Case studies of risk-informed inservice inspection of nuclear piping systems

In the present paper, a probabilistic failure analysis is used to find failure probabilities of piping segments, and a probabilistic risk assessment model is employed to obtain risks to a nuclear power plant should these failures occur. The multiplication of the piping failure probability and the consequence for that particular failure results in the risk contribution of the pipe. The degrees of risk for different piping segments can then be ranked, and their results can be used as the basis for planning a risk-informed inservice inspection program. Numerical studies are offered with special emphases on: (1) the status and experience with RI-ISI applications in Taiwan; (2) the comparison of risk-rankings performed with three different methods developed in the US; (3) aspects of the probabilistic fracture mechanics calculation including the flaw size distributions and stress corrosion cracking model. The results indicate the proposed method can indeed be adopted for planning a cost effective inservice inspection program.     

Experimental investigation of the failure behavior of notched wall-thinned pipes

The present study performed full-scale pipe tests using 100A Schedule 80 pipe specimens with simulated notched and circular wall thinning to investigate the failure behavior of notched wall-thinned pipes. The tests were conducted under both monotonic and cyclic bending moments at a constant internal pressure of 10 MPa at room temperature. The failure pattern, load carrying capacity, deformation ability, and fatigue strength of the notched wall-thinned pipes were evaluated by comparing results to those of circular wall-thinned pipes. The investigation showed that the effect of the type of thinning on the failure behavior was more sensitive under cyclic loading conditions than under monotonic loading conditions. The load carrying capacity of pipes with notched wall thinning was approximately the same or slightly less than that of pipes with circular wall thinning when the thinning area was subjected to tensile stress. However, when the thinning area was subjected to compressive stress, the load carrying capacity of pipes with notched wall thinning was greater than that of pipes containing circular wall thinning. The deformation ability and fatigue strength increased proportionally with the axial length of the thinning defect, and thus these properties were significantly reduced in notched wall-thinned pipes.     

Probabilistic fracture failure analysis of nuclear piping containing defects using R6 method

Failure analysis of in-service nuclear piping containing defects is an important subject in the nuclear power plants. Considering the uncertainties in various internal operating loadings and external forces, including earthquake and wind, flaw sizes, material fracture toughness and flow stress, this paper presents a probabilistic assessment methodology for in-service nuclear piping containing defects, which is especially designed for programming. A general sampling computation method of the stress intensity factor (SIF), in the form of the relationship between the SIF and the axial force, bending moment and torsion, is adopted in the probabilistic assessment methodology. This relationship has been successfully used in developing the software, Safety Assessment System of In-service Pressure Piping Containing Flaws (SAPP-2003), based on a well-known engineering safety assessment procedure R6. A numerical example is given to show the application of the SAPP-2003 software. The failure probabilities of each defect and the whole piping can be obtained by this software.     

Probabilistic fracture mechanics analysis of nuclear structural components: a review of recent Japanese activities

This paper describes a review of recent Japanese activities on probabilistic fracture mechanics (PFM) analyses. Japan Atomic Energy Research Institute (JAERI) has sponsored research committees on PFM organized by Japan Society of Mechanical Engineers (JSME) and Japan Welding Engineering Society (JWES) for more than 10 years. The purpose of the continuous activity is to establish standard procedures for evaluating failure probabilities of Japanese nuclear structural components such as PV&P and steam generator tube, combining the state-of-the-art knowledge on structural integrity of nuclear structural components and modern computer technology such as parallel processing. This paper shows two topics of the newest results of JWES committee, PFM analysis of aged reactor pressure vessel considering embedded cracks and PFM analysis of piping considering seismic loading, and one topic by JAERI itself, development of PTS analysis code for transient loading (PASCAL).     

Development of expert system for nuclear piping integrity

The objective of this paper is to develop a nuclear piping integrity expert system (NPIES) for nuclear piping integrity. This paper describes the structure and the development strategy of the NPIES system. The NPIES system consists of five parts; user interface, database, knowledge base, expert and integrity evaluation parts. The user interface part is developed to connect the user and the NPIES system effectively. In the database part, nuclear piping material properties are stored; the unknown material properties are stored in the knowledge base part. Various rules for inferring material properties are stored in the knowledge base part. The most appropriate evaluation method for a given input condition is recommended in the expert part. Finally, the integrity evaluation part is developed for the evaluation of piping integrity effectively.     

