CANDU6慢化剂系统丧失冷却情况下的温度分析

压力管卧式重水反应堆(CANDU6)具有相互独立的冷却剂系统和慢化剂系统。慢化剂系统将堆芯高能裂变中子慢化到能维持持续裂变所需的热中子水平,并将慢化中子过程中产生的热量带出。在反应堆大修期间,需要对再循环冷却水系统(RCW)进行检修,则需要并投入其备用系统,但是RCW备用系统仅对反应堆冷却剂系统进行冷却,不提供慢化剂系统热交换器冷却水。所以在RCW备用系统投入的情况下,慢化剂系统丧失冷却。为判断在此情况下慢化剂的温度变化情况,本文对CANDU6大修期间慢化剂系统丧失冷却情况下的温度变化进行分析并与试验结果进行比较,评估是否会由于温度过高而导致系统失效。     

空间核反应堆电源热工水力特性研究综述

随着空间探索领域的快速发展，研究高功率、安全、可靠的空间核反应堆电源将变得愈发重要。本文针对国内外空间核反应堆电源的热工水力关键问题，即空间堆系统稳态和事故瞬态研究、堆芯单冷却剂通道及全堆芯的三维流动换热、静态与动态热电转换装置分析、热工水力特性试验研究等进行研究，分析了空间核反应堆电源热工水力研究的趋势。本文结果可为空间核反应堆电源设计分析及热工水力安全特性研究提供帮助和指导。     

空间核反应堆电源热工水力特性研究综述

随着空间探索领域的快速发展,研究高功率、安全、可靠的空间核反应堆电源将变得愈发重要。本文针对国内外空间核反应堆电源的热工水力关键问题,即空间堆系统稳态和事故瞬态研究、堆芯单冷却剂通道及全堆芯的三维流动换热、静态与动态热电转换装置分析、热工水力特性试验研究等进行研究,分析了空间核反应堆电源热工水力研究的趋势。本文结果可为空间核反应堆电源设计分析及热工水力安全特性研究提供帮助和指导。     

水下航行器微型核电源堆芯设计

热管冷却反应堆因其体积小、功率密度高、使用寿命长、环境适应性强的特点，在飞行器、水下航行器动力系统等领域具有广泛的应用前景，具有重要研究意义。本文在调研国内外相关研究的基础上，针对水下航行器静默式微型核电源，提出了一种热功率2.4 MW、长寿命、低噪声的锂热管冷却堆芯设计方案。采用蒙特卡罗程序进行中子学计算，得到堆芯功率分布、反应性反馈与临界安全特性；采用MCNP5与点燃耗程序ORIGEN2的耦合程序MCORE计算堆芯寿期。结果表明，堆芯最大功率峰因子为1.42，堆芯寿期达到14 a，堆芯参数符合设计要求，为该型核电源的设计与应用提供了一定的参考。     

空间核热推进粒子球床堆慢化剂温度效应分析

空间核热推进(SNTP)粒子球床堆(PBR)呈现正的慢化剂温度效应,影响反应堆运行安全。基于PBR堆芯物理模型,采用蒙特卡洛中子-光子输运程序(MCNP)对PBR堆芯慢化剂温度系数进行计算。从中子平衡的角度分析慢化剂升温前后堆芯内中子能谱、中子产生率、中子吸收率和中子泄漏率的变化。结果表明:7Li H升温后,堆芯总的中子消失率(吸收率和泄漏率)的增量要比中子产生率的增量少得多,使得PBR堆芯表现出正的慢化剂温度效应,且低温时正温度系数值较大。     

堆芯应急冷却热块再淹没过程实验研究

实验模拟核反应堆堆芯失水后堆芯熔融物和被加热压力容器壁等热块再淹没时的应急冷却过程。实验研究发现,液滴飞溅对热块钢板起到了预冷作用,在淹没液位上升的过程中,热块纵向导热越来越强,被淹没位置具有很高的中心冷却速率;热块被淹没位置的中心冷却速率并不随浸没速率单调变化,而是在一定区间内呈起伏变化,这说明在某个淹没速率下存在一个最小中心冷却速率的区间,因此在进行应急冷却时要避免这个区间;在高温情况下,冷却的初始温度对中心冷却速率影响不大。     

物理-热工耦合计算方法在熔盐堆稳态分析中的应用

针对美国橡树岭国家实验室（ORNL）熔盐堆（MSR）实验的堆芯设计,采用物理分析程序MCNP进行三维堆芯功率分布计算。针对以石墨作为慢化剂的堆芯结构,开发了并联多通道程序来进行堆芯热工水力分析。在此基础上,把物理和热工分析程序进行耦合,用ORNL技术报告中的相关内容来验证物理 热工耦合分析的可行性和准确性。结果表明,本工作的耦合计算方法可获得熔盐堆堆芯功率分布、温度分布、压降和流量分配。熔盐堆耦合程序的研发对熔盐堆概念设计、运行分析有重要意义。     

