Investigation on the interaction effect of two parallel axial through-wall cracks existing in steam generator tube

It is commonly required that steam generator tubes wall-thinned in excess of 40% should be plugged. However, the plugging criterion is known to be too conservative for some locations and types of defects and its application is confined to a single crack. In the previous study, the conservatism of the present plugging criterion of steam generator tubes was reviewed and a crack coalescence model applicable to steam generator tubes with two collinear axial through-wall cracks was proposed. Since parallel axial cracks are more frequently detected during in-service inspections than collinear axial cracks, the studies on parallel axial cracks spaced in circumferential direction are necessary. The objective of this paper is to investigate the interaction effect between two parallel axial through-wall cracks existing in a steam generator tube. Finite element analyses were performed and a new failure model of the steam generator tube with these types of cracks is suggested. Interaction effects between two adjacent cracks were investigated to explain the deformation behavior of cracked tubes.     

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.     

Restraining effect of support plates on the limit loads for circumferential cracks in the steam generator tube

The steam generator in a nuclear power plant is a large heat exchanger that uses heat from a reactor to generate steam to drive the turbine generator. Rupture of a steam generator tube can result in release of fission products to environment outside. Therefore, an accurate integrity assessment of the steam generator tubes with cracks is of great importance for maintaining the safety of a nuclear power plant. The steam generator tubes are supported at regular intervals by tube support plates and rotations of the tubes are restrained. Although it has been reported that the limit load for a circumferential crack is significantly affected by boundary condition of the tube, existing limit load solutions do not consider the restraining effect of tube support plate correctly. In addition, there are no limit load solutions for circumferential cracks in U-bend region with the effect of tube support plate. This paper provides detailed limit load solutions for circumferential cracks in top of tube sheet and the U-bend regions of the steam generator tube with the actual boundary conditions to simulate the restraining effect of the tube support plate. Such solutions are developed based on three-dimensional (3-D) finite element analyses. The resulting limit load solutions are given in a polynomial form, and thus can be simply used in practical integrity assessment of the steam generator tubes.     

Safety and availability of steam generator tubes affected by secondary side corrosion

The outside diameter stress corrosion cracking at tube support plates became the dominating ageing mechanism in steam generator tubes made of Inconel 600. A variety of maintenance approaches were developed and implemented world-wide to enable safe and reliable plant operation with affected tubes. Despite different philosophical and physical backgrounds involved, all applied approaches satisfy relevant regulatory requirements. The main goal followed in this paper is to quantify the degree of safety which is achieved through the implementation of selected maintenance approaches. A method is proposed which measures the operational safety and availability through three efficiency parameters: probability of steam generator tube rupture; predicted accidental leak rates through the defects in the tube bundle; and number of plugged tubes. An original probabilistic model quantifies the probability of tube rupture, while procedures available in literature were used to evaluate the accidental leak rates. A numerical example is based on data from the Kr ko NPP (PWR 623 MWe). The maintenance strategies analyzed are: (a) traditional defect depth (40%) plugging criterion; (b) alternate plugging criterion (bobbin coil voltage as defined by EPRI and US NRC); (c) combination of traditional and alternate plugging criteria; and (d) no plugging at all. Advantages of the defect specific approaches (b) and (c) over the traditional one (a) are clearly shown. The efficiency of the traditional approach (a) is shown to be comparable to the no plugging at all approach (d). Finally, a sensitivity analysis aimed at ranking of the input parameters is presented. Uncertain failure models are shown to be the major contributor to the scatter of obtained results.     

含缺陷Inconel 690蒸汽发生器传热管的强度及堵管准则研究

针对缺陷对传热管强度的影响以及传热管判废准则问题展开研究,研制了适用于小管径蒸汽发生器传热管极限载荷及爆破压测试的实验装置,对含体积型缺陷及面型缺陷的Inconel 690蒸汽发生器传热管进行了实验研究,并采用有限元法对极限载荷及爆破压进行了估算.在此基础上,研究了传热管的堵管准则,提出了两级评定方法.该评定方法可根据缺陷的深度、轴向及环向长度来综合评价.     

