STP1329: The Actual Properties of WWER-440 Reactor Pressure Vessel Materials Obtained by Impact Tests of Subsize Specimens Fabricated out of Samples Taken from the RPV

    Korolev, YN
    Head of mechanical test Group, Head of Lab., Leading research scientist, Director and Deputy Director, ORM Russian Research Center “Kurchatov Institute”, Moscow,

    Kryukov, AM
    Head of mechanical test Group, Head of Lab., Leading research scientist, Director and Deputy Director, ORM Russian Research Center “Kurchatov Institute”, Moscow,

    Nikolaev, YA
    Head of mechanical test Group, Head of Lab., Leading research scientist, Director and Deputy Director, ORM Russian Research Center “Kurchatov Institute”, Moscow,

    Platonov, PA
    Head of mechanical test Group, Head of Lab., Leading research scientist, Director and Deputy Director, ORM Russian Research Center “Kurchatov Institute”, Moscow,

    Shtrombakh, YI
    Head of mechanical test Group, Head of Lab., Leading research scientist, Director and Deputy Director, ORM Russian Research Center “Kurchatov Institute”, Moscow,

    Langer, R
    Dipl. Engineer and Senior Engineer, Siemens AG KWU NT13, Erlangen,

    Leitz, C
    Dipl. Engineer and Senior Engineer, Siemens AG KWU NT13, Erlangen,

    Rieg, C-Y
    Chief de la Division Electricite de France, Immeuble France-Evry-Boukvard France — BP.128,

    Pages: 15    Published: Jan 1998


    Abstract

    This paper presents the results of a study on degradation due to irradiation occurring in WWER-440 reactor pressure vessel (RPV) steel, using subsize Charpy specimens (5×5×27.5 mm3). An application of subsize specimens for estimation of irradiation embrittlement of RPV steel has been substantiated. Comparison between ductile-to-brittle transition temperatures (DBTT) for full-size and subsize Charpy specimens has been carried out. The relation between these two values has been clarified. The results of testing trepans and templates cut out from WWER-440 reactor pressure vessels are considered. The main results of the program TACIS-91/1.1. are discussed.

    Keywords:

    reactor pressure vessel, radiation embrittlement, subsize impact test, ductile-to-brittle transition temperature, post irradiation annealing


    Paper ID: STP37989S

    Committee/Subcommittee: E10.02

    DOI: 10.1520/STP37989S


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