MOX燃料组件装入现役M310堆芯对堆芯核设计的影响研究

国际上的MOX燃料技术目前已较为成熟,且已有在压水堆中运行的工程经验。本文对MOX燃料组件的中子学性能进行了分析,对其在我国现役M310堆芯应用的可行性进行了研究,得到了M310堆芯由全部使用UO2燃料组件向使用30%的MOX燃料组件过渡的堆芯燃料管理方案,并对使用MOX燃料组件的堆芯的部分中子学参数进行了初步分析。结果表明:使用30%的MOX燃料组件的堆芯可达到与全UO2堆芯相当的循环长度;堆芯反应性控制能力可满足要求;慢化剂温度系数、Doppler温度系数、Doppler功率系数、氙和钐的动态特性均趋向使堆芯运行更加安全和稳定。本文的研究结果可为MOX燃料在M310堆芯中应用的进一步研究提供参考。     

微型核反应堆技术发展趋势研究

微型核反应堆采用四代非轻水堆、热管堆以及三代轻水堆等固有安全性高的堆型,可以为偏远海岛和矿区、边防哨所和基地、应急救灾、太空和深海探索等创新场景提供长续航高可靠能源,具有广阔的应用前景,是实现国家战略的重要技术支撑之一。本研究总结了微型核反应堆的定义和主要研发堆型,描述了微型核反应堆固有安全性高、易于模块化和扩展、可运输性、便于部署、自主运行等创新技术特征,分析了微型核反应堆新型燃料、主回路一体化、新型热电转换装置、非能动安全系统、智能运维以及核能与其他能源耦合等关键技术的发展趋势,可为制定适用于我国的微型核反应堆发展技术路线提供支撑。      

重水冷却钍基长寿命模块化小堆概念设计研究

基于传统压水堆（PWR）技术,提出一种重水冷却的钍基长寿命模块化小堆（RMSMR）的概念设计方案,采用二维模型系统分析并对比了PWR和RMSMR的燃料类型、慢化剂类型等参数,获得反应堆各项中子学参数的变化机理;然后基于二维计算结果提出了最终的三维堆芯设计方案,并开展了初步的中子物理和热工安全分析。研究表明,RMSMR在设计上采用三区燃料布置来展平功率,采用钍-铀燃料维持了负空泡系数,通过布置增殖包层提高了堆芯的转换比（CR）;RMSMR采用了重水冷却剂可以使中子能谱硬化,从而提高CR,减小寿期反应性波动,增加堆芯寿期;RMSMR能够在100 MW电功率下维持6 a的安全运行。本文研究可为新型反应堆的设计发展提供借鉴。      

A dynamic simulation of a nuclear power station for control analysis

Nonlinear simulation has been applied to the control study of an organic-cooled Canadian CANDU reactor. The study was carried out as an integral part of an overall design study. The control characteristics of the power plant were studied for reactor start-up, pump failures, reactor trips and various rates of power change. The response of the power plant to these disturbances is discussed.     

Investigation of burnable poisons effects in reactor core design

Burnable absorber rods (BAR) and chemical shim are the main control poisons that are used in the core for improving the reactor behavior and satisfying the safety criteria during the core life time. These poisons have several constraints, criteria, advantages and also disadvantages from the safety and operation points of view; and these characteristics depend on the concentration and distribution of mentioned poisons in the reactor core. Therefore, understanding their effects on the reactor core behavior, especially the mutual interaction between them, is a crucial issue in reactor core design procedure. In this study, the influences of the burnable poisons on the main parameters of the reactor such as multiplication factor, burnup, soluble poison concentration, moderator temperature coefficient and power peaking factor over the reactor life time are investigated. The VVER-1000 reactor was selected for this investigation.     

混合堆增殖乏燃料组件中子学特性初步研究

本文主要对聚变-裂变混合堆增殖乏燃料在压水堆组件中使用的可能性进行了初步研究。根据聚变 裂变混合堆增殖乏燃料的特点,给出了的聚变-裂变混合堆增殖乏燃料压水堆组件设计方案,分析组件的燃料温度系数、慢化剂温度系数等参数。结果表明：聚变 裂变混合堆乏燃料组件的特性与全铀组件的特性相似。在相同的易裂变同位素质量百分比情况下,本文给出的组件设计方案的功率不均匀系数更小。研究结果可为未来实现聚变 裂变混合堆和压水堆联合循环系统提供技术支持。     

Thermo-hydraulic analysis for SCWR during power-raising phase of startup

The study of thermal characteristics during startup is one of the most important aspects for safety analysis of supercritical water-cooled reactor(SCWR).According to the given sliding pressure mode of SCWR,thermal analysis on temperature-raising phase and power-raising phase of startup are carried out.Considering the radial heterogeneity of power distribution,thermal characteristics for different assemblies during startup are also put forward.The results show that,during temperature-raising phase with core power increased only,the temperature of moderator,coolant and fuel cladding in inner assemblies are increased with little amplitude.During power-raising phase with core power and feed-water flow rate increased,the coolant temperature keeps unchanged,but the moderator temperature is decreased.With a greater variation of power,fuel cladding temperature shows a greater increase.Furthermore,considering the uneven distribution of radial power,thermo-hydraulic characteristics with uneven cladding temperature distribution shows a certain horizontal heterogeneity for different fuel assemblies,which becomes serious as flow rate and power increase.By adjusting flow rate distribution in different fuel assemblies or changing power setting during startup,the cladding temperature difference could be effectively reduced,which provides a certain reference for startup optimization of SCWR.     

