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Title: Mechanical properties of neutron-irradiated nickel-containing martensitic steels: II. Review and analysis of helium-effects studies

Authors:
 [1];  [1];  [1]
  1. ORNL
Publication Date:
Research Org.:
Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); High Flux Isotope Reactor; Metals-Processing Laboratory Users Facility; Shared Research Equipment Collaborative Research Center
Sponsoring Org.:
USDOE Office of Science (SC)
OSTI Identifier:
991169
DOE Contract Number:
DE-AC05-00OR22725
Resource Type:
Journal Article
Resource Relation:
Journal Name: Journal of Nuclear Materials; Journal Volume: 357; Journal Issue: 1-3
Country of Publication:
United States
Language:
English

Citation Formats

Klueh, Ronald L, Sokolov, Mikhail A, and Hashimoto, Naoyuki. Mechanical properties of neutron-irradiated nickel-containing martensitic steels: II. Review and analysis of helium-effects studies. United States: N. p., 2006. Web. doi:10.1016/j.jnucmat.2006.05.049.
Klueh, Ronald L, Sokolov, Mikhail A, & Hashimoto, Naoyuki. Mechanical properties of neutron-irradiated nickel-containing martensitic steels: II. Review and analysis of helium-effects studies. United States. doi:10.1016/j.jnucmat.2006.05.049.
Klueh, Ronald L, Sokolov, Mikhail A, and Hashimoto, Naoyuki. Sun . "Mechanical properties of neutron-irradiated nickel-containing martensitic steels: II. Review and analysis of helium-effects studies". United States. doi:10.1016/j.jnucmat.2006.05.049.
@article{osti_991169,
title = {Mechanical properties of neutron-irradiated nickel-containing martensitic steels: II. Review and analysis of helium-effects studies},
author = {Klueh, Ronald L and Sokolov, Mikhail A and Hashimoto, Naoyuki},
abstractNote = {},
doi = {10.1016/j.jnucmat.2006.05.049},
journal = {Journal of Nuclear Materials},
number = 1-3,
volume = 357,
place = {United States},
year = {Sun Jan 01 00:00:00 EST 2006},
month = {Sun Jan 01 00:00:00 EST 2006}
}
  • Microstructural evolutions in tempered martensitic steels (TMS) under neutron-irradiation, at fusion relevant He/dpa ratios and dpa rates, were characterized using a novel in situ He-implanter technique. F82H-mod3 was irradiated at 500 C in HFIR to a nominal 9 dpa and 190 or 380 appm He in both in the as-tempered (AT) and 20% cold-worked (CW) conditions. In all cases, a high number density of 1-2 nm He-bubbles were observed, along with fewer but larger 10 nm void-like faceted cavities. The He-bubbles form preferentially on dislocations and various interfaces. A slightly larger number of smaller He bubbles were observed in themore » CW condition. The lower He/dpa ratio produced slightly smaller and fewer He-bubbles. Comparisons of these observations to the results in nano-structured ferritic alloy (NFA) MA957 provide additional evidence that TMS may be susceptible to He-embrittlement as well as void swelling at fusion relevant He concentrations, while NFA are much more resistant to these degradation phenomena.« less
  • Chromium-molybdenum martensitic (ferritic) steels such as 9 Cr-1 Mo-V-Nb and 12 Cr-1 Mo-V-W are candidates for fast reactor and fusion reactor applications. In a fast reactor, the effect of neutron irradiation is caused by displacement damage, that is, by the interstitials and vacancies that are created by the high-energy neutrons. Increases in strength occur for irradiation up to [approximately]450 C. This hardening is largely attributed to the dislocation loops that form from the agglomeration of the interstitials. Precipitates that form during irradiation can also contribute to the hardening. At higher temperatures, most of the displacement damage anneals out. Irradiation effectsmore » expected in the first wall of a fusion reactor differ from those in a fast reactor. In addition to displacement damage, large amounts of transmutation helium will also be produced. The simultaneous effects of displacement damage and helium can be simulated by irradiating nickel-doped ferritic steels in a mixed-spectrum fission reactor. Helium is produced by transmutation reactions between thermal neutrons and nickel, and displacement damage is formed by the fast neutrons of the spectrum. Results using this technique indicate that hardening occurs as in a fast reactor, but the helium causes a strength increase in addition to that caused by displacement damage alone. This effect of helium could have a significant effect on other properties, especially toughness, and must be considered in the design of fusion reactors.« less
  • Tensile and creep properties have been determined on specimens of type 316 stainless steel irradiated in the High Flux Isotope Reactor in the range 380 to 785$sup 0$C. Irradiation of type 316 in this reaction partially simulates fusion reactor irradiation, with displacement damage levels up to 120 dpa and helium contents up to 6000 appM achieved in two years. Samples irradiated in the annealed condition to about 100 dpa and 4000 appM helium showed an increased yield strength between 350 and 600$sup 0$C and, except at 350$sup 0$C, a reduced ultimate tensile strength compared with values for the unirradiated material.more » Samples irradiated in the 20 percent-cold-worked condition showed decreases in both yield and ultimate tensile strengths at all test temperatures. The irradiated samples of both annealed and cold-worked material exhibited little strain hardening. Total elongations were small and became zero for tests at 650$sup 0$C. Tensile tests at 575$sup 0$C and creep-rupture tests at 550$sup 0$C showed strong effects of fluence on strength and ductility for helium contents above about 30 appM. Optical metallography showed extensive carbide precipitation at all temperatures and precipitation of a second phase, believed to be sigma, at the high temperatures. (NL)« less
  • 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 inmore » 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.« less