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Journal Articles

Microstructural stability of ODS steel after very long-term creep test

Oka, Hiroshi; Tanno, Takashi; Yano, Yasuhide; Otsuka, Satoshi; Kaito, Takeji; Tachi, Yoshiaki

Journal of Nuclear Materials, 547, p.152833_1 - 152833_7, 2021/04

 Times Cited Count:1 Percentile:81.22(Materials Science, Multidisciplinary)

In order to evaluate the stability of nano-sized oxide particles and matrix structure of ODS cladding tube, which are the determinants of their high temperature strength, the microstructural observation was carried out after internal pressurized creep test at 700$$^{circ}$$C for over 45,000 hours. The specimens were the as-received and crept specimens of 9Cr-ODS steel with tempered martensite and 12Cr-ODS steel with recrystallized ferrite. Small platelet was cut out from the crept pressurized tube, then thinned to foil. Microstructural observation was conducted with TEM JEOL 2010F. As a result of the observation, it was confirmed that the size and number density of the nano-sized particles were almost unchanged even after the creep test. In addition, the tempered martensite structure, which is one of the determinants of the creep strength of 9Cr-ODS steel, was not significantly different between the as-received and crept specimen, indicating the stability of their matrix structure.

Journal Articles

Development of ODS tempered martensitic steel for high burn up fuel cladding tube of SFR

Otsuka, Satoshi; Tanno, Takashi; Oka, Hiroshi; Yano, Yasuhide; Tachi, Yoshiaki; Kaito, Takeji; Hashidate, Ryuta; Kato, Shoichi; Furukawa, Tomohiro; Ito, Chikara; et al.

2018 GIF Symposium Proceedings (Internet), p.305 - 314, 2020/05

Oxide Dispersion Strengthened (ODS) steel has been developed worldwide as a high-strength and radiation-tolerant steel used for advanced nuclear system. Japan Atomic Energy Agency (JAEA) has been developing ODS steel as the primary candidate material of Sodium cooled Fast Reactor (SFR) high burn-up fuel cladding tube. Application of high burn-up fuel to SFR core can contribute to improvement of economical performance of SFR in conjunction with volume and hazardousness reduction of radioactive waste. This paper described the current status and future prospects of ODS tempered martensitic steel development in JAEA for SFR fuel application.

Journal Articles

Method to reduce long-lived fission products by nuclear transmutations with fast spectrum reactors

Chiba, Satoshi*; Wakabayashi, Toshio*; Tachi, Yoshiaki; Takaki, Naoyuki*; Terashima, Atsunori*; Okumura, Shin*; Yoshida, Tadashi*

Scientific Reports (Internet), 7(1), p.13961_1 - 13961_10, 2017/10

 Times Cited Count:18 Percentile:94.32(Multidisciplinary Sciences)

Transmutation of long-lived fission products (LLFPs: $$^{79}$$Se, $$^{93}$$Zr, $$^{99}$$Tc, $$^{107}$$Pd, $$^{129}$$I, and $$^{135}$$Cs) into short-lived or non-radioactive nuclides by fast neutron spectrum reactors without isotope separation has been proposed as a solution to the problem of radioactive wastes disposal. Despite investigation of many methods, such transmutation remains technologically difficult. To establish an effective and efficient transmutation system, we propose a novel neutron moderator material, yttrium deuteride (YD$$_{2}$$), to soften the neutron spectrum leaking from the reactor core. Neutron energy spectra and effective half-lives of LLFPs, transmutation rates, and support ratios were evaluated with the continuous-energy Monte Carlo code MVP-II/MVP-BURN and the JENDL-4.0 cross section library. With the YD$$_{2}$$ moderator in the radial blanket and shield regions, effective half-lives drastically decreased from 10$$^{6}$$ to 10$$^{2}$$ years and the support ratios reached 1.0 for all six LLFPs. This successful development and implementation of a transmutation system for LLFPs without isotope separation contribute to developing a self-consuming cycle of LLFPs using fast spectrum reactors to reduce radioactive waste.