Updating of risk-informed ISI programme for Ignalina RBMK-1500 nuclear power plant in Lithuania: Results and challenges

The paper presents a risk-informed in-service inspection (RI-ISI) pilot study project of 300 mm piping at Ignalina nuclear power plant (INPP) RBMK-1500 reactor, located in Lithuania. The RI-ISI study investigates optimal 300 mm piping ISI strategies with respect to risk and required resources. In total 1240 stainless steel welds were analyzed, assuming inter-granual stress corrosion cracking (IGSCC) to be the main damage mechanism. Pipe break frequency was estimated by probabilistic fracture mechanics methods and combined with safety barriers, provided by probabilistic safety assessment (PSA) study.After 3 years of operation, updating of RI-ISI was performed by taking into account new statistical data on pipe defects. Comparison with previous RI-ISI program was performed. The paper includes discussion on uncertainties in the study and robustness of RI-ISI programs.     

管道系统的功能性评定准则研究

管道系统的功能性是不同于管道系统压力边界完整性的一项要求，美国核管理委员会（NRC）提出了管道系统功能性的2种评定准则。为了探讨功能性评定准则的来源以及应用，通过研究经典文献中有关功能性评定准则的内容，阐述了2种评定准则的来历和依据，分析了2种功能性评定准则的特点，指出了使用功能性评定准则的注意事项。通过一个管道系统功能性评定的实例，提出2种功能性评定准则在不同的核电厂设计阶段的应用策略。对于新建的核电厂，尽量使用C级限值来保证管道系统的功能性，如果是已建造的核电厂，则可以用D级限值附加5个条件来保证管道系统的功能性。      

Collapse moment estimation by support vector machines for wall-thinned pipe bends and elbows

The collapse moment due to wall-thinned defects is estimated through support vector machines with parameters optimized by a genetic algorithm. The support vector regression models are developed and applied to numerical data obtained from the finite element analysis for wall-thinned defects in piping systems. The support vector regression models are optimized by using both the data sets (training data and optimization data) prepared for training and optimization, and its performance verification is performed by using another data set (test data) different from the training data and the optimization data. In this work, three support vector regression models are developed, respectively, for three data sets divided into the three classes of extrados, intrados, and crown defects, which is because they have different characteristics. The relative root mean square (RMS) errors of the estimated collapse moment are 0.2333% for the training data, 0.5229% for the optimization data and 0.5011% for the test data. It is known from this result that the support vector regression models are sufficiently accurate to be used in the integrity evaluation of wall-thinned pipe bends and elbows.     

Probabilistic analysis of core shroud cracks

Traditional limit load analysis and fracture mechanics analysis have been applied to evaluate the integrity of the degraded nuclear power plant components. Although these methodologies are generally accepted by the regulatory authorities in the nuclear industry, conservatism introduced by the uncertainties of inspection, material property, crack geometry, applied loading, neutron environment, etc. is recognized to have great impact on the evaluation accuracy. A probabilistic analysis may overcome this shortcoming and reveal some additional insight to the problem. The purpose of the present study is to apply probabilistic methods to analyze the degraded core shroud, and to predict the quantitative risk of the cracked shroud. In the analysis, the loading condition, crack growth rate, material properties and existing defects are all considered random. A sample analytical result shows that, based on some previously observed data and under certain assumptions, the crack-through probability of the studied core shroud is in the order of 10⁻⁷ after 13 cycles of operation. The probability will increase considerably through operation cycles or operation years if no repair action is taken.     

核安全一级高温管道系统结构分析与安全评估方法研究

为解决复杂核安全一级高温管道系统结构分析与评定工程问题，在管道分析软件与核级高温评定规范ASME-NH之间建立了一座桥梁。首先，对管道结构（直管及弯管）在不同载荷作用下的应力状态解析解进行了详细推导分析，并且与有限元数值解进行了误差分析。结果显示，给出的直管及弯管结构应力状态解析解具有很好的准确性。随后，将一维管线力学分析模型与截面三维应力状态解析解相结合，给出了高温管道系统结构分析、评定方法及应用步骤，将ASME-NH-3650规范内容明确化。      

Probability of crack-induced failure in BWR recirculation piping

The Lawrence Livermore National Laboratory (LLNL) has estimated the probability of double-ended guillotine break (DEGB) in the reactor coolant piping of Mark I boiling water reactor (BWR) plants. Two causes of pipe break are considered: crack growth at welded joints and the seismically-induced failure of component supports. For the former a probabilistic fracture mechanics model is used, for the latter a probabilistic support reliability model. This paper describes a probabilistic model developed to account for effects of intergranular stress corrosion cracking (IGSCC). The IGSCC model, based on experimental and field data compiled from several sources, correlates times to crack initiation and crack growth rates for Types 304 and 316NG stainless steel against material-specific ‘damage parameters’ which consilidate the separate effects of coolant environment (temperature, dissolved oxygen content, level of impurities), stress (including residual stress), and degree of sensitization. Application of this model to actual BWR recirculation piping shows that IGSCC clearly dominates the probability of failure in 304SS piping, mainly due to cracks that initiate within a few years after plant operation has begun. Replacing Type 304 piping with 316NG reduces failure probabilities by several orders of magnitude.     

