P.C. Stangeby

The Plasma Boundary of Magnetic Fusion Devices

Part 1: An introduction to the subject of the plasma boundary. Part 2: Introduction to fluid modelling of the boundary plasma. Part 3: Plasma boundary research.

Chapter 4: Power and particle control

Progress, since the ITER Physics Basis publication (ITER Physics Basis Editors et al 1999 Nucl. Fusion 39 2137–2664), in understanding the processes that will determine the properties of the plasma edge and its interaction with material elements in ITER is described. Experimental areas where significant progress has taken place are energy transport in the scrape-off layer (SOL) in particular of the anomalous transport scaling, particle transport in the SOL that plays a major role in the interaction of diverted plasmas with the main-chamber material elements, edge localized mode (ELM) energy deposition on material elements and the transport mechanism for the ELM energy from the main plasma to the plasma facing components, the physics of plasma detachment and neutral dynamics including the edge density profile structure and the control of plasma particle content and He removal, the erosion of low- and high-Z materials in fusion devices, their transport to the core plasma and their migration at the plasma edge including the formation of mixed materials, the processes determining the size and location of the retention of tritium in fusion devices and methods to remove it and the processes determining the efficiency of the various fuelling methods as well as their development towards the ITER requirements. This experimental progress has been accompanied by the development of modelling tools for the physical processes at the edge plasma and plasma–materials interaction and the further validation of these models by comparing their predictions with the new experimental results. Progress in the modelling development and validation has been mostly concentrated in the following areas: refinement in the predictions for ITER with plasma edge modelling codes by inclusion of detailed geometrical features of the divertor and the introduction of physical effects, which can play a major role in determining the divertor parameters at the divertor for ITER conditions such as hydrogen radiation transport and neutral–neutral collisions, modelling of the ion orbits at the plasma edge, which can play a role in determining power deposition at the divertor target, models for plasma–materials and plasma dynamics interaction during ELMs and disruptions, models for the transport of impurities at the plasma edge to describe the core contamination by impurities and the migration of eroded materials at the edge plasma and its associated tritium retention and models for the turbulent processes that determine the anomalous transport of energy and particles across the SOL. The implications for the expected performance of the reference regimes in ITER, the operation of the ITER device and the lifetime of the plasma facing materials are discussed.

Plasma boundary phenomena in tokamaks

Effects of the plasma boundary can have a substantial influence on the behaviour of the entire plasma in tokamaks. Progress in the field, particularly that over the last decade, is reviewed, with emphasis on experimental observation. Simple modelling for interpretation is also included.

Erosion/deposition issues at JET

Deposition and H-isotope retention in JET is highly asymmetric, with deposition predominantly in the inner divertor, where flaking deposits form on water-cooled louvres shadowed from the plasma. The asymmetry implies drift in the SOL of JET from outboard to inboard under normal operating conditions, which has been measured. In order to model the amounts of deposition, assumptions have to be made about the transport at the inner target. Analysis of divertor tiles shows that material from the main chamber travels along the SOL to the inner divertor wall, from where carbon is preferentially removed leaving a beryllium-rich film. The carbon travels to shadowed areas (such as the louvres) where deposits with high H-isotope content accrue. The analysis indicates that chemical processes must be important.

Dust measurements in tokamaks (invited)

Dust production and accumulation present potential safety and operational issues for the ITER. Dust diagnostics can be divided into two groups: diagnostics of dust on surfaces and diagnostics of dust in plasma. Diagnostics from both groups are employed in contemporary tokamaks; new diagnostics suitable for ITER are also being developed and tested. Dust accumulation in ITER is likely to occur in hidden areas, e.g., between tiles and under divertor baffles. A novel electrostatic dust detector for monitoring dust in these regions has been developed and tested at PPPL. In the DIII-D tokamak dust diagnostics include Mie scattering from Nd:YAG lasers, visible imaging, and spectroscopy. Laser scattering is able to resolve particles between 0.16 and 1.6 microm in diameter; using these data the total dust content in the edge plasmas and trends in the dust production rates within this size range have been established. Individual dust particles are observed by visible imaging using fast framing cameras, detecting dust particles of a few microns in diameter and larger. Dust velocities and trajectories can be determined in two-dimension with a single camera or three-dimension using multiple cameras, but determination of particle size is challenging. In order to calibrate diagnostics and benchmark dust dynamics modeling, precharacterized carbon dust has been injected into the lower divertor of DIII-D. Injected dust is seen by cameras, and spectroscopic diagnostics observe an increase in carbon line (CI, CII, C(2) dimer) and thermal continuum emissions from the injected dust. The latter observation can be used in the design of novel dust survey diagnostics.

