Strategies for Treating Weld Residual Stresses in Probabilistic Fracture Mechanics Codes

Risk based treatment of degradation and fracture in nuclear power plants has emerged as an important topic in recent years. One degradation mechanism of concern is stress corrosion cracking. Stress corrosion cracking is strongly driven by the weld residual stresses (WRS) which develop in nozzles and piping from the welding process. The weld residual stresses can have a large uncertainty associated with them. This uncertainty is caused by many sources including material property variations of base and welds metal, weld sequencing, weld repairs, weld process method, and heat inputs. Moreover, often mitigation procedures are used to correct a problem in an existing plant, which also leads to uncertainty in the WRS fields. The WRS fields are often input to probabilistic codes from weld modeling analyses. Thus another source of uncertainty is represented by the accuracy of the predictions compared with a limited set of measurements. Within the framework of a probabilistic degradation and fracture mechanics code these uncertainties must all be accounted for properly. Here we summarize several possibilities for properly accounting for the uncertainty inherent in the WRS fields. Several examples are shown which illustrate ranges where these treatments work well and ranges where improvement is needed. In addition, we propose a new method for consideration. This method consists of including the uncertainty sources within the WRS fields and tabulating them within tables which are then sampled during the probabilistic realization. Several variations of this process are also discussed. Several examples illustrating the procedures are presented.

The Effects of Warm Pre-Stressing of a Pre-Cracked Pressurised Thermal Shock Disk Specimen Under a LUCF Loading Cycle

A finite element analysis has been performed to investigate the effects of warm prestressing of a pre-cracked PTS-D (Pressurized Thermal Shock Disk) specimen, for comparison with the experimental work conducted by the Belgium SCK-CEN organisation under the European NESC VII project. The specimen was loaded to a maximum loading at −50 °C, unloaded at the same temperature, cooled down to −150 °C, and then re-loaded to fracture at −150 °C. This is a loading cycle known as a LUCF cycle. The temperature-dependant tensile stress-strain data was used in the model and the finite element software ABAQUS was used in the analysis. The finite element results were used to derive the apparent fracture toughness by three different methods: (1) Chell’s displacement superposition method; (2) the local stress matching method; and (3) Wallin’s empirical formula. The apparent fracture toughness values were derived at the deepest point of the semi-elliptical crack for a 5% un-prestressed fracture toughness of 43.96 MPam1/2 at −150 °C. The detailed results were presented in the paper.

Crack Propagation Simulation: A Local Deformation-Based Mesh Mapping Method for Crack Modeling and a Direct Method for Residual Stress Representation in Welds

Cracks found in a nuclear power plant reactor coolant system (RCS), such as primary water stress corrosion cracking (PWSCC) and intergranular stress corrosion cracking (IGSCC), usually have natural crack front shapes that can be very different from the idealized semi-elliptical or rectangular shapes considered in engineering handbooks and other analytical solutions based on limited shapes. Simplifications towards semi-elliptical shape or rectangular shape may potentially introduce unnecessary conservatism when the simplified shape has to contain the actual crack shape. On the other hand, it is very time-consuming to create a three-dimensional (3D) finite element (FE) model to simulate crack propagation in a natural shape using existing public-domain software like ABAQUS or ANSYS. In this study, a local deformation-based mesh-mapping (LDMM) method is proposed to model cracks with a natural front shape in any 3D structures. This methodology is first applied to model circumferential surface cracks with a natural crack front shape in the cross-sectional plane of a cylinder. The proposed new method can be applied to simulate both shallow and deep cracks. Also discussed in this paper is a direct method to reproduce welding residual stresses in the crack model using temperature fields combined with other sustained loads to predict crack propagations. With this novel LDMM method, natural crack fronts and non-planar crack faces can be easily modeled. The proposed new method can be used to generate a high-quality finite element model that can be used for both linear-elastic fracture mechanics (LEFM) and elastic-plastic fracture mechanics (EPFM) analyses. The study case illustrates that the proposed LDMM method is easy to implement and more efficient than the existing commercial software.

Weld Residual Stress in Large Diameter Nuclear Nozzles

Weld residual stresses in nuclear power plant can lead to cracking concerns caused by stress corrosion. These are large diameter thick wall pipe and nozzles. Many factors can lead to the development of the weld residual stresses and the distributions of the stress through the wall thickness can vary markedly. Hence, understanding the residual stress distribution is important to evaluate the reliability of pipe and nozzle joints with welds. This paper represents an examination of the weld residual stress distributions which occur in various different size nozzles. The detailed weld residual stress predictions for these nozzles are summarized. Many such weld residual stress solutions have been developed by the authors in the last five years. These distributions will be categorized and organized in this paper and general trends for the causes of the distributions will be established. The residual stress field can therefore feed into a crack growth analysis. The solutions are made using several different constitutive models such as kinematic hardening, isotropic hardening, and mixed hardening model. Necessary fabrication procedures such as repair, overlay and post weld heat treatment are also considered. Some general discussions and comments will conclude the paper.

