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Kondo, Keietsu; Miwa, Yukio*; Okubo, Nariaki; Kaji, Yoshiyuki; Tsukada, Takashi
Journal of Nuclear Materials, 417(1-3), p.892 - 895, 2011/10
Times Cited Count：4 Percentile：32.59(Materials Science, Multidisciplinary)For the purpose to suppress the degradation of corrosion resistance induced by irradiation in austenitic stainless steels (SSs), aluminum-doped type 316L SS (316L/Al) was fabricated, and its electrochemical corrosion property was estimated after Ni-ion irradiation at the temperature range from 330
C to 550C. And it was revealed that aluminum addition to SSs was effective in the case of irradiation at elevated temperature. 316L/Al irradiated at 550C up to 12 dpa showed high corrosion resistance in the vicinity of grain boundaries (GBs) and in grains, while the severe GB etching and local corrosion in grains were observed in irradiated 316L and 316 SS. It is supposed that the aluminum enrichment, which is caused by radiation induced segregation at GBs and by radiation induced precipitation such as Ni3Al in grains, was enhanced by high-temperature irradiation, and contributes to compensate the lost corrosion resistance by the chromium depletion.
Doda, Norihiro; Ohshima, Hiroyuki; Kamide, Hideki; Watanabe, Osamu*; Okubo, Yoshiyuki*
Proceedings of 14th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-14) (CD-ROM), 13 Pages, 2011/09
Toward the commercialization of fast reactors, a design study of Japan Sodium-cooled Fast Reactor (JSFR) is being performed, in which fully natural circulation system is adopted as the decay heat removal system. A new evaluation method of core hot spot which can be applied to natural circulation decay heat removal has been developed. The new method consists of three-step thermal hydraulics analyses in order to consider the effects of physical phenomena particular to natural circulation, such as inter-fuel-assembly heat transfer and flow redistribution in the core due to buoyancy force. From the viewpoint of calculation cost reduction, we have also developed a simplified model substituting for the third step analysis (subchannel analysis). The new method was applied to the evaluations of a loss-of-external-power event and of a sodium leakage accident in a secondary loop of a large scale reactor.
Okubo, Tsutomu; Oki, Shigeo; Ogura, Masashi*; Okubo, Yoshiyuki*; Kotake, Shoji*
Proceedings of 2011 International Congress on Advances in Nuclear Power Plants (ICAPP '11) (CD-ROM), p.479 - 486, 2011/05
A conceptual design study and related R&D on a commercial-base large-scale Japan Sodium-cooled Fast Reactor (JSFR) have been carried out in the framework of the Fast Reactor Cycle Technology development (FaCT) project. As a next generation plant, JSFR adopts a number of innovative technologies in order to achieve economic competitiveness, enhanced reliability and safety. This paper describes the current results of the ongoing conceptual design study on the JSFR core. The most important point in the core design is to achieve a high core average burn-up around 150 GWd/t, assuming the ODS steel utilization as the cladding material. Another design target for the breeding ratio is intended to have some flexibility and is set at from around 1.0 to 1.2 under the design philosophy of the compatible fuel assembly among them. Also, the fuel composition is considered to have some variation range based on the wide variety of the spent fuel composition expected to be treated during the LWR to FBR transition period. The core design study performed in the FaCT project has clarified the feasibility of the JSFR core concept, which is based on the high internal conversion ratio type core using a large fuel rod diameter around 10 mm and satisfies a number of design targets and requirements including ones mentioned above.
Doda, Norihiro; Ohshima, Hiroyuki; Kamide, Hideki; Watanabe, Osamu*; Okubo, Yoshiyuki*
Proceedings of 8th International Topical Meeting on Nuclear Thermal-Hydraulics, Operation and Safety (NUTHOS-8) (CD-ROM), 11 Pages, 2010/10
Toward the commercialization of fast reactors, design study of JSFR is being performed. Adoption of fully natural circulation system is being examined as the decay heat removal system. In order to confirm feasibility of the system, we are developing a new evaluation method of core hot spot in transition from rated operation to natural circulation decay heat removal conditions, which requires uncertainty factor assessment for the natural circulation conditions as well as for the rated operation conditions. In this paper, we focus on effects of wire-spacer contact and pellet- cladding eccentricity on the peak cladding temperature as typical uncertainty factors and evaluated these two effects under natural circulation conditions quantitatively.
Ohshima, Hiroyuki; Doda, Norihiro; Kamide, Hideki; Watanabe, Osamu*; Okubo, Yoshiyuki*
Nihon Kikai Gakkai Rombunshu, B, 76(763), p.448 - 450, 2010/03
Toward the commercialization of fast reactors, a design study of Japan Sodium Cooled Fast Reactor is being performed. In this design study, the adoption of decay heat removal system operated by fully natural circulation is being examined from viewpoints of economic competitiveness and passive safety. This paper describes a new evaluation method of core hot spot that is necessary for confirming feasibility of the fully natural circulation decay heat removal system. The new method consists of three analysis steps in order to include effects of thermal hydraulic phenomena particular to the natural circulation decay heat removal and therefore it enables more rational hot spot evaluation rather than conventional ones. This method was applied to a hot spot evaluation of loss-of-external-power event and the result was compared with those by conventional simulations. It was confirmed that the proposed method can estimate the hot spot with reasonable degree of conservativeness.
