Masayuki Asano

Fatigue crack growth behavior of weld heat-affected zone of type 304 stainless steel in high temperature water

Abstract The fatigue crack growth behavior of the weld heat-affected zone (HAZ) of type 304 stainless steel in high temperature water which simulates the boiling-water reactor environment was investigated to clarify the effects of welding residual stress, cyclic frequency f and thermal aging on crack growth rate. A lower crack growth rate of the HAZ than of the base metal was observed in both the high temperature water and the ambient air caused by the compressive residual stress. The crack closure point was measured in the high temperature water. The effect of the welding residual stress on the crack growth rate of the HAZ can be evaluated separately from the environmental effect through the crack closure behavior. The high temperature water increased the crack growth rate at a cyclic frequency of 0.0167 Hz but did not affect it much at 3 and 5 Hz. The crack growth behavior of the thermally aged HAZ at 400 °C for 1800 h was almost the same as that of the unaged material tested at 0.0167 and 5 Hz in the high temperature water.

Development of a flaw evaluation handbook of the high pressure institute of Japan

This paper introduces a handbook, which is edited by the High Pressure Institute of Japan to support engineers who evaluate flaws detected in nuclear power components according to the Japanese fitness-for-service code. The handbook is written in Japanese and contains basic information on fracture mechanics as well as the specific flaw evaluation procedures and material properties data stipulated in the code. The main features of the handbook are summarized in the paper.

Effect of thermal aging on fracture toughness of RPV steel

The effect of thermal aging on mechanical properties and fracture toughness was investigated on pressure vessel steel of light water reactors. Submerged are welded plates of ASME SA508 C1.3 steel were isothermally aged at 350°C, 400°C and 450°C for up to 10,000 hrs. Tensile, Charpy impact and fracture toughness testings were conducted on the base metal and the weld heat affected zone (HAZ) material to evaluate whether thermal aging induced by the plant operation is critical for the integrity of the pressure vessel or not. Tensile properties of the base metal was not changed by thermal aging as far as the thermal aging conditions were concerned. Relatively distinct degradation was observed in fracture toughness JIC and J-resistance properties of both the base metal and the weld HAZ material, while only slight changes were observed in Charpy impact properties for both of them. However, it was concluded that the effect of thermal aging estimated by 40–80 years of plant operation on fracture toughness of both materials is small.

Ductile fracture analysis of carbon steel pipe with a circumferential through-wall crack

Abstract It is necessary to make clear the pipe fracture conditions based on elastic-plastic fracture mechanics to assess the leak before break situation of carbon steel pipes for LWR plants. The aim of the present work is to discuss the effects of pipe size, initial crack length and fracture toughness on the estimated fracture load and mode of carbon steel pipes with a circumferential through-wall crack. As an analytical method, the R6-Rev.3 approach was applied to the pipe fracture analyses considering its simplicity in the application. The results indicate that the net-section collapse attainment becomes difficult with increasing diameter and decreasing thickness of carbon steel pipes. The degree of the net-section collapse attainment decreases with increasing crack length up to some critical size and then increases. The predicted fracture load is more sensitive to the materials J– R curve than to the elastic-plastic fracture toughness J IC . And a simple limit load analysis based on the yield stress is appropriate to evaluate the fracture load, as long as a proper margin was included.

Crack Propagation Analysis Procedure Using FEM Applied to the Three-Dimensional Stress Field

Recently, several cracks have been found on the weld joints of Boiling Water Reactor (BWR) core shrouds during inspection. In order to ensure the continuous operation of nuclear power plants, it is necessary to assess the structural integrity of core shrouds with cracks on the weld joints. In general, a crack propagates in a complicated manner according to three-dimensional stress field and it is difficult to predict crack propagation direction and crack shape change. Usually, half ellipsoid crack shape is assumed and the propagation of the crack is calculated in the constant direction for assessment. In this study, crack propagation analysis procedure using the Finite Element Method (FEM) is developed for general shaped crack, and the procedure is verified by experiments. In this procedure, it is assumed that the crack propagates according to the maximum J-integral under three-dimensional stress fields and the re-mesh technique is used in the FEM analysis in order to calculate crack shape variation during propagation. The fatigue crack propagation tests under cyclic tensile load were performed to verify the analysis procedure. The specimens are made of a plate from 316SS and designed to generate non-uniform stress distribution on the crack front in order to observe continuous crack propagation direction change.Copyright

A study of the structural integrity of the core support structure of a fast breeder reactor

This paper reports on the core support structure of a fast breeder reactor supports the fuel assemblies, supplies sodium coolant to the fuel assemblies, and maintains the insertability of control rods even during an earthquake. The core support structure is designed as a box fabricated of welded plates, ribs, and cylinders that distribute the load in a diverse manner, in order to reduce the weight and to fulfill safety-related functions. This box structure was not adopted in the Monju prototype reactor; thus, a method to evaluate the structural integrity of this structure must be developed. To prepare design guidelines, structural integrity was studied in accordance with the requirements and features of the box structure. From the results of these experiments, the crack growth rate was evaluated and incorporated into the structural integrity evaluation method. Finally, the structural integrity of the core support structure of the Japanese demonstration reactor is evaluated by this method.

On a Conventional Stress Intensity Factor Calculation Technique of Surface Cracks

This paper describes one conventional stress intensity factor calculation procedure of surface cracks by Boundary Integral Equation Method. Using both stress and displacement type singular elements, the procedure utilizes displacement and traction solutions to give accurate stress intensity factor in the analysis with coarse mesh data. The numerical results show that the present approach improves the stability of tractions near crack tip and gives accurate stress intensity factor solutions in a sense of engineering evaluation.