Creep failure of a reactor pressure vessel lower head under severe accident conditions

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A severe accident in a nuclear power plant could result in the relocation of large quantities of molten core material onto the lower head of he reactor pressure vessel (RPV). In the absence of inherent cooling mechanisms, failure of the RPV ultimately becomes possible under the combined effects of system pressure and the thermal heat-up of the lower head. Sandia National Laboratories has performed seven experiments at 1:5th scale simulating creep failure of a RPV lower head. This paper describes a modeling program that complements the experimental program. Analyses have been performed using the general-purpose finite-element code ABAQUS-5.6. In order ... continued below

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8 p.

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Pilch, M. M.; Ludwigsen, J. S.; Chu, T. Y. & Rashid, Y. R. August 1998.

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  • Sandia National Laboratories
    Publisher Info: Sandia National Labs., Albuquerque, NM (United States)
    Place of Publication: Albuquerque, New Mexico

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Description

A severe accident in a nuclear power plant could result in the relocation of large quantities of molten core material onto the lower head of he reactor pressure vessel (RPV). In the absence of inherent cooling mechanisms, failure of the RPV ultimately becomes possible under the combined effects of system pressure and the thermal heat-up of the lower head. Sandia National Laboratories has performed seven experiments at 1:5th scale simulating creep failure of a RPV lower head. This paper describes a modeling program that complements the experimental program. Analyses have been performed using the general-purpose finite-element code ABAQUS-5.6. In order to make ABAQUS solve the specific problem at hand, a material constitutive model that utilizes temperature dependent properties has been developed and attached to ABAQUS-executable through its UMAT utility. Analyses of the LHF-1 experiment predict instability-type failure. Predicted strains are delayed relative to the observed strain histories. Parametric variations on either the yield stress, creep rate, or both (within the range of material property data) can bring predictions into agreement with experiment. The analysis indicates that it is necessary to conduct material property tests on the actual material used in the experimental program. The constitutive model employed in the present analyses is the subject of a separate publication.

Physical Description

8 p.

Notes

INIS; OSTI as DE98005080

Source

  • 1998 ASME/JSME joint pressure vessel and piping (PVP) conference, San Diego, CA (United States), 26-30 Jul 1998

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  • Other: DE98005080
  • Report No.: SAND--98-1735C
  • Report No.: CONF-980708--
  • Grant Number: AC04-94AL85000
  • Office of Scientific & Technical Information Report Number: 656714
  • Archival Resource Key: ark:/67531/metadc702863

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Office of Scientific & Technical Information Technical Reports

Reports, articles and other documents harvested from the Office of Scientific and Technical Information.

Office of Scientific and Technical Information (OSTI) is the Department of Energy (DOE) office that collects, preserves, and disseminates DOE-sponsored research and development (R&D) results that are the outcomes of R&D projects or other funded activities at DOE labs and facilities nationwide and grantees at universities and other institutions.

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Creation Date

  • August 1998

Added to The UNT Digital Library

  • Sept. 12, 2015, 6:31 a.m.

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  • Oct. 3, 2017, 4:27 p.m.

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Pilch, M. M.; Ludwigsen, J. S.; Chu, T. Y. & Rashid, Y. R. Creep failure of a reactor pressure vessel lower head under severe accident conditions, article, August 1998; Albuquerque, New Mexico. (digital.library.unt.edu/ark:/67531/metadc702863/: accessed June 23, 2018), University of North Texas Libraries, Digital Library, digital.library.unt.edu; crediting UNT Libraries Government Documents Department.