Creep properties of Zircaloy-4 for nuclear fuel cladding FEA simulation

    Research output: Chapter in Book/Report/Conference proceedingConference abstract in proceedingsScientific

    Abstract

    Zirconium alloys are commonly used as cladding tube material for nuclear water reactors. To improve the understanding of the creep damage accumulation in thin walled fuel cladding tubes made of Zircaloy-4, data collation (tensile, creep strain and rupture) and material modelling has been performed for use in finite element analysis (FEA). In literature there are two distinct areas of creep modelling: creep strain response to short power transients and long term creep strain evolution for storage purposes. In this paper the short term creep strain response is mainly targeted for FEA simulation of fuel-cladding interaction. In addition, the model performance in predicting long term creep strain is verified from the available public domain data. The creep rupture models are optimized for predicting biaxial deformation (hoop strain) of thin walled tubes. The relevant temperature range is selected for postulated system disturbances, i.e. power transients between 300 and 600°C. For the preliminary FEA simulations the material is assumed to be un-irradiated, cold worked and stress relived. The base material models (constitutive equations) do not at this stage incorporate the effect of anisotropy, however two methods of incorporating irradiation effect are presented. The main models applied for this work are the Wilshire equations (WE) for rupture and the logistic creep strain prediction (LCSP) model for strain.
    Original languageEnglish
    Title of host publicationBaltica IX. International Conference on Life Management and Maintenance for Power Plants - Abstracts
    EditorsPertti Auerkari, Juha Veivo
    Place of PublicationEspoo
    PublisherVTT Technical Research Centre of Finland
    Pages46
    ISBN (Print)978-951-38-8029-3, 978-951-38-8030-9
    Publication statusPublished - 2013
    MoE publication typeNot Eligible
    EventBALTICA IX - International Conference on Life Management and Maintenance for Power Plants - Helsinki-Stockholm, Finland
    Duration: 11 Jun 201313 Jun 2013

    Publication series

    SeriesVTT Technology
    Number107
    ISSN2242-1211

    Conference

    ConferenceBALTICA IX - International Conference on Life Management and Maintenance for Power Plants
    CountryFinland
    CityHelsinki-Stockholm
    Period11/06/1313/06/13

    Fingerprint

    Nuclear fuel cladding
    Creep
    Finite element method
    Zirconium alloys
    Constitutive equations
    Logistics
    Anisotropy

    Cite this

    Holmström, S., Andersson, T., Tulkki, V., & Laukkanen, A. (2013). Creep properties of Zircaloy-4 for nuclear fuel cladding FEA simulation. In P. Auerkari, & J. Veivo (Eds.), Baltica IX. International Conference on Life Management and Maintenance for Power Plants - Abstracts (pp. 46). Espoo: VTT Technical Research Centre of Finland. VTT Technology, No. 107
    Holmström, Stefan ; Andersson, Tom ; Tulkki, Ville ; Laukkanen, Anssi. / Creep properties of Zircaloy-4 for nuclear fuel cladding FEA simulation. Baltica IX. International Conference on Life Management and Maintenance for Power Plants - Abstracts. editor / Pertti Auerkari ; Juha Veivo. Espoo : VTT Technical Research Centre of Finland, 2013. pp. 46 (VTT Technology; No. 107).
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    title = "Creep properties of Zircaloy-4 for nuclear fuel cladding FEA simulation",
    abstract = "Zirconium alloys are commonly used as cladding tube material for nuclear water reactors. To improve the understanding of the creep damage accumulation in thin walled fuel cladding tubes made of Zircaloy-4, data collation (tensile, creep strain and rupture) and material modelling has been performed for use in finite element analysis (FEA). In literature there are two distinct areas of creep modelling: creep strain response to short power transients and long term creep strain evolution for storage purposes. In this paper the short term creep strain response is mainly targeted for FEA simulation of fuel-cladding interaction. In addition, the model performance in predicting long term creep strain is verified from the available public domain data. The creep rupture models are optimized for predicting biaxial deformation (hoop strain) of thin walled tubes. The relevant temperature range is selected for postulated system disturbances, i.e. power transients between 300 and 600°C. For the preliminary FEA simulations the material is assumed to be un-irradiated, cold worked and stress relived. The base material models (constitutive equations) do not at this stage incorporate the effect of anisotropy, however two methods of incorporating irradiation effect are presented. The main models applied for this work are the Wilshire equations (WE) for rupture and the logistic creep strain prediction (LCSP) model for strain.",
    author = "Stefan Holmstr{\"o}m and Tom Andersson and Ville Tulkki and Anssi Laukkanen",
    year = "2013",
    language = "English",
    isbn = "978-951-38-8029-3",
    series = "VTT Technology",
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    Holmström, S, Andersson, T, Tulkki, V & Laukkanen, A 2013, Creep properties of Zircaloy-4 for nuclear fuel cladding FEA simulation. in P Auerkari & J Veivo (eds), Baltica IX. International Conference on Life Management and Maintenance for Power Plants - Abstracts. VTT Technical Research Centre of Finland, Espoo, VTT Technology, no. 107, pp. 46, BALTICA IX - International Conference on Life Management and Maintenance for Power Plants, Helsinki-Stockholm, Finland, 11/06/13.

