W.H. Meyer

Experimental vertical stability studies for ITER performance and design guidance

United States Department of Energy (DE-FC02-04ER54698, DEAC52- 07NA27344, and DE-FG02-04ER54235)

A TANGENTIALLY VIEWING VISIBLE TV SYSTEM FOR THE DIII-D DIVERTOR

A video camera system has been installed on the DIII-D tokamak for 2-D spatial studies of line emission in the lower divertor region. The system views the divertor tangentially from an outer port at approximately the height of the X-point. At the tangency plane the entire divertor from inner wall to outside the DIII-D bias ring is viewed with spatial resolution of approximately 1 cm. The image contains information from approximately 90 degrees of toroidal angle. In a recent upgrade, remotely controllable filter changers were added which have produced images from nominally identical shots using a series of spectral lines. Software was developed to calculate the response function matrix using distributed computing techniques and assuming toroidal symmetry. Standard sparse matrix algorithms are then used to invert the 3-D images onto a poloidal plane. Spatial resolution of the inverted images is 2 cm; higher resolution simply increases the size of the response function matrix. Initial results from a series of experiments with multiple identical shots show that the emission from CII and CIII, which appears along the inner scrape-off layer above and below the X-point during ELMing H-mode, moves outward and becomes localized near the X-point in Partially Detached Divertor (PDD) operation.

Development of a radiative divertor for DIII-D

Abstract We have used experiments and modeling to develop a new radiative divertor configuration for DIII-D. Gas puffing experiments with the existing open divertor have shown the creation of a localized (∼ 10 cm diameter) radiation zone which results in substantial reduction (3–10) in the divertor heat flux while τ E remains ∼ 2 times ITER-89P scaling. However, n e increases with D 2 puffing, and Z eff increases with neon puffing. Divertor structures are required to minimize the effects on the core plasma. The UEDGE fluid code, benchmarked with DIII-D data, and the DEGAS neutrals transport code are used to estimate the effectiveness of divertor configurations; slots reduce the core ionization more than baffles. The overall divertor shape is set by confinement studies which indicate that high triangularity (δ ≈ 0.8) is important for high τ E VH-modes. Results from engineering feasibility studies, including diagnostic access, will be presented.

Generation of high power 140 GHz microwaves with an FEL for the MTX experiment

We have used the improved ETA-II linear induction accelerator (ETA-III) and the IMP steady-state wiggler to generate high power (1-2 GW) microwaves at 140 GHz. The FEL was used in an amplifier configuration with a gyrotron driver. Improved control of energy sweep and computerized magnetic alignment in ETA-III resulted in small beam corkscrew motion (<1.5 mm) at 6 Mev, 2.5 kA. Reduction of wiggler errors (<0.2%), improved electron beam matching, and tapered wiggler operation resulted in peak microwave power (single-pulse) of up to 2 GW. These pulses were transported to the MTX tokamak for microwave absorption experiments. In addition, the FEL was run in a burst mode, generating 50-pulse bursts of microwaves; these results are discussed elsewhere.<<ETX>>

Power balance in DIII-D during single-null ELMing H-mode plasmas

Abstract We account for ≥85% of the injected power in DIII-D during single-null diverted ELMing H-mode discharges. Core plasma radiation accounts for ≤18% with the rest flowing through the scrape-off layer to the inboard and outboard divertors. The total power roughly splits in an in: out ratio of 1:1.3. Of the power entering the outboard divertor 75% leaves as direct heat flux to the divertor target plates with the rest in the form of radiation. In the inboard divertor, however, 75% of the power leaves as radiation with the rest as divertor target plate heat flux and charge exchange losses. The in: out ratio remains constant over a range of injected power and plasma current and even as divertor conditions change due to gas puffing.

