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# Finite element methods on composite meshes for tuning plasma equilibria in tokamaks

- Holger Heumann
^{1, 2}, - Francesca Rapetti
^{2}Email author and - Xiao Song
^{1, 2}

**8**:8

https://doi.org/10.1186/s13362-018-0050-7

© The Author(s) 2018

**Received:**13 February 2018**Accepted:**12 August 2018**Published:**5 September 2018

## Abstract

The fundamental concept of magnetically confined nuclear fusion devices is the magnetohydrodynamic (MHD) equilibrium: the pressure gradient due to highly energetic charged particles is balanced by the Lorenz force due to strong magnetic fields. Hence, numerical methods for MHD equilibria are also fundamental for fusion engineering applications. We rely here on a finite element method on composite meshes for the simulation of axisymmetric equilibria in tokamaks, torus-shaped nuclear fusion devices. One mesh with Cartesian quadrilaterals covers the domain accessible by the plasma and one mesh with triangles discretizes the region outside the chamber. The two meshes overlap in a narrow region. This approach gives the flexibility to achieve easily and at low cost higher order regularity for the approximation of the principal unknown, the poloidal magnetic flux, while preserving accurate meshing of the geometric details in the exterior. We show that higher order regularity allows to formulate appropriate optimal control problems that help to find a special type of equilibria, called snowflake equilibria, that are a very promising concept to mitigate high heat loads due to plasma escaping particles.

## Keywords

- Mortar element method
- Thermonuclear fusion
- Optimal control
- Non-linear elliptic partial differential equations

## MSC

- 35J60
- 65N30
- 65N21

## 1 Introduction

The possibility of using composite meshes in finite element (FE) simulations of industrial problems is a recurrent topic [8, 12, 28, 29, 33]. Composite meshes are involved as soon as the global discretization of a partial differential equation combines discretizations on local (overlapping or non-overlapping) subdomains, each suitably triangulated by non-matching grids. The reason for using composite meshes are various: fitting the geometry or the local smoothness of the solution, resolving multiple scales in regions with irregular data, using fast solvers on structured grids or a divide-and-conquer/domain decomposition approach to very large problems on parallel machines.

In the present case, we are looking for a simple and practical approach to introduce in certain parts of the computational domain FE functions that are not only continuous, but have also first order, second order or higher order continuous derivatives. In general it is very difficult to introduce FE spaces over simplicial unstructured meshes with such properties. On the other hand, if we work with Cartesian meshes this becomes very simple. It is sufficient to use tensor products of spline spaces with sufficiently high regularity. So, as it is naive to expect that technical devices can be entirely triangulated with Cartesian meshes, we introduce composite meshes involving Cartesian meshes in those subdomains where we want high regular FE representations and triangular unstructured meshes in those subdomains where we want conformity with the geometry.

The industrial application we consider concerns the free-boundary plasma equilibrium in tokamaks for nuclear fusion [3], which derives from the magnetohydrodynamic (MHD) equilibrium, described by the force balance and Maxwell’s equations in the eddy-current approximation. Nuclear fusion is a highly exothermic reaction in which two light atomic nuclei fuse to form a heavier nucleus. The peaceful use of such reactions for energy production on earth is a multinational research effort with high impact on the long-term perspective of energy production and consumption. The most promising technology to achieve this goal is currently the tokamak, a torus shaped reactor that uses strong magnetic fields to confine plasma and to achieve the extreme conditions to start the fusion reaction. The upcoming ITER (International Thermonuclear Experimental Reactor) tokamak, jointly build by China, Europe, India, Russia, South Korea and USA, will be the largest magnetic confinement experiment. It aims at demonstrating the principle of producing more energy from the fusion process than is used to initiate it, something that has not yet been achieved in any fusion reactor. Achieving this dream of creating a small sun on earth, that provides an almost endless amount of clean and sustainable energy, is a main motivation of this joint research work. The various open problems require interdisciplinary research and are a steady source of interesting and challenging questions, requiring also a high expertise in the field of applied and computational mathematics.

