Bulletin of the American Physical Society
2005 47th Annual Meeting of the Division of Plasma Physics
Monday–Friday, October 24–28, 2005; Denver, Colorado
Session FI1: The Road to Burning Plasmas |
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Chair: Stephen Wolfe, Massachusetts Institute of Technology Room: Adam's Mark Hotel Plaza Ballroom ABC |
Tuesday, October 25, 2005 9:30AM - 10:00AM |
FI1.00001: The Physics of Edge Resonant Magnetic Perturbation in Hot Tokamak Plasmas Invited Speaker: Resonant magnetic perturbation (RMP) experiments in \hbox{DIII-D} have demonstrated that the dynamics of pedestal plasmas, especially the coupling between transport and stability, are strongly modified by small changes in the edge magnetic topology. Edge localized RMPs alter the plasma rotation, $E_r$ well, and ELM behavior. Large Type~I ELMs have been completely eliminated at ITER relevant collisionalities, $\nu_e*<$0.1, with $dB_r(11,3)/B\sim$3$\times$10$^{-4}$ and a broad ($\sim$14\% in normalized poloidal magnetic flux $\psi_N$) vacuum stochastic layer, resulting in an externally controlled quiescent H-mode (ECQH-mode). During ECQH-modes, changes in the edge particle and energy confinement compared to an ELMing H-mode contradict theoretical expectations. In contrast, at $\nu_e^*\sim$1 with $dB_r(11,3)/B$ about equal to the intrinsic field-errors (1$\times$10$^{-5}$), large Type~I ELMs are suppressed without altering the confinement, pedestal profiles, or \hbox{H-mode} $E_r$ well by increasing the frequency and amplitude of small intermittent transport events. This level of RMP produces isolated island chains spanning the pedestal, combined with a thin ($\sim$3\% in $\psi_N$) vacuum stochastic separatrix region. Detailed calculations of the edge magnetic topology reveal the formation of a complex web of homoclinic tangles during the RMPs. Heat and particle flux measurements confirm the existence of these tangles and demonstrate a clear correlation between the RMPs and tangle formation. These results demonstrate that relatively small RMPs can be used to control the dynamics of the pedestal over a wide range of parameters and that they may provide an attractive ELM control option for future devices such as ITER with low $\nu_e^*$ pedestals. [Preview Abstract] |
Tuesday, October 25, 2005 10:00AM - 10:30AM |
FI1.00002: The effect of shaping on plasma performance on the National Spherical Torus Experiment Invited Speaker: NSTX has explored the effects of strong shaping on plasma performance as determined by many diverse topics such as: global MHD modes (e.g. ideal external kinks and resistive wall modes), Edge Localized Modes (ELMs), bootstrap current drive, divertor flux expansion, and heat transport. Precise plasma shape control has been achieved on NSTX using real-time equilibrium reconstruction. NSTX has simultaneously achieved $\kappa \sim 2.8$ and $\delta \sim 0.8$. Ideal MHD theory predicts increased stability at high values of shaping factor $S = q_{95} [ I_p / a B_t]$, which has been observed at large values of $S \sim 34$ on NSTX. The relationship between shape and stability is examined in detail. Improved shaping capability has been crucial to achieving $\beta_t \sim 40 \% $. ELM behavior is observed to depend on plasma shape factor. A description of the ELM regimes attained as shape is varied will be presented. High $\kappa$ is predicted to increase the bootstrap fraction at fixed $I_p$. The achievement of high $\kappa$, as well as H-mode triggered early in the current ramp, has enabled operation with 1s pulses with $I_p =$ 1MA. Detailed analysis of the noninductive current fraction as well as empirical analysis of the achievable plasma pulse length as elongation is varied will be presented. Data is presented showing a reduction in peak divertor heat load due to increasing in flux expansion. Tokamak global transport scaling relations (e.g ITER89P, ITERH98) indicate that confinement increases with increasing $\kappa$. Comparisons are made between these scaling relations and confinement on NSTX. [Preview Abstract] |
Tuesday, October 25, 2005 10:30AM - 11:00AM |
