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Title: 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

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 ofmore » 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.« less
 [1] ;  [1] ;  [1] ;  [1] ;  [2] ;  [2] ;  [2] ;  [3] ;  [3]
  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)
Publication Date:
Report Number(s):
Journal ID: ISSN 0022-3115; PII: S0022311515301264
Grant/Contract Number:
AC52-06NA25396; NU-11-3150; FG03-94ER54275
Accepted Manuscript
Journal Name:
Journal of Nuclear Materials
Additional Journal Information:
Journal Volume: 468; Journal Issue: C; Journal ID: ISSN 0022-3115
Research Org:
Los Alamos National Lab. (LANL), Los Alamos, NM (United States)
Sponsoring Org:
Country of Publication:
United States
36 MATERIALS SCIENCE; 11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS; ferritic; irradiation; cladding; reactor
OSTI Identifier:
Alternate Identifier(s):
OSTI ID: 1359391