By symmetry considerations the free-boundary plasma equilibrium problem can be reduced to a scalar semi-linear elliptic one for the flux of the poloidal magnetic field (see Sect. 2.1 for details and references or [3, Sect. 1.2]). As the magnetic field and the current density are tangential to the level sets of the poloidal flux, the precise calculation of the level set distribution for the poloidal flux is fundamental in tokamak science. Hence, it is important to have good approximations not only of the poloidal magnetic flux but also of its derivatives.

*limiter configuration*, or is fully detached from the limiter. The latter is called

*divertor configuration or X-point configuration*as the plasma boundary contains a saddle point of the poloidal field. In the very early tokamak devices the plasma was always attached to the limiter, while later the focus shifted towards devices that create free standing plasmas that are completely detached from material that would have to resist extremely high temperatures, otherwise. This allows to maintain plasmas at much higher temperatures. But even though a direct contact of plasma and material is avoided there are nevertheless high temperature heat loads on some parts of the material inside the tokamak. Particles that escape from the plasma hit the limiter at places where the limiter intersects with the contour lines of the poloidal flux. But since the toroidal component of the particle velocity is very large, the impact place is in the vicinity of the X-point. This may damage the device since the particles deposit their significant energy on a relatively small area leading to unacceptably high heat flux densities. To mitigate the heat flux impact several strategies are proposed. One of these strategies aims at the exploration of the so-called snowflake configuration [30, 31]. A snowflake configuration (see Fig. 2 right) is obtained when there is a point where not only the gradient of the poloidal flux vanishes but also its second order derivatives. This second condition implies that contour line through this point has more than four different branches. It thus results that the heat flux load is distributed over a larger impact area.

The outline is the following: The Sects. 2 and 3 deal with the free-boundary plasma equilibrium problem and the numerical methods, respectively. We present the model (Sect. 2.1), the free-boundary equilibrium problem, and its role for the operation of tokamaks (Sect. 2.2). In Sect. 3.1 we recall the classical mortar element method (MEM) for overlapping meshes and introduce a modified method (MEM-M) that simplifies the implementation by avoiding integrals over cut elements. Next, in Sect. 3.2 we present the MEM-M Galerkin formulation for the plasma equilibrium problem. The related optimal control problem for tuning plasma equilibria appears in Sect. 3.3. We explain how an existing implementation of Newton’s method for the free-boundary equilibrium problem can be easily extended to solve efficiently the corresponding optimal control problems. The section on results, Sect. 4, starts with a validation of the convergence of the MEM and MEM-M and ends with a case study for finding snowflake configurations for the future tokamak CFETR. Sect. 5 gives a brief summary and draws the conclusions.

## 2 Formulation of the problem

*p*is the plasma kinetic pressure,

**B**is the magnetic induction,

**J**is the current density and

*μ*the magnetic permeability. The magnetodydrodymic equilibrium (1) is a fundamental concept for nuclear fusion and we refer to standard text books, e.g. [3, 13, 15, 16, 35] and [22] for the details. Nevertheless, to keep this contribution concise, we give in the subsequent section a brief introduction following the lines of [21, Sect. 2].

