The influence of reactor core parameters on effective breeding coefficient K_(eff)

The values of effective breeding coefficient Keff in a reactor core of nuclear power plant are calculated for different values of parameters （core structure, fuel assembly component） by using the Monte Carlo method. The obtained values of Keff are compared and analysed, which can provide theoretical basis for reactor design.

Feasibility neutronic design for the reactor core configurations of a 5 MWth transportable block-type HTR

Small long-life transportable high temperature gas-cooled reactors(HTRs) are interesting because they can safely provide electricity or heat in remote areas or to industrial users in developed or developing countries.This paper presents the neutronic design of the U-Battery,which is a 5 MWth block-type HTR with a fuel lifetime of 5–10 years.Assuming a reactor pressure vessel diameter of less than 3.7 m,some possible reactor core configurations of the 5 MWth U-Battery have been investigated using the TRITON module in SCALE 6.The neutronic analysis shows that Layout 12×2B,a scattering core containing 2 layers of 12 fuel blocks each with 20% enriched235U,reaches a fuel lifetime of 10 effective full power years(EFPYs).When the diameter of the reactor pressure vessel is reduced to 1.8 m,a fuel lifetime of 4 EFPYs will be achieved for the 5 MWth U-Battery with a 25-cm thick graphite side reflector.Layouts 6×3 and 6×4 with a 25-cm thick BeO side reflector achieve a fuel lifetime of 7 and 10 EFPYs,respectively.The comparison of the different core configurations shows that,keeping the number of fuel blocks in the reactor core constant,the annular and scattering core configurations have longer fuel lifetimes and lower fuel cost than the cylindrical ones.Moreover,for the 5 MWth U-Battery,reducing the fuel inventory in the reactor core by decreasing the diameter of fuel kernels and packing fraction of TRISO particles is more effective to lower the fuel cost than decreasing the 235U enrichment.

Numerical Analysis of Magnetic Force of Dry-Type Air-Core Reactor

This paper presents a coupled magnetic-circuit method for computing the magnetic force of air-core reactor under short-time current. The current and the magnetic flux density are computed first and then the magnetic force is obtained. Thus, the dynamic stability performance of air-core reactor can be analyzed at the design stage to reduce experimental cost and shorten the lead-time of product development.

Calculation and Design of Dry-type Air-core Reactor

Based on the method of compound and additional conditions under the conditions of the equal temperature rise and the equal potential drop (P.D.) of resistance, the application of design software of dry-type air-core reactor is introduced in this thesis. The analytical methods of the inductance are also given. This approach is proved entirely feasible in theory through the simplification with Bartky transformation, and is able to quickly and accurately calculate reactor inductance. This paper presents the analytical methods of the loss of dry-type air-core reactor as well.

Development of an Evaluation Methodology for Fuel Discharge in Core Disruptive Accidents of Sodium-Cooled Fast Reactors

The purpose of the present study is to develop a methodology to evaluate fuel discharge through the CRGT （control-rod guide tube） during CDAs （core-disruptive accidents） of SFRs （sodium-cooled fast reactors）, since fuel discharge will decrease the core reactivity and CRGTs have a potential to provide an effective discharge path. Fuel discharge contains multi-component fluid dynamics with phase changes, and, in the present study, the SFR safety analysis code SIMMER （Sn, implicit, multifield, multicomponent, Eulerian recriticality） was utilized as a technical basis. First, dominant phenomena affecting fuel discharge through the CRGT are identified based on parametric calculations by the SIMMER code. Next, validations on the code models closely relating to these phenomena were carried out based on experimental data. It was shown that the SIMMER code with some model modifications could reproduce the experimental results appropriately. Through the present study, the evaluation methodology for the molten-fuel discharge through the CRGT was successfully developed.

