Since the unstable start-ups and shutdown processes of nuclear power plants, the effect of strain rate on tensile and creep-fatigue (CF) properties has attracted attention. This paper aims to study the strain rate dependent tensile and CF behavior of 316H stainless steel. The tensile and CF tests were independently performed with different strain rates of 0.0001, 0.0002, 0.0005, 0.001 and 0.005 s-1 at 600 degrees C. Due to the dynamic strain aging (DSA), the tensile stress-strain curve appears serrated flow phenomenon and shows negative strain rate sensitivity (SRS), which means the tensile strength of the material does not increase significantly with an increase of the strain rate, and even decreases. Meanwhile, it was found that peak tensile stress and inelastic strain increase in CF tests under low strain rates. The CF life of 316H steel was predicted using the hysteresis energy method. At the same time, based on microstructure and fracture observation, different dominant damage mechanisms were discussed.

针对考虑共因失效的多失效模式可靠性问题,基于分析法设计的失效准则和响应面法建立非连续的功能函数,采用点估计法计算功能函数的前四阶中心矩,并根据高阶矩标准化技术在标准正态空间中计算可靠度.通过对控制棒驱动机构(CRDM)承压壳体多失效模式可靠性建模分析,验证了该方法在核电设备结构可靠性分析中的适用性.结果表明,对于基本随机变量偏度较大的情况,点估计法比传统基于摄动理论的一次可靠性方法精度更高,也更适用于功能函数复杂的多失效模式可靠性分析.

非能动余热排出热交换器(PRHR HX)是先进反应堆非能动堆芯余热排出系统的关键设备之一,置于内置换料水箱中.水箱内水蒸发或水箱泄漏等故障失水后液位降低,PRHR HX与管外水传热面积和管外流场均发生变化,导致堆芯非能动余热排出能力下降.为分析半液位下PRHR HX换热性能,建立半液位PRHR HX实验装置和数值计算模型,获得稳态和瞬态升温过程的换热量、温度分布和流场特性.实验结果表明,稳态沸腾换热阶段半液位结构的总换热量较全液位时的下降12％～22％,仍具有较强的换热能力.瞬态升温过程中管外换热模式从纯单相对流换热开始,先后逐渐出现局部过冷沸腾和局部饱和沸腾,并最终迅速由局部饱和沸腾转变为全管长饱和沸腾.水箱内逐渐升温过程中出现显著的热分层现象,结合数值模拟分析发现此时管外流体在高度上呈现多层回流流动结构.

The input ground motion for the seismic design of nuclear power plant (NPP) equipment is required to envelop both the required response spectrum (RRS) and target power spectral density (PSD). However, no unified algorithm exists to generate the target PSD, whether at home or abroad. Based on the generation methodology for target PSD suggested in Section 3.7.1 of the Standard Review Plan (SRP) of U.S. Nuclear Regulatory Commission, this paper proposes an improved method for target PSD synthesis with the original iteration process optimized. The proposed method is applied successfully to the two RRS cases for nuclear equipment. The result shows that the improved method for target PSD generation is highly compatible with RRS and provides simple and rapid calculation, while its convergence accuracy is similar to the calculations by random vibration theory (RVT). This method is considered as the target PSD test basis for the input artificial ground motion (AGM) for the seismic design of NPP equipment.

液压阻尼器是一种对速度反应灵敏的振动控制装置,具有抗冲击和减振性能,被广泛使用在核电厂中来支撑主设备.大型液压阻尼器的正常运行对保证核电厂的安全运行有着重要意义.本研究借助(Advanced modeling en-vironment for performing simulation of engineering systems)AMESim软件建立液压阻尼器的仿真模型并验证其准确性,针对阻尼器的不同工况开展仿真研究,以内泄漏为例对其进行故障模拟,并对阻尼器在正常工况、存在外泄漏工况、存在内泄漏工况、存在阻尼阀堵塞工况四种不同工况下的监测信号进行定性分析,得出其中的逻辑关系.为进一步的液压阻尼器故障诊断系统研究及实际应用提供理论基础和指引.

