A study on nuclear analysis of the divertor region of the CFETR

This paper presents the nuclear analysis performance of the Chinese Fusion Engineering Test Reactor(CFETR)divertor region using the MCNP-5 Monte Carlo N-particles code in a 3D geometry model.We assessed the nuclear responses of the divertor region component systems and evaluated their shielding capability,which can support the development strategy of the physical and engineering design of the CFETR.Model specification based on the latest CAD model of the CFETR divertor has been integrated into the CFETR MCNP reference model with a major/minor radius R=7.2 m/a=2.2 m in the 22.5° model,and a fusion-power range of around 1-1.5 GW.The nuclear heating and radiation damage of the divertor system are enhanced compared to that of the ITER and the earlier CFETR design.The initial nuclear responses of the toroidal field coil and vacuum vessel systems showed that the shielding of the current divertor design is not sufficient and optimization work has been carried out.We also carried out calculations and analysis using a hypothetical operating scenario of over 14 years.An excellent improvement in the nuclear performance has been obtained by the improved additional shielding block in the divertor region when referring to the ITER design limit,which can support the design of the future update of the divertor region systems of the CFETR.     

Conceptual design and heat transfer performance of a flat-tile water-cooled divertor target

The divertor target components for the Chinese fusion engineering test reactor (CFETR) and the future experimental advanced superconducting tokamak (EAST) need to remove a heat flux of up to ～20 MW m-2.In view of such a high heat flux removal requirement,this study proposes a conceptual design for a flat-tile divertor target based on explosive welding and brazing technology.Rectangular water-cooled channels with a special thermal transfer structure (TTS)are designed in the heat sink to improve the flat-tile divertor target's heat transfer performance(HTP).The parametric design and optimization methods are applied to study the influence of the TTS variation parameters,including height (H),width (W*),thickness (T),and spacing (L),on the HTP.The research results show that the flat-tile divertor target's HTP is sensitive to the TTS parameter changes,and the sensitivity is T ＞ L ＞ W* ＞ H.The HTP first increases and then decreases with the increase of T,L,and W* and gradually increases with the increase of H.The optimal design parameters are as follows:H =5.5 mm,W* =25.8 mm,T =2.2 mm,and L =9.7 mm.The HTP of the optimized flat-tile divertor target at different flow speeds and tungsten tile thicknesses is studied using the numerical simulation method.A flat-tile divertor mock-up is developed according to the optimized parameters.In addition,high heat flux (HHF)tests are performed on an electron beam facility to further investigate the mock-up HTP.The numerical simulation calculation results show that the optimized flat-tile divertor target has great potential for handling the steady-state heat load of 20 MW m-2 under the tungsten tile thickness＜5 mm and the flow speed ≥7 m s-1.The heat transfer efficiency of the flat-tile divertor target with rectangular cooling channels improves by ～ 13％ and ～30％ compared to that of the flat-tile divertor target with circular cooling channels and the ITER-like monoblock,respectively.The HHF tests indicate that the flat-tile divertor mock-up can successfully withstand 1000 cycles of 20 MW m-2 of heat load without visible deformation,damage,and HTP degradation.The surface temperature of the flat-tile divertor mock-up at the 1000th cycle is only ～930 ℃.The fiat-tile divertor target's HTP is greatly improved by the parametric design and optimization method,and is better than the ITER-like monoblock and the fiat-tile mock-up for the WEST divertor.This conceptual design is currently being applied to the engineering design of the CFETR and EAST flat-tile divertors.     

Evaluation and optimization of tritium breeding,shielding and nuclear heating performances of the helium cooled solid breeder blanket for CFETR

Chinese Fusion Engineering Test Reactor (CFETR) is a new test tokamak device being designed in China aimed to bridge gaps between ITER and DEMO. As one of the candidate tritium breeding blankets, a conceptual design scheme of the helium cooled solid breeder blanket has been proposed and a series of preliminary analyses have been carried out to access the performances. However, both the required global tritium breeding ratio for CFETR not less than 1.2 and its poor working conditions under intense radiation need further thorough neutronics analyses and optimizations during the design phase. In this work, first, three-dimensional neutronics model of CFETR was built, and the neutron wall loading and global TBR were obtained. The nuclear and thermal calculation results were automatically coupled, which could make the neutronics calculation results more accurate and guaranteed they could always satisfy the corresponding thermal limits during the whole process. Then, the tritium breeding and shielding performances of both the outboard and inboard equatorial blanket modules were optimized for the comprehensive optimal schemes. The influences of Be/W armors on the shielding performance and TBR were also investigated. Finally, the nuclear heating rates and the neutron flux densities in different components were calculated based on the obtained comprehensive optimal scheme. In this paper, the neutronics analyses and optimizations verified that the optimized conceptual design could well meet the tritium self-sufficiency and neutron shielding requirements, and this could provide a valuable reference for the further thermal-hydraulic analysis and structural optimization of the CFETR helium cooled solid breeder blanket.     

聚变反应堆焊缝超声检测标准比较

ITER(国际热核聚变实验堆)及CFETR(中国聚变工程实验堆)大型聚变装置的役前及在役无损检测分别参照标准RCC-MR-2007第3部分及标准NB/T 47013-2015执行。从适用范围、表面准备、探头参数规定、标准试块和对比试块、扫查方式、缺陷表征和验收等几个方面,将RCC-MR-2007中对接接头焊缝常规超声检测与NB/T 47013-2015中焊缝超声检测标准的相关内容进行对比,使设计人员更加深入地理解相应标准,并为设计优化提供参考。     

Hefei utility negative ions test equipment with RF source: commissioning and first results

In order to study the key technology and physics of RF driven negative ion source for neutral beam injector in China, the Hefei utility negative ions test equipment with RF source was developed at Institute of Plasma Physics, Chinese Academy of Sciences (ASIPP). Its negative ion source can be equipped with single or double RF drivers. There is a plasma expansion chamber with depth of 19 mm and an enhanced filter field. A three electrodes negative ion accelerator was employed to extract and accelerate the negative ions, which are plasma grid, extraction grid and ground grid. And there are several diagnostic tools for the plasma and beam parameters measurement. The characteristics of plasma generation, negative ion production and extraction were studied on the test equipment. The negative ion beam was extracted from the RF driven negative ion source for the first time. The detailed structure and main results are presented in this article.     

