Development of probability of detection functions for evaluating integrity of axial outside diameter stress corrosion cracks in steam generator

The reliability of an eddy current testing (ECT) inspection system depends upon the inspection technique and quality of analyst. In evaluating the integrity of a steam generator (SG) tube, degradation detection and sizing accuracy are considered performance measures of the nondestructive evaluation (NDE) system. A probability of detection (POD) model serves as a functional measure of the ability of an NDE system to detect degradation. It is one of the inputs in the operational assessment, and it is used to estimate the degradation during service via ECT of the SG tube. In this study, the POD functions of the inspection technique and analyst were obtained to quantitatively analyze the ECT bobbin probe for axial outside diameter stress corrosion cracks in SG tubes. This should serve to evaluate the integrity of the SG tubes. The depth and amplitude of defects were used as parameters of the POD model. Hit (detection) and miss (no detection) binary data obtained from destructive and nondestructive inspection of cracked tubes were also used.     

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.     

快堆蒸汽发生器热力参数对泄漏探测系统响应特性的影响

研究建立了水泄漏引起的钠水反应产物在快堆蒸汽发生器和取样支路传输扩散的一维数学模型,分析了蒸汽发生器流量、钠温度和取样支路流量对泄漏探测系统响应特性的影响。模型计算和实验结果表明：蒸汽发生器流量的增加将缩短系统的响应时间,但却降低了蒸汽发生器钠出口处的氢离子浓度,使系统探测水泄漏的灵敏度降低;蒸汽发生器钠温度对系统的响应时间影响不大,钠温升高,OH^-离子的离解速率加快,探测系统的灵敏度提高;增大取样支路流量可改善系统的响应特性。     

Probabilistic modelling of steam generator tube pitting corrosion

Pitting corrosion is a serious form of degradation in steam generator (SG) tubing of some nuclear stations. The nature and extent of the pitting process is assessed through inspection programs, typically using various eddy current (EC) techniques, while the impact of pitting is minimized through deposit removal maintenance activities such as water lancing and chemical cleaning of SGs. This paper presents a probabilistic model of SG tube pitting corrosion that incorporates trends observed from a large EC inspection database from a nuclear generating station. The pitting occurrence process is modelled as a stochastic Poisson process and the pit size is treated as a random variable. The model is statistically calibrated with the available EC inspection data. The model is applied to estimate the probability of tube leakage, forced outage rate and the distribution of the number of tubes plugged per SG in a given operating interval. The proposed model is useful in optimizing strategies for the life-cycle management of SGs.     

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.     

Calcul d'energie de fautes d'empilement dans le dioxyde d'uranium

Energy computations of some (100), (110) and (111), planar defects were performed using an ionic bond model for stoichiometric uranium dioxyde. The repulsive contribution to the fault was estimated in two different ways, i.e. using the Born-Mayer classical treatment, or potentials derived from shell model calculations. The stability of the various defect configurations has been studied; on the basis of our numerical values, we may conclude that dislocation dissociation is unlikely in stoichiometric uranium dioxyde.     

Secondary side corrosion in steam generator tubes: lessons learned in France from the in-service inspection results

Non-destructive testing (NDT) has proved to be very important in the maintenance of steam generator tubing. This is particularly true in the case of secondary side corrosion, because this type of degradation leads to various morphologies which are often complex (intergrranular attack) (IGA), intergranular stress corrosion cracking (IGSCC), or a mixture of both. Their detection and characterization by the usual NDT techniques have been achieved through numerous laboratory studies, which were conducted in order to determine the performance and limitations of NDT. Pulled tube examination in a hot laboratory was very valuable, for both NDT and fracture mechanics aspects. The eddy current bobbin coil probe, used for multipurpose inspection of tubes, allows the detection of IGA-SCC at the tube support plate elevation. In France, the use of rotating probes is not required for that type of degradation, since the repair criterion is based on bobbin coil results only. The bobbin coil is also used for detection of IGSCC occurring in free spans, within sludge deposits. The eddy current rotating probe allows, in that case, characterization of main cracks. Concerning the outer diameter initiated circumferential cracks which occur at the top of the tube sheet, only the rotating probe is used. An ultrasonic (UT) inspection was performed several times, in order to obtain information on UT capabilities. The goal of tube inspection is obviously knowledge of the status of steam generators, but also to follow up degradations and to estimate their revolution, and to verify the beneficial effect of some corrective measures, e.g. boric acid injection.     