Assessment of Axial Power Peaking Factors in GHARR-1 LEU Core: A Decadal Simulation Analysis

This study aims to thoroughly investigate the axial power peaking factors (PPF) within the low-enriched uranium (LEU) core of the Ghana Research Reactor-1 (GHARR-1). This study uses advanced simulation tools, like the MCNPX code for analysing neutron behavior and the PARET/ANL code for understanding power variations, to get a clearer picture of the reactor’s performance. The analysis covers the initial six years of GHARR-1’s operation and includes projections for its whole 60-year lifespan. We closely observed the patterns of both the highest and average PPFs at 21 axial nodes, with measurements taken every ten years. The findings of this study reveal important patterns in power distribution within the core, which are essential for improving the safety regulations and fuel management techniques of the reactor. We provide a meticulous approach, extensive data, and an analysis of the findings, highlighting the significance of continuous monitoring and analysis for proactive management of nuclear reactors. The findings of this study not only enhance our comprehension of nuclear reactor safety but also carry significant ramifications for sustainable energy progress in Ghana and the wider global context. Nuclear engineering is essential in tackling global concerns, such as the demand for clean and dependable energy sources. Research on optimising nuclear reactors, particularly in terms of safety and efficiency, is crucial for the ongoing advancement and acceptance of nuclear energy.

基于磁热耦合的超大容量干式空心限流电抗器金属结构件温度场研究及试验验证

针对500kV/6000A的干式空心限流电抗器金属结构件发热问题,依托于上海远东-亭卫500kV变电站加装串抗工程,采用磁热耦合仿真分析与试验验证相结合的方法,研究了额定电流6000A,单台阻抗7Ω,单台容量252MVar的大容量限流电抗器金属结构件温度场分布特性。研究结果表明:一是限流电抗器的金属结构件温升试验与仿真结果基本吻合,验证了仿真计算的准确性;二是采用特殊结构设计的铝吊臂通流,热点温升为92K,满足超大容量限流电抗器长期通流的使用要求;三是采用低涡流设计结构的不锈钢支撑结构件,热点温升不超过100K,满足支撑强度要求。

压水堆核电厂换料堆芯装载优化专家系统SEDRIO/INCORE研制

依赖于专家知识建立了大亚湾核电站和秦山第二核电厂换料堆芯装载知识库,在此基础上进行换料堆芯装载方案启发式优化搜索。应用已用于工程设计的二维细网堆芯燃料管理程序系统(INCORE)进行装载方案评价,采用循环长度和堆芯功率峰因子综合指标计算装载方案的价值并评价其优劣程度。用该系统分别对大亚湾核电站二号堆第四循环和秦山第二核电厂第四循环堆芯优化方案搜索计算。结果表明,无论从堆芯径向功率峰因子还是从循环长度指标来看,专家系统SEDRIO/INCORE搜索得到的装载方案都明显优于参考方案。

Neutronics physics analysis of a large fluoride-salt-cooled solidfuel fast reactor with Th-based fuel

Fast reactors based on thorium fuel have enhanced inherent safety. Fluoride salt performs well as a coolant in high-temperature nuclear systems. In this paper,we present a reference core for a large fluoride-salt-cooled solid-fuel fast reactor(LSFR) using thorium–uranium fuel cycle. Neutronics physics of the LSFR reference core is investigated with 2D and 3D in-core fuel management strategy. The design parameters analyzed include the fuel volume fraction, power density level and continuous removal of fission products with 3D fuel shuffling that obtains better equilibrium core performance than 2D shuffling. A self-sustained core is achieved for all cases,and the core of 60% fuel volume fraction at 50 MW/m^3 power density is of the best breeding performance(average breeding ratio 1.134). The LSFR core based on thorium fuel is advantageous in its high discharge burn-up of 20–30% fissions per initial heavy metal atom, small reactivity swing over the whole lifetime(to simplify the reactivity control system), the negative reactivity temperature coefficient(intrinsically safe for all cases) and accepted cladding peak radiation damage. The LSFR reactor is a good alternative option for the deployment of a self-sustained thorium-based nuclear system.