Both experimental and theoretical works had been done to study the effects of mixture parameters on the performance of corrugated plate (CP) dryer. A water-air experimental branch had been established to simulate water-steam separation in a CP dryer. The effect of ambient relative humidity on the dryer experiment is analyzed. The diameter distribution of water droplets in air-water mixture at (CP) dryer entrance was measured by laser particle size analyzer. The effects of droplet diameter, moisture content and mixture flow velocity on the separation efficiency, pressure drop and the re-entrainment were studied. At the same time, a theoretical model for trajectory tracking of water particle in CP channel was established to illustrate experimental results. The conclusion shows that: 1) this effect of ambient relative humidity cannot be ignored or error could be introduced to experimental results. 2) The separation efficiency increases with the increase of inlet particle size, increases with the increase of inlet moisture content and increases first and then decreases with the increase of inlet velocity. 3) Critical airflow speed of re-entrainment decreases as the inlet moisture content increases. 4) The pressure drop is exponentially increasing with inlet airflow speed.

在反应堆一回路系统设计中,冷却剂流动的阻力是反应堆一回路系统设计的重要依据.因此通过分析或试验的方法预测出准确的一回路冷却剂流动阻力尤为重要.反应堆一回路系统设备主要由反应堆本体和蒸汽发生器等组成,本文针对反应堆本体和蒸汽发生器流阻开展了计算方法研究及试验验证,得到了一套反应堆系统流阻计算方法,并通过与电厂热态调试实测数据的对比验证,证明了此套分析方法的准确性.

为解决我国核电厂蒸汽发生器二次侧三维两相热工水力分析程序缺乏的问题,针对压水堆核电厂蒸汽发生器结构特点,以CFD软件中多孔介质模型为基础,采用UDF对CFD软件进行了二次开发,完成了蒸汽发生器二次侧三维两相热工水力分析程序的开发.基于模化分析采用小规模管束进行流动传热试验,模拟蒸汽发生器一、二次侧流场,获得试验体试验工况下的一、二次侧流场数据.采用开发的程序对试验件建模分析,结果表明:该程序能够实现对蒸汽发生器试验件的模拟,预测结果与试验测量结果符合良好,证明了该程序计算结果的可靠性.

核电厂蒸汽管线的安全阀支管设计不当容易引发流场与声场耦合的声共振现象,对大管径主管上的旁支管结构开展实验研究,研究在10～65 m/s流速区域内,不同的旁支管长度、直径、并排结构对声共振强度、特征频率以及流动特征的影响.实验结果表明,在大管径主管下,二阶水力模态特征显著,其声共振峰值对应的斯特鲁哈尔数(St)约为一阶水力模态声共振峰值对应的St的2倍;旁支管长度的增加显著降低了声共振的强度,旁支管直径的减小则会抑制二阶水力模态的发生,且会改变流动特征并大幅降低声共振强度;并排双旁支管声共振特征与单旁支管相似,但其一阶水力模态一阶声模态控制区的声共振强度与控制流速范围显著增大.

为提高旋叶分离器中气-液混合物的分离效率,针对内筒具有正三角形排布疏水孔的旋叶分离器开展空气-水性能试验和计算流体动力学(CFD)数值模拟研究,分析每一排疏水孔分离水比例变化规律和疏水孔布置高度对分离性能的影响.研究结果表明,当空气表观速度较低(≤14.4 rn/s)时,每一排疏水孔分离水比例随着内筒高度的升高呈先下降后增加的趋势;当空气表观速度较大(＞14.4 m/s)时,每一排疏水孔分离水比例随内筒高度升高呈单调下降的趋势.研究对比了第1排疏水孔布置距离旋叶分别是0.8倍、1.0倍、1.2倍内简直径3种位置时分离性能的变化,结果表明当空气表观速度较低(＜ 12.9 m/s)时,疏水孔位于1.2倍内简直径位置时旋叶分离效率最高;空气表观速度较高(≥12.9 m/s)时,疏水孔位于0.8倍内简直径位置时旋叶分离效率最高.

针对抗振条-传热管大间隙的4跨传热管直管束开展了流致振动试验研究.传热管束转角正三角形排列,3处抗振条将直管束分为4跨,中间其中1跨的局部区域受到横向流体的冲刷.试验测试获得了管间流速在3.3～14.7 m/s区间内传热管振动位移和振动频率响应特性.结果表明,随着管间流速逐渐增大,传热管在来流方向和升力方向的振动频率依次增大,传热管的振动模态从抗振条1处有效支撑 、2处未有效支撑的状态,转换为3处抗振条均有效支撑的状态.试验观测到传热管流弹失稳,其临界流速为14.5 m/s,与5种经验关系式预测结果的对比表明,Chen关系式能较好地预测流弹失稳的发生,预测结果较保守,与试验值间的相对偏差为21.4％.

