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Journal Articles

Initial sintering kinetics of non-stoichiometric CeO$$_{2-x}$$

Watanabe, Masashi; Seki, Takayuki*

Materials Science & Engineering B, 272, p.115369_1 - 115369_6, 2021/10

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

The effect of oxygen non-stoichiometry on the initial sintering behavior of CeO$$_{2}$$ was investigated. It was found that the initial sintering of the stoichiometric and hypo-stoichiometric composition was controlled by the grain boundary diffusion. The activation energies of cation diffusion were derived from initial sintering data. Moreover, it is suggested that the cation diffusion was caused by a vacancy mechanism.

Journal Articles

Leaching behavior of radionuclides from samples prepared from spent fuel rod comparable to core debris in the 1F NPS

Onishi, Takashi; Maeda, Koji; Katsuyama, Kozo

Journal of Nuclear Science and Technology, 58(4), p.383 - 398, 2021/04

 Times Cited Count:1 Percentile:0.01(Nuclear Science & Technology)

Journal Articles

Adsorption behavior of cesium on hybrid microcapsules in spent fuel solution

Onishi, Takashi; Koyama, Shinichi; Mimura, Hitoshi*

Nihon Ion Kokan Gakkai-Shi, 31(3), p.43 - 49, 2020/10

Journal Articles

Release behavior of radionuclides from MOX fuels irradiated in a fast reactor during heating tests

Tanaka, Kosuke; Sato, Isamu*; Onishi, Takashi; Ishikawa, Takashi; Hirosawa, Takashi; Katsuyama, Kozo; Seino, Hiroshi; Ohno, Shuji; Hamada, Hirotsugu; Tokoro, Daishiro*; et al.

Journal of Nuclear Materials, 536, p.152119_1 - 152119_8, 2020/08

 Times Cited Count:0 Percentile:0.01(Materials Science, Multidisciplinary)

In order to obtain the release rate coefficients from fuels for fast reactors (FRs), heating tests and the subsequent analyses of the fission products (FPs) and actinides that are released were carried out using samples of uranium-plutonium mixed oxide (MOX) fuel pellets irradiated at the experimental fast reactor Joyo. Three heating tests targeting temperatures of 2773, 2973 and 3173 K were conducted using an FP release behavior test apparatus equipped with a high-frequency induction furnace and solid FP sampling systems consisting of a thermal gradient tube (TGT) and filters. Irradiated fuel pellets were placed into a tungsten crucible, then loaded into the induction furnace. The temperature was raised continuously at a heating rate of 10 K/s to the targeted temperature and maintained for 500 s in a flowing argon gas atmosphere. The FPs and actinides released from the MOX fuels and deposited in the TGT and filters were quantified by gamma-ray spectrometry and inductively coupled plasma mass spectrometry (ICP-MS) analysis. Based on the analysis, the release rates of radionuclides from MOX fuels for FR were obtained and compared with literature data for light water reactor (LWR) fuels. The release rate coefficients of FPs obtained in this study were found to be similar to or lower than the literature values for LWR fuels. It was also found that the release rate coefficient data for actinides were within the range of variation of literature values for LWR fuels.

Journal Articles

Concerning aging of nuclear fuel material use facilities Examination of measures to improve safety assessment methods

Sakamoto, Naoki; Fujishima, Tadatsune; Mizukoshi, Yasutaka

Hozengaku, 19(2), p.125 - 126, 2020/07

The five post-irradiation examination facilities in JAEA's Oarai research and development institute have been operated for over 40 years in order to investigate the irradiation performance of fast reactor fuel materials. The equipment associated with these facilities has been managed to maintain secure from the problems occurred in the process of aging. Therefore, we established a safety assessment method for aging facilities in 2002, and we have been conducting maintenance management of facilities since then. In this study, improvement plans of the safety assessment method are considered in order to solve the issues detected as a result of analysis of past maintenance information.

Journal Articles

Adsorption of platinum-group metals and molybdenum onto aluminum ferrocyanide in spent fuel solution

Onishi, Takashi; Sekioka, Ken*; Suto, Mitsuo*; Tanaka, Kosuke; Koyama, Shinichi; Inaba, Yusuke*; Takahashi, Hideharu*; Harigai, Miki*; Takeshita, Kenji*

Energy Procedia, 131, p.151 - 156, 2017/12

 Times Cited Count:6 Percentile:97.92

no abstracts in English

Journal Articles

Fabrication and short-term irradiation behaviour of Am-bearing MOX fuels

Kihara, Yoshiyuki; Tanaka, Kosuke; Koyama, Shinichi; Yoshimochi, Hiroshi; Seki, Takayuki; Katsuyama, Kozo