Slow Crack Growth Resistance of Parent and Joint Materials From PE4710 Piping for Safety-Related Nuclear Power Plant Piping

HDPE pipes are considered by the nuclear industry as a potential replacement option to currently employed metallic piping for service-water applications. The pipes operate under high temperatures and pressures. Hence HDPE pipes are being evaluated from perspective of design, operation, and service life requirements before routine installation in nuclear power plants. Various articles of the ASME Code Case N-755 consider the different aspects related to material performance, design, fabrication, and examination of HDPE materials. Amongst them, the material resistance (part of Article 2000) to the slow crack growth (SCG) from flaws/cracks present in HDPE pipe materials is an important concern. Experimental investigations have revealed that there is a marked difference (almost three orders less) in the time to failure when the notch/flaw is in the butt-fusion joint, as opposed to when the notch/flaw is located in the parent HDPE material. As part of ongoing studies, the material resistance to SCG was investigated earlier for unimodal materials. The current study investigated the SCG in parent and butt-fusion joint materials of bimodal HDPE (PE4710) pipe materials acquired from two different manufacturers. The various stages of the specimen deformation and failure during the creep test are characterized. Detailed photographs of the specimen side-surface were used to monitor the specimen damage accumulation and SCG. The SCG was tested using a large specimen (large creep frame) as well as using a smaller size specimen (PENT frame) and the results were compared. Further, the effect of polymer orientation or microstructure in the bimodal HDPE pipe on the SCG was studied using specimens with axial and circumferential notch orientations in the parent pipe material.

NRC/EPRI Residual Stress Validation Program Phase I: Experimental Specimen Modeling and Measurement

The US Nuclear Regulatory Commission (NRC) and the Electric Power Research Institute (EPRI) are working cooperatively under a memorandum of understanding to validate welding residual stress predictions in pressurized water reactor primary cooling loop components containing dissimilar metal welds. These stresses are of interest as DM welds in pressurized water reactors are susceptible to primary water stress corrosion cracking (PWSCC) and tensile weld residual stresses are one of the primary drivers of this stress corrosion cracking mechanism. The NRC/EPRI weld residual stress (WRS) program currently consists of four phases, with each phase increasing in complexity from lab size specimens to component mock-ups and ex-plant material. This paper describes the Phase 1 program, which comprised an initial period of learning and research for both FEA methods and measurement techniques using simple welded specimens. The Phase 1 specimens include a number of plate and cylinder geometries, each designed to provide a controlled configuration for maximum repeatability of measurements and modeling. A spectrum of surface and through-wall residual stress measurement techniques have been explored using the Phase 1 specimens, including incremental hole drilling, ring-core, and x-ray diffraction for surface stresses and neutron diffraction, deep-hole drilling, and contour method for through-wall stresses. The measured residual stresses are compared to the predicted stress results from a number of researchers employing a variety of modeling techniques. Comparisons between the various measurement techniques and among the modeling results have allowed for greater insight into the impact of various parameters on predicted versus measured residual stress. This paper will also discuss the technical challenges and lessons learned as part of the DM weld materials residual stress measurements.

STYLE: Study on Transferability of Fracture Material Properties From Small Scale Specimens to a Real Component

This paper describes numerical and experimental investigations on transferability of material properties obtained by testing of small scale specimens to a real component. The presented study is related to the experimental and analytical work performed on Mock-up3, which is one of three unique large scale Mock-ups tested within the European project STYLE. Mock-up3 is foreseen to investigate transferability of material data, in particular fracture mechanics properties. An important part of this work is to study constraint effects on different small scale specimens and to compare their fracture behaviour with the fracture behaviour of a large scale (component like) structure. The Mock-Up3 is an original part of a surge line made of low alloy steel 20 MnMoNi 5 5 (which corresponds to SA 508 Grade 3, Cl. 1). The goal of the test is to introduce stable crack growth of an inner surface flaw until a break through the wall occurs. To design such a test reliable fracture mechanics material properties must be available. Usually, these material data are obtained by testing small specimens, which are subsequently used for the assessment of a large scale structure (component). This is being done under the assumption that these “small scale” material properties are fully transferable to “large scale” components. It is assumed that crack initiation in the ductile tearing regime is rather independent of the crack shape, a/W ratio, loading condition or size of the specimen (constraint effects). In order to check the aforementioned assumption and to improve understanding of the physical process leading to failure of cracked components comprehensive experimental and analytical work is being undertaken in STYLE. This paper summarizes Up-To-Date available results, which have been achieved during the first 15 months of the project.