“玲龙一号”小堆堆芯与安全设计

“玲龙一号”(ACP100)作为一款多用途小型模块化反应堆,是我国核电技术自主研发的标志性成果。“玲龙一号”在研发过程、堆芯和安全设计中的关键技术,主要涵盖堆芯核设计、热工水力设计、安全设计理念、固有安全设计、事故应对策略等,通过引入“全非能动”的安全设计理念,同时融合确定论与概率论的分析方法,大幅提升了“玲龙一号”的安全性,全面满足并超越了三代核电安全标准。

快中子反应堆堆芯物理分析方法的研究现状与发展建议

快中子反应堆在核能可持续发展中扮演着重要角色,是各核强国都在致力发展的重点堆型。快堆物理计算是快堆核设计的基础,其计算方法的研究和相应计算程序的开发是快堆理论研究和数值模拟技术发展中极其重要的环节。本文对国内外快堆物理计算方法,特别是近20年来的发展变化进行了系统梳理,以对国内外专用和通用快堆物理计算程序的总结为线索,介绍了快堆物理分析理论体系的发展情况,对其中体现出的一致性特点和最近几年发展的趋势进行了分析,并对我国快堆堆芯物理分析方法的发展给出了建议,为我国快堆物理计算理论的进步和自主化的物理分析软件研发提供参考。

钠冷快堆小栅板联箱压降对组件流量分配影响研究

钠冷快堆堆芯采用大栅板联箱、小栅板联箱和组件的三级流量分配方式,小栅板联箱的压降影响组件的流量分配,进而影响堆芯的安全,因此进行钠冷快堆小栅板联箱压降对组件流量分配影响研究有重要意义。根据小栅板联箱压降造成组件流量分配偏差的机理,提出了理论计算模型和堆芯组件优化设计的方法,并针对中国实验快堆(CEFR)堆芯进行了组件压降的优化设计,通过优化设计降低了CEFR燃料组件流量分配负偏差。结果表明,在进行钠冷快堆堆芯热工水力设计时,需要结合实际堆芯布置分析组件压降设计值的优化方向,并进行敏感性分析,以确定组件的最优设计压降,将小栅板联箱压降对组件流量分配影响降低到最低程度。本文结果可为钠冷快堆堆芯热工水力设计提供参考。

数值反应堆堆芯与E级高性能计算的科学内涵

反应堆是一复杂的系统过程,是由中子场、温度场、流场、应力场、化学场等多个物理过程相互耦合的装置,这些物理场涉及从微观核反应到宏观能量释放的多尺度作用机理。随着E级(1 000PFLOPS,百亿亿次/每秒)计算机的问世,核能发展的总趋势正从传统工程驱动模式向以高性能数值模拟为主转变。当前四代堆设计立足小型化和精密化,高分辨率数值模拟对提升核装置性能和降低裕量作用突出。为研究解决当前模拟软件与计算机之间存在的浮点效率低、移植周期长、模式通用难和规模扩展难等问题的办法,突破软件和硬件之间存在的编程墙和性能墙,本文通过解读美国NEAMS、CASL和ECP计划,结合团队近年在数值反应堆和高性能计算关键技术突破方面的经验,提出基于并行中间件的集成共性、发展个性的技术路线,探索一条快速提升我国自主CAE软件整体水平的途径,供业内同行探讨,以在国产超级计算机上实现核装置的精细化建模和多物理、多尺度、多过程耦合计算。

月球基地用反应堆电源方案研究

针对月球基地反应堆电源应用需求及月球的特殊使用环境,提出了一种锂回路冷却快堆方案。采用MCNP程序对该方案的核特性进行深入研究,详细计算分析了反应性、燃耗、通量及功率分布等堆芯物理量,并对反应堆的掉落临界安全和屏蔽进行了详细的计算分析。研究结果表明:锂冷回路冷却快堆方案具有堆芯结构紧凑、质量轻、导热效率高、堆芯固有安全性高、功率输出性能好、停堆深度深等优点,适合用作月球探索活动的能量源。

