Xuebing Peng

Design feasibility analysis of the robot for EAST tokamak flexible in-vessel inspection

EAST is a full superconducting tokamak fusion experiment device with ‘D’ shape vacuum vessel and toroidal coils and actively cooled plasma facing components (PFCs), which aims at studying the scientific and engineering issues under steady state operation. It is difficult but very important to know well the operating state of the components in the vacuum vessel during plasma operation campaign for the help of guiding operation plan and understanding physical phenomena. A robotic system (called flexible in-vessel inspection robot, FIVIR) is proposed to inspect the surface of plasma facing components to know the performance of first wall, such as erosion and desquamation. Because of the geometric requirement and the intensity of ports usage arrangement, two FIVIRs are planned installing toroidal symmetry on EAST. Each FIVIR is a series-wound robot with ten degrees of freedom consisting of main robot and end-effector. The main robot has the primary function that transports the end-effector and associated process tools into the vacuum chamber in the equatorial plane at R=1.94m and back to the storage port, which has the benefit that easy control of the FIVIR and easy position calculation of the end-effector. All the joints in the FIVR are driven by actuators. The workspace of the robot is analyzed to see whether it can reach any point of the plasma facing surface in the distance of 15mm and the range of ±90°along toroidal direction.

Evaluation of performance for the EAST upgraded divertor targets during type I ELMy H-mode

The long-pulse high-confinement (H-mode) plasma regime is considered to be a preferable scenario in future fusion devices, and in the period of normal operation during H-mode, edge-localised modes (ELMs) are one of the most serious threats to the performance and capability of divertor targets. The EAST recently achieved a variety of H-mode regimes with ELMs. For the purpose of studying the performance of the EAST upgraded divertor during type I ELMs, a series of simulations were performed by using three-dimensional (3D) finite element code. To make a visible outcome of the direct ELM impact on the divertor targets, a preliminary evaluation system with three indices to exhibit the influence has been developed. The indices that comprise temperature evolution, thermal penetration depth and crack initiation life, which could reveal the process of micro-crack formation, are calculated in both low and high-power scenarios for type I ELMs. The initial results indicate that the transient heat load has a significant influence in a very short thickness layer along the direction perpendicular to the plasma-facing surface throughout its duration. The conclusion could offer a pertinent guide to the next-step high-power long-pulse operation in EAST and would also be helpful for scientifically studying the damage and fatigue mechanism of the divertor in ITER and future fusion power reactors.

Conceptual Design of High Temperature Water-Cooled Divertor Plasma-Facing Unit for Fusion Reactor

Tokamak, a type of fusion experimental device, is considered as the most promising device on which net fusion power could be outputted for electricity production, i.e., the fusion reactor. Divertor, as one of the core components in Tokamak, has to sustain very high heat flux from high temperature plasma, up to tens of MWs, so that the design of divertor plasma-facing unit (PFU) is quite important. Up to now, the most advanced mature PFU technology is the ITER W/Cu PFU, which is a monoblock structure with tungsten, Cu and CuCrZr as the plasma-facing material, interlayer and heat sink, respectively. However, due to high activation of Cu element by neutron irradiation, CuCrZr is not appropriate for the material of heat sink anymore for future fusion reactor. In the paper, a design of PFU with a low activation material (named CLAM, China Low Activation Martensitic steel) as the heat sink was proposed based on ITER monoblock structure, i.e., W/CLAM PFU. With the temperature operation window as the design limit, thermal analysis of W/CLAM PFU was done initially with the same structural dimensions as ITER monoblock. Then the structural dimensions were optimized for the purpose of improving the heat loads removing capability, more than 10 MW/m2. As a consequence, thermal stresses were calculated for the W/CLAM PFU, where stress in CLAM heat sink was found beyond the material limit. The issue was discussed and the solution was preliminary proposed. This can provide the necessary theoretical basis for improvement of heat flux handling capacity to divertor PFU in future application.