NEA/NSC/R(2017)3, p.341 - 350, 2017/11

In order to investigate the effect of the addition of americium to MOX fuels on the irradiation behaviour, the "Am-1" program is being conducted at the JAEA. The Am-1 program consists of two short-term irradiation tests of 10-min and 24-h irradiation periods, and a steady-state irradiation test. The short-term irradiation tests and their post irradiation examinations (PIEs) have been successfully completed. To date, the data for PIE of the Am-MOX fuels focused on the microstructural evolution and redistribution behaviour of Am at the initial stage of irradiation have been obtained and reported. In this paper, the results obtained from the Am-1 program are reviewed and detailed descriptions of the fabrication and inspection techniques for the Am-MOX fuels prepared for the program are provided. PIE data for the Am-MOX fuels at the initial stage of irradiation have been accumulated. In this paper, unpublished PIE data for the Am-MOX fuels are also presented.

Journal Articles

Oxidation and reduction behaviors of a prototypic MgO-PuO$$_{2-x}$$ inert matrix fuel

Miwa, Shuhei; Osaka, Masahiko

Journal of Nuclear Materials, 487, p.1 - 4, 2017/04

 Times Cited Count:4 Percentile:51.35(Materials Science, Multidisciplinary)

Oxidation and reduction behaviors of prototypic MgO-based inert matrix fuels (IMFs) containing PuO$$_{2-x}$$ were experimentally investigated by means of thermogravimetry. The oxidation and reduction kinetics of the MgO-PuO$$_{2-x}$$ specimen were determined. The oxidation and reduction rates of the MgO-PuO$$_{2-x}$$ were found to be low compared with those of PuO$$_{2-x}$$. It is note that the changes in O/Pu ratios of MgO-PuO$$_{2-x}$$ from stoichiometry were smaller than those of PuO$$_{2-x}$$ at high oxygen partial pressure. From these results, it can be said that MgO matrix lower the oxygen supply and release of PuO$$_{2-x}$$, which is preferable as the minor actinides incineration devices, since the high oxygen potentials of minor actinide oxides can cause certain problems in terms of thermochemical aspects such as enlarged cladding inner-surface corrosion.

Journal Articles

High temperature physicochemical properties of irradiated fuels

Ishikawa, Takashi; Onishi, Takashi; Hirosawa, Takashi; Tanaka, Kosuke; Katsuyama, Kozo

Proceedings of 54th Annual Meeting of Hot Laboratories and Remote Handling (HOTLAB 2017) (Internet), 10 Pages, 2017/00

Journal Articles

Development of a method of periodic safety review to cope with the aging degradation of hot laboratories

Tamaoki, Yuichi; Omori, Tsuyoshi; Fujishima, Tadatsune; Mizukoshi, Yasutaka; Sakamoto, Naoki

Proceedings of 53rd Annual Meeting of Hot Laboratories and Remote Handling Working Group (HOTLAB 2016) (Internet), 6 Pages, 2016/11

Post irradiation examinations are conducted in hot laboratories in Oarai Research and Development Center of the Japan Atomic Energy Agency in order to develop fuels and materials for fast reactors. These facilities, the majority of which were constructed in the 1970s, have accumulated operating experience over a period of more than 40 years. Continuous operational safety requires the maintenance of important equipment such as electronic devices, manipulators, in-cell cranes, as well as air conditioning and ventilation systems. A method for the periodic safety review for hot laboratories based on the periodic safety review method employed for preventive maintenance at commercialized power reactors in Japan has been developed. In this paper, the status of implementation of the periodic safety review for hot laboratories using the safety review method are introduced.

Journal Articles

Oxygen potential of a prototypic Mo-cermet fuel containing plutonium oxide

Miwa, Shuhei; Osaka, Masahiko; Nozaki, Takahiro*; Arima, Tatsumi*; Idemitsu, Kazuya*

Journal of Nuclear Materials, 465, p.840 - 842, 2015/10

 Times Cited Count:1 Percentile:12.13(Materials Science, Multidisciplinary)

Oxygen potential of a prototypic Mo-cermet fuel containing PuO$$_{2-x}$$ was experimentally determined. It was shown that the oxygen potentials of Mo-cermet fuel containing PuO$$_{2-x}$$ were the same as those of pure PuO$$_{2-x}$$. It was also confirmed that the gradual oxidation of the Mo matrix occurred only above the oxygen potential of Mo/MoO$$_{2}$$. It is concluded that the oxidation-reduction behavior of the Mo-cermet fuel can be evaluated individually for each phase of actinides oxide and Mo matrix. Better phase structures of the Mo-cermet fuel for taking full advantage of the oxidation-reduction controllability were suggested by the confinement of the actinides oxide phase with Mo.