Burst mode FEL with the ETA-III induction linac

Pulses of 140 GHz microwaves have been produced at a 2 kHz rate using the ETA-III induction linac and IMP wiggler. The accelerator was run in bursts of up to 50 pulses at 6 MeV and greater than 2 kA peak current. A feedback timing control system was used to synchronize acceleration voltage pulses with the electron beam, resulting in sufficient reduction of the corkscrew and energy sweep for efficient FEL operation. Peak microwave power for short bursts was in the range 0.5-1.1 GW, which is comparable to the single-pulse peak power of 0.75-2 GW. FEL bursts of more than 25 pulses were obtained.<<ETX>>

Remote control of alcator C-mod from Lawrence Livermore National Laboratory

Operation ofa tokamak from a remote site has been demonstrated for the first time. The Alcator C-Mod tokamak, located at the Massachusetts Institute of Technology, was operated over the Internet from a remote control room set up at Lawrence Livermore National Laboratory in California. Prescription of the physics parameters such as plasma current, density, shape, heating power, and active diagnostics was accomplished entirely from the remote site using the same interface as when operating from the C-Mod control room. Engineering control of subsystems (e.g., vacuum, cooling, and power supply limits) remained under local control, providing appropriate equipment and personnel security. Although the principal purpose for running this experiment from a distance was to demonstrate the remote operation, it was planned as a productive physics run. The operation was highly successful; important new physics data were obtained, and valuable insight was gained into the potential of remote operation as well as its limitations.

Validation of the thermal transport model used for ITER startup scenario predictions with DIII-D experimental data

Control simulations of ITER startup using 2D free-boundary equilibrium and 1D transport codes rely on accurate predictions of the electron and ion temperature profiles that determine the electrical conductivity and pressure profiles during the current rise. We present results of validation studies that apply the transport model used by the ITER team to DIII-D discharge evolution and compare predictions with data from similarity experiments. Results presented here detail difficulties and sensitivities associated with the modelling of time-dependent current profile evolution required to asses performance of the poloidal-field coil system and controllers on ITER.

Wide-angle ITER-prototype tangential infrared and visible viewing system for DIII-D

An imaging system with a wide-angle tangential view of the full poloidal cross-section of the tokamak in simultaneous infrared and visible light has been installed on DIII-D. The optical train includes three polished stainless steel mirrors in vacuum, which view the tokamak through an aperture in the first mirror, similar to the design concept proposed for ITER. A dichroic beam splitter outside the vacuum separates visible and infrared (IR) light. Spatial calibration is accomplished by warping a CAD-rendered image to align with landmarks in a data image. The IR camera provides scrape-off layer heat flux profile deposition features in diverted and inner-wall-limited plasmas, such as heat flux reduction in pumped radiative divertor shots. Demonstration of the system to date includes observation of fast-ion losses to the outer wall during neutral beam injection, and shows reduced peak wall heat loading with disruption mitigation by injection of a massive gas puff.

Fast ion transport during applied 3D magnetic perturbations on DIII-D

Measurements show fast ion losses correlated with applied three-dimensional (3D) fields in a variety of plasmas ranging from L-mode to resonant magnetic perturbation (RMP) edge localized mode (ELM) suppressed H-mode discharges. In DIII-D L-mode discharges with a slowly rotating magnetic perturbation, scintillator detector loss signals synchronized with the applied fields are observed to decay within one poloidal transit time after beam turn-off indicating they arise predominantly from prompt loss orbits. Full orbit following using M3D-C1 calculations of the perturbed fields and kinetic profiles reproduce many features of the measured losses and points to the importance of the applied 3D field phase with respect to the beam injection location in determining the overall impact on prompt beam ion loss. Modeling of these results includes a self-consistent calculation of the 3D perturbed beam ion birth profiles and scrape-off-layer ionization, a factor found to be essential to reproducing the experimental measurements. Extension of the simulations to full slowing down timescales, including fueling and the effects of drag and pitch angle scattering, show the applied RMPs in ELM suppressed H-mode plasmas can induce a significant loss of energetic particles from the core. With the applied fields, up to 8.4% of the injected beam power is predicted to be lost, compared to 2.7% with axisymmetric fields only. These fast ions, originating from minor radii , are predicted to be primarily passing particles lost to the divertor region, consistent with wide field-of-view infrared periscope measurements of wall heating in RMP ELM suppressed plasmas. Edge fast ion (FIDA) measurements also confirm a large change in edge fast ion profile due to the fields, where the effect was isolated by using short 50 ms RMP-off periods during which ELM suppression was maintained yet the fast ion profile was allowed to recover. The role of resonances between fast ion drift motion and the applied 3D fields in the context of selectively targeting regions of fast ion phase space is also discussed.