Characterization and comparative analysis of the tensile properties of five tempered martensitic steels and an oxide dispersion strengthened ferritic alloy irradiated at ≈295 °C to ≈6.5 dpa

JOURNAL OF NUCLEAR MATERIALS

卷:468

Article

Maloy, S. A.¹  Saleh, T. A.²  Anderoglu, O.¹  Romero, T. J.³  Odette, G. R.⁴  Yamamoto, T.⁴  Li, S.⁴  Cole, J. I.⁵  Fielding, R.⁶ 

[1] Los Alamos Natl Lab, MST 8, Los Alamos, NM 87545 USA

[2] Los Alamos Natl Lab, MST 16, Los Alamos, NM 87545 USA

[3] Los Alamos Natl Lab, C IIAC, Los Alamos, NM 87545 USA

[4] Univ Calif Santa Barbara, Dept Mech Engn, Santa Barbara, CA 93106 USA

[5] Idaho Natl Lab, Nucl Sci User Facil, Idaho Falls, ID 83415 USA

[6] Idaho Natl Lab, Fuel Fabricat & Characterizat Dept, Idaho Falls, ID 83415 USA

关键词: Ferritic;    Irradiation;    Cladding;    Reactor;   

DOI  :  10.1016/j.jnucmat.2015.07.039

来源: Elsevier

【 摘 要 】

Tensile test results at 25 and 300 degrees C on five 9-12Cr tempered martensitic steels and one 14Cr oxide dispersion strengthened alloy, that were side-by side irradiated to 6.5 dpa at 295 degrees C in the Advanced Test Reactor (ATR), are reported. The engineering stress strain curves are analyzed to provide true stress-strain constitutive sigma(epsilon) laws for all of these alloys. In the irradiated condition, the sigma(epsilon) fall into categories of: strain softening, nearly perfectly plastic and strain hardening. Increases in yield stress (Delta sigma(y)) and reductions in uniform strain ductility (e(u)) are observed, where the latter can be understood in terms of the alloy's sigma(epsilon) behavior. Increases in the average sigma(epsilon) in the range of 0-10% strain are smaller than the corresponding Delta sigma(y), and vary more from alloy to alloy. The data are also analyzed to establish relations between Delta sigma(y), and coupled changes in the ultimate stresses as well as the effects of both test temperature and the unirradiated yield stress (sigma(yu)). The latter shows that higher sigma(yu) correlates with lower Delta sigma(y). In five out of six cases the effects of irradiation are generally consistent with previous observations on these alloys. However, the particular heat of the 12Cr HT-9 tempered martensitic steel in this study has a much higher e(u) than observed for earlier heats. The reasons for this improved behavior are not understood and may be microstructural in origin. However, it is noted that the new heat of HT-9, which was procured under modern quality assurance standards, has lower interstitial nitrogen than previous heats. Notably lower interstitial solute contents correlate with improved ductility and homogenous deformation in broadly similar steels. Published by Elsevier B.V.

【 授权许可】

Free