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.     

SG LOCA摇晃动力响应数值分析

经过合理的简化与等效处理，建立了国内某3代核电站的蒸汽发生器（SG）非线性有限元模型，将其与反应堆冷却剂环路（RCL）串联，开展了SG失水事故（LOCA）摇晃动力响应数值分析，得到了作用在SG传热管上的应力极值及其随管径的变化规律，并获得了作用在上部支承上的载荷。将本文方法与传统解耦法进行对比，结果表明:SG的解耦对摇晃动力响应有较大影响，应采用与RCL耦联的计算方式。      

Native defect properties in β-SiC: Ab initio and empirical potential calculations

There is considerable ambiguity regarding the formation of native defects and their clusters in silicon carbide (SiC), since different empirical potentials give different results, particular for the stability of interstitial configurations. Density functional theory (DFT) is used to study the formation and properties of native defects in β-SiC. The DFT results are compared with those calculated by molecular dynamics (MD) simulations using the Tersoff potentials, with modified cut-off distances and parameters obtained from the literature. The formation energy of vacancies and antisite defects obtained by DFT calculations are in good agreement with those given by the Tersoff potential, regardless of the cut-off distances, but for interstitials there is a disparity between the two methods, depending on the cut-off distances used in the Tersoff potential. The present results provide guidelines for evaluating the quality and fit of empirical potentials for large-scale simulations of irradiation damage (displacement cascades) and point defect migration (recombination or annealing) in SiC.     

Development of a hybrid ECT sensor for JSFR SG double-wall tubes

A new design has been adopted for the steam generator (SG) tubes of the Japan Sodium-cooled Fast Reactor (JSFR) using double-wall tubes. This paper estimates and assesses the effectiveness of detecting defects in SG double-wall tubes of the JSFR by using combined high-frequency eddy current testing (ECT) and low-frequency remote field eddy current sensors. We confirm that the proposed hybrid ECT sensor is highly sensitive to small defects, fatigue cracks, and other defects even when located under support plates of tubes. The parameters of the hybrid ECT sensor are designed and optimized to detect small defects using accurate numerical simulations based on the finite element method, using an in-house developed code. The sensitivity and high performance of the hybrid ECT sensor was validated with experimental measurements.     

基于RELAP5的大功率非能动核电厂SGTR事故分析研究

本文应用RELAP5/mod3.3程序对大功率非能动核电厂进行建模,开展了蒸汽发生器传热管破裂事故(SGTR)分析研究,研究就事故造成的最大质量释放和破损SG最大水体积两种工况分别进行了计算。通过对两种工况计算结果的分析,发现虽然在不同工况条件下,系统参数变化和事故发展序列存在一定差异,但总体来讲,在SGTR事故过程中即使操纵员不干预,大功率非能动核电厂保护系统和非能动设计措施将会触发自动的响应措施,可终止蒸汽发生器(SG)传热管的泄漏,并将反应堆冷却剂系统(RCS)稳定在安全状态,能够防止SG发生满溢和自动降压系统动作,最终使放射性后果在可接受剂量水平限值范围内。     

Object Kinetic Monte Carlo calculations of irradiated Fe-Cr dilute alloys: The effect of the interaction radius between substitutional Cr and self-interstitial Fe

