cosRMC复杂曲面开发及在聚变中子学的应用

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。     

DAG-OpenMC在聚变中子学分析中的应用研究

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。     

托卡马克聚变中子源的数值模型研究

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。     

中国聚变工程试验堆中子学分析中的GEANT4应用研究

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。     

COSINE堆用蒙卡分析软件cosRMC研发与应用

为进一步提升核电软件自主化能力,研发了核电厂设计与安全分析一体化软件包COSINE。其中cosRMC为堆用三维中子-光子-电子输运蒙卡软件,已具备输运计算、燃耗计算、群常数产生、敏感性及不确定性分析、可视化建模等功能,可用于堆芯设计分析、确定论校核计算以及辐射屏蔽计算。本文从cosRMC的计算功能以及软件在先进非能动型压水堆(AP1000)与中国聚变工程实验堆(CFETR)中的典型应用对cosRMC软件的研发现状进行介绍。其中,AP1000堆芯的模拟结果显示,21种燃料组件及全堆芯模型的增殖因子绝对值最大偏差为89.9×10~(-5),功率分布计算结果绝对值最大偏差为2.1%;CFETR的模拟结果显示,氚增殖比的最大绝对值偏差为0.6%,cosRMC网格权窗功能可以有效解决模拟过程中的深穿透问题。cosRMC软件计算功能可满足压水堆、聚变堆等大型复杂模型的计算需求,软件具有较高的计算精度,同时可视化建模工具可有效提升建模效率及正确性。     

CFETR第一壁及赤道面外包层中子辐照损伤初步分析

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。     

双功能液态锂铅包层氚增殖性能分析

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%。     

基于中国聚变工程试验堆的双功能液态铅锂包层活化分析和废物处理

使用中子输运设计与安全评价软件系统(SuperMC)和聚变评价数据库JEFF3.2,根据中国聚变工程试验堆(CFETR)第一阶段设计要求,对双功能液态铅锂包层中各部件的活化特性进行计算和分析。采用燃耗输运耦合方法计算了聚变堆赤道面内、外包层中各部件放射性活度、衰变余热、剂量率和潜在生物危害随停堆冷却时间的变化,并根据欧洲聚变堆安全和环境评估策略中核废料处理标准,分析了聚变堆退役后氚增殖包层的废料处理问题。分析结果表明:在功率200 MW时正常运行10 a条件下,包层中各部件在经过50 a冷却后均可达到简单回收标准,满足CFETR第一阶段放射性废物处理要求。     

聚变堆固态产氚包层中子学优化方法研究与应用

为提升聚变堆包层产氚性能,更好地满足氚自持要求,首先,基于中子微扰理论与模拟退火算法开发了适用于聚变堆产氚包层（TBB）中子学优化新算法与新程序。其次,选取中国聚变工程实验堆（CFETR）氦冷固态包层,完成了全堆中子学性能优化的示范性应用。最后,对优化后的包层方案进行了热工、流体、结构的三维有限元校核。结果表明：(1)相比于传统包层中子学优化算法,本文所提出的优化算法具有更好的优化效果与更高的优化效率;(2)本文所开发的智能优化程序可更好地满足聚变堆TBB中子学优化与设计的需求,可为包层设计提供算法理论基础与程序支撑。     

CFETR氦冷陶瓷增殖包层中子学分析

为满足中国聚变工程实验堆(CFETR)包层的应用要求,本文提出氦冷陶瓷增殖(HCCB)包层方案。为验证HCCB包层设计方案的合理性与可行性,采用三维蒙特卡罗粒子输运程序MCNP,计算和分析了HCCB包层方案的氚增殖比、中子壁负载、中子通量密度、核热、辐照损伤等中子学特性。结果表明,HCCB包层方案满足氚自持要求,中子通量密度和核热分布合理,屏蔽性能良好,基本满足设计要求。     

Analysis of Time-Dependent Tritium Breeding Capability of Water Cooled Ceramic Breeder Blanket for CFETR

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.     

