Nuclear Fusion Tokamak Research Articles

Effects of H on Electronic Structure and Ideal Tensile Strength of W: A First-Principles Calculation

We investigate the structure, energetics, and the ideal tensile strength of tungsten (W) with hydrogen (H) using a first-principles method. Both density of states (DOS) and the electron localization function (ELF) reveal the underlying physical mechanism that the tetrahedral interstitial H is the most energetically favorable. The first-principles computational tensile test (FPCTT) shows that the ideal tensile strength is 29.1 GPa at the strain of 14% along the [001] direction for the intrinsic W, while it decreases to 27.1 GPa at the strain of 12% when one impurity H atom is embedded into the bulk W. These results provide a useful reference to understand W as a plasma facing material in the nuclear fusion Tokamak.

First-principles study on dissolution and diffusion properties of hydrogen in molybdenum

Employing a first-principles method, we have investigated dissolution and diffusion properties of hydrogen (H) in molybdenum (Mo), one of potential candidates for plasma facing materials in a nuclear fusion Tokamak. We show that single H atom is energetically favorable sitting at the tetrahedral interstitial site (TIS) instead of octahedral interstitial site and diagonal interstitial site. This can be confirmed by the electron localization function result. Bader charge analysis suggests that the bonding between H and surrounding Mo is mainly ionic mixed with slight covalent component. Double H atoms tend to be paired up at the two neighboring TIS’s along the 〈1 1 0〉 direction with the distance of ∼0.221 nm and the binding energy of 0.03 eV. This suggests a weak attractive interaction between H atoms, with the implication that self-trapping of H and thus formation of the H 2 molecules are quite difficult in an intrinsic Mo environment. We demonstrate that the diffusion barrier of H that jumps between the TIS’s is 0.16 eV, and the dissolved concentration of H in the intrinsic Mo is 2.6 × 10 −8 at a typical temperature of 600 K. The diffusion coefficients of H, D, and T are different due to the different masses, which are calculated to be 1.3 × 10 −7 m 2/s, 9.2 × 10 −8 m 2/s, and 7.5 × 10 −8 m 2/s at 600 K.

The ideal tensile strength and deformation behavior of a tungsten single crystal

We employ first-principles total energy method based on the density functional theory with the generalized gradient approximation to investigate the ideal tensile strengths of a bcc tungsten (W) single crystal systemically. The ideal tensile strengths are shown to be 29.1, 49.2 and 37.6GPa for bcc W in the [001], [110] and [111] directions, respectively. The [001] direction is shown to be the weakest direction due to the occurrence of structure transition at the lower strain and the [110] direction is strongest. The results can provide a useful reference for W as a PFM in the nuclear fusion Tokamak.

Effect of He on the structure and bonding properties of W: A first-principles computational tensile test

Using a first-principles computational tensile test (FPCTT), we have investigated the effect of helium (He) on the structure and bonding properties of tungsten (W), which is a promising plasma-facing material in nuclear fusion Tokamak. Density of states results reveal the underlying reason that the substitutional site for He is the most energetically favorable, while the tetrahedral interstitial site is more favorable than the octahedral interstitial one. The FPCTT shows that the ideal tensile strength is 29.1GPa at the strain of 14% along the [001] direction for intrinsic W, while it decreases to 28.2GPa at the same strain when one impurity He atom is introduced. A local bond-breaking region around He forms in the tensile process due to the presence of He, which suggests He will have a large effect on the bonding properties of W.

Study of the theoretical tensile strength of Fe by a first-principles computational tensile test

This paper employs a first-principles total-energy method to investigate the theoretical tensile strengths of bcc and fcc Fe systemically. It indicates that the theoretical tensile strengths are shown to be 12.4, 32.7, 27.5 GPa for bcc Fe, and 48.1, 34.6, 51.2 GPa for fcc Fe in the [001], [110] and [111] directions, respectively. For bcc Fe, the [001] direction is shown to be the weakest direction due to the occurrence of a phase transition from ferromagnetic bcc Fe to high spin ferromagnetic fcc Fe. For fcc Fe, the [110] direction is the weakest direction due to the formation of an instable saddle-point ‘bct structure’ in the tensile process. Furthermore, it demonstrates that a magnetic instability will occur under a tensile strain of 14%, characterized by the transition of ferromagnetic bcc Fe to paramagnetic fcc Fe. The results provide a good reference to understand the intrinsic mechanical properties of Fe as a potential structural material in the nuclear fusion Tokamak.