FI1.00003: Operation of Alcator C-Mod with high-Z plasma facing components and implications Invited Speaker: High-Z Plasma Facing Components (PFCs) are likely necessary for a tokamak reactor due to their low tritium (T) retention, capability to handle high heat fluxes with low erosion, and robustness to nuclear damage and activation. ITER is considering using all high-Z PFCs to reduce the T retention projected from current carbon PFC experiments. Recent C-Mod experiments, utilizing molybdenum PFCs, provide unique experience regarding the effect of high-Z PFCs on: 1) plasma performance; 2) necessity of a low-Z wall-coating (boronization); {\&} 3) hydrogenic retention. After boron was removed from vessel {\&} molybdenum PFC surfaces RF-heated H-modes were readily achieved although the resultant enhancement in energy confinement was small (H89 $\sim $ 1). Particle confinement was `good,' causing core Mo radiation to rapidly rise after the H-mode transition, cooling the plasma, reducing confinement and/or causing a back H/L transition. Ohmic H-modes had better confinement (H89 $\sim $ 1.5). Post-boronization the situation was changed; Mo sources and core levels were reduced $\sim $ x10 with H89 reaching 2. Under these conditions a world-record volume-average plasma pressure of 1.8 atmospheres at 5.4 T was achieved at the ITER $\beta _{N}$. The positive effects of boronization are found to last a limited time, correlated with the input energy. Intra- and inter-shot boronization techniques were developed with the latter being the most successful. Wall fuel retention was significant (up to 50{\%} of D$_{2}$ pulsed in) both for boronized and un-boronized PFCs. Scaling fuel retention to an ITER-size device gives of order 50g/pulse. Planned, localized disruptions were developed to thermally desorb the retained H/D from PFC surfaces. This initial comparison indicates that high-Z operation, without boronization, carries some risk for poor confinement performance and implies that boronization (or other low-Z wall coating), not presently planned for ITER, might be required for high-Z PFCs. With or without boronization, the H/D retention could be large; but disruptive techniques to remove the D/T show promise. [Preview Abstract] |
Tuesday, October 25, 2005 11:00AM - 11:30AM |
FI1.00004: Thirty Minutes Plasma Sustainment by Real Time Magnetic Axis Swing for Effective Divertor Load Dispersion in the Large Helical Device Invited Speaker: Achieving steady-state plasma operation at high plasma temperature is one of the important goals of the world-wide magnetic fusion research. We report here a successful high temperature $\sim $ 2 keV, steady-state plasma sustainment operation on the Large Helical Device (LHD) where the high temperature plasmas were created and maintained for more than 30 minutes with the world record 1.3 GJ of auxiliary heating power injection. By using the magnetic axis swing technique developed on LHD, the heat load to the divertor plates was effectively dispersed. The heat load along the divertor heat exhaust region was largely redistributed and the divertor tile temperatures maintained at acceptable levels by sweeping the magnetic axis position by only 3 cm (R $\sim $ 3.66 - 3.69m) in heliotron configuration and the experimental observation of heat load was well explained by modeling calculation. The LHD steady-stead plasma was mainly heated and sustained by the hydrogen minority heating at Ion cyclotron Range of Frequencies (ICRF) heating while additional Electron Cyclotron (EC) and Neutral Beam Injection (NBI) heating methods were also used. The sustained central ion temperature was around 2 keV or higher, and the line-averaged electron density was around 0.7$\sim $0.8 x 10$^{19}$ m$^{-3}$. The average input power was 680 kW and the plasma duration was 31 min 45 sec, and the total input energy to the plasma reached 1.3 GJ. An innovative liquid stub tuner was developed enabling the real time antenna coupling control. This successful long operation shows that the heliotron configuration has a high potential as a steady-state fusion reactor. [Preview Abstract] |
Tuesday, October 25, 2005 11:30AM - 12:00PM |