Journal Articles

Thermophysical properties of americium-containing barium plutonate

Tanaka, Kosuke; Sato, Isamu; Hirosawa, Takashi; Kurosaki, Ken*; Muta, Hiroaki*; Yamanaka, Shinsuke*

Journal of Nuclear Science and Technology, 52(10), p.1285 - 1289, 2015/10

 Times Cited Count:2 Percentile:22.47(Nuclear Science & Technology)

Polycrystalline specimens of americium-containing barium plutonate have been prepared by mixing the appropriate amounts of (Pu$$_{0.91}$$Am$$_{0.09}$$)O$$_{2}$$ and BaCO$$_{3}$$ powders followed by reacting and sintering at 1600 K under the flowing gas atmosphere of dry-air. The sintered specimens had a single phase of orthorhombic perovskite structure and were crack-free. The elastic moduli were determined from the longitudinal and shear sound velocities. The Debye temperature was also determined from the sound velocities and lattice parameter measurements. The thermal conductivity was calculated from the measured density at room temperature, literature values of heat capacity, and thermal diffusivity measured by laser flash method in vacuum. The thermal conductivity of americium-containing barium plutonate was roughly independent of the temperature and was almost the same magnitude as that of BaPuO$$_{3}$$ and BaUO$$_{3}$$.

Journal Articles

Early-in-life fuel restructuring behavior of Am-bearing MOX fuels

Tanaka, Kosuke; Sasaki, Shinji; Katsuyama, Kozo; Koyama, Shinichi

Transactions of the American Nuclear Society, 113(1), p.619 - 621, 2015/10

In order to evaluate the microstructural change behavior of Am-MOX fuels at the initial stage of irradiation, detailed investigations using image analysis were performed on X-ray Computed Tomography (X-ray CT) images and on ceramographs from fuels irradiated in both B11 and B14.

Journal Articles

Chemical form consideration of released fission products from irradiated fast reactor fuels during overheating

Sato, Isamu; Tanaka, Kosuke; Koyama, Shinichi; Matsushima, Kenichi*; Matsunaga, Junji*; Hirai, Mutsumi*; Endo, Hiroshi*; Haga, Kazuo*

Energy Procedia, 82, p.86 - 91, 2015/07

 Times Cited Count:2 Percentile:22.47(Nuclear Science & Technology)

Experiments simulating overheating conditions of fast reactor severe accidents have been previously carried out with irradiated fuels. For the present study, the chemical forms of the fission products (FPs) included in the irradiated fuels were evaluated by thermochemical equilibrium calculations. At temperatures of 2773 K and 2973 K, the most stable forms of Cs, I, Te, Sb, Pd and Ag are gaseous compounds. Cs and Sb detected in the thermal gradient tube (TGT) in the experiments can take gaseous chemical forms of elemental Cs, CsI, Cs$$_{2}$$MoO$$_{4}$$, CsO and elemental Sb, SbO, SbTe, respectively. By comparing experimental results and the estimations, it is seen CsI thermochemically behaves in a manner that traps it in the TGT, while elemental Cs trends to move as fine particles. The moving behavior of the gaseous FPs will obey not only thermochemical principles, but also those of particle dynamics.

Journal Articles

Removal of zirconium from spent fuel solution by alginate gel polymer

Onishi, Takashi; Koyama, Shinichi; Mimura, Hitoshi*

Progress in Nuclear Energy, 82, p.69 - 73, 2015/07

 Times Cited Count:4 Percentile:41.12(Nuclear Science & Technology)

no abstracts in English

Journal Articles

Influence of boron vapor on transport behavior of deposited CsI during heating test simulating a BWR severe accident condition

Sato, Isamu; Onishi, Takashi; Tanaka, Kosuke; Iwasaki, Maho; Koyama, Shinichi

Journal of Nuclear Materials, 461, p.22 - 28, 2015/06

We observed one of the simplified processes by conducting primitive experiments. CsI was heated at 1323 K to be vaporized and deposited on sampling parts with a temperature range of 1023 - 423 K and then B$$_{2}$$O$$_{3}$$ was vaporized at 1973 K to be reacted with Cs/I there. After heating tests, each sampling part was soaked into alkali water to dissolve the surface-deposits for ICP-MS analysis. The results showed that CsI deposited at the sampling parts kept above approx. 850 K was striped by B$$_{2}$$O$$_{3}$$ vapour. This behaviour will be thermodynamically discussed to study the Cs/I/B chemistry in the severe accidents.

Journal Articles

Influence of boron vapor on transport behavior of deposited CsI during heating test simulating a BWR severe accident