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。     

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。     

AP1000机组首次临界试验验证计算

首次临界试验是压水堆核电厂调试启动过程的关键环节，旨在确认核反应堆堆芯能按照设计要求达到预期的临界运行状态。本文利用西安交通大学自主研发的NECP-Bamboo程序系统对AP1000机组堆芯的首次临界试验的设计结果进行了验证计算，并与AP1000堆芯的核设计结果进行了比较。计算结果表明：预估临界状态下的硼浓度的偏差为-15 ppm，控制棒积分价值的最大偏差为-52 pcm，硼微分价值的偏差不超过0.2 pcm/ppm，反应性温度系数的偏差不超过1 pcm/K。本文计算结果的精度与高保真计算程序KENO（概率论方法）和VERA（确定论方法）的计算精度相当，为确保AP1000堆芯调试启动阶段的核安全提供了进一步的数据支撑。     

反应堆堆芯先进中子学模拟软件SCAP-N研发

堆芯中子学计算是反应堆设计分析的基础,为提高堆芯中子学计算的模拟分辨率与计算精度,开发了反应堆堆芯先进中子学模拟软件（SCAP-N）。该程序首先根据轴向特征对堆芯进行分层,并逐层进行二维堆芯非均匀输运计算,再采用超级均匀化方法（SPH）获得栅元等效均匀化截面,最后进行三维堆芯逐棒（pin-by-pin）输运计算,获得堆芯有效增殖因子与精细棒功率分布。为提高程序计算效率,采用分布式/共享式（MPI/OPENMP）混合并行方式对程序进行了并行化开发。利用虚拟反应堆（VERA）系列基准例题及美国先进非能动压水堆（AP1000）启动物理试验实测数据对程序进行了测试验证。结果表明,相比于商用核设计程序系统,SCAP-N程序采用的逐棒输运技术能够提高堆芯中子学的计算精度。与同类型高精度中子学程序相比,SCAP-N具有更高的计算效率,可进一步提高核电厂的经济性及运行灵活性。      

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。     

HFETR三维堆芯输运燃料管理程序的应用

阐述了高通量工程试验堆(HFETR)三维堆芯输运燃料管理计算软件HEFT的研发背景,介绍了HEFT主要程序模块HEFT-lat、LINK、HEFT-core、HEFT-int的功能及应用方向,给出了栅元参数计算方法、考验装置计算方法,以及HEFT程序对HFETR零功率临界堆芯、最近19炉段堆芯、考验组件的校算结果,对具有实测值的有效增殖系数keff、停堆棒位、中子注量率、燃耗进行对比分析。结果表明,HEFT程序所采用的栅元计算、堆芯计算的模型和方法正确,可应用于HFETR堆芯燃料管理。     

欠热条件下非能动余热排出热交换器传热试验数值模拟

针对大型先进压水堆非能动余热排出热交换器设计和安全分析计算模型存在的重要缺陷,以AP1000的非能动余热排出热交换器为原型,采用3根C型管进行了非能动余热排出热交换器传热试验。然后采用流体计算软件对欠热试验工况进行了数值模拟,通过多次计算得到了传热管外传热计算可采用的传热关系式,选取的传热模型下的计算结果与试验结果符合较好。利用传热模型验证了AP1000的设计工况,发现AP1000非能动余热排出热交换器的设计能带走堆芯余热。本文研究可为大型先进压水堆设计和安全分析提供技术支撑。     

舰船用反应堆停堆氙毒特性仿真研究

结合船用堆的特点，对核电站反应堆正方形燃料组件堆芯仿真软件进行修改和移植，开发可用于研究船用堆非干净六边形燃料组件堆芯中毒碘坑的堆芯仿真软件。应用该软件，对燃耗为30MW•d的某反应堆进行了碘坑仿真，并与点模型仿真结果进行了比较。结果表明：点模型的仿真结果偏小，用本软件进行仿真，平衡氙毒计算值与实测值间的偏差为-0.8%，最大氙毒计算值与实测值间的偏差为-4.3%，碘坑计算值与实测值间的偏差为-0.5%。本软件仿真结果符合实际运行规律和物理规律，因此，本软件可准确模拟碘坑中毒，对提高船用堆机动性和安全运行有重要的理论意义和应用价值。     

AP1000冷管段小破口失水事故分析

基于压水堆最佳估算程序RELAP5/MOD3.4,对AP1000的冷却剂系统和非能动堆芯冷却系统进行建模分析,得到了系统压力、破口流量、燃料包壳温度等关键参数的瞬态变化,计算结果与西屋公司采用NOTRUMP程序计算的结果基本一致。分析表明：AP1000的非能动专设安全设施能有效地对一回路进行冷却和降压,防止堆芯过热,验证了AP1000发生冷管段小破口失水事故后的安全性。     