Passive residual heat removal heat exchanger (PRHR HX), as a key component of the passive residual heat removal system (PRHRS) in the advanced reactors of CAP1400 and AP1000, should have sufficient heat transfer capacity to remove core decay heat during the accidents. In this paper, experiments and numerical simulations are constructed to explore the secondary side flow and heat transfer characteristics of a miniaturized PRHR HX that comprises 6 x 7 C-shaped tubes located in the center of a rectangular tank. In the three-dimensional (3D) numerical simulation, the C-shaped tubes are simplified as porous media and the drift flux model with the distributed resistance method is employed to simulate the two-phase flow in the secondary side of PRHR HX. The heat removal rate and mass transfer rate in the tank are estimated using empirical correlations. In the test, both the primary side and secondary side temperatures are measured. The bubbles generated on the tube outer walls are recorded using a high-speed camera. The local outer wall temperatures, total heat transfer rate and local water temperatures are all compared with experimental values to verify the simulation. The maximum deviation of the calculated heat transfer rate is less than 4.5%. The details about the secondary side flow conditions and bubble behaviors are analyzed. (C) 2020 Elsevier Ltd. All rights reserved.

The structural integrity of welded joints in the reactor pressure vessel (RPV) is directly related to the safety of nuclear power plants. The RPV is made from SA508-III steel in a pressurized water reactor. In this study, we investigated the effects of temperature on the tensile and fracture toughness properties of Chinese SA508-III welded joint in different sampling areas in order to provide reference data for structural integrity assessments of RPVs. The specimens used in tensile and fracture toughness tests were fabricated from the base metal (BM), weld metal (WM), and the heat-affected zone (HAZ) in the welded joint. The representative testing temperatures included the ambient temperature (20 degrees C), upper shelf temperature (100 degrees C), and service temperature (320 degrees C). The results showed that temperature greatly affected the fracture toughness (J(IC)) values for the SA508-III welded joint. The J(IC) values for BM and HAZ both decreased remarkably from 20 degrees C to 320 degrees C. The fracture morphologies showed that the BM and HAZ in the welded joint exhibited fully ductile fracture at 20 degrees C, whereas partial cleavage fracture was mixed in ductile fracture mode at 100 degrees C and 320 degrees C. The WM exhibited the ductile and cleavage fracture mixed mode at various temperatures, and the J(IC) values showed slight changes. (C) 2020 Korean Nuclear Society, Published by Elsevier Korea LLC.

The control rod drive mechanism coil plays an important role in controlling the movement of control rod. This paper investigates the thermal aging mechanism of the coil by analyzing the collected accelerated life data. The Arrhenius-Lognormal Regression Model and Arrhenius-Weibull Regression Model are introduced, and then the parameters and some quantities of interest as well as their confidence intervals are estimated. Empirical study shows that the Arrhenius-Lognormal Regression Model provides a better fit to the data.

在不同液滴粒径工况下,对等比例旋叶分离器的旋叶倾角和上升通道高度进行结构敏感性分析.建立基于欧拉法和Realizable K-Epsilon湍流模型的空气-液滴两相流动的数学模型,通过计算流体力学软件对冷态工况下5种不同结构的旋叶分离器流场进行数值模拟,得到了不同液滴粒径下的分离效率变化曲线和液滴质量流量径向分布曲线,同时还通过冷态试验验证数值计算模型.结果表明:当液滴粒径等于5 μm或大于100 μm时,旋叶分离器效率对旋叶倾角和上升通道高度结构不敏感;当液滴粒径在5～100 μm时,18°旋叶的旋叶分离器分离效率大于 30°旋叶,上升通道高度等于其一倍直径时旋叶分离器分离效率最优;其中当液滴粒径等于30 μm时,旋叶分离器分离效率差值最大,结构敏感性最为显著.