Fluctuation-driven transport in the DIII-D boundary

Cross-field fluctuation-driven transport is studied in edge and scrape-off layer (SOL) plasmas in the DIII-D tokamak using a fast reciprocating Langmuir probe array allowing local measurements of the fluctuation-driven particle and heat fluxes. Two different non-diffusive mechanisms that can contribute strongly to the cross-field transport in the SOL of high-density discharges are identified and compared. The first of these involves intermittent transport events that are observed at the plasma separatrix and in the SOL. Intermittence has qualitatively similar character in L-mode and ELM-free H-mode. Low-amplitude ELMs observed in high-density H-mode produce in the SOL periods with cross-field transport enhanced to L-mode levels and featuring intermittent events similar to those in L-mode. The intermittent transport events are compatible with the concept of plasma filaments propagating across the SOL due to E×B drifts. The intermittent character of the transport in the SOL is also in agreement with predictions of the non-linear numerical simulations performed with an imposed driving flux. Another type of non-diffusive transport is often seen in high-density H-modes with prolonged ELM-free periods, where the transport near the separatrix is dominated by quasi-coherent modes driving particle and/or heat fluxes exceeding L-mode levels. These modes may play an important role by providing particle and/or heat exhaust between ELMs.

First tests of molybdenum mirrors for ITER diagnostics in DIII-D divertor

Metallic mirrors will be used in ITER for optical diagnostics working in different spectral ranges. Their optical properties will change with time due to erosion, deposition, and particle implantation. First tests of molybdenum mirrors were performed in the DIII-D divertor under deposition-dominated conditions. Two sets of mirrors recessed 2cm below the divertor floor in the private flux region were exposed to a series of identical, lower-single-null, ELMing (featuring edge localized modes) H-mode discharges with detached plasma conditions in both divertor legs. The first set of mirrors was exposed at ambient temperature, while the second set was preheated to temperatures between 140 and 80°C. During the exposures mirrors in both sets were additionally heated by radiation from the plasma. The nonheated mirrors exhibited net carbon deposition at a rate of up to 3.7nm∕s and suffered a significant drop in reflectivity. Net carbon deposition rate on the preheated mirrors was a factor of 30–100 lower and their...

Plasma?surface interaction, scrape-off layer and divertor physics: implications for ITER

Recent research in scrape-off layer (SOL) and divertor physics is reviewed; new and existing data from a variety of experiments have been used to make cross-experiment comparisons with implications for further research and ITER. Studies of the region near the separatrix have addressed the relationship of profiles to turbulence as well as the scaling of the parallel power flow. Enhanced low-field side radial transport is implicated as driving parallel flows to the inboard side. The medium-n nature of edge localized modes (ELMs) has been elucidated and new measurements have determined that they carry ~10?20% of the ELM energy to the far SOL with implications for ITER limiters and the upper divertor. The predicted divertor power loads for ITER disruptions are reduced while those to main chamber plasma facing components (PFCs) increase. Disruption mitigation through massive gas puffing is successful at reducing PFC heat loads. New estimates of ITER tritium retention have shown tile sides to play a significant role; tritium cleanup may be necessary every few days to weeks. ITERs use of mixed materials gives rise to a reduction of surface melting temperatures and chemical sputtering. Advances in modelling of the ITER divertor and flows have enhanced the capability to match experimental data and predict ITER performance.

Calculation of observable quantities using a divertor impurity interpretive code, DIVIMP

Our picture of divertor edge plasma behaviour is largely theoretical with little experimental confirmation. Questions exist with regard to impurity production mechanisms, sputtering yields, plasma ion temperatures, parallel temperature gradients, the existence of plasma flow reversal, cross-field diffusion coefficients, etc. Spatial distributions of impurity ions of successive charge states in the edge, when interpreted with an impurity production/transport code, provide a basis for discriminating amongst the various theoretical possibilities. In this paper, for an example case, a wide range of models/assumptions about the divertor edge plasma behaviour is examined with regard to the effect on the spectroscopic intensities of C + through C 5+ in the vicinity of the divertor targets. It is shown that a single spectroscopic viewing location can be sufficient to discriminate amongst most of the models/assumptions made about the divertor plasma behaviour. More refined discrimination would require additional viewing locations and diagnostics, such as Doppler temperature measurements.

Analysis of a multi-machine database on divertor heat fluxesa)

A coordinated effort to measure divertor heat flux characteristics in fully attached, similarly shaped H-mode plasmas on C-Mod, DIII-D, and NSTX was carried out in 2010 in order to construct a predictive scaling relation applicable to next step devices including ITER, FNSF, and DEMO. Few published scaling laws are available and those that have been published were obtained under widely varying conditions and divertor geometries, leading to conflicting predictions for this critically important quantity. This study was designed to overcome these deficiencies. Analysis of the combined data set reveals that the primary dependence of the parallel heat flux width is robustly inverse with Ip, which all three tokamaks independently demonstrate. An improved Thomson scattering system on DIII-D has yielded very accurate scrape off layer (SOL) profile measurements from which tests of parallel transport models have been made. It is found that a flux-limited model agrees best with the data at all collisionalities, while...