Sato, Isamu; Arima, Tatsumi*; Nishina, Masahiro*; Tanaka, Kosuke; Onose, Shoji; Idemitsu, Kazuya*
JAEA-Research 2012-006, 66 Pages, 2012/05
As one of the important properties for fuel manufacturability and burning behavior, the diffusion behavior of actinides in oxide fuels was investigated by both the experimental and the molecular dynamics simulation (MD). Using diffusion couples consisted of an Am containing mixed oxide fuel and a UO fuel, the diffusion tests were performed. The diffusion coefficients were estimated to be 10m/s 10m/s. In addition, the difference between Pu and Am diffusion coefficients was vanishingly small. The temperature dependence of bulk diffusion coefficients of actinides in mixed oxide fuels could be evaluated by MD. An evaluation technique for the grain boundary diffusion could be established based on the coincidence site lattice theory. The practical diffusion coefficients were obtained by combining data from the experiments with those predicted from MD. The practical diffusion coefficients obtained was discussed for use of a fuel behavior analysis code.
Miwa, Shuhei; Osaka, Masahiko; Usuki, Toshiyuki; Sato, Isamu; Tanaka, Kosuke; Hirosawa, Takashi; Yoshimochi, Hiroshi; Onose, Shoji
Proceedings of International Conference on Toward and Over the Fukushima Daiichi Accident (GLOBAL 2011) (CD-ROM), 8 Pages, 2011/12
A new fast reactor (FR) cycle concept was previously proposed that incorporates MgO-based inert matrix fuels (IMFs) containing minor actinides harmonious with the existing FR cycle technologies. A basic study of MgO-based IMFs was made regarding their fabrication, characterization and reprocessing in terms of applicability to existing FR cycle technology. It was concluded from these basic investigations of MgO-based IMFs that the existing FR cycle technologies can be applied to those for MgO-based IMFs, and the basic technologies of MgO-based IMFs containing minor actinides harmonious with the existing FR cycle technologies were established.
Tanaka, Kosuke; Kawamata, Kazuo; Yoshimochi, Hiroshi; Sozawa, Shizuo; Onose, Shoji; Niimi, Motoji; Asaka, Takeo
Proceedings of 1st Asian Symposium on Material Testing Reactors (ASMTR 2011), p.71 - 76, 2011/02
Post irradiation examination (PIE) facilities have been operated for about 40 years at the Oarai Research and Development Center of the Japan Atomic Energy Agency to investigate the performance and soundness of irradiated fuels and materials. The JMTR Hot Laboratory (JMTR-HL) was founded in 1971 mainly to examine the objects irradiated in the Japan Material Testing Reactor (JMTR). The Alpha-Gamma Facility (AGF) was constructed as the first laboratory to perform PIE of plutonium-bearing fuels for Japanese fast reactor development programs. This facility started hot operation in 1971 and has performed physical, metallurgical, and chemical examinations of irradiated fuels including uranium plutonium mixed oxide fuels. A renewal plan for the JMTR-HL and AGF is now in progress, associated with re-operation of the JMTR.
Hayashi, Kimio; Nakagawa, Tetsuya; Onose, Shoji; Ishida, Takuya; Nakamichi, Masaru; Takatsu, Hideyuki; Nakamura, Mutsumi*; Noguchi, Tsuneyuki*
Journal of Nuclear Materials, 386-388, p.1083 - 1086, 2009/04
no abstracts in English
Hayashi, Kimio; Nakagawa, Tetsuya; Onose, Shoji; Ishida, Takuya; Nakamichi, Masaru; Katsuyama, Kozo; Iwamatsu, Shigemi; Hasegawa, Teiji; Kodaka, Hideo; Takatsu, Hideyuki; et al.
JAEA-Technology 2009-007, 168 Pages, 2009/03
In-pile functional tests of breeding blankets have been planned by Japan Atomic Energy Agency (JAEA), using a test blanket module (TBM) which will be loaded in the International Thermonuclear Experimental Reactor (ITER). In preparation for the in-pile functional tests, JAEA has been being performed irradiation experiments of lithium titanate (LiTiO), which is the first candidate of solid breeder materials for the blanket of the demonstration reactor (DEMO) under designing in Japan. The present report describes (1) results of a detailed design and trial fabrication tests of a dismantling apparatus for irradiation capsules which were used in irradiation experiments by the Japan Materials Testing Reactor (JMTR) of JAEA, and (2) results of a preliminary investigation of a glove box facility for post-irradiation examinations (PIEs). In the detailed design of the dismantling apparatus, datailed specifications and the installation methods were examined, based on results of a conceptual design and basic design. In the trial fabrication, cutting tests were curried out by making a mockup of a cutting component. Furthermore, a preliminary investigation of a glove box facility was carried out in order to secure a facility for PIE work after the capsule dismantling, which revealed a technical feasibility.
