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聚变装置工程模型极其复杂,使得中子学分析的建模十分繁琐和耗时。开源蒙特卡罗程序OpenMC通过集成DAGMC(Direct Accelerated Geometry Monte Carlo),可以直接基于CAD模型进行粒子输运模拟计算,该特性可显著提高复杂工程模型的建模与分析效率。以中国聚变工程试验堆(China Fusion Engineerging Test Reactor,CFETR)为对象,开展OpenMC在聚变中子学分析中的应用研究。基于CFETR一维柱壳模型验证OpenMC与MCNP计数结果的一致性。根据等离子体空间分布特点,基于源扩展接口自定义源类和源函数准确描述复杂聚变中子源。利用DAG-OpenMC的CAD几何功能成功建立了CFETR的三维模型,并计算获得了中子壁负载分布、氚增殖率和核热沉积等物理量。结果表明:DAG-OpenMC与MCNP的计算结果具有极好的一致性。在建立复杂的聚变堆工程模型时,基于CAD几何功能极大地提高了建模效率。DAG-OpenMC在聚变中子学应用中关键问题的验证表明了其处理复杂工程结构条件下聚变中子学问题的可行性。 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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《核技术》2017,(2)
在托卡马克实验装置中,D-T聚变反应释放出的14 Me V高能中子,与周围部件接触会引起包层材料活化、热负载过高等一系列问题,因此在包层设计和优化过程中,相关的中子学计算显得尤为重要。为了研究不同描述的中子源模型对中国聚变工程实验堆(China Fusion Engineering Test Reactor,CFETR)中子学计算的影响,使用基于蒙特卡罗方法的MCNP(Monte Carlo N Particle Transport Code)程序来模拟中子输运过程,分别计算点源、均匀体源、基于L、H、A模约束的中子源模型对不同中子学计算的影响。结果表明,不同描述的中子源模型对氚增殖比影响较小,对中子壁负载和功率密度分布影响比较明显。 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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利用D-T聚变中子源转化长寿命裂变产物的中子学可行性研究 总被引:4,自引:0,他引:4
从中子学角度研究长寿命裂变产物在Tokamak型D-T聚变堆包层中转化的可行性.提出了用可裂变Pu增殖中子的混合包层转化方案,研制了相应的燃耗计算程序及数据库,并对所提方案进行了计算和分析.结果表明,在可预见的聚变堆芯技术条件下,所研究的概念性包层可对长寿命裂变产物进行有效转化. 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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双功能液态锂铅(DFLL)包层作为一种高性能的产氚包层,是中国聚变工程实验堆(CFETR)的候选包层之一。氚增殖比(TBR)是产氚包层核心设计参数之一,是评估聚变堆氚自持性能的重要指标,有必要对其进行详细分析。本文介绍了DFLL包层氚增殖性能数值分析与实验验证工作。其中数值分析采用中子输运设计与安全评价软件系统SuperMC建立了全堆三维中子学模型,计算分析了包层不同位置处的TBR值,并对影响包层氚增殖性能的相关因素,如第一壁材料、钨护甲、包层增殖区厚度、6 Li富集度等进行了敏感性分析及参数优化;实验验证工作利用强流聚变中子源HINEG,开展了DFLL中子学实验模块不同位置产氚率(TPR)测量。研究结果显示,经过优化的DFLL包层TBR设计参数可达到1.208,满足CFETR第一阶段的氚自持要求;实验结果与理论计算结果的最大偏差为8%以内符合,实验结果的测量不确定度在2σ内优于9.8%。 相似文献
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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... 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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Clement Fausser Antonella Li Puma Franck Gabriel Rosaria Villari 《Fusion Engineering and Design》2012,87(5-6):787-792
Neutronic studies of European demonstration fusion power plant (DEMO) have been so far based on plasma physics low confinement mode (L-mode). Future tokamaks, nevertheless, may likely use alternative confinement modes such as high or advanced confinement modes (H&A-mode). Based on analytical formulae used in plasma physics, H&A-modes D-T neutron sources formulae are proposed in this paper. For that purpose, a tokamak random neutron source generator, TRANSGEN, has been built generating bidimensional (radial and poloidal) neutron source maps to be used as input for neutronics Monte-Carlo codes (TRIPOLI-4 and MCNP5). The impact of such a source on the neutronic behavior of the European DEMO-2007 Helium-cooled lithium–lead reactor concept has been assessed and compared with previous results obtained using a L-mode neutron source. An A-mode neutron source map from TRANSGEN has been used with the code TRIPOLI-4. Assuming the same fusion power, results show that main reactor global neutronic parameters, e.g. tritium breeding ratio and neutron multiplication factor, evolved slightly when compared to present uncertainties margin. However, local parameters, such as the neutron wall loading (NWL), change significantly compared to L-mode shape: from ?22% to +11% for NWL. 相似文献
