In-service monitoring of piping and vessels in nuclear power plants: report on BMU Study SR 2218

The BMU-Study SR 2218 is made with a view to describing and assessing fatigue monitoring as commonly applied today to piping and vessels in nuclear power plants. First, the fundamentals with regard to strain, instrumentation and calculation of fatigue are compiled and the current regulations listed. With reference to the available literature, own experience and a survey conducted among operators and manufacturers of the now common types of installations, the points of measurement, the methods of measuring, the acquisition systems for measured data and the methods of analysis are described, as are the results obtained from the measurements so far. After a careful review of these results, proposals are presented for improving the acquisition of measured data and, most important of all, for analysis and assessment. In the nuclear power plants in Germany those areas which are of relevance to fatigue have been identified by means of temperature measurements taken in the recent past in the various types of installations. Many thorough analyses have been made on the basis of transients measured in order to assess the relevance of the areas in terms of fatigue. The degrees of fatigue established are normally D=0.2–0.3, that is <D=0.5 for loads occurring in service. It can be stated, in summary, that the degrees of fatigue obtained from fatigue monitoring systems are more realistic than those obtained from the approach adopted in the design stage. As the requirements of the nuclear power plants are specific and varied, there can be no universal and flexible system that is adequate for all applications. That is why it will always be necessary to find solutions for each individual case. This paper gives an overview about the content and the results of the study.     

Implementation of finite element based fatigue monitoring system at Heavy Water Plant Kota

A finite element based fatigue monitoring system has been developed for on-line monitoring of fatigue degradation of components used in various plants. The system can take care of the fluctuations of the process fluid temperature, pressure and flow rate. This can also account for the system induced loads such as axial forces and bending moments. The system converts the plant transients to temperature/stress responses using finite element method, computes the fatigue usage factor and updates the information continuously. This system is capable of analysing several components of a plant using a P.C. 486. As a part of life extension programme of Heavy Water Plant Kota, this system has been installed for monitoring fatigue degradation of important components. This system is successfully under operation since mid 1996.     

Research and developments of guidelines for thermal load modeling

In order to realize reliable and economical plants, Japanese commercialized fast reactors adopt innovative component design such as simplification of a reactor vessel, integration of intermediate heat exchangers with primary pumps, shortening of pipes and reduction of loop numbers. Thermal load, one of the principal loads in fast reactor components, becomes critical in those components. For realistic evaluation and mitigation of thermal loads “Interim Guidelines for Thermal Load Modeling” were developed. These guidelines are referred from “Elevated Temperature Structural Design Guide for Commercialized Fast Reactor (FDS)”.Current guidelines deal with system thermal transient load and thermal striping load. As for system thermal transient load, two kinds of modeling method were provided. Concerning thermal striping load, thermal fatigue evaluation method was developed considering attenuation effects of thermal stress related to frequency of temperature fluctuation.Interpretation documents of the guidelines and exemplars of application problems were also provided to support designers.     

Fatigue — determination of a more realistic usage factor

The ability to use a suitable counting method for determining the stress range spectrum in elastic and simplified elastic-plastic fatigue analyses is of crucial importance for enabling determination of a realistic usage factor. Determination of elastic-plastic strain range using the K_e factor from fictitious elastically calculated loads is also important in the event of elastic behaviour being exceeded. This paper thus examines both points in detail. A fatigue module with additional options, which functions on this basis is presented. The much more realistic determination of usage factor presented here offers various economic benefits depending on the application.     

An assessment of acoustic emission for nuclear pressure vessel monitoring

Recent research has greatly improved our understanding of the basic mechanisms of deformation and fracture that generate detectable acoustic emission signals in structural steels. A critical review of the application of acoustic emission (AE) to the fabrication, proof testing and in-service monitoring of nuclear pressure vessels is presented in the light of this improved understanding. The detectability of deformation and fracture processes in pressure vessel steels is discussed, and recommendations made for improving source location accuracy and the development of quantitative source assessment techniques.

Published data suggest that AE can make an important contribution to weld fabrication monitoring, and to the detection of defects in lower toughness materials during vessel proof testing. In high toughness materials, however, the signals generated during ductile crack growth may frequently be too weak for reliable detection. The feasibility of AE for continuous monitoring has not yet been adequately demonstrated because of high background noise levels and uncertainty about AE signal strengths from the defect growth processes that occur in service. In-service leak detection by AE shows considerable promise.

