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JAEA Reports

Research on diffusion behavior in oxide fuels (Joint research)

Sato, Isamu; Arima, Tatsumi*; Nishina, Masahiro*; Tanaka, Kosuke; Onose, Shoji; Idemitsu, Kazuya*

JAEA-Research 2012-006, 66 Pages, 2012/05

JAEA-Research-2012-006.pdf:13.55MB

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$$_{2}$$ fuel, the diffusion tests were performed. The diffusion coefficients were estimated to be 10$$^{-12}$$m$$^{2}$$/s $$sim$$ 10$$^{-14}$$m$$^{2}$$/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.

Journal Articles

MgO-based inert matrix fuels for a minor actinides recycling in a fast reactor cycle

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.

Journal Articles

Status of PIE technology development in JAEA-Oarai

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.

Journal Articles

Design and trial fabrication of a dismantling apparatus for irradiation capsules of solid tritium breeder materials

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

 Times Cited Count:0 Percentile:100(Materials Science, Multidisciplinary)

no abstracts in English

JAEA Reports

Investigation and design of the dismantling process of irradiation capsules containing tritium, 2; Detailed design and trial fabrication of capsule dismantling apparatus and investigation of glove box facility

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

JAEA-Technology-2009-007.pdf:31.88MB

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 (Li$$_{2}$$TiO$$_{3}$$), 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.

Journal Articles

Neutron irradiation effect on isotopically tailored $$^{11}$$B$$_{4}$$C

Morohashi, Yuko; Maruyama, Tadashi*; Donomae, Takako; Tachi, Yoshiaki; Onose, Shoji

Journal of Nuclear Science and Technology, 45(9), p.867 - 872, 2008/09

 Times Cited Count:9 Percentile:42.43(Nuclear Science & Technology)

JAEA Reports

Investigation and design of the dismantling process of irradiation capsules containing tritium, 1; Conceptual investigation and basic design

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

JAEA-Technology-2008-010.pdf:11.31MB

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-$$gamma$$ 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.

Journal Articles

Effects of fast reactor irradiation conditions on tensile and transient burst properties of ferritic/martensitic steel claddings

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

 Times Cited Count:11 Percentile:35.12(Nuclear Science & Technology)

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.

Journal Articles

Neutron irradiation effects on $$^{11}$$B$$_{4}$$C and recovery by annealing

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

 Times Cited Count:4 Percentile:68.53(Materials Science, Ceramics)

Use of moderator materials in Fast Breeder Reactor (FBR) is effective for transmutation technology, and $$^{11}$$B$$_{4}$$C is one of the candidates. Up to now, the behavior of $$^{10}$$B$$_{4}$$C as the Control rod material is well known, but that of $$^{11}$$B$$_{4}$$C is hardly investigated. In this paper, the radiation effects of $$^{11}$$B$$_{4}$$C 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 $$^{11}$$B$$_{4}$$C pellets. But, bubble nucleation was found in grain and along grain boundaries of $$^{11}$$B$$_{4}$$C. And, it was shown that the conductivity of $$^{11}$$B$$_{4}$$C was higher than that of $$^{10}$$B$$_{4}$$C. During the annealing from room temperature to 1400$$^{circ}$$C, three recovery stages were found on thermal conductivity. It was suggested that, the recovery of B$$_{4}$$C was related to the dispersion behavior of helium. Judging from these results, as $$^{11}$$B$$_{4}$$C was mechanically more stable compared with $$^{10}$$B$$_{4}$$C under irradiation, it was shown that $$^{11}$$B$$_{4}$$C had high applicability for a moderator.

Journal Articles

Tensile and transient burst properties of advanced ferritic/martensitic steel claddings after neutron irradiation

Yano, Yasuhide; Yoshitake, Tsunemitsu; Yamashita, Shinichiro; Akasaka, Naoaki; Onose, Shoji; Takahashi, Heishichiro*

Journal of Nuclear Materials, 367-370(1), p.127 - 131, 2007/08

 Times Cited Count:10 Percentile:37.96(Materials Science, Multidisciplinary)

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.

Journal Articles

Interaction among dislocation and complex oxide particles in ODS steels haevily-irradiated at high temperature

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 500-700$$^{circ}$$C to fast fluences ranging from 19.8 to 20.8 $$times$$ 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 600$$^{circ}$$C, have not changed. In addition, large regions in all specimens have retained high dislocation density, contained negligible cavitation.

Journal Articles

Present and future status of distributed database for nuclear materials, Data-free-way

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.

Journal Articles

Development of distributed material knowledge base system based on XML

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.

JAEA Reports

None

*; *; ; Onose, Shoji

JNC-TY9400 2002-011, 87 Pages, 2002/08

JNC-TY9400-2002-011.pdf:7.54MB

no abstracts in English

JAEA Reports

First selection of the LLFP iodine compound by the literature on this subject

; ; Onose, Shoji; ; Nakamura, Yasuo

JNC-TN9420 2002-003, 19 Pages, 2002/03

JNC-TN9420-2002-003.pdf:0.66MB

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 $$^{129}$$I and $$^{99}$$Tc. Therefore, it was tried to select iodine chemical compounds fitted for transmutation by means of literature survey, because the half-life of $$^{129}$$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$$_{2}$$, KI, NiI$$_{2}$$, CuI, RbI, YI$$_{3}$$, MoI$$_{2}$$, BaI$$_{2}$$ were selected as target materials from 32 candidates.

Journal Articles

Determination on Irradiation Temperature using SiC Temperature Monitors

; Onose, Shoji

Dai-3-Kai Shoshago Shiken Ni Kansuru Nikkan Semina, 0 Pages, 1999/00

None

Journal Articles

Post-irradiation creep rupture properties of FBR grade 316 SS structural material

; Abe, Yasuhiro; Ukai, S.; Onose, Shoji

Journal of Nuclear Materials, 272, p.173 - 178, 1999/00

None

Journal Articles

Fabrication and Thermal Conductivity of Boron Carbide /Copper Cermet

; Onose, Shoji

Journal of Nuclear Science and Technology, 36(4), p.380 - 385, 1999/00

 Times Cited Count:23 Percentile:15.41(Nuclear Science & Technology)

None

Journal Articles

Effect of Temperature Change on Void Swelling in P,Ti-modified 316 Stainless Steel

Akasaka, Naoaki; Onose, Shoji; Ukai, S.;

8th International Conference of Fusion Reactor Materials, 0 Pages, 1997/10

None

Journal Articles

Effect of Fast Neutron Irradiation on the Properties of Boron Carbide Pellet

Maruyama, Tadashi; Onose, Shoji; Kaito, Takeji; Horiuchi, Hiroto

Journal of Nuclear Science and Technology, 34(10), p.1006 - 1014, 1997/10

None

48 (Records 1-20 displayed on this page)