Comparison of critical circumferential through-wall crack lengths in welds between pieces of straight pipes to welds between straight pipes and bends, with and without internal pressure at force- and displacement-controlled bending load

In many research projects methods to calculate critical circumferential through-wall cracks have been developed and verified. During the last years, the differentiation between force- and displacement-controlled loading has been shown to be of significant importance. So it was looked at with more interest in new analytical methods to calculate the critical crack length. Most of the approaches applied in the safety analysis of piping systems assume defect at welds connecting pieces of straight pipes. But in nearly all cases in modern power plants the true position of the welds in the piping system is not correctly represented, as in those systems only few welds connect parts of straight pipes. Most of the connections are situated between pipes and bends, bends with elongated ends, nozzles or T-parts. This paper presents a non-linear finite element (FEM) study covering an essential part of the relevant piping parameters of nuclear power plants primary and secondary system. It compares defects in circumferential welds between straight pipes to those joining pipes to elbows. In the case of displacement controlled loading, e.g. as due to restrained thermal expansion, which is one of the most severe load cases for most of the welds, we find, that the calculated J-integral values, and so the critical crack length are of comparable size. At force-controlled loading the codes require stronger limitations to the allowable forces and moments. In the regime of allowable loads, we find that the critical crack sizes in welds near bends are not significantly longer than the ones connecting straight pipes. In the cases where we have to consider in the safety analysis of piping systems, it is a realistic approach to use the methods accepted for welds between pipes to calculate the critical crack length in welds near bends.     

Low cycle fatigue and ductile fracture for Japanese carbon steel piping under dynamic loadings

Dynamic fracture behavior of circumferentially cracked pipe is important to evaluate the structural integrity of nuclear piping from the viewpoint of the LBB concept under seismic conditions. Fracture tests have been conducted for Japanese carbon steel (STS410) circumferentially through-wall cracked pipes that are subjected to monotonic or cyclic bending loads at room temperature. In the monotonic-loading tests, the maximum load to failure increases slightly with increasing loading rate. The failure cycles can be expressed simply by ratio of the load amplitude to the plastic collapse load. Fracture analysis has been also conducted to model the pipe tests. A new equation for calculating ΔJ for a circumferentially through-wall cracked pipe subjected to bending has been proposed. The failure cycles under cyclic loads are satisfactorily evaluated using an elastic-plastic fracture mechanics parameter ΔJ.     

Trends in the evaluation of the structural integrity of RPVs

The aim of this paper is to review recent trends, improvements and validations of methodologies for the assessment of reactor pressure vessel (RPV) integrity against the risk of leak or catastrophic failure, mainly deriving from the possible presence of crack-like defects at critical locations in the vessel wall.The first part of the work gives an overview of the input parameters, namely loading conditions, material properties and possible crack shape and dimensions, which are needed for a comprehensive fracture analysis of RPVs, discussing recent findings and still open questions about them.The next two sections are concerned with reviews of the presently available fracture approaches, related to both brittle and ductile fracture behaviour, and of probabilistic fracture mechanics methodologies.As conclusion, present limitations of methodologies for evaluation of RPV structural integrity and areas which need further improvements are outlined.     

基于ASME和RSE-M规范的含平面缺陷奥氏体不锈钢核级管道剩余寿命预测方法的数值对比研究

核级管道在加工和安装环节可能存在不同的缺陷。此外,由于核电厂运行条件的影响,管道中可能存在少量缺陷,如裂缝。需要合理预测评估含缺陷管道的剩余寿命,以便安排更换方案,避免对核电厂的效率造成严重影响。本文根据ASME和RSE-M规范,在应力强度因子计算、裂纹扩展分析和裂纹稳定性评价等环节,通过数值对比研究了含有平面缺陷的奥氏体不锈钢核级管道的剩余寿命评估方法,为类似工作提供参考。      

Fuzzy probability on reliability study of nuclear power plant probabilistic safety assessment: A review