It is recommended that further tests are carried out with realistic defects, and under realistic conditions of loading (including thermal shock and fatigue) and of environment.     

Electrical connectors for blanket modules in ITER

Blanket electrical connectors (E-straps, ES) are low-impedance electrical bridges crossing gaps between blanket modules (BMs) and vacuum vessel (VV). Similar ES are used between two parts on each BM: the first wall panel (FW) and shield block (SB). The main functions of E-straps are to: (a) conduct halo currents intercepting some rows of BM, (b) provide grounding paths for all BMs, and (c) operate as electrical shunts which protect water cooling pipes (branch pipes) from excessive halo and eddy currents. E-straps should be elastic enough to absorb 3-D imposed displacements of BM relative VV in a scale of ±2 mm and at the same time strong enough to not be damaged by EM loads. Each electrical strap is a package of flexible conductive sheets made of CuCrZr bronze. Halo current up to 137 kA and some components of eddy currents do pass through one E-strap for a few tens or hundreds milliseconds during the plasma vertical displacement events (VDE) and disruptions. These currents deposit Joule heat and cause rather high electromagnetic loads in a strong external magnetic field, reaching 9 T. A gradual failure of ES to conduct Halo and Eddy currents with low enough impedance gradually redistributes these currents into branch pipes and cause excessive EM loads. When branch pipes will be bent so much that will touch surrounding structures, the Joule heating in accidental electrical contact spots will cause local melting and may lead to a water leak.The paper presents and compares two design options of E-straps: with L-shaped and Z-shaped elastic elements. The latter option was developed in 2012 on the basis of more thoughtful analysis of bi-directional cyclic loading conditions influencing a fatigue lifetime. Detail comparative simulations of current and field patterns and subsequent analysis of the fatigue strength and technological assessment allowed make a final choice for the E-strap design in ITER.     

Leak area and leakage rate prediction for probabilistic risk assessment of concrete containments under severe core conditions

The fundamental gap in knowledge for estimating release for probabilistic risk assessment of concrete containments subject to beyond design basis loads is in estimating leak areas and leakage rates. By evaluating the available literature and carefully studying the test results, several generic rules are postulated for leak areas and leakage rates of concrete containments. These rules, coupled with theory-based leakage flow equations and empirically-based crack roughness constants, provide a realistic estimate of leak rates through liner tears in concrete containments.     

核电厂金属疲劳监测过程中环境影响疲劳评价方法研究

轻水反应堆中金属部件的环境影响疲劳寿命（EAF）问题近年来一直是国内外核安全设备研究和监管所关注的重要问题。本文对日本和美国的相应试验研究数据进行了深入分析,对环境影响疲劳修正系数F_en的计算公式进行了分析讨论,总结出不同金属材料F_en计算结果的保守性以及关键影响因素的敏感性。结合我国的实际情况,借助F_en修正计算公式,给出了一种核电厂金属疲劳监测过程中开展环境影响疲劳的评价方法。结果表明：现阶段可基于核电厂监测的真实瞬态数据采用F_en修正公式对环境影响因子进行估算,对于同类型的金属材料,不同试验环境拟合的F_en公式计算结果非常接近,但EAF问题在金属部件疲劳寿命评估过程中不可忽视。     

Fluid-structure interaction analysis for pressurizer surge line subjected to thermal stratification

Serious mechanical damages such as cracks and plastic deformations due to excessive thermal stress caused by thermal stratification have been experienced in several nuclear power plants. In particular, the thermal stratification in the pressurizer surge line has been addressed as one of the significant safety and technical issues. In this study, a detailed unsteady computational fluid dynamics (CFD) analysis involving conjugate heat transfer analysis is performed to obtain the transient temperature distributions in the wall of the pressurizer surge line subjected to stratified internal flows either during out-surge or in-surge operation. The thermal loads from CFD calculations are transferred to the structural analysis code which is employed for the thermal stress analysis to investigate the response characteristics, and the fatigue analysis is ultimately performed. In addition, the thermal stress and fatigue analysis results obtained by applying the realistic temperature distributions from CFD calculations are compared with those by assuming the simplified temperature distributions to identify some requirements for a realistic and conservative thermal stress analysis from a safety point of view.     

