Engineering topics

Schaffer, M J

Publications and source records attributed to Schaffer, M J.

8 records

DOE OSTI2012

The limits and challenges of error field correction for ITER

Significant progress has been made in interpreting the effects of non-axisymmetric 'error' fields on a plasma through ideal MHD stability and a dominant 'least stable' ideal mode through which the fields couple to the tearing resonant surface. However, in contrast to expectations from such theories, experiments have found limited success in correcting error fields, with single correction coil arrays giving benefits of between 0% and {approx}50% correction (in terms of improvement to a low density locked mode limit), dependent on the structure of the error and correcting fields. With additional coils up to {approx}70% is possible. It was unclear whether this represented an intrinsic stability or control limit, or higher order toroidal or poloidal harmonic effects. Thus, studies on the DIII-D tokamak explored correction of a proxy error field, using two differently structured coil arrays. This enabled the principles of error correction to be tested at high amplitudes and operational densities, with known pure n=1 fields. Results showed substantial residual effects from the corrected n=1 field, with improvements of only {approx}50% in the low density locked mode limit. This suggests that n=1 error fields must couple to more than one surface in the plasma, and this is conjectured to be through more than one ideal mode, thereby requiring precise correction. For ITER, updated predictions of field error have been obtained and compared with revised scalings for tearing mode thresholds, indicating 50% or better error field correction will be needed. This will likely require more than one well coupled correction coil array and sets a challenge for theory to model the behavior, in order to clarify the plasma response and braking mechanisms, and so the effectiveness of ITER's correction coils and the possible need for support from its edge localized mode control coils.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2012

Reactor-relevant quiescent H-mode operation using torque from non-axisymmetric, non-resonant magnetic fields

Results from recent experiments demonstrate that quiescent H-mode (QH-mode) sustained by magnetic torque from non-axisymmetric magnetic fields is a promising operating mode for future burning plasmas. Using magnetic torque from n=3 fields to replace counter-I{sub p} torque from neutral beam injection (NBI), we have achieved long duration, counter-rotating QH-mode operation with NBI torque ranging from counter-I{sub p} to up to co-I{sub p} values of 1-1.3 Nm. This co-I{sub p} torque is 3 to 4 times the scaled torque that ITER will have. These experiments utilized an ITER-relevant lower single-null plasma shape and were done with ITER-relevant values of {nu}{sub ped}{sup *} and {beta}{sub N}{sup ped}. These discharges exhibited confinement quality H{sub 98y2}=1.3, in the range required for ITER. In preliminary experiments using n=3 fields only from a coil outside the toroidal coil, QH-mode plasmas with low q{sub 95}=3.4 have reached fusion gain values of G={beta}{sub N}H{sub 89}/q{sub 95}{sup 2}=0.4, which is the desired value for ITER. Shots with the same coil configuration also operated with net zero NBI torque. The limits on G and co-I{sub p} torque have not yet been established for this coil configuration. QH-mode work to has made significant contact with theory. The importance of edge rotational shear is consistent with peeling-ballooning mode theory. Qualitative and quantitative agreements with the predicted neoclassical toroidal viscosity torque is seen.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2011

Three-dimensional magnetohydrodynamics simulations of counter-helicity spheromak merging in the Swarthmore Spheromak Experiment

Recent counter-helicity spheromak merging experiments in the Swarthmore Spheromak Experiment (SSX) have produced a novel compact torus (CT) with unusual features. These include a persistent antisymmetric toroidal magnetic field profile and a slow, nonlinear emergence of the n = 1 tilt mode. Experimental measurements are inconclusive as to whether this unique CT is a fully merged field-reversed configuration (FRC) with strong toroidal field or a partially merged ''doublet CT'' configuration with both spheromak- and FRC-like characteristics. In this paper, the SSX merging process is studied in detail using three-dimensional resistive MHD simulations from the Hybrid Magnetohydrodynamics (HYM) code. These simulations show that merging plasmas in the SSX parameter regime only partially reconnect, leaving behind a doublet CT rather than an FRC. Through direct comparisons, we show that the magnetic structure in the simulations is highly consistent with the SSX experimental observations. We also find that the n = 1 tilt mode begins as a fast growing linear mode that evolves into a slower-growing nonlinear mode before being detected experimentally. A simulation parameter scan over resistivity, viscosity, and line-tying shows that these parameters can strongly affect the behavior of both the merging process and the tilt mode. In fact, merging in certain parameter regimes is found to produce a toroidal-field-free FRC rather than a doublet CT.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2011