深穿透跨尺度辐射场分析软件NECP-MCX研发及应用

西安交通大学核工程计算物理实验室自主研发了深穿透跨尺度辐射场分析软件NECP-MCX。针对大空间伽马射线辐射输运模拟、聚变堆停堆剂量模拟和点源屏蔽问题等新应用场景下的新问题与新挑战,在NECP-MCX中研发了对应的新方法与新功能。针对km尺度的伽马射线辐射输运问题,提出一致性共轭驱动重要性抽样(CADIS)-下次事件估计器(NEE)耦合方法,该方法能够精确高效地获得km尺度距离处的光子通量密度,计算效率比传统的NEE高6.8倍;针对聚变堆停堆剂量问题,采用粒子输运-燃耗-活化-源项耦合分析方法,获得PF线圈、TF线圈、真空室和偏滤器处停堆剂量随停堆时间的变化;对于点源屏蔽问题,提出首次碰撞源(FCS)-CADIS方法,解决CADIS方法对点源进行源偏倚的局限性,FCS-CADIS方法的计算效率比CADIS方法高2倍。     

基于RELAP5的CFETR氦冷陶瓷增殖包层模块热工安全分析

本文对中国聚变工程实验堆(CFETR)氦冷陶瓷增殖(HCCB)包层进行热工安全分析。采用大型反应堆瞬态分析程序RELAP5对HCCB包层建模,并进行稳态分析和假设事故的模拟。计算结果表明,CFETR HCCB包层在真空室内氦气泄漏和增殖区盒内氦气泄漏事故中均未出现结构材料熔化,同时各部分的压强变化情况均未超出设计阈值,包层系统在事故发生后均能有效快速地排出余热。CFETR HCCB包层的设计满足热工安全方面的要求。     

RAMI Analysis Program Design and Research for CFETR (Chinese Fusion Engineering Testing Reactor) Tokamak Machine

Chinese Fusion Engineering Testing Reactor (CFETR) is a test reactor which shall be constructed by National Integration Design Group for Magnetic Confinement Fusion Reactor of China with an ambitious scientific and technological goal. The reactor has the equivalent scale compared with ITER, but has the complementary function to ITER. CFETR is a demonstration of long pulse or steady-state operation with duty cycle time not less than 0.3–0.5 and the full cycle of tritium self-sustained with TBR not less than 1.2. At the same time it will be exploring options for DEMO blanket and divertor with an easy changeable core by remote handling way. To be able to reach its scientific and technological objectives, as one of technical risks control methods, RAMI analysis need to be done during the hold lifetime of CFETR, from conception design to decommissioning. Base on stating of CFETR lifetime and preliminary operational programme, the RAMI analysis program and process are designed and discussed, it consists of five major steps: (1) functional analysis are performed, (2) calculating reliability block diagrams, (3) analyzing failure mode, effects and criticality analysis, (4) risk mitigation actions are taken to ensure every system is compatibility with RAMI objectives, (5) All the RAMI analysis are integrated as the final RAMI analysis reports to be reviewed in the system final design review. Along with the elements of the analysis the vacuum vessel (VV) system was performed to provide as examples, detailed showing how the CFETR RAMI analysis is carried out. CFETR RAMI analysis guidelines were designed and established, after constantly revised and improved these analysis criteria and programs will become the basis standards for CFETR RAMI analysis. Preliminary RAMI analysis of CFETR VV system was obtained, which will be updated with the VV system design progresses.     

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.     

CAP1400数值反应堆系统关键技术研究及示范应用

本文系统介绍了“大型先进压水堆及高温气冷堆核电站”国家科技重大专项课题“CAP1400数值反应堆关键技术研究”的主要研究成果。课题首先分别开发了基于确定论方法和蒙特卡罗方法的高保真堆芯物理计算程序,然后开发了pin by pin先进子通道分析程序和基于精细网格的燃料棒性能分析程序,以此为基础建立了物理 热工 燃料性能多物理耦合的CAP1400数值反应堆系统。利用国际基准题VERA、AP1000启动物理实验参数对数值反应堆系统进行了验证和确认,并进一步实现了CAP1400大型先进压水堆的启动物理参数、循环模拟分析和部分功率能力分析的示范应用。数值结果表明,所开发的数值反应堆关键分析软件具有很高的计算精度,可直接服务于CAP1400的设计验证、物理启动和运行支持。     

Neutronics Comparison Analysis of the Water Cooled Ceramics Breeding Blanket for CFETR

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.     

Effects of kinetic profiles on neutron wall loading distribution in CFETR

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.     

Load following control and global stability analysis for PWR core based on multi-model,LQG, IAGA and flexibility idea

The work is to design a nonlinear Pressurized Water Reactor (PWR) core load following control system and analyze the global stability of this system. On the basis of modeling a nonlinear PWR core, linearized models of the core at five power levels are chosen as local models of the core to substitute the nonlinear core model in the global range of power level. The combination control strategy of the Linear Quadratic Gaussian (LQG) control and the Proportional Integral Derivative (PID) control with an optimization process of Improved Adaptive Genetic Algorithm (IAGA) proposed is used to contrive a combined controller with the robustness of a core local model as a local controller of the nonlinear core. Based on the local models and local controllers, the flexibility idea of modeling and control is presented to design a decent controller of the nonlinear core at a random power level. A nonlinear core model and a flexibility controller at a random power level compose a core load following control subsystem. The combination of core load following control subsystems at all power levels is the core load following control system. The global stability theorem is deduced to define that the core load following control system is globally asymptotically stable within the whole range of power level. Finally, the core load following control system is simulated and the simulation results show that the control system is effective.     

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。