were evaluated using an iterative method to execute an orbitfollowing MonteCarlo code and a onedimensional transport code. The EF effect on the loss of fast ions strongly depends on the operation scenario as well as on the error field. The electric field is very significant in the loss of fast ions in a 9 MA ITER operation scenario with a higher safety factor and in the error field associated with TBMs. The EF effect in the error field of TF ripple is very small in any operation scenario. The electric field is much more significant for the loss of NBI fast ions than for that of alpha particles. The radial electric field changes the toroidal precession of fast ions and consequently alters their condition of resonance with the error field, which may account for the EF effect on the loss of fast ions in ITER with TBMs.Nuclear Fusion 05/2015; 55(5):053010. DOI:10.1088/00295515/55/5/053010 · 3.06 Impact Factor  [Show abstract] [Hide abstract]
ABSTRACT: An assessment of ITER plasma parameters is carried out for the low activation phase that is required for commissioning the basic ITER systems including plasma control, heating and current drive. Such an operation is analysed for hydrogen, helium and deuterium plasmas for full field and current, as well as with magnetic field and plasma current reduced to half of their design values, B0 = 2.65 T, Ip = 7.5 MA. Both hydrogen and deuterium neutral beam injection (NBI) are considered. We assess the possible domain for safe operation, and the possible target plasmas for commissioning the NBI, electron cyclotron heating (ECH) and ion cyclotron heating (ICH) systems, taking into account the constraints imposed by NB shinethrough loss, Greenwald limit and access to Hmode operation. Simulations with the Automated System for Transport Analysis (ASTRA) show that for 33 MW of NBI with 20 MW of ECH, Hmode access is marginal for hydrogen plasmas. Good Hmode confinement, expected at PNB + PEC + PIC > 1.5 PLH, is more likely for the helium and deuterium cases. It is found that plasma parameters, such as normalized beta, plasma density and current flattop duration, for full power/half field/half current operation can be similar to those required for the DT long pulse operation. Preliminary assessment is also made of the maximum of tritium and neutron yield achievable in a single shot at the deuterium phase of ITER operation.Nuclear Fusion 12/2013; 53(12):3026. DOI:10.1088/00295515/53/12/123026 · 3.06 Impact Factor  [Show abstract] [Hide abstract]
ABSTRACT: The energetic ion loss has been assessed using the F3DOFMC code for a 15 MA inductive scenario with Q = 10 and the latest information on the first wall geometry, the implementation of ferritic inserts (FI) and the ELM mitigation/control coils. Alpha particles and NB ions generated by the neutral beam injectors with the injection energy of 1 MeV are well confined and the heat load on the first wall is negligibly small and allowable for the magnetic background by the toroidal field coils and FI. However, an increase in the loss of these energetic ions is observed when the magnetic field by the ELM coils is applied. The increase in the loss fraction is larger for NB ions than for alpha particles under the ELM coil field. The origin of the expelled NB ions is dominantly trapped ions generated in the peripheral region due to a highdensity plasma of the 15 MA scenario.Nuclear Fusion 09/2012; 52(9). DOI:10.1088/00295515/52/9/094008 · 3.06 Impact Factor 
Article: Effects of ELM mitigation coils on energetic particle confinement in ITER steadystate operation
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ABSTRACT: The effects of edgelocalized mode (ELM) mitigation coils (ELM coils) on the loss of NBIproduced fast ions and fusionproduced alpha particles are investigated using an orbit following Monte Carlo code. The ELM mitigation coil field (EMC field) may cause a significant loss of fast ions produced by NBI on the order of 16.0–17.0% for a 9 MA steadystate ITER scenario. A significant transitparticle loss occurs in the case of the toroidal mode number n = 4 in which magnetic surfaces are ergodic near the plasma periphery. When the number of ELM coils in each toroidal row is nine, the main toroidal mode n = 4 is accompanied by a complementary mode nc = 5. Concerning the resonance of fastion trajectories, the antiresonant surfaces of n = 4 are very close to the resonant surfaces of nc = 5 and vice versa. Since the effect