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Inoue, Masaki; Suto, Mitsuo; Koyama, Shinichi; Otsuka, Satoshi; Kaito, Takeji
JAEA-Research 2013-009, 78 Pages, 2013/10
JAEA-Research-2013-009.pdf:3.75MB
In order to exammine the applicability for advanced aqueous reprocessing system, the martensitic oxide dispersion strengthened ferritic steel (9Cr-ODS steel), which is the primary candidate material for high burnup fuel pin cladding tube in fast reactor cycle, was evaluated for the amount of corrosion products in the dissolution process. The quantity of corrosion products was calculated to investigate the influence of both various chemical processes and waste glass (vitrified high level radioactive wastes) by use of the results of a maximum cladding temperature fuel subassembly and the sum of all fuel subassemblies, respectively. The experimental results of immersion tests in flowing liquid sodium loops and fuel pin irradiation tests in fast reactors were reviewed to consider the effect of outer and inner corrosions in high burnup fuel pins on corrosion products. This work revealed that the sum of corrosion products depends largely on the mass transfer behavior in flowing liquid sodium.
Otsuka, Satoshi; Kaito, Takeji; Tanno, Takashi; Yano, Yasuhide; Koyama, Shinichi; Tanaka, Kenya
Journal of Nuclear Materials, 442(1-3), p.S89 - S94, 2013/09
Times Cited Count：14 Percentile：72.14(Materials Science, Multidisciplinary)The manufacturing tests of 11-12Cr ODS tempered martensitic steels were carried out, and their ferritic/martensitic duplex structures were quantitatively evaluated by three types of methods, i.e. high temperature XRD, EPMA and metallography. It was demonstrated that excessive formation of residual-alpha ferrite provided by increasing Cr can be suppressed by appropriately controlling the concentration of ferrite-forming element and austenite-forming element on the basis of the parameter "chemical driving force of 
to 
reverse transformation" as a useful indication. The 11Cr-ODS steel containing a small portion of residual-alpha ferrite was successfully manufactured. In the as-received condition, this 11Cr-ODS steel is shown to have the satisfactory creep strength and ductility as high as the 9Cr-ODS steel while 0.2% proof strength at 973K is lower than in the 9Cr-ODS steel.
Tanno, Takashi; Otsuka, Satoshi; Yano, Yasuhide; Kaito, Takeji; Oba, Yojiro*; Onuma, Masato*; Koyama, Shinichi; Tanaka, Kenya
Journal of Nuclear Materials, 440(1-3), p.568 - 574, 2013/09
Times Cited Count：17 Percentile：78.22(Materials Science, Multidisciplinary)This study carried out mechanical tests and microstructure characterizations of several 9Cr and 11Cr-ODS tempered martensitic steels, and discussed the appropriate chemical composition range of 11Cr-ODS tempered martensitic steel from the viewpoint of high-temperature strength improvement. It was shown that the residual -ferrite fraction in 11Cr-ODS steel was successfully controlled to the same level as the 9Cr-ODS steel by selecting the matrix chemical compositions on the basis of the multi-component phase diagram. The tensile strength decreased with decreasing W content from 2.0 to 1.4 wt%. On the other hand, creep strength at 973 K did not degrade by the decreasing W content. Both tensile strength and creep strength increased with increasing population of the nano-sized oxide particles. Small angle X-ray scattering analysis revealed that titanium and excess oxygen contents were key parameters in order to improve the dispersion condition of nano-sized oxide particles.
Otsuka, Satoshi; Kaito, Takeji; Yano, Yasuhide; Yamashita, Shinichiro; Ogawa, Ryuichiro; Uwaba, Tomoyuki; Koyama, Shinichi; Tanaka, Kenya
Journal of Nuclear Science and Technology, 50(5), p.470 - 480, 2013/05
Times Cited Count：5 Percentile：38.62(Nuclear Science & Technology)Four experimental fuel assemblies (EFAs) containing 9Cr-ODS steel cladding fuel pins were previously irradiated in the BOR-60. One of the EFAs achieved the best data, a peak burn-up of 11.9at% and a neutron dose of 51 dpa, without any microstructure instability or any fuel pin rupture. On the other hand, in another EFA (peak burn-up, 10.5at%; peak neutron dose, 44 dpa), peculiar irradiation behaviors such as microstructure instability and fuel pin rupture occurred. The combined effects of matrix Cr heterogeneity (presence of metallic Cr inclusions) and high-temperature irradiation were concluded to be the main cause of the peculiar microstructure change of 9Cr-ODS steel cladding tubes in the BOR-60 irradiation tests. They contributed to the fuel pin rupture.