Journal Articles

Application of Laser-Induced Breakdown Spectroscopy (LIBS) analysis to molten alloy production process

Ozu, Akira; Tachi, Yoshiaki; Arita, Yuji*

Reza Kenkyu, 42(12), p.913 - 917, 2014/12

Laser-induced breakdown spectroscopy (LIBS) analysis has been applied to the molten alloy production process, in which simulated metals (Zr, Cu, Sm, Ce) are used instead of nuclear metallic fuels contained minor actinide (MA), with the aim of in-situ monitoring the elementary composition of the surface of the molten alloy in a chamber and vapor particles generated from the surface of the molten alloy. The variation in the ratio of elementary composition of the surface of the molten alloy in the crucible was successfully observed depending on temperature of the crucible. The elementary composition of the vapor particles appeared in the molten alloy chamber was also measured. The practical experimental results show that LIBS technique is very useful for investigating the elementary composition in the process and understanding the behavior of molten alloy in the crucible.

JAEA Reports

X-ray CT basic data about inspection of irradiated fuel assembly

Ishimi, Akihiro; Tachi, Yoshiaki; Katsuyama, Kozo; Misawa, Susumu*

JAEA-Data/Code 2014-012, 72 Pages, 2014/08


The Fuels Monitoring Section (FMS) of Japan Atomic Energy Agency (JAEA) has carried out examination of the fuel assemblies irradiated at Experimental Fast Reactor Joyo to verify about deformation and damage using X-ray computed tomography (CT) technique. This technique can observe deformation and internal information of the irradiated fuel assembly without dismantling and thus can apply to inspections of the irradiated fuel assembly in Fukushima Daiichi Nuclear Power Plant (1F). In order to obtain X-ray CT basic data for 1F fuel assembly inspection, the simulated specimens were made and the X-ray CT examinations were performed in the Fuels Monitoring Facility (FMF). This paper compiled the data about the X-ray CT examination of the simulated specimens.

Journal Articles

Reduction in degree of absorber-cladding mechanical interaction by shroud tube in control rods for the fast reactor

Donomae, Takako; Katsuyama, Kozo; Tachi, Yoshiaki; Maeda, Koji; Yamamoto, Masaya; Soga, Tomonori

Journal of Nuclear Science and Technology, 48(4), p.580 - 584, 2011/04

One of the challenges in developing a long-life control rod is to restrain absorber-cladding mechanical interaction (ACMI). Its lifetime was limited by ACMI, which is induced by the swelling and relocation of B$$_{4}$$C pellets. To restrain ACMI, a shroud tube was inserted into the gap between the B$$_{4}$$C pellets and the cladding tube. And sodium was selected as bonding material instead of helium to restrain increases in the pellet temperature. As a result of these improvements, the estimated lifetime of the control rod at Joyo was doubled. In this paper, the results of post irradiation examination are reported.

Journal Articles

Candidate iodides for LLFP transmutation in FR core

Tachi, Yoshiaki; Wakabayashi, Toshio*

Transactions of the American Nuclear Society, 103(1), p.268 - 269, 2010/11

It is very attractive technique to transmute long-lived fission products such as iodine-129 included in spent fuel by using Fast Reactors (FRs) to minimize environmental burden and toxic risk due to high level waste disposal. Iodine has low melting point (386K) and low boiling points (457K) compared with FR core temperature. Furthermore, it is corrosive against iron based materials. Then, it is the most important issue to search suitable chemical forms for FR core conditions, having high stability at elevated temperature and good compatibility with cladding material. From the viewpoint of melting point, nuclides generation by neutron irradiation, reactivity with stainless steel, easy fabrication method and recyclability, 5 kinds of iodides of CuI, MgI$$_{2}$$, YI$$_{3}$$, RbI and YI$$_{3}$$ were selected as the candidate chemical form of iodine for transmutation in FR. In order to evaluate stability of the candidate iodides at elevated temperature, TG-DTA of CuI, RbI, BaI$$_{2}$$ and YI$$_{3}$$ were performed. According to TG-DTA results, no significant mass change showed in BaI$$_{2}$$ beyond those temperatures. Mass of CuI, RbI and YI$$_{3}$$ started to be lost severely at the lower temperature than their melting points. Compatibility test between iodides and cladding materials were conducted. Cladding materials in contact with iodides were heated in a capsule filled with pure Ar gas at 873K for 5000h maximum. Results of cross-section observation showed that MgI$$_{2}$$ and YI$$_{3}$$ made pit corrosion on SUS316 and ODS. It appeared that the upper part of the inner surface test capsules with RbI, YI$$_{3}$$ were degraded severely. Based on the experimental results, the most applicable iodide for transmutation by FR is BaI$$_{2}$$ from the viewpoints of stability at elevated temperature and compatibility with cladding material.

