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聚变发电反应堆概念设计研究 总被引:11,自引:24,他引:11
吴宜灿 汪卫华 刘松林 李静惊 王红艳 陈红丽 陈明亮 张士杰 黄群英 黄德所 郑善良 曾勤 胡丽琴 柏云清 章毛连 李艳芬 李春京 冯岩 宋勇 龙鹏成 FDS课题组 《核科学与工程》2005,25(1):76-85
在广泛分析聚变能相关领域研究发展状况和国际热核聚变实验堆(ITER)物理与技术基础上,提出了一个考虑了技术可行性的聚变发电反应堆概念(称之为FDS Ⅱ)。这个概念具有ITER参数适量外推的等离子体物理与技术水平的聚变堆芯和具有发展潜力的液态锂铅氚增殖包层,在对这个概念进行中子学、热工水力学、力学、安全与环境影响和经济学等一系列计算分析的基础上,给出了初步的概念设计和进一步设计优化的共性原则。 相似文献
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W/Cu和W/C涂层的循环热负荷实验 总被引:1,自引:0,他引:1
用等离子体喷涂和热压方法制作了W /Cu梯度功能材料和W /C涂层 ,其中W/C涂层具有多层的钨 (W )、铼 (Re)扩散阻挡势垒。为了试验这些复合材料能否经受聚变等离子体破裂时的高热负荷 ,用大功率ND :YAG激光进行了模拟实验。结果表明 :在 1 0 0~ 4 0 0MW /m2 的瞬时 (脉冲宽度为 4ms)热负荷作用下 ,经过 2 0 0~ 70 0次热循环 ,未发现W /Cu复合体的开裂。其中在 1 2 3MW /m2 的功率密度下作用70 0次后 ,发现等离子体喷涂试样表面的再结晶现象和严重的晶界腐蚀 ,由于激光的冷效应 ,晶粒生长的趋势并不明显 ,再结晶层的晶粒呈垂直于表面的柱状结构。W/C试样的退火实验表明 ,钨涂层的再结晶温度稍高于 1 4 0 0℃。在更高的功率密度下 (3 98MW /m2 )出现了明显的腐蚀坑 ,坑内呈疏松的蜂窝结构 ,坑的边缘形成了沉积区 ,能谱分析表明沉积区集聚了大量的金属杂质。等离子体喷涂试样比热压试样更易产生晶界的断裂和裂纹。在同等的热负荷条件下 ,W /CuFGM的质量损失低于石墨材料 相似文献
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聚变发电反应堆双冷液态锂铅包层模块结构设计与分析 总被引:9,自引:8,他引:1
给出聚变发电反应堆FDS Ⅱ模块式液态锂铅包层(DLL)结构方案,以低活化马氏体(RAFM)钢为结构材料,采用液态金属LiPb作为增殖材料和冷却剂,使用碳化硅流道插件作为电绝缘和热绝缘。包层的设计特点体现在:从增殖区、冷却剂流道、屏蔽包层、母管、机械连接、维修装配等几个方面全局考虑包层设计,结构布置完整;独有的“”形隔板设计使氦气冷却回路容易实现,增殖流道简单,可简化制造工艺,提高可靠性。同其他液态锂铅包层相比,DLL包层在冷却剂系统、制造、装配上可成就较高的可行性。 相似文献
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托卡马克(Tokamak)聚变装置中子学分析中,聚变中子源描述是重要的输入参数,其准确性直接影响分析结果的可靠性。通过分析ITER和欧洲聚变示范堆(EU DEMO)中子学分析中所采用的聚变中子源模型,提出了一种完整描述Tokamak中L-mode、H-mode等离子体的D-D、D-T聚变中子源的数值模型。在中国聚变工程实验堆(CFETR)的工程集成设计平台上,编写了基于蒙特卡罗算法的程序SCG求解该数值模型,实现了读取(零维)等离子体参数、输出可供典型中子学软件MCNP直接读取的中子源定义文件的功能。以CFETR氦冷球床包层的中子学分析模型为基准,在相同的L-mode等离子体D-T聚变工况下,相较于采用EU DEMO源子程序,采用本模型计算得到的中子壁负载差异最大值为2.02%,包层氚增殖率差异为0.18%,全堆能量增益因子的差异为0.23%。结果表明,本模型与其他源描述的差异较小,可应用于CFETR的中子学分析。 相似文献
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China Fusion Engineering Test Reactor is a new tokamak device which is proposed by China National Integration Design Group. The fusion power is 50–200 MW and its plasma major radius and plasma minor radius are 5.7 and 1.6 m. The helium cooled lithium ceramic (HECLIC) blanket, as a key component of the tokamak, has the basic function to provide tritium breeding and plasma limiter. The blanket also provides main thermal and nuclear shielding of the vacuum vessel and ex-vessel components such as magnetic coils during plasma operations. With the development of the numerical simulation technology, more and more design parameters can be obtained by this method. Numerical simulation has been used for design and optimization, because some parameters are very hard to obtain though theoretical calculation. In this study, the simulation methods are investigated for HECLIC blanket design. Besides, design flow of the blanket is discussed and related analysis is also introduced to improve the design. 相似文献
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K.M. Feng Author Vitae G.S. Zhang G.Y. Zheng Z. Zhao T. Yuan Z.Q. Li G.Z. Sheng C.H. Pan 《Fusion Engineering and Design》2009,84(12):2109-2113
The basic definition and development strategy of the DEMO plant based on the Chinese fusion power plant (FPP) program are presented briefly. A conceptual design study of fusion HCSB-DEMO reactor with a fusion power of 2550 MW and a neutron wall loading of 2.3 MW/m2 is performed recently. Three sets parameters of core plasma for different DEMO design objectives are proposed. A helium-cooled blanket system with ceramic breeder (Li4SiO4), the structure material of low-activation ferritic steel (LAF/M) and Be neutron multiplier based on Chinese ITER HCSB-TBM design foundation are considered. The design parameters, preliminary analyses and the basic structure as well as development strategy of HCSB-DEMO reactor are introduced. 相似文献
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A conceptual blanket design for UWMAK-II based on breeding in LiAlO2 and helium cooling for a D-T fusion reactor is described. The reactor is a Tokamak with 316 stainless steel as the primary structural material, a major radius of 13 m and a minor radius of 5 m. The power output is 5000 MW(th) and the maximum temperature in the stainless steel structure is 650°C. This reactor design study is one of a series performed to evaluate the merits of various fusion reactor design concepts. In this paper the mechanical and the thermal hydraulics problem associated with the blanket for this reactor is described. Special attention has been given to the need for repairing and replacing the first wall of the blanket. Other problems which may arise from such a blanket design are also discussed. 相似文献