Morohashi, Yuko; Maruyama, Tadashi*; Donomae, Takako; Tachi, Yoshiaki; Onose, Shoji
Journal of Nuclear Science and Technology, 45(9), p.867 - 872, 2008/09
Hayashi, Kimio; Nakagawa, Tetsuya; Onose, Shoji; Ishida, Takuya; Kodaka, Hideo; Katsuyama, Kozo; Kitajima, Toshio; Takahashi, Kozo; Tsuchiya, Kunihiko; Nakamichi, Masaru; et al.
JAEA-Technology 2008-010, 68 Pages, 2008/03
In-pile functional tests of breeding blankets for fusion reactors have been planned by Japan Atomic Energy Agency (JAEA), using a test blanket module (TBM) which will be loaded in ITER. The present report describes a conceptual investigation and a basic design of the dismantling process for irradiation capsules which were used in irradiation experiments by the Japan Materials Testing Reactor (JMTR) of JAEA. In the present design, the irradiation capsule is cut by a band saw; the released tritium is recovered safely by a purge-gas system, and is consolidated into a radioactive waste form. Furthermore, adoption of the inner-box enclosing the dismantling apparatus has brought a prospect to be able to utilize an existing hot cell (beta- cell) equipped with usual wall material permeable to tritium, without extensive refurbishing of the cell. Thus, the present study has indicated the feasibility of the present dismantling process for the irradiated JMTR capsules containing tritium.
Yano, Yasuhide; Yoshitake, Tsunemitsu; Yamashita, Shinichiro; Akasaka, Naoaki; Onose, Shoji; Watanabe, Seiichi*; Takahashi, Heishichiro
Journal of Nuclear Science and Technology, 44(12), p.1535 - 1542, 2007/12
The effects of fast neutron irradiation have been investigated on the mechanical properties of 11Cr-0.5Mo-2W, Nb, V ferritic/martensitic (F/M) stainless steel (PNC-FMS) and 10.5Cr-1.5Mo, Nb, V F/M stainless steel (HT9M) claddings, especially tensile and transient burst properties. These two F/M claddings were irradiated in the experimental fast reactor JOYO using the PFB090 fuel test assembly. Post irradiation tensile and temperature-transient-to-burst tests were carried out for defueled cladding specimens. The results of mechanical tests for PNC-FMS cladding showed that there was no significant degradation in tensile and transient burst strengths even after fast neutron irradiation. However, these strengths for HT9M cladding tended to shift to lower values than those of as-received specimens. This different behavior of tensile and transient burst strengths was attributed to martensite structural stability which was related to the stable solid solution elements.
Donomae, Takako; Tachi, Yoshiaki; Sekine, Manabu*; Morohashi, Yuko; Akasaka, Naoaki; Onose, Shoji
Journal of the Ceramic Society of Japan, 115(1345), p.551 - 555, 2007/09
Use of moderator materials in Fast Breeder Reactor (FBR) is effective for transmutation technology, and BC is one of the candidates. Up to now, the behavior of BC as the Control rod material is well known, but that of BC is hardly investigated. In this paper, the radiation effects of BC pellets, neutron irradiated in the experimental fast reactor JOYO were studied. From the experimental results, it was observed that no macro-cracks were recognized in the irradiated BC pellets. But, bubble nucleation was found in grain and along grain boundaries of BC. And, it was shown that the conductivity of BC was higher than that of BC. During the annealing from room temperature to 1400C, three recovery stages were found on thermal conductivity. It was suggested that, the recovery of BC was related to the dispersion behavior of helium. Judging from these results, as BC was mechanically more stable compared with BC under irradiation, it was shown that BC had high applicability for a moderator.
Yano, Yasuhide; Yoshitake, Tsunemitsu; Yamashita, Shinichiro; Akasaka, Naoaki; Onose, Shoji; Takahashi, Heishichiro*
Journal of Nuclear Materials, 367-370(1), p.127 - 131, 2007/08
The effects of fast neutron irradiation on tensile and transient burst properties of advanced ferritic/martensitic steel claddings were investigated. Specimens were irradiated in the experimental fast reactor JOYO using the material irradiation rig at temperatures between 773 and 1013 K to fast neutron doses ranging from 11 to 102 dpa. The post-irradiation tensile and temperature-transient-to-burst tests were carried out. The results of mechanical tests showed that there was no significant degradation in tensile and transient burst strengths after neutron irradiation below 873 K. This was attributed to grain boundary strengthening caused by precipitates that preferentially formed on prior-austenite grain boundaries. Both strengths at neutron irradiation above about 903 K up to 102 dpa decreased due to recovery of lath martensite structures and recrystallization.
