升功率工况下波动管热分层数值分析

以CPR1000稳压器波动管为研究对象,采用CFD方法,使用FLUENT软件,对反应堆功率增加瞬态工况下波动管热分层现象进行数值模拟研究,得到了波动管内热分层流体的流场和温度场分布,探讨了涡流效应对热分层分布的影响。结果表明：瞬态工况下波动管热分层与传统观念下的稳态热分层相比有很大不同,最显著的是T型三通区域,由于受到涡流效应的影响,流体热分层呈环形左右分布,而不再是稳态热分层的上下分布。本研究得到的瞬态工况下的温度分布结果可作为瞬态热应力分析的温度载荷,为后续的力学分析和疲劳分析奠定了基础。     

AP1000波动管涌动瞬态抑制的运行策略综述

抑制涌动瞬态能够显著地减少稳压器波动管的热应力水平,对保持波动管在电站寿期的完整性和提高其可靠性有重要的意义,除设计改进外,有效的运行策略对抑制涌动瞬态具有非常大的作用。本文分析影响AP1000波动管涌动瞬态的运行工况和主要参数,建立连续的波动管涌出工况和减少系统温差最大值的总体策略,综述机组正常运行及启停工况下的有关操作。     

瞬态工况下压水堆稳压器波动管热分层现象数值模拟

利用计算流体动力学软件ANSYS/CFX,对秦山核电二期扩建工程2×650 MW压水堆核电站四号机组核岛厂房的稳压器波动管进行了三维全尺寸非稳态计算。建立了波动管整体和不同截面的热分层瞬态,对管内热分层流动与换热进行了研究。研究结果表明:同一截面内高温层流体和低温层流体的升温方式不同;不同截面位置的管内流动温度分布特性差别较大,但均呈现分层流体温差先增大后减小的趋势。计算结果可为后续波动管热应力分析及寿命评价提供一定基础。     

压水堆核电厂稳压器波动管热分层现象数值分析

为分析评价压水堆核电厂稳压器波动管热分层现象对波动管结构完整性的影响,采用计算流体力学(CFD)分析方法,对稳压器波动管热分层现象进行了数值模拟.研究了波动管内的流体流动,得到了稳压器波动管的传热特性、流体流场和温度分布,分析了稳压器波动管波动热分层现象与波动流速之间的关系.研究结果表明:波动流速在一定范围内变化时,管道最大截面温差随着波动流速的增大而增大.并且得到了不同波动流速下管道最大截面温差及其出现的位置,指出了热分层现象发生时波动管的薄弱环节.     

稳压器波动管考虑热分层影响的疲劳分析

在核电厂中,稳压器波动管及波动管热段三通是保证核电厂反应堆冷却剂压力边界完整性的重要设备.其属于核安全1级设备,承受内压、自重、热胀、地震及各种正常加异常工况下的温度和压力瞬态,特别对于压水堆核电厂的波动管,还会承受热分层导致的总体和局部载荷.热分层现象的反复出现增加了管道及接管嘴处出现疲劳失效(贯穿管壁裂纹)的可能性.本文阐述了对波动管热分层实施温度测量的方案,及对测量结果的分析处理;建立分析热分层整体应力和局部应力,以及波动管疲劳分析的计算模型;确立合理且切实可行的波动管疲劳分析所需的分析瞬态.上述方法已在"300 MWe PWR NPP稳压器波动管热分层"课题研究得到鉴定,并在实际的寿命管理等工程项目中发挥了重要作用.     

压水堆稳压器波动管热分层的分析研究

热分层是管道水平管段中相对滞止或缓慢流动的冷、热流体因缺少混合而产生的不均匀温度分布现象.通过稳压器波动管热分层现象产生的原因和机理分析,并对稳压器波动管热分层现象进行数值模拟,建立了不同稳压器内部不同截面的热分层瞬态.     

反应堆Ⅱ类瞬态下燃料棒PCI研究

介绍反应堆Ⅱ类瞬态下燃料棒芯块与包壳相互作用(PCI)分析方法和PCI热-力学计算理论模型,在此基础上对海南核电厂降功率燃料管理方案进行PCI评价,并对影响PCI失效裕量的因素进行分析。结果表明,所有瞬态条件下包壳的应变能密度与技术限值相比较都有裕量;瞬态局部线功率越大,瞬态发生前局部燃耗越深,PCI失效裕量越小;瞬态发生前,降功率时间越长,PCI失效裕量越小;降功率后再升功率,裕量得到一定程度恢复。     

管型对波动管热分层现象的影响分析

为分析评价压水堆核电厂稳压器波动管管型对热分层现象的影响,提出采用螺纹管来减弱热分层的措施。利用计算流体力学（CFD）分析方法,对升温、升压阶段波动管原型和改进模型的热分层现象进行数值模拟,得到两种模型不同波动流速下沿波动管轴线方向的截面最大温差分布以及流场分布。对比分析结果表明：波动管结构由光管改为螺纹管后流场紊动加强并出现涡流,冷热流体间的混合增强,与原型相比可使波动管的截面温差减小约1/3,从而有效地减弱热分层的影响。     