Fault tree analysis (FTA) is a graphical model which has been widely used as a deductive tool for nuclear power plant (NPP) probabilistic safety assessment (PSA). The conventional one assumes that basic events of fault trees always have precise failure probabilities or failure rates. However, in real-world applications, this assumption is still arguable. For example, there is a case where an extremely hazardous accident has never happened or occurs infrequently. Therefore, reasonable historical failure data are unavailable or insufficient to be used for statistically estimating the reliability characteristics of their components. To deal with this problem, fuzzy probability approaches have been proposed and implemented. However, those existing approaches still have limitations, such as lack of fuzzy gate representations and incapability to generate probabilities greater than 1.0E-3. Therefore, a review on the current implementations of fuzzy probabilities in the NPP PSA is necessary. This study has categorized two types of fuzzy probability approaches, i.e. fuzzy based FTA and fuzzy hybrid FTA. This study also confirms that the fuzzy based FTA should be used when the uncertainties are the main focus of the FTA. Meanwhile, the fuzzy hybrid FTA should be used when the reliability of basic events of fault trees can only be expressed by qualitative linguistic terms rather than numerical values.     

Proposal of limit moment equation applicable to planar/non-planar flaw in wall thinned pipes under bending

In this paper, a limit bending moment equation applicable to all types of planar and non-planar flaws in wall-thinned straight pipes under bending was proposed. A system to rationally classify the planar/non-planar flaws in wall-thinned pipes was suggested based on experimental observations focused on the fracture mode. The results demonstrate the importance of distinguishing between axial and circumferential long flaws in wall-thinned pipes.     

Reliability analysis of pipe whip impacts

A probability-based approach is presented as the integration of probabilistic methods and deterministic modelling based on the finite element method. An existing finite element software package was linked to an existing probabilistic package to analyse the complex mechanics that occur during the transient non-linear analysis of impact problems. This methodology is applied to a pipe whip analysis of a group-distribution-header, which results from a guillotine break, and subsequent impact with the adjacent building wall; this is a postulated accident for the Ignalina Nuclear Power Plant RBMK-1500 reactors. The uncertainties of material properties, component geometry data and loads were taken into consideration. The probabilities of failure of the impacted header and of the header support-wall were estimated given uncertainties in material properties, geometrical parameters and loading. The software ProFES was used for the probabilistic analysis and the finite element software NEPTUNE for deterministic structural integrity evaluation. The Monte Carlo Simulation, First Order Reliability method and Response Surface method were used in the probabilistic analysis.     

Normalizing the influence of flaw length on failure pressure of straight pipe with wall-thinning

Burst tests using wall-thinned pipe of carbon steel for high-temperature use were conducted in order to examine the influence of length of wall-thinning on burst pressure. Then, three-dimensional elastic-plastic large deformation finite element analyses (EP-FEA) were performed to accurately predict the burst pressure obtained by the tests. The failure pressure corresponding to the burst pressure in tests was defined as the maximum pressure during the analysis including the instability condition after the peak of pressure. The results showed that the failure pressure obtained by EP-FEA agreed well with the experimental results. Finally, failure pressures of wall-thinned pipes with various sizes, thicknesses, flaw lengths and depths were examined by EP-FEA with the same procedure of analysis as validated in this paper. The results showed that, from the standpoint of influence of flaw length on failure pressure, it is preferable to normalize flaw length by pipe mean radius of the unflawed section R rather than by shell parameter (Rt)^0.5, where t is the thickness of the unflawed section.     

Validation of simplified fracture mechanics methods by testing of real components

The evaluation of integrity of structural components is often based on the proof of leak-before-break (LBB). Leak-before-break behaviour in piping constitutes a fail-safe condition. Which means that, during multiplied loading conditions, a defect results at first in a leakage. The crack length which leads to the leakage is smaller than the critical through-wall crack length. Simplified fracture mechanics concepts are used for the demonstration of LBB. For this the conservative, safe calculation of the critical through-wall crack length for ductile failure is necessary. To validate simplified calculation methods for circumferential cracks (flow stress concept (FSC); plastic limit load (PLL)) and for axial cracks (Battelle approach (BMI); Ruiz approach (RUIZ)) all available experiments on real structural components, especially on pipes, were analysed and evaluated by the mentioned simplified methods (approximately 460 experiments). The methods were adapted by application of correction factors, mainly on the flow stress, to result in conservative (safe) and realistic (as near as possible to the experiments) predictions. Depending on method (FSC, PLL, BMI, RUIZ), crack orientation (circumferential and axial cracks) and type of material (ferritic and austenitic material) different definitions of flow stresses were established.