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Xinli Gao Tiejun Zu Wenxi Tian Suizheng Qiu Guanghui Su Hongchun Wu 《Fusion Engineering and Design》2013,88(12):3185-3193
A Water-cooled Pressure Tube Energy production blanket (WPTE) for fusion driven subcritical reactor has been designed to achieve 3000 MW thermal power with self-sustaining tritium cycle. Pressurized water has great advantages in energy production; however the high pressure may cause some severe structural design issues. This paper proposes a new concept of water-cooled blanket. To solve the problem of the high pressure of the coolant, the pressure tube was adopted in the design and in the meantime, the thickness of the first wall can be significantly reduced as result of adopting pressure tube. The numerically simulating and calculating of temperature, stress distribution and flow analyses were carried out and the feasibility of using water as coolant was discussed. The results demonstrated the engineering feasibility of the water-cooled fusion–fission hybrid reactor blanket module. 相似文献
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In this paper, one standard water cooled ceramic breeder blanket sector has been modeled for the Chinese fusion engineering test reactor using RELAP5/MOD3.3 with details of anisotropic structures, positions and nuclear heat of the blanket modules. The multi-pipe manifolds of the current sector design scheme has been designed and analyzed. And an optimized scheme was proposed to further reduce the pressure drop, uniform the flow distribution, and prevent overheating. Also the fusion power excursion transients were simulated to evaluate the system heat removal and recovery ability. The results indicated that high-transient heat flux up to 0.8 MW/m2 can cause sub-cooled boiling of the coolant in the first wall area of certain modules. Coolant returns to single phase soon after the end of the transient. According to the analysis, it is suggested that the blanket modules surrounding plasma have as similar structure design features as possible and sizes of the modules should be kept relatively small so as to obtain a reasonable pressure drop. 相似文献
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Robert W. Conn Gerald L. Kulcinski Charles W. Maynard 《Nuclear Engineering and Design》1976,39(1):5-44
UWMAK-II is a conceptual design study of a low ß, circular Tokamak fusion power reactor. The aim of the study has been to perform a self-consistent analysis of a probable future fusion power system based on the philosophy that design decisions, wherever possible, should be conservative and should be based on present technology. As such, this system will not be the smallest, the least expensive, or the optimum Tokamak reactor. Rather, it represents a feasible system which we use to assess the technological problems uncovered and to examine possible solutions. The plasma is designed to generate 5000 MW(th) during a pulse and 1709 MW(e) continuously based upon a burn cycle with a 90 min burn and a 6.5 min rejuvenation period. The plasma carries a current of 14.9 MA and is designed with a double null poloidal divertor for impurity control and particle pumping. In addition, a low Z liner in the form of a carbon curtain is included to eliminate any source of high Z impurities. Plasma heating to ignition involves the use of neutral beam heating for a 10 sec period during which 200 MW of 500 keV deuterium atoms are injected into the plasma.The blanket design employs helium cooling and the solid lithium-bearing compound, lithium aluminate (Li2Al2O4) for breeding tritium. The structural material is 316 stainless steel and beryllium is used as a neutron multiplier. The neutron radiation environment produces radiation damage that considerably influences blanket and system performance. The most significant effect is the loss of ductility which appears to limit the usable lifetime of the blanket first wall to about 2 yr at a 14 MeV neutron wall loading of 1.16 MW/m2. The solid breeder blanket minimizes the tritium inventory but because of the low fractional burnup in the plasma and the need for roughly a one day reserve of fuel, the inventory is 17.7 kg. Induced radioactivity levels in the structure are of the order of 1 Ci/W(th) at shutdown after two years of operation. The main contributors to the activity are
