Heli Talja

Comparison of approaches for estimating pipe rupture frequencies for risk-informed in-service inspections

Abstract This paper describes the comparative study of two approaches to estimate pipe leak and rupture frequencies for piping. One method is based on a probabilistic fracture mechanistic model while the other one is based on statistical estimation of rupture frequencies from a large database. In order to be able to compare the approaches and their results, the rupture frequencies of some selected welds have been estimated using both of these methods. This paper highlights the differences both in methods, input data, need and use of plant specific information and need of expert judgement. The study focuses on one specific degradation mechanism, namely the intergranular stress corrosion cracking. This is the major degradation mechanism in old stainless steel piping in BWR environment, and its growth is influenced by material properties, stresses and water chemistry.

Consistency of fracture assessment criteria for the NESC-1 thermal shock test

Abstract In this report, a best estimate analysis of the NESC Spinning Cylinder test is performed and the consistency of the state of the art application of the different fracture parameters and assessments criteria is examined. Direct fracture toughness measurements best describes the fracture behavior of the NESC Spinning Cylinder in the simulated pressurized thermal shock loading. The concept based on the ASME RT NDT reference temperature (either Pellini or Cv test) does not provide consistent description of the NESC1 fracture assessment. In general the fracture mode, i.e. ductile initiation and tearing followed by cleavage event, was successfully estimated by 3D FE based fracture assessment, combined with Master Curve description of fracture toughness.

Fracture behaviour of large scale pressure vessels in the hydrotest

Abstract Within the Nordic Countries a four-year research programme in the area of elastic-plastic fracture mechanics was initiated in 1985. This programme aims to assess the leak-before-break (LBB) criteria for pressure vessels and piping. The main experimental effort of the programme is to rupture large size pressure vessels, one having dimensions resembling those of a nuclear reactor pressure vessel, under internal pressure. Artificial flaws were made on the inner wall of the vessels. The dimensions of the flaws were defined by calculations so that the LBB condition was just anticipated during the test. For the time being two tests have been performed. The first test with a large pressure vessel was pressurized by water at 60°C, which was the lowest acceptable temperature for the hydrotest. In this paper experimental details including flaw preparation, instrumentation and material characterization are described. The fracture behaviour as well as experimental results of the tests are reported and compared to the analytical solutions of the analyses.

Comparative analyses concerning integrity of a VVER-440 reactor pressure vessel

Comparative analyses have been performed with 3-D finite element (FE) models for a six loop cladded reactor pressure vessel with an assumed circumferential surface crack. In the analyses, a thermomechanical transient due to a loss of coolant with high pressure injection has been considered. The transient is characterized by an axisymmetric cooling phase which is followed by an asymmetric, strip-like cooling period. For the fracture assessment, stress intensity factors calculated from local J-integral values have been compared. It has been shown that with different FE programs and meshes, the crack loading can be calculated within a 10% scatter. Furthermore, simplified fracture methods show a much larger scatter, mainly due to approximation of the non-linear stress distribution by membrane and bending stress.

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

Abstract 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

Abstract 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.

Fracture behaviour simulation of flawed full scale pressure vessel

Abstract In 1985 a four year research programme in elastic-plastic fracture mechanics was initiated within the Nordic countries. The aim of the programme was to verify the methods used for fracture analysis of real structures. A large cylindrical pressure vessel, having dimensions resembling those of a nuclear reactor pressure vessel, was tested in 1986. An artificial sharp axial surface flaw was made on the inner wall of the vessel. One of the circumferential welds intersected the crack at its midpoint. Failure of the vessel occurred as local rupture in the weld area. A maintenance deck, which was located around the midsection of the pressure vessel, was partially removed so as to prevent interference with the expected vessel deformations upon pressurization. After the test, two three-dimensional nonlinear finite element analyses were performed taking into account the existence of the circumferential weld in the ligament. In the first case, the original flawed structure was modelled allowing for different stress-strain curves of the base and weld material. In the second analysis, the model included a short through-the-wall crack in the weldment. Additionally, a simple two-dimensional analysis was made assuming the crack to be infinitely long in the axial direction. A three-dimensional analysis was repeated without considering the effect of the maintenance deck. For fracture mechanics evaluation, J- integrals along the crack front were calculated. In this paper, results of the three-dimensional analyses are reported and compared to experimental findings.

Elasto-plastic analysis of a cracked ductile cylindrical pressure vessel

Abstract The deformation of a long cylindrical pressure vessel made of strain hardening material with a long longitudinal crack is analyzed under internal pressure by an FEM code. Stress distributions, the shape of the crack opening, the plastic zone, the J-integral and the CTOD are calculated. The J CTOD ratio is found to be practically independent of the amount of pressure load, as well as of the crack height. Comparisons are made with a fully plastic strip yield model which seems to give a reasonable estimate of crack opening. This model suggests that the CTOD is linearly dependent on the radius of cylinder. This implies that the size of the pressure vessel is an important parameter when considering the tolerance of flaws.

Nordic numerical round robin for a side-grooved CT-specimen

Abstract The reliability of fracture mechanics assessments based on finite element method calculations has to be confirmed before making safety assessments for critical components like nuclear pressure vessels. Calculations for simulation of fracture mechanics tests and numerical round robin programmes are useful methods in such verification. In this paper the results of Nordic finite element round robin calculations for a side-grooved CT-specimen are presented and compared with experimental results. The round robin consisted of two parts. In the first part, two-dimensional calculations assuming plane strain and plane stress behaviour were made. In the second part, three-dimensional calculations were performed for best-estimate analysis of the specimen behaviour. The differences between the solutions were comparatively small. Only one two-dimensional solution, where lower order finite elements were used, deviated clearly from the other ones. A good agreement was seen between two-dimensional plane strain results and experimental results. Three-dimensional calculations correspond very accurately with each other and with the experiment.

Computational methods assuring nuclear power plant structural integrity and safety: an overview of the recent activities at VTT

Numerical, simplified engineering and standardised methods are applied in the safety analyses of primary circuit components and reactor pressure vessels. The integrity assessment procedures require input relating both to the steady state and transient loading actual material properties data and precise knowledge of the size and geometry of defects. Current procedures hold extensive information regarding these aspects. It is important to verify the accuracy of the different assessment methods especially in the case of complex structures and loading. The focus of this paper is on the recent results and development of computational fracture assessment methods at VTT Manufacturing Technology. The methods include effective engineering type tools for rapid structural integrity assessments and more sophisticated finite-element based methods. An integrated PC-based program system MASI for engineering fracture analysis is described. A summary of the verification of the methods in computational benchmark analyses and against the results of large scale experiments is presented.