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Effect of various spatial and energy distributions of fusion neutron source on the calculation of neutron wall loading of Tokamak D-D fusion device has been investigated by means of the 3-D Monte Carlo code MCNP.A realistic Monte Carlo source model was developed based on the accurate representation of the spatial distribution and energy spectrum of fusion neutrons to solve the complicated problem of tokamak fusion neutron source modelling.The results show that those simplified source models will introduce significant uncertainties.For accurate estimation of the key nuclear responses of the tokamak design and analyses,the use of the realistic source is recommended.In addition,the accumulation of tritium produced during D-D plasma operation should be carefully considered. 相似文献
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《等离子体科学和技术》2016,18(8):865-869
Attaining tritium self-sufficiency is an important mission for the Chinese Fusion Engineering Testing Reactor(CFETR) operating on a Deuterium-Tritium(D-T) fuel cycle. It is necessary to study the tritium breeding ratio(TBR) and breeding tritium inventory variation with operation time so as to provide an accurate data for dynamic modeling and analysis of the tritium fuel cycle. A water cooled ceramic breeder(WCCB) blanket is one candidate of blanket concepts for the CFETR. Based on the detailed 3D neutronics model of CFETR with the WCCB blanket,the time-dependent TBR and tritium surplus were evaluated by a coupling calculation of the Monte Carlo N-Particle Transport Code(MCNP) and the fusion activation code FISPACT-2007.The results indicated that the TBR and tritium surplus of the WCCB blanket were a function of operation time and fusion power due to the Li consumption in breeder and material activation.In addition, by comparison with the results calculated by using the 3D neutronics model and employing the transfer factor constant from 1D to 3D, it is noted that 1D analysis leads to an over-estimation for the time-dependent tritium breeding capability when fusion power is larger than 1000 MW. 相似文献
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《等离子体科学和技术》2016,18(7):775-780
In order to investigate the nuclear response to the water-cooled ceramic breeder blanket models for CFETR, a detailed 3D neutronics model with 22.5otorus sector was developed based on the integrated geometry of CFETR, including heterogeneous WCCB blanket models,shield, divertor, vacuum vessel, toroidal and poloidal magnets, and ports. Using the Monte Carlo N-Particle Transport Code MCNP5 and IAEA Fusion Evaluated Nuclear Data Library FENDL2.1,the neutronics analyses were performed. The neutron wall loading, tritium breeding ratio, the nuclear heating, neutron-induced atomic displacement damage, and gas production were determined.The results indicate that the global TBR of no less than 1.2 will be a big challenge for the watercooled ceramic breeder blanket for CFETR. 相似文献
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吴宜灿 刘超 宋钢 王永峰 李桃生 汪建业 蒋洁琼 赵柱民 宋勇 胡丽琴 黄群英 李亚洲 王文 王志刚 王刚 季翔 王亮 王为田 于前锋 黄国强 程雄卫 王飞鹏 张思纬 李雅男 韩运成 宋婧 龙鹏程 FDS团队 《核科学与工程》2016,(1):77-83
强流氘氚聚变中子源HINEG(High Intensity D-T Fusion Neutron Generator)研发分两期:HINEG-Ⅰ为直流脉冲双模式,已成功产生中子强度1.1×10~(12)n/s的氘氚聚变中子,并实现连续稳定运行;HINEG-Ⅱ中子强度设计指标为10~(14)~10~(15)n/s量级,重点突破强流离子源和高载热氚靶技术。HNEG中子源可开展中子学方法程序与核数据、辐射屏蔽与防护、材料活化与辐照损伤机理和部件中子学性能等核能与核安全研究,同时也可在核医学与放射治疗、中子照相等领域拓展核技术应用研究。本文简要介绍HINEG总体设计方案与关键技术研究进展。 相似文献