Yano, Yasuhide; Kaito, Takeji; Otsuka, Satoshi; Tanno, Takashi; Uwaba, Tomoyuki; Koyama, Shinichi
JAEA-Data/Code 2012-022, 51 Pages, 2012/09
JAEA-Data-Code-2012-022.pdf:2.32MB
It is necessary to develop core materials for fast reactors in order to achieve high-burnup. Ferritic steels are expected to be good candidate core materials to achieve this objective because of their excellent void swelling resistance. Therefore, oxide dispersion strengthened (ODS) ferritic steel and 11Cr-ferritic/martensitic steel (PNC-FMS) have been respectively developed for cladding and wrapper tube materials in Japan Atomic Energy Agency. In this study, various physical properties of PNC-FMS wrapper materials were measured and equations and future standard measurement technique of physical properties for the design and evaluation were conducted.
Otsuka, Satoshi; Kaito, Takeji; Yano, Yasuhide; Yamashita, Shinichiro; Ogawa, Ryuichiro; Uwaba, Tomoyuki; Koyama, Shinichi; Tanaka, Kenya
Proceedings of International Conference on Toward and Over the Fukushima Daiichi Accident (GLOBAL 2011) (CD-ROM), 6 Pages, 2011/12
This paper describes evaluation results of in-reactor integrity of 9Cr and 12Cr-ODS steel cladding tubes and the plan for reliability improvement in homogeneous tube production. A fuel assembly in the BOR-60 irradiation test including 9Cr and 12Cr-ODS fuel pins has achieved the highest burn-up, i.e. peak burn-up of 11.9at% and peak neutron dose of 51dpa, without any fuel pin rupture and microstructure instability. In another fuel assembly containing 9Cr and 12Cr-ODS steel fuel pins whose peak burn-up was 10.5at%, one 9Cr-ODS steel fuel pin failed near the upper end of the fuel column. A peculiar microstructure change occurred in the vicinity of the ruptured area. The primary cause of this fuel pin rupture and microstructure change was shown to be the presence of metallic Cr inclusions in the 9Cr-ODS steel tube, which had passed an ultrasonic inspection test for defects. In the next stage from 2011 to 2013, the fabrication technology of full pre-alloy 9Cr-ODS steel cladding tube will be developed.
Naoe, Takashi; Futakawa, Masatoshi; Koyama, Tomofumi*; Kogawa, Hiroyuki
Jikken Rikigaku, 6(3), p.301 - 307, 2006/09
A mercury target for spallation neutron source is subject to pressure waves caused by proton bombarding mercury. The pressure wave propagation induces the cavitation in mercury that imposes pitting damage on the target vessel. In this paper, single micro-bubble behavior in mercury was evaluated using numerical calculation on the basis of bubble dynamics given by Rayleigh-Plesset. Impact pressure loading tests using an electro-Magnetic IMpact Testing Machine (MIMTM) were performed to measure the impact pressure and acoustic vibration. Additionally, in order to visualize micro-bubble behavior in mercury, high-speed video camera observation was carried out. As the result, we confirmed that the maximum bubble radius and lifetime of micro-bubble are dependent on the imposed pressure and the pressure saturate time and that the acoustic vibration with high frequency components above 15 kHz is exited by the micro-bubble collapse.
Teshigawara, Makoto; Harada, Masahide; Saito, Shigeru; Oikawa, Kenichi; Maekawa, Fujio; Futakawa, Masatoshi; Kikuchi, Kenji; Kato, Takashi; Ikeda, Yujiro; Naoe, Takashi*; et al.
Journal of Nuclear Materials, 356(1-3), p.300 - 307, 2006/09
Times Cited Count：9 Percentile：53.38(Materials Science, Multidisciplinary)We adopted silver-indium-cadmium (Ag-In-Cd) alloy as a material of decoupler for decoupled moderator in JSNS. However, from the heat removal and corrosion protection points of view, the Ag-In-Cd alloy is needed to clad between Al alloys (Al5083). We attempted to obtain good bonding conditions for between Al5083 and ternary Ag-In-Cd alloys by HIPing tests. The good HIP condition was found for small test piece (
20mm). Though a hardened layer due to the formation of AlAg
was found in the bonding layer, the rupture strength of the bonding layer was more than 20 MPa, which was the calculated design stress. Bonding tests of a large size piece (200
20030 mm
), which simulated the real scale, were also performed according to the results of small size tests. The result also gave good bonding and enough required-mechanical-strength, however the rupture strength of the large size test was smaller than that of small one.
