共查询到18条相似文献,搜索用时 250 毫秒
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《原子能科学技术》2021,(1)
中子辐照条件下材料结构与性能是中国聚变工程实验堆(CFETR)以及未来聚变反应堆工程设计的重要依据。钨材料是CFETR拟全面使用的壁材料,但中子辐照导致钨硬度升高和韧性大幅下降,严重影响材料的服役性能,进而影响CFETR运行的安全性和稳定性。在目前缺乏聚变中子源进行辐照实验的情况下,开展聚变堆材料中子辐照模拟研究显得愈发重要和紧迫。在国家磁约束核聚变能发展研究专项的支持下,本文以钨为模型材料,构建金属材料聚变中子辐照模拟平台,解决中子辐照模拟的共性关键技术问题,实现中子级联损伤→辐照微结构→力热性能的多尺度模拟,籍此预测聚变中子辐照条件下材料的行为。 相似文献
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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
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中国聚变工程实验堆(CFETR)是我国自主设计和研制的重大科学工程,CFETR旨在与ITER相衔接和补充,为研制DEMO级别聚变堆电站提供必要的技术。蒙特卡罗方法在聚变中子学与屏蔽设计等方面具有重要作用。本文基于自主化蒙特卡罗程序cosRMC,研究了蒙特卡罗复杂曲面建模的数学模型和计算方法,开发了复杂曲面建模功能,并通过PPCS(power plant conceptual study)模型验证了该功能实现的正确性。然后构建了CFETR的三维精细化模型,并利用该模型对CFETR包层设计中的关键中子学参数进行计算分析。结果表明,cosRMC对中子学参数氚增殖比、中子壁载荷和核热沉积的计算结果与MCNP的计算值吻合良好,相对偏差均小于5%,满足工程设计需求。研究证明了cosRMC应用于聚变堆包层中子学分析的正确性和有效性。CFETR中子学参数的计算分析,也为其设计和优化提供了参考。 相似文献
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《原子能科学技术》2020,(4)
中国聚变工程实验堆(CFETR)是我国自主设计和研制的重大科学工程,CFETR旨在与ITER相衔接和补充,为研制DEMO级别聚变堆电站提供必要的技术。蒙特卡罗方法在聚变中子学与屏蔽设计等方面具有重要作用。本文基于自主化蒙特卡罗程序cosRMC,研究了蒙特卡罗复杂曲面建模的数学模型和计算方法,开发了复杂曲面建模功能,并通过PPCS(power plant conceptual study)模型验证了该功能实现的正确性。然后构建了CFETR的三维精细化模型,并利用该模型对CFETR包层设计中的关键中子学参数进行计算分析。结果表明,cosRMC对中子学参数氚增殖比、中子壁载荷和核热沉积的计算结果与MCNP的计算值吻合良好,相对偏差均小于5%,满足工程设计需求。研究证明了cosRMC应用于聚变堆包层中子学分析的正确性和有效性。CFETR中子学参数的计算分析,也为其设计和优化提供了参考。 相似文献
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中子学分析对聚变堆尤其是其氚增殖包层的设计和安全运行具有重要意义,基于蒙特卡罗方法的模拟是聚变中子学分析的常用手段。以中国聚变工程试验堆(China Fusion Engineering Test Reactor,CFETR)为研究对象,研究蒙特卡罗程序GEANT4在聚变中子学分析中的应用,开展截面库基准测试计算,验证G4NDL截面库在聚变中子学分析中的适用性。采用编程方式和借助McCAD转换方式在GEANT4中分别建立CFETR一维柱壳模型和三维模型,并设置中子源和计数方式,实现了GEANT4中CFETR中子学分析模型的建立。在GEANT4中自主开发了新的物理过程,设置反射面边界,计算获得了中子壁负载。结果表明:GEANT4与MCNP计算结果差异小于1%,验证了反射面设置的有效性和GEANT4在聚变中子学工程分析中应用的可行性。 相似文献
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中国聚变工程试验堆(CFETR)先进材料辐照考验样品所在胶囊结构较为复杂,其内部填充氦气,胶囊肋条尺寸、位置以及胶囊内部填充材料对样品温度影响大。基于STAR-CCM+程序建立CFETR先进小样品辐照装置内胶囊全尺寸模型,针对样品的目标温度,对胶囊的肋条和填充材料进行了调整。对于胶囊内整体样品释热率较低的情况,采用释热率较大的钨材料作为填充材料,可以明显提高整体样品温度;对于局部样品释热率差别较大的情况,调整局部肋条的尺寸和位置,能够很好控制样品间的温度,使样品计算温度满足目标温度范围。结果表明:采用上述方法进行优化后,样品中心温度能够满足目标温度范围,且满足入高通量工程试验堆(HFETR)辐照的热工安全,保证整个辐照任务能够顺利开展。 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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聚变堆等未来先进核能系统要求材料在强流高能中子辐照下长期保持良好的结构稳定性和机械性能。为适应未来先进核能技术发展的需要,中国科学院核能安全技术研究所•凤麟团队牵头研发了具有我国自主知识产权的中国抗中子辐照钢--CLAM钢。CLAM钢的设计考虑了未来核能清洁性的要求,以及苛刻服役环境中材料抗辐照、耐高温、耐腐蚀等性能要求。通过中子学计算分析设计了低活化成分范围,基于选择性纳米相析出进行了抗辐照、耐高温性能优化设计。针对材料的抗辐照性能,利用国内外中子、离子、电子及等离子体辐照设施开展了系列辐照考验研究,通过多角度表征辐照前后材料的微观结构和宏观性能,综合评估了材料的辐照性能,并与国际上同类材料在相近或相同条件下的辐照性能进行了对比分析,结果表明CLAM钢具有良好的抗辐照性能。 相似文献
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Successful development of fusion energy will require the design of high-performance structural materials that exhibit dimensional stability and good resistance to fusion neutron degradation of mechanical and physical properties. The high levels of gaseous (H, He) transmutation products associated with deuterium–tritium (D–T) fusion neutron transmutation reactions, along with displacement damage dose requirements up to 50–200 displacements per atom (dpa) for a fusion demonstration reactor (DEMO), pose an extraordinary challenge. One or more intense neutron source(s) are needed to address two complementary missions: (1) scientific investigations of radiation degradation phenomena and microstructural evolution under fusion-relevant irradiation conditions (to provide the foundation for designing improved radiation resistant materials), and (2) engineering database development for design and licensing of next-step fusion energy machines such as a fusion DEMO.A wide variety of irradiation facilities have been proposed to investigate materials science phenomena and to test and qualify materials for a DEMO reactor. Some of the key technical considerations for selecting the most appropriate fusion materials irradiation source are summarized. Currently available and proposed facilities include fission reactors (including isotopic and spectral tailoring techniques to modify the rate of H and He production per dpa), dual- and triple-ion accelerator irradiation facilities