Failure probability assessment of wall-thinned nuclear pipes using probabilistic fracture mechanics

The integrity of nuclear piping system has to be maintained during operation. In order to maintain the integrity, reliable assessment procedures including fracture mechanics analysis, etc., are required. Up to now, this has been performed using conventional deterministic approaches even though there are many uncertainties to hinder a rational evaluation. In this respect, probabilistic approaches are considered as an appropriate method for piping system evaluation. The objectives of this paper are to estimate the failure probabilities of wall-thinned pipes in nuclear secondary systems and to propose limited operating conditions under different types of loadings. To do this, a probabilistic assessment program using reliability index and simulation techniques was developed and applied to evaluate failure probabilities of wall-thinned pipes subjected to internal pressure, bending moment and combined loading of them. The sensitivity analysis results as well as prototypal integrity assessment results showed a promising applicability of the probabilistic assessment program, necessity of practical evaluation reflecting combined loading condition and operation considering limited condition.     

Matrix displacement methods in fracture mechanics analysis of reactor vessels

The technology of fracture mechanics is developing rapidly in response to increased requirements for integrity of engineering structures. It enables structural engineers to evaluate brittle failure resistance of structures within appropriate regimes of temperature, materials and geometry. The evaluation includes the combined effects of material toughness, flaw characteristics, environment and service loadings. Calculations of stress intensity factors associated with the flaws, geometry and applied loading form the basis of fracture analysis and control procedures for reactor vessels.     

Crack initiation and arrest in a pressurized thermal shock test for a model pressure vessel made of VVER-440 reactor pressure vessel steel

A joint pressure vessel integrity research programme involving three partners is being carried out during 1990–1995. The partners are the Central Research Institute of Structural Materials “Prometey” from Russia, IVO International Ltd (IVO) from Finland, and the Technical Research Centre of Finland (VTT). The main objective of the research programme is to increase the reliability of the VVER-440 reactor pressure vessel safety analysis. This is achieved by providing material property data for the VVER-440 pressure vessel steel, and by producing experimental understanding of the crack behaviour in pressurized thermal shock loading for the validation of different fracture assessment methods. The programme is divided into four parts: pressure vessel tests, material characterization, computational fracture analyses, and evaluation of the analysis methods. The testing programme comprises tests on two model pressure vessels with artificial axial outer surface flaws. The first model vessel had circumferential weld seam at the mid-length of the vessel. A special embrittling heat treatment is applied to the vessels before tests to simulate the fracture toughness at the end-of-life condition of a real reactor pressure vessel. The sixth test on the first model led to crack initiation followed by arrest. After the testing phase, material characterization was performed. Comparison of calculated and experimental data generally led to a good correlation, although the work is being continued to resolve the discrepancies between the measured initiation and arrest properties of the material.     

Validation of experimental and computational fracture assessment methods for flawed pressure components

Safety and integrity assessments of pressure boundary components require reliable knowledge of the material property values and the validated experimental and computational analysis methods. To improve the accuracy and validity of the experimental and computational fracture assessment methods, a four year Nordic research programme under the auspices of the Nordic Liaison Committee of Atomic Energy was initiated in 1985 and is now under completion. The main technical objective of the programme was to clarify how catastrophic failure can be prevented in pressure vessels and pipings.Experiments with small fracture mechanics specimens and pressure vessels were performed to validate the computational fracture assessment analysis. Two tests were conducted on a decommissioned full-scale chemical reactor pressure vessel from an oil refinery plant, and were extensively instrumented, e.g. by utilizing a 64-channel acoustic emission monitoring system. The scattering of their material property values were determined by numerous fracture mechanics samples. In addition, as a part of the experimental work, the reactor pressure vessel was repaired by welding after the first test. The repair was carried out without postweld heat treatment and welding was done by applying the temper-bead technique. Residual stresses were measured during and after welding.Different fracture assessment methods were developed and subsequently applied to the tested components. Inter-laboratory round robin programmes with the participation of several laboratories were arranged to examine elastic-plastic finite element calculations and fracture mechanics testing.     

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).     

