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

NameVTT Technology
PublisherVTT
Number107
ISSN (Print)2242-1211
ISSN (Electronic)2242-122X

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).
@inbook{056bfaba956f4b4d80e0923ed61037ba,
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",
publisher = "VTT Technical Research Centre of Finland",
number = "107",
pages = "46",
editor = "Pertti Auerkari and Juha Veivo",
booktitle = "Baltica IX. International Conference on Life Management and Maintenance for Power Plants - Abstracts",
address = "Finland",

}

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).