Space-State Modeling and Control of Tokamak Reactors

Nuclear fusion has the potential to produce unlimited, clean energy, which presents itself as a reliable energy supply but it also helps to stop the threat of climate change that faces the world nowadays. However, to sustain the pulse duration long enough to produce the necessary energy, new controls have to be developed, composing a new application area of Control Engineering, with new and interesting challenges for the control community. In this sense, this paper deals with the modeling of tokamak nuclear fusion reactors. In order to control the creation of unstable modes in fusion processes, it is necessary to derive numerical models suitable for control strategies. The model presented in this paper addresses flux and energy conservation issues, discussing the mechanisms behind the creation of uncontrollable modes. The dynamics of the system is given by means of the energy functions which are solved for the currents in the structure, plasma current and plasma position. Thus, the equations for the state variables are derived based on the Hamiltonian equation of motion. In order to solve this system numerically, this model is linearized around an operation point by taking a Newton-Raphson step. Besides, the system output is completed by considering the equations for the flux and the poloidal field. Finally, the resulting low-order linear model is modified so as to obtain the corresponding state-space model which is verified by means of numerical experiments.

Boundary-only integral equation approach based on polynomial expansion of plasma current profile to solve the Grad–Shafranov equation

A new type of boundary element method (BEM) has been applied to solve the Grad–Shafranov equation and to give a distribution of magnetic flux function in a tokamak nuclear fusion device. The quantity μ0rjφ related to the plasma current profile is expanded into a two-dimensional polynomial. Using the particular solution of the Grad–Shafranov equation with this inhomogeneous polynomial source and applying Green's second identity, the domain integral related to the plasma current is transformed into an equivalent boundary integral. Domain discretization is not required in this formulation, thus preserving all the advantages of the BEM. Numerical computations of all boundary integrals are only required in the initial stage of the eigenvalue iteration, so that the number of eigenvalue iterations does not hamper the total computing time. Test calculations demonstrated that the present method provides stable and accurate solutions.

Study of thermal and cooling load for KSTAR thermal shield design

In order to derive the detailed design of cryostat thermal shield cryopanel, which plays a role to make the Tokamak Nuclear Fusion Equipment work at both stable and efficient conditions, the commercially available software package FLUENT 5.3, whose accuracy is already verified, was utilized. This study investigated the effects of thermal sources on the temperatures of lid, body, base, and EH-Port cryopanel by the numerical technique whose grid generations cover the solid and gas region of the panel. The physical model of the thermal shield cryopanel is that the 10 mm diameter of the pipe with 1 mm thickness is soldered on the stainless steel panel with 4 mm thickness. The heat fluxes to the panel are assumed to be by thermal radiation in the vacuum space and by conduction through the supporters. The inlet conditions of helium gas are 20 atmospheric pressures and 60 K temperature.

An innovative intelligent system for sensor validation in tokamak machines

A sensor validation strategy based on soft computing techniques to isolate and classify some faults occurring in the measurement system of a Tokamak fusion plant is described. Particular attention is focused on the system used to measure vertical stress in the mechanical structure of a Tokamak nuclear fusion plant during fusion experiments. The strategy adopted is based on a modular structure comprising two stages. The first stage consists of a neural network which acts as a symptom model able to estimate directly some suitable features of the expected sensor responses, thus allowing the most frequently occurring sensor faults to be isolated. The second stage consists of a fault classifier implemented via a fuzzy inference system, in order to exploit the knowledge of the experts. The proposed strategy was validated at the Joint European Torus (JET), on several experiments. A comparison was made with both traditional sensor monitoring techniques and validation performed manually by experts. A great improvement was achieved, in terms of both fault detection and classification capabilities, and the degree of automation achieved.

Analyzing thermal creep strain of a tokamak first-wall steel

It is widely anticipated that one formulation of ferritic steel, HT9, will be utilized in the fabrication of the first wall of a nuclear fusion (tokamak) reactor. In this respect, the thermal creep behavior of this material is an important consideration in the final selection decision. Using short-term experimental data, relations which describe the dependence of the thermal creep strain of this alloy on the operational parameters (e.g., time, stress and temperature) have been obtained. This allows estimates to be made of the long-term (design) values of the thermal creep strain for any specified combination of operational parameters.

Unipolar arcing in tokamak nuclear fusion devices treated as a type of open-circuit corrosion

Conceptual analogies between interfacial events involved in unipolar arcs ( e.g. in tokamak nuclear fusion devices) and in the open-circuit corrosion of metals are delineated, following brief descriptions of the two phenomena. Fundamentally, both involve conjugate electronation and deelectronation reactions occurring on the “cathodic” and “anodic” sites respectively which exist on the same equipotential surface, i.e. the metal wall in the unipolar arcs and the dissolving metal in the active corrosion reaction. Both phenomena involve electrode-electrolyte interfaces across which there is an electrochemical charge transfer but no net current flow, i.e. conjugate cathodic and anodic currents of equal magnitudes are involved.