Naoki Ogawa

Residual Stress Evaluation of Dissimilar Weld Joint Using Reactor Vessel Outlet Nozzle Mock-Up Model (Report-1)

In recent years, the occurrence of primary water stress corrosion cracking (PWSCC) in Alloy 600 weld regions of PWR plants has increased. In order to evaluate the crack propagation of PWSCC, it is required to estimate stress distribution including residual stress and operational stress through the wall thickness of the Alloy 600 weld region. In a national project in Japan for the purpose of establishing residual stress evaluation method, two test models were produced based on a reactor vessel outlet nozzle of Japanese PWR plants. One (Test model A) was produced using the same welding process applied in Japanese PWR plants in order to measure residual stress distribution of the Alloy 132 weld region. The other (Test model B) was produced using the same fabrication process in Japanese PWR plants in order to measure stress distribution change of the Alloy 132 weld region during fabrication process such as a hydrostatic test, welding a main coolant pipe to the stainless steel safe end. For Test model A, residual stress distribution was obtained using FE analysis, and was compared with the measured stress distribution. By comparing results, it was confirmed that the FE analysis result was in good agreement with the measurement result. For mock up test model B, the stress distribution of selected fabrication processes were measured using the Deep Hole Drilling (DHD) method. From these measurement results, it was found that the stress distribution in thickness direction at the center of the Alloy 132 weld line was changed largely during welding process of the safe end to the main coolant pipe.Copyright

Measurement of Welding Residual Stresses of Reactor Vessel by Inherent Strain Method : Measurement of Residual Stresses of Pipe-Plate Penetration Joint

This study aims to ensure the safety of nuclear power plants. The accidents involving leaks from the welded zones at the pipe penetration part of a reactor vessel or at a coolant pipe are reported at home and abroad. One of the main causes is the welding residual stress. So, it is important to know the welding residual stress for maintaining high safety of the plants, the estimation of plants life cycle and the plan of maintenance. The welded joints of the nuclear power plants have complex shapes, and the welding residual stresses also have complex distributions three-dimensionally. In this study, the inherent strain method combined with finite element method is used to measure the welding residual stresses accurately. The mock-up is idealized for the welded joint at the pipe penetration part of the actual reactor vessel. The inherent strain method is applied to measure the residual stresses. In this method, the inherent strains are unknowns. When the residual stresses are distributed complexly in a three-dimensional stress-state, the number of unknowns becomes very large. So, the inherent strains are expressed with some functions to decrease the number largely. The theory, the experiment process and the analysed results are explained. The characteristics of the distributions of residual stresses and their production mechanisms are discussed. The inherent strain method gives the most probable values and the deviations of the residual stresses. The deviations are small enough for the most probable values. It assures the high reliability of the estimated results.

Measurement of welding residual stresses of reactor vessels by inherent strain method – Diagnosis of inherent strain distribution function

The fundamental objective of this study is to ensure the safety of nuclear reactors. A few accidents involving leaks from welded zones at the pipe penetration part of reactor vessels or at coolant pipes have been reported at home and abroad. One of the main causes is welding residual stress. Therefore, it is very important to know the welding residual stress in order to maintain the high safety of the plant, estimate the plant life cycle and design an effective maintenance plan. Welded joints of nuclear reactor vessels have complex shapes, and the welding residual stresses also have three-dimensional (3D) complex distributions. In this study, inherent strain-based theory and method are applied to measure the welding residual stresses. The inherent strain method is an analytical method as an inverse problem, using the least squares method, based on the finite element method. So the method gives the most probable value and deviation of residual stress. The reliability of the estimated result is discussed. In this method, inherent strains are unknowns. When residual stresses are distributed complexly in a 3D stress-state, the number of unknowns becomes very large. So, the inherent strain distribution is expressed with an appropriate function to decrease largely the number. A mock-up is idealized for a welded joint at the pipe penetration part of an actual reactor vessel. The inherent strain method is applied to the measure the residual stress of the joint. In this paper, the applicability of the inherent strain distribution function is diagnosed. Ten kinds of functions are applied to estimate the residual stress, and the accuracy and reliability of the analysed results are judged from three points of view, i.e. residuals, unbiased estimate of variance of errors and welding mechanics. The most suitable function is selected, which brings the most reliable result.

Evaluation on Constraint Effect of Reactor Pressure Vessel Under Pressurized Thermal Shock

In recent years, the integrity of reactor pressure vessel (RPV) under pressurized thermal shock (PTS) accident has become controversial issue since the larger shift of RTNDT in some higher fluence surveillance data raised a concern on RPV integrity. Under PTS condition, the combination of thermal stress due to a temperature gradient and mechanical stress due to internal pressure causes considerable tensile stress inside the wall of RPV. Currently, RPV integrity is assessed by comparing stress intensity factor on a crack tip under PTS condition and a reference toughness curve based on the fracture toughness data of irradiated compact specimens. Since PTS loading is large enough to cause plastic deformation, a crack tip behavior on the inner surface of RPV can be explained by elastic-plastic fracture mechanics using the J-integral.In this study, 3D elastic plastic finite element analyses were performed to assess the crack tip behavior on surface of a RPV under Loss of coolant Accident, which causes one of the most severe PTS condition. In order to quantify the constraint effect on a surface crack, J-Q approach was applied. The constraint effect of a surface crack was compared with a compact specimen and its influence on the fracture toughness was assessed.As a result, the difference of constraint effect was clearly obtained. And it is recommended to consider constraint effects in the evaluation of structural integrity of RPV under PTS.Copyright

