Divertor Plasma Facing Components Research Articles

High heat flux actively cooled plasma facing components development, realisation and first results in Tore Supra

The development, design, manufacture and testing of actively cooled high heat flux (HHF) plasma facing components (PFCs) has been an essential part of the Tore Supra programme towards long powerful tokamak operation. The Tore Supra PFC programme has culminated in the installation and operation of a toroidal pump limiter, since 2002, which already allowed to reach new world records in steady state operation (1 GJ injected in a 6 min discharge). The HHF PFCs development and manufacturing was achieved through a long lead development and industrialisation programme (about 10 years) marked out with a number of challenges. The major technical topics cope with bonding technology analysis involving an adequate material selection and procurement, repair processes development and implementation, development of destructive and non-destructive testing methods, and more generally industrialisation assessment. All these lessons are relevant to the ITER divertor PFCs manufacturing, although the technical solution adopted for Tore Supra (flat tiles concept) is different from the one proposed for the ITER divertor (monoblock concept). The routine operation of the actively cooled toroidal pumped limiter (TPL), capable to sustain up to 10 MW m −2 of nominal convected heat flux, is described. Up to now, the limiter fulfills its objectives in terms of heat exhaust. However, the thermographic monitoring exhibits unexpected behavior of the surface temperature. Particle exhaust control displays a complex pattern, due to the high fraction of the injected deuterium, which remains in the wall. The first experimental results with a full actively cooled wall gives access to ITER relevant information on wall conditioning, hydrogen plasma density and vacuum vessel inventory control, carbon erosion and redeposition and capability of in situ monitoring in a completely actively cooled environment.

A mature industrial solution for ITER divertor plasma facing components: Hypervapotron cooling concept adapted to Tore Supra flat tile technology

The use of flat tile technology to handle heat fluxes in the range of 20 MW/m 2 with components relevant for fusion experiment applications is technically possible with the hypervapotron cooling concept. This paper deals with recent high heat flux performances operated with success in two different electron gun facilities and highlights the high potential of this technology for ITER divertor application.

Manufacturing and testing in reactor relevant conditions of brazed plasma facing components of the ITER divertor

A fabrication route based on brazing technology has been developed for the realization of the high heat flux components for the ITER vertical target and Dome-Liner. The divertor vertical target is armoured with carbon fiber reinforced carbon and tungsten in the lower straight part and in the upper curved part, respectively. The armour material is joined to heat sinks made of precipitation hardened copper–chromium–zirconium alloy. The plasma facing units of the dome component are based on a tungsten flat tile design with hypervapotron cooling. An innovative brazing technique based on the addition of carbon fibers to the active brazing alloy, developed by Ansaldo Ricerche for applications in the field of the energy production, has been used for the carbon fiber composite to copper joint to reduce residual stresses. The tungsten–copper joint has been realized by direct casting. A proper brazing thermal cycle has been studied to guarantee the required mechanical properties of the precipitation hardened alloy after brazing. The fabrication route of plasma facing components for the ITER vertical target and dome based on the brazing technology has been proved by means of thermal fatigue tests performed on mock-ups in reactor relevant conditions.

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The European DEMO power reactor is currently under conceptual design within the EUROfusion Consortium. One of the most critical activities is the engineering of the plasma-facing components (PFCs) covering the plasma chamber wall, which must operate reliably in an extreme environment of neutron irradiation and surface heat and particle flux, while also allowing sufficient neutron transmission to the tritium breeding blankets. A systems approach using advanced numerical analysis is vital to realising viable solutions for these first wall and divertor PFCs. Here, we present the system requirements and describe bespoke thermo-mechanical and thermo-hydraulic assessment procedures which have been used as tools for design. The current first wall and divertor designs are overviewed along with supporting analyses. The PFC solutions employed will necessarily vary around the wall, depending on local conditions, and must be designed in an integrated manner by analysis and physical testing.

Tungsten: an option for divertor and main chamber plasma facing components in future fusion devices

The tungsten programme in ASDEX Upgrade is pursued towards a full high-Z device. The spectroscopic diagnostic of W has been extended and refined and the cooling factor of W has been re-evaluated. The W coated surfaces now represent a fraction of 65% of all plasma facing components (24.8 m2). The only two major components that are not yet coated are the strikepoint region of the lower divertor as well as the limiters at the low field side. While extending the W surfaces, the W concentration and the discharge behaviour have changed gradually pointing to critical issues when operating with a W wall: anomalous transport in the plasma centre should not be too low, otherwise neoclassical accumulation can occur. One very successful remedy is the addition of central RF heating at the 20–30% level. Regimes with low ELM activity show increased impurity concentration over the whole plasma radius. These discharges can be cured by increasing the ELM frequency through pellet ELM pacemaking or by higher heating power. Moderate gas puffing also mitigates the impurity influx and penetration, however, at the expense of lower confinement. The erosion yield at the low field side guard limiter can be as high as 10−3 and fast particle losses from NBI were identified to contribute a significant part to the W sputtering. Discharges run in the upper W coated divertor do not show higher W concentrations than comparable discharges in the lower C based divertor. According to impurity transport calculations no strong high-Z accumulation is expected for the ITER standard scenario as long as the anomalous transport is at least as high as the neoclassical one.