### 2.1 The free-boundary plasma equilibrium problem

Tokamaks are predominantly axial symmetric devices, hence it is convenient to formulate (1) in a cylindrical coordinate system \((r,\varphi,z)\) in order to consider only a section at \(\varphi= \text{ constant}\) of the tokamak, generally referred to as *poloidal section*. Working in a poloidal section, the scalar field *p* does not depend on the angle *φ*, thus ∇*p* belongs to the poloidal \((r,z)\)-plane. We introduce \(\mathbb {H} = [0,\infty] \times[-\infty,\infty]\), the positive half plane, to denote the poloidal plane that contains the tokamak centered at the origin. The classical primal unknowns for toroidal plasma equilibria described by (1) are the *poloidal magnetic flux*
\(\psi=\psi(r,z)\), the pressure *p* and the *diamagnetic function*
*f*. The poloidal magnetic flux \(\psi:= r\mathbf {A}\cdot\mathbf {e}_{\varphi}\) is the scaled toroidal component (*φ*-component) of the magnetic vector potential **A**, such that \(\mathbf {B}= \operatorname{curl}\mathbf {A} \), and \(\mathbf {e}_{\varphi}\) the unit vector for the *φ* coordinate. The diamagnetic function \(f = r \mathbf {B}\cdot\mathbf {e}_{\varphi}\) is the scaled toroidal component of the magnetic field **B**. It can be shown that both the pressure *p* and the diamagnetic function *f* are constant on *ψ* contour lines, i.e. \(p=p(\psi)\) and \(f=f(\psi)\).

*i*th coil \(\mathcal {C}_{i} \subset\mathbb {H}\) and

*μ*is a functional of

*ψ*that reads

*ψ*: \(\mathcal {P}(\psi)\) is a functional of the poloidal flux

*ψ*. The different characteristic shapes of \(\mathcal {P}(\psi)\) are illustrated in Fig. 2: the boundary of \(\mathcal {P}(\psi)\) either touches the boundary of \(\mathcal {L}\) (limiter configuration) or the boundary contains one or more saddle points of

*ψ*(divertor configuration). The saddle points of

*ψ*, denoted by \((r_{\mathrm{X}}, z_{\mathrm{X}})\)=\((r_{\mathrm{X}}(\psi),z_{\mathrm{X}}(\psi))\), are called X-points of

*ψ*. The plasma domain \(\mathcal {P}(\psi)\) is the largest subdomain of \(\mathcal {L}\) bounded by a closed

*ψ*-contour line in \(\mathcal {L}\) and containing the magnetic axis \((r_{\mathrm{{max}}} ,z_{\mathrm{{max}}})\). The magnetic axis is the point \((r_{\mathrm{{max}}},z_{\mathrm {{max}}})=(r_{\mathrm{{max}}}(\psi), z_{\mathrm{{max}}}(\psi))\), where

*ψ*has its global maximum in \(\mathcal {L}\). For convenience, we introduce also the coordinates \((r_{\mathrm{{bdp}}},z_{\mathrm{{bdp}}})=(r_{\mathrm{{bdp}}}(\psi ),z_{\mathrm{{bdp}}}(\psi))\) of the point that determines the plasma boundary. Note that \((r_{\mathrm {{bdp}}},z_{\mathrm{{bdp}}} )\) is either an X-point of

*ψ*or the contact point with the limiter \(\partial\mathcal {L}\). The Fig. 4 presents the actual geometric setting of 3 different tokamaks showing the big variety of designs.

*Grad–Shafranov–Schlüter*equation [17, 25, 32]. The domain of \(p'\) and \(f f'\) is the interval \([\psi_{\mathrm{{bdp}}},\psi_{\mathrm{{max}}}]\) with the scalar values \(\psi_{\mathrm{{max}}}\) and \(\psi_{\mathrm {{bdp}}}\) being the flux values at the

*magnetic axis*and at the boundary of the plasma:

*ψ*, a fixed domain \([0,1]\). They are usually given as piecewise polynomial functions. Another frequent a priori model is

*β*is related to the poloidal beta [3, p. 15], whereas

*α*and

*γ*describe the peakage of the current profile.

### 2.2 The plasma equilibrium and tokamak experiments

^{1}

*ψ*values on \(N_{\mathrm{desi}}+1\) given points \((r_{i},z_{i})\). However, from the definition of the equilibrium problem it is clear that the stationary points of the magnetic flux

*ψ*have a very important role and it would be very beneficial to formulate objective functionals for these stationary points. Moreover, the location of the X-point has a big influence, where the extremely hot impurities released from the plasma core hit the walls of the reactor. Very recently it was discovered that the so-called snowflake configuration, with degenerated X-points or with many X-points close to each other (see Fig. 7) can have very positive effects for the heat load mitigation, and hence, engineers are interested in preparing tokamak scenarios with such configurations.