简化反应堆一回路系统自然循环的动态比例特性研究

在反应堆安全领域,合适的比例分析对非能动系统实验台架的设计起到了关键作用。为深入了解比例缩放时自然循环瞬态过程的变化机理,基于简化反应堆一回路系统,分别采用H2TS（双向分层比例分析）和DSS（动态比例分析）方法进行了自然循环的比例分析,计算了升降功率工况下的自然循环,对比分析了不同尺度下关键参数的动态变化。结果表明,基于RELAP5的计算结果与实验结果基本一致,5%初始功率以下的阶跃变化不会造成大的流量波动;基于两种比例分析方法所得缩比模型下的计算结果均可基本反映原型参数变化;所有工况下,自然循环流量和冷热段温差在初始阶段05个循环周期内存在较大的波动,之后则相对平稳。     

针对热分层现象的稳压器波动管结构评价方法研究及分析程序应用开发

稳压器波动管热分层现象对核电厂安全运行具有潜在威胁。根据热分层发生机理,采用Fr数来判断热分层现象是否发生,研究热交换系数的计算方法,并将热分层引起的三维热应力解耦成一维总体应力和二维局部应力。根据RCC-M规范的要求,采用一维和二维组合的应力计算方法,将热分层产生的应力和其他载荷产生的应力叠加,进行结构完整性评定。配套本文提出的分析评价方法,对SYSTUS程序和ROCOCO程序进行应用开发。采用本文的分析评价方法和配套的分析程序,对秦山二期扩建工程稳压器波动管热分层进行分析评价,结果表明:稳压器波动管在热分层效应下结构完整性仍然满足RCC-M规范要求。     

Stress analysis of a 900 MW pressurizer surge line including stratification effects

Following temperature monitoring programmes performed on 900 MW pressurized water reactor pressurizer surge lines, it has been reported that those lines are stratified in steady state, owing to their geometry. The highest temperature difference occurs during reactor heat-up and cool-down, reaching 110°C. Obviously, this phenomenon was not considered in nuclear steam supply system (NSSS) design transients and stress reports.Based on Electricité de France and FRAMATOME experiences, such as temperature measurements on site and mock-up, and thermal hydraulic computations, NSSS transients are updated. Stratification conditions are defined in different cross-sections of the line, using pressurizer temperature, hot leg temperature and flow rate, through the Froude number. A complete stress analysis of surge lines is performed including the updated transients and bending moment increase due to stratification. First of all different sensibility studies are carried out in order to simplify assumptions.Using a two-dimensional-one-dimensional method developed by FRAMATOME, the usage factor is then computed in different cross-sections, distinguishing upper and lower parts. In the presence of stratification, the surge line is subjected to thermal stresses following thermal shocks and to bending moment variation. These two load types are studied vs. time in order to reduce conservatism present in usual analyses.     

Fluid-structure interaction analysis for pressurizer surge line subjected to thermal stratification

Serious mechanical damages such as cracks and plastic deformations due to excessive thermal stress caused by thermal stratification have been experienced in several nuclear power plants. In particular, the thermal stratification in the pressurizer surge line has been addressed as one of the significant safety and technical issues. In this study, a detailed unsteady computational fluid dynamics (CFD) analysis involving conjugate heat transfer analysis is performed to obtain the transient temperature distributions in the wall of the pressurizer surge line subjected to stratified internal flows either during out-surge or in-surge operation. The thermal loads from CFD calculations are transferred to the structural analysis code which is employed for the thermal stress analysis to investigate the response characteristics, and the fatigue analysis is ultimately performed. In addition, the thermal stress and fatigue analysis results obtained by applying the realistic temperature distributions from CFD calculations are compared with those by assuming the simplified temperature distributions to identify some requirements for a realistic and conservative thermal stress analysis from a safety point of view.     

稳压器波动管热分层分析

为评价热分层对稳压器波动管结构完整性的影响,从理论上分析了稳压器波动管热分层发生的条件.以百万千瓦级三环路压水堆核电厂核反应堆启堆为例,建立了热分层瞬态,研究了热分层应力计算方法,从理论上将一个复杂的三维应力分析问题简化为一维和二维组合问题.结合ANSYS程序功能,提出了波动管热分层应力计算的工程方法.     

Structural evaluation of thermal stratification for PWR surge line

Recent observations at operating plants and subsequent US NRC requirements have identified flow stratification in surge lines as a phenomenon that must be considered in the design basis of surge lines. To address these concerns, the stratified loading conditions were included in the design of YGN 3 and 4 surge line as a design basis transient and pipe temperature and displacement measurement were taken during YGN 3 pre-core hot functional testing to determine the degree of surge line flow stratification. The measured displacements and temperatures were extensively reviewed and evaluated in detail: (1) to verify the validity of the thermal hydraulic model used to predict the pipe top-to-bottom temperature differentials; (2) to analytically correlate measured surge line temperatures and displacements; and (3) to confirm the validity of the stratified flow analysis procedure. This paper shows that the stratified flow phenomenon is generic and therefore generic loadings can be developed and evaluated for the surge line analyses.     

Structural evaluation of a piping system subjected to thermal stratification

Piping systems in nuclear power plants are often designed for pressure, mechanical loads originating from deadweight and seismic events and operating thermal transients using the equations in the ASME Boiler and Pressure Vessel Code, Section III. In the last few decades a number of failures in piping have occurred due to thermal stratification caused by the mixing of hot and cold fluids under certain low flow conditions. Such stratified temperature fluid profiles give rise to circumferential metal temperature gradients through the pipe leading to high stresses causing fatigue damage. A simplified method has been developed in this work to estimate the stresses caused by the circumferential temperature distribution from thermal stratification. It has been proposed that the equation for the peak stress in the ASME Section III piping code include an additional term for thermal stratification.