I. Konkashbaev

Comprehensive physical models and simulation package for plasma/material interactions during plasma instabilities.

Damage to plasma-facing components (PFCs) from plasma instabilities remains a major obstacle to a successful tokamak concept. The extent of the damage depends on the detailed physics of the disrupting plasma, as well as on the physics of plasma-material interactions. A comprehensive computer package called High Energy Interaction with General Heterogeneous Target Systems (HEIGHTS) has been developed and consists of several integrated computer models that follow the beginning of a plasma disruption at the scrape-off layer (SOL) through the transport of the eroded debris and splashed target materials to nearby locations as a result of the deposited energy. The package can study, for the first time, plasma-turbulent behavior in the SOL and predict the plasma parameters and conditions at the divertor plate. Full two-dimensional (2-D) comprehensive radiation magnetohydrodynamic (MHD) models are coupled with target thermodynamics and liquid hydrodynamics to evaluate the integrated response of plasma-facing materials. Factors that influence the lifetime of plasma-facing and nearby components, such as loss of vapor cloud confinement and vapor removal due to MHD effects, damage to nearby components due to intense vapor radiation, melt splashing, and brittle destruction of target materials, are also modeled and discussed.

An assessment of disruption erosion in the ITER environment

The behavior of divertor materials during a major disruption in ITER is very important to successful and reliable operation of the reactor. Erosion of material surfaces due to a thermal energy dump can severely limit the lifetimes of plasma-facing components and thus diminish the reactors economic feasibility. A comprehensive numerical model has been developed and used in this analysis, which includes all major physical processes taking place during plasma/material interactions. Models to account for material thermal evolution, plasma/vapor interaction physics, and models for hydrodynamic radiation transport in the developed vapor cloud are implemented in a self-consistent manner to realistically assess disruption damage. The extent of self-protection from the developed vapor cloud in front of the incoming plasma particles is critically important in determining the overall disruption lifetime. Models to study detailed effects of the strong magnetic field on the behavior of the vapor cloud and on the net erosion rate have been developed and analyzed. Candidate materials such as beryllium and carbon are both considered in this analysis. The dependence of divertor disruption lifetime on disruption physics and reactor conditions was analyzed and discussed.

Lifetime evaluation of plasma-facing materials during a tokamak disruption☆

Erosion losses of plasma-facing materials in a tokamak reactor during major disruptions, giant ELMS, and large power excursions are serious concerns that influence component survivability and overall lifetime. Two different mechanisms lead to material erosion during these events: surface vaporization and loss of the melt layer. Hydrodynamics and radiation transport in the rapidly developed vapor-cloud region above the exposed area are found to control and determine the net erosion thickness from surface vaporization. A comprehensive self-consistent kinetic model has been developed in which the time-dependent optical properties and the radiation field of the vapor cloud are calculated in order to correctly estimate the radiation flux at the divertor surface. The developed melt layer of metallic divertor materials will, however, be free to move and can be eroded away due to various forces. , Physical mechanisms that affect surface vaporization and cause melt layer erosion are integrated in a comprehensive model. It is found that for metallic components such as beryllium and tungsten, lifetime due to these abnormal events will be controlled and dominated by the evolution and hydrodynamics of the melt layer during the disruption. The dependence of divertor plate lifetime on various aspects of plasma/material interaction physics is discussed.

Comprehensive model for disruption erosion in a reactor environment

Abstract A comprehensive disruption erosion model which takes into account the interplay of major physical processes during plasma-material interaction has been developed. The model integrates with sufficient detail and in a self-consistent way, material thermal evolution response, plasma-vapor interaction physics, vapor hydrodynamics and radiation transport in order to realistically simulate the effects of a plasma disruption on plasma-facing components. Candidate materials such as beryllium and carbon have been analyzed. The dependence of the net erosion rate on disruption physics and various parameters was analyzed and is discussed.

Comprehensive modeling of ELMs and their effect on plasma-facing surfaces during normal tokamak operation☆

During the normal operation of the high confinement regime (H-mode) in next generation tokamaks, edge-localized modes (ELMs) are a serious concern for divertor plasma-facing components. The periodic relaxation of edge pressure gradient results in pulses of energy and particles transported across the Separatrix to the scrape-off-layer (SOL) and eventually to the divertor surface. ELMs could, therefore, result in cyclic thermal stresses, excessive target erosion, and consequently shorter divertor lifetime. In this study a comprehensive two-fluid model has been developed to integrate SOL parameters during ELMs with divertor surface evolution (melting, vaporization, vapor cloud dynamics, and macroscopic spallation) for different ELM parameters. Calculations were performed using the HEIGHTS numerical simulation package. Initial results indicate that high-power ELMs in ITER-like machines can cause serious damage to divertor components, may terminate plasma in disruptions, and may affect subsequent plasma operations due to extensive contamination.