Object Kinetic Monte Carlo models allow for the study of the evolution of the damage created by irradiation to time scales that are comparable to those achieved experimentally. Therefore, the essential Object Kinetic Monte Carlo parameters can be validated through comparison with experiments. However, this validation is not trivial since a large number of parameters is necessary, including migration energies of point defects and their clusters, binding energies of point defects in clusters, as well as the interaction radii. This is particularly cumbersome when describing an alloy, such as the Fe-Cr system, which is of interest for fusion energy applications. In this work we describe an Object Kinetic Monte Carlo model for Fe-Cr alloys in the dilute limit. The parameters used in the model come either from density functional theory calculations or from empirical interatomic potentials. This model is used to reproduce isochronal resistivity recovery experiments of electron irradiated dilute Fe-Cr alloys performed by Abe and Kuramoto. The comparison between the calculated results and the experiments reveal that an important parameter is the capture radius between substitutional Cr and self-interstitial Fe atoms. A parametric study is presented on the effect of the capture radius on the simulated recovery curves.     

Sensitivity analysis of the MASLWR helical coil steam generator using TRACE

Accurate simulation of transient system behavior of a nuclear power plant is the goal of systems code calculations, and the evaluation of a code's calculation accuracy is accomplished by assessment and validation against appropriate system data. These system data may be developed either from a running system prototype or from a scaled model test facility, and characterize the thermal hydraulic phenomena during both steady state and transient conditions. The identification and characterization of the relevant thermal hydraulic phenomena, and the assessment and validation of thermal hydraulic systems codes, has been the objective of multiple international research programs. The validation and assessment of the best estimate thermal hydraulic system code TRACE against the Multi-Application Small Light-Water Reactor (MASLWR) Natural Circulation (NC), helical coil Steam Generator (SG), Nuclear Steam Supply System (NSSS) design is a novel effort, and is the topic of the present paper. Specifically, the current work relates to the assessment and validation process of TRACE code against the NC database developed in the OSU-MASLWR test facility. This facility was constructed at Oregon State University under a U.S. Department of Energy grant in order to examine the NC phenomena of importance to the MASLWR reactor design, which includes an integrated helical coil SG. Test series have been conducted at this facility in order to assess the behavior of the MASLWR concept in both normal and transient operation and to assess the passive safety systems under transient conditions. In particular the test OSU-MASLWR-002 investigated the primary system flow rates and secondary side steam superheat, used to control the facility, for a variety of core power levels and Feed Water (FW) flow rates. This paper illustrates a preliminary analysis, performed by TRACE code, aiming at the evaluation of the code capability in predicting NC phenomena and heat exchange from primary to secondary side by helical SG in superheated condition and to evaluate the fidelity of various methods to model the OSU-MASLWR SG in TRACE. The analyses of the calculated data show that the phenomena of interest of the OSU-MASLWR-002 test are predicted by the code and that one of the reasons of the instability of the superheat condition of the fluid at the outlet of the SG is the equivalent SG model used to simulate the different group of helical coils. The SNAP animation model capability is used to show a direct visualization of selected calculated data.     

Severe accident analysis for a typical PWR using the MELCOR code

Using the MELCOR code, we simulated and analyzed a severe accident at a Chinese pressurized reactor 1000-MW (CPR1000) power plant caused by station blackout (SBO) with failure of the steam generator (SG) safety relief valve (SRV). The CPR1000 response and results for three different scenarios were analyzed: (i) seal leakage and an auxiliary feed water (AFW) supply; (ii) no seal leakage or AFW supply; and (iii) seal leakage but no AFW supply. The results for the three scenarios are compared with those for a simple SBO accident. According to our calculations, the SG SRV stuck in the open position would greatly accelerate the sequence for a severe accident. For an SBO accident with the SRV stuck open without seal leakage or an AFW supply, the pressure vessel would fail at 9576 s and the containment system would fail at 124,000 s. If AFW is supplied, pressure vessel failure would be delayed nearly 30000 s and containment failure would delay at least 50000 s. When seal leakage exists, pressure vessel failure is delayed about 50 s and containment failure time would delay about 30000 s. The results will be useful in gaining an insight into the detailed processes involved and establishing management guidelines for a CPR1000 severe accident.     