Miwa, Yukio; Kaji, Yoshiyuki; Okubo, Nariaki; Kondo, Keietsu; Tsukada, Takashi
Nihon Kikai Gakkai M&M 2007 Zairyo Rikigaku Kanfarensu Koen Rombunshu (CD-ROM), p.236 - 237, 2009/07
In core structural materials of next generation reactors, materials' degradation behavior by neutron irradiation damage and thermal (cyclic) stress should be considered with fair accuracy in design process, because the materials are used under higher temperature gradients and higher neutron flux fields than those in the present light water reactors. In the current experiential design rules, service lives of core structural components were determined by the materials degradation such as the increase of ductile-to-brittle transition temperature after post irradiation examination data. However, other materials degradations such as irradiation-assisted stress corrosion cracking (IASCC), which occurs by the degradation synergistically interacting with radiation hardening, local chemical composition change, swelling and radiation creep, should be considered reasonably in the design process of the next generation reactors, because of the anticipation of the beneficial effects by synergy of radiation damage. To predict material failure by IASCC with reasonable accuracy, in this study, each material degradation phenomenon with different dose dependence was modeled with consideration of radiation induced stress relaxation. In this paper, the models obtained by ion-irradiation experiments and compared by data from neutron irradiation experiments were presented, and the concept of our new evaluation method and the programming code for the failure simulation were outlined.
Kaji, Yoshiyuki; Miwa, Yukio; Kondo, Keietsu; Okubo, Nariaki; Tsukada, Takashi
Proceedings of 2009 International Congress on Advances in Nuclear Power Plants (ICAPP '09) (CD-ROM), P. 9359, 2009/05
In this paper, we describe the simulation results of the irradiation assisted stress corrosion cracking (IASCC) behavior at the flaws considering the radiation induced stress relaxation (RISR) with residual stress introduced by the welding process for a long operation period.
Okubo, Nariaki; Miwa, Yukio; Kondo, Keietsu; Kaji, Yoshiyuki
Journal of Nuclear Materials, 386-388, p.290 - 293, 2009/04
Times Cited Count：4 Percentile：30.49(Materials Science, Multidisciplinary)no abstracts in English
Miwa, Yukio; Kaji, Yoshiyuki; Okubo, Nariaki; Kondo, Keietsu; Tsukada, Takashi
Journal of Solid Mechanics and Materials Engineering (Internet), 2(1), p.145 - 155, 2008/00
In core structural materials of next generation reactors, materials' degradation behavior by neutron irradiation damage and thermal (cyclic) stress should be considered with fair accuracy in design process (including maintenance and repair plans), because the materials are used under higher temperature gradients and higher neutron flux fields than those in the present light water reactors. In the current experiential design rules, service lives of core structural components were determined by the materials degradation such as the increase of ductile-to-brittle transition temperature after post irradiation examination data. However, other materials degradations such as irradiation-assisted stress corrosion cracking (IASCC) should be considered reasonably in the design process of the next generation reactors, because of the anticipation of the beneficial effects by synergistics of these radiation damage such as radiation hardening, local chemical composition change, swelling and radiation creep. To predict material failure by IASCC with reasonable accuracy, in this study, each material degradation phenomenon with different dose dependence was modeled with consideration of radiation induced stress relaxation. The models were integrated to simulate the failure behavior for the reactor operation period. In this paper, the models obtained by ion-irradiation experiments were presented, and the concept of new evaluation method and the programming code for the failure simulation were outlined.
Kondo, Keietsu; Miwa, Yukio; Okubo, Nariaki; Kaji, Yoshiyuki; Tsukada, Takashi
Proceedings of 13th International Conference on Environmental Degradation of Materials in Nuclear Power Systems (CD-ROM), 11 Pages, 2007/00
The effect of residual stress on corrosion behavior in type 316L austenitic stainless steel was examined by ion irradiated specimens. Ion irradiation was performed on specimens both undeformed and deformed plastically by bending constrait at 330C to average displacement damage from 1 to 45dpa. It was observed in EPR testing that deformed specimens showed higher corrosion resistance than undeformed specimens. Three-dimensional atom probe analysis was conducted on irradiated specimens. It was found that the enrichment of Ni, Si and the depletion of Cr at dislocations, and the degree of segregation was greater in undeformed specimen than in deformed specimen. It could be suggested that radiation induced segregation behavior of solute atoms as a consequence of diffusion and annihilation of irradiation defects at sink is affected by residual stress, and this also might affect the corrosion resistance.
