TACR压力管与排管间辐射换热分析

辐射换热是钍基先进CANDU型反应堆(TACR)压力管和排管间换热的主要途径.本文以灰体辐射模型和电网络分析方法为基础建立了TACR压力管和排管间辐射换热的计算模型,利用该模型计算了给定温度边界和热流密度边界的情况下,压力管和排管间的辐射换热能力.计算结果表明,该模型可以用于TACR压力管和排管间辐射换热能力的计算.     

TACR排管容器研究中多孔介质法的应用

以钍基先进重水堆(TACR:Thorium Based Advanced CANDU Reactor) 慢化剂系统排管容器作为研究对象,提出了一种适用于计算排管容器内流场和温度场的非结构网格多孔介质方法,并将其计算结果与精细数值计算法所获得的结果进行了比较,二者符合较好,由此验证了其方法与计算结果的正确性.     

TACR热传输系统热工水力瞬态分析

钍燃料的利用对于缓解核燃料资源短缺具有重要意义,坎杜型反应堆(Canadian Deuterium Uranium,CANDU)在堆芯布置、中子利用效率及先进燃料循环方面具有较高的灵活性,使得其在CANDU反应堆中引入钍燃料循环更具现实意义。CANDU型反应堆中钍基燃料应用关键基础技术研究是加拿大与我国正在开展的合作课题,其中开发自主的CANDU堆堆芯热工水力设计和安全分析程序是钍基燃料应用必不可少的设计工作之一。本文针对CANDU型反应堆热传输系统结构特点,采用FORTRAN程序设计语言开发了适用于CANDU型反应堆热传输系统的热工水力瞬态分析程序CANTHAC(CANDU Thermal-Hydraulic Analysis Code)。利用CANTHAC对钍基先进CANDU堆(Thorium-based Advanced CANDU Reactor,TACR)进行了瞬态分析,计算工况包括满功率稳态、无保护蒸汽发生器(Steam Generator,SG)二次侧给水温度降低事故及完全失流事故。其中,满功率稳态计算结果与清华大学设计的钍基先进CANDU堆TACR设计值吻合较好,相对误差不超过2%,在可接受范围内;无保护SG二次侧给水温度降低事故及完全失流事故在计算条件下所得的燃料温度及系统压力等关键热工水力参数均在安全限值内,满足安全准则要求。程序为模块化编程,便于移植和改进,具有一定的通用性,为进一步研究工作奠定了基础。     

钍基先进核能系统压力管与排管之间的非能动热开关设计

提出了一种用于钍基先进核能系统压力管与排管之阃的非能动热开关设计方案,该热开关基于形状记忆合金原理,当压力管的温度达到340℃的时候．热开关中的形状记忆台金部分开始变形,推动导热部件移动．使得压力管和排管之间通过导热部件接触。从而将堆芯内的热量从压力管传到排管和慢化剂,再通过慢化荆冷却系统把热量带出并最终释放到周围环境之中．理论计算表明．24片的热开关组件可以满足设计要求。     

CANDU-6核电机组压力管寿命管理

介绍了CANDU反应堆压力管的使用条件、压力管寿命与电站寿命的关系、AECL多年来对压力管的改进工作 ,论述了影响压力管使用寿命的因素和秦山三期压力管寿命管理的思路和主要措施     

CANDU及RBMK压力管锆合金的氢致延迟断裂研究

采用紧凑拉伸试样(CT），在恒定载荷、不同氢含量、不同温度条件下，测量了CANDU堆和RBMK堆Zr-2．5Nb压力管材料氢致延迟开裂速率。用金相显微镜和扫描电镜观察断口及氢化物形貌，并测量临界应力场强度因子及开裂速率，对材料的微结构及氢化物分布进行分析。结果表明，氢致延迟断裂(DHC)生长呈阶梯状。与CANDU压力管比较，RBMK压力管的DHC开裂速率将近低一个数量级。其原因是RBMK压力管的屈服强度比CANDU压力管低得多。     

加拿大CANDU型重水堆核电厂的现状

从50年代起,加拿大原子能有限公司（AECL）就开始研究重水慢化天然铀反应堆。一个由科学家、工程师、电业界代表、制造商和私人顾问组成的高级专家小组,就加拿大核电厂的选型问题进行了专门研究,最后得出的结论是重水慢化反应堆是加拿大核电厂采用堆型的最佳选择。其理由一方面是考虑到重水堆的中子经济性好,可有效地利用铀资源,本国在重水研究堆上已积累了很多经验;另一方面是考虑到加拿大自身的国力。当时它既不能建造铀同位素浓缩厂（需要巨额投资及复杂的技术）,也无力制造压水堆所必需的大型压力容器和其他重要设备。因此,它选择了一条投资少、见效快、容易自立更生发展核能的道路,即发展用重水作慢化剂、天然铀作燃料的压力管式反应堆,简称为坎杜型（CANDU）重水堆核电站。这种电站,只要国内铀资源较为丰富,在具有相对适中工业制造基础的国家就可独立建造。 CANDU堆是卧式结构,在装有低温低压重水慢化剂的卧式圆柱形排管式堆容器内,有数百根压力管贯穿于排管式堆容器内管,压力管内装有燃料元件棒束,高温高压重水冷却剂流经压力管,将燃料裂变所产生的热量带到蒸     

