Design and fabrication of the vacuum vessel for the tokamak fusion test reactor

1982 ◽  
Vol 20 (4) ◽  
pp. 1173-1176 ◽  
Author(s):  
W. G. Reddan
Keyword(s):  
Vacuum Vessel ◽  
Test Reactor

R&D of Full-Scale Partial Vacuum Vessel Mockup for Future Fusion Engineering Test Reactor in China

2015 ◽  
Vol 34 (3) ◽  
pp. 666-670 ◽  
Author(s):  
Ma Jianguo ◽  
Wu Jiefeng ◽  
Liu Zhihong ◽  
Fan Xiaosong
Keyword(s):  
Full Scale ◽  
Vacuum Vessel ◽  
Fusion Engineering ◽  
Test Reactor ◽  
Partial Vacuum

Neutronic Analyses for CFETR With Modular Helium Cooled Lithium Ceramic Blanket

2017 ◽  
Author(s):  
Kun Xu ◽  
Minyou Ye ◽  
Yuntao Song ◽  
Mingzhun Lei ◽  
Shifeng Mao
Keyword(s):  
Fusion Energy ◽  
Time Interval ◽  
Vacuum Vessel ◽  
Tritium Breeding ◽  
Thermal Shield ◽  
Sector Model ◽  
Self Sufficiency ◽  
Radial Profiles ◽  
Single Standard ◽  
Test Reactor

China Fusion Engineering Test Reactor (CFETR) is a superconducting tokamak proposed by national integration design group for magnetic confinement fusion reactor of China to bridge the R&D gaps between ITER and DEMO. Since the launch of CFETR conceptual design, a modular helium cooled lithium ceramic blanket concept had been under development by the blanket integration design team of the Institute of Plasma Physics of the Chinese Academy of Sciences, to complete CFETR in demonstrating its fusion energy production ability, tritium self-sufficiency and the remote maintenance strategy. To validate the feasibility, the neutronic analyses for CFETR with this modular helium cooled lithium ceramic blanket were performed. The 1-D neutronic study for CFETR was done in the first place to give a preliminary and quick demonstration of the overall neutronic performance. Meanwhile, the neutronic analyses for a single standard helium cooled lithium ceramic blanket module were done in several times to give more insight for the material and geometry parameters of intra-module structures. Therefore, the principles for neutronic design and the module level optimized parameters were produced, based on which the design of practical blanket modules planted in tokamak vacuum vessel was completed. In the end, the 3-D neutronic analysis for CFETR was done utilizing the MCNP code, in which the 11.25 degree sector model (consist of blanket modules, manifold, support plate, shield, divertor, vacuum vessel, thermal shield and TF coils) was generated with the McCad automated conversion tool from the reference CAD model for analysis, the bi-dimensional (radial and poloidal) neutron source map was plugged via general source definition card to stimulate the D-T fusion neutrons. The concerned neutronics parameters of CFETR, mainly including the tritium breeding ratio to characterize tritium self-sufficiency, the energy multiplication factor to characterize power generation, as well as, the inboard mid-plane radial profiles of neutron flux densities, helium production rate, displacement damage rate and the energy deposition to characterize the shielding performance, were produced. In principle, the neutronics performance of CFETR with modular helium cooled lithium ceramic blanket is promising. The tritium breeding capability meets the design target and, by referring to that for ITER and the EU DEMO fusion power plant, the inboard mid-plane shielding is effective to fulfill the radiation design requirement of the superconducting TF-coils, resulting in a compulsory warm-up time interval of ∼2 FPY for TF-coils. The nuclear heating loads to other CFETR components were generated. As an outcome of this work, the applicability of McCad on CFETR neutronic modeling is demonstrated.

scholarly journals Development of Vacuum Vessel Design and Analysis Module for CFETR Integration Design Platform

2016 ◽  
Vol 2016 ◽  
pp. 1-10 ◽  
Author(s):  
Chen Zhu ◽  
Minyou Ye ◽  
Xufeng Liu ◽  
Shenji Wang ◽  
Shifeng Mao ◽  
...  
Keyword(s):  
Engineering Design ◽  
Main Idea ◽  
Fe Analysis ◽  
Design Analysis ◽  
Future Application ◽  
Vacuum Vessel ◽  
Development Environment ◽  
Analysis Workflow ◽  
Fusion Engineering ◽  
Test Reactor

