Creep-fatigue design of Gen IV high temperature reactor plants

    Research output: Chapter in Book/Report/Conference proceedingChapter or book articleProfessional

    Abstract

    The service conditions expected for the GEN IV reactors pose significant challenges to structural materials selection and qualification efforts. The components will undergo varied service conditions which include exposure to high temperatures and high neutron doses compared to the service conditions in today's commercial reactors. A considerable damage mechanism for materials used in GEN IV components is predicted to be creep-fatigue (CF) damage, which arises due to startup and shutdown or power transients during normal operation. Design codes, such as RCC-MRx and ASME III NH, for GEN IV nuclear reactors use interaction diagram based method for CF assessment. In the interaction diagram the fatigue damage is expressed as the ratio of design cycles over the allowable amount of cycles in service and the creep damage as the ratio of time in service over the design life. When standard laboratory CF tests with relatively large strain ranges and short hold or creep periods are assessed using the design code CF assessment procedures, the design codes seem to provide sufficient level of conservatism for safe design. But does the sufficient level of conservatism still remain with parameters relevant to power plant conditions? Can this type of test results and assessment and modelling methods emerging from them be extrapolated to GEN IV relevant conditions, where the stress and strain levels are lower, hold periods are significantly longer and lifetimes of components are expected to be about 60 years or more? Are there enough data and information on the long-term microstructural evolution and its effect on the creep and cyclic properties of e.g. P91 steel?
    Original languageEnglish
    Title of host publicationMaterials Science and Technology - Nuclear Materials, Advanced Course
    EditorsHannu Hänninen, Timo Kiesi
    Place of PublicationHelsinki
    PublisherAalto University
    Pages65-77
    ISBN (Print)978-952-60-6579-3
    Publication statusPublished - 2015
    MoE publication typeD2 Article in professional manuals or guides or professional information systems or text book material

    Publication series

    SeriesAalto University Publication Series Science + Technology
    Number16/2015
    ISSN1799-4896

    Fingerprint

    High temperature reactors
    Creep
    Fatigue of materials
    Fatigue damage
    Microstructural evolution
    Nuclear reactors
    Power plants
    Neutrons
    Steel

    Keywords

    • Gen IV reactors
    • design codes
    • creep-fatigue

    Cite this

    Pohja, R. (2015). Creep-fatigue design of Gen IV high temperature reactor plants. In H. Hänninen, & T. Kiesi (Eds.), Materials Science and Technology - Nuclear Materials, Advanced Course (pp. 65-77). Helsinki: Aalto University. Aalto University Publication Series Science + Technology, No. 16/2015
    Pohja, Rami. / Creep-fatigue design of Gen IV high temperature reactor plants. Materials Science and Technology - Nuclear Materials, Advanced Course. editor / Hannu Hänninen ; Timo Kiesi. Helsinki : Aalto University, 2015. pp. 65-77 (Aalto University Publication Series Science + Technology; No. 16/2015).
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    abstract = "The service conditions expected for the GEN IV reactors pose significant challenges to structural materials selection and qualification efforts. The components will undergo varied service conditions which include exposure to high temperatures and high neutron doses compared to the service conditions in today's commercial reactors. A considerable damage mechanism for materials used in GEN IV components is predicted to be creep-fatigue (CF) damage, which arises due to startup and shutdown or power transients during normal operation. Design codes, such as RCC-MRx and ASME III NH, for GEN IV nuclear reactors use interaction diagram based method for CF assessment. In the interaction diagram the fatigue damage is expressed as the ratio of design cycles over the allowable amount of cycles in service and the creep damage as the ratio of time in service over the design life. When standard laboratory CF tests with relatively large strain ranges and short hold or creep periods are assessed using the design code CF assessment procedures, the design codes seem to provide sufficient level of conservatism for safe design. But does the sufficient level of conservatism still remain with parameters relevant to power plant conditions? Can this type of test results and assessment and modelling methods emerging from them be extrapolated to GEN IV relevant conditions, where the stress and strain levels are lower, hold periods are significantly longer and lifetimes of components are expected to be about 60 years or more? Are there enough data and information on the long-term microstructural evolution and its effect on the creep and cyclic properties of e.g. P91 steel?",
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    Pohja, R 2015, Creep-fatigue design of Gen IV high temperature reactor plants. in H Hänninen & T Kiesi (eds), Materials Science and Technology - Nuclear Materials, Advanced Course. Aalto University, Helsinki, Aalto University Publication Series Science + Technology, no. 16/2015, pp. 65-77.

    Creep-fatigue design of Gen IV high temperature reactor plants. / Pohja, Rami.

    Materials Science and Technology - Nuclear Materials, Advanced Course. ed. / Hannu Hänninen; Timo Kiesi. Helsinki : Aalto University, 2015. p. 65-77 (Aalto University Publication Series Science + Technology; No. 16/2015).

    Research output: Chapter in Book/Report/Conference proceedingChapter or book articleProfessional

    TY - CHAP

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    AB - The service conditions expected for the GEN IV reactors pose significant challenges to structural materials selection and qualification efforts. The components will undergo varied service conditions which include exposure to high temperatures and high neutron doses compared to the service conditions in today's commercial reactors. A considerable damage mechanism for materials used in GEN IV components is predicted to be creep-fatigue (CF) damage, which arises due to startup and shutdown or power transients during normal operation. Design codes, such as RCC-MRx and ASME III NH, for GEN IV nuclear reactors use interaction diagram based method for CF assessment. In the interaction diagram the fatigue damage is expressed as the ratio of design cycles over the allowable amount of cycles in service and the creep damage as the ratio of time in service over the design life. When standard laboratory CF tests with relatively large strain ranges and short hold or creep periods are assessed using the design code CF assessment procedures, the design codes seem to provide sufficient level of conservatism for safe design. But does the sufficient level of conservatism still remain with parameters relevant to power plant conditions? Can this type of test results and assessment and modelling methods emerging from them be extrapolated to GEN IV relevant conditions, where the stress and strain levels are lower, hold periods are significantly longer and lifetimes of components are expected to be about 60 years or more? Are there enough data and information on the long-term microstructural evolution and its effect on the creep and cyclic properties of e.g. P91 steel?

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    A2 - Hänninen, Hannu

    A2 - Kiesi, Timo

    PB - Aalto University

    CY - Helsinki

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    Pohja R. Creep-fatigue design of Gen IV high temperature reactor plants. In Hänninen H, Kiesi T, editors, Materials Science and Technology - Nuclear Materials, Advanced Course. Helsinki: Aalto University. 2015. p. 65-77. (Aalto University Publication Series Science + Technology; No. 16/2015).