Torsion test technique for interfacial shear evaluation of F82H RAFM HIP-joints

Nozawa, Takashi; Ogiwara, Hiroyuki*; Kannari, Jun*; Kishimoto, Hirotatsu*; Tanigawa, Hiroyasu

Fusion Engineering and Design, 86(9-11), p.2512 - 2516, 2011/10

https://doi.org/10.1016/j.fusengdes.2011.01.110

A hot isostatic press (HIP) process is a key technology to fabricate a first wall of the blanket system utilizing a reduced-activation ferritic/martensitic (RAFM) steel such as F82H. A primary objective of this study is to characterize interfacial properties of HIPed F82H joints by torsion to identify the feasibility of this test method. It is apparent that the absorption energies of the HIP joints varied by the processing conditions, although the maximum shear strength was not much different. According to the fracture surfaces, it is indicated that the reduction of the absorption energy was due to the oxide formed on the interface of the HIP joint and this was consistent with the results of charpy impact tests. In conclusion, the torsion test method enables to precisely evaluate the shear properties of the HIPed joint interface and becomes one of promising powerful techniques for inspection of the HIP joints.

Super ODS steels R&D for fuel cladding of next generation nuclear systems, 1; Introduction and alloy design

Kimura, Akihiko*; Kasada, Ryuta*; Iwata, Noriyuki*; Kishimoto, Hirotatsu*; Zhang, C. H.*; Isselin, J.*; Dou, P.*; Lee, J. H.*; Muthukumar, N.*; Okuda, Takanari*; et al.

Proceedings of 2009 International Congress on Advances in Nuclear Power Plants (ICAPP '09) (CD-ROM), p.9220_1 - 9220_8, 2009/05

Cladding material development is essential for realization of highly efficient high burn-up operation of next generation nuclear systems, where high performance is required for the materials, that is, high strength at elevated temperature, high resistance to corrosion and high resistance to irradiation. Oxide dispersion strengthening (ODS) ferritic steels are considered to be most adequate for the cladding material because of their high strength at elevated temperature. In this work, "Super ODS steel" that has better corrosion resistance than 9Cr-ODS steel, has been developed for application to cladding of a variety of next generation nuclear systems. In the following ten papers, the recent experimental results of "Super ODS steel" R&D will be presented, indicating that many unexpected preferable features were found in the mechanical properties of nano-sized oxide dispersion high-Cr ODS ferritic steel. A series of paper begins with alloy design of "Super ODS steel". Corrosion issue requires Cr concentration more than 14wt.%, but aging embrittlement issue requires less than 16wt.%. An addition of 4wt.%Al is effective to improve corrosion resistance of 16wt.%Cr-ODS steel in supercritical water (SCW) and lead-bismuth eutectic (LBE), while it is detrimental to high-temperature strength. Additions of 2wt.%W and 0.1wt.%Ti are necessary to keep high strength at elevated temperatures. An addition of small amount of Zr or Hf results in a significant increase in creep strength at 700 C in Al added ODS steels. Tube manufacturing was successfully done for the super ODS steel candidates. "Super ODS steel" is promising for the fuel cladding material of next generation nuclear systems, and the R&D is now ready to proceed to the next stage of empirical verification.

Super ODS steels R&D for fuel cladding of next generation nuclear systems, 8; Ion irradiation effects at elevated temperatures

Kishimoto, Hirotatsu*; Kasada, Ryuta*; Kimura, Akihiko*; Inoue, Masaki; Okuda, Takanari*; Abe, Fujio*; Onuki, Somei*; Fujisawa, Toshiharu*

Proceedings of 2009 International Congress on Advances in Nuclear Power Plants (ICAPP '09) (CD-ROM), p.9219_1 - 9219_8, 2009/05

The Super ODS steels, having excellent high-temperature strength and highly corrosion resistant, are considered to increase the energy efficiency by higher temperature operation and extend the lifetime of next generation nuclear systems. High-temperature strength of the ODS steels strongly depends on the dispersion of oxide particles, therefore, the irradiation effect on the dispersed oxides is critical in the material development. In the present research, ion irradiation experiments were employed to investigate microstructural stability under the irradiation environment at elevated temperatures. Ion irradiation experiments were performed with 6.4 MeV Fe ions irradiated at 650 C up to a nominal displacement damage of 60 dpa. Microstructural investigation was carried out using TEM and EDX. No significant change of grains and grain boundaries was observed by TEM investigation after the ion irradiation. Main oxide particles in the 16Cr-4Al-0.1Ti (SOC-1) ODS steel were (Y, Al) complex oxides. (Y, Ti) complex oxides were in 16Cr-0.1Ti (SOC-5) and 15.5Cr-2W-0.1Ti (SOCP-3). (Y, Zr) complex oxides were in 15.5Cr-4Al-0.6Zr (SOCP-1). No significant modification of these complex oxides was detected after the ion irradiation up to 60 dpa at 650 C. The stable complex oxides are considered to keep highly microstructural stability of the Super ODS steels under the irradiation environments.