Final design of W7-X divertor plasma facing components—tests and thermo-mechanical analysis of baffle prototypes

The plasma facing components (PFCs) of the W7-X are designed in detail. The current design of the target plates, baffle plates and wall protection is presented which takes into account the requirements of the plasma heating, diagnostic systems and mounting. Prototypes of baffle elements were tested with heat loading to investigate the long term behaviour. The experimental results are compared with finite element calculations of the temperature and stress distributions in the elements. Based on these activities, the fabrication of the W7-X divertor PFCs and the graphite covered wall protection for W7-X can be initiated.

The thermophysical and mechanical properties of the copper heat sink material intended for use in ITER

Two copper-based alloys are currently under consideration as the heat sink material for the divertor, limiter and first-wall plasma facing components (PFCs) of ITER. During pulsed operation, these alloys will be subjected to cyclic stresses and periods of steady loading: it is anticipated that fatigue, creep and combined creep-fatigue effects will be important in limiting the operational lifetime of these PFCs, in particular the copper alloy components. A wide range of thermophysical and mechanical properties measurements have therefore been made, including under vacuum, in order to provide material characterisation of these important alloys, to add to the ITER materials database and to guide future design efforts.

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.

Lifetime of Irradiated Iter Divertor Heat Sink

Comparative lifetime analysis of the ITER divertor Plasma Facing Component (PFCs) - Carbon Fiber Composite (CFC) monoblock with a heat sink tube made from CuCrZr bronze has been performed. The aim of this analysis was to evaluate the irradiation effects in CFC and CuCrZr on the monoblock lifetime. The decreasing of CFC thermal conductivity and changing of CuCrZr mechanical properties such as tensile and yield strength and elongation during the monoblock life were taken into consideration. Cumulative fatigue damage for both irradiated and unirradiated states of the heat sink tube was evaluated.

Acceptable Mechanical Properties of CuCrZr Cooling Tubes of ITER Divertor Plasma Facing Components

Comparative lifetime analysis of the ITER divertor Plasma Facing Components (PFCs) - Carbon Fiber Composite (CFC) monoblock with a cooling tube made from CuCrZr has been performed. The aim of this analysis was to evaluate acceptable mechanical properties of the CuCrZr cooling tube, such as ultimate tensile strength (UTS) and yield strength (YS), that will satisfy the ITER lifetime requirements. These properties can vary depending on the manufacturing cycle and heat treatment. Three sets of mechanical properties were used in the analysis, and two variants of the CFC monoblock design were considered.

Acceptable Mechanical Properties of CuCrZr Cooling Tubes of ITER Divertor Plasma Facing Components

Comparative lifetime analysis of the ITER divertor Plasma Facing Components (PFCs) - Carbon Fiber Composite (CFC) monoblock with a cooling tube made from CuCrZr has been performed. The aim of this analysis was to evaluate acceptable mechanical properties of the CuCrZr cooling tube, such as ultimate tensile strength (UTS) and yield strength (YS), that will satisfy the ITER lifetime requirements. These properties can vary depending on the manufacturing cycle and heat treatment. Three sets of mechanical properties were used in the analysis, and two variants of the CFC monoblock design were considered.

ITER divertor maintenance L7 R&D project — results and perspectives

The availability of International Thermonuclear Experimental Reactor (ITER) (or an ‘ITER-like’) reactor will be strongly influenced by the effectiveness of the in-vessel components remote handling strategy. In the present reactor concept, the relevant key components are the divertor cassettes, located in the lower region of the vacuum vessel. Due to erosion and damage to the divertor plasma facing components, estimated scheduled replacement of the cassettes will be required eight times during the machine lifetime. Moreover, for such an experimental tokamak where completely new plasma regimes will be established, unscheduled interventions cannot be excluded a priori. To test and optimise the divertor maintenance scenario, the so called ITER L7 R&D project has been realised at the ENEA Brasimone laboratories, with a full scale simulation of the in-vessel cassette maintenance environment and of the hot cell refurbishment operations. The basic demonstration of the validity of the current scenario has been established, and now new activities are in progress to optimise many aspects of the operations (procedures, hardware improvements, reliability, etc.). Based on the background and the results of these activities, this paper discusses the lessons learned during the project implementation, and identifies key points of the current strategy that should be maintained for any new design of ITER or an ‘ITER-like’ reactor.

Gallium cooled ITER divertor cassette

This paper involves the results of the conceptual design analyses of an advanced divertor plates system, altematWe to the current high heat flux loaded divertor plasma facing components for ITER. I...

Competition of bulk trapping and surface erosion in the kinetics of tritium inventory and permeation in plasma protection metals

A simplified transient model is presented to describe the migration of implanted tritium in the presence of trap sites across the bulk of metallic substrates whose thickness is decreasing with time due to erosion. The subject is relevant for quantifying the tritium inventory in — and permeation through — plasma facing armours in the next generation of D-T fuelled tokamak devices (i.e., the International Thermonuclear Experimental Reactor). This paper describes the equations of the physical model and the main assumptions used to simplify the complex analysis, and surveys the influence of several parameters such as the implantation flux, the erosion rate, the armour temperature, the armour thickness, the density and trapping energy of neutron-induced traps, etc., which are all expected to play a key role in the phenomena investigated. The examples presented to show the applicability of the model include the results of a study performed for beryllium armours exposed to heat and particle loads similar to those expected on the ITER divertor plasma facing components and comparison is made with cases where erosion does not play any role.