Assessment of potential heat flux overload of target and first wall components in Wendelstein 7-X finite-beta magnetic configurations and choice of locations for temperature monitoring
Introduction
Within the framework of the European roadmap to the realization of fusion energy [1], the construction and operation of the Wendelstein 7-X (W7-X) stellarator machine represents one of the big milestones. The W7-X is a large stellarator with superconducting coils, operated at the Greifswald site of the Max-Planck-Institut für Plasmaphysik [2]. The first goal of the W7-X operation has been to show the feasibility of building a superconducting modular stellarator with the required precision which has been successfully shown [3,4,5]. A second primary objective is to demonstrate the accessibility of plasma parameters close to those of a future Fusion Power Plant (FPP) [6,7], which is ongoing work [8,9]. The last goal will be to prove the possibility of high-power steady-state operation.
W7-X can be operated in different magnetic configurations, controlled by the currents in the different types of field coil [10]. Those configurations envisaged for high-power operation are characterised by a chain of natural magnetic islands at the plasma boundary. In any toroidal cross section, the magnetic flux surfaces in the island region form an O point in the centre of each island and an X point between two adjacent islands [11]. In the 3D view the corresponding O and X points of each cross section are connected by field lines closing upon themselves after a low number of toroidal revolutions. The islands are intersected by the target plates (island divertor concept) [9,12,13]. Following the typical five-fold toroidal symmetry and the up-down flip symmetry (stellarator symmetry), the targets are arranged in ten identical divertor units (Fig. 1).
So far, W7-X was operated with uncooled test divertor units (TDU) [14]. For future operation phases, a water-cooled high heat flux (HHF) divertor will be installed [15]. It is then planned to operate W7-X in steady-state discharges of up to 30 min with 10 MW of heating power. Adjacent to the target plates, where lower heat loads are expected, so called baffles are installed (Fig. 1). The remaining surface of the plasma vessel is covered partly by wall protection tiles of the same design as the baffles [16], partly by steel panels [17]. In the future stellarator FPPs, almost all the plasma vessel internal wall will be covered by breeding blanket modules [[18], [19], [20], [21]] aimed at removing the thermal power generated by the fusion plasma, shielding the magnets from neutron and gamma radiation and ensuring the tritium breeding of the plant.
For high-power long-pulse operation of W7-X, it is essential to protect the Plasma Facing Components (PFCs) listed above from heat loads exceeding the design specifications. This is particularly important for the baffles and targets, which are the most loaded PFCs because of the convective heat power deposited by charged particles. During the first divertor operation phase of W7-X, baffle loads above the design values were derived from infrared (IR) camera images [22], presumably because anomalous transport perpendicular to the magnetic field is higher than assumed during the design of the PFCs. At the same time, refined thermomechanical analysis of the baffles indicated that the maximum thermal load to these components should be reduced [23]. Whereas the target plates and part of the baffles is well monitored by IR cameras (Fig. 2), there are locations in which baffle or wall protection tiles cannot be observed.
It was therefore decided to install thermocouples in the heat sinks of selected baffle and wall protection tiles in order to avoid thermal overload to these components.
We are using field line diffusion (FLD) [24] to simulate the convective power load to PFCs. Whereas this has been done before mostly for the W7-X vacuum reference configurations [25,26] and only for a few cases with finite plasma pressure, and with a focus on the target loads [27,28], we shall here investigate configurations with finite plasma pressure. These are the so-called finite-β configurations, where β = p/(B2/(2μ0) is the ratio between plasma pressure and magnetic pressure. The magnetohydrodynamic equilibria with finite plasma pressure in a toroidally confined plasma are radially shifted toward the torus outboard side relative to the vacuum flux surfaces, such that we expect potential overload to occur on baffle tiles on the torus outboard side.
In this paper, we shall describe the methodology adopted to select a limited number of baffle positions for temperature monitoring. The basic philosophy is described in section 2, together with the approach to model the heat load onto PFCs for some magnetic configurations and to assess its statistical significance. Section 3 reports the main outcomes of the overload calculations whereas in section 4 we present the locations selected for temperature monitoring. Lastly, conclusions are given in section 5 and a complete overview of the overload calculations carried out is reported in the Appendix.
