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1.
水冷陶瓷增殖剂(WCCB)包层作为中国聚变工程试验堆(CFETR)候选包层之一,承担着氚增殖、核热提取、屏蔽等重要涉核功能,其中子学设计的可靠性直接影响CFETR氚自持目标的实现。为验证中子学设计工具,即MCNP和FNEDL3.0数据库,在WCCB包层中子学设计中的可靠性,基于研制出的WCCB包层模块,在DT中子环境下开展中子学实验,对以产氚率(TPR)为代表的中子学参数进行了模拟值(C)和实验值(E)对比分析。结果表明,模块中轴线位置处TPR的C/E为0.97?1.08,而模块边缘位置处TPR的C/E为0.65?0.82;模块钛酸锂层边缘区197Au(n,γ)198Au反应率的C/E为0.72?0.90,表明模块边缘区存在非期望的散射中子,导致该区TPR模拟值和实验值偏离较大。  相似文献   
2.
为解决中国聚变工程实验堆316L不锈钢焊缝超声波检测时,探头扫查空间受限、检测信噪比低的难题,提出了基于双晶面阵探头的相控阵超声检测方案.通过CIVA仿真,分析了不同聚焦参数下DMA探头的声场特征,确定对接接头的检测工艺.参考NB/T47013.3-2015《承压设备无损检测第3部分:超声检测》附录I,设计并制作了对比试块,验证了检测工艺下的声束覆盖和φ2 mm侧横孔信噪比.结果表明,DMA探头可以在有限的扫查空间内实现焊缝声束全覆盖,对比试块中不同位置的φ2 mm侧横孔信噪比大于15 dB.试验结果可为316L类不锈钢对接焊缝相控阵超声检测工艺制定提供参考.  相似文献   
3.
In a fusion power plant, the integration of the blanket system in the design progress is of vital importance to address the fundamental function of sufficient tritium production, reliable nuclear heat extraction, and permanent components protection due to the complex assembling systems into the tokamak vessel. Some progress activities of the blanket system design and integration for China fusion engineering test reactor (CFETR) are developed. The integration work involves the design of the breeding blanket, back plate support, shielding blanket, and supporting structures. To guarantee normal operation of the reactor, the design and arrangement of cooling pipes are very critical. The layout of the complex cooling pipes inside the blanket system is designed and integrated. Interfaces between main connecting components are designed for blanket system integration. In this work, the U‐shaped HCCB blanket is utilized as the adaption breeding module into the integrated blanket system. The three‐dimensional (3D) neutronic analyses verified that the integrated design of the blanket system could well meet the requirements of tritium self‐sufficiency and neutron shielding. The nuclear heat generation of main components in the reactor is obtained with a nuclear energy multiplication factor of 1.35. The primary principle during the integration is to accommodate all the components allocated on the vacuum vessel. This work is the preliminary integration design and validation for CFETR blanket system, and further detailed design will be performed around these obtained references for better fusion feasibility and performance in the next‐step design stage.  相似文献   
4.
CFETR which stands for “China Fusion Engineering Test Reactor” is a new tokamak device. Its magnet system includes the Toroidal Field (TF) winding, Center solenoid winding (CS) and Poloidal Field (PF) winding. The main goal of the project is to build a fusion engineering Tokamak reactor with its fusion power is 50–200 MW and should be self-sufficiency by blanket.In order to ensure the maintenance ports design and maintenance method, this article discussed the concept design of the magnet system based on different maintenance port cases. The paper detailed studied the magnet system of CFETR including the electromagnetic analysis and parameters for TF (CS)PF. Besides, the volt-seconds of ohmic field are presented as detailed as possible in this paper. In addition, the calculations and optimizations of equilibrium field which should guarantee the plasma discharge of single null shape is carried out. The design work reported here illustrates that the present maintenance ports will not have a great impact on the design of the magnet system. The concept design of the magnet system can meet the requirement of the physical target.  相似文献   
5.
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.  相似文献   
6.
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.  相似文献   
7.
The hybrid scenario is a projection for CFETR operation with high plasma current and density. Therefore, the energetic particles (EPs) generated by fusion reactions can destabilize Alfvén eigenmodes (AEs), which could result in significant EPs loss and redistribution. Both the eigenvalue code NOVA-K and the wrapped local stability code TGLFEP are used to analyze AE stability. The simulation indicates the beta-induced Alfvén eigenmodes with n>5 in the core region are the most unstable. The NOVA-K code is used to benchmark the critical density gradient calculated by TGLFEP. In addition, the EPtran code is employed to predict EP transport induced by unstable AEs and turbulence, which reduce EP density in the core and drive approximately 30% EP transport from the core to the edge, thus the EP density profile flattens and EPs with lower energy deposit near the edge.  相似文献   
8.
中国聚变工程实验堆(Chinese Fusion Engineering Testing Reactor,CFETR)的包层和偏滤器第一壁面向堆芯等离子体,第一壁辐照损伤分析对于托克马克安全运行至关重要。赤道面外包层较其它包层距离堆芯等离子体中心更近,其结构材料承受中子辐照大。因此,进行中子辐照损伤评估十分必要。基于此目的,采用计算机辅助设计(Computer Aided Design,CAD)模型和蒙特卡罗中子学建模转换接口Mc CAD完成中子学建模,并用蒙特卡罗方法的粒子输运程序计算第一壁和氦冷固态外包层结构材料辐照损伤。此外,对比了铍和钨作为面向等离子体材料两种情况下第一壁的受损情况。计算结果表明,氦冷固态包层模型下结构材料可以满足CFETR一期的运行要求。  相似文献   
9.
A hybrid superconducting central solenoid employs Bi‐2212 high‐temperature superconductors and Nb3Sn low‐temperature superconductors under the design of the Institute of Plasma Physics, Chinese Academy Of Science for further upgrade to CFETR, namely, the China Fusion Engineering Testing Reactor. The conductor type of both parts is cable‐in‐conduit conductors. This paper mainly focuses on stability study of the inner high‐temperature superconductors part whose conductor works under a peak magnetic field of 16.79 T, and the maximum operating current of each turn is 50 kA. The simulation based on a 1‐D simplified model is performed using the code THEA (thermal hydraulic and electric analysis of superconducting cable). Firstly, a brief analysis of stability considering the AC loss during current ramp‐up is studied. Then, the stability margins in cases of different perturbations with varied lengths and durations are calculated, and a qualitative explanation of the result is proposed. Besides, the inlet pressure and pressure drop crucially influence the convection heat transfer between strands and helium; thus, the effect of these two factors on stability margin is discussed. All these results will provide important references for further optimization of this hybrid magnet. Copyright © 2017 John Wiley & Sons, Ltd.  相似文献   
10.
实现氚自持、建立完整的氚循环系统并保证氚安全是中国聚变工程实验堆(CFETR)的主要目标之一。在CFETR氦冷固态包层及其辅助系统设计过程中,需对系统级氚输运行为进行详细分析,包括氚滞留量、释放量、浓度的动态变化等。基于已建立的动态氚分析程序TriSim-Dynamic,在此基础上进行修改完善,利用该程序对CFETR氦冷固态包层及其辅助系统氚动态输运进行分析模拟,得到了冷却剂及提氚吹扫气中氚浓度、氚分压,管壁及结构材料中氚盘存量,氚通过包层结构材料和辅助系统管壁向真空室、水冷系统及建筑的渗透通量动态变化,并将其稳态值与已进行基准校核的稳态氚分析程序TriSim-SA及理论解析解进行比较,以初步验证分析结果的准确性,数据结果也对CFETR氚安全分析提供了一定的参考。  相似文献   
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