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In the structural integrity assessment of an embrittled reactor pressure vessel (RPV) based on probabilistic fracture mechanics (PFM), it is crucial to consider both of the mean value and the uncertainty of the embrittlement prediction. Typically, the embrittlement prediction uncertainty is given by a normal distribution, and its standard deviation is determined from the residuals of the measured and predicted values of all data used to develop the embrittlement prediction method. Therefore, the same uncertainty is assumed regardless of neutron fluence and the available data. To overcome this issue, the Japan Atomic Energy Agency recently developed the embrittlement prediction method based on machine learning and Bayesian statistics, i.e., the Bayesian nonparametric (BNP) method, and introduced it to the PFM analysis code. The BNP method can predict the probability distribution according to data scarcity and provide more reasonable uncertainty because it estimates significant uncertainties where the scatter of actual measured data is large and the number of data is small. In this study, PFM analysis for a Japanese model RPV in a pressurized water reactor is conducted using the PFM analysis of structural components in aging light water reactors, where the embrittlement probability distribution calculated using the BNP method is used. Furthermore, the effects of different embrittlement prediction methods and uncertainties on the failure frequency of RPVs are investigated in the PFM analyses, and the results are presented. • Embrittlement prediction based on machine learning was applied to probabilistic fracture mechanics analysis. • The probability distribution of superposition of normal distributions according to the number and variation of the data was used. • The effect of different embrittlement prediction methods and uncertainties on the failure frequency of RPVs was investigated. [ABSTRACT FROM AUTHOR]

The fitness for service code of the Japan Society of Mechanical Engineers (JSME) provides a fracture evaluation methodology based on linear elastic fracture mechanics (LEFM) for boiling water reactor (BWR) internal components irradiated to a neutron fluence of 3.0 × 1024 n/m2 or more. On the other hand, it is known that austenitic stainless steel maintains its ductility and toughness even after neutron irradiation up to the fluence expected for BWR internal components. Accordingly, it is preferable to apply the ductile fracture evaluation methodology for BWR internal components. The present study examined the applicability of damage mechanics analysis based on the Gurson‒Tvergaard‒Needleman (GTN) model for irradiated stainless steel. The changes of mechanical properties of the material due to neutron irradiation were simulated by cold working. As a result, in terms of material strength properties the applicability of the GTN model was confirmed for irradiated stainless steel using the simulated material for neutron fluence up to about 1.5 × 1025 n/m2. • Damage mechanics analysis based on the Gurson‒Tvergaard‒Needleman (GTN) model for irradiated stainless steel was investigated. • Changes of mechanical properties of the material due to neutron irradiation were simulated by cold working. • Fracture tests and analyses were performed with plate specimens. • Applicability of GTN model was examined by comparing the maximum loads and fracture behaviors between the tests and analyses. • Applicability of the GTN model was confirmed for irradiated stainless steel using the simulated material for neutron fluence up to about 1.5 × 1025 n/m2. [ABSTRACT FROM AUTHOR]

Most of reactor vessel internals (RVIs) have been constructed by austenitic stainless steels (ASSs) to support and protect core from high temperature coolant and radiation particles. However, if they are operated in the harsh environment for a long time, ASS materials may undergo significant changes in micro-structural features, macroscopic deformation and strengths due to age-related degradation mechanisms (ARDMs). In this study, user subroutines were developed based on material constitutive models of recently revised technical reports to reflect irradiation embrittlement, irradiation enhanced creep and void swelling behaviors. Benchmark analyses for a simple rod and detailed numerical analyses for a RVI sub-components assembly with welds were conducted via coupling the subroutines. With regard to the latter analyses, not only typical normal operating sequences but also three fuel cycles associated with specific core loading patterns were taken into account with a constant initial weld residual stress distribution. Adequacy of the subroutines was verified through analytical solutions generated from temperature and radiation damage dependent equations. Complex neutron dose, temperature and pressure profiles of the assembly were successfully determined from thermal and mechanical finite element analyses. Structural integrity assessment in terms of the ductility, embrittlement and irradiation assisted stress corrosion cracking susceptibility ratio after 40-years of operation led to acceptable in spite of the ARDMs. Contributions of factors and interaction effects were further quantified and discussed via design of experiment approach. • Five subroutines were developed to reflect irradiation induced degradations. • Uniform operating, fuel cycle loads and weld residual stress were incorporated. • The contribution of irradiation creep was the most influential to loss of ductility. • Stress relief reduced irradiation assisted stress corrosion cracking susceptibility. [ABSTRACT FROM AUTHOR]

It is known that irradiation damage can cause a reduction in fracture toughness and an increase in the strength of austenitic stainless steels used for the internal structures of nuclear power plant reactors. A change in the mechanical properties of material, such as fracture toughness and strength, affects the fracture behavior of the components. Actual reactor internal structures, such as the top guides and core shrouds in BWR plants, are irradiated from the core in a single direction. This causes the attenuation of neutron flux, which results in the distribution of neutron fluence in the components, meaning that the distribution of mechanical properties is in the through-wall direction of the components. It is therefore possible for the fracture mode of the components to change in the through-wall direction. It would be valuable for more accurate assessment of the structural integrity of the nuclear components to make the effect of the mechanical properties distribution on the fracture mode clear. However, it is difficult to prepare large-scale test specimens with the mechanical properties distribution induced by uni-directional neutron irradiation. In this study, the mechanical properties distribution was simulated by applying the gradient of the cold working ratio using Type 316L stainless steel plate with an uneven thickness. The targeted range of the cold working ratio was designed by assuming BWR reactor internal structures, such as the top guides. The mechanical properties distribution was evaluated based on the results of hardness, tensile and fracture toughness tests. It was found that the macroscopic mechanical properties of irradiated stainless steels can be simulated by using the cold working and the macroscopic distribution of the mechanical properties of irradiated stainless steels can also be simulated by using the cold working ratio with a gradient. Furthermore, fracture tests were performed using test plates with the mechanical properties distribution assuming irradiated components with a crack. In some cases, the maximum loads obtained by the tests were almost the same value in spite of the different cold working ratio in the ligament. Based on the results of the fracture tests, it was considered that the dominant mechanical properties for the fracture load were those at the tip of the slit, not those in the ligament. [ABSTRACT FROM AUTHOR]