对CANDU燃料组件结构的一维瞬态辐射程序开发与验证

在气冷CANDU式燃料组件之中,辐射换热也是不容忽视的一部分。特别是在出现了系统失压/失流事故时,辐射换热将会成为保证燃料安全的主要冷却手段。本文中针对CANDU式压力管编制了针对压力管几何条件下的一维辐射换热瞬态程序。程序中采用将燃料元件棒转化为同心圆环的方式简化辐射角的计算,并加入了隔层辐射模型,使模型更加贴近实际。采用分别将程序中的几个模块的计算结果与CFX计算结果对比的方式来达到程序验证的目的,验证结果显示程序RHTPB具有良好的表现,能够满足于反应堆安全计算的需要。     

红沿河核电项目除氧器超压问题分析及改进措施

红沿河核电站现有除氧器在主蒸汽旁路工况下会发生超压问题,为了定量把握超压问题的实质,对现有除氧器在汽轮机旁路工况下的压力全变化过程进行了详细分析。分析表明,现有除氧器将在切入旁路工况5.2 s后内部压力超过设计压力,引起安全阀起跳并向大气排放蒸汽。为了提高系统的安全性,文章提出了将正常抽汽管道和旁路蒸汽管道分离设置,并在现有除氧器内增设鼓泡管系统的改进设计方案,利用新建立的数值分析模型对改进方案进行了理论验算。计算结果表明,改进后的方案可保证在主蒸汽旁路工况50 s内除氧器内部压力不会超过设计压力。     

ATWS事故工况下的应急初始条件和应急行动水平在核电厂的应用

本文描述了在未能紧急停堆的预期瞬变(ATWS)事故工况下应急初始条件及应急行动水平在PWR核电厂和CANDU核电厂的应用,并对这两种类型核电厂在ATWS事故工况下相同应急初始条件的应急行动水平的不同进行了比较.     

Contact Conductance between Cladding/Pressure Tube and Pressure Tube/Calandria Tube of Advanced Thermal Reactor (ATR)

In some postulated accidents with coincident loss of emergency coolant injection of the Advanced Thermal Reactor (ATR), the rate of heat transfer to the heavy water moderator that acts as heat sink for the decay heat depends on the contact conductance between cladding and pressure tube, and the same between pressure and calandria tubes. Experiments were performed to assess these contact conductances for clean plates and plates with simulated crud of Fe₂O₃ powder that is the main ingredient of the crud, and the applicable correlations were also studied. Test specimens were cut from actual pressure tube made of Zr-2.5%Nb and calandria tube made of Zircaloy-2 and flattened to sizes. The artificial waviness of various kinds of height and wave length of 10 mm was machined on the surface of the pressure tube specimen. The ranges of contact pressure, roughness, specimen temperature and gas pressure were from 0.5 to 7 MPa, 4.8 to 100 μm, 400 to 840 K and 0.001 Torr to atmospheric respectively. The experimental results without crud compare favorably with Tachibana's correlation. The experimental contact conductances with crud are found to be lowered. Since Tachibana's correlation was not applicable to the present situation, the empirical correlation was proposed. The correlation fitted well with the experimental values.     

Core coolability of an ATR by heavy water moderator in situations beyond design basis accidents

In the case of a loss-of-coolant accident (LOCA) with coincident loss of emergency coolant injection (LOECI), core cooling is generally very severe. However, as the ATR plant has heavy water at about 60°C in the core, decay heat can be removed by the heavy water cooling system. Separate-effects tests relating to heavy water cooling were conducted with each setup. The important thermal hydraulics was radiation heat transfer, ballooning of a pressure tube, contact conductance between the pressure tube and a calandria tube and critical heat flux of the calandria tube. Constants and correlations obtained by the tests were incorporated into several codes to assess the core cooling. Long term core cooling capability with the heavy water cooling system was assessed. The core was cooled without melting under the postulated events due to inherent characteristics of the ATR.     

反应堆直接安注结构热工影响分析

建立了反应堆压力容器直接安注的实验模型,进行传热和流动实验。实验对比研究了文丘里管和直管两种不同安注管结构下压力容器不同区域的流动混合、传热系数以及相应的温度变化,并通过可视化实验对比验证了不同安注模型的流场,对不同结构下安注管的阻力特性进行了分析。研究结果表明,不同安注管在流动阻力上存在较大差异。研究结果为反应堆压力容器安注管的设计选择提供了重要依据。     

针翅管传热与压降特性研究

以润滑油为换热介质,对整体针翅管传热与阻力特性进行了理论分析与试验研究,研究结果可为针翅管的优化设计提供参考。在换热介质纵向冲刷换热管的条件下,对不同针翅长度的3种整体针翅管与光管进行了传热与阻力试验。结果表明：整体针翅管对润滑油换热具有很好的强化能力,在本试验范围内,整体针翅管对油流体扰动强烈,换热强度是同条件下光管的2~6倍;针翅长度是影响针翅管压降的主要因素,在雷诺数达300时,压降曲线出现转折。     