The Effects of Loadings on Welding Residual Stresses and Assessment of Fracture Parameters in a Welding Residual Stress Field

The level of welding residual stress is an important consideration in the ECA of a structure or component such as a pipeline girth weld. Such a consideration is further complicated by their variation under load and the complexity involved in the proper assessment of fracture mechanics parameters in a welding residual stress field. In this work, 2D axi-symmetric FEA models for simulation of welding residual stresses in pipe girth welds were first developed. The modelling method was validated using experimental measurements from a 19-pass girth weld. The modeling method was used on a 3-pass pipe girth weld to predict the residual stresses and variation under various static and fatigue loadings. The predicted relaxation in welding residual stress is compared to the solutions recommended in the defect assessment procedure BS 7910. Fully circumferential internal cracks of different sizes were introduced into the FE model of the three-pass girth weld. Two methods were used to introduce a crack. In one method the crack was introduced instantaneously and the other method introduced the crack progressively. Physically, the instantaneously introduced crack represents a crack originated from manufacturing or fabrication processes, while the progressively growing crack simulates a fatigue crack induced during service. The J-integral values for the various cracks in the welding residual stress field were assessed and compared. This analysis was conducted for a welding residual stress field as a result of a welding simulation rather than for a residual stress field due to a prescribed temperature distribution as considered by the majority of previous investigations. The validation with the 19-pass welded pipe demonstrated that the welding residual stress in a pipe girth weld can be predicted reasonably well. The relaxation and redistribution of welding residual stresses in the three-pass weld were found to be significantly affected by the magnitude of applied loads and the strain hardening models. The number of cycles in fatigue loading was shown to have little effect on relaxation of residual stresses, but the range and maximum load together governed the relaxation effect. A significant reduction in residual stresses was induced after first cycle but subsequent cycles had no marked effect. The method of introducing a crack in a FE model, progressively or instantaneously, has a significant effect on J-integral, with a lower value of J obtained for a progressively growing crack. The path-dependence of the J-integral in a welding residual stress field is discussed.

Fast Three-Dimensional Finite Element Analysis of Weld Overlay Application on a Formed Feeder Elbow

This paper presents a fast finite element analysis (FEA) model to efficiently predict the residual stresses in a feeder elbow in a CANDU nuclear reactor coolant system throughout the various stages of the manufacturing and welding processes, including elbow forming, Grayloc hub weld, and weld overlay application. The finite element (FE) method employs optimized FEA procedure along with three-dimensional (3-D) elastic-plastic technology and large deformation capability to predict the residual stresses due to the feeder forming and various welding processes. The results demonstrate that the fast FEA method captures the residual stress trends with acceptable accuracy and, hence, provides an efficient and practical tool for performing complicated parametric 3-D weld residual stress studies.

Overview of RPV Sub-Project of Perform 60

PERFORM 60 (Prediction of the effects of radiation for reactor pressure vessel and in-core materials using multi-scale modelling — 60 years foreseen plant lifetime) is a 48-month project of the 7th Framework of the European Atomic Energy Community (EURATOM) being carried out under the auspices of the Directorate General Research, Technology and Development (DG.RTD) of the European Commission. Launched in March 2009, and building on the achievements of PERFECT, a EURATOM 6th Framework project, PERFORM 60 has as its main objective the development of multi-scale modelling tools integrated onto a common software platform, aimed at predicting for PWRs (i) the effects of irradiation on RPV materials (low alloy bainitic steels), (ii) the combined effects of irradiation and corrosion on internals (austenitic stainless steels). Accordingly, PERFORM 60 is based on two main technical sub-projects: SP1 (RPV) and SP2 (Internals). An integration work package within both SP1 and SP2 serves to facilitate software development. A Users’ Group (SP3) supports the main technical sub-projects and allows representatives of constructors, utilities, regulators and research organizations from Europe and further afield to receive the information and training needed to make their own appraisal as to the validity of the developed tools. A significant effort is also being made to train young researchers in the field of physical modelling of materials degradation due to neutron irradiation. Against this background, the paper provides an overview of SP1, highlighting the various models and methods being developed, building on the achievements of PERFECT, to describe the evolution of flow properties of low-alloy steels with irradiation and address their subsequent effects on cleavage fracture behaviour.