Prediction of crack coalescence of steam generator tubes in nuclear power plants

Prediction of failure pressures of cracked steam generator tubes of nuclear power plants is an important ingredient in scheduling inspection and repair of tubes. Prediction is usually based on nondestructive evaluation (NDE) of cracks. NDE often reveals two neighboring cracks. If the cracks interact, the tube pressure under which the ligament between the two cracks fails could be much lower than the critical burst pressure of an individual equivalent crack. The ability to accurately predict the ligament failure pressure, called “coalescence pressure,” is important. The failure criterion was established by nonlinear finite element model (FEM) analyses of coalescence of two 100% through-wall collinear cracks. The ligament failure is precipitated by local instability of the ligament under plane strain conditions. As a result of this local instability, the ligament thickness in the radial direction decreases abruptly with pressure. Good correlation of FEM analysis results with experimental data obtained at Argonne National Laboratory’s Energy Technology Division demonstrated that nonlinear FEM analyses are capable of predicting the coalescence pressure accurately for 100% through-wall cracks. This failure criterion and FEA work have been extended to axial cracks of varying ligament width, crack length, and cases where cracks are offset by axial or circumferential ligaments.     

In-service inspection of PWR steam generators by eddy currents; French achievements in defect detection and evaluation

In a nuclear plant the steam generator tubes must be efficiently inspected. The highest possible detection sensitivity is necessary to get a clear decision for plugging the rejected tubes. For this reason multi-frequency eddy current examinations have been developed at the Commissariat à l'Energie Atomique and industrialized by Intercontrôle. Defects are characterised case by case as the needs arose on the site: sludge height determination; measurement of tube deformation under pressure of oxides; determination of tube degradation due to the pressure of foreign bodies; detection of cracks under coatings; detections of cracks at the end of flanging. Problems unsolved by standard probes were dealt with by rotating-head machines. This report sums up part of the research undertaken in the field of eddy current testing.     

高温气冷堆蒸汽发生器两相流不稳定性预报

论述了蒸汽发生器立式上升流动螺旋管内高压汽－水两相流不稳定性试验研究。研究结果表明，螺旋管中存在压力降型、密度波型和热力型脉动。采用无因次分析法得到了预测系统稳定的经验关系式，并得到了判断系统稳定性的界限图。同时对蒸汽发生器立式下降流动螺旋管与立式上升流动螺旋管的不稳定性进行了比较。最后对实际蒸汽发生器两相流不稳定性进行了预报。     

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.     

螺旋管管束流体诱发振动的实验研究

根据200MW高温气冷堆蒸汽发生器的结构参数建立了实验研究模型,并针对相邻两层旋向相同的螺旋管管束,在横向流的冲刷下,对管子的动态响应进行了实验研究。实验结果表明,螺旋管管束的流体诱发振动响应与直管相比存在一定的差别;200MW高温气冷堆蒸汽发生器的设计避免了管束的大振幅振动。     

Overview of steam generator tube degradation and integrity issues

The degradation of steam generator tubes in pressurized water nuclear reactors, and, in particular, the stress corrosion cracking (SCC) of Alloy 600 tubes, continues to be a serious problem. Primary water SCC is commonly observed at the roll transition zone (RTZ), at U-bends and tube denting locations, and occasionally in plugs and sleeves. Outer-diameter SCC (ODSCC) and intergranular attack (IGA) commonly occur near tube support plate (TSP) crevices, near the tube sheet in crevices, or under sludge piles, and occasionally in the free span. A particularly troubling recent trend has been the increasing occurrence of axial and circumferential cracking at the RTZ on both the primary and secondary sides. Outer-diameter stress corrosion cracking in TSP crevices, commonly consisting of segmented axial cracks with interspersed uncracked ligaments, is also becoming more common. Despite recent advances in inservice inspection (ISI) technology, a clear need still exists for quantifying and improving the reliability of ISI methods with respect to the probability of detection of the various types of flaws and their accurate sizing. These improvements are necessary to permit an accurate assessment of the consequences of leaving degraded tubes in service over the next reactor operating cycle. Degradation modes such as circumferential cracking, intergranular attack, and ODSCC at the TSP have affected a large number of tubes. New regulatory guidance is being developed that requires the development and implementation of a steam generator management program that monitors tube condition against accepted performance criteria to ensure that the tubes will perform the required safety function over the next operating cycle. In formulating new guidance for the implementation of alternate repair criteria, the U.S. Nuclear Regulatory Commission is also evaluating the contribution to overall plant risk from severe accidents. Preliminary analyses are being performed for a postulated severe-accident scenario that involves station blackout and loss of primary feedwater, in order to determine the probability of failure for degraded tubes.     