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

Fast beam-ion losses were studied in DIII-D in the presence of a scaled mockup of two Test Blanket Modules (TBM) for ITER. Heating of the protective tiles on the front of the TBM surface was found when neutral beams were injected and the TBM fields were engaged. The fast-ion core confinement was not significantly affected. Different orbit-following codes predict the formation of a hot spot on the TBM surface arising from beam-ions deposited near the edge of the plasma. The codes are in good agreement with each other on the total power deposited at the hot spot predicting an increase in power with decreasing separation between the plasma edge and the TBM surface. A thermal analysis of the heat flow through the tiles shows that the simulated power can account for the measured tile temperature rise. The thermal analysis, however, is very sensitive to the details of the localization of the hot spot which is predicted to be different among the various codes.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2011

Measurement and modeling of three-dimensional equilibria in DIII-D

A detailed experiment-theory comparison reveals that linear ideal MHD theory is in quantitative agreement with external magnetic and internal soft x-ray measurements of the plasma response to externally applied non-axisymmetric fields over a broad range of beta and rotation. This result represents a significant step toward the goal of advancing the understanding of three-dimensional tokamak equilibria. Both the magnetic and soft x-ray measurements show the driven plasma perturbation increases linearly with the applied perturbation, suggesting the relevance of linear plasma response models. The magnetic and soft x-ray measurements are made at multiple toroidal and poloidal locations, allowing well resolved measurements of the global structure. The comparison also highlights the need to include kinetic effects in the MHD model once beta exceeds 80% of the kink mode limit without a conducting wall. Two distinct types of response fields are identified by the linear ideal MHD model: one that consists of localized currents at the rational surfaces that cancel the applied resonant field and another that is excited by the components of the external field that couple to the kink mode. Numerical simulations show these two fields have similar amplitudes in ITER-shaped DIII-D discharges where n = 3 fields are used to suppress edge localized modes.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2011

ITER Test Blanket Module Error Field Simulation Experiments at DIII-D

Experiments at DIII-D investigated the effects of magnetic error fields similar to those expected from proposed ITER test blanket modules (TBMs) containing ferromagnetic material. Studied were effects on: plasma rotation and locking, confinement, L-H transition, the H-mode pedestal, edge localized modes (ELMs) and ELM suppression by resonant magnetic perturbations, energetic particle losses, and more. The experiments used a purpose-built three-coil mock-up of two magnetized ITER TBMs in one ITER equatorial port. The largest effect was a reduction in plasma toroidal rotation velocity v across the entire radial profile by as much as Delta upsilon/upsilon similar to 60% via non-resonant braking. Changes to global Delta n/n, Delta beta/beta and Delta H(98)/H(98) were similar to 3 times smaller. These effects are stronger at higher beta. Other effects were smaller. The TBM field increased sensitivity to locking by an applied known n = 1 test field in both L-and H-mode plasmas. Locked mode tolerance was completely restored in L-mode by re-adjusting the DIII-D n = 1 error field compensation system. Numerical modelling by IPEC reproduces the rotation braking and locking semi-quantitatively, and identifies plasma amplification of a few n = 1 Fourier harmonics as the main cause of braking. IPEC predicts that TBM braking in H-mode may be reduced by n = 1 control. Although extrapolation from DIII-D to ITER is still an open issue, these experiments suggest that a TBM-like error field will produce only a few potentially troublesome problems, and that they might be made acceptably small.

Schaffer, M J↗

DOE OSTI2011

Fast-ion effects during test blanket module simulation experiments in DIII-D

Fast beam-ion losses were studied in DIII-D in the presence of a scaled mock-up of two test blanket modules (TBM) for ITER. Heating of the protective tiles on the front of the TBM surface was found when neutral beams were injected and the TBM fields were engaged. The fast-ion core confinement was not significantly affected. Different orbit-following codes predict the formation of a hot spot on the TBM surface arising from beam ions deposited near the edge of the plasma. The codes are in good agreement with each other on the total power deposited at the hot spot, predicting an increase in power with decreasing separation between the plasma edge and the TBM surface. A thermal analysis of the heat flow through the tiles shows that the simulated power can account for the measured tile temperature rise. The thermal analysis, however, is very sensitive to the details of the localization of the hot spot, which is predicted to be different among the various codes.

Kramer, G↗

DOE OSTI2011

Toroidally Resolved Structure of Divertor Heat Flux in RMP H-mode Discharges on DIII-D

As shown on DIII-D edge localized modes (ELMs) can be either completely eliminated or mitigated with resonant magnetic perturbation (RMP) fields. Two infrared cameras, separated 105 degrees toroidally, were used to make simultaneous measurements of ELM heat loads with high frame rates. Without the RMP fields ELMs display a variety of different heat load dynamics and a range of toroidal variability that is characteristic of their 3D structure. Comparing radial averages there is no asymmetry between two toroidal locations. With RMP-mitigated ELMs, the variability in the radially averaged power loads is significantly reduced and toroidal asymmetries in power loads are introduced. In addition to RMP ELM suppression scenarios an RMP scenario with only very small ELMs and very good confinement has been achieved.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