Fracture Evaluation of Ni-Based Alloy Weld Joint of Cylindrical Model Subjected to 4-Point Bending or Inner Pressure

For the purpose of establishing fracture evaluation method of nickel based alloy weld of nuclear power plants, fracture tests using pipe models (8B and 14B for bending, 12B for inner pressure) with an alloy 132 weld joint have been performed at room temperature and high temperature (325°C). The predicted loads calculated by limit load evaluation method using the measured and code regulated flow stresses were compared with the maximum test loads. And the predicted bending loads of the pipes at 325°C (8B and 14B) and at room temperature (8B) with the initial surface crack whose depth is 75% of the pipe thickness were in good agreement with the maximum test loads. Also the predicted inner pressure of the pipe at room temperature (12B) agreed with the measured maximum pressure. Only for one case of the 14B pipe subjected to the bending load at room temperature, the predicted load by limit load evaluation method has 20% unconservative difference from the measured data, on the other hand, the predicted load by J-T analysis made this difference smaller and conservative.Copyright

Proposal for Update on Evaluation Procedure for Reactor Pressure Vessels Against Pressurized Thermal Shock Events in Japan

The evaluation procedure for the reactor pressure vessel integrity of Japanese PWR plants against Pressurized Thermal Shock (PTS) events is prescribed in the Japan Electric Association Code, JEAC 4206, “Method of Verification Tests of the Fracture Toughness for Nuclear Power Plant Components” since 1991. The current procedure was developed based on the PTS verification test program, which was conducted as Japanese national project and the related studies in 1980’s.Since much progress has been made on fracture mechanics, fracture toughness, in-service inspection techniques/results and so on, it is preferred to advance the current procedure for more credible evaluation by reflecting the latest knowledge.This paper describes the outline of the studies to update the current procedure.© 2014 ASME

Residual Stress Evaluation of Dissimilar Weld Joint Using Mock-Up Models of Bottom Mounted Instrument Nozzle

In this paper benchmark analyses was performed to evaluate the accuracy of the FE analysis method for estimating residual stress of a dissimilar weld joint of a bottom mounted instrument (BMI) nozzle. The analyses simulated each fabrication and operational process of the weld joint nozzle. For confirmation of calculation accuracy a mock-up models of BMI nozzle were produced using the same fabrication process in Japanese PWR plants and the stress distribution of the model was measured using inherent strain method. From the limitation of the geometry of the BMI nozzle smaller strain gauges were applied to decrease the measurement error. As a result a better agreement between the calculation and measurement was obtained than that of the former investigation and the circumferential stress is confirmed dominant component in the weld region.Copyright

Sensitivity Analysis of SCC Crack Extension Simulation

For the preliminary investigation of realistic SCC crack extension behavior, several crack extension simulation methods were applied to example problems. As for a mesh division procedure, two kinds of methods were employed. One is a conventional one which needs new mesh generation of finite element (FE) models following crack extension and the other one is X-FEM without new FE mesh generation. Also the effects of crack extension contained within the weld material and the crack growth rate dependence on the crack extension direction at the weld joint were considered. Under hypothetical liner stress distribution, the stress intensity factors from the FE analyses by the conventional mesh division procedure and X-FEM agreed well with those from the stress intensity factor K equations of a hand book in the case of semi-elliptical and semi circular cracks. As for the crack extension analysis, consideration of the direction dependent crack growth rate and limited crack extension within the hypothetical shaped weld groove boundary gave a large effect on the crack extension amount in the thickness direction.© 2011 ASME

Fracture Evaluation Using Limit Analysis for Complex Structure (Pressure Load)

The sensitivity study of limit analyses was performed for a cracked weld joint between a pipe and a nozzle. As limit analysis methods, the twice elastic slope method in Sec. III basis, the lower bound asymptotic method and net section criterion for straight pipes were chosen. Evaluation method, FE model’s range and material properties were the parameters for the analyses and those effects or sensitivities on failure load were investigated. Based on the results, the key issues for limit analyses for complex structures, especially a pipe with a nozzle were summarized.Copyright

Sensitivity Analysis of Residual Stress Simulation of Dissimilar Metal Joint of Safe End Nozzle and Key Issues for Standard Procedure to Maintenance Rules

For evaluation of PWSCC crack extension behavior detailed residual stress evaluation such a stress distribution on the evaluation points’ thickness is needed. In order to establish the residual stress evaluation procedure, Japan Nuclear Energy Safety Organization (JNES) had a seven years’ project from 2001 to 2007. In this project several sensitivity analyses were performed for understanding key points for the numerical simulation of residual stress. And also residual stress measurements were carried out for validation of the simulation methods. In this paper focused on the safe end nozzles, the results of the sensitivity analyses and the measurement result of the residual stress of the U-groove weld joint were introduced. Based on these result, the key issues of a guideline of a standard residual stress analysis procedure was proposed.Copyright