A quantitative methodology for reactor vessel pressurized thermal shock decision making

The recent operating experience of the Pressurized Water Reactor (PWR) Industry has focused increasing attention on the issue of reactor vessel pressurized thermal shock (PTS). Previous reactor vessel integrity concerns have led to changes in vessel and plant system design and to operating procedures, and increased attention to the PTS issue is causing consideration of further modifications. Events such as excess feedwater, loss of normal feedwater, and steam generator tube rupture have led to significant primary system cooldowns. Each of these cooldown transients occurred concurrently with a relatively high primary system pressure. Consideration of these and other postulated cooldown events has drawn attention to the impact of operator action and control system effects on reactor vessel PTS.A methodology, which couples event sequence analysis with probabilistic fracture mechanics analyses, was developed to identify those events that are of primary concern for reactor vessel integrity. Operating experience is utilized to aid in defining the appropriate event sequences and event frequencies of occurrence for the evaluation.Once the specific event sequences of concern are identified, detailed thermal-hydraulic and structural evaluations can be performed to determine the conditions required to minimize the extension of postulated flaws or enhance flaw arrest in the reactor vessel. This paper addresses key aspects of the thermal-hydraulic and fracture mechanics analyses of the reactor vessel. The effects of incomplete mixing of safety injection flow in the primary cold leg and vessel downcomer and the application of warm prestressing are emphasized. The results of these analyses are being used to define further modifications in vessel and plant system design and to operating procedures.Previous design considerations that have evolved as a result of reactor vessel integrity evaluations are mentioned. These include the development of realistic design analysis tools and selection of plant system modifications. Modifications that are being developed or are under consideration are also mentioned. These include vessel fluence reductions, additional modifications to operating procedures, increased use of probabilistic event sequence and fracture mechanics analysis methods, enhanced material fracture toughness, and reductions in the severity or frequency of occurrence of dominant reactor vessel PTS transients.     

Short circumferential through-wall cracked pipes subjected to stretch-bending load

The methods for assessment of elastic–plastic fracture behaviour of cracked components include the net section plastic collapse concept, the J-integral approach, and the two-parameter R-6 failure assessment diagram, Revision 3. These failure assessment methods are usually used to obtain fracture behaviour prediction with a reasonable degree of accuracy without carrying out complicated full-length numerical fracture analysis. In the current work, fracture experiments on stainless steel pipes with short circumferential through-wall cracks under stretch-bending load were conducted. Stretch-bending load refers to the loading situation where axial load is generated that is proportional or related to the applied bending load. The J-integral values derived from the experimental load-point load–displacement data under stretch-bending and pure bending conditions are compared to investigate the effect of axial load on the J–resistance curves. The results show clear dependence of crack resistance force on axial load for short circumferential cracks. Crack resistance force decreased noticeably for increased stretch-bending loading compared to pure bending loading.     

Fracture Toughness of DCI Casks and Prediction by Small Specimen Test

Abstract

Spent nuclear fuel transport and/or storage containers (casks) must maintain their structural integrity even when subjected to hypothetical accidents during transport or handling accidents at storage facilities. For ductile cast iron (DCI) to be used as a cask containment boundary material, adequate fracture toughness must be demonstrated at service temperatures and Impact loading conditions of concern. In Japan, comprehensive studies of the fracture toughness of heavy section DCI have been undertaken by a number of research organisations to provide the safety assurance for the DCI casks. In the present study, the fracture toughness data were used to develop a lower bound trend curve for heavy section DCI and to examine the prediction methods by small specimen tests. The fracture toughnesses K_Ic, K_IIc and K_IIIc were also obtained to study the safety assessment of DCI casks under different loading mode conditions.     

Comparison of crack initiation life estimation procedures under creep-fatigue conditions

This paper describes the application of “high temperature structural integrity assessment procedures” developed in the UK and Japan to creep-fatigue crack initiation in welded Type 316 features tests. The components were subjected to both fatigue and creep-fatigue loading at 630 °C. The loadings are representative of those on the upper seal gimbal joint in an advanced gas cooled reactor (AGR), except that the tests were isothermal and the imposed dwell times were reduced. It is demonstrated that application of the procedures gives accurate predictions of the observed crack initiation in the weldment, based on two different advanced inelastic constitutive models (BE and CRIEPI models) and best estimate materials data. Application of simplified assessment methods based on elastic analysis is shown to be conservative. Where appropriate, contrasts between the UK and the Japanese assessment procedures and inelastic modelling techniques have been highlighted.     

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.     