同心环形管内沸腾两相流动与传热实验研究

对窄缝为2．1mm的同心环形管,试验研究了外管加热条件下水的沸腾两相流动阻力与传热特性,得到了以下结果：窄缝环形管内两相流动的阻力较普通圆管内大,沸腾换热得到了较明显的强化,换热系数弓压力、热平衡干度、工质流量、加热负荷均有关系,且与缝隙宽度和加热方式有关;提出了环形管强化传热的微液膜蒸发机理与汽泡扰动机理的物理解释;得到了环形管内流动摩擦阻力系数与传热系数的实验关联式。     

Severe accident development modeling and evaluation for CANDU

Romania as UE member got new challenges for its nuclear industry. Romania operates since 1996 a CANDU nuclear power reactor and since 2007 the second CANDU unit. In EU are operated mainly PWR reactors, so, ours have to meet UE standards. Safety analysis guidelines require to model nuclear reactors severe accidents.Starting from previous studies, a CANDU degraded core thermal hydraulic model was developed. The initiating event is a LOCA, with simultaneous loss of moderator cooling and the loss of emergency core cooling system (ECCS). This type of accident is likely to modify the reactor geometry and will lead to a severe accident development. When the coolant temperature inside a pressure tube reaches 1000 °C, a contact between pressure tube and calandria tube occurs and the decay heat is transferred to the moderator. Due to the lack of cooling, the moderator, eventually, begins to boil and is expelled, through the calandria vessel relief ducts, into the containment. Therefore the calandria tubes (fuel channels) uncover, then disintegrate and fall down to the calandria vessel bottom. All the quantity of calandria moderator is vaporized and expelled, the debris will heat up and eventually boil. The heat accumulated in the molten debris will be transferred through the calandria vessel wall to the shield tank water, which surrounds the calandria vessel. The thermal hydraulics phenomena described above are modeled, analyzed and compared with the existing data.     

快堆钠-空气热交换器翅片管传热及阻力特性试验研究

钠-空气热交换器是钠冷快堆事故余热排出系统的重要设备之一,与外界环境一起构成事故下反应堆余热的最终热阱。由于钠-空气热交换器的换热管采用垂直布置的翅片管结构,空气在不同位置处冲刷换热管的流速以及角度不同,导致其传热特性及阻力特性与传统翅片管换热器有很大不同。本文以钠-空气热交换器工程设计需求为研究背景,设计了两种试验件分别进行空气冲刷角度为90°和30°时翅片管束传热与流动阻力特性试验研究。试验结果表明：对于相同管排,空气冲刷角度为90°时的翅片管的换热系数及阻力系数明显大于空气冲刷角度为30°时的翅片管;对于相同空气冲刷角度下的不同换热管排,第2排翅片管的换热系数最大。本文研究为钠-空气热交换器的设计及优化提供了理论依据。     

Experimental Study on Heat Transfer and Resistance Characteristics for Finned Tube of Fast Reactor Sodium-air Heat Exchanger

Sodium-air heat exchanger is one of the important equipment in the decay heat removal system of sodium-cooled fast reactor, which together with the external environment constitutes the final heat trap for the residual heat of a reactor. Due to the vertically arranged finned tube structure adopted by the heat exchanger tubes of the sodium-air heat exchanger and the difference of the flow direction and velocity of the air at the different positions, the heat transfer and resistance characteristics are very different from of traditional finned tube heat exchanger. In this paper, based on the engineering design requirements of sodium-air heat exchanger, two kinds of experimental specimens were designed to study the heat transfer and flow resistance characteristics of finned tube bundles with air flow angles of 90° and 30°. The experimental results show that the heat transfer and resistance coefficients of the same finned tube with air flow angles of 90° are significantly greater than those of the finned tube with air flow angles of 30°. For the same air flow direction, the second finned tube has the largest heat transfer coefficient. The research provides a theoretical basis for the design and optimization of sodium-air heat exchanger.     

不同管径与倾角下单管和管束外含空气蒸汽冷凝对比分析

为评估不同传热管结构参数下单管与管束外含空气蒸汽冷凝传热规律的差异,基于外径12～19 mm、倾角0°～90°的单管和3×3管束在压力0.2～1.6 MPa、空气质量份额12%～87%的参数范围内开展了试验研究。结果表明：不同压力范围内,管径和倾角对单管和管束的影响呈现不同的规律。在压力小于0.8MPa时,管束冷凝传热受管径和倾角影响的规律与单管总体一致,两者的冷凝传热系数均随管径和倾角的减小而增大。在0.8～1.6 MPa时,管束冷凝传热受管径和倾角的影响与单管存在明显差异。结合不凝性气体影响蒸汽冷凝传热的机制对所呈现的一致性和差异性规律进行了分析。