Progress on coupling UEDGE and Monte-Carlo simulation codes

Our objective is to develop an accurate self-consistent model for plasma and neutral sin the edge of tokamak devices such as DIII-D and ITER. The tow-dimensional fluid model in the UEDGE code has been used successfully for simulating a wide range of experimental plasma conditions. However, when the neutral mean free path exceeds the gradient scale length of the background plasma, the validity of the diffusive and inertial fluid models in UEDGE is questionable. In the long mean free path regime, neutrals can be accurately and efficiently described by a Monte Carlo neutrals model. Coupling of the fluid plasma model in UEDGE with a Monte Carlo neutrals model should improve the accuracy of our edge plasma simulations. The results described here used the EIRENE Monte Carlo neutrals code, but since information is passed to and from the UEDGE plasma code via formatted test files, any similar neutrals code such as DEGAS2 or NIMBUS could, in principle, be used

Scrape-Off Layer Plasmas for ITER with 2nd X-Point and Convective Transport Effects

Plasma fluxes to the divertor region in ITER near the magnetic separatrix have been modeled extensively in the past. The smaller, but potentially very important fluxes to the main chamber and outer divertor regions are the focus of the present paper. Two main additions to the usual transport modeling are investigated: namely, convective radial transport from intermittent, rapidly propagating ''blob'' events, and inclusion of the magnetic flux-surface region beyond the second X-point that actually contacts the main-chamber wall. The two-dimensional fluid transport code UEDGE is use to model the plasma, while the energy spectrum of charge-exchange neutrals to the main chamber wall is calculated by DEGAS 2 Monte Carlo code. Additionally, the spatial distribution of Be sputtered from the main chamber wall is determined in the fluid limit

Simulation of Main-Chamber Recycling in DIII-D with the UEDGE Code

This report demonstrates a computer simulation model for single-null diverted plasma configurations that include simultaneous interaction of the scrape-off layer (SOL) plasma with toroidally symmetric main-chamber limiter surfaces and divertor plate surfaces. The simulations use the UEDGE code which treats the SOL plasma and recycled neutrals as two-dimensional toroidally symmetric fluids. The spatial domain can include field lines that intersect main chamber surfaces in the far scrape-off layer, which allows the model to include simultaneous plasma contact with both divertor and main chamber targets. Steady-state simulation results for low-density L-mode plasma discharges in DIII-D show that total core fueling increases by about 70 percent when the separatrix-baffle gap is reduced from 6 cm to 3 cm. The additional core fueling is due to neutrals which originate from the ion particle flux incident on the upper outer divertor baffle

Pulsed lower-hybrid wave penetration in reactor plasmas

Providing lower-hybrid power in short, intense (GW) pulses allows enhanced wave penetration in reactor-grade plasmas. We examine nonlinear absorption, ray propagation, and parametric instability of the intense pulses. We find that simultaneously achieving good penetration while avoiding parametric instabilities is possible, but imposes restrictions on the peak power density, pulse duration, and/or rf spot shape. In particular, power launched in narrow strips, elongated along the field direction, is desired. 4 refs., 4 figs

Role of Poloidal E×B\mathbf{E}\times\mathbf{B} Drift in Divertor Heat Transport in DIII-D

Simulations for DIII-D high confinement mode plasmas with the multifluid code UEDGE show a strong role of poloidal $\mathbf{E}\times\mathbf{B}$ drifts on divertor heat transport, challenging the paradigm of conduction limited scrape-off layer (SOL) transport. While simulations with reduced drift magnitude are well aligned with the assumption that electron heat conduction dominates the SOL heat transport, simulations with drifts predict that the poloidal convective $\mathbf{E}\times\mathbf{B}$ heat transport dominates over electron heat conduction in both attached and detached conditions. Since poloidal $\mathbf{E}\times\mathbf{B}$ flow propagates across magnetic field lines, poloidal transport with shallow magnetic pitch angles can reach values that are of the same order as would be provided by sonic flows parallel to the field lines. These flows can lead to strongly convection dominated divertor heat transport, increasing the poloidal volume of radiative power front, consistent with previous measurements at DIII-D. Due to these convective flows, the Lengyel integral approach, assuming zero convective fraction, is expected to provide a pessimistic estimate for radiative capability of impurities in the divertor. For the DIII-D simulations shown here, the Lengyel integral approach underestimates the radiated power by a factor of 6, indicating that for reliable DIII-D divertor power exhaust predictions, full 2D calculations, including drifts, would be necessary.Comment: Paper submitted into the Contributions to Plasma Physics in the special issue of the 17th International Workshop on Plasma Edge Theory in Fusion Device