Assessment of actual loading conditions of steam generators and heat exchangers according to probabilistic criteria

Classical design codes are based on nominal conditions, e.g. pressure, temperature, material strength and tube diameter for tubes under pressure; local loads, hot spots, deviations in strength and geometry, etc. are compensated by the use of safety factors. More recent methods — in particular those for the assessment of steam generators and heat exchangers for nuclear applications — call for designs to meet actual loading conditions. Thus, detailed investigations are required to determine realistic operational characteristics of the components to be assessed. However, a worst case treatment on the basis of the above criteria would yield an over-conservative design which in special cases might even be worse than the classical one, e.g. in the case of heated tubes due to increased thermal stresses. As a consequence, with all the data or the spectra of data available, the assessment has to be executed according to probabilistic criteria resulting in a design with an acceptably low and, above all, known failure rate adjusted to fit into the overall plant concept.In components to which the above criteria have been applied before — mainly pressure vessels — the design variables pressure and temperature are generally considered uncorrelated and the temperature is in most cases assumed to be constant. For heat exchangers and steam generators these simplications are not acceptable on principle. On the contrary, even deviations in geometry (like tube tolerances) result in flow rate deviations with corresponding deviations in temperature and pressure. Thus a new procedure for assessment according to actual loading conditions, e.g. by the application of probabilistic criteria, is proposed. This procedure can be used for heat exchangers and steam generators. With this new method, components of nuclear power plants like steam generators and heat exchangers may be designed to meet the low failure probability required for satisfying the overall reliability concept of the plant.     

Extreme value risk analysis in the structural design of reactor components

The tendency to confuse “uncertainties” associated with design assumptions and parameters and compensated by the safety factor with objective ‘risks of failure’ implicit in the design, has been characteristic of the approach to probability-based structural design on all levels. However, a clear differentiation between uncertainty and risk is required to remove the lack of correlation between design safety analysis and risk analysis implicit in the present approach to the design of major structural and mechanical components of nuclear reactors as well as other structures.In a recent paper [5] the author has used the definition of the safety factor as a random variable (distribution of a quotient) to construct a probability model that justifies the introduction of the asymptotic distributions of extreme values as the physically relevant distributions of the design parameters governing ultimate load failure on which a realistic risk assessment can be based. Realistic reliability and risk assessment of reactor components subject to fatigue and creep, i.e. design conditions that exceed in practical importance that of ultimate load failure, can be based on the use of the third asymptotic distribution of smallest values.In the case of structural components working under complicated conditions it becomes necessary to perform full-scale tests reproducing, as closely as possible, the anticipated operational and, whenever necessary, critical limiting conditions to be provided in the design as well as in the associated reliability and risk assessment. The economic necessity of keeping the number of such full-scale tests to a minimum which, in the case of larger components, is usually a single or very small number of tests, raises the problem of integration of the test results into the framework of a reliability and risk assessment.     

Intelligent system for transient data collection and fatigue monitoring of pressurised water reactors nuclear steam supply system

Electricité de France (EDF), the French national electricity company, is operating 54 standardised pressurised water reactors. This about 500 reactor-years experience in nuclear stations operation and maintenance area has allowed EDF to develop its own strategy for monitoring of age-related degradations of NPP systems and components relevant for plant safety and reliability. After more than fifteen years of experience in regulatory transient data collection and seven years of successful fatigue monitoring prototypes experimentation, EDF decided to design a new system called SYSFAC (acronym for SYstème de Surveillance en FAtigue de la Chaudière) devoted to transient logging and thermal fatigue monitoring of the reactor coolant pressure boundary. The system is fully automatic and directly connected to the on-site data acquisition network without any complementary instrumentation. A functional transient detection module and a mechanical transient detection module are in charge of the general transient data collection. A fatigue monitoring module is aimed towards a precise surveillance of five specific zones particularly sensible to thermal fatigue. After a first step of preliminary studies, the industrial phase of the SYSFAC project is currently going on, with hardware and software tests and implementation. The first SYSFAC system will be delivered to the pilot power plant by the beginning of 1996. The extension to all EDF’s nuclear 900 MW is planned after one more year of feedback experience.     

