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

Thermal Analysis of the FSP-1 Experiment in the Advanced Test Reactor

2016 ◽  
Author(s):  
Grant L. Hawkes ◽  
Nicolas E. Woolstenhulme
Keyword(s):  
Thermal Analysis ◽  
Flow Instability ◽  
National Laboratory ◽  
Safety Evaluation ◽  
Thermal Safety ◽  
Primary Coolant ◽  
Coolant Pump ◽  
Safety Margins ◽  
Enriched Uranium ◽  
Test Reactor

The U.S. High Performance Research Reactor Conversions fuel development team is focused on developing and qualifying the uranium-molybdenum (U-Mo) alloy monolithic fuel to support conversion of domestic research reactors to low enriched uranium. Several previous irradiations have demonstrated the favorable behavior of the monolithic fuel. The Full Scale Plate 1 (FSP-1) fuel plate experiment will be irradiated in the northeast (NE) flux trap of the Advanced Test Reactor (ATR). This fueled experiment contains six aluminum-clad fuel plates consisting of monolithic U-Mo fuel meat. Three different types of fuel plates with matching pairs for a total of six plates were analyzed. These three types of plates are: full burn, intermediate power, and thick meat. A thermal analysis has been performed on the FSP-1 experiment to be irradiated in the ATR at the Idaho National Laboratory (INL). A thermal safety evaluation was performed to demonstrate that the FSP-1 irradiation experiment complies with the thermal-hydraulic safety requirements of the ATR Safety Analysis Report (SAR). The ATR SAR requires that minimum safety margins to critical heat flux and flow instability be met in the case of a loss of commercial power with primary coolant pump coast-down to emergency flow. The thermal safety evaluation was performed at 26 MW NE lobe power to encompass the expected range of operating power during a standard cycle. Additional safety evaluations of reactivity insertion events, loss of coolant event, and free convection cooling in the reactor and in the canal are used to determine the response of the experiment to these events and conditions. This paper reports and shows that each safety evaluation complies with each safety requirement of the ATR SAR.

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.

scholarly journals Thermal Predictions of the AGR-3/4 Experiment With Post Irradiation Examination Measured Time-Varying Gas Gaps

2017 ◽  
Vol 3 (4) ◽  
Author(s):  
Grant L. Hawkes ◽  
James W. Sterbentz ◽  
John T. Maki ◽  
Binh T. Pham
Keyword(s):  
Thermal Analysis ◽  
Fast Neutron ◽  
Neutron Fluence ◽  
National Laboratory ◽  
Element Model ◽  
Post Irradiation ◽  
Measured Time ◽  
Dimensional Measurements ◽  
Test Reactor ◽  
Experiment Individual

A thermal analysis was performed for the advanced gas reactor test experiment (AGR-3/4) with post irradiation examination (PIE) measured time (fast neutron fluence) varying gas gaps. The experiment was irradiated at the advanced test reactor (ATR) at the Idaho National Laboratory (INL). Several fuel irradiation experiments are planned for the AGR Fuel Development and Qualification Program, which supports the development of the very high-temperature gas-cooled reactor under the advanced reactor technologies project. The AGR-3/4 test was designed primarily to assess fission product transport through various graphite materials. Irradiation in the ATR started in December 2011 and finished in April 2014. Forty-eight (48) tristructural-isotropic-fueled compacts were inserted into 12 separate capsules for the experiment. The purpose of this analysis was to calculate the temperatures of each compact and graphite layer to obtain daily average temperatures using PIE-measured time (fast neutron fluence) varying gas gaps and compare with experimentally measured thermocouple (TC) data. PIE-measured experimental data were used for the graphite shrinkage versus fast neutron fluence. PIE dimensional measurements were taken on all the fuel compacts, graphite holders, and all of the graphite rings used. Heat rates were input from a detailed physics analysis for each day during the experiment. Individual heat rates for each nonfuel component were input as well. A steady-state thermal analysis was performed for each daily calculation. A finite element model was created for each capsule.

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