Characteristics relevant to ductile failure of bimetallic welds and evaluation of transferability of fracture properties

Life management and structural integrity assessment of bimetallic welds in its state-of-the-art form relies on practical methods derived on the basis of years of experience in operation and simplistic strength of materials analyses. The complex conditions and properties of the weldment, as resulting from the elaborate interaction of different microstructures with gradients in material properties, have limited the ability of currently existing methods to construct the assessment on the basis of actual failure mechanisms of bimetallic welds. Current work addresses the assessment procedure by combining experimental and numerical fracture mechanics comprising a micro-mechanical evaluation of the relevant damage mechanisms. The studied dissimilar ferrite (SA508)–austenite (AISI 304) circumferencial weld is one with a Ni-enriched buttering layer.The experimental work comprises tensile and fracture mechanical characterization of the different microstructural zones of the bimetallic weld. Tensile properties are determined with microstructure specific flat bar specimens as well as round bar specimens enabling better inference of true stress–strain curves. Fracture resistance curves are established by applying small-specimen testing techniques. Different crack configurations are modeled by finite element analysis (FEA) to assess the relationships between fracture types, toughness and local near crack tip constraint parameters. Transferability and characterization question are considered by determining J–Q-trajectories and employing small-scale yielding corrections (SSYCs). On the basis of the experimental and numerical results and a fractographical investigation, the micromechanics of fracture are interpreted. Differences in strain hardening capacities of microstructural zones are found to most severely affect the toughness transitions of the weld and the associated failure modes. Two prime failure types are noted, one for cracks located at outer heat affected zone (HAZ) resulting in an unstable crack deflection towards the fusion line (FL) and another type associated with cracks positioned near the fusion line, wherein a low-toughness ductile fracture process results. Small fracture mechanics specimen is found applicable for fracture resistance determination of bimetallic weldments.     

反应堆压力容器接管嘴内隅角应力强度因子计算研究

本文针对反应堆压力容器接管嘴内隅角,采用含真实裂纹的三维有限元法对温度与压力作用下应力强度因子的计算进行了研究。以某工程压力容器接管嘴内隅角为例,用含真实裂纹的三维有限元法和目前使用的简化工程算法对压力与热载荷作用下的接管嘴内隅角应力强度因子进行了计算,并对两种方法的计算结果进行对比分析。结果表明:当简化工程算法得到的应力强度因子接近规范限值时,应对热载荷引起的应力强度因子进行详细有限元计算,以规避简化工程算法的不保守性给压力容器带来的快速断裂风险。     

Ductile crack growth in pre-cracked CVN specimens: numerical studies

This study describes plane strain, finite element analyses to model ductile crack extension in pre-cracked Charpy specimens subjected to static and impact loading. The Gurson–Tvergaard (GT) dilatant plasticity model for voided materials describes the degradation of material stress capacity. Fixed-size, computational cell elements defined over a thin layer along the crack plane provide an explicit length scale for the continuum damage process. Outside of this layer, the material remains undamaged by void growth, consistent with metallurgical observations. The finite strain constitutive models include the effects of high strain rates on the material flow properties. Parametric studies focusing on numerically generated R-curves quantify the relative influence of impact velocity, material strain rate sensitivity, and properties of the computational cells (thickness and the initial cell porosity). In all cases, impact loading elevates significantly the R-curve by increasing the amount of background plasticity. The strong effects of impact loading on the driving force for cleavage fracture are illustrated through evolution of the Weibull stress. The analyses suggest a negligible, additional effect of tearing on the Weibull stress under impact loading. Validation of the computational cell approach to predict loading rate effects on R-curves is accomplished by comparison to static and impact experimental sets of R-curves for three different steels.     

Reactor pressure vessel structural integrity research

Development continues on the technology used to assess the safety of irradiation embrittled nuclear reactor pressure vessels (RPVs) containing flaws. Fracture mechanics tests on RPV steel, coupled with detailed elastic-plastic finite element analyses of the crack-tip stress fields, have shown that (1) constraint relaxation at the crack-tip of shallow surface flaws results in increased data scatter but no increase in the lower-bound fracture toughness, (2) the nil-ductility temperature (NDT) performs better than the reference temperature for nil-ductility transition (RT_NDT) as a normalizing parameter for shallow flaw fracture toughness data, (3) biaxial loading can reduce the shallow flaw fracture toughness, (4) stress based dual-parameter fracture toughness correlations cannot predict the effect of biaxial loading on shallow flaw fracture toughness because in-plane stresses at the crack-tip are not influenced by biaxial loading, and (5) an implicit strain based dual-parameter fracture toughness correlation can predict the effect of biaxial loading on shallow flaw fracture toughness. Experimental irradiation investigations have shown that (1) the irradiation induced shift in Charpy V-notch vs. temperature behavior may not be adequate to conservatively assess fracture toughness shifts due to embrittlement, and (2) the wide global variations of initial chemistry and fracture properties of a nominally uniform material within a pressure vessel may confound accurate integrity assessments that require baseline properties.     

