This paper describes the evaluation of partial safety factors (PSF’s) for parameters related to flaw evaluation of pipes which have a circumferential surface flaw, and proposes the important matter which should be pay attention in the setup of the safety factors used in flaw evaluation. PSF’s were evaluated considering randomness of flaw size, a fracture resistance curve (J-R curve) and applied loads using load and resistance factor design method (LRFD). The limit state function is expressed by fracture resistance (resistance-related parameter) and applied J integral (load-related parameter). The measure parameters in the reliability assessment are the flaw size and the J-R curve, and PSF’s of them are larger than those of applied loads. Since the material properties used in the flaw evaluation are generally set to the engineering lower limit of their variation (e.g., 95% lower confidence limit), variation of the flaw size is considered to have important role on flaw evaluation. Therefore, when setting up the safely factors used in Rules on Fitness-for-Service (FFS), it is necessary to take into consideration not only the influence of variation of loads or material strength but the influence of variation of flaw size.
The Reduced-Moderation Water Reactor (RMWR) is being developed at Japan Atomic Energy Agency and demonstration of the core heat removal performance is one of the most important issues. However, operation of the full-scale bundle experiment is difficult technically because the fuel rod bundle size is larger, which consumes huge electricity. Hence, it is expected to develop an analysis code for simulating RMWR core thermal-hydraulic performance with high accuracy. Subchannel analysis is the most powerful technique to resolve the problem. A subchannel analysis code NASCA (Nuclear-reactor Advanced Sub-Channel Analysis code) has been developed to improve capabilities of analyzing transient two-phase flow phenomena, boiling transition (BT) and post BT, and the NASCA code is applicable on the thermal-hydraulic analysis for the current BWR fuel. In the present study, the prediction accuracy of the NASCA code has been investigated using the reduced-scale rod bundle test data, and its applicability on the RMWR has been improved by optimizing the mechanistic constitutive models.
In this study, severe accident analyses were conducted for the Advanced Boiling Water Reactor (ABWR) using a detailed nodalization and the impact of MAAP nodalization on the severe accident analysis was investigated. The results of the analysis obtained using a detailed nodalization shows that the PCV temperature is more challenging in comparison to the results of the original nodalization. However, it was found that the structural temperature of the PCV head flange can be maintained below the ultimate temperature of the head gasket when passive mitigation systems are successful, even if no reactor well injection is conducted.
This paper provides failure probability assessment results for piping systems affected by stress corrosion cracking (SCC) and pipe wall thinning in nuclear power plants. On the basis of the results, considerations for applying the leak-before-break (LBB) concept in actual plants are presented. The failure probability for SCC satisfies the target failure probability even if conservative conditions are assumed. Moreover, for pipe wall thinning analysis, pre-service inspection is important for satisfying the target failure probability because the initial wall thickness affects the accuracy of the wall thinning rate. The pipe wall thinning analysis revealed that the failure probability is higher than the target probability if the bending stress in the pipe is large.
This paper reports the results of the study on the failure modes and limit loads of piping in nuclear power plants subjected to cyclic seismic loading. By investigating the past fracture tests and earthquake resistance tests, it became clear that dominant failure mode of piping was fatigue, and the effect of ratchet strain was negligible. Until now, the stress generated with the acceleration of an earthquake was classified into the primary stress. However, the relationship between the input acceleration and the seismic response displacement of the pipe observed from earthquake resistance tests is non-linear, and increasing rate of displacement is lower than that of input acceleration in elastic-plastic stress condition. Therefore, the seismic loading can be treated as displacement controlled loading. To evaluate the reliability-based critical acceleration, a limit state function was defined taking the variations in the fatigue strength or some parameters into consideration. By using the limit state function, the reliability was evaluated for the typical piping of boiling water reactor (BWR) plants subjected to cyclic seismic loading, and a partial safety factors were calculated. Based on these results, a fatigue curve corresponding to the target reliability was proposed.
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