With the current approaches it is not obvious how to formulate objective functionals for such tasks. The gradients or Hessians of the Galerkin approximation of *ψ* are non-smooth across element boundaries. Point evaluations of these gradients and Hessians are not well defined. Therefore, we prefer to work with higher order regular Galerkin methods. As this is easy with Cartesian meshes, we are interested in combining Cartesian meshes covering the burning chamber with triangle meshes covering the remaining parts of the computational domain.

## 3 Numerical methods

To simplify the presentation of the optimal control formulation and the main ingredients for a implementation we focus first on the details of the Mortar Element Method. To keep the discussion concise we elaborate the MEM for a linear elliptic problem.

### 3.1 A mortar element method (MEM) with overlapping meshes

*ψ*in the bounded domain \(\Omega\subset\mathbb {R}^{n}\) with boundary \(\Gamma= \partial\Omega\):

*f*and the Dirichlet data \(\psi_{0}\) are given. Let \(L^{2}(\Omega)\), be the functional space of measurable functions on Ω that are square integrable in Ω and \({H}^{1}(\Omega) = \{u \in L^{2}(\Omega), \nabla u \in L^{2}(\Omega )^{2}\}\) the Hilbert space endowed with the norm \(\|u\|^{2}_{{H}^{1}(\Omega)} = \|u\|^{2}_{\Omega} + |u|^{2}_{{H}^{1}(\Omega)}\) where \(|u|^{2}_{{H}^{1}(\Omega)} = \| \nabla u \|^{2}_{\Omega}\). Let \(\Omega^{\mathrm{in}}\subset\Omega\) be a subdomain with \(\Omega ^{\mathrm{in}}\cap \Gamma= \emptyset\) and \(\Omega^{\mathrm{ex}}= \Omega\setminus \Omega^{\mathrm{in}}\) the complement of \(\Omega^{\mathrm{in}}\) in Ω. Further, the boundary of \(\Omega^{\mathrm{in}}\), \(\gamma:= \partial\Omega^{\mathrm{in}}\), is the interface between \(\Omega^{\mathrm{ex}}\) and \(\Omega^{\mathrm{in}}\). To formulate (7) as a variational problem in a domain decomposition framework, let us introduce the functional space

*γ*. Then we introduce the space

- 1.
Assembling of the stiffness matrices associated to \(\mathsf {a}^{\mathrm{ex}}_{{s}}(\cdot,\cdot)\) and \(\mathsf{a}^{\mathrm {in}}_{{t}}(\cdot,\cdot)\) involves products of basis functions defined on different meshes. Similarly, assembling of the load vectors corresponding to \(\mathsf {\ell}^{\mathrm{ex}}_{{s}}(f,\cdot)\) and \(\mathsf{\ell}^{\mathrm {in}}_{{t}}(f,\cdot)\) involves integration over intersections of elements from different meshes.

- 2.
The stability of \(\text{MEM}_{s,t}\) requires the projections \(\Pi^{\mathrm{ex}}\) and \(\Pi^{\mathrm{in}}\) to be stable in \(H^{\frac{1}{2}}\). The obvious choice of \(L^{2}\) projections involves again surface integrals of products of basis functions defined on different meshes.

In the following we will introduce two mortar-like mappings different from the standard \(L^{2}\) projection, that allow to choose \(s=t=0\) in (9) and hence avoid the expensive assembling of the stiffness matrix for basis functions on two different meshes.