Journal Articles

Effect of high dose/high temperature irradiation on the microstructure of heat resistant 11Cr ferritic/martensitic steels

Yamashita, Shinichiro; Yano, Yasuhide; Tachi, Yoshiaki; Akasaka, Naoaki

Journal of Nuclear Materials, 386-388, p.135 - 139, 2009/04

 Times Cited Count:10 Percentile:60.19(Materials Science, Multidisciplinary)

The heat resistant 11Cr ferritic/martensitic steels were irradiated at 400-670 $$^{circ}$$C up to 100 dpa in FFTF and JOYO. The microstructures of unirradiated 11Cr ferritic/martensitic steels consist of laths, dislocation, and carbide. Almost of the prior austenitic boundaries (PABs) were partially decorated with carbides. It was observed from the results of post irradiation microstructural examinations that the irradiation-induced microstructures were classified into the following three types depending on irradiation temperature; (1) When irradiated at 400-450 $$^{circ}$$C, both dislocation loops and cavities with less than 30 nm in diameter were formed in the ferrite phase. On the other hand, the void swelling was about 0.05%. (2) In the case of irradiation at moderate temperature (500-600 $$^{circ}$$C), the precipitates formation M$$_{23}$$C$$_{6}$$ carbide was primarily dominated. It was a most noticeable microstructural feature that the carbides; M$$_{23}$$C$$_{6}$$ and M$$_{6}$$C grew and covered the PABs at this temperature range. (3) Finally, when irradiation temperature was above 650 $$^{circ}$$C microstructures were drastically-changed. Microstructural observations revealed that formation and growth of equi-axial grain occurred in addition to recovery of laths, growth of carbides simultaneously at high temperature. This remarkable microstructural change might be closely related to a severe degradation in the mechanical properties.

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:56.05(Nuclear Science & Technology)

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:30.7(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

Research and development of minor actinide-containing fuel and target in a future integrated closed cycle system

Osaka, Masahiko; Serizawa, Hiroyuki; Kato, Masato; Nakajima, Kunihisa; Tachi, Yoshiaki; Kitamura, Ryoichi; Miwa, Shuhei; Iwai, Takashi; Tanaka, Kenya; Inoue, Masaki; et al.

Journal of Nuclear Science and Technology, 44(3), p.309 - 316, 2007/03

 Times Cited Count:25 Percentile:85.69(Nuclear Science & Technology)

no abstracts in English

JAEA Reports

Investigation of transmutation target for long-lived fission products; Basic examinations of iodide candidates, 2

Donomae, Takako; Tachi, Yoshiaki; Matsumoto, Shinichiro

JAEA-Research 2006-033, 35 Pages, 2006/07


no abstracts in English

JAEA Reports

Development of Am-dispersed MgO target; Fabrication and characterization of simulated Am-dispersed MgO target

Miwa, Shuhei; Osaka, Masahiko; Tachi, Yoshiaki; Tanaka, Kenya

JAEA-Research 2005-003, 25 Pages, 2006/01


In the technology for transmutation of Minor Actinides (MAs) to stable and short-lived nuclides by irradiation in the reactor, one promising candidate for the MA-containing ceramic form is a composite of MA compound and inert matrix, so called "target". In this study, surrogate target which is the combination of CeO$$_{2}$$ as a substitution of (Pu, Am)O$$_{2}$$ with MgO as the inert matrix is fabricated by a simple process based on commercial manufacturing technology for the present fuel. Structural analyses are performed by XRD, SEM observation and EDS analysis. From the results of analysis, the targets with high density, high homogenous dispersion of sphere, non reaction phase are obtained. Thermal stability and thermal conductivity of fabricated target are investigated. Dissolubility to nitride solution, thermal stability and thermal conductivity of fabricated target are investigated.

Journal Articles

Simple fabrication process for CeO$$_{2}$$-MgO composite as surrogate for actinide-containing target for use in nuclear fuel

Osaka, Masahiko; Miwa, Shuhei; Tachi, Yoshiaki

Ceramics International, 32(6), p.659 - 663, 2006/00

 Times Cited Count:20 Percentile:80.46(Materials Science, Ceramics)

A simple fabrication process for the fabrication of CeO$$_{2}$$-MgO composite as surrogate for actinide-containing target for the use in nuclear field is established. The process is adaptable to the present manufacturing technology for the nuclear fuel. The manufactured target have good characteristics, e.g. high density, good appearance etc. In particular, although the shape of CeO2 is not ideal sphere as expected but elliptic, thermal conductivity measurement results show that the elliptic shape is not disadvantageous. It means that the present simple process is comparable to that of rather complicated one, advanced sol-gel supported process.