    Creep properties of Zircaloy-4 for nuclear fuel cladding FEA simulation. / Holmström, Stefan; Andersson, Tom; Tulkki, Ville; Laukkanen, Anssi.

    Baltica IX. International Conference on Life Management and Maintenance for Power Plants - Abstracts. ed. / Pertti Auerkari; Juha Veivo. Espoo : VTT Technical Research Centre of Finland, 2013. p. 46 (VTT Technology; No. 107).

    Research output: Chapter in Book/Report/Conference proceedingConference abstract in proceedingsScientific

    TY - CHAP

    T1 - Creep properties of Zircaloy-4 for nuclear fuel cladding FEA simulation

    AU - Holmström, Stefan

    AU - Andersson, Tom

    AU - Tulkki, Ville

    AU - Laukkanen, Anssi

    PY - 2013

    Y1 - 2013

    N2 - Zirconium alloys are commonly used as cladding tube material for nuclear water reactors. To improve the understanding of the creep damage accumulation in thin walled fuel cladding tubes made of Zircaloy-4, data collation (tensile, creep strain and rupture) and material modelling has been performed for use in finite element analysis (FEA). In literature there are two distinct areas of creep modelling: creep strain response to short power transients and long term creep strain evolution for storage purposes. In this paper the short term creep strain response is mainly targeted for FEA simulation of fuel-cladding interaction. In addition, the model performance in predicting long term creep strain is verified from the available public domain data. The creep rupture models are optimized for predicting biaxial deformation (hoop strain) of thin walled tubes. The relevant temperature range is selected for postulated system disturbances, i.e. power transients between 300 and 600°C. For the preliminary FEA simulations the material is assumed to be un-irradiated, cold worked and stress relived. The base material models (constitutive equations) do not at this stage incorporate the effect of anisotropy, however two methods of incorporating irradiation effect are presented. The main models applied for this work are the Wilshire equations (WE) for rupture and the logistic creep strain prediction (LCSP) model for strain.

    AB - Zirconium alloys are commonly used as cladding tube material for nuclear water reactors. To improve the understanding of the creep damage accumulation in thin walled fuel cladding tubes made of Zircaloy-4, data collation (tensile, creep strain and rupture) and material modelling has been performed for use in finite element analysis (FEA). In literature there are two distinct areas of creep modelling: creep strain response to short power transients and long term creep strain evolution for storage purposes. In this paper the short term creep strain response is mainly targeted for FEA simulation of fuel-cladding interaction. In addition, the model performance in predicting long term creep strain is verified from the available public domain data. The creep rupture models are optimized for predicting biaxial deformation (hoop strain) of thin walled tubes. The relevant temperature range is selected for postulated system disturbances, i.e. power transients between 300 and 600°C. For the preliminary FEA simulations the material is assumed to be un-irradiated, cold worked and stress relived. The base material models (constitutive equations) do not at this stage incorporate the effect of anisotropy, however two methods of incorporating irradiation effect are presented. The main models applied for this work are the Wilshire equations (WE) for rupture and the logistic creep strain prediction (LCSP) model for strain.

    M3 - Conference abstract in proceedings

    SN - 978-951-38-8029-3

    SN - 978-951-38-8030-9

    T3 - VTT Technology

    SP - 46

    BT - Baltica IX. International Conference on Life Management and Maintenance for Power Plants - Abstracts

    A2 - Auerkari, Pertti

    A2 - Veivo, Juha

    PB - VTT Technical Research Centre of Finland

    CY - Espoo

    ER -

    Holmström S, Andersson T, Tulkki V, Laukkanen A. Creep properties of Zircaloy-4 for nuclear fuel cladding FEA simulation. In Auerkari P, Veivo J, editors, Baltica IX. International Conference on Life Management and Maintenance for Power Plants - Abstracts. Espoo: VTT Technical Research Centre of Finland. 2013. p. 46. (VTT Technology; No. 107).