J. Bucalossi

WEST Physics Basis

With WEST (Tungsten Environment in Steady State Tokamak) (Bucalossi et al 2014 Fusion Eng. Des. 89 907-12), the Tore Supra facility and team expertise (Dumont et al 2014 Plasma Phys. Control. Fusion 56 075020) is used to pave the way towards ITER divertor procurement and operation. It consists in implementing a divertor configuration and installing ITER-like actively cooled tungsten monoblocks in the Tore Supra tokamak, taking full benefit of its unique long-pulse capability. WEST is a user facility platform, open to all ITER partners. This paper describes the physics basis of WEST: the estimated heat flux on the divertor target, the planned heating schemes, the expected behaviour of the L-H threshold and of the pedestal and the potential W sources. A series of operating scenarios has been modelled, showing that ITER-relevant heat fluxes on the divertor can be achieved in WEST long pulse H-mode plasmas.

Numerical modelling for divertor design of the WEST device with a focus on plasma?wall interactions

In the perspective of operating tungsten monoblocks in WEST, the ongoing major upgrade of the Tore Supra tokamak, a dedicated modelling effort has been carried out to simulate the interaction between the edge plasma and the tungsten wall. A new transport code, SolEdge2D–EIRENE, has been developed with the ability to simulate the plasma up to the first wall. This is especially important for steady state operation, where thermal loads on all the plasma facing components, even remote from the plasma, are of interest. Moreover, main chamber tungsten sources are thought to dominate the contamination of the plasma core. We present here in particular new developments aimed at improving the description of the interface between the plasma and the wall, namely a way to treat sheath physics in a more faithful way using the output of 1D particle in cell simulations. Moreover, different models for prompt redeposition have been implemented and are compared. The latter is shown to play an important role in the balance between divertor and main chamber sources.

DEMO reactor design using the new modular system code SYCOMORE

A demonstration power plant (DEMO) will be the next step for fusion energy following ITER. Some of the key design questions can be addressed by simulations using system codes. System codes aim to model the whole plant with all its subsystems and identify the impact of their interactions on the design choices. The SYCOMORE code is a modular system code developed to address key questions relevant to tokamak fusion reactor design. SYCOMORE is being developed within the European Integrated Tokamak Modelling framework and provides a global view (technology and physics) of the plant. It includes modules to address plasma physics, divertor physics, breeding blankets, shield design, magnet design and the power balance of plant. The code is coupled to an optimization framework which allows one to specify figures of merit and constraints to obtain optimized designs. Examples of pulsed and steady-state DEMO designs obtained using SYCOMORE are presented. Sensitivity to design assumptions is also studied, showing that the operational domain around working points can be narrow for some cases.

The WEST project: validation program for WEST tungsten coated plasma facing components

The W—for tungsten—Environment in Steady-state Tokamak (WEST) project is based on an upgrade of the Tore Supra tokamak from a carbon limiter to an X-point divertor device. A new set of actively cooled tungsten coated plasma facing components will cover a part of the vessel to provide a fully metallic environment. This paper deals with the validation program performed for tungsten coatings (≥15 μm) on a CuCrZr substrate. The first step was dedicated to the qualification under high heat flux tests of the coating on small inertially cooled samples. To study the thermal behavior and the non-uniformity, the second step was dedicated to the validation of the coating on large inertially cooled samples with geometry and shape (540 × 120 mm) representative of the WEST coated components. The last step was dedicated to the optimization of the coating and to the high heat flux tests up to 10.5 MW m−2 on relevant coated actively cooled prototypes. Non-uniformity and thickness of the coating (15 and 30 μm) correspond to specifications. As no delamination was observed, coatings of 15 and 30 μm were qualified with regard to their application on WEST coated components. In order to decrease the risk of coating delamination under thermal loading, it was decided to cover the upper divertor and baffle targets with the thinnest coating option of 15 μm.

Coupling between a multi-physics workflow engine and an optimization framework

A generic coupling method between a multi-physics workflow engine and an optimization framework is presented in this paper. The coupling architecture has been developed in order to preserve the integrity of the two frameworks. The objective is to provide the possibility to replace a framework, a workflow or an optimizer by another one without changing the whole coupling procedure or modifying the main content in each framework. The coupling is achieved by using a socket-based communication library for exchanging data between the two frameworks. Among a number of algorithms provided by optimization frameworks, Genetic Algorithms (GAs) have demonstrated their efficiency on single and multiple criteria optimization. Additionally to their robustness, GAs can handle non-valid data which may appear during the optimization. Consequently GAs work on most general cases. A parallelized framework has been developed to reduce the time spent for optimizations and evaluation of large samples. A test has shown a good scaling efficiency of this parallelized framework. This coupling method has been applied to the case of SYCOMORE (System COde for MOdeling tokamak REactor) which is a system code developed in form of a modular workflow for designing magnetic fusion reactors. The coupling of SYCOMORE with the optimization platform URANIE enables design optimization along various figures of merit and constraints

