蒸汽发生器传热管堵头有效性分析

蒸汽发生器传热管是核电厂—回路压力边界的薄弱环节,传热管的完整性直接影响到整个一次侧的安全。当传热管出现裂纹、腐蚀或磨损等缺陷时,在评定确认可能会发生一次侧流体进入二次侧情况下,需要对传热管进行堵管。利用有限元法对某蒸汽发生器传热管的滚压堵头进行分析评定,模拟计算在堵管时以及堵管后堵头、传热管接触力情况,通过计算及分析确认堵头在极限运行工况的有效性,计算显示此堵头满足强度要求。     

含矩形缺陷圆柱形管的剩余强度评定方法

磨损是蒸汽发生器传热管一种常见的缺陷类型,磨损缺陷是影响传热管安全性的重要因素之一,需要根据管材的具体结构尺寸,制定适用的结构完整性评价方法。为了确定含矩形缺陷蒸汽发生器传热管的剩余强度,本文针对含矩形缺陷的传热管试样进行了内压爆破试验,并依据试验结果检验和评价了NB20013、BS7910、API579和Janelle规定的含局部减薄缺陷承压构件中剩余强度系数RSF计算式的精度。分析表明:当缺陷深度比a/t≥70%时,按以上方法评价含缺陷管的安全性时都可能偏于不保守。最后,基于BS7910建立了适用于单个矩形缺陷改进型RSF计算式,其计算值既具有很高的计算精度,又具有满意的可靠性。     

蒸汽发生器传热管在泵致脉动压力载荷下的动力学响应研究

泵致脉动压力是核电站中引起主设备部件疲劳失效的主要原因之一。本文建立了蒸汽发生器传热管的泵致脉动压力载荷表达式，并建立不同弯曲半径的传热管有限元模型，对蒸汽发生器传热管在泵致脉动压力载荷下的动力学响应进行了研究。结果表明：34、64、94、114、124、144排传热管附近的频率、振型对泵致脉动压力最为敏感；包络泵致脉动压力作用下，最大应力出现在32排传热管上；传热管在泵致脉动压力载荷作用下，泵致脉动压力载荷的轴频频率对结构响应的贡献最大。本文分析结果为蒸汽发生器传热管在泵致脉动压力载荷下的磨损分析提供了参考。     

快堆蒸汽发生器沸腾传热恶化引起的热应力计算和分析

快堆管壳式直流蒸汽发生器发生沸腾传热恶化是不可避免的,由此引起的传热管管壁温度波动会使传热管受到疲劳破坏。研究蒸汽发生器的沸腾传热恶化及热疲劳破坏的实验昂贵,难度较大。本文依据国外已发表的实验结果,建立蒸汽发生器沸腾传热恶化发生时传热管管壁温度及热应力的分析模型,应用数值方法求解,对蒸汽压力、质量流速、钠汽温差变动的影响进行了讨论并给出了主要结论。     

蒸汽发生器传热管结垢厚度的涡流检测方法与应用

蒸汽发生器二回路中有较多的沉积物存在并危害传热管安全,利用涡流检测方法可以对传热管二次侧泥渣进行有效检测。通过模拟传热管结垢的不同厚度并进行实验,可获得厚度与幅值的对应关系。本文描述了对蒸汽发生器传热管结垢的检测方法及幅值与厚度的对应关系,为统计蒸汽发生器传热管外壁结垢情况提供了较为有效的参考基准量。     

ACP1000蒸汽发生器I-690TT合金U形传热管评定技术

为了固化核电厂蒸汽发生器传热管的全流程制造工艺和关键工艺参数,保证传热管批量化制造时质量的稳定性,提出了一整套评定技术方案。该技术方案可对核电厂I-690TT合金U形传热管的化学成分、机械性能、金相组织的均匀性及无损检测方法的有效性进行全面验证,并在ACP1000蒸汽发生器传热管国产化研制过程中成功应用。     

蒸汽发生器传热管的降质及对其完整性的评估

蒸汽发生器传热管是反应堆冷却剂压力边界的主要组成部分，这就意味着必须保持传热管的完整性。然而，运行经验表明，蒸汽发生器传热管会出现各种降质。这些降质可能会导致管子的泄漏或破裂，使反应堆冷却剂丧失，并提供了直接通向二回路和释放到环境中去的途径。本文将介绍几种已知的传热管降质，传热管完整性性能准则．并对蒸汽发生器传热管完整性进行评估。     