of resonance on fastion trajectories dominates that of antiresonance, a synergy effect of the main and complementary modes effectively enlarges the resonant regions. In a single nmode EMC field, the resonant and antiresonant regions are well separated. The peak heat load due to the loss of NBproduced fast ions near the upper ELM coils is as high as 1.0–1.5 MW m−2, which exceeds the allowable level in ITER. Rotation of the EMC field is essential for ITER to alleviate the local peak heat load. Most loss particles hit the inner side of the torus of the dome in the ITER divertor. The loss of alpha particles is also increased by the effect of the EMC field. The loss is still acceptably low at less than 1.0%.Nuclear Fusion 12/2011; 52(1):013012. DOI:10.1088/00295515/52/1/013012 · 3.06 Impact Factor 
Article: Integrated modelling of steadystate scenarios and heating and current drive mixes for ITER
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ABSTRACT: Recent progress on ITER steadystate (SS) scenario modelling by the ITPAIOS group is reviewed. Codetocode benchmarks as the IOS group's common activities for the two SS scenarios (weak shear scenario and internal transport barrier scenario) are discussed in terms of transport, kinetic profiles, and heating and current drive (CD) sources using various transport codes. Weak magnetic shear scenarios integrate the plasma core and edge by combining a theorybased transport model (GLF23) with scaled experimental boundary profiles. The edge profiles (at normalized radius ρ = 0.8–1.0) are adopted from an edgelocalized modeaveraged analysis of a DIIID ITER demonstration discharge. A fully noninductive SS scenario is achieved with fusion gain Q = 4.3, noninductive fraction fNI = 100%, bootstrap current fraction fBS = 63% and normalized beta βN = 2.7 at plasma current Ip = 8 MA and toroidal field BT = 5.3 T using ITER day1 heating and CD capability. Substantial uncertainties come from outside the radius of setting the boundary conditions (ρ = 0.8). The present simulation assumed that βN (ρ) at the top of the pedestal (ρ = 0.91) is about 25% above the peeling–ballooning threshold. ITER will have a challenge to achieve the boundary, considering different operating conditions (Te/Ti ≈ 1 and density peaking). Overall, the experimentally scaled edge is an optimistic side of the prediction. A number of SS scenarios with different heating and CD mixes in a wide range of conditions were explored by exploiting the weakshear steadystate solution procedure with the GLF23 transport model and the scaled experimental edge. The results are also presented in the operation space for DT neutron power versus stationary burn pulse duration with assumed poloidal flux availability at the beginning of stationary burn, indicating that the long pulse operation goal (3000 s) at Ip = 9 MA is possible. Source calculations in these simulations have been revised for electron cyclotron current drive including parallel momentum conservation effects and for neutral beam current drive with finite orbit and magnetic pitch effects.Nuclear Fusion 08/2011; 51(10):103006. DOI:10.1088/00295515/51/10/103006 · 3.06 Impact Factor  [Show abstract] [Hide abstract]
ABSTRACT: The NEMO (NEutral beam MOdelling) code for simulating neutral beam ionization during neutral beam injection (NBI) in tokamak plasmas has been developed for implementation in integrated modelling frameworks and is presented. Integrated modelling of fusion plasmas is becoming increasingly important, both for preparation and analysis of experiments in large devices. Moreover, it should play a crucial role for the design of future fusion reactors. In a modern context, integrated modelling requires codes that are (i) flexible in terms of their interfaces, i.e. can be adapted for different simulation environments, (ii) machine independent; i.e. they should not contain hard coded information on a particular device to be simulated; (iii) optimized for speed of execution, (iv) verified and validated. The NEMO code has been specially designed to meet these requirements. The code is based on the physics concept outlined by Feng et al (1995 Comput. Phys. Commun. 88 161–72) and is a completely modular program: it works with any input NBI geometry and can be coupled to any external Fokker–Planck calculation for evaluating the distribution function of the injected species, i.e. it can provide source terms for both Monte Carlo codes and codes using finite difference/elements methods. The NEMO code has already been integrated with the CRONOS integrated modelling suite (Artaud et al 2010 Nucl. Fusion 50 043001) and the European Integrated Tokamak Modelling Task Force (ITMTF)*. The basics of the code are described in this paper along with an illustration of its integration in the ITMTF simulation platform. A crucial aspect is the verification of the code, the results of benchmarks carried out with other NBI codes for JET and ITER discharges are thereby presented.Nuclear Fusion 05/2011; 51(6):063019. DOI:10.1088/00295515/51/6/063019 · 3.06 Impact Factor  [Show abstract] [Hide abstract]