Flow induced vibrations in steam generator and other heat exchangers can result in tube failure and lead to large economic losses. Tube supports such as tube support plates (TSP) or anti-vibration bars (AVB) are commonly used to suppress the flow induced vibration. Knowledge on the tube dynamic characteristics and instability phenomena under these support conditions is therefore essential, when fluid flows across the multi-span tubes. In this paper, flow induced vibration of a four-span rotated triangular tube array with a pitch-to-diameter ratio of 1.4 under three sets of AVB or TSP supports is experimentally studied. Part of the middle segment of the tube is subjected to cross flow. The vibration displacements and frequency spectra of a tube on first row are analyzed. It was found that the vibration mode for four-span tube bundle with TSP is different from that of the tube supported by AVB when fluid elastic instability (FEI) happens. The effective FEI critical velocities for the tube with TSP and that with AVB is 6.8 m/s and 3.0 m/s, respectively. Five correlations for predicting the onset of FEI of the tube bundle tested are compared with experimental results. Weaver and El-Kashlan correlation (1981b) is best. The effective FEI critical velocities predicted by Weaver and El-Kashlan correlation (1981b) are 15% and 27% higher than the experimental results for tube with TSP and tube with AVB, respectively. The vortex shedding responses are analyzed for the two tube-support structures. The Strouhal number for non-resonant vortex shedding response is calculated and compared with expectations from literature. The turbulent buffeting response for tube under different supports are compared with each other and discussed.

In pressurized water reactor, fretting corrosion has become the main reason for the failure of 690TT heat exchanger tubes. The effect of temperature on the fretting corrosion behavior between 690TT tube and 405 stainless steel (SS) bar has been studied during 106 fretting cycles. The overall average coefficient of friction (COF) values descends with an increase in test temperature, while the width of worn scar becomes wider. The severest fretting corrosion happens when the test temperature is at 100 °C. The wear mechanism differs at different test temperatures, from adhesive wear at room temperature to abrasive wear and delamination at 100 °C, to abrasive wear at 200 °C. Deformation slips, high residual strain concentration, and micro-cracks are found which are disadvantageous for the further service performance of the tubes.

In advanced nuclear power plant AP1000, the passive residual heat removal heat exchanger (PRHR HX) is a key part of the passive safety system. It can transfer decay heat from the reactor core to the water in the in-containment refueling water storage tank (IRWST) during the accidents. The secure operation of the C-shaped PRHR HX has a great impact on the reactor safety. In the present work, the heat transfer performance of a reduced scale PRHR HX was investigated by experiments and numerical simulations. A system including a 6 x 7 C-shaped tube bundle and a rectangular tank was studied. A three-dimensional numerical model using porous media method was developed to explore the transient single phase thermal hydraulic characteristics of the PRHR HX. The distributed resistances caused by the C-shaped tube bundle were calculated by empirical correlations. Both the heat transfer coefficients from the hot fluid in the tube to the inside surface of tube and that from the outer tube surface to the water in the IRWST were calculated by empirical correlations published in literatures. Tests were performed on the 6 x 7 C-shaped tube bundle installed in the middle of the tank. The outer wall temperature and bulk water temperature in the tank were measured by thermocouples. The numerical methods were validated by the experimental results. The calculated results of the local temperature distribution and heat transfer rate were both well consistent with the experimental values. The deviation of the heat transfer rate was smaller than 5%. (C) 2019 Elsevier Ltd. All rights reserved.

乏燃料贮存格架是核电厂贮存乏燃料的重要设备,在满载条件下和地震/跌落事故中,都应保持稳定和安全状态.本文基于LS-DYNA对乏燃料贮存格架进行了跌落事故冲击分析,考虑了最重重物从可能最高处意外跌落的情况.分析时考虑了碰撞、几何大变形、材料非线性等非线性因素.分析发现,浅跌落情况下贮存格架变形较大,但为局部变形,冲击载荷不影响贮存格架的安全功能.深跌落情况下,组件跌落在支座上方时支座承受的载荷最大,并在许用载荷范围内.为确保核电厂安全性和可靠性,基于分析得到的现象,设计和开展了乏燃料贮存格架浅跌落和深跌落试验.试验采用了等比例贮存腔和真实燃料组件的管座.同时对试验件进行了跌落分析,并与试验结果进行了对比,验证了分析技术的保守性和准确性.掌握的分析技术可应用于所有压水堆核电站乏燃料贮存格架的跌落事故分析.

Fretting is one of the main causes of failure of heat-transfer tubes in steam generators.It is important to understand the fretting fatigue of heat transfer tubes made of 690 alloy.By combining the finite element model and the self-compiled program,the fretting fatigue crack initiation life of 690 alloy in contact with the anti-vibration bar was analyzed to find the effect of contact pressure.T he results show that the crack initiation life under contact pressure is much lower than plain fatigue life.The life reduction is related to fretting contact status and fretting wear.An empirical formula with simple analytical expression was proposed to estimate the fretting fatigue life considering con-tact pressure.The empirical formula predicts a relatively conservative estimation for the fatigue life of 690 alloy and can be conveniently implemented in design and life prediction.