Effect of cladding on the stress intensity factors in the reactor pressure vessel

In general, reactor pressure vessels (RPV) are cladded with stainless steel to prevent corrosion and radiation embrittlement. The ASME Sec. XI specifies that a subclad crack which may be found during the in-service inspection must be considered as a semi-elliptical surface crack when the thickness of cladding is less than 40% of the crack depth. In order to refine the fracture assessment procedures for such subclad cracks, three-dimensional finite element analyses were applied for various subclad cracks embedded in the base metal. A total of 18 crack geometries were analyzed, and the results were compared with those for idealized semi-elliptical surface cracks for two different loading conditions, i.e. internal pressure and pressurized thermal shock. The resulting stress intensity factors for subclad cracks were 50–70% less than those for idealized surface cracks. It has been proven that the condition specified on the ASME Sec. XI is overly conservative for subclad cracks which are assumed to be surface cracks.     

The segmented expanding cone-mandrel test revisited as material characterization and component test for fuel claddings

This paper presents an assessment of the segmented expanding mandrel (SEM) test for material characterization and structural integrity assessment of nuclear fuel claddings. The loading is induced by expanding segments, which are placed inside a cladding tube, radially to simulate cracked fuel that expands thermally. Experimental results are presented for zircaloy-2 cladding tubes for different number of segments. The tests are analysed with semi-analytical models and two-dimensional finite element models. A complex stress field with stress concentrations occurs at the edge of the segments, which simulates pellet cladding interaction. The variation of the stresses and the strength of the stress and strain concentrations increases with fewer segments and increases strongly with higher friction coefficient between segments and the cladding tube. By increasing the number of segments and ensuring a low friction coefficient, the deformation is close to axi-symmetric and the SEM can be used to determine qualitatively the material properties such as fracture criteria and stress-strain curves, but the test is more appropriate for assessment of how defects and microstructure affect ductility. For simulation of mechanical pellet-cladding interaction it is important that the friction coefficient is representative. The resulting stress concentrations promote failure at lower loads and need to be taken into account for the integrity assessment. The SEM test can then be used as a relatively simple test for assessment of pellet fuel cladding interaction.     

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。      

Probabilistic modeling for fracture mechanic studies of reactor vessels with characterization of uncertainties

The research presented in this paper addresses some important probabilistic fracture mechanics issues raised in recent efforts for assessing pressurized thermal shock risk (PTS) in pressurized water reactors (PWRs). The common objectives addressing these issues were to (1) provide a framework including uncertainty characterization and treatment for PTS studies; (2) use the screening criteria for scenario selection developed and reported by the authors; (3) develop a simple computation approach for PTS analyses; and (4) support other researchers in their efforts to develop tools and techniques for reactor vessel failure assessment.The PTS evaluation consists of three disciplines: probabilistic risk assessment, thermal hydraulics, and probabilistic fracture mechanics. The major contribution of this research has been to provide a framework for the fracture mechanics aspects of pressurized thermal shock analysis including uncertainty characterization and treatment. The uncertainty analysis proposed in this research makes a distinction between aleatory and epistemic uncertainties for proper propagation of uncertainties.To ease the computational burden incurred by introducing thermal hydraulics uncertainties, a simple and reasonably accurate fracture mechanics computational approach that performs fracture mechanics calculations rapidly and conservatively is proposed. The approach is tested by developing a mathematical routine in form of a simple Mathematica™ routine called MATH-PFM.     

Degraded piping program - phase II progress

This paper summarizes the results from the NRC Degraded Piping Program over the last year. The objective of the NRC Degraded Piping Program - Phase II, is to verify limit-load analyses and develop elastic-plastic fracture mechanics analyses methods for cracked (degraded) nuclear piping under a variety of loading conditions. These analyses are used in leak-before-break evaluations. Since experimental efforts are conducted at LWR temperatures, failure modes and metallurgical phenomena of concern are also being assessed.     

Analytical assessment of an inclad defect in a PWR vessel belt line

The French approach to the assessment of the integrity of PWR vessels requires, in particular, that existence of large margins with respect to fast fracture shall be demonstrated for all kinds of defects which can be produced during manufacturing, taken with envelope sizes. The case of defects in the cladding with one tip against the base metal interface raises several difficult problems, mainly on the effect of residual stresses in the cladding, and on the choice of a relevant criterion for the risk of initiating cleavage cracking in the base material. The behaviour of a typical defect has been computed with elastic-plastic analyses and the criteria of the local approach of fracture: the effect of residual stresses is negligible and the margins with respect to fast fracture are much larger than those indicated previously by LEFM computations with plasticity corrections. The values obtained ensure that the integrity of the vessel would not be affected if such defects had occurred during manufacturing.