ABSTRACT: The possibility of stationary TypeI ELMy Hmode operation in helium plasma is studied with hydrogen NBI and ELM pacing with hydrogen pellets. A possible extension of the operational domain is also discussed.  [Show abstract] [Hide abstract]
ABSTRACT: The final design of the ITER poloidalfield coil system converged after recent design reviews leading up to the start of construction. Significant effort has been spent on modeling details of the baseline 1 5MA inductive discharge and recent exploration of advanced inductive (or hybrid) and steady state scenarios. However, the early operation of ITER will rely on non activation scenarios to validate performance of heating and current drive systems, diagnostics and plasma control before going to full operation. In this paper, we present results from a new effort to define and develop a low activation scenario and demonstrate its performance. 
Article: On the heating mix of ITER
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ABSTRACT: This paper considers the heating mix of ITER for the two main scenarios. Presently, 73MWof absorbed power are foreseen in the mix 20/33/20 for ECH, NBI and ICH. Given a sufficient edge stability, Q = 10—the goal of scenario 2—can be reached with 40MW power irrespective of the heating method but depends sensitively inter alia on the Hmode pedestal temperature, the density profile shape and on the characteristics of impurity transport. ICH preferentially heats the ions and would contribute specifically with ΔQ < 1.5. The success of the Q = 5 steadystate scenario 4 with reduced current requires discharges with improved confinement necessitating weakly or strongly reversed shear, fbs > 0.5, and strong offaxis current drive (CD). The findings presented here are based on revised CD efficiencies γ for ECCD and a detailed benchmark of several CD codes. With ECCD alone, the goals of scenario 4 can hardly be reached. Efficient offaxis CDis only possible with NBI. With beams, inductive discharges with fni > 0.8 can be maintained for 3000 s. The conclusion of this study is that the present heating mix of ITER is appropriate. It provides the necessary actuators to induce in a flexible way the best possible scenarios. The development risks of NBI at 1MeV can be reduced by operation at 0.85MeV.Plasma Physics and Controlled Fusion 11/2010; 52(12). DOI:10.1088/07413335/52/12/124044 · 2.19 Impact Factor  [Show abstract] [Hide abstract]
ABSTRACT: By using a fully three dimensional magnetic field orbitfollowing MonteCarlo code, the energetic ion confinement was investigated for the current conceptual design of the ferromagnetic components in ITER which will be employed for reducing the toroidal magnetic field (TF) ripple. The ferromagnetic insert is effective in the reference standard scenario with Q=10 (Scenario No. 2) and steady state scenario with Q=5 (Scenario No. 4) to improve the energetic ion confinement. Overcompensation appears at half of the full toroidal magnetic field and its effect becomes stronger when the quantity of the ferromagnetic insert is increased in order to more reduce the TF ripple at the full toroidal magnetic field. Though the current design is acceptable, whether to increase the ferromagnetic insert to achieve lower TF ripple amplitude at the full field operation depends on how prospected are possibilities of lower field operations. Planned test blanket modules do not induce large loss (Fusion Engineering and Design 01/2009; 84(1):2432. DOI:10.1016/j.fusengdes.2008.08.040 · 1.15 Impact Factor 
Article: Magnetic fluctuation profile measurement using optics of motional Stark effect diagnostics in JT60U