Status and progress of research on a removed-from-service steam generator

Under an NRC directed group sponsored project (including French, Italian, Japanese, and EPRI participation) a steam generator removed from service is the subject of extensive research. The generator now serves as a vehicle for studies involving validation of the accuracy and reliability of current nondestructive examination (NDE) characterization during inservice inspections, determination of remaining integrity of service defected steam generator tubes, determination of failure consequences (leak rate) of defects, and demonstration of cleaning and decontamination techniques. Program objectives are to provide inputs to regulatory guides on inservice inspection and tube plugging criteria.During the past year dilute chemical reagent decontamination of the steam generator channelhead using modified LOMI and Candecon processes, each on one side, has been completed. In addition to decontamination effectiveness other factors such as corrosivity during the process, methods for waste handling, and potential for affecting return to service of the component were evaluated. Following decontamination a subcontracted effort was conducted to remove a large number of plugs placed in the generator during its service life. In under three weeks 969 explosive type plugs were removed. This provided a new level of experience in large scale plug removal. The plug removal was to optimize access to defected steam generator tubes for nondestructive primary side characterization. The first of these nondestructive examinations was conducted toward fiscal year end, employing state-of-the-art eddy current technology. A map of the generator condition, as obtained from 100% eddy current examination, is being formed. In parallel with the above endeavors, extensive efforts have been made toward characterizing the secondary side of the generator. The tubesheet surface and the inner row U-bend regions have been extensively examined. Innovative photographic approaches have provided success in documenting generator conditions in the sludge pile area.     

Life prediction of steam generator tubing due to stress corrosion crack using Monte Carlo Simulation

The failure of steam generator tubing is one of the main accidents that seriously affects the availability and safety of a nuclear power plant. In order to estimate the probability of the failure, a probabilistic model was established to predict the whole life-span and residual life of steam generator (SG) tubing. The failure investigated was stress corrosion cracking (SCC) after the generation of one through-wall axial crack. Two failure modes called rupture mode and leak mode based on probabilistic fracture mechanics were considered in this proposed model. It took into account the variance in tube geometry and material properties, and the variance in residual stresses and operating conditions, all of which govern the propagations of cracks. The proposed model was numerically calculated by using Monte Carlo Simulation (MCS). The plugging criteria were first verified and then the whole life-span and residual life of the SG tubing were obtained. Finally, important sensitivity analysis was also carried out to identify the most important parameters affecting the life of SG tubing. The results will be useful in developing optimum strategies for life-cycle management of the feedwater system in nuclear power plants.     

蒸汽发生器实验回路倒流特性仿真分析

针对立式倒U型管蒸汽发生器传热管内出现的倒流现象,基于RELAP5/MOD3.3程序,采用新的控制体划分方案对蒸汽发生器实验段进行建模,模拟实验回路中发生的倒流现象。通过与实验数据进行对比分析,验证建模方案的正确性。在此基础上,分析倒流现象对蒸汽发生器实验段流动传热的影响。结果表明:倒流现象发生在较短管内,对于单个U型管,倒流管的流量高于正流管。倒流发生后,系统进入相对稳定状态,但蒸汽发生器实验段的换热功率和进出口腔室负压降绝对值显著降低。     

Computational investigation of the effect of hydraulic and nondestructive test parameters on the strength reliability of the heat exchanger tubes of a steam generator

The results of calculations of the probability of a leak appearing in the tube band of steam generator in a VVéR-440 reactor system during operation are presented. The MAVR-1.1 computer code is used to calculate the probability of the formation of a leak and rupture of one of the heat exchanger tubes. The binomial distribution is used to determine the probability of the number of tubes that do not satisfy the plugging criterion. A leak in a tube bank is calculated as a sum of leaks in individual tubes. The probability of such a leak is calculated as a random sum. The calculations show that the parameters of test measures (pressure of the hydraulic tests, reliability of nondestructive testing for defects) and the sequence in which they are performed have a large effect on the failure probability of a tube bank during reactor operation. The computational results and the experience gained in operating steam generators show that the algorithm and the method developed for computing the leak probability could be helpful for estimating the strength reliability of heat exchanger tubes. __________ Translated from Atomnaya énergiya, Vol. 102, No. 4, pp. 216–221, April, 2007.     

