Review of The Test Programs Performed on Trepan Materials Taken from Decommissioned Reactor Pressure Vessels

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    Abstract

    The number of decommissioned nuclear reactors is constantly increasing while in the meantime long term operation is considered for many other reactors. Some vessels offer an opportunity not only to validate the current regulatory procedures based on surveillance data but also to explore a number of issues that are critical for long term operation. This report aims to review the literature data dedicated to testing of materials taken from decommissioned vessels. It also highlights the main outcome as well as identify gaps or issues that should or could be addressed. Six reactor pressure vessels were considered: BR3 (B), Chooz-A (F-B), Greisfwald and Gundremmingen (D), Novovoronesh (RU) and Trawsfynydd (UK). The test program efforts were not all equal in terms of addressed issues and were often limited to a couple of objectives. In particular, the main focus was to determine the actual embrittlement state of the vessels in comparison to expectations. An international test program on decommissioned reactor vessels will aim to resolve a number of issues that are important for RPV ageing and embrittlement assessment. Focusing on the RPV materials, several possible R&D topics were identified. • Chemical composition : distribution in the vessel, presence of segregated zones . • Neutron exposure : dosimetry versus calculations, through-wall thickness attenuation, ex-vessel dosimetry. • Mechanical properties : comparison to surveillance data and accelerated MTR data, effect of WPS, direct fracture toughness measurements. • Microstructural characterization • Thermal ageing : long term exposure versus accelerated data. • Flux/spectrum effects : vessel versus surveillance versus MTR. • Validation of the regulatory procedures in particular in the presence of non-homogeneities. Finally, some key specifications that ideally should be considered in an international test program were outlined, in particular with respect to materials (forging, weld and heat affected zone), their chemical composition (Cu, Ni, P contents), their neutron exposure and associated dosimetry, their irradiation conditions (temperature, fluence), the results of the surveillance programs and the availability of the archive materials. All these elements would greatly contribute to a better understanding of reactor vessel embrittlement to the benefit of the regulatory authorities as well as the utilities and research organizations.
    Original languageEnglish
    PublisherSCK CEN
    Number of pages52
    StatePublished - 28 Apr 2017

    Publication series

    NameSCK•CEN Reports
    PublisherStudiecentrum voor Kernenergie
    No.BLG-1125

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