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ABSTRACT: Motional Stark effect (MSE) diagnostics in JT60U works as polarimeter to measure the pitch angle of magnetic field as well as beamemissionspectroscopy (BES) monochromator simultaneously at 30 spatial channels. Fluctuation in the BES signal using MSE optics (MSE/BES) contains fluctuations in not only the density but also the pitch angle (or the magnetic field). Correlation analysis of the magnetic fluctuation between two spatial channels is applied to highbeta plasma with a magnetohydrodynamic activity at frequency of about 0.9 kHz. It has been found that the magnetic fluctuation measured by the MSE/BES is spatially localized near the magnetic flux surface having safety factor and that the phase of the fluctuation is inverted at about the surface, suggesting magnetic island structure by tearing mode. The phase of the magnetic fluctuation measured by the MSE/BES at outside of the q=2 surface is consistent with that by the pickup coil placed outside the plasma.The Review of scientific instruments 11/2008; 79(10):10F533. DOI:10.1063/1.2965780 · 1.61 Impact Factor  [Show abstract] [Hide abstract]
ABSTRACT: Aiming at optimization of current profile in highβ plasmas for higher confinement and stability, a realtime control system of the minimum of the safety factor (qmin) using the offaxis current drive has been developed. The offaxis current drive can raise the safety factor in the centre and help to avoid instability that limits the performance of the plasma. The system controls the injection power of lowerhybrid waves, and hence its offaxis driven current in order to control qmin. The realtime control of qmin is demonstrated in a highβ plasma, where qmin follows the temporally changing reference qmin,ref from 1.3 to 1.7. Applying the control to another highβ discharge (βN = 1.7, βp = 1.5) with m/n = 2/1 neoclassical tearing mode (NTM), qmin was raised above 2 and the NTM was suppressed. The stored energy increased by 16% with the NTM suppressed, since the resonant rational surface was eliminated. For the future use for current profile control, current density profile for offaxis neutral beam current drive (NBCD) is for the first time measured, using the motional Stark effect diagnostic. Spatially localized NBCD profile was clearly observed at the normalized minor radius ρ of about 0.6–0.8. The location was also confirmed by multichordal neutron emission profile measurement. The total amount of the measured beam driven current was consistent with the theoretical calculation using the ACCOME code. The CD location in the calculation was inward shifted than the measurement.Nuclear Fusion 02/2008; 48(4):045002. DOI:10.1088/00295515/48/4/045002 · 3.06 Impact Factor  [Show abstract] [Hide abstract]
ABSTRACT: Steady state operation is preferable for fusion reactors. The possibility of extending the pulse length in ITER is considered taking into account the capabilities of the planned electroncyclotron current drive (ECCD) and lowhybrid current drive (LHCD). The ECCD efficiency for current drive at different locations is assessed. The possibility of extending the pulse length by the increase in the current drive efficiency due to the synergetic effect for combined ECCD and LHCD at the same location is assessed. The calculated synergetic effect of ECCD and LHCD on the current drive efficiency is less than 10% for ITER parameters. Long pulse operation with the energy multiplication factor Pfus/Paux = Q > 5 and duration t > 3000 s will be possible in the case of enhanced confinement with respect to the ELMy Hmode scaling HH98y,2 ~ 1.3–1.4.Nuclear Fusion 01/2008; 48(1):015002. DOI:10.1088/00295515/48/1/015002 · 3.06 Impact Factor 
Conference Paper: Benchmarking of Neutral Beam Current Drive Codes as a Basis for the Integrated Modeling for ITER
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ABSTRACT: This paper discusses the results of a benchmark study in which the predictions of numerical codes for neutral beam current drive and heating were compared using the parameters of the reference ITER steady state scenario, as a collaboration work of in the frame of the ITPASSO group. The models employed in the benchmarked codes for each physics related to NB heating and current drive, such as the beam model, beam stopping cross section, fast ion solver, orbit effects and electron shielding, are reviewed and examined through comparison.Fusion Energy 2008, International Atomic Energy Agency (2009); 01/2008  [Show abstract] [Hide abstract]