Yamashita, Shinichiro; Akasaka, Naoaki; Onose, Shoji
Zairyo Kaihatsu No Tameno Kenbikyoho To Oyo Shashinshu, P. 133, 2006/03
Oxide Dispersion Strengthened (ODS) ferritic steel dealt with this study was a MA957 (Fe-0.015C-14Cr-0.3Mo-1.0Ti-0.25Y2O3). The objectives of this study were to understand oxide particle stability of ODS steel during irradiation and interaction among dislocation and oxide particles, reflecting to advanced nuclear reactor design of next generation. Development of some nuclear energy generating systems has been proposed and supported intensively under several international collaboration programs (Generation IV International Forum (GIF), Advanced Fuel Cycle Initiative (AFCI), International Nuclear Energy Research Initiative (I-NERI) etc).Current research issue on ODS ferritic steels is considered to be poverty of experience and understanding on their practical neutron-irradiation behaviors at the temperature higher than 600C.In this research, a MA957, most familiar but primitive 14CrODS ferritic steel contained the highly textured-anisotropic grain structures, was irradiated at 500-700C to fast fluences ranging from 19.8 to 20.8 1026 n/m2 (E 0.1MeV) in the experimental fast reactor JOYO. The dose achieved varied from 99 to 104 dpa. TEM observation and micro-hardness measurement were carried out to clarify the irradiation effects on microstructural evolution of 14CrODS ferritic steel at elevated temperature and high dose. Microstructural examination revealed that all of the highly textured- anisotropic grain structures, following heavy irradiation at the temperature above 600C, have not changed. In addition, large regions in all specimens have retained high dislocation density, contained negligible cavitation.
Fujita, Mitsutane*; Xu, Y.*; Kaji, Yoshiyuki; Tsukada, Takashi; Mashiko, Shinichi*; Onose, Shoji*
RIST News, (38), p.3 - 14, 2004/11
The distributed materials database system named "Data-Free-Way(DFW)" has been developed with the collaboration of three organizations: the National Institute for Materials Science, the Japan Atomic Energy Research Institute, and the Japan Nuclear Cycle Development Institute over the Internet since 1990. At present, the development of a distributed knowledge based system, in which knowledge extracted from DFW is expressed, is planned with the collaboration of three organizations as we add data into DFW and make DFW open for the public use. Network technique and presentation and acquisition technique of the information developed rapidly and these techniques brought about a revolution in the society and our daily life changed. This paper describe the present status of DFW and future direction of the material databases with the transition of information technology.
Kaji, Yoshiyuki; Tsukada, Takashi; Fujita, Mitsutane*; Kinugawa, Junichi*; Yoshida, Kenji*; Mashiko, Shinichi*; Onose, Shoji*; Iwata, Shuichi*
2003-Nen Johogaku Shimpojiumu Koen Rombunshu, p.89 - 92, 2003/01
The distributed material database system named 'Data-Free-Way' has been developed by four organizations (the National Institute for Materials Science, the Japan Atomic Energy Research Institute, the Japan Nuclear Cycle Development Institute, and the Japan Science and Technology Corporation) under a cooperative agreement. In order to create additional values of the system, knowledge base system, in which knowledge extracted from the material database is expressed, is planned to be developed for more effective utilization of Data-Free-Way. XML (eXtensible Markup Language) has been adopted as the description method of the retrieved results and the meaning of them. This paper will describe the description method of knowledge extracted from the material database with XML and the distributed material knowledge base system.
*; *; ; Onose, Shoji
JNC-TY9400 2002-011, 87 Pages, 2002/08
no abstracts in English
; ; Onose, Shoji; ; Nakamura, Yasuo
JNC-TN9420 2002-003, 19 Pages, 2002/03
The Partitioning and Transmutation (P&T) for radionuclides included in high level has been researched in many countries. This technology for the radionuclides consists of partitioning them to several groups according to their half-lives and purposes of utilization and transmutating minor actinides (MA) and long lived fission products (LLFP) to short lived or stable nuclides. Japan Nuclear Cycle Development Institute (JNC) made a plan to develop this technology in the Feasibility Study for Fast Reactors and Related Fuel Cycle (FS), in cooperation with basic research groups. The main objective of JNC is to transmutate MA and LLFP in fast reactor. And this research was planned to carry out, taking into account not only reduction of environmnental burden and nuclear non-proliferation but also technical realization and economics. As a part of the research, the development of the elements for irradiation tests has just stated. According to the gained results of FS. The LLFPs, which have a possibility to realize the transmutation from the view point of nuclear physics, are I and Tc. Therefore, it was tried to select iodine chemical compounds fitted for transmutation by means of literature survey, because the half-life of I is long and the effect of radiation is comparatively hard. The literature survey was performed from the viewpoint of five properties, that is, nuclear physics, thermal phase change, chemical stability, fabrication, applicability to cycle use. As a result, 8 chemical compounds, namely, MgI, KI, NiI, CuI, RbI, YI, MoI, BaI were selected as target materials from 32 candidates.
; Onose, Shoji
Dai-3-Kai Shoshago Shiken Ni Kansuru Nikkan Semina, 0 Pages, 1999/00
; Abe, Yasuhiro; Ukai, S.; Onose, Shoji
Journal of Nuclear Materials, 272, p.173 - 178, 1999/00
; Onose, Shoji
Journal of Nuclear Science and Technology, 36(4), p.380 - 385, 1999/00
Akasaka, Naoaki; Onose, Shoji; Ukai, S.;
8th International Conference of Fusion Reactor Materials, 0 Pages, 1997/10
Maruyama, Tadashi; Onose, Shoji; Kaito, Takeji; Horiuchi, Hiroto
Journal of Nuclear Science and Technology, 34(10), p.1006 - 1014, 1997/10