Plasma facing conponents of EAST

EAST plasma facing components (PFCs) have the function of protecting the vacuum vessel, heating systems and diagnostic components from the plasma particles and heat loads, and also additional to this particles and heat loads handling. They are installed in the vacuum vessel together with in-vessel coils, cryopump and diagnostic components. The design, fabrication and assembly have been finished. The PFCs are designed up-down symmetry to accommodate with both double null and single null plasma configuration. All PFCs use graphite tile for plasma facing surfaces affixed to copper alloy heat sink. A special deep hole drilling technology was developed to drill cooling channels directly on heat sink for high efficient heat removal. All Heat sink are installed onto the base alignment rails through stainless steel supports. As the benchmark of assembly for PFCs, the base rails are installed and measured precise based on a new alignment method integrating the optical instruments and a mechanical template. And so is a mechanical check template for checking the surface of first wall. As indicated, all the first wall components were fabricated and assembled successfully and meet the design requirement for the plasma operation.

Structural Analysis of Wendelstein 7-X Nonplanar Coil Type 1 in Self-Field Test Conditions

Abstract The Wendelstein 7-X (W7-X), the largest modular stellarator in the world, is in operation at Max Planck Institute for Plasma Physics in Greifswald, Germany. The magnet system of the W7-X consists of 50 nonplanar and 20 planar superconducting coils, which are supported by a massive central support structure. All superconducting coils have been subjected to gravity and electromagnetic force due to the interaction between self-field and the coil current in the test conditions in Saclay, France. Each coil is equipped with a few mechanical sensors. Some of the sensors have indicated considerable deviation from the numerical prediction. The nonplanar coil Type 1 is an example of such deviations. This technical note presents structural analyses performed to verify the numerical modeling by checking the stresses in the measurement points. In order to find the reason from the finite element model, three factors are considered: mesh refinement, increasing the region of mesh refinement, and changing the element supports. The results show that the three factors have no impact on the stresses at the measurement points. Finally, special attention has been paid to the sensors during commissioning of the W7-X, which revealed that lack of information about boundary conditions or temperature fluctuations could be the reason for the original discrepancies.

Preliminary Design and Verification of Divertor Module of CFETR System Code

A system code for integrated simulation and optimization is being developed for China Fusion Engineering Test Reactor (CFETR). The main function of CFETR system code is to achieve consistent physical objectives with achievable engineering parameters. As one of the most important works for tokamak, many aspects must be considered in the divertor design, such as high-heat loads removal, particle pumping, electromagnetic forces sustaining, remote handling compatibility, and so on. Inevitably, it is difficult to ensure all parameters are optimum due to the fact that some divertor parameters may be conflicting or highly interdependent. An underdeveloped divertor module code could address these issues. The purposes of developing such a module code are to explore key elements of the divertor system, simplify the design process, and determine the main engineering parameters by iteration of a self-consistent workflow. The divertor module code incorporates submodules with input and output ports, material and criterion databases, and simulation and optimization CAE tool-box, including CATIA, ANSYS, and OPTIMUS. This paper presents a typical design workflow for divertor module and a simplified simulation case to validate the feasibility of this workflow.

Preliminary design and verification of divertor module for CFETR system code

An integrated simulation and optimization system code is being developed for China Fusion Engineering Test Reactor (CFETR). The main function of the CFETR system code is to achieve consistent physical objectives with achievable engineering parameters. As one of the most important works in a tokamak R&D, the design of divertor must consider many aspects, such as high heat loads removal, particles pumping, electromagnetic forces sustaining, remote handling compatibility, etc. However, it is difficult to ensure all parameters are optimum due to some divertor parameters may be conflict or highly interdependent. An underdeveloped divertor module code may address these issues. The purposes of developing this module code are to explore key elements of divertor system, simplify design process and determine main engineering parameters by iteration of a self-consistent workflow. The divertor module code incorporates submodules with input, output, material and criterion database, simulation and optimization CAE tool-box including CATIA, ANSYS and OPTIMUS. This paper also presents some simulation work to verify this workflow.