Probability-based assessment and maintenance of welded joints damaged by fatigue

This paper presents a probabilistic reliability assessment procedure for steel components damaged by fatigue. The study combines the structural reliability theory with a maintenance strategy. The fatigue assessment model is based on a modelisation of the fatigue phenomenon issued from the principles of fracture mechanics theory. The safety margin includes the crack growth propagation and allows to treat fatigue damage in a general manner. Damaging cycles and non damaging cycles are distinguished. The sensitivity study of the different parameters shows that some variables can be taken as deterministic. Applications are made on a welded joint ‘stiffener/bottom-plate' of a typical steel bridge. The model is then used for taking into account inspection results. Non destructive inspection (NDI) techniques are also used for updating failure probabilities. The results show their ability to be inserted in a maintenance strategy for optimizing the next inspection time. This has led to define cost functions related to the total maintenance cost; this cost is then minimized for determining the optimal next inspection time. An example of welded joint cracked by fatigue highlights the different concepts. The approach presented in the paper is not only restrained to fatigue problems, but can be applied to a wide variety of degrading phenomena.     

A computational fatigue analysis of cyclic thermal shock in notched specimens

Thermal shock induced fatigue plays a role in the assessment of the lifetime of different components in the primary cooling circuit of a nuclear plant. In spite of the implementation of substantial and costly safety factors, a few, unexpected cases of fatigue failure have occurred. Here we report on a laboratory experiment which mimics the thermal loading observed in such components. A finite element thermal stress analysis using a calibrated, elasto-plastic, combined kinematic-isotropic cyclic hardening material model is presented. The distribution of transient stresses and strains in the specimens subjected to cyclic thermal shock, are used to predict the number of cycles to crack initiation with a fatigue curve that has been calibrated experimentally with data from equivalent specimens under pure mechanical fatigue. Our results indicate that cyclic thermal shock induced ratcheting occurs locally near the tip of the notch in the specimens, and the potential impact on the number of cycles to crack initiation is explored.     

Dynamic qualification of complex structural components of nuclear power plants

The safety requirements and the lack of accessibility for any future repair, impose the design requirement that the integrity of reactor components of nuclear power plants be assured for the lifetime of the plant. To meet this design requirement it is essential to qualify the component, i.e. prove its capability to perform the design function for the design life. In performing its design function, the component is subjected to both static and dynamic loads. The qualification for static loads is rather simple and reliable, but qualification for dynamic loads is complex and often uncertain. This is because analytical tools are often inadequate for a realistic dynamic qualification and exact structurally simulated experimental models are almost always difficult to build. In such a situation, methods using tests on simple experimental set-ups supplemented by conservative analytical back-ups must be evolved. This paper highlights the intricacies involved in the conservative dynamic qualification of the complex components by considering the example of the moderator sparger tube. This component is a perforated tube submerged in water and excited by flow. For such a case, a completely analytical or a totally experimental qualification is not possible. This paper describes a procedure by which the required dynamic characteristics such as added mass, damping and fluid forces are generated from simple experiments and the component is qualified by analysis using these data.     

Conducting thermomechanical fatigue test in air at light water reactor relevant temperature intervals

In Light Water Reactors (LWR), many structural components are made of austenitic stainless steels (SS). These components are subject to extreme conditions, such as large temperature gradients and pressure loads during service. Hence, the fatigue and fracture behavior of austenitic SS under these conditions has evoked consistent interest over the years. Most studies dealing with this problem in the past, investigated the isothermal fatigue (IF) condition, which is not the case in the service, and less attention has been paid to thermomechanical fatigue (TMF). Moreover, the existing codes of practice and standards for TMF testing are mainly derived from the high temperature TMF tests (T_mean > 400 °C). This work presents the development of a facility to perform TMF tests under LWR relevant temperature interval in air. The realized testing parameters and tolerances are compared with the recommendations of existing codes of practice and standards from high temperature tests. The effectiveness of the testing facility was verified with series of TMF and IF tests performed on specimens made out of a commercial austenitic SS TP347 pipe material. The results revealed that the existing tolerances in standards are quite strict for the application of lower temperature ranges TMF tests. It was found that the synchronous, in-phase (IP) TMF tested specimens possess a higher lifetime than those subjected to the asynchronous, out-of-phase (OP) TMF and IF at T_max in the investigated strain range for austenitic SS. Nevertheless, the fatigue lifetime of all the test conditions was similar in the engineering scale.     