CANDU堆发展的回溯与展望

ＣＡＮＤＵ堆是世界上达到充分成熟且成功发展的少数几种堆型之一。这种堆的设计概念的基本出发 是使用天然铀燃料，这一选择决定了其它几个有利的选择，例如采用重水慢化剂，不停堆换料以及计划机控制。采用重水慢化剂和增大输出功率的需求决定了压力管式堆芯结构的设计方案。     

Fuel assembly design study for a reactor with supercritical water

The European concept of the High Performance Light Water Reactor (HPLWR) differs from current light water reactors in a higher system pressure beyond the critical point of water, as well as a higher heat-up of the coolant within the core and thus higher core outlet temperatures, leading to a significant increase in turbine power and thermal efficiency of the power plant. The motivation to develop a novel fuel assembly for the HPLWR is caused by the high variation of coolant density in the core by more than a factor of seven. A systematic design study shows that a square fuel assembly with two rows of fuel rods and a central moderator box is best to minimize the structural material, to optimize the moderator to fuel ratio and to reduce differences of fuel rod power. Using neutronic and thermal-hydraulic analyses, a detailed mechanical design of a fuel assembly of the HPLWR has been worked out. Moreover, concepts for the head piece, the foot piece, the steam plenum and the lower mixing plenum, including the lower core plate, have been developed to account for the individual flow paths of this reactor. These allow a leak-tight counter current flow of moderator water and coolant as well as uniform mixing of different mass flows. The assembly design concept can be used as a general key component for any advanced core design of this reactor.     

Experimental study of neutronic parameters in Tehran research reactor mixed-core

In this work, general characteristics of a typical mixed core, including HEU & LEU fuel is studied. The study is performed in the Tehran research reactor (TRR). In this study the neutronic parameters, reactivity feedback coefficients and kinetic parameters are investigated. The reference core designated for such study is the equilibrium core (No. 61) with an average bun-up of 27% & 36% for SFE's & CFE's, respectively. The MTR_PC package is used for neutronic analysis. In this research, experimental and computational results for the reference and mixed core are compared. Meantime, the obtained values for neutronic parameters are mostly below the adopted safety criteria and they are in good agreement with the experimental results. However β_eff and ℓ_p are a little bit higher in the mixed core with respect to the reference core, but in practice, these small changes will not cause substantial impacts on the dynamic behaviour of the reactor core. The absolute values of the fuel temperature, moderator density and void coefficients of reactivity, are less in the mixed core and only the moderator temperature coefficient is higher. The calculated values of power defect, based on the reactivity coefficients; in both core configurations are in good agreement with the experimental values.     

热管反应堆燃料经济性影响因素初步探索

热管反应堆通过高温热管从堆芯直接导出热量，系统设计本身就极为简化，较为适宜作为小型核电源的技术选型。燃料经济性是反应堆技术路线选型的重要依据，为详细研究热管反应堆设计对其燃料循环经济性影响，本文初步建立热管反应堆燃料经济性影响因素分析模型，以eVinci反应堆为例，开展了燃料循环经济性影响因素探索研究，获得了总体方案功率规模、堆芯运行温度等因素对热管堆燃料经济性的影响变化趋势。结果表明受燃料价格、铀装量、富集度等综合影响，热管反应堆燃料经济性相对较好的优选热功率规模区间在约1~5 MW。提高堆芯运行温度可使燃料经济性大幅提升，经济性最佳功率区间向高功率规模扩展。      

Studies on Influence of Sodium Void Reactivity Effect on the Concept of the Core and Safety of Advanced Fast Reactor

The paper is devoted to studies on the influence of the sodium void reactivity effect (SVRE) on the safety and technical and economical characteristics of the BN-1200-type reactor. Different core options are considered for application to this reactor. These core options differ in design, dimensions, and, hence, SVRE value. It is shown by the analysis that the most flattened core with sodium plenum at the top assures reactor self-protection under beyond-design-basis accident conditions. Sodium plenum abandonment and core height increase causing an SVRE value increase deteriorate reactor self-protection, but at the same time, improve some technical and economical characteristics of the reactor. Self-protection means the possibility to avoid rapid core meltdown under conditions of the above-listed beyond-design accidents. The possibility of controlling beyond-design accidents (for instance, by restoring the power supply of the main pumps in a rather short time) is taken into account. Issues of choosing the optimal core design under these conditions are discussed.     

大型先进压水堆核电站堆芯装载方案设想

核电站堆芯装载方案是反应堆堆芯设计的重要基础,它首先必须满足核安全的要求,同时还要尽可能地提高经济性。通过分析国内、外百万千瓦级核电站的堆芯装载,对反应堆输出功率、燃料组件数、堆芯平均线功率密度进行比较,给出我国大型先进压水堆核电站示范工程反应堆堆芯装载方案的设想,为技术决策提供参考。