An integration design platform is under development for the design of the China Fusion Engineering Test Reactor (CFETR). It mainly includes the integration physical design platform and the integration engineering design platform. The integration engineering design platform aims at performing detailed engineering design for each tokamak component (e.g., breeding blanket, divertor, and vacuum vessel). The vacuum vessel design and analysis module is a part of the integration engineering design platform. The main idea of this module is to integrate the popular CAD/CAE software to form a consistent development environment. Specifically, the software OPTIMUS provides the approach to integrate the CAD/CAE software such as CATIA and ANSYS and form a design/analysis workflow for the vacuum vessel module. This design/analysis workflow could automate the process of modeling and finite element (FE) analysis for vacuum vessel. Functions such as sensitivity analysis and optimization of geometric parameters have been provided based on the design/analysis workflow. In addition, data from the model and FE analysis could be easily exchanged among different modules by providing a unifying data structure to maintain the consistency of the global design. This paper describes the strategy and methodology of the workflow in the vacuum vessel module. An example is given as a test of the workflow and functions of the vacuum vessel module. The results indicate that the module is a feasible framework for future application.

Preliminary Neutronics Design and Analysis of the Bit Helium Cooling Ceramics Blanket for CFETR

2013 ◽  
Author(s):  
Jia Li ◽  
Songlin Liu ◽  
Xuebing Ma ◽  
Yong Pu ◽  
Xiangcun Chen
Keyword(s):  
Calculation Model ◽  
Vacuum Vessel ◽  
Concept Design ◽  
Reactor Model ◽  
Plasma Parameters ◽  
Blanket Design ◽  
Self Sufficiency ◽  
Nuclear Heat ◽  
Tube Type ◽  
Test Reactor

CFETR is a Tokamak fusion engineering test reactor whose concept design is being developed in China. It is a key issue for breeding blanket design to attain tritium self-sufficiency as one of important missions of CFETR. This paper presents a preliminary neutronics design and analysis employing a BIT (breeder inside tube) type helium cooling ceramics blanket (HCCB) design concept as one of CFETR blanket design candidates. Firstly, 1D reactor model was designed using ceramic breeder Li4SiO4 and beryllium in pebble for multiplier. The primary blanket parameters were optimized to yield the higher tritium breeder ratio (TBR), including the thickness of outboard breeder blanket, enrichment of Li-6 and ratio of Li4SiO4 to Be. Secondly, based on the optimized blanket parameters and plasma parameters, a detailed 3D neutronics calculation model of 22.5° reactor sector was developed, including blanket modules, shield, divertor, vacuum vessel and TF coil. The gap between blanket modules had been taken into account. Finally, a set of nuclear analyses were carried out addressing the key neutronics issues by Monte Carlo neutron-photon transport code MCNP version 5 and the FENDL-2.1 data library. The preliminary analysis results showed that the global TBR could achieve 1.21 which satisfied the tritium self-sufficiency demand. Nuclear heat, neutronic flux, and distribution of neutron wall loading (NWL) were also analyzed as source terms of the blanket thermal-hydraulics design and reactor nuclear response.

Thermal Analysis of the Tokamak Fusion Test Reactor Vacuum Vessel

1985 ◽  
Vol 7 (1) ◽  
pp. 111-124 ◽  
Author(s):  
S. Z. Fixler ◽  
G. W. Gilchrist ◽  
J. Bialek
Keyword(s):  
Thermal Analysis ◽  
Vacuum Vessel ◽  
Test Reactor

scholarly journals The First Decommissioning of a Fusion Reactor Fueled by Deuterium-Tritium

2003 ◽  
Author(s):  
Charles A. Gentile ◽  
Erik Perry ◽  
Keith Rule ◽  
Michael Williams ◽  
Robert Parsells ◽  
...  
Keyword(s):  
Fusion Reactor ◽  
Field Work ◽  
Vacuum Vessel ◽  
The Novel ◽  
Diamond Wire ◽  
Princeton University ◽  
Innovative Process ◽  
The World ◽  
Test Reactor ◽  
Viable Model

The Tokamak Fusion Test Reactor (TFTR), at the Plasma Physics Laboratory of Princeton University (PPPL) was the first fusion reactor fueled by a mixture of deuterium and tritium (D-T) to be decommissioned in the world. The decommissioning was performed over a period of three years and was completed safely, on schedule and under budget [1]. Provided is an overview of the project and detail of various factors which led to the success of the project. Discussion will cover management of the project, engineering planning before the project started and during the field work was being performed, training of workers in the field, the novel adaptation of tools from other industry, and the development of an innovative process for the use of diamond wire to segment the activated / contaminated vacuum vessel. The success of the TFTR decommissioning provides a viable model for the decommissioning of D-T burning fusion devices in the future.

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