Session XO4: ITER II

Measuring Error Fields in ITER Before Its First Plasma

It is planned to measure the non-axisymmetric field errors in ITER prior to its first plasma operation with the aims of identifying their sources and enabling their effective correction. ITER presents challenges for such measurements because of its scale, its superconducting coils and its use of ferromagnetic components to reduce the toroidal field ripple. For the toroidal field errors, it is proposed to measure the trajectories of an electron beam launched toroidally from a set of points on a contour in a poloidal plane. The trajectories would be measured at several toroidal locations by photogrammetry using fluorescence produced either on screens or in a dilute gas in the vessel. For each of the poloidal field coils it is proposed to measure the distribution of its field normal to a toroidal surface which lies outside the boundary of any plasma but inside all sources of the applied field. The partial flux loops in ITER which span the inner poloidal circumference of the vacuum vessel at several toroidal locations provide a way to measure the normal field from which it would be possible to determine the field in the plasma region. Analyses are being performed to determine the number of measurements and the accuracy required to resolve the Fourier components of the error fields of interest.   

Error Field Measurement Techniques for ITER Using Plasma Response

The plasma response to external magnetic field asymmetries is a potential tool for detection and correction of the intrinsic error field, but MARS-F modeling and DIII-D data show that the ideal MHD response to error fields is very small in low beta, ohmic plasmas. This indicates that simple proportional feedback control based on a linear plasma response may not be appropriate for error field correction during ITER's initial operational phase. However, modeling and experimental data suggest that the nonlinear response of a tearing mode may be useful for error field measurement and correction at low beta. We discuss several possible approaches, including the onset threshold of an induced tearing mode, open-loop manipulation of a saturated island, and feedback control of the tearing mode. We also assess the possible use of plasma rotation as an error field diagnostic in plasmas with some neutral beam injection.   

Analysis of TBM Error Fields with IPEC

The DIII-D mock-up coils for ITER Test Blanket Modules (TBMs) produce highly localized error fields. The TBM mock-up fields caused Neoclassical Toroidal Viscosity (NTV) damping of toroidal rotation, as expected from their non-resonant nature. However, the TBM mock-up fields also caused direct resonant plasma locking. The plasma locking could be mitigated by n=1 error-field correction consistent with predictions by Ideal Perturbed Equilibrium Code (IPEC). This extreme case demonstrates that the control of the dominant error field is indeed central in error field corrections. In addition, generalized NTV analysis with IPEC showed that the damping in both the core and the edge can be explained by low and high toroidal harmonic perturbations, respectively. This indicates that the control of the dominant error field may also be able to mitigate NTV damping, and remains as an important future experiment. This work was supported by the US DOE Contract {\#}DE-AC02-09CH11466 and {\#}DE-FC02-04ER54682.   

Fast Ion Effects During Test Blanket Module Simulation Experiments in DIII-D

The fast beam-ion confinement in the presence of a scaled mock-up of two Test Blanket Modules (TBM) for ITER was studied in DIII-D. The TBM on \hbox{DIII-D} has four vertically arranged protective carbon tiles with thermocouples placed at the back of each tile. Temperature increases of up to 200$^{\rm o}$C were measured for the two tiles closest to the midplane when the TBM fields were present. These measurements agree qualitatively with results from the full orbit-following beam-ion code, SPIRAL, that predict beam-ion losses to be localized on the central two carbon tiles when the TBM fields present. Within the experimental uncertainties no significant change in the fast-ion population was found in the core of these plasmas which is consistent with SPIRAL analysis. These experiments indicate that the TBM fields do not affect the fast-ion confinement in a harmful way which is good news for ITER.   

Numerical Analysis of Resonant Magnetic Perturbations ELM Control in ITER

We analyze the effect of the RMP coils on the magnetic topology in ITER since this is known to affect the stability of Type-I ELMs. Studies are done for different ITER scenarios (H-mode, steady-state and other) with various coil current amplitudes and phasing. We examine the effect of various RMP configurations on the magnetic topology in ITER using vacuum field modeling. This is done by calculating field line characteristics such as the Chirikov magnetic island overlap parameter and the magnetic field line loss. Results show that we can achieve a level of stochastization equivalent to that in DIII-D RMP ELM suppressed discharges using the nominal ELM control coil currents planned in ITER. We have also included the effect of the ITER Error Field Correction coils in this analysis. The results indicate the robustness of the ITER RMP coils. Our analysis of coil failure shows that the RMP coils will perform well even if current is lost in 10\% of the coils.   