Study of the pores inside tungsten coating after thermal cycling for fusion device

In the next fusion devices, all the plasma facing components will consist of bulk tungsten or tungsten coating on carbon. This paper focuses on the behaviour of tungsten coated on carbon fibre composite designed for the WEST project (Bucalossi et al 2011 Fusion Eng. Des. 86 684-688) under intensive thermal cycling delivered by an electron beam. The use of scanning electron microscope has allowed in particular, the observation of several pore lines inside the coating. These pore lines have different aspects depending on the observed zone according to the localisation of the electron beam, accentuated lines with more numerous enlarged pores in zone exposed to the electron beam. An analogous trend is also observed for JET tungsten-coated samples under similar thermal cycles despite their different properties due to an alternative manufacturing method of the substrate. A systematic and attentive comparison on the coating changes after the application of the electron beam heating is presented. The observed comportments as the formation of the pore lines or the pore shapes are assumed to be inherent to simultaneous diffusion processes. In association with the pore line formation, a migration of the carbon substrate towards the surface is presumed and discussed.

Tungsten covered graphite and copper elements and ITER-like actively cooled tungsten divertor plasma facing units for the WEST project

After a brief introduction giving some insight of the WEST project, we present the three types of plasma facing units (PFUs) developed for the WEST project taking into account the envisaged main scenarios: (1) high power short pulse scenario (a few seconds) where the objective is to maximize the power handling of the PFUs, up to 20 MW m−2, (2) high fluence scenario (a few 100 s) on actively cooled ITER-like tungsten (W) PFUs, up to 10 MW m−2 during 1000 s. For the graphite PFUs, the high heat flux tests have been done at GLADIS (ion beam test facility), and for the CuCrZr PFUs on the JUDITH (electron beam test facility). The tests were successful, as no damage occurred for the different load cases. This confirms that the modelling done during the design phase is appropriate to describe these PFUs. Series productions are expected to be achieved by the end of 2015 for the graphite and CuCrZr PFUs, and few ITER-like W PFUs are expected at the beginning of 2016. The lower divertor will be complemented with ITER-like W PFUs as soon as available from our partners so that different fabrication procedures could be evaluated in a real industrial process and a real tokamak environment.

Radiation Driven Bifurcations in Fusion Plasmas

Operation of high performance fusion plasmas relies on self-organised properties to reach appropriate working points that are compatible with both high confinement performance to achieve a burning plasma, and controlled ageing of the confinement device. The latter conditions requires a trade-off between simplicity of the operation point and reaching conditions that can be sustained in steady state. The issue of heat flux control at the plasma edge and onto the plasma facing components is an example of this synergy. We address in this framework the problem of radiative divertor operation. The simplified 1D problem is recast in Hamiltonian formalism, the effective energy being invariant. This property is most efficient to address bifurcations and critical points leading to no-solution regions of the parameter space. Analytical investigation of these solutions indicates that taking into account the radiative front location and constraints on the upstream temperature reduces the operation space. Furthermore, one finds that radiative divertor operation tends to lead to operation at reduced plasma pressure, unless stable conditions and hot upstream plasma temperature can be sustained at vanishing divertor temperature.

Wall surface temperature calculation in the SolEdge2D-EIRENE transport code

A thermal wall model is developed for the SolEdge2D-EIRENE edge transport code for calculating the surface temperature of the actively-cooled vessel components in interaction with the plasma. This is a first step towards a self-consistent evaluation of the recycling of particles, which depends on the wall surface temperature. The proposed thermal model is built to match both steady-state temperature and time constant of actively-cooled plasma facing components. A benchmark between this model and the Finite Element Modelling code CAST3M is performed in the case of an ITER-like monoblock. An example of application is presented for a SolEdge2D-EIRENE simulation of a medium-power discharge in the WEST tokamak, showing the steady-state wall temperature distribution and the temperature cycling due to an imposed Edge Localised Mode-like event.

ICRF heating scenarios in JET with emphasis on 4He plasmas for the non-activated phase of ITER

In the initial phase of ITER operation, 4He plasmas could be used in order to avoid activating the machine. The main ICRH scenarios foreseen for ITER 4He plasmas are (3He)4He and (H)4He. ICRH experiments have been carried out on JET using 4He plasmas to validate these scenarios. At the same time, conditions for access to H-mode in plasmas of various isotope compositions from dominantly 4He to dominantly D have been studied. Experiments have also been carried out for the first time in 4He plasmas with the ICRF power added to 4He neutral beam injection at the third harmonic of 4He in order to produce a 4He tail for alpha particle studies.