Structural Integrity Assessment of a Nuclear Vessel through ASME and Master Curve Approaches Using Irradiation Embrittlement Predictions

脆化 材料科学 断裂韧性 反应堆压力容器 结构完整性 韧性 核工程 脆性 结构材料 压力容器 复合材料 法律工程学 结构工程 冶金 工程类
作者
Diego Ferreño,David Fernández-Rucoba,J.A. Casado,Roberto Bascones,Estela Ruiz,Álvaro Rodríguez-Prieto,Roberto Carlos Miguel,Enrique G. Poncela
出处
期刊:Journal of Testing and Evaluation [ASM International]
卷期号:48 (6): 4748-4766 被引量:1
标识
DOI:10.1520/jte20180569
摘要

Abstract The assessment of the structural integrity of nuclear vessels is based on a series of procedures developed in the 1970s and 1980s. On one hand, curves that, according to the American Society of Mechanical Engineers code, describe the relationship between steel toughness and temperature in the ductile-to-brittle transition region, based on the reference temperature concept RTNDT, were adopted in 1972. On the other hand, the material embrittlement derived from the exposure of steel to neutron irradiation is determined through the model included in “Regulatory Guide 1.99 Rev. 2,” published in 1988. Since then, there have been enormous advances in this field. For example, the Master Curve, based on the reference temperature T0, describes the relationship between toughness and temperature in the transition zone more realistically and with much more robust microstructural and mechanical foundations and uses the elastic-plastic fracture toughness KJc. Moreover, improved models have been developed to estimate the embrittlement of steel subjected to neutron irradiation, such as ASTM E900, Standard Guide for Predicting Radiation-Induced Transition Temperature Shift in Reactor Vessel Materials. This study is aimed at comparing the results obtained using traditional procedures to the improved alternatives developed later. For this purpose, the behavior of the steel of a nuclear vessel that is currently under construction has been experimentally characterized through RTNDT and T0 parameters. In addition, the material embrittlement has been quantified using “Regulatory Guide 1.99 Rev. 2” and ASTM E900. These experimental results have been transferred to the assessment of the structural integrity of the vessel to determine the pressure-temperature limit curves and size of the maximum admissible defect as a function of the operation time of the plant. The results have allowed the implicit overconservatism present in the traditional procedures to be quantified.

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