Design Issues for an Intense Neutron Calibration Source for ITER

ITER needs to be able to calibrate its neutron diagnostics, in-situ. We are developing a design for a compact 10$^{11}$ n/sec DD neutron source using multiple deuterium ion beams impacting on a thin-film deuterated liquid propane jet target. This will be substantially higher flux than any commercial non-radioisotope neutron source. The system will be cooled by liquid nitrogen, and the beam source is differentially pumped from the target region, with no intervening windows. It must operate for weeks at a time, without maintenance, inside of ITER, with a minimum of interfering structures. We envision the substantial power ($<$ 100 kW), cooling, and gas feed will enter the ITER vessel from the outside, via umbilical lines. The system should be movable, ideally fitting on the end of an ITER manipulator arm. Possible solutions to the overall design issues will be discussed. This work is supported by the Office of Fusion Energy Sciences, and DOE/LANL contract DE-AC52-06NA25396 and LANL-ITER contract FIA-09-035.   

Electron Cyclotron Emission Diagnostics on ITER

Electron cyclotron emission (ECE) will be employed on ITER to measure the radial profile of electron temperature and non thermal features of the electron distribution as well as measurements of ELMs, magnetic islands, high frequency instabilities, and turbulence. There are two quasioptical systems, designed with Gaussian beam analysis. One view is radial, primarily for temperature profile measurement, the other views at a small angle to radial for measuring non-thermal emission. Radiation is conducted to by a long corrugated waveguide to a multichannel Michelson interferometer which provides wide wavelength coverage but limited time response as well as two microwave radiometers which cover the fundamental and second harmonic ECE and provide excellent time response. Measurements will be made in both X and O mode. In-situ calibration is provided by a novel hot calibration source. We discuss spatial resolution and the implications for physics studies.   

PTRANSP Simulations of ITER Steady-State and Hybrid Discharges

ITER PTRANSP simulations, in which the temperatures, toroidal angular frequency and currents are evolved, are carried out starting with initial profiles obtained from TSC studies. Simulations are carried out using both the new Multi-Mode model and GLF23 transport models to investigate the prediction of plasma rotation and the sensitivity of fusion power production to pedestal height and impurity concentration. It is shown that the time history associated with heating and current drive sources impacts the time dependence of the evolution of the plasma rotation profile and fusion power production. Various mixes of heating (with associated current and torque drives, and fueling) and variations in the parameters associated with these source are considered in order to examine various possibilities and contingencies for ITER steady state and hybrid discharges. It is found, using the Multi-Mode model for a steady state discharge, the fusion power is 185 MW corresponding to a fusion $ Q = 7.1 $, and for a hybrid discharge the fusion power exceeds 400 MW corresponding to $ Q = 8.6 $.   

Time-Dependent Simulations of Alcator C-Mod ITER-like Discharges

Alcator C-Mod is performing ITER-like experiments to benchmark projections to 15 MA ELMy H-mode Inductive ITER discharges. The main focus has been on the transient ramp phases. For most of the experiments presented, the plasma current in C-Mod is 1.3 MA, toroidal field is 5.4 T, aspect ratio is 3.0, and the plasma elongation is 1.75-1.85. The rampup experiments have current rampup times of 500 to 650 ms, which are equivalent to a 100 s rampup in ITER, based on the scaling of $\Delta $t$_{ramp}$/$<$T$_{e}>^{3/2}$a$^{2}$. Ohmic and ion cyclotron (ICRF) heated discharges are examined. The rampdown phase examined in experiments on C-Mod have durations of 350, 600, and 750 ms, equivalent to ITER rampdown times of 70, 120, and 180 s, respectively. Both ELM-free and EDA H-modes have been sustained in the rampdowns. Simulations of the C-Mod discharges have used the Tokamak Simulation Code, augmented with ICRF source deposition from TRANSP. The Coppi-Tang energy transport model is used to provide the best fit to the experimental electron temperature profile. The Bohm/gyro-Bohm and the CDBM energy transport models are also being examined for C-Mod ITER-like discharges. Supported by DE-AC02-09CH11466, DE-FC02-99ER54512.