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 Article · Thu Aug 06 00:00:00 EDT 2015 · Journal of Nuclear Materials

DOI:https://doi.org/10.1016/j.jnucmat.2015.07.039· OSTI ID:1227101

Maloy, Stuart A. ^[1]; Saleh, Tarik A. ^[1]; Anderoglu, Osman ^[1]; Romero, Tobias J. ^[1]; Odette, G. Robert ^[2]; Yamamoto, Takuya ^[2]; Li, S. ^[2]; Cole, James I. ^[3]; Fielding, Randall ^[3]

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1. Los Alamos National Lab. (LANL), Los Alamos, NM (United States)
2. Univ. of California, Santa Barbara, CA (United States)
3. Idaho National Lab. (INL), Idaho Falls, ID (United States)

Tensile test results at 25 and 300 °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 °C in the Advanced Test Reactor (ATR), are reported. The engineering stress–strain curves are analyzed to provide true stress–strain constitutive σ(ε) laws for all of these alloys. In the irradiated condition, the σ(ε) fall into categories of: strain softening, nearly perfectly plastic and strain hardening. Increases in yield stress (Δσ_y) and reductions in uniform strain ductility (e_u) are observed, where as the latter can be understood in terms of the alloy's σ(ε) behavior. Increases in the average σ(ε) in the range of 0–10% strain are smaller than the corresponding Δσ_y, and vary more from alloy to alloy. The data are analyzed to establish relations between Δσ_y and coupled changes in the ultimate stresses as well as the effects of both test temperature and the unirradiated yield stress (σ_yu). The latter shows that higher σ_yu correlates with lower Δσ_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. As a result, notably lower interstitial solute contents correlate with improved ductility and homogenous deformation in broadly similar steels.

• Research Organization: Los Alamos National Lab. (LANL), Los Alamos, NM (United States)
• Sponsoring Organization: USDOE
• Grant/Contract Number: AC52-06NA25396; NU-11-3150; FG03-94ER54275
• OSTI ID: 1227101
• Alternate ID(s): OSTI ID: 1359391
• Report Number(s): LA-UR-15-23495; PII: S0022311515301264
• Journal Information: Journal of Nuclear Materials, Vol. 468, Issue C; ISSN 0022-3115
• Publisher: Elsevier
• Country of Publication: United States
• Language: English

Related Subjects

36 MATERIALS SCIENCE
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS
ferritic
irradiation
cladding
reactor