Kondo, Keietsu; Nemoto, Yoshiyuki; Miwa, Yukio; Kaji, Yoshiyuki; Tsukada, Takashi; Nagai, Yasuyoshi*; Hasegawa, Masayuki*; Okubo, Tadakatsu*; Hono, Kazuhiro*
JAEA-Research 2006-013, 39 Pages, 2006/12
JAEA-Research-2006-013.pdf:4.57MB
There has been an increasing number of stress corrosion cracking (SCC) incidents on low carbon austenitic stainless steels used in boiling water reactor (BWR) environments. To reveal the acceleration factor of intergranular crack propagation from the viewpoint of solute distribution in stainless steels, the material extracted from a core shroud of Japanese BWR was analyzed by the three dimensional atom probe (3DAP), which has the highest spatial resolution among the various microanalytical techniques. It was revealed by statistical analysis on 3DAP data that solute elements, such as Fe, Cr, Ni, Mo, Mn, Si, are randomly distributed in matrix of the shroud material. This result means that solute was not segregated or precipitated and was not form spinodal decomposition during the service. The concentration profile in the vicinity of grain boundary obtained from 3DAP dataset showed the random distribution of Cr. This result shows that degradation of the corrosion resistance induced by depletion of Cr was not responsible for the crack propagation along grain boundaries in low carbon stainless steel. On the other hand, enrichment of Mo and Si was observed at grain boundary. The width of the enriched zone was about 2 nm across the grain boundary, and the concentration of those elements could be much higher than the concentration obtained by field emission transmission electron microscopy/energy dispersive X-ray spectroscopy (FE-TEM/EDS). Therefore, it is necessary to study about the effects of enrichment of Mo and Si as a potential contributor to SCC.
Chikazawa, Yoshitaka; Okano, Yasushi; Hori, Toru*; Okubo, Yoshiyuki*; Shimakawa, Yoshio*; Tanaka, Toshihiko*
Journal of Nuclear Science and Technology, 43(8), p.829 - 843, 2006/08
Times Cited Count：4 Percentile：30.68(Nuclear Science & Technology)In phase II of the feasibility study of commercialized fast reactor cycle systems, we make a concept of a small sodium cooled reactor for a power source of a city with various requirements, such as, safety and economical competitiveness. Various reactor concepts are surveyed and a tank type reactor whose intermediate heat exchanger and primary main pumps are arranged in series is selected. In this study, a compact long life core and a simple reactor structure designs are pursued. The core type is three regional Zr concentration with one Pu enrichment core, the reactor outlet temperature achieves 550C and the reactor electric output increases from 150MWe to 165MWe. The construction cost is much higher than the economical goal in the case of FOAK. But the construction cost in the case of NOAK is estimated to be 85.6% achieving the economical goal.
Aizawa, Kosuke; Chikazawa, Yoshitaka; Shiraishi, Tadashi*; Sakata, Hideyuki*; Okubo, Yoshiyuki*
no journal, ,
no abstracts in English
Okubo, Nariaki; Miwa, Yukio; Kondo, Keietsu; Kaji, Yoshiyuki
no journal, ,
no abstracts in English
Kondo, Keietsu; Okubo, Nariaki; Miwa, Yukio; Kaji, Yoshiyuki; Tsukada, Takashi
no journal, ,
no abstracts in English
Wakai, Eiichi; Takada, Fumiki; Takaya, Shigeru; Kato, Shoichi; Kitazawa, Sin-iti; Okubo, Nariaki; Suzudo, Tomoaki; Fujii, Kimio; Yoshitake, Tsunemitsu; Kaji, Yoshiyuki; et al.
no journal, ,
no abstracts in English
Kondo, Keietsu; Miwa, Yukio; Okubo, Nariaki; Kaji, Yoshiyuki; Tsukada, Takashi
no journal, ,
no abstracts in English
Miwa, Yukio; Kondo, Keietsu; Okubo, Nariaki; Kaji, Yoshiyuki
no journal, ,
This study proposed an evaluation method of damage occurrence of core structural materials by simulation model. Irradiation assisted stress corrosion cracking (IASCC) is one of the damage modes among aged core components. IASCC was thought to be caused by material degradation such as losses of workability due to radiation hardening and corrosion resistance due to radiation-induced segregation and residual stress due to welding, then the influence of residual stress on both radiation hardening and radiation induced segregation was examined by ion-irradiation experiments. Not only the residual stress change due to radiation creep was affected by residual stress and dose, but also the material degradation was affected. This was modeled and the damage occurrence by IASCC in a shroud type component during reactor operating life was estimated by means of FEM calculation based on the models of the material degradation and residual stress.
Ohshima, Hiroyuki; Kamide, Hideki; Tanaka, Masaaki; Watanabe, Osamu*; Okubo, Yoshiyuki*
no journal, ,
In the design study of commercialized sodium-cooled fast reactor, the adoption of decay heat removal by entire natural circulation is being examined from the viewpoint of enhancing economical competitiveness and safety. In this study, an evaluation method is proposed, in which the hot spot in the core can be rationally evaluated under transient conditions from rated operation to natural circulation decay heat removal.
Kondo, Keietsu; Miwa, Yukio; Okubo, Nariaki; Kaji, Yoshiyuki; Tsukada, Takashi
no journal, ,
no abstracts in English