Design and thermal‐hydraulic evaluation of helium‐cooled ceramic breeder blanket for China fusion engineering test reactor

China Fusion Engineering Test Reactor, a fusion tokamak device, is proposed to provide complementary technology and experience for ITER and the future fusion power plant. A helium‐cooled ceramic breeder blanket concept is adopted as the candidate tritium breeding blanket for China Fusion Engineering Test Reactor. Detailed design of the blanket structure located at the outboard equatorial plane is presented. The coolant flow characters in the blanket were calculated by the theoretical method and the finite element method. The comparison of the calculated results was done, and it has a good agreement between theoretical results and simulation results. The results show that the pressure drop is 0.13 MPa and the total temperature rise is 194.6°C.     

Radiation shielding design of the CFETR polarimeter interferometer and CO2 dispersion interferometer

A three-wave based laser polarimeter/interferometer and a CO₂ laser dispersion interferometer are used to determine the electron and current density profiles on a Chinese fusion engineering test reactor (CFETR). Radiation shielding is designed for the combination of polarimeter/interferometer and CO₂ dispersion interferometer. Furthermore, neutronics models of the two systems are developed based on the engineering-integrated design of CFETR polarimeter/interferometer and CO₂ dispersion interferometer and the major material components of CFETR. The polarimeter/interferometer and CO₂ dispersion interferometer's neutron and photon transport simulations were performed using the Monte Carlo neutral transport code to determine the energy deposition and neutron energy spectrum of the optical mirrors. The energy depositions of the first mirrors on the polarimeter/interferometer are reduced by three orders with the whole shielding. Since the mirrors of CO₂ dispersion interferometer are very close to the diagnostic first wall, shielding space is limited and the CO₂ dispersion interferometer energy deposition is higher than that of the polarimeter/interferometer. The dose rate after shutdown 10⁶ s in the back-drawer structure has been estimated to be 83 μSv h⁻¹ when the radiation shield is filled in the diagnostic shielding modules, which is below the design threshold of 100 μSv h⁻¹. Radiation shielding design plays a key role in successfully applying polarimeter/interferometer and CO₂ dispersive interferometer in CFETR.     

Estimation of China Fusion Engineering Test Reactor performance and burning fraction in different pellet fueling scenarios by a multi-species radial transport model

Tritium self-sufficiency in future deuterium–tritium fusion reactors is a crucial challenge. As an engineering test reactor, the China Fusion Engineering Test Reactor requires a burning fraction of 3% for the goal to test the accessibility to the future fusion plant. To self-consistently simulate burning plasmas with profile changes in pellet injection scenarios and to estimate the corresponding burning fraction, a one-dimensional multi-species radial transport model is developed in the BOUT++ framework. Several pellet-fueling scenarios are then tested in the model. The results show that the increased fueling depth improves the burning fraction by particle confinement improvement and fusion power increase. Nevertheless, by increasing the depth, the pellet cooling-down may significantly lower the temperature in the core region. Taking the density perturbation into consideration, the reasonable parameters of the fueling scenario in these simulations are estimated as pellet radius ${r}_{{\rm{p}}}=3\,{\rm{mm}},$ injection rate $=\,4\,\mathrm{Hz},$ and pellet injection velocity $=\,1000\mbox{--}2000\,{\rm{m}}\,{{\rm{s}}}^{-1}$ without drift or $450\,{\rm{m}}\,{{\rm{s}}}^{-1}$ with high-field-side drift.     

CFETR氦冷固态增殖剂包层产氚性能研究及优化

氦冷固态增殖剂包层是中国聚变工程实验堆（CFETR）的3种候选包层概念之一。本文基于中国核工业西南物理研究院提出的一种氦冷固态增殖剂包层概念，通过蒙特卡罗输运程序MCNP5建立了包层三维中子学模型，探究了不同几何布置方案及结构设计参数对包层产氚性能的影响，得到了全堆氚增殖比（TBR）及极向各包层模块产氚分布，并由优化后的模型得到了包层模块核热分布。结果表明，优化后的TBR达到1.177，满足氚自持的最低要求。     

硅酸锂球床传热传质特性与吹扫气体流动的DEM-CFD耦合模拟

氦冷固态增殖包层是中国聚变工程实验堆（CFETR）的3种候选包层概念之一,氚增殖球床是包层的核心部件,采用硅酸锂颗粒作为氚增殖材料。球床结构对氚在球床内的输运行为及流动和传热均有重要影响。本文基于离散单元法（DEM）生成了满足氚增殖球床填充率要求的随机堆积结构,通过CFD计算获取了球床结构下氚在吹扫气体内的等效扩散系数及吹扫气体的流动特性,包括速度分布、压力分布及进出口压降;开展了外加热流及有内热源两种工况下球床等效导热系数的模拟。计算结果表明,球床结构下氚在吹扫气体内的等效扩散系数为二元气体扩散系数的40%;受球床结构影响,球床内存在流动迟滞区,壁面出现流动加速;拟合得到Ergun方程的黏性阻力系数C1=87;有内热源工况下的球床等效导热系数低于外加热流工况下的球床等效导热系数。