Modeling and simulation of melt-layer erosion during plasma disruption

Abstract Metallic plasma-facing components (PFCs) e.g. beryllium and tungsten, will be subjected to severe melting during plasma instabilities such as disruptions, edge-localized modes and high power excursions. Because of the greater thickness of the resulting melt layers relative to that of the surface vaporization, the potential loss of the developing melt-layer can significantly shorten PFC lifetime, severely contaminate the plasma and potentially prevent successful operation of the tokamak reactor. Mechanisms responsible for melt-layer loss during plasma instabilities are being modeled and evaluated. Of particular importance are hydrodynamic instabilities developed in the liquid layer due to various forces such as those from magnetic fields, plasma impact momentum, vapor recoil and surface tension. Another mechanism found to contribute to melt-layer splashing loss is volume bubble boiling, which can result from overheating of the liquid layer. To benchmark these models, several new experiments were designed and performed in different laboratory devices for this work; the results are examined and compared. Theoretical predictions (A ∗ THERMAL-S and SPLASH codes) are generally in good agreement with the experimental results. The effect of in-reactor disruption conditions, which do not exist in simulation experiments, on melt-layer erosion is discussed.

Tritium behavior in eroded dust and debris of plasma-facing materials

Tritium behavior in plasma-facing components (PFCs) of future tokamak reactors such as ITER is an essential factor in evaluating and choosing the ideal plasma-facing materials (PFMs). One important parameter that influences tritium buildup and release in candidate materials is the effect of material porosity on tritium diffusion and retention. Diffusion in porous materials, for example, consists of three different processes: along grain boundaries, along microcrystallite boundaries, and in pure crystallite structures. Such diffusion processes have strong nonlinear behavior due to temperature, solubility, and existing trap sites. Therefore, a realistic model for tritium diffusion in porous and neutron-irradiated materials must account for both nonlinear and multidimensional effects. A tritium transport computer model Tritium Accumulation in Porous Structure (TRAPS) has been developed to evaluate and predict the kinetics of tritium transport in porous media. This two-dimensional model incorporates tritium diffusion and trapping processes that also account for hydrogen-isotope solubility limits in PFMs. This model is being coupled with the computer model Tritium In Compound System (TRICS) which has been developed to study the effects of surface erosion to tritium behavior in PFCs.

Erosion of melt layers developed during a plasma disruption

Material erosion of plasma-facing components during a tokamak disruption is a serious problem that limits reactor operation and economical reactor lifetime. In particular, metallic low-Z components such as Be will be subjected to severe melting during disruptions and edge localized modes (ELMs). Loss of the developed melt layer will critically shorten the lifetime of these components, severely contaminate the plasma, and seriously inhibit successful and reliable operation of the reactor. In this study mechanisms responsible for melt-layer loss during a disruption are modeled and evaluated. Implications of melt-layer loss on the performance of metallic facing components in the reactor environment are discussed.

New critical assessments of chamber and wall response to target implosion in inertial fusion reactors.

Abstract The chamber walls in inertial fusion energy (IFE) reactors are exposed to harsh conditions following each target implosion. Key issues of the cyclic IFE operation include intense photon and ion deposition, wall thermal and hydrodynamic evolution, wall erosion and fatigue lifetime, and chamber clearing and evacuation to ensure desirable conditions prior to next target implosion. Several methods for wall protection have been proposed in the past, each having its own advantages and disadvantages. These methods include use of solid bare walls, gas-filled cavities, and liquid walls/jets. Detailed models have been developed for reflected laser light, emitted photons, neutrons, and target debris deposition and interaction with chamber components and have been implemented in the comprehensive heights software package. The hydrodynamic response of gas-filled cavities and photon radiation transport of the deposited energy have been calculated by means of new and advanced numerical techniques for accurate shock treatment and propagation. Photon radiation transport models are developed for either the gas-filled cavity or in the evolving vapor cloud layer above the wall surface. The focus of this work is to examine the overall wall response and lifetime due to various erosion mechanisms.

Theory and models of material erosion and lifetime during plasma instabilities in a tokamak environment

Surface and structural damage to plasma-facing components (PFCs) due to the frequent loss of plasma confinement remains a serious problem for the tokamak reactor concept. The deposited plasma energy causes significant surface erosion, possible structural failure and frequent plasma contamination. Surface damage consists of vaporization, spallation, and liquid splatter of metallic materials. Structural damage includes large temperature increases in structural materials and at the interfaces between surface coatings and structural members. To evaluate the lifetimes of plasma-facing materials and nearby components and to predict the various forms of damage that they experience, comprehensive models (contained in the HEIGHTS computer simulation package) are developed, integrated self-consistently and enhanced. Splashing mechanisms, such as bubble boiling and various liquid magnetohydrodynamic instabilities and brittle destruction mechanisms of non-melting materials, are being examined. The design requirements and implications of plasma-facing and nearby components are discussed, along with recommendations to mitigate and reduce the effects of plasma instabilities on reactor components.