基于声发射技术的蒸汽发生器传热管点腐蚀研究

蒸汽发生器传热管的腐蚀是影响核动力装置安全运行的重要问题之一,传热管的腐蚀以点腐蚀的危害最为常见。利用声发射仪器,对蒸汽发生器传热管进行腐蚀实验时的信号进行采集和分析,并对腐蚀点进行了准确定位。实验结果表明,传热管的点腐蚀经历3个阶段:发展期、平稳期和迅速发展期。声发射技术能比其它任何无损检测方法更早地发现传热管腐蚀损伤,可对蒸汽发生器的安全和运行情况进行在线实时监测,具有重要的意义。     

传热管泄漏对蒸汽发生器动态特性的影响

以大亚湾核电站蒸汽发生器为研究对象,建立了基于漂移流理论的蒸汽发生器一维动态数学模型及传热管泄漏模型,并进行了蒸汽发生器不同工况下的稳态仿真。在验证所建立漂移流模型和传热管泄漏模型的基础上,研究了不同工况下传热管泄漏位置及泄漏流量对蒸汽发生器关键参数的影响。研究结果表明,所建立的漂移流模型和传热管泄漏模型能准确反映不同泄漏情况下蒸汽发生器质量含汽率及蒸汽压力等关键参数的变化规律,泄漏发生在热端沸腾段入口处时各参数变化最显著,泄漏量为冷却剂流量的5%时出口质量含汽率由0.261降到0.163。基于漂移流理论传热管泄漏对蒸汽发生器动态特性影响的成功预测,为蒸汽发生器传热管泄漏事故的监测与防范措施的制定提供一定参考。     

基于流热固耦合的核电蒸汽发生器传热管热应力数值模拟

以大亚湾核电站蒸汽发生器为原型,基于相似模化原理建立了蒸汽发生器简化物理模型。采用两流体模型及热弹性力学基本关系式分别描述气液两相流沸腾相变过程和热应力变化规律。利用CFX对一、二回路侧流体流动传热及与传热管的耦合换热过程进行了数值模拟,并在ANSYS WORKBENCH中实现了流体温度场载荷向结构的传递,进而对传热管进行稳态热分析和热应力分析。计算结果表明：二回路出口质量含汽率为24.5%,冷却剂出口温度为296.2 ℃,均与大亚湾蒸汽发生器实际运行参数相符;传热管热应力与其壁面温差分布一致,且沿壁厚方向先减小后增大,并存在中性层,传热管最大热应力为54.5 MPa。研究结果为蒸汽发生器的优化设计及安全运行提供了一定的理论支撑。     

Plastic limit load analysis for steam generator tubes with local wall-thinning

This paper introduces the study of experimental and numerical analysis for plastic limit loads of Inconel 690 steam generators (SG) tubes with local wall-thinning defects. Meanwhile, the effect of the three dimensions of a local wall-thinning defect on the plastic limit load of SG tubes is analyzed.A test facility which can test both burst pressure and plastic limit load of SG tubes was established and SG tubes with 3 typical types of defects were tested by using the facility. A regularization method for local wall-thinning defect is proposed and the finite element method was used to analyze the plastic limit load of SG tubes with defects. Compared with the experimental results of SG tubes with real defects, the calculated values of plastic limit load for SG tubes with regularized defects are conservative.Based on finite element method, the effect of the three dimensions of local wall-thinning defects on plastic limit loads of defected Inconel 690 SG tubes has been got. The studied results show that the defect depth of a local wall-thinning defect is the main factor influencing the plastic limit load of SG tubes, on the other hand, both the longitudinal length and the circumferential length of a defect have effect on the plastic limit load of SG tubes.It is found that in some cases, when the longitudinal length and the circumferential angle of a local wall-thinning defect exceed some extent, the effect of the longitudinal length and the circumferential angle on plastic limit load can be ignored.     