that enable greatly accelerated irradiation studies with fusion-relevant H and He production rates per dpa within microscopic volumes, D–Li stripping reaction and spallation neutron sources, and plasma-based sources.The advantages and limitations of the main proposed fusion materials irradiation facility options are reviewed. Evaluation parameters include irradiation volume, potential for performing accelerated irradiation studies, capital and operating costs, similarity of neutron irradiation spectrum to fusion reactor conditions, temperature and irradiation flux stability/control, ability to perform multiple-effect tests (e.g., irradiation in the presence of a flowing coolant, or in the presence of complex applied stress fields), and technical maturity/risk of the concept. Ultimately, it is anticipated that heavy utilization of ion beam and fission neutron irradiation facilities along with sophisticated materials models, in addition to a dedicated fusion-relevant neutron irradiation facility, will be necessary to provide a comprehensive and cost-effective understanding of anticipated materials evolution in a fusion DEMO and to therefore provide a timely and robust materials database. 相似文献
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《等离子体科学和技术》2016,18(2):179-183
China Fusion Engineering Test Reactor(CFETR) is an ITER-like fusion engineering test reactor that is intended to fill the scientific and technical gaps between ITER and DEMO.One of the main missions of CFETR is to achieve a tritium breeding ratio that is no less than 1.2to ensure tritium self-sufficiency.A concept design for a water cooled ceramics breeding blanket(WCCB) is presented based on a scheme with the breeder and the multiplier located in separate panels for CFETR.Based on this concept,a one-dimensional(1D) radial built breeding blanket was first designed,and then several three-dimensional models were developed with various neutron source definitions and breeding blanket module arrangements based on the 1D radial build.A set of nuclear analyses have been carried out to compare the differences in neutronics characteristics given by different calculation models,addressing neutron wall loading(NWL),tritium breeding ratio(TBR),fast neutron flux on inboard side and nuclear heating deposition on main in-vessel components.The impact of differences in modeling on the nuclear performance has been analyzed and summarized regarding the WCCB concept design. 相似文献
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沙建军 《等离子体科学和技术》2003,5(5):1965-1976
SiC with fiber-reinforced composites, which are presently considered as the primary structural materials in some fusion reactor conceptual designs, are more attractive and competitive for structural materials in a fusion energy system because of its excellent chemical and mechanical properties such as high fracture toughness, induced-low activation, afterheat under 14MeV neutron irradiation environment at elevated temperature, and good compatibility with coolant and breeder materials. Thus it is important to investigate the research progress of advanced SiC composite, including transmuted helium gas, radiation swelling, radiation effects on mechanical properties, irradiation-enhanced creep, fatigue, physical properties associated with fusion design and their critical issues. This report summarized these results and addressed the major critical issues under irradiation conditions. 相似文献
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The China Fusion Engineering Test Reactor (CFETR) is under design,which aims to bridge the gaps between ITER and the future fusion power plant.The neutron wall loading (NWL) depends on the neutron source distribution,which depends on the density and temperature profiles.In this paper,we calculate the NWL of CFETR and study the effects of density and temperature profiles on the NWL distribution along the first wall.Our calculations show that for a 200 MW fusion power,the maximum NWL is at the outer midplane and the vaule is about 0.4 MW m-2.The density and temperature profiles have little effect on the NWL distribution.The value of NWL is determined by the total fusion power. 相似文献