Journal Articles

Development of minor actinide containing fuel/target for the use in a future integrated system of fast reactor and accelerator driven system

Osaka, Masahiko; Serizawa, Hiroyuki*; Kato, Masato; Inoue, Masaki; Nakajima, Kunihisa*; Tachi, Yoshiaki; Kitamura, Ryoichi; Oki, Shigeo; Miwa, Shuhei; Iwai, Takashi*; et al.

Proceedings of International Conference on Nuclear Energy System for Future Generation and Global Sustainability (GLOBAL 2005) (CD-ROM), 6 Pages, 2005/10

Development of minor actinide containing fuel/target, i.e., (Pu,Am)O$$_{2}$$-MgO, (Pu,Np)O$$_{2}$$-MgO, (U,Pu,Np)O$$_{2}$$, (U,Pu,Np)N and (Pu,Np,Zr)N, for the use in a future integrated system of fast reactor and accelerator driven system is underway as a collaborative work between JAERI and JNC. The present statuses of fabrication test and property measurements are given. Irradiation test in the experimental fast reactor JOYO is also mentioned.

JAEA Reports

Results of fundamental research and development of partitioning and transmutation technology of long-lived-nuclides for fast breeder reactor cycle

Yamashita, Kiyonobu; Ozawa, Masaki; Ikegami, Tetsuo; Harada, Hideo; Osaka, Masahiko; Oki, Shigeo; Tachi, Yoshiaki; Furutaka, Kazuyoshi; Nakamura, Shoji

JNC TN9420 2004-001, 106 Pages, 2005/03


Research Evaluation Committee carried out a pre-evaluation of Research and Development of Partitioning and Transmutation Technology of Long-Lived-Nuclides namely, partitioning, nuclear data, reactor physics, fuel, creative transmutation technology, in Aug. 2000. Following results are obtained from the research and development. For example, two extract ant systems, capable of recovering all actinides in spent fuel, were newly nominated in partitioning technology. Also, neutron capture cross sections of 7 nuclides of MA and Fission Products (FP) were determined in nuclear data measurements. Some of those measurements are for the first time in the world. An advanced measurement system of a full solid angle Bi4Ge3O12 detector etc. was developed to measure the energy dependence of the neutron capture cross sections. These achievements in the first phase are summarized in the report to promote the research and development in the second phase effectively.

Journal Articles

Interstitial atom behavior in neutron irradiated beta-silicon nitride.

Akiyoshi, Masafumi; Akasaka, Naoaki; Tachi, Yoshiaki; Yano, Toyohiko*

Abstract p303,22-P-02, 303 Pages, 2003/00

Focusing on the cover layer materials (as the Radon Barrier Materials), which could have the effect to restrain the radon from scattering into the air and the effect of the radiation shielding, we produced the radon barrier materials with crude bentonite on an experimental basis, using the rotary type comprehensive unit for grinding and mixing, through which we carried out the evaluation of the characteristics thereof.

JAEA Reports


*; Tachi, Yoshiaki; *; *; *; *; *

JNC TY9400 2001-003, 111 Pages, 2000/11


no abstracts in English

Journal Articles

Application of the distributed database (Data-Free-Way) on the analysis of mechanical properties in neutron irradiated 316 stainless steel

Fujita, Mitsutane*; Kinugawa, Junichi*; Tsuji, Hirokazu; Kaji, Yoshiyuki; Tachi, Yoshiaki*; Saito, Junichi*; Shimura, Kazuki*; Nakajima, Ritsuko*; Iwata, Shuichi*

Fusion Engineering and Design, 51-52, p.769 - 774, 2000/11

 Times Cited Count:1 Percentile:12.53(Nuclear Science & Technology)

no abstracts in English

JAEA Reports

Properties of lithium and its handling

; Kano, Shigeki; Tachi, Yoshiaki; *

JNC TN9410 2000-013, 89 Pages, 2000/09


Lithium is one of goodcoolants because of high boiling point (1317$$^{circ}$$C), small specific gravity (0.47 at 600$$^{circ}$$C) and large specific heat (1cal/g/$$^{circ}$$C). Therefore if lithium will be used in fast reactor for coolant, the heat efficiency of reactor will largely increase. Here the fundamental properties of lithium and the results of examination on chemical reaction, combustion and extinction are shown. These examinations were also carried out on sodium to compare with lithium. The differences between both are that lithium reacts more moderately with water, not explosive, and is not combustible but after ignition burns at higher temperature and longer.

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