关于蒸汽发生器传热管应力腐蚀的几个问题

本文探讨了以下5个问题：（1）传热管发生氯致应力腐蚀的原因；（2）发生氯致应力腐蚀裂纹机制的分析；（3）没有涡流探伤显示信号的其他传热管的可用性：（4）是否可以先堵已切割的传热管；（5）美国的堵管准则和俄罗斯的堵管准则的比较。     

Influence of flow stress choice on the plastic collapse estimation of axially cracked steam generator tubes

The influence of the choice of flow stress on the plastic collapse estimation of axially cracked steam generator (SG) tubes is considered. The plastic limit and collapse loads of thick-walled tubes with external axial semi-elliptical surface cracks are investigated by three-dimensional non-linear finite element (FE) analyses. The limit pressure solution as a function of the crack depth, length and tube geometry has been developed on the basis of extensive FE limit load analyses employing the elastic–perfectly plastic material behaviour and small strain theory. Unlike the existing solutions, the newly developed analytical approximation of the plastic limit pressure for thick-walled tubes is applicable to a wide range of crack dimensions. Further, the plastic collapse analysis with a real strain-hardening material model and a large deformation theory is performed and an analytical approximation for the estimation of the flow stress is proposed. Numerical results show that the flow stress, defined by some failure assessment diagram (FAD) methods, depends not only on the tube material, but also on the crack geometry. It is shown that the plastic collapse pressure results, in the case of deeper cracks obtained by using the flow stress as the average of the yield stress and the ultimate tensile strength, can become unsafe.     

Evaluation of the burst characteristics for axial notches on SG tubings

Some events of steam generator tubes have been reported in some nuclear power plants around the world. Main causes of the leakage are from various types of corrosion in the steam generator (SG) tubing. Primary water stress corrosion cracking (PWSCC) of steam generator tubing have occurred in many tubes in Korean plants, and they were repaired using sleeves or plugs. In order to develop proper repair criteria, it is necessary to ascertain the leak behavior of the tubings. A high-pressure leak and burst testing system was manufactured. Various types of electro-discharged-machined (EDM) notches having different lengths were machined on the o.d. of test tubes to study SG tube behavior. Leak rate and ligament rupture pressure as well as the burst pressure were measured for the tubes at room temperature. Rupture pressure of the part through-wall defect tubes depends on the defect depth and length. Water flow rates after the rupture were independent of the flaw types; tubes having 20–60 mm long EDM notches showed similar flow rates regardless of the initial defect depth. A fast pressurization rate generated a lower burst pressure than the case of a slow pressurization.     

Coalescence evaluation of collinear axial through-wall cracks in steam generator tubes

To maintain the structural integrity of steam generator tubes, usually, 40% of wall thickness plugging criterion has been adopted. However, since the criterion is applicable only for the steam generator tube containing a single crack, the interaction effect of multiple cracks cannot be considered. In this paper, the coalescence pressure of tube with dual cracks is evaluated based on detailed three-dimensional elastic–plastic finite element analyses. In terms of the crack configuration, collinear axial through-wall cracks with various length, distance and ratio between individual cracks are selected. The applicability of failure pressure prediction models recently proposed by the authors was verified by comparing the finite element analyses results with corresponding experimental data for tubes with two identical cracks. Further, in order to quantify the effect of crack length ratio on failure behavior, the failure pressure prediction model was used expansively for tubes containing different-sized cracks and a coalescence evaluation diagram was developed.     

Predicting the accumulated number of plugged tubes in a steam generator using statistical methodologies

A steam generator (SG) plays a significant role not only with respect to the primary-to-secondary heat transfer but also as a fission product barrier to prevent the release of radionuclides. Tube plugging is an efficient way to avoid releasing radionuclides when SG tubes are severely degraded. However, this remedial action may cause the decrease of SG heat transfer capability, especially in transient or accident conditions. It is therefore crucial for the plant staff to understand the trend of plugged tubes for the SG operation and maintenance. Statistical methodologies are proposed in this paper to predict this trend. The accumulated numbers of SG plugged tubes versus the operation time are predicted using the Weibull and log–normal distributions, which correspond well with the plant measured data from a selected pressurized water reactor (PWR). With the help of these predictions, the accumulated number of SG plugged tubes can be reasonably extrapolated to the 40-year operation lifetime (or even longer than 40 years) of a PWR. This information can assist the plant policymakers to determine whether or when a SG must be replaced.     

