Fatigue properties of F82H irradiated at 523 K to 3.8 dpa

Miwa, Yukio; Jitsukawa, Shiro; Yonekawa, Minoru

doi:10.1016/j.jnucmat.2004.04.029

Cited by 15 publications

(18 citation statements)

References 22 publications

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“…Post-irradiation fatigue tests were performed for F82H at room temperature. 39) Irradiation was carried out in the JMTR to 3.8 dpa at 523 K. Neutron irradiation did not cause any changes in fatigue life, except for the results with smallest plastic strain range of 0.04%. The number of cycles to failure decreased to one seventh that of the unirradiated specimen at the strain range.…”

Section: Fatigue Propertiesmentioning

confidence: 97%

Current Status of Reduced-Activation Ferritic/Martensitic Steels R&D for Fusion Energy

Kimura

2005

Mater. Trans.

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Reduced-activation ferritic/martensitic (RAF/M) steels have been considered to be the prime candidate for the fusion blanket structural material. The irradiation data obtained up to now indicates rather high feasibility of the steels for application to fusion reactors because of their high resistance to degradation of material performance caused by both the irradiation-induced displacement damage and transmutation helium atoms. The martensitic structure of RAF/M steels consists of a large number of lattice defects before the irradiation, which strongly retards the formation of displacement damage through absorption and annihilation of the point defects generated by irradiation. Transmutation helium can be also trapped at those defects in the martensitic structure so that the growth of helium bubbles at grain boundaries is suppressed. The major properties of the steels are well within our knowledge, and processing technologies are mostly developed for fusion application. RAF/M steels are now certainly ready to proceed to the next stage, that is, the construction of International Thermo-nuclear Experimental Reactor Test Blanket Modules (ITER-TBM).Oxide dispersion strengthening (ODS) steels have been developed for higher thermal efficiency of fusion power plants. Recent irradiation experiments indicated that the steels were quite highly resistant to neutron irradiation embrittlement, showing hardening accompanied by no loss of ductility. High-Cr ODS steels whose chromium concentration was in the range from 14 to 19 mass% showed high resistance to corrosion in supercritical pressurized water. It is shown that the 14Cr-ODS steel is susceptible to neither hydrogen nor helium embrittlement.A combined utilization of ODS steels with RAF/M steels will be effective to realize fusion power early at a reasonable thermal efficiency.

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Section: Fatigue Propertiesmentioning

confidence: 97%

Current Status of Reduced-Activation Ferritic/Martensitic Steels R&D for Fusion Energy

Kimura

2005

Mater. Trans.

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“…6(d)) increases with an increase in tantalum. Cyclic softening might be a serious concern especially at low strain range under irradiation and under thermomechanical loading conditions [26] since deformation localization can lead to local instabilities. A comparative study is made on the softening behavior of the RAFM steels with varying W and Ta contents.…”

Section: Resultsmentioning

confidence: 99%

Effect of tungsten and tantalum on the low cycle fatigue behavior of reduced activation ferritic/martensitic steels

Shankar

Mariappan

Nagesha

et al. 2012

Fusion Engineering and Design

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“…Cavities size in SiC matrix resulting from triple ion beam irradiation (Si, He, H) tended to be greater than that in the specimen irradiated by dual ion beam (Si, He) at 800°C. Moreover cavities were observed on SiC fibres after triple-beam irradiation and not after dual-beam irradiation at 1000°C [52,53]. In the case of 1300°C irradiation, cavities on grain boundaries became larger and fine cavities appeared in grains but synergetic effect of H was not obvious.…”

Section: Transmutation Gasesmentioning

confidence: 91%

“…Carbon interphase showed no evidence of microstructural changes and no He bubbles, due to the larger He diffusion coefficient in carbon. More recently the role of H on the synergistic effects of displacement damage, helium and hydrogen implantation on microstructure and mechanical properties of ultra-high-purity stoichiometric beta-SiC and Hi-Nicalon S-CVI composite has been investigated qualitatively [52][53][54][55]. Microstructural observations of composites irradiated to 10 dpa at 800, 1000 and 1300°C with a He/dpa and H/dpa respectively equal to 130 and 40 appm/dpa showed that special distribution of cavities in SiC composites depended on grain structure of SiC and irradiation conditions in this temperature region [52].…”

Section: Transmutation Gasesmentioning

confidence: 99%