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Title: Behavior of an oxide dispersion strengthened ferritic steel irradiated in Phenix

Abstract

This paper deals with the irradiation behavior of the oxide dispersion strengthened (ODS) ferritic alloy DT2203Y05, a 13% Cr ferritic alloy strengthened by a fine dispersion of yttrium and titanium oxides. This alloy was irradiated up to 81 dpa in Phenix as fuel pin cladding. The profilometry measurements confirm its high swelling resistance. Few voids, mainly associated with oxides, are observed at low irradiation temperatures, but this alloy is severely embrittled by irradiation. A few cracks are observed in the lower 2/3 of the fissile column, and the longitudinal tensile tests show hardening and severe ductility loss induced by irradiation along the whole fuel column. Transmission electron microscopy observations show that the radiation-induced strong ductility loss results from a fine and uniform {alpha}{prime} precipitation and dislocation loop formation at low temperatures (as in classic ferritic steels) and from important {chi} precipitation at high temperatures. The radiation-induced embrittlement of this ODS ferritic steel results also from the recoil dissolution of oxides leading to the total dissolution of the finer oxides and to the formation of small oxide shells around the larger initial oxides. The oxide dispersion, which is characterized before irradiation by a broad size distribution with oxide-free bands leading tomore » an heterogeneous dislocation loop distribution under irradiation, is probably responsible for the apparition of cracks in the lower 2/3 of the fuel column and for the channel fracture.« less

Authors:
; ; ; ;  [1]
  1. CEA Saclay, Gif sur Yvette (France)
Publication Date:
OSTI Identifier:
679527
Report Number(s):
CONF-960643-
Journal ID: ISSN 1050-7515; TRN: 99:009460
Resource Type:
Conference
Resource Relation:
Conference: 18. symposium on effects of radiation on materials, Hyannis, MA (United States), 25-27 Jun 1996; Other Information: PBD: 1999; Related Information: Is Part Of Effects of radiation on materials: 18. international symposium; Nanstad, R.K. [ed.] [Oak Ridge National Lab., TN (United States)]; Hamilton, M.L.; Garner, F.A. [eds.] [Pacific Northwest National Lab., Richland, WA (United States)]; Kumar, A.S. [ed.] [Univ. of Missouri, Rolla, MO (United States)]; PB: 1261 p.; ASTM special technical publication 1325
Country of Publication:
United States
Language:
English
Subject:
36 MATERIALS SCIENCE; 21 NUCLEAR POWER REACTORS AND ASSOCIATED PLANTS; PHENIX REACTOR; PHYSICAL RADIATION EFFECTS; YTTRIUM OXIDES; TITANIUM OXIDES; STEEL-CR13; FUEL CANS; SWELLING; TENSILE PROPERTIES; EMBRITTLEMENT; MICROSTRUCTURE

Citation Formats

Dubuisson, P, Schill, R, Hugon, M P, Grislin, I, and Seran, J L. Behavior of an oxide dispersion strengthened ferritic steel irradiated in Phenix. United States: N. p., 1999. Web.
Dubuisson, P, Schill, R, Hugon, M P, Grislin, I, & Seran, J L. Behavior of an oxide dispersion strengthened ferritic steel irradiated in Phenix. United States.
Dubuisson, P, Schill, R, Hugon, M P, Grislin, I, and Seran, J L. Fri . "Behavior of an oxide dispersion strengthened ferritic steel irradiated in Phenix". United States.
@article{osti_679527,
title = {Behavior of an oxide dispersion strengthened ferritic steel irradiated in Phenix},
author = {Dubuisson, P and Schill, R and Hugon, M P and Grislin, I and Seran, J L},
abstractNote = {This paper deals with the irradiation behavior of the oxide dispersion strengthened (ODS) ferritic alloy DT2203Y05, a 13% Cr ferritic alloy strengthened by a fine dispersion of yttrium and titanium oxides. This alloy was irradiated up to 81 dpa in Phenix as fuel pin cladding. The profilometry measurements confirm its high swelling resistance. Few voids, mainly associated with oxides, are observed at low irradiation temperatures, but this alloy is severely embrittled by irradiation. A few cracks are observed in the lower 2/3 of the fissile column, and the longitudinal tensile tests show hardening and severe ductility loss induced by irradiation along the whole fuel column. Transmission electron microscopy observations show that the radiation-induced strong ductility loss results from a fine and uniform {alpha}{prime} precipitation and dislocation loop formation at low temperatures (as in classic ferritic steels) and from important {chi} precipitation at high temperatures. The radiation-induced embrittlement of this ODS ferritic steel results also from the recoil dissolution of oxides leading to the total dissolution of the finer oxides and to the formation of small oxide shells around the larger initial oxides. The oxide dispersion, which is characterized before irradiation by a broad size distribution with oxide-free bands leading to an heterogeneous dislocation loop distribution under irradiation, is probably responsible for the apparition of cracks in the lower 2/3 of the fuel column and for the channel fracture.},
doi = {},
journal = {},
issn = {1050-7515},
number = ,
volume = ,
place = {United States},
year = {1999},
month = {10}
}

Conference:
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