Structural analysis of electrosleeved tubes under severe accident transients

A flow stress model was developed for predicting failure of electrosleeved PWR steam generator tubing under severe accident transients. The electrosleeve, which is nanocrystalline pure nickel, loses its strength at temperatures greater than 400°C during severe accidents because of grain growth. A grain growth model and the Hall–Petch relationship were used to calculate the loss of flow stress as a function of time and temperature during the accident. Available tensile test data, as well as high-temperature failure tests, on notched electrosleeved tube specimens were used to derive the basic parameters of the failure model. The model was used to predict the failure temperatures of electrosleeved tubes with throughwall and part-throughwall axial cracks in the parent tube during a postulated severe accident transient.     

蒸汽发生器传热管的微动磨损损伤及预测模型

蒸汽发生器是核电厂中能量转换的关键装备,内部高速流经的高温、高压流体引起传热管流激振动,造成传热管微动磨损损伤,严重时发生管道破裂。文章介绍了传热管典型的微动磨损失效案例,相应的模拟实验研究结果,以及机械磨损与冲蚀-腐蚀共同作用的损伤机制。采用工作率模型可对传热管的磨损失效进行合理的寿命预测评估,该预测模型已经在核电厂安全评估方面应用。     

双弹性管流固耦合振动的数值模拟

为研究反应堆结构中诸如燃料棒、蒸汽发生器和其他换热器等管束类结构的流固耦合振动问题,利用有限体积法离散大涡模拟的流体控制方程及有限元方法离散结构动力学方程,结合动网格技术,建立了三维流体诱发弹性管束振动的数值模型,实现了计算结构动力学与计算流体力学之间的双向耦合。得到横流作用下单管的振动响应,并与已有的实验数据比较,证明了本文模型的合理性;对横流作用下的两串列管、两并列管的流固耦合振动进行了数值模拟,着重研究了节径比为1.2、1.6、2、3、4的两弹性管在不同流速作用下的动力学响应及流场特性;得到串列管、并列管的临界间距与临界流速。     

Analysis of Overheating Rupture in Heat-Transfer Tubes Causing Corrosive High-Temperature Reaction

In the steam generator of a liquid metal fast breeder reactor, a defect penetrating through heat-transfer tube will cause high-pressure water/steam to spout into the low-pressure sodium filling the space outside the tube, to initiate sodium-water reactions. If the leak exceeds an intermediate level (~2kg/s), the reaction jet may rupture adjoining tubes with overheating in the event of insufficient cooling available inside the tubes. Such phenomenon of overheating tube rupture presents a serious problem to the economy and safety of steam generator. With a view to clarifying the failure behavior of steam generator heat-transfer tubes under such condition a model of the phenomenon is derived through a series of tests on sodium-water reactions making use of a test loop representing the scale model of an actual fast breeder steam generator. Comparison of actual test data with analysis based on the model has yielded the following information: The failure behavior of gas-pressurized tubes fall into two categories: (a) by creep failure—occurring upon increase of cumulative damage with tube wall wastage caused by the reaction jet and (b) by ductile failure accompanied by creep—upon tube heating with the reaction jet to the extent of lowering tube wall strength below the hoop stress exerted by tube pressure. Analysis of the two categories of failure results in estimation of the percentage difference between analyzed and measured times to failure of 35–50% in the case of creep failure and of 20–50% in the case of ductile failure accompanied by creep. In practical application to steam generators in order to provide a safety margin a time factor—i.e., the safety factor indicating multiple of actual time to failure—of 3 is adopted against 1.5–2 indicated from test to be the actually applicable value.     

CPR1000蒸汽发生器二次侧三维稳态热工水力分析

采用三维稳态分析软件GENEPI,对CPR1000蒸汽发生器二次侧管束区进行了热工水力计算,利用多孔介质及局部阻力系数来表征传热管及各几何部件的复杂结构和压降影响,得到了二次侧管束区流场、温度场等的分布情况。计算结果表明:管束区最大干度为0.3;将典型传热管的动能数据提供给流致振动软件进行计算分析,结果显示在本工况下,传热管的流致振动在可接受范围内;对管板附近的流场及温度场进行分析,预测了此模型及工况下的泥渣沉积区域,为排污管的设计提供了输入数据。计算结果验证了CPR1000蒸汽发生器二次侧管束区设计的合理性。