) and
). Afterheat levels are slightly above 1% of thermal power but the afterheat power density is low, less than 0.1 w/g. The power cycle involves a He---Na intermediate heat exchanger followed by a sodium—steam system. The sodium intermediary is used to minimize tritium leakage through the power cycle and to provide a working fluid for thermal energy storage such that continuous electrical output is produced despite a pulse plasma cycle. A materials resource study has been completed for a UWMAK-II type system and beryllium appears to present a particular problem with regard to adequate resources. Other materials that could present problems of procurement include chromium and nickel. A preliminary economic analysis has been carried out to identify major cost areas and this is described. 相似文献
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Dale M. Meade 《Journal of Fusion Energy》1994,13(2-3):145-154
Temperatures, densities and confinement of deuterium plasmas confined in tokamaks have been achieved within the last decade that are approaching those required for a D-T reactor. As a result, the unique phenomena present in a D-T reactor plasma (D-T plasma confinement, alpha confinement, alpha heating and possible alpha driven instabilities) can now be studied in the laboratory. Recent experiments on the Tokamak Fusion Test Reactor (TFTR) have been the first magnetic fusion experiments to study plasmas with reactor fuel concentrations of tritium. The injection of 20 MW of tritium and 14 MW of deuterium neutral beams into the TFTR produced a plasma with a T/D density ratio of 1 and yielded a maximum fusion power of 9.2 MW. The fusion power density in the core of the plasma was 1.8 MW m–3 approximating that expected in a D-T fusion reactor. In other experiments TFTR has produced 6.4 MJ of fusion energy in one pulse satisfying the original 1976 goal of producing 1 to 10 MJ of fusion energy per pulse. A TFTR plasma with T/D density ratio of 1 was found to have 20% higher energy confinement time than a comparable D plasma, indicating a confinement scaling with average ion mass, A, of E. The core ion temperature increased from 30 keV to 37 keV due to a 35% improvement of ion thermal conductivity. Using the electron thermal conductivity from a comparable deuterium plasma, about 50% of the electron temperature increase from 9 keV to 10.6 keV can be attributed to electron heating by the alpha particles. At fusion power levels of 7.5 MW, fluctuations at the Toroidal Alfvén Eigenmode frequency were observed by the fluctuation diagnostics. However, no additional alpha loss due to the fluctuations was observed. These D-T experiments will continue over a broader range of parameters and higher power levels.Work supported by U.S. Department of Energy Contract No. DE-AC02-76-CHO-3073. 相似文献
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The water cooled ceramic breeder (WCCB) blanket employing pressurized water as a coolant is one of the breeding blanket candidates for the China Fusion Engineering Test Reactor (CFETR).Some updating of neutronics analyses was needed,because there were changes in the neutronics performance of the blanket as several significant modifications and improvements have been adopted for the WCCB blanket,including the optimization of radial build-up and customized structure for each blanket module.A 22.5 degree toroidal symmetrical torus sector 3D neutronics model containing the updated design of the WCCB blanket modules was developed for the neutronics analyses.The tritium breeding capability,nuclear heating power,radiation damage,and decay heat were calculated by the MCNP and FISPACT code.The results show that the packing factor and 6Li enrichment of the breeder should both be no less than 0.8 to ensure tritium self-sufficiency.The nuclear heating power of the blanket under 200 MW fusion power reaches 201.23 MW.The displacement per atom per full power year (FPY) of the plasma-facing component and first wall reach 0.90 and 2.60,respectively.The peak H production rate reaches 150.79 appm/FPY and the peak He production reaches 29.09 appm/FPY in blanket module # 3.The total decay heat of the blanket modules is 2.64 MW at 1 s after shutdown and the average decay heat density can reach 11.09 kW m-3 at that time.The decay heat density of the blanket modules slowly decreases to lower than 10 W m-3 in more than ten years. 相似文献