ABSTRACT: Integrated simulations are performed to establish a physics basis, in conjunction with present tokamak experiments, for the operating modes in the International Thermonuclear Experimental Reactor (ITER). Simulations of the hybrid mode are done using both fixed and freeboundary 1.5D transport evolution codes including CRONOS, ONETWO, TSC/TRANSP, TOPICS and ASTRA. The hybrid operating mode is simulated using the GLF23 and CDBM05 energy transport models. The injected powers are limited to the negative ion neutral beam, ion cyclotron and electron cyclotron heating systems. Several plasma parameters and source parameters are specified for the hybrid cases to provide a comparison of 1.5D core transport modelling assumptions, source physics modelling assumptions, as well as numerous peripheral physics modelling. Initial results indicate that very strict guidelines will need to be imposed on the application of GLF23, for example, to make useful comparisons. Some of the variations among the simulations are due to source models which vary widely among the codes used. In addition, there are a number of peripheral physics models that should be examined, some of which include fusion power production, bootstrap current, treatment of fast particles and treatment of impurities. The hybrid simulations project to fusion gains of 5.6–8.3, βN values of 2.1–2.6 and fusion powers ranging from 350 to 500 MW, under the assumptions outlined in section 3. Simulations of the steady state operating mode are done with the same 1.5D transport evolution codes cited above, except the ASTRA code. In these cases the energy transport model is more difficult to prescribe, so that energy confinement models will range from theory based to empirically based. The injected powers include the same sources as used for the hybrid with the possible addition of lower hybrid. The simulations of the steady state mode project to fusion gains of 3.5–7, βN values of 2.3–3.0 and fusion powers of 290 to 415 MW, under the assumptions described in section 4. These simulations will be presented and compared with particular focus on the resulting temperature profiles, source profiles and peripheral physics profiles. The steady state simulations are at an early stage and are focused on developing a range of safety factor profiles with 100% noninductive current.Nuclear Fusion 08/2007; 47(9):1274. DOI:10.1088/00295515/47/9/026 · 3.06 Impact Factor  [Show abstract] [Hide abstract]
ABSTRACT: The multichannel motional Stark effect (MSE) diagnostic system in JT60U has been upgraded to measure density fluctuation profile. A 16channel fastsampling digitizer has been added in order to measure photomultipliertube signals at measurement frequency of 0.5–1 MHz. The new system works as a MSE and beam emission spectroscopy diagnostic. Spatially resolved electron density fluctuation profile measurement in various operation regimes is presented. In the core plasma, density fluctuation induced by rotation of tearing mode islands was observed. Temporal evolution of the fluctuation frequency agrees with that measured by Mirnov coils (poloidal and toroidal mode numbers: 2 and 1, respectively). The phases of the fluctuations on either side of the q = 2 surface are inverted, which is consistent with electron cyclotron emission. These measurements show that the density fluctuation is caused by a rotating magnetic island structure induced by the tearing mode. In the scrapeoff layer of a Hmode plasma with edgelocalizedmode (ELM), i. e., ELMy Hmode outward propagation of strong intermittent emission corresponding to ELM crash was also observed. The propagation velocity is 0.69–2.2 km/s along the MSE measurement points, the time lag and distance between adjacent channels being 67±35 μs and 70 mm, respectively.Review of Scientific Instruments 10/2006; 77(10):10E91410E9144. DOI:10.1063/1.2229264 · 1.61 Impact Factor 
Article: Benchmarking of Lower Hybrid Current Drive Codes with Applications to ITERRelevant Regimes