Implementation of the GTNEUT 2D Neutrals Transport Code for Routine DIII-D Analyses

The Georgia Tech Neutral Transport (GTNEUT) code is being implemented to provide a tool for routine analysis of the effects of neutral atoms on edge phenomena in DIII-D. GTNEUT can use an arbitrarily complex two-dimensional grid to represent the plasma edge geometry. The grid generation capability built into the UEDGE code, which utilizes equilibrium fitting data taken from experiment, is being adapted to produce geometric grids for the complex 2D geometries in the DIII-D plasma edge. The process for using experimental measurements supplemented by plasma edge calculations to provide the required background plasma parameters for the GTNEUT calculation will be systematized once the geometric grid generation is complete

Transport of Dust Particles in Tokamak Devices

Recent advances in the dust transport modeling in tokamak devices are discussed. Topics include: (1) physical model for dust transport; (2) modeling results on dynamics of dust particles in plasma; (3) conditions necessary for particle growth in plasma; (4) dust spreading over the tokamak; (5) density profiles for dust particles and impurity atoms associated with dust ablation in tokamak plasma; and (6) roles of dust in material/tritium migration

Simulations of carbon sputtering in fusion reactor divertor plates

The interaction of edge plasma with material surfaces raises key issues for the viability of the International Thermonuclear Reactor (ITER) and future fusion reactors, including heat-flux limits, net material erosion, and impurity production. After exposure of the graphite divertor plate to the plasma in a fusion device, an amorphous C/H layer forms. This layer contains 20-30 atomic percent D/T bonded to C. Subsequent D/T impingement on this layer produces a variety of hydrocarbons that are sputtered back into the sheath region. We present molecular dynamics (MD) simulations of D/T impacts on amorphous carbon layer as a function of ion energy and orientation, using the AIREBO potential. In particular, energies are varied between 10 and 150 eV to transition from chemical to physical sputtering. These results are used to quantify yield, hydrocarbon composition and eventual plasma contamination

Models and applications of the UEDGE code

The transport of particles and energy from the core of a tokamak to nearby material surfaces is an important problem for understanding present experiments and for designing reactor-grade devices. A number of fluid transport codes have been developed to model the plasma in the edge and scrape-off layer (SOL) regions. This report will focus on recent model improvements and illustrative results from the UEDGE code. Some geometric and mesh considerations are introduced, followed by a general description of the plasma and neutral fluid models. A few comments on computational issues are given and then two important applications are illustrated concerning benchmarking and the ITER radiative divertor. Finally, we report on some recent work to improve the models in UEDGE by coupling to a Monte Carlo neutrals code and by utilizing an adaptive grid

A novel flexible field-aligned coordinate system for tokamak edge plasma simulation

Tokamak plasmas are confined by a magnetic field that limits the particle and heat transport perpendicular to the field. Parallel to the field the ionised particles can move freely, so to obtain confinement the field lines are “closed” (ie.form closed surfaces of constant poloidal flux) in the core of a tokamak. Towards, the edge, however, the field lines intersect physical surfaces, leading to interaction between neutral and ionised particles, and the potential melting of the material surface. Simulation of this interaction is important for predicting the performance and lifetime of future tokamak devices such as ITER. Field-aligned coordinates are commonly used in the simulation of tokamak plasmas due to the geometry and magnetic topology of the system. However, these coordinates are limited in the geometry they allow in the poloidal plane due to orthogonality requirements. A novel 3D coordinate system is proposed herein that relaxes this constraint so that any arbitrary, smoothly varying geometry can be matched in the poloidal plane while maintaining a field-aligned coordinate. This system is implemented in BOUT++ and tested for accuracy using the method of manufactured solutions. A MAST edge cross-section is simulated using a fluid plasma model and the results show expected behaviour for density, temperature, and velocity. Finally, simulations of an isolated divertor leg are conducted with and without neutrals to demonstrate the ion-neutral interaction near the divertor plate and the corresponding beneficial decrease in plasma temperature