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.

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.

Concept Design of Bi-2212 Superconducting Joint for the Central Solenoid Magnet of China Fusion Engineering Test Reactor

2020 ◽  
Vol 89 (9) ◽  
pp. 094201
Author(s):  
Haojun Yang ◽  
Kun Lu ◽  
Xiongyi Huang ◽  
Jian Rong ◽  
Yuntao Song
Keyword(s):  
Concept Design ◽  
Fusion Engineering ◽  
Test Reactor

Benchmark of Advanced Burner Test Reactor model using MCNPX 2.6.0 and ERANOS 2.1

2011 ◽  
Vol 53 (6) ◽  
pp. 633-644 ◽  
Author(s):  
Kenneth Allen ◽  
Travis Knight ◽  
Samuel Bays
Keyword(s):  
Reactor Model ◽  
Test Reactor

Conceptual Design and Neutronics/Thermal-Hydraulic Coupling Optimization Analyses of Two Typical Helium Cooled Solid Breeder Blanket Modules for CFETR Phase II

2018 ◽  
Author(s):  
Shijie Cui ◽  
Dalin Zhang ◽  
Wenxi Tian ◽  
G. H. Su ◽  
Suizheng Qiu
Keyword(s):  
Conceptual Design ◽  
Phase Ii ◽  
Fusion Energy ◽  
Operation Condition ◽  
The Past ◽  
Self Sufficiency ◽  
Hydraulic Coupling ◽  
The Future ◽  
Test Reactor ◽  
Solid Breeder Blanket

Chinese Fusion Engineering Test Reactor (CFETR) is a new test Tokamak device which is now being designed in China to make the transition from the International Thermonuclear Experimental Reactor (ITER) to the future Fusion Power Plant (FPP). Breeding blanket is the key component of fusion reactor which is mainly responsible for the tritium self-sufficiency and fusion energy conversion. In the past few years, three kinds of blanket conceptual design schemes have been proposed and tested in parallel for CFETR Phase I, in which the helium cooled solid breeder (HCSB) blanket concept is acknowledged as the most promising one. However, nowadays, the design phase of CFETR has gradually changed from I to II aiming for the future DEMO operation condition, the main parameters of which are quite different from the previous one. Therefore, it’s necessary to perform conceptual design and various analyses for the HCSB blanket under the new working condition. In this work, firstly, a new conceptual design scheme of HCSB blanket for Phase II is put forward. Then, the radial build arrangements, of the two typical blanket modules are optimized by using the NTCOC. This work can provide valuable references for further conceptual design and neutronics/thermal-hydraulic coupling analyses of the HCSB blanket for CFETR Phase II.

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

ITER-FEAT vacuum vessel and blanket design features and implications for the R&D programme

2001 ◽  
Vol 41 (3) ◽  
pp. 265-275 ◽  
Author(s):  
K Ioki ◽  
W Dänner ◽  
K Koizumi ◽  
V.A Krylov ◽  
A Cardella ◽  
...  
Keyword(s):  
Vacuum Vessel ◽  
Design Features ◽  
Blanket Design

scholarly journals Economic Dispatch Model of Nuclear High-Temperature Reactor with Hydrogen Cogeneration in Electricity Market

Energies ◽  
2021 ◽  
Vol 14 (24) ◽  
pp. 8289
Author(s):  
James Richards ◽  
Cristian Rabiti ◽  
Hiroyuki Sato ◽  
Xing L. Yan ◽  
Nolan Anderson
Keyword(s):  
High Temperature ◽  
Electricity Market ◽  
Fossil Fuels ◽  
Economic Dispatch ◽  
Sale Price ◽  
Nuclear Heat ◽  
Time Histories ◽  
Price Duration ◽  
Test Reactor ◽  
Industry Sectors

Hydrogen produced without carbon emissions could be a useful fuel as nations look to decarbonize their electricity, transport, and industry sectors. Using the iodine–sulfur (IS) cycle coupled with a nuclear heat source is one method for producing hydrogen without the use of fossil fuels. An economic dispatch model was developed for a nuclear-driven IS system to determine hydrogen sale prices that would make such a system profitable. The system studied is the HTTR-GT/H2, a design for power and hydrogen cogeneration at the Japan Atomic Energy Agency’s High Temperature Engineering Test Reactor. This study focuses on the development of the economic model and the role that input data plays in the final calculated values. Using a historical price duration curve shows that the levelized cost of hydrogen (LCOH) or breakeven sale price of hydrogen would need to be 98.1 JPY/m3 or greater. Synthetic time histories were also used and found the LCOH to be 67.5 JPY/m3. The price duration input was found to have a significant effect on the LCOH. As such, great care should be used in these economic dispatch analyses to select reasonable input assumptions.

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