Radiation induced phase instability of precipitates in reduced-activation ferritic/martensitic steels

Tanigawa, Hiroyasu; Sakasegawa, Hideo; Ogiwara, Hiroyuki*; Kishimoto, Hirotatsu*; Koyama, Akira*

Journal of Nuclear Materials, 367-370(1), p.132 - 136, 2007/08

https://doi.org/10.1016/j.jnucmat.2007.03.155

It was previously reported that reduced-activation ferritic/martensitic steels (RAFs), such as F82H-IEA and JLF-1, showed a variety of changes in its mechanical property after neutron irradiation at 573K up to 5dpa, and have possible correlation with precipitation. The effects of irradiation on precipitation were also reported previously. In this study, irradiation effects on precipitation were investigated in detail utilizing ion irradiation in which irradiation condition could be controlled with high accuracy. F82H IEA heat, JLF-1 HFIR heat, and aged F82H-IEA (873K30k h) were used for experiments. The specimens were irradiated at DuET facility, Inst. of Advanced Energy, Kyoto University up to 10 dpa at 573K with 6.4MeV Fe ion. Cross sectional TEM thin film specimens of ion irradiated region were made utilizing focused ion beam (FIB) processor with micro-sampling system at JAERI. These thin film specimens were made to contain both irradiated region and non-irradiated region beneath irradiated region. Size distribution and aspect ratio of precipitates were analyzed on each region. It turned out that the finer precipitates were dominant in irradiated region of F82H compared to that in non-irradiated region, but fewer and larger precipitates were dominant in irradiated region of JLF-1. These results confirmed the presence of irradiation effects on precipitate evolution even at 573K, which was observed in neutron irradiated RAFs.

Evaluation on fracture resistance of advanced SiC/SiC composites using single- and double-notched specimens

Nozawa, Takashi; Kato, Yutai*; Kishimoto, Hirotatsu*; Koyama, Akira*

no journal, , 

One of the advantages for highly-crystalline and stoichiometric silicon carbide (SiC) composites (advanced SiC/SiC) is improved stability of chemical, physical and mechanical properties at high-temperatures. Besides, it has been revealed that significant contribution from the optimized fiber/matrix (F/M) interface enables to give good quasi-ductility beyond matrix cracking. Though the importance of the F/M interfacial role is recognized, understanding the mechanism of crack propagation is insufficient. Meanwhile, the existing test methods to determine fracture resistance of composites, i.e., a critical energy required to propagate a crack, are not fully established because of very limited considerations on the influence of irreversible energies emerged by interaction at the F/M interface and microcrack forming. This study aims to develop a fracture resistance test methodology and to quantify the crack resistance of advanced SiC/SiC composites.

Microstructural development of multi-pass TIG welded F82H steels under dual-ion irradiation

Ogiwara, Hiroyuki; Tanigawa, Hiroyasu; Mizui, Tomohiro*; Kishimoto, Hirotatsu*; Koyama, Akira*

no journal, , 

Reduced-activation ferritic/martensitic steels for first wall and blanket structural component applications in a fusion reactor required joining by welding, and effects of displacement damage and helium production on mechanical properties and microstructures are important to these applications. In the fabrication of blanket modules, the joints of a first wall/side walls will be applied to a multi-pass tungsten inert gas (TIG) welding. The objectives of this work are to clarify the helium effects on swelling behavior and the microstructural evolution in the region welded by a multi-pass TIG welding. F82H steels were irradiated at 470 C up to high dose 20 dpa by using 6.4 MeV Fe and/or energy-degraded 1.0 MeV He. The damage rate is 3.010 dpa/s, and the helium injection rate is 1510 appm He/s. Microstructure and Vickers hardness profiles across base metal, heat affected zone (HAZ) and fusion zone (FZ) were examined before irradiation experiments. The amount of hardness in FZ increased in increments of number in welding passes. The swelling resistance varied with the type considered due to the phase transformation that occur during the heating and cooling cycles of the fusion welding process. In dual-ion irradiated FZ, cavities were observed to a region from one pass to fourth passes and not fifth passes, and amount of swelling decreased in increments of number in welding passes. The tempered zone offered the largest amount of swelling across HAZ.

Super ODS steel research and development towards highly efficient nuclear systems, 5; Irradiation effects

Kishimoto, Hirotatsu*; Kasada, Ryuta*; Kimura, Akihiko*; Onuki, Somei*; Okuda, Takanari*; Abe, Fujio*; Inoue, Masaki; Fujisawa, Toshiharu*

no journal, , 

Irradiation effects on Super ODS steels will be summarized for the project named as "Super ODS Steel Research and Development towards Highly Efficient Nuclear Systems".

Interfacial shear properties of HIP joints of reduced-activation ferritic/martensitic steel

Nozawa, Takashi; Noh, S.; Kishimoto, Hirotatsu*; Tanigawa, Hiroyasu

no journal, , 

A hot isostatic press (HIP) process is a key technology to fabricate a first wall with cooling channels of the fusion blanket system utilizing a reduced-activation ferritic/martensitic (RAFM) steel. A primary objective of this study is to provide a reasonable and comprehensive method to determine interfacial shear properties of HIP joints during the torsional fracture process. By considering hardening during torsional process, we could propose a reasonable and realistic solution to define the interfacial debond shear strength of the HIPed interface by torsion.

Fabrication of tungsten and carbon clad plates by sinter bonding methods

Kishimoto, Hirotatsu*; Matano, Minoru*; Asakura, Yuki*; Fukumoto, Masakatsu; Kubo, Hirotaka

no journal, ,