核电蒸汽发生器接管嘴外表面裂纹应力强度因子计算

核电蒸汽发生器（SG）接管嘴处由于其结构的特殊性,易在制造及服役过程中产生缺陷。为评价该处缺陷的安全性,需要工程可用的应力强度因子解。本文以核电SG接管嘴外表面裂纹为研究对象,采用有限元方法（FEM）及RSE-M规范计算获得了不同方向及尺寸裂纹在内压、弯矩和温度载荷下的等效应力强度因子值,并分析了不同载荷作用下等效应力强度因子在裂纹前沿的分布规律。将计算结果与RSE-M规范的直管应力强度因子解进行比较,发现RSE-M规范的直管应力强度因子计算方法可保守地应用于SG接管嘴处裂纹,并且随着裂纹深度的增加保守度增大。为实现SG接管嘴处缺陷安全的准确评价,基于有限元计算和RSE-M影响系数法给出了适用于SG接管嘴外表面裂纹的应力强度因子计算方法,该方法可以为SG的设计与维护提供指导。      

Chemical cleaning of PWR steam generators with a low temperature process

Corrosion products generated in the steam, feedwater and condensate systems of a PWR will be transferred by the feedwater into the secondary side of the steam generators (SG). Up to several hundred kilograms of deposits may collect on the surfaces of the SG. These deposits not only reduce the efficiency of the SG by deterioration of the heat transfer, but also cause an acceleration of the corrosion of the SG tubes and blockage of support plate passages.The chemical removal of the corrosion products opens the possibility to eliminate causes of defects on heat transfer tubes by removing the corrosion products and trapped impurities.A low temperature cleaning process, based on the EPRI developed SGOG iron and copper solvents, was applied in November 1990 to remove the hard deposits from the tube sheets of a two-loop plant. The sludge containing over 60% copper was removed with the application of one iron removal step and several copper removal steps. Over 95% of the available sludge was removed. The corrosion of the unalloyed and low alloy materials was extremely low. The Incoloy 600 tubes showed no corrosion.In addition aspects of crevice cleaning at elevated temperatures are mentioned.     

Predicting growth rate of wall thinning for severely degraded SG tubes using a statistical methodology

Based on the requirements of TSTF-449 or NEI 97-06, operational assessment (OA) should be performed to guarantee the steam generator (SG) tube integrity. OA is a forward looking evaluation of the SG tube conditions. One of main evaluations for the OA is to estimate the growth rate of tube degradation prior to the next SG tube inspection. Therefore, the majority of this paper is to predict the growth rate of wall thinning for the SG tubes by way of a statistical methodology. The wall thinning of degraded SG tubes predicted by the present model agrees well with the plant measured one. The relative errors between the predictions and measurements are less than 10%. In addition, the present model would over-predict the wall thinning in most cases, revealing that this methodology could provide a useful and conservative tool for the PWR plant staff to execute the OA for SG tubes.     

Vibration analysis of a dummy fuel rod continuously supported by spacer grids

The fuel rods in the pressurized water reactor are continuously supported by a spring system called a spacer grid (SG), which is one of the main structural components for the fuel rod cluster (fuel assembly). The fuel rods have a vibration behavior within the reactor due to coolant flow. Since the vibration, which is called flow-induced vibration, can wear away the surface of the fuel rod, it is important to understand its vibration characteristics. In this paper, a modal testing and a finite element (FE) analysis using ABAQUS on a dummy fuel rod continuously supported by Optimized H Type (OHT) and New Doublet (ND) spacer grids are performed to obtain the vibration characteristics such as natural frequencies and mode shapes and to verify the FE model used. The results from the test and the FE analysis are compared according to modal assurance criteria values. The natural frequency differences between the two methods as well as the mode comparison results for the rod with the OHT SG are better than those with the ND SG. That is, in the case of the ND grid model using beam-spring elements, there was a large discrepancy between the two methods. Thus, we tried to modify the FE model for the ND SG considering the contact phenomena between the fuel rod and the SG. The results of the new model showed a good agreement with the experiment compared with those of a beam-spring model.