Section snippets
Calculation of thermal loads to wall components
In order to calculate the thermal loads onto W7-X target and wall components for a certain magnetic configuration, the following procedure was used. Whereas for a vacuum configuration (i. e., the changes in magnetic field due to plasma currents are negligible) the magnetic field inside the W7-X plasma vessel was calculated from the currents in the field coils (represented as filaments), for a finite-β plasma the magnetic field of a magnetohydrodynamic equilibrium was calculated by the VMEC [30]
Results of overload analysis
In order to investigate the overloads arising onto baffle and divertor tiles in finite-β configurations, FLD calculations for the magnetic configurations listed in Table 2 are performed. In particular, 7 different vacuum field configurations are chosen (called standard, low shear (orig./mod.), outward shifted, low iota, high mirror and high iota) and, for each of them, three different values of β are chosen for VMEC calculations. Configurations with finite plasma current are not covered in this
Choice of thermocouple locations
As discussed in section 2.4, the wall protection tiles with overloads in some of the magnetic configurations considered are clustered in a limited number of locations, and the same very few tiles within each cluster are predicted to receive the highest load even in different magnetic configurations. Those are the obvious choice for the thermocouples placement. In addition, care is taken on the one hand side to provide temperature monitoring in those critical locations that are not visible in
Conclusion
In the framework of W7-X R&D activities, an assessment of the overloads arising onto divertor, baffle and heat shield in finite-β magnetic configurations is reported in this paper. To this purpose a calculation procedure, aimed at finding out those tiles where the predicted convective heat flux exceeds the limit, is applied. The study is carried out assuming, for the baffle and heat shield, limits of 0.50 MW/m2 and 0.25 MW/m2. A maximum convected heat power of 8.0 MW is considered.
Results show
Declaration of Competing Interest
The authors declare that they have no known competing financial interests or personal relationships that could have appeared to influence the work reported in this paper.
CRediT authorship contribution statement
G. Bongiovì: Formal analysis, Investigation, Writing - original draft, Visualization. D. Böckenhoff: Methodology, Software, Resources, Writing - review & editing. A. Carls: Resources. M. Endler: Project administration, Conceptualization, Resources, Supervision, Writing - review & editing, Visualization. J. Fellinger: Project administration, Supervision, Writing - review & editing. J. Geiger: Resources.
Declaration of Competing Interest
The authors report no declarations of interest.
Acknowledgments
Authors want to thank M. Krause for providing the CAD views of the baffles and shields.
This work has been carried out within the framework of the EUROfusion Consortium and has received funding from the Euratom research and training programme 2014-2018 and 2019-2020 under grant agreement No 633053. The views and opinions expressed herein do not necessarily reflect those of the European Commission.
References (35)
- et al.
Stable stellarators with medium β and aspect ratio
Physics Letters A
(1986) Progress in the design and development of a test divertor (TDU) for the start of W7-X operation
Fus. Eng. Des.
(2009)Lessons learned from the design and fabrication of the baffles and heat shields of Wendelstein 7-X
Fus. Eng. Des.
(2013)The procurement and testing of the stainless steel in-vessel panels of the Wendelstein 7-X Stellarator
Fus. Eng. Des.
(2011)Preliminary structural assessment of the HELIAS 5-B breeding blanket
Fus. Eng. Des.
(2019)Progresses in the structural assessment of the central region of a HELIAS 5-B breeding blanket half sector
Fus. Eng. Des.
(2020)Overview of fatigue life assessment of baffles in Wendelstein 7-X
Fus. Eng. Des.
(2018)Service oriented architecture for scientific analysis at W7-X. An example of a field line tracer
Fus. Eng. Des.
(2013)Three-dimensional free boundary calculations using a spectral Green’s function method
Fus. Eng. Des.
(1986)European Research Roadmap to the Realisation of Fusion Energy
EUROfusion
(2018)