Impact Analysis of Helium Cooled Solid Blanket Structures on the Tritium Breeding Performance in CFETR

Chinese Fusion Engineering Test Reactor (CFETR) is a test tokamak reactor to bridge the gap between ITER and future fusion power plant. As its objectives are to demonstrate generation of fusion power and to realize tritium self-sufficiency, the tritium breeding ratio (TBR) is a key design parameter. In the blanket design and optimization, the structures such as the first wall (FW), cooling plate (CP), stiffening plate (SP), cap and some other design parameters in detailed 3-D model have significant impacts on the tritium breeding performance. Based on a helium cooled solid breeder blanket option for CFETR, the impact analysis of the helium cooled solid blanket structures on tritium breeding performance was performed in this paper. Firstly, the detailed 3D neutronics model was built by using of a CAD to Monte Carlo Geometry conversion tool McCad. Then based on the detailed 3D neutronics model, the impact analyses of the blanket structures on tritium breeding performance were carried out, which include the FW, CP, SP, cap and side wall. By the sensitivity study of the blanket structures on the TBR, it gave the TBR variation trend and references for the blanket design and optimization.     

中国聚变工程试验堆包层的核热耦合效应研究

本文以中国聚变工程试验堆（CFETR）的氦冷固态包层和水冷固态包层为研究对象,基于蒙特卡罗程序MCNP和计算流体力学程序FLUENT,利用3D-1D-2D耦合方法和伪材料方法,分别对200 MW的氦冷固态包层和水冷固态包层及1.5 GW的水冷固态包层方案进行了核热耦合计算分析。研究结果表明,金属铍的热散射效应和轻水密度是聚变包层核热耦合效应的主要来源,核热耦合效应对氦冷固态包层的影响可忽略,对水冷固态包层的氚增殖比和温度分布有一定程度的影响。     

NAPTH: Neutronics analysis code system for the fusion–fission hybrid reactor with pressure tube type blanket

The fusion–fission hybrid reactor is considered as a potential path to the early application of fusion energy. A new concept with pressure tube type blanket has recently been proposed for a feasible hybrid reactor. In this paper, a code system for the neutronics analysis of the pressure tube type hybrid reactor is developed based on the two-step calculation scheme: the few-group homogeneous constant calculation and the full blanket calculation. The few-group homogeneous constants are calculated using the lattice code DRAGON4. The blanket transport calculation is performed by the multigroup Monte Carlo code. A link procedure for fitting the cross sections is developed between these two steps. An additional procedure is developed to calculate the burnup, power distribution, energy multiplication factor, tritium breeding ratio and neutron multiplication factor. From some numerical results, it is found that the code system NAPTH is reliable and exhibits good calculation efficiency. It can be used for the conceptual design of the pressure tube type hybrid reactor with precise geometry.     

Neutronics design analyses of fusion power reactors based on a novel integral approach

A novel integral approach was applied for the nuclear design analyses performed for the European DEMO Conceptual Study including, for the first time, the automatic generation of analysis models for Monte Carlo calculations from available CAD geometry data by the new McCad conversion tool. Starting from neutronics pre-analyses to define the radial reactor build, a generic neutronics CAD model of the DEMO reactor was constructed serving as basis for the generation of DEMO reactor models employing the HCLL (Helium Cooled Lithium Lead) and HCPB (Helium Cooled Pebble Bed) blankets for the tritium breeding and the power production. The HCLL and HCPB DEMO reactor models were converted to the MCNP geometry representation by the newly developed McCad software tool. The nuclear analyses performed on the basis of MCNP Monte Carlo calculations using the converted models showed that both DEMO reactor variants could satisfy the requirements for a sufficient shielding and tritium breeding performance. As a major outcome of this work it is concluded that the newly established integral approach for neutronics analyses, including the automatic generation of analysis geometry models by McCad, is mature for applications to reactor design analyses and studies.