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.     

Simulation of stress corrosion crack growth in steam generator tubes

Assuming a small axial surface crack inside a steam generate (S/G) tube, stress corrosion crack growth is simulated by using finite element method. Pressure difference and residual stresses induced from the roll expansion are considered as applied forces and Scott's crack growth equation based on the stress intensity factor is used. Stress intensity factor distribution along crack front, variation of crack shape and crack growth rate are obtained during the crack growth. From the results, it is noted that for the given residual stress distribution, variation curve of the crack aspect ratio during the crack growth is uniquely determined. In addition, the curve shows nearly constant crack aspect ratio during the initial crack growth stage. When adjacently growing two small cracks are coalesced to form a longer crack, the growth rate of crack depth is increasing but that of crack length is decreasing, and the crack aspect ratio is converging to the original variation curve during the subsequent crack growth.     

Modelling the early development of secondary side stress corrosion cracks in steam generator tubes using incomplete random tessellation

A thorough understanding of the secondary side stress corrosion cracking of Inconel 600 in steam generator (SG) tubes seems to be still somewhat in the future. Especially the early phase of the development of cracks, also called the initiation phase, is beyond the present state-of-the-art explanations. An effort was, therefore, made to propose modelling and visualisation of the kinetics of secondary side stress corrosion crack initiation and growth on the grain-size scale:

An incomplete random tessellation is used to approximate the random planar grain structure.

The crack initiation is modelled by random processes, taking into account the most important factors such as proximity of the aggressive medium and the orientation of the grain boundaries relative to the stress field.

The stochastic process describing crack growth accounts for crack branching, coalescence and interference between neighbouring cracks.

Several numerical examples are provided to demonstrate the versatility of the proposed method. Reasonable qualitative agreement with metallographic results is shown.     

Stress identification in steam generator tubes from profile measurements

An identification method devoted to the determination of stresses in tubes, by means of profile measurements, available from on site non-destructive evaluations, is presented here. From the only furnished data (the radial displacement component w on the inner wall), the computation of the strain, and consequently the stresses in the elastic-plastic range, is made within the framework of the shell theory. For this purpose, we need to determine the associated curvature w″: this step is an ill-posed problem because of the lack of continuity with respect to the discrete data. This difficulty is overridden by means of an appropriate regularization procedure. The predictive ability of the method has been tested by comparison with direct simulations; we present an industrial application. This diagnosis tool has been applied successfully to PWR steam generator tubes (in the roll expansion transition zone) and vessel closure head penetrations.     

Simulation of the fluid-structure-interaction of steam generator tubes and bluff bodies

The accuracy of computational fluid dynamics in simulating the cross-flow around a steam generator and the feasibility of a full scale coupled CFD/FEA fluid-structure-interaction (FSI) analysis is examined through successive validations.The study begins with a comparison between experiment and computation of flow within a stationary tube bank. Results from the simulation of an individual tube experiencing two-degree-of-freedom flow-induced vibration (at a Reynolds number of 3800) are then shown to compare favorably to experimental results. Finally, free vibration of a single cantilevered hydrofoil is simulated with comparison of mean square acceleration at resonant and non-resonant velocities, respectively. The magnitudes and frequencies of vibration are shown to be accurately captured.     

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.     

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.     

Effect of shot peening on primary water stress corrosion cracking of Alloy 600 steam generator tubes in an operating PWR plant

The effect of shot peening on the primary stress corrosion cracking behavior of thermally treated Alloy 600 steam generator tubes in an operating pressurized water reactor (PWR) plant was analyzed based on pulled tube examinations and in-service inspection eddy current test (ISI-ECT) data. The evaluation was focused on the shape of crack, evolution of the number of new cracks and cracked tube fraction, and variation of crack length and the corresponding eddy current amplitude before and after shot peening. The shape of the crack was changed from a half-elliptical type before shot peening, to an elliptical one with bulging after peening. It was concluded that the shot peening was not effective for retarding both crack initiation and growth for this plant.     

PWR steam generator R&D in France: thermal-hydraulic and vibratory aspects

Concomitant with the launching of the French pressurized water reactor (PWR) nuclear power program, a large research and development (R&D) effort was initiated, devoted to the steam generators (SGs). This program, managed cooperatively by Framatome, the SG designer and manufacturer; Electricité de France (EDF), the French electrical utility; and the Commissariat à l'Energie Atomique (CEA), the French Atomic Energy Commission, primarily responsible for nuclear research; was focused on four main objectives:

1. (1) To obtain a better understanding of the physical phenomena existing in these steam generators and leading to SG performance alterations or operating life reductions.

2. (2) To test and validate improved design solutions for the model 51 Framatome steam generator, which was the first one designed under Westinghouse license.

3. (3) To test and validate new Framatome SG designs.

4. (4) To test and validate new, high-performance design tools.

This vast R&D program covers the following theses:

• - SG thermal-hydraulics,

• - SG tube vibration and wear,

• - SG materials (production, corrosion, etc.),

• - Primary and secondary fluid chemistry,

• - SG technology (manufacturing processes, NDT, etc.),

• - SG in-service inspection, and

• - SG maintenance.

These themes are too numerous to be dealt with in a single article. Consequently, the present article will focus on only the first two themes.     

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.     