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ABSTRACT: This paper discusses the results of a benchmark exercise in which the predictions of several simulation models for lower hybrid current drive (LHCD) were compared using parameters typical of the steady state operating scenario in the ITER device (the socalled Scenario 4). The most complete LHCD simulation models that were used combined 2D velocity space Fokker Planck solvers with toroidal ray tracing packages. These models also predicted the highest LHCD efficiencies with 2.0 – 2.6 MA of driven current for 30 MW of coupled LHRF power. Codes that solved the Fokker Planck equation using a Green's function approach and then computed the driven LH current using a wave induced RF flux based on 1D parallel velocity damping along ray trajectories were found to predict LH currents 3540% lower than the 2D Fokker Planck models. This discrepancy is understood in terms of the approximate nature of the 1D wave induced flux that fails to properly capture 2D velocity space effects that occur in LHCD owing to the significant distortion of the electron distribution function. We also used an orbitfollowing Monte Carlo code to study the possible parasitic damping of LH waves on fusion generated alpha particles. The effect of magnetic field ripple and fast ion anomalous transport on the alpha population was considered. It was found that for a large anomalous diffusion coefficient (1m 2 /s), the absorption on fusion alphas can be as high as 7.7% using a LHRF source frequency of 3.7 GHz. This result gives some confidence in the source frequency choice of 5.0 GHz in order to minimize the possibility of this parasitic interaction.  [Show abstract] [Hide abstract]
ABSTRACT: It is found that no current is driven in a central region of a tokamak plasma once the central current density becomes nearly zero ("current hole"), in spite of high electric conductivity, at the current drive by a toroidal electric field and a radiofrequency wave in experiments on the JT60U tokamak. This is a new, stiff, selforganized structure of a magnetic field in an axisymmetric toroidal plasma.Physical Review Letters 09/2005; 95(7):075001. DOI:10.1103/PhysRevLett.95.075001 · 7.51 Impact Factor  [Show abstract] [Hide abstract]
ABSTRACT: Evolution of the current density profile associated with magnetic island formation in a neoclassical tearing mode plasma is measured for the first time in JT60U by using a motional Stark effect diagnostic. As the island grows, the current density profile turns flat at the radial region of the island and a hollow structure appears at the rational surface. As the island shrinks the deformed region becomes narrower and finally diminishes after the disappearance of the island. In a quiescent plasma without magnetohydrodynamic instabilities, on the other hand, no deformation is observed. The observed deformation in the current density profile associated with the tearing mode is reproduced in a time dependent transport simulation assuming the reduction of the bootstrap current in the radial region of the island. Comparison of the measurement with a calculated steadystate solution also shows that the reduction and recovery of the bootstrap current at the island explains the temporal behaviours of the current density and safety factor profiles. From the experimental observation and simulations, we reach the conclusion that the bootstrap current decreases within the island Opoint.Nuclear Fusion 08/2005; 45(9):1101. DOI:10.1088/00295515/45/9/010 · 3.06 Impact Factor  [Show abstract] [Hide abstract]
ABSTRACT: Rapid frequency sweeping modes observed in reversed magnetic shear (RS) plasmas on the Japan Atomic Energy Research Institute Tokamak 60 Upgrade (JT60U) [ H. Ninomiya and the JT60 Team, Phys. Fluids B 4, 2070 (1992) ] have been identified as reversedshearinduced Alfvén eigenmodes (RSAEs), which are ideal magnetohydrodynamic Alfvén eigenmodes (AEs) localized to the region of minimum safety factor, qmin, and are excited by negativeionbased neutral beam injection. The chirping and subsequent saturation of the mode frequency are consistent with theoretical predictions for the transition from RSAEs to toroidal Alfvén eigenmodes (TAEs). The previously observed rapid frequency sweeping modes in ion cyclotron wave heated plasmas in JT60U can also be similarly explained. The observed AE amplitude is largest during the transition from RSAEs to TAEs, and fast ion loss is observed when the AE amplitude is largest at this transition. It is preferable to operate outside the transition range of qmin, e.g., 2.4<qmin<2.7 for the n = 1 AE to avoid substantial fast ion loss in RS plasmas.Physics of Plasmas 08/2005; 12(8):0825090825097. DOI:10.1063/1.1938973 · 2.14 Impact Factor
Publication Stats
2k  Citations  
187.70  Total Impact Points  
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Institutions

19992009

Japan Atomic Energy Agency
 Nuclear Science and Engineering Directorate
Muramatsu, Niigata, Japan
