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    Structural integrity assessment of a nuclear vessel through ASME and master curve approaches using irradiation embrittlement predictions

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    Identificadores
    URI: http://hdl.handle.net/10902/19364
    DOI: 10.1520/JTE20180569
    ISSN: 0090-3973
    ISSN: 1945-7553
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    Autoría
    Ferreño Blanco, DiegoAutoridad Unican; Fernández-Rucoba, David; Casado del Prado, José AntonioAutoridad Unican; Báscones, Roberto; Ruiz Martínez, EstelaAutoridad Unican; Rodríguez, Álvaro; Miguel, Roberto; Poncela, Enrique Gomez
    Fecha
    2020-11
    Derechos
    © ASTM
    Publicado en
    Journal of testing and evaluation Volume 48, Issue 6, 1 November 2020
    Editorial
    American Society for Testing and Materials
    Enlace a la publicación
    10.1520/JTE20180569
    Resumen/Abstract
    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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    UNIVERSIDAD DE CANTABRIA

    Repositorio realizado por la Biblioteca Universitaria utilizando DSpace software
    Contacto | Sugerencias
    Metadatos sujetos a:licencia de Creative Commons Reconocimiento 4.0 España