*E*is the trivial extension operator. Let us introduce two mappings \(\Pi^{\mathrm{ex}}_{f} \psi^{\mathrm{in}}\) for \(\psi^{\mathrm{in}}= \psi^{\mathrm{in}}_{\circ}+ \psi^{\mathrm{in}}_{\gamma}\), with \(\psi^{\mathrm{in}}_{\circ}\in \mathcal{V}^{\mathrm{in}}_{\circ}\), \(\psi^{\mathrm{in}}_{\gamma}\in \mathcal{V}^{\mathrm{in}}_{\gamma}\) and \(\Pi^{\mathrm{in}} _{f} \psi^{\mathrm{ex}}\) for \(\psi^{\mathrm{ex}}= \psi^{\mathrm {ex}}_{\circ}+ \psi^{\mathrm{ex}}_{\gamma}\), with \(\psi^{\mathrm{ex}}_{\circ}\in\mathcal{V}^{\mathrm{ex}}_{\circ}\), \(\psi^{\mathrm{ex}}_{\gamma}\in\mathcal{V}^{\mathrm{ex}}_{\gamma}\). The mapping \(\Pi^{\mathrm{ex}}_{f}\) is defined as:

### 3.2 MEM-M Galerkin formulation for the equilibrium problem

The MEM-M Galerkin formulation (12) is similar to the formulation in [21] with the difference that we treat here the scaling parameter *λ* as an unknown in order to match with a prescribed total plasma current \(I_{p}\). Working here with the *free-boundary equilibrium problem with fixed plasma current* is due to the fact that in most applications the total plasma current \(I_{p}\) will be a prescribed quantity.

We want to stress that an implementation of the MEM-M Galerkin formulation (12) requires quadrature rules for the approximation of the integrals. Moreover, since the problem is non-linear we also need an iteration scheme. It is the non-linearity due to the unknown plasma domain \(\mathcal {P}(\psi^{\mathrm{in}})\) that makes these steps non-standard, and we refer to [3, 20] for the details for FE spaces over triangular meshes. The case of FE over Cartesian meshes is presented in [21, Sect. 4.4], and we mention here only that we use a Newton method for the fully discretized system as opposed to a discretization of the Newton method for the weak formulation. In many cases these two approaches are identical, but in the free-boundary setting it makes a difference. The difference is not too important for finding approximate solutions of (12) but it is essential for the optimal control formulation in the following section.

Let \(\mathbf {y}^{\mathrm{ex}}\) and \(\mathbf {y}^{\mathrm{in}}\) represent the vector of the values of degrees of freedom of \(\psi^{\mathrm{ex}}\in\mathcal{V}^{\mathrm{ex}}\) and \(\psi^{\mathrm{in}}\in \mathcal{V}^{\mathrm{in}}\). Then we have the decomposition \(\mathbf {y}^{\mathrm{ex}}= (\mathbf {y}^{\mathrm{ex}}_{\circ}, \mathbf {y}^{\mathrm {ex}}_{\gamma})\) and \(\mathbf {y}^{\mathrm{in}}= (\mathbf {y}^{\mathrm{in}}_{\circ}, \mathbf {y}^{\mathrm{in}}_{\gamma})\), where \(\mathbf {y}^{\mathrm{ex}}_{\circ}\) (resp., \(\mathbf {y}^{\mathrm{in}}_{\circ}\) ) and \(\mathbf {y}^{\mathrm {ex}}_{\gamma}\) (resp., \(\mathbf {y}^{\mathrm{in}}_{\gamma}\) ) are the degrees of freedom in \(\mathcal{V}^{\mathrm{ex}}_{\circ}\) (resp., \(\mathcal{V}^{\mathrm{in}}_{\circ}\)) and \(\mathcal{V}^{\mathrm {ex}}_{\gamma}\) (resp., \(\mathcal{V}^{\mathrm{in}}_{\gamma}\)). Further let **u** represent the vector of coil currents. The weak formulation (12) yields

