Mingzhun Lei

Conceptual Design of a Helium-Cooled Ceramic Breeder Blanket for CFETR

Abstract The China Fusion Engineering Test Reactor (CFETR) is a superconducting tokamak reactor proposed by the China National Integration Design Group. The aim is to develop China’s next-step fusion device. A helium-cooled ceramic breeder (HCCB) blanket concept has been put forward by the blanket integration design team of the Institute of Plasma Physics of the Chinese Academy of Sciences. The present blanket configuration is based on the upper port dimension and maintenance scheme. The HCCB blanket comprises a U-shaped first wall, cap, breeder unit, middle plate, and back plate. The breeder unit includes a cooling plate, beryllium pebble bed, and lithium silicate pebble bed. The structure of the HCCB blanket concept is obviously different from the ITER HCCB test blanket module of China. The CFETR HCCB blanket will meet the requirements of CFETR long-pulse or steady-state operation with a duty cycle time greater than 0.3. The maintenance scheme of the blanket is introduced in this paper. Three-dimensional neutronic analysis results show that the tritium breeding ratio can satisfy the design requirement. Besides, the thermal-hydraulic behavior of the first wall has been studied using ANSYS CFX code.

In-vessel Components

Why we need various in-vessel components in a tokamak? Firstly it is necessary to know the goal and characteristics of a tokamak experimental device or fusion reactor. The device is used for the study of high temperature (tens of millions centigrade degrees) plasma and controllable nuclear fusion or to generate some Gigawatt net fusion energy. So, obviously two critical issues would come up. One is how the components facing the plasma can survive or how to protect the permanent components installed in the vacuum vessel (VV) from so high temperature plasma and the other is how to extract the nuclear heat from the hot plasma in the VV and transfer it to the outside of the device. These are what plasma facing components (PFCs) basically do. In addition to the PFCs, there are also many other components installed in the VV for the purpose of measuring the parameters of the plasma and magnetic field, of diagnosing the behavior of the plasma and the PFCs, etc. All these components installed in the tokamak VV are the so-called in-vessel components.

Preliminary Thermal Mechanical Analysis of the Equatorial Thermal Shield for ITER

Thermal analysis of the equatorial thermal shield for ITER is conducted in order to confirm that the cooling tube design was reasonable under both the plasma operational and the baking operational conditions. The structural performance was analyzed by means of the finite element software ANSYS. A comparison of the results with design requirements shows that the results of the simulation are within allowable design requirements, which indicates the feasibility and reliability of the equatorial thermal shield structure.

Analysis and optimization of the manifolds of the HCCB blanket for CFETR

ABSTRACT Thermal-hydraulic performance is a challenging issue in helium-cooled ceramic breeder (HCCB) blanket design due to the extremely complicated working environment and the strict limits of materials temperature. The heat loads deposited on the HCCB blanket comprises of severe surface heat flux from plasma and the volumetric nuclear heat from neutron irradiation, which can be exhausted by the built-in cooling channels of the components. High pressure helium with 8 MPa, distributed from the coolant manifolds is employed as coolant in the blanket. The design and optimization of the manifolds configuration was performed to guarantee the accurate flow control of helium coolant. The flow distribution in the coolant manifolds was investigated based on the structural improvement of manifolds aiming at overall uniform mass flow rates and better flow streamline distribution without obvious vortexes. The peak temperature of different functional materials in the blanket under normal operating condition is below allowable material limits. It is found that the components in the current blanket module could be cooled effectively under the intense thermal loads due to the updated design and optimization analysis of manifolds.

The Design and Test of a Be Window for the ITER Radial X-Ray Camera

Beryllium (Be) window is a key component of the ITER radial X-ray camera (RXC). The Be window presented in this paper has a mechanical clamping structure, the thickness of the Be foil is 80 μm, and the X-ray threshold of the 80 μm Be foil is 1.24 keV. A honeycomb support is designed and applied to strengthen the Be foil to prevent it from breakage when it is exposed to 1 atm perssure. Based on analysis results, the hole diameter of the support is chosen as 4 mm. A metal seal is used to isolate the vacuum on two sides of the Be window, the hollow metal sealing ring ensures the He leakage rate of the Be window being lower than 6×10−10 Pam3s−1. Baking (240°C, 2 h) and vibration(3.3 Hz, 2 h) tests are carried out and the feasibility of the Be windows sealant in these situations is tested. The Be window has good stability that can save maintenance cost as well as enhancing the safety of the RXC.

Thermal-Structural Coupled Analysis of ITER Torus Cryo-Pump Housing

An ITER torus cryo-pump housing (TCPH), which encloses a torus cryo-pump, is connected to a vacuum vessel (VV) by a set of associated double bellows. There are complicated loads due to two different operating states (pumping and regeneration) and foreseeable accidents with the cryo-pump. This paper describes a thermal-structural coupled analysis of the present TCPH according to the allowable stress criteria of RCC-MR, in which the worst cases and outcomes of various load combinations are obtained. Meanwhile, optimization of the structure has been carried out to obtain positive analysis results and an adequate safety margin.