Three-dimensional study on steady thermohydraulics characteristics in secondary side of steam generator

Steam generator (SG), as the primary-to-secondary heat exchanger and pressure boundary of primary loop, should be integrated and perform well in heat transfer ability. Flow characteristics of the secondary side fluid of SG are essential to analyze U-tube wastage caused by the flow-induced vibration and thermal stress. In this paper, secondary side two-phase flow was simulated based on the porous media model. Additional momentum and energy source terms were appended to the momentum and energy equations for porous media region, respectively. The additional momentum source contained the resistances of downcomer, tube bundle, support plate and separator. The additional energy source included the heat transfer from primary side to secondary side fluid. Solving the governing equations by ANSYS FLUENT solver yielded the distributions of velocity, temperature, pressure, density and quality, which can be used in the analysis of flow-induced vibration and separators. The thermal-hydraulic characteristics of hot side differed from these of cold side considerably. The minimum flow quality of cold side was 0.07, while the maximum one of hot side was 0.71; the average flow quality of outlet was 0.272. The flow rate in the gap of the hot side was 1.02 times of that of the cold side.     

Primary water stress corrosion cracking behaviors in the shot-peened Alloy 600 TT steam generator tubings

The effect of compressive residual stress on the primary water stress corrosion cracking behavior was investigated, based on the J-1 and J-2 nuclear power plant data. The following analyses were performed such as: (i) Weibull slope; (ii) crack growth rate; (iii) average crack length; (iv) crack length distribution. Alloy 600 TT exhibits strong heat to heat variations in its sensitivity to PWSCC. Crack growth rate was retarded after shot-peening. The compressive residual stress induced by shot-peening was more effective on new, short cracks, than on existing, long cracks. However, whether the ‘new’ cracks were initiated after peening is an unresolved issue, due to the present ECT sensitivity limit.     

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.     

A case study on detection and sizing of defects in steam generator tubes using eddy current testing

Eddy current testing (ECT) method is widely used to detect various types of defects occurring in nuclear steam generator tubes. Therefore, the reliability of its detection and sizing accuracy for defects should be validated. For this purpose, two tubes with defect signals were pulled from an operating steam generator and destructively examined. The defect type was a circumferential crack for one tube and an intergranular attack (IGA) for the other tube. The plus point coil probe showed a better capability to detect and size both a circumferential crack and a volumetric IGA than pancake and bobbin coil probe. The destructive results are correlated with the ECT results obtained during the in-service inspection.     

Techniques for processing remote field eddy current signals from bend regions of steam generator tubes of prototype fast breeder reactor

Steam generator (SG) is one of the most critical components of sodium cooled fast breeder reactor. Remote field eddy current (RFEC) technique has been chosen for in-service inspection (ISI) of these ferromagnetic SG tubes made of modified 9Cr–1Mo steel (Grade 91). Expansion bends are provided in the SGs to accommodate differential thermal expansion. During ISI using RFEC technique, in expansion bend regions, exciter–receiver coil misalignment, bending stresses, probe wobble and magnetic permeability variations produce disturbing noise hindering detection of defects. Fourier filtering, cross-correlation and wavelet transform techniques have been studied for noise reduction as well as enhancement of RFEC signals of defects in bend regions, having machined grooves and localized defects. Performance of these three techniques has been compared using signal-to-noise ratio (SNR). Fourier filtering technique has shown better performance for noise reduction while cross-correlation technique has resulted in significant enhancement of signals. Wavelet transform technique has shown the combined capability of noise reduction and signal enhancement and resulted in unambiguous detection of 10% of wall loss grooves and localized defects in the bend regions with SNR better than 7 dB.     

核电厂蒸汽发生器一次侧向二次侧泄漏率取值研究北大核心CSCD

合理确定蒸汽发生器一次侧向二次侧泄漏率取值,并据此制定核电厂运行策略,对核电厂的安全及稳定运行意义重大。本文根据泄漏率数值使用目的,将泄漏率分为用于辐射防护设计的泄漏率取值、用于核电厂运行控制的泄漏率控制值、用于保证蒸汽发生器传热管完整性的泄漏率保护阈值三大类,并探讨了各类取值的确定依据。完成了对国内外核电厂蒸汽发生器一次侧向二次侧泄漏率取值情况的调研分析,结合研究情况,提出了我国核电厂蒸汽发生器一次侧向二次侧泄漏率取值及控制的建议。     

Comprehensive investigation of the corrosion state of the heat exchanger tubes of steam generators. Part I. General corrosion state and morphology

The present work, constituting the first part of a series of two, deals with a systematic investigation of the general corrosion state of 22 heat exchanger tubes originating from different steam generators of the Paks NPP (Hungary). While the passivity of the inner surface of the stainless steel tube specimens was studied by voltammetry, the morphology and chemical composition of the oxide layer formed on the surfaces were analyzed by SEM–EDX method. Based on the measured corrosion characteristics (corrosion rate, thickness and chemical composition of the protective oxide layer) a strong dependence of these parameters on the decontamination history of the steam generators was revealed. It is well documented that the chemical decontamination carried out by a non-regenerative version of the AP-CITROX procedure does exert, on the long run, a detrimental effect on the corrosion resistance of steel surfaces. Therefore, process restrictions and modifications to minimize corrosion damages have be defined.     

Flow distribution in the inlet plenum of steam generator

The flow distribution in a 1/5th and 1/8th scale models of inlet plenum of steam generator (SG) has been studied by a combination of experiments and Computational Fluid Dynamics (CFD) simulations. The distribution of liquid sodium in the inlet plenum of the SG strongly affects the thermal as well as mechanical performance of the steam generator. Various flow distribution devices have been used to make the flow distribution uniform in axial as well as tangential direction in the window region. Experiments have been conducted to measure the radial velocity distribution using Ultrasonic Velocity Profiler (UVP) and Particle Image Velocimetry (PIV) under a variety of conditions. CFD modeling has been carried out for various configurations to give more insight into the flow distribution phenomena. The various flow distribution devices have been compared on the basis of a non-uniformity index parameter.