Determination of equivalent single crack based on coalescence criterion of collinear axial cracks

In a nuclear power plant the steam generator tubes cover a major portion of the primary pressure-retaining boundary. Thus, very conservative approaches have been taken in the light of steam generator tube integrity. According to the present criteria, tubes wall-thinned in excess of 40% should be plugged whatever the cause. However, many analytical and experimental results have shown that no safety problems exist even with thickness reductions greater than 40%. The present criterion was developed about 20 years ago when wear and pitting were dominant causes for steam generator tube degradation, and it is based on tubes with single cracks regardless of the fact that the appearance of multiple cracks is more common in general. The objective of this study is to review the conservatism of the present plugging criteria of steam generator tubes and to propose a new coalescence model for two adjacent through-wall cracks existing in steam generator tubes. Using the existing failure models and experimental results, we reviewed the conservatism of the present plugging criteria. In order to verify the usefulness of the proposed new coalescence model, we performed finite element analysis and some parametric studies. Then, we developed a coalescence evaluation diagram.     

压水堆蒸汽发生器一、二次侧稳态流场耦合分析

蒸汽发生器（SG）在运行过程中主要面临流致振动所导致的传热管破裂事故,而流致振动分析需以SG内的三维两相流场作为输入条件。采用多孔介质模型,对SG二次侧流场进行求解,同时耦合一、二次侧换热,获得SG二次侧速度场、温度场、压力场及流动含气率分布,并获得传热管一维的一、二次侧流体温度和换热系数及传热管温度分布。由于一次侧向二次侧释热极不均匀,SG内流场分布及汽水分离器内的含气率分布极不均匀;汽水分离器内的最大、最小含气率分别为0.62和0.05,该参数可为汽水分离器负载设计提供依据。通过计算还获得弯管区速度分布,该分布可为传热管的流致振动磨损评估提供输入条件。     

Risk–benefit analyses of SG tube maintenance based on probabilistic fracture mechanics

As an application of probabilistic fracture mechanics (PFM), a risk–benefit analysis was performed for the purpose of optimizing maintenance activities of steam generator (SG) tubes used in pressurized water reactors (PWRs). The probabilities of the SG tube leakage and rupture are defined as risks in this study. A model was made modifying pc-PRAISE (Piping Reliability Analysis Including Seismic Events) to evaluate the risks during 60 year operations due to stress corrosion cracking (SCC) of the tubes under various maintenance strategies for SG tubes.In the risk analysis, parameters such as inspection accuracy, inspection interval, sampling inspection and crack propagation law were selected for sensitivity analysis. Based on the risk analysis, a risk–benefit analysis was conducted when implementing two maintenance strategies taking both costs and revenues for 60 year operations into account. In the risk–benefit analysis, the expected cost of leakage or rupture was calculated by multiplying ‘probability of leakage or rupture’ by ‘expected loss of leakage or rupture accident’. To justify whether it is worthwhile implementing the maintenance strategies or not, the net present value (NPV) was calculated as an index, which is one of the most fundamental financial indices for decision-making based on the discounted cash flow (DCF) method.The results demonstrated that in the risk analysis, the risks are influenced significantly by the crack propagation law, accuracy of inspection and sampling inspection. In the risk–benefit analysis, it was suggested that investment to improve inspection accuracy would reduce the total costs of 60 year operations significantly and increase the NPV.Although the analysis was mainly conducted for SG tubes made of Inconel 600 mill anneal (MA) material, the analysis was also carried out for Inconel 690 thermal treatment (TT) material, making assumptions on its crack initiation and crack propagation law. In addition, the effect of introducing maintenance criteria, namely, operation with a crack justified by certain criteria, on NPV was evaluated.     

TA16传热管的断裂韧度测试方法研究

近年来，以蒸汽发生器传热管为代表的小尺寸管材的断裂韧度评价方法受到了日益关注。本文设计了一种可用于TA16传热管断裂韧性测试的含径向裂纹C形试样，基于弹塑性有限元分析获得试样的应力强度因子K和J积分的计算式。采用规则化法完成了TA16传热管的断裂韧度试验，试验结果表明，不同试样得到的J阻力曲线和条件启裂韧度J_Q的分散性均较小，J_Q均值为32.875 MPa•mm，标准差为1.377 MPa•mm。