Naoe, Takashi*; Futakawa, Masatoshi; Koyama, Tomofumi*; Kogawa, Hiroyuki; Ikeda, Yujiro
Jikken Rikigaku, 5(3), p.280 - 285, 2005/09
no abstracts in English
Koyama, Tomofumi*; Futakawa, Masatoshi; Kogawa, Hiroyuki; Ishikura, Shuichi*
Nihon Kikai Gakkai Kanto Shibu Ibaraki Koenkai (2002) Koen Rombunshu, p.5 - 6, 2002/09
no abstracts in English
Yamashita, Shinichiro; Yano, Yasuhide; Otsuka, Satoshi; Kaito, Takeji; Akasaka, Naoaki; Inoue, Masaki; Yoshitake, Tsunemitsu; Nishinoiri, Kenji; Koyama, Shinichi; Tanaka, Kenya
no journal, ,
In this work, neutron irradiation behaviour of ODS ferritic steel cladding tubes developed for fast reactor (FR) was investigated to understand the effect of neutron irradiation on their microstructures. Chemical compositions of the ODS cladding tubes examined were Fe-0.13C-8.84Cr-1.97W-0.20Ti-0.34YO
(9Cr-ODS) and Fe-0.04C-11.34Cr-1.89W-0.25Ti-0.23YO (12Cr-ODS). These ODS cladding tubes were irradiated, without fuel condition, at 731-1089 K to fast fluences ranging from 3.2 to 6.610
n/m
(E 
0.1 MeV) in the experimental fast reactor JOYO. Microstructural stability of these cladding tubes was evaluated using transmission electron microscope (TEM). Density of the tube specimens before and after irradiation was measured by a conventional immersion method with water, indicating that no significant swelling occurred for all the irradiated specimens. TEM observations show that the radiation-induced defect cluster formation during neutron irradiation was suppressed. It was highly possible due to high density defect sink site such as initially-existed dislocation introduced during tube fabrication process, interface between precipitates including oxide and each matrix. In addition, it revealed that oxide particles, which are closely related with high temperature strength under the practical reactor operation, were stable up to the maximum doses of this irradiation test from the analyses of TEM micrographs.
Yano, Yasuhide; Otsuka, Satoshi; Kaito, Takeji; Uwaba, Tomoyuki; Koyama, Shinichi; Tanaka, Kenya
no journal, ,
no abstracts in English
Yano, Yasuhide; Yamashita, Shinichiro; Otsuka, Satoshi; Kaito, Takeji; Koyama, Shinichi; Tanaka, Kenya
no journal, ,
no abstracts in English
Tanno, Takashi; Otsuka, Satoshi; Kaito, Takeji; Yano, Yasuhide; Koyama, Shinichi; Tanaka, Kenya; Oba, Yojiro*; Onuma, Masato*
no journal, ,
The 9Cr-ODS martensitic steel is under development as first candidate material for fast reactor long life fuel cladding. It should have high temperature strength, irradiation resistance and productibility. The 11Cr-ODS steels with additional corrosion resistance by Cr enrichment are also under development. And the evaluation of their applicability for cladding material is in progress. Trial manufacture, creep test and nano/macro structure evaluation of several ODS steels were carried out to modify the chemical composition and manufacturing process of the 11Cr-ODS steel. Uni-axial creep strength of manufactured 11Cr-ODS steels were equal to or higher than that of 9Cr-ODS steel. The strength of 11Cr-ODS steels with 0.3wt% Ti was slightly higher than that with 0.2wt% Ti. The size and number density of dispersed-oxide depended on the amount of added Ti and extra-oxygen.
Uwaba, Tomoyuki; Yano, Yasuhide; Kaito, Takeji; Nemoto, Junichi*; Otsuka, Satoshi; Tanno, Takashi; Koyama, Shinichi
no journal, ,
Bulging type spacer pad forming on a wrapper tube was investigated as a development of wrapper tubes for fast reactor fuels. The pad forming tests were performed using the real wrapper tubes. Elastic-plastic analyses simulating the pad forming were also performed with FEM models to evaluate the residual stress and strain caused by the forming.
Yano, Yasuhide; Kaito, Takeji; Uwaba, Tomoyuki; Otsuka, Satoshi; Tanno, Takashi; Koyama, Shinichi
no journal, ,
no abstracts in English
Kaito, Takeji; Yano, Yasuhide; Uwaba, Tomoyuki; Otsuka, Satoshi; Tanno, Takashi; Koyama, Shinichi
no journal, ,
no abstracts in English