Effect on the Flow Behaviors by Adding Internals in a Riser Reactor

Riser reactor is a key unit in the Fluid Catalytic Cracking (FCC), and it has important influences on increasing the yield coefficient of gas and oil. In this paper, the behaviors of gas-solid two-phase flow in the traditional y-type riser reactor are investigated by numerical simulation. The calculated particle concentration distribution is in good agreement with the experimental data, which verified the advanced models and calculating methods. The non-uniform distribution, such as core-annulus flow, may result in the unreasonable matching relationship of catalyst-to-oil ratio. An optimized riser with cuneal internals is proposed and the comparison of two different structures of riser reactor is presented. The comparison results show that the cuneal internals in the riser both can block effectively the slip down of the particles near wall region and weaken core-annulus flow structure due to the redistribution of particles. The results also prove that the particle concentration distribution becomes uniform along the axial and radial direction in the optimized riser by adding cuneal internals, which would be benefits for the catalytic cracking reactions.

基于小波分析的空心电抗器匝间短路磁场探测方法研究

为了解决目前干式空心电抗器匝间短路检测方法灵敏度和准确度不足的问题,根据故障发展期探测线圈感应电压的异常变化,提出一种新的基于小波分析的空心电抗器匝间短路磁场探测方法。首先,对电抗器故障发展期探测线圈电压信号进行仿真分析,得到电抗器不同位置匝间短路时探测线圈信号变化特征。再基于信号分解小波系数中有用信号与噪声相对比值最大的原则,逐层自适应选取最优小波基函数,利用小波变换对信号进行阈值去噪,提取故障特征量从而实现对电抗器匝间短路发展期及时准确地判断,避免故障进一步发展。通过搭建电抗器匝间短路故障检测系统实验平台,验证了基于小波分析磁场探测法的灵敏度和可靠性,为电抗器的安全稳定运行提供有效保障。

Preliminary study of the tight lattice pressured heavy water reactor loaded with Pu/U and Th/U mixed fuels

To improve nuclear fuel utilization efficiency and prolong fuel cycle burn-up, a tight pitch lattice pressured heavy water reactor was investigated as an alternative of next generation of power reactors. It is shown that the high conversion ratio and negative coolant void reactivity coefficient are challenges in the reactor core physics designs. Various techniques were proposed to solve these problems. In this work, a tight pitch lattice and mixed fuel assemblies pressured heavy water reactor concept was investigated. By utilizing numerical simulation technique, it is demonstrated that reactor core mixed with Pu/U and Th/U assemblies can achieve high conversion ratio (0.98), long burn-up (60 GWD/t) and negative void reactivity coefficients.

钠冷快堆堆芯捕集器设计优化数值研究

堆芯捕集器是为有效分散堆芯熔融物并防止压力容器下封头与熔体大规模接触而被破坏的非能动预防和缓解装置。在钠冷快堆(SFR)中,堆芯捕集器的结构直接影响碎片床的堆积形状和分布,进而影响碎片床的再临界性和长期衰变热去除能力。本文针对堆芯捕集器的结构设计优化开展数值研究,重点关注其烟囱结构设计对碎片床形成和分布的影响机理及规律。基于无量纲刚度系数和无量纲阻尼系数改进离散元法(DEM),通过改变堆芯捕集器烟囱顶盖垂直投影边长、顶盖倾斜角度和烟囱间距,研究碎片颗粒的运动和碎片床的形成行为。结果表明,堆芯捕集器的烟囱顶盖垂直投影边长、烟囱顶盖倾角和烟囱间距对碎片床的堆积形状和分布均有重要影响,碎片颗粒的二次散射对于改善碎片床的均匀性至关重要。

TOPAZ-2型热离子反应堆电源中子物理代理模型开发

为建立TOPAZ-2型热离子反应堆电源中子物理快速计算方法,以对全范围仿真系统的堆芯计算提供新的辅助手段,采用神经网络方法,对堆芯的有效增殖系数keff和功率分布进行预测。通过MCNP得到不同转鼓角度组合下的高精度中子物理计算结果,将转鼓角度组合和对应的keff、功率分布结果分别作为输入和输出数据,建立代理模型。研究结果表明,代理模型得到的keff和功率分布都与MCNP结果符合较好,且计算时长大幅缩减。因此,研究建立的代理模型能够用于TOPAZ-2型热离子反应堆电源中子物理快速计算。