*λ*in one vector \(\mathbf {y}:=(\mathbf {y}^{\mathrm{ex}}_{\circ}, \mathbf {y}^{\mathrm {ex}}_{\gamma}, \mathbf {y}^{\mathrm{in}}_{\circ}, \mathbf {y}^{\mathrm{in}}_{\gamma}, \lambda)\) we can recast (14) in the form

### 3.3 The inverse problem for tuning plasma equilibria

*N*unknowns of the poloidal flux approximation in \(\mathcal{V}^{\mathrm {ex}}\) and \(\mathcal{V}^{\mathrm{in}}\). The components of the control variable \(\mathbf {u}\in\mathbb {R}^{M}\) are the currents in the

*M*different coils.

_{y}and

_{u}denote differentiation with respect to

**y**and

**u**, respectively. Splitting \(\mathbf {p}= ( \mathbf {p}^{\mathrm{ex}}_{\circ}, \mathbf {p}^{\mathrm{ex}}_{\gamma},\mathbf {p}^{\mathrm{in}}_{\circ}, \mathbf {p}^{\mathrm{in}}_{\gamma}, \mathbf {p}_{\lambda})^{T}\) analogously to \(\mathbf {y}=(\mathbf {y}^{\mathrm {ex}}_{\circ}, \mathbf {y}^{\mathrm{ex}}_{\gamma}, \mathbf {y}^{\mathrm {in}}_{\circ}, \mathbf {y}^{\mathrm{in}}_{\gamma}, \lambda)^{T}\) we can provide the more detailed depictions

*inexact*Newton method, than an

*exact*Newton method.

**Y**is equal to the sensitivity \(\mathbf {y}_{\mathbf {u}}(\mathbf {u}) = -\mathbf {b}_{\mathbf {y}}^{-1}(\mathbf {y}(\mathbf {u}),\mathbf {u}) \mathbf {b}_{\mathbf {u}}(\mathbf {y}(\mathbf {u}),\mathbf {u})\). In general, it is recommended to avoid the explicit calculation of these sensitivities and adjoint methods were introduced for exactly that purpose. However, as we have a very few number of control parameters, this is not an issue. To motivate Algorithm 1 we introduce

The iteration scheme (18) for the constraint optimization problem (16) involves first order derivatives of \({\mathbf {b}}(\mathbf {y},\mathbf {u})\) and first and second order derivatives of \(R(\mathbf {u})\) and \({C}(\mathbf {y})\). The derivative \({\mathbf {b}}_{\mathbf {y}}(\mathbf {y},\mathbf {u})\) is already available from the Newton iterations for (15) and as we have explicit expressions for \({\mathbf {b}}(\mathbf {y},\mathbf {u})\), \(R(\mathbf {u})\) and \({C}(\mathbf {y})\) that are algebraic in **u** and **y** we can also provide the remaining derivatives.

## 4 Results and discussion

We highlight that, to our knowledge, there is no theory yet available, that justifies rigorously convergence of the MEM-M. Only for lowest order Lagrangian elements we have a convergence assertion in \(L^{\infty}\) [24, Theorem 1]. Therefore, we present here first an experimental validation of the MEM-M, and continue afterwards, in Sect. 4.2, with the application.

All the numerical results are based on the MATLAB/Octave library FEEQS.M developed by one of the authors. This library utilizes in large parts vectorization. Therefore, the running time is comparable to C/C++ implementations.