FI1.00005: Advanced Tokamak Research in Long Time Scales on JT-60U Invited Speaker: In order to realize a steady-state tokamak fusion reactor, sustainment of advanced tokamak plasmas with a large fraction of bootstrap current ($f_{BS}$) and high normalized beta ($ \beta_N$) is required. Important characteristic time scales to be considered are the current diffusion time ($\tau_R$) and the wall saturation time ($\tau_{wall}$). Recent JT-60U experiments extended the duration of advanced tokamak plasmas longer than $ \tau_R$ and approaching $\tau_{wall}$, which enabled us to study control issues in long time scales. In a reversed shear plasma, a very high $f_{BS}$ of 75\% was maintained for 7.4 s ($2.7 \tau_R$) together with very high confinement $HH_{y2} \sim 1.7$ at $\beta_N \sim 1.7$ and $q_{95} \sim 8.6$. Stationary conditions in the current and pressure profiles have been obtained for the first time in a bootstrap-current-dominant plasma relevant to the steady-state tokamak reactor. In a high $\beta_p$ H-mode plasma with N-NB injection, a weak shear $q$ profile with $q_{min} \sim 1.5$, $q_{95} \sim 4.5$, $f_{BS} \sim 43$-50\% and $\beta_N$ of 2.4 was successfully maintained for 5.8 s ($2.8 \tau_R$) under nearly full non-inductive current drive condition, approaching to requirements for the ITER steady-state operation scenario. In a high $\beta_p$ H-mode plasma with $q(0)$ close to 1, $\beta_N$ of 2.5 and $f_{BS}$ of $\sim 30$-35\% have been successfully maintained for 15.5 s ($ \sim 9.5 \tau_R$) at $q_{95} \sim 3.4$ with extended pulse length of NB. The figure of merit of fusion performance $ \beta_{N} H_{89}/q_{95}^2$ was kept 0.4-0.5. The current profile reached a stationary state with $q(0) \sim 1$ without appearance of sawteeth or neoclassical tearing modes. A slight decrease in energy confinement was observed in a later phase, which can be attributed to the increase in the particle recycling and the plasma density, suggesting importance of particle control in long pulse plasmas. The divertor pumping was effective for density control under the saturation of wall inventory in repeated long pulse ($\sim 30$ s) ELMy H-mode discharges. [Preview Abstract] |
Tuesday, October 25, 2005 12:00PM - 12:30PM |
FI1.00006: JET helps prepare for ITER operation Invited Speaker: The main focus of the JET programme (2006-10) in preparation of ITER operation is a new ITER-like ICRH antenna (total RF power increased to $\sim $15MW), a new ITER-like first wall (beryllium in the main chamber, tungsten in the divertor, and possibly CFC at the strike points), upgraded NB power (to 35MW/20s or 17.5MW/10s), and an improved diagnostic and control capability. Mass flows for ITER Scenarios with the ITER-like first wall will be optimised, particularly to minimise in-vessel tritium inventory, since this must be controlled strictly in ITER and has been shown on JET with a carbon first wall to depend sensitively on plasma conditions. Higher power will allow confinement scalings to be resolved for normalised parameters closer to ITER (beta dependence of ELMy H-modes, confinement of improved H-modes at low $\rho $*) and offers the prospect of high beta operation at high current and density, and new fully non-inductive, high performance, ITB discharges sustained to long pulse by real time current and pressure profile control, particularly in bootstrap current dominated regimes. Together, the first wall and increased heating power place strict constraints on the optimisation of ITER scenarios for long pulse operation with low melt damage. Large ELMs (in excess of 1MJ; marginally accessible on JET at present) and disruptions could cause melt severe damage which must be studied and controlled. The testing and optimisation of techniques for ELM mitigation (impurity seeding, demonstrated on JET; use of a new high frequency pellet injector (10-60Hz) to prevent large ELMs, demonstrated on ASDEX Upgrade) and disruption mitigation (fast gas injection from a new disruption mitigation valve, demonstrated on DIII-D) will be even more relevant under the ITER-like edge plasma conditions accessible with the increased power. Acknowledgement : Contributors to EFDA-JET Workprogramme [Preview Abstract] |
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