Progress on NSTX center stack upgrade

The national spherical torus experiment (NSTX) will be upgraded to provide increased toroidal field, plasma current and pulse length. This involves the replacement of the so-called center stack, including the inner legs of the toroidal field (TF) coil, the Ohmic heating (OH) coil, and the inner poloidal field (PF) coils. In addition the increased performance of the upgrade requires qualification of remaining existing components for higher loads. Initial conceptual design efforts were based on worst-case combinations of possible currents that the power supplies could deliver. This proved to be an onerous requirement and caused many of the outer coils support structures to require costly heavy reinforcement. This has led to the planned implementation of a digital coil protection system (DCPS) to reduce design-basis loads to levels that are more realistic and manageable. As a minimum, all components must be qualified for the increase in normal operating loads with headroom. Design features and analysis efforts needed to meet the upgrade loading are discussed. Mission and features of the DCPS are presented.     

General review of recent activities in acoustic emission technology in Italy

This paper describes the present situation in Italy in the field of Acoustic Emission researches and applications.Information on the level of instrumentation development is given. Both multichannel and multiparameter systems for large structure examination in real time and data logging systems for continuous surveillance purposes are considered.The expertise accumulated in the application of AE to pressure vessel examination during hydrotest is mentioned, this being oriented to pressure components of conventional power stations and chemical plants.Particular attention is recently paid to mechanical fatigue tests. These were conducted on intermediate PWR nuclear pressure vessel, reduced scale offshore nodes and full scale prototype aircraft.A considerable activity has been carried out on application of AE technique to the detection of fluid leakages in power plant components. Both intrusive and non-intrusive methods have been considered. Many boilers and pre-heaters of thermal power plants have been instrumented for an on-line AE monitoring during operation. The problem of the loose part monitoring has been also considered.Several basic researches for material characterization by AE have been also conducted. Different composite material, carbon and austenitic steels, metal alloys have been studied.     

New generation of monitoring systems with on-line diagnostics

For more than 15 years, systems for monitoring the integrity of the primary system and for the diagnosis of nuclear power plant components have been manufactured by Siemens (KWU). These systems record the vibrational behaviour of reactors and their internals and of principal components of the primary system (SÜS), indicate the presence of loose parts or parts that have become detached within the pressure boundary of the primary system (KÜS), detect and locate leaks in piping, tanks and vessels (ALÜS, FLÜS), and evaluate the fatigue of nuclear power plant components as a result of the pressure and thermal stresses to which they are subjected (FAMOS).     

Application of advanced ultrasonic test techniques coupled with fatigue and leakage monitoring to assure integrity of BWR feedwater nozzles

As a result of feedwater nozzle cracking observed in Boiling Water Reactor (BWR) plants, several design modifications were implemented to eliminate the thermal cycling that led to crack initiation. BWR plants with these design changes have successfully operated for over ten years without any recurrence of cracking. To provide further assurance of this, the U.S. Nuclear Regulatory Commission (NRC) issued NUREG-0619, which established periodic ultrasonic testing (UT) and liquid penetration testing (PT) requirements. While these inspections are useful in confirming structural integrity, they are time consuming and can lead to significant radiation exposure to plant personnel. In particular, the PT requirement poses problems since it is difficult to perform the inspections with the feedwater sparger in place and also leads to additional personnel exposure. Clearly, an inspection and monitoring program that eliminates the PT examination and still verifies the absence of surface cracking would be extremely valuable in limiting costs as well as radiation exposure. This paper describes a program involving the application of advanced UT techniques coupled with fatigue and leakage monitoring to assure integrity of BWR feedwater nozzles. The inspection methods include: (1) scanning with optimized transducers and techniques from the outside vessel wall surface to inspect the nozzle inner radius region, and (2) scanning from the nozzle forging outside-diameter to inspect the nozzle bore region. Methods of analyzing the data using 3-D graphics displays have been developed that show crack location, size, and maximum depth of penetration into the nozzle inner surface. These techniques have been developed to the point where they are now considered a reliable alternative to the liquid penetrant requirements of NUREG-0619. An important supplement to the UT program is the use of automated fatigue, leakage and crack growth monitoring to verify the absence of cracking. This approach provides for a continuous assessment of the integrity of the nozzle structure by tracking the actual fatigue duty, measuring thermal sleeve bypass leakage and performing crack growth predictions based on actual thermal duty. Collectively, the monitoring and inspection program provides technically sound assurance of nozzle integrity and a firm basis for plant operational planning.