### 4.1 Experimental validation

For validation of the MEM, we consider a rectangular domain \(\Omega= [-1,1]^{2}\) and define \(\Omega^{\mathrm{in}}\) as the polygon with vertices \((-0.125, 0.5)\), \((0.375, 0.25)\), \((0.375, -0.375)\), \((0, -0.5)\), \(( -0.375, -0.375)\), and \((-0.5, 0.25)\). The meshes \(\mathcal{T}^{\mathrm{in}}\) and \(\mathcal {T}^{\mathrm{ex}}\) for the interior and exterior domain will be a Cartesian mesh and a triangular mesh. For simplicity we prefer to take \(\Omega^{\mathrm{ex}}_{h}= \Omega ^{\mathrm{ex}}= \Omega \setminus\Omega^{\mathrm{in}}\). For the numerical test, we take \({D}=1\) and choose the data \(f(x,y)\) and \(\psi_{0}\) in agreement with \(\psi(x,y) = \cos(\pi x) \sin(\pi y) \) as solution of (7). If \(h_{\mathrm{ex}}\) (\(h_{\mathrm{in}}\)) is the maximal diameter of elements in \(\mathcal{T}^{\mathrm{ex}}\) (\(\mathcal{T}^{\mathrm {in}}\)), and \(p_{\mathrm{ex}}\) (\(p_{\mathrm{in}}\)) the local polynomial degree of the FE spaces \(\mathcal{V}^{\mathrm{ex}}\) (\(\mathcal {V}^{\mathrm{in}}\)), one has optimal convergence if, for a smooth solution, the approximation error in the \(H^{1}(\Omega^{\mathrm{ex}}_{h})\) and \(H^{1}(\Omega^{\mathrm {in}}_{h})\)-norms behaves as \(O(h^{p-1})\), with \(h = \max(h_{\mathrm{ex}},h_{\mathrm{in}})\) and \(p = \min (p_{\mathrm{ex}},p_{\mathrm{in}})\) (in \(L^{2}(\Omega^{\mathrm{ex}}_{h})\) and \(L^{2}(\Omega^{\mathrm {in}}_{h})\)-norms one dares to obtain \(O(h^{p})\)). To keep the presentation as clear as possible we show in the following figures always the maximum between the error in \(\Omega^{\mathrm {ex}}_{h}\) and that in \(\Omega^{\mathrm{in}}_{h}\).

We consider two different pairings of FE spaces \(\mathcal{V}^{\mathrm {ex}}\)–\(\mathcal{V}^{\mathrm{in}}\). The first denoted with P1-Q1 uses lowest order linear FEs over \(\mathcal{T}^{\mathrm{ex}}\) and lowest order bilinear FEs over \(\mathcal{T}^{\mathrm{in}}\). The second pair, denoted with P2–Q3 uses quadratic FEs over \(\mathcal{T}^{\mathrm{ex}}\) and bicubic FEs over \(\mathcal{T}^{\mathrm{in}}\). The elements of P2-Q3 are not only continuous on \(\Omega^{\mathrm {in}}_{h}\) and \(\Omega^{\mathrm{ex}}_{h}\) but have also continuous gradients on \(\Omega^{\mathrm{in}}_{h}\).

Next, we study the convergence rates for MEM-M when \(\Omega^{\mathrm {in}}_{h}\) has a large overlap with \(\Omega^{\mathrm{ex}}_{h}\). For this we fix \(\Omega ^{\mathrm{in}}_{h}\) to be the square \([-0.76,0.65]\times[-0.76,0.78]\) (see Fig. 8, right). Note that both \(\gamma^{\mathrm{ex}}\) and \(\gamma^{\mathrm{in}}\) remain fixed during the refinements in \(\Omega^{\mathrm{in}}_{h}\). Once again, the MEM-M yields optimal convergence rate in the \(H^{1}\) norm (see Fig. 10). Moreover in the case of larger overlap we observe even optimal convergence in the \(L^{2}\)-norm. There is no qualitative difference between MEM-M based on the \(L^{2}\)-projection or on the interpolation. More detailed numerical tests of the MEM-M can be found in [34].

*s*and

*t*set to zero (MEM

_{0,0}), the convergence rates in the \(H^{1}\) and \(L^{2}\) norms are not optimal in the case of minimal overlap between \(\Omega^{\mathrm {in}}_{h}\) and \(\Omega^{\mathrm{ex}}_{h}\) (see Fig. 11). The MEM

_{0,0}does not yield convergence in the case of a large overlap between \(\Omega^{\mathrm{in}}_{h}\) and \(\Omega^{\mathrm{ex}}_{h}\).

Our experimental results for MEM_{0,0} are not very surprising. All available convergence assertions assume \(s+t=1\), which leads to the cumbersome integration over cut elements, that we prefer to avoid.

### 4.2 A case study

*ψ*to be constant on the prescribed points \(\mathbf {x}_{1}, \ldots,\mathbf {x}_{N_{\mathrm{desi}}}\). The objective functional \({C}_{2}\) forces

*ψ*to have a stationary point at \(\mathbf {x}_{0}\). Using \({C}_{1}\) alone for the formulation of the optimal control problem (16), is the standard approach to find a certain configuration of plasma currents that give an equilibrium boundary that is close to the prescribed points \(\mathbf {x}_{i}\).

*w*. We denote by

**y**is the vector of degrees of freedom of \((\psi^{\mathrm{ex}},\psi^{\mathrm{in}}) \in \mathcal{V}^{\mathrm{ex}} \times\mathcal{V}^{\mathrm{in}}\). The current density profiles (6) use the parameter values \(\alpha= 1\), \(\beta =1.2\), \(\gamma= 1.1\) and \(r_{0}=6.65\). The total plasma current has value \(I_{p} = 11 10^{6} A\). The iteration stops if the relative increments in Algorithm 1 are smaller than \(tol= 10^{-11}\). In Fig. 13 we see that for a sufficiently large value of

*w*our approach is capable to create snowflake-like plasma equilibrium configurations.

## 5 Conclusions

Handling the heat where the plasma touches the vessel wall is one of the outstanding challenges for magnetically confined fusion energy research. Indeed, the predicted heat load on the ITER vessel walls will be greater than that on the soil beneath a launching rocket. Current experiments are trying to find magnetic field configurations which can provide the most effective heat load reduction. One of such advanced configurations is the so called snowflake configuration, where the plasma boundary is the flux contour-line that passes through a degenerated saddle point of the poloidal flux.

The main control parameters for shaping the contour lines of the poloidal flux, and hence for shaping the form and position of a plasma equilibrium in a tokamak are the currents in the surrounding poloidal field coils. Optimal control formulations combined with finite element methods are an obvious and fairly established mean [3, 20] to determine currents that ensure a certain desired form and position of the plasma. But the low regularity, e.g. lack of well-defined pointwise derivatives, of standard \(H^{1}\)-conforming finite elements, seems to be an obstacle to define good objective functions for finding the snowflake configurations characterized by degenerated saddle points.

We therefore, presented here an extension of this approach that combines the optimal control formulation with a mortar-type FE method. The mortar-type FE method has the advantage that we can introduce higher order regular FE in the places where we have objective functions involving point wise values of flux derivatives. This is achieved in combining FE on Cartesian meshes with FE on triangular meshes.

The examples for the tokamak CFETR approve the viability and flexibility of the presented approach.

## Declarations

### Acknowledgements

We thank the anonymous reviewers for their valuable input. We also thank Jean-Charles Gilbert for the critical comments on Sect. 3.3.

### Availability of data and materials

Not applicable.

### Funding

Not applicable.

### Authors’ contributions

All authors contributed to the writing of the final version of this paper. The main concept of the proposed method was developed by HH, FR contributed with her expertise in high order FEs and mortar element methods and XS provided simulations for CFETR and HL-2M. All authors read and approved the final manuscript.

### Competing interests

The authors declare that they have no competing interests.

**Open Access** This article is distributed under the terms of the Creative Commons Attribution 4.0 International License (http://creativecommons.org/licenses/by/4.0/), which permits unrestricted use, distribution, and reproduction in any medium, provided you give appropriate credit to the original author(s) and the source, provide a link to the Creative Commons license, and indicate if changes were made.

## Authors’ Affiliations

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