Scrape-off layer modeling of radiative divertor and high heat flux experiments on DIII-D
We use a new multispecies 1D fluid code, NEWT-ID, to model DIII-D scrape-off layer (SOL) behavior during radiative divertor and high heat flux experiments. The separatrix location and the width of the SOL are uncertain, and affect the comparison of the data in important ways. The model agrees with many of the experimental measurements for a particular prescription for the separatrix location. The model cannot explain the recent data on the separatrix Ti with a conventional picture of ion and electron power flows across the separatrix. Radial transport of particles and heat in some form is required to explain the peak heat flux data before and after gas puffing. For argon puffing in the private flux region, entrainment is poor in the steady state. The calculations suggest that strike point argon puffing in a slot divertor geometry results in substantially better entrainment. Self-consistent, steady-state solutions with radiated powers up to 80% of the SOL power input are obtained in 1D. We discuss significant radial effects which warrant the development of a code which can treat strongly radiating impurities in 2D geometries.
References (13)
- T.W. Petrie et al.
- C.P.C. Wong
- A.H. Futch et al.
- Y.L. Igitkhanov et al.
J. Nucl. Mater.
(1990) - S.L. Allen
- D.N. Hill
Rev. Sci. Instr.
(1990)
Cited by (12)
Carbon influx in He and D plasmas in DIII-D
1999, Journal of Nuclear MaterialsDifferences in the carbon behavior between He and D plasmas during VH-mode, L-mode and L-mode with excess gas puffing are reported and inferences on the importance of the various carbon sources during these modes of operation are discussed. During a VH-mode phase, VUV and visible charge exchange spectroscopy indicates that for both He and D operation the carbon behavior is very similar. In the edge plasma, carbon build up is quite rapid, and the carbon influx represents a large fraction of the total plasma density increase until the termination of the VH phase. During cold divertor operation induced by puffing the primary fueling gas, D and He discharges show a difference in the carbon behavior. The core carbon density is seen to be approximately constant during a D discharge as it transitions from an attached to a cold divertor. However in a He discharge, the core carbon density disappears soon after the cold divertor transition. Arguments are made that the primary carbon source in the ELM free H-mode period is physical sputtering by ion impact at the divertor strike point. In L-mode, both attached and cold divertor, the primary source is from the divertor region and two possibilities for this source are chemical sputtering or charge exchange neutral sputtering. Existing data supports charge exchange neutrals as dominant.
A review on impurity transport in divertors
1997, Journal of Nuclear MaterialsPower exhaust is one of the most crucial requirements for future fusion reactors, like ITER. It is widely recognized that impurity injection is needed to significantly reduce the heat load to the divertor plates. By controlled injection of light impurities, the compatibility of high confinement core plasma with strong radiative divertor has been successfully demonstrated in many tokamaks. However, the core impurity contamination was high (Zeff ∼ 3) compared to the value required for ITER (Zeff < 1.6). Therefore, a scheme for impurity retention in the divertor region should be established for fusion research. This paper reviews the recent progress in experiments and simulations which have been made for the purpose of understanding impurity transport in divertors. The issues contained in the paper are impurity generation, shielding and cross field diffusion. As for the impurity generation, chemical sputtering and wall source are discussed with emphasis on the characteristics of their transport and shielding. Impurity control with plasma flow induced by gas puffing and divertor pumping, and adequately designed divertor geometry is also presented.
An inexact Newton algorithm for solving the tokamak edge plasma fluid equations on a multiply-connected domain
1995, Journal of Computational PhysicsNewton's method is combined with a preconditioned conjugate gradient-like algorithm and finite volume discretization to solve the steady-state two-dimensional tokamak edge plasma fluid equations. A numerical evaluation of the Jacobian is employed. Mesh sequencing, pseudo-transient continuation, and adaptive damping are used to increase the radius of convergence. The computations are performed on a multiply-connected curvilinear geometry in a fully coupled manner. The preconditioned conjugate gradient-like algorithm is shown to have a significant storage advantage over the previously used banded Gaussian elimination, while maintaining the excellent convergence characteristics of the overall algorithm. Simulations of a high recycling divertor and a gaseous divertor on the DIII-D tokamak geometry are used to demonstrate algorithm performance.
Development of a radiative divertor for DIII-D
1995, Journal of Nuclear MaterialsWe 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 τE remains ∼ 2 times ITER-89P scaling. However, ne increases with D2 puffing, and Zeff 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 τE VH-modes. Results from engineering feasibility studies, including diagnostic access, will be presented.
Divertor design for the tokamak physics experiment
1995, Journal of Nuclear MaterialsIn this paper we discuss the divertor design for the planned TPX tokamak, which will explore the physics and technology of steady state (1000 s pulses) heat and particle removal in high confinement (up to 4 × L-mode), high beta (up to βN = 5) divertor plasmas sustained by non-inductive current drive. TPX will operate in the double-null divertor configuration, with actively cooled graphite targets forming a deep (0.57 m) slot at the outer strike point. The peak heat flux on the highly tilted (74° from normal) re-entrant divertor plate (tilted to recycle ions back toward the separatrix) will be in the range of 4–6 MW/m2 with 17.5 MW of auxiliary heating power. The combination of pumping and gas puffing (D2 plus impurities), along with higher heating power (45 MW maximum) will allow testing of radiative divertor concepts at ITER-like power densities.
The effect of ELMs on edge plasma scaling in DIII-D
1992, Journal of Nuclear MaterialsIn this paper we report results of scaling studies aimed at determining how the divertor conditions vary with plasma current, toroidal field, and neutral beam heating power in H-mode discharges with ELMs in the DIII-D tokamak. We find that ELMs produce relatively more direct particle losses (50% or more of the total) than energy losses (≤20%). The time-average peak divertor heat flux in these plasmas is found to scale as d α (PNBIIp)(Bp,mp/Bp,div). The linear power dependence suggests that the plasma sheath at the targets is primarily responsible for limiting the parallel energy flow, while the Ip variation may mean that the radial energy transport in the SOL decreases with increasing plasma current, just as it does in the core.