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As a part of round robin analysis to develop a finite element elastic-plastic seismic analysis procedure for nuclear safety class 1 components, a series of parametric analyses was carried out on the simulated pressurizer surge line system model to investigate sensitivity of the analysis results to finite element analysis variables. The analysis on the surge line system model considered dynamic effect due to the seismic load corresponding to PGA 0.6 g and elastic-plastic material behavior based on the Chaboche combined hardening model. From the parametric analysis results, it was found that strains such as accumulated equivalent plastic strain and equivalent plastic strain are more sensitive to the analysis variables than von Mises effect stress. The parametric analysis results also identified that finite element density and ovalization option in the elbow elements have more significant effect on the analysis results than the other variables.

The NUREG/CR-6909, which had provided the technical background for the Environmental Fatigue Evaluation (EFE), was revised in May 2018 to incorporate the Japan Nuclear Energy Safety organization (JNES)’s extensive fatigue ε-N data and to address concerns from interested stakeholders over the original NUREG/CR-6909. But the revised NUREG/CR-6909 sets limit of the temperature ranges to 325°C for the application of the environmental fatigue correction factor (Fen). Therefore, the revised NUREG/CR-6909-based EFE is not applicable to the materials whose temperatures are higher than 325°C. In this paper, Monte Carlo simulations have been performed to determine the transformed temperatures over 325°C for the materials which are not covered in the revised NUREG/CR-6909. To determine the distributions of fatigue lives in LWR environments, Monte Carlo simulations were performed using the fatigue life models in the LWR environments and transformed parameters provided in the revised NUREG/CR-6909 and JNES-SS-1005. The transformed parameters which affect the fatigue lives in the LWR environments were selected randomly using the uniform distribution. And a total of 100,000 simulations were performed for each material. The distributions of fatigue lives in LWR environments obtained from Monte Carlo simulations were fit to lognormal curves to estimate the cumulative distributions. The estimated cumulative distributions were individually determined in the temperature ranges of 100 ∼ 325°C and 325 ∼ 350°C to investigate the temperature effects (FT,effect) over 325°C. The temperature effect is defined as ratio of fatigue lives in the LWR environments in the temperature range of 100 ∼ 325°C to that in the temperature range of 325 ∼ 350°C in this paper. As a result of reviewing the temperature effects between the revised NUREG/CR-6909 and JNES-SS-1005, the transformed temperatures over 325°C for each material have been suggested for the application of the revised NUREG/CR-6909.

A numerical analysis has been performed to estimate the effect of thermal stratification in the safety injection piping system. The Direct Vessel Injection (DVI) system is used to perform the functions of Emergency Core Cooling and Residual Heat Removal for an APR1400 nuclear power plant (Korea’s Advanced Power Reactor 1400 MW-Class). The thermal stratification is anticipated in the horizontally routed piping between the DVI nozzle of the reactor vessel and the first isolation valve. Non-axisymmetric temperature distribution across the pipe diameter induced by the thermal stratification leads to differential thermal growth of the piping causing the global bending stress and local stress. Thermal hydraulic analysis has been performed to determine the temperature distribution in the DVI piping due to the thermal stratification. Piping stress analysis has also been carried out to evaluate the integrity of the DVI piping using the thermal hydraulic analysis results. This paper provides a methodology for calculating the global bending stresses and local stresses induced by the thermal stratification in the DVI piping and for performing fatigue evaluation based on Subsection NB-3600 of ASME Section III.Copyright © 2012 by ASME

The seismic analysis for the OPR1000 (previously entitled Korea Standard Nuclear Power Plant, now Optimized Power Reactor 1000 MW-Class) was carried out based on the SSE to assure the integrity of the reactor coolant pressure boundary and to secure a safe shutdown of reactor as well as the OBE to ensure the functionality of all the necessary systems for the continued operation of the reactor without undue risk to the health and safety of the public. In fact, the OBE with some supplementary functions, however, tends to control the design of nuclear power plants instead of the SSE due to its conservative analysis methodology. As a result of analysis work by means of more sophisticated techniques and continuously raised questions by the nuclear industry about the adequacy of using the OBE with at least one-half the maximum vibratory ground acceleration of the SSE as a design input, the OBE elimination design is exceptionally allowed through the alternative design rule by KINS (Korea Institute of Nuclear Safety: Independent nuclear regulatory expert organization), as long as the OBE is established as less than or equal to one-third of the SSE. Evaluated in this paper was the effect of the application of the OBE elimination-related alternative design rule on the KEPIC (Korea Electric Power Industry Code)/ASME Code Class 1 piping systems of an OPR1000-based nuclear power plant. The SSE (e.g. 0.3g) at least three-folds larger than the OBE (e.g. 0.1g) ground acceleration could decrease the flexibility of piping systems by adding more restraints to mitigate the increased SSE-induced seismic loads. Therefore, the flexibility of four piping subsystems, whose acquisition is one of the main purposes of the OBE elimination, was evaluated for an APR1400-based nuclear power plant (Korea’s Advanced Power Reactor 1400 MW-Class), which has practically adopted the application of the OBE elimination for the first time in Korea.Copyright © 2010 by ASME

The fatigue effects have to be revaluated in preparing the license renewal application for the continued operation of an old vintage nuclear power plant. This paper presents a complete fatigue analysis for a branch piping with the effect of thermal stratification, induced by turbulent penetration, and environmental factors on fatigue. Three-dimensional computational fluid dynamics and finite element analyses were performed for the branch line to evaluate the thermal stratification loading. Proposed is a supplementary methodology of considering the effect of environmental factor on the combined conventional peak stress intensity range, based on the NB-3600 of ASME Section III Code, with thermal stratification loading. It can be used for safety enhancement of old vintage nuclear power plants.Copyright © 2006 by ASME