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dc.contributor.authorFerreño Blanco, Diego 
dc.contributor.authorFernández-Rucoba, David
dc.contributor.authorCasado del Prado, José Antonio 
dc.contributor.authorBáscones, Roberto
dc.contributor.authorRuiz Martínez, Estela 
dc.contributor.authorRodríguez, Álvaro
dc.contributor.authorMiguel, Roberto
dc.contributor.authorPoncela, Enrique Gomez
dc.contributor.otherUniversidad de Cantabriaes_ES
dc.date.accessioned2020-10-20T08:06:11Z
dc.date.available2021-11-30T03:45:11Z
dc.date.issued2020-11
dc.identifier.issn0090-3973
dc.identifier.issn1945-7553
dc.identifier.urihttp://hdl.handle.net/10902/19364
dc.description.abstractABSTRACT: 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.es_ES
dc.description.sponsorshipThis project was carried out with the financial support of Sociedad para el Desarrollo Regional de Cantabria (SODERCAN) and Equipos Nucleares S.A. (ENSA), to whom the authors would like to express their gratitude.es_ES
dc.format.extent20 p.es_ES
dc.language.isoenges_ES
dc.publisherAmerican Society for Testing and Materialses_ES
dc.rights© ASTMes_ES
dc.sourceJournal of testing and evaluation Volume 48, Issue 6, 1 November 2020es_ES
dc.titleStructural integrity assessment of a nuclear vessel through ASME and master curve approaches using irradiation embrittlement predictionses_ES
dc.typeinfo:eu-repo/semantics/articlees_ES
dc.relation.publisherVersion10.1520/JTE20180569es_ES
dc.rights.accessRightsopenAccesses_ES
dc.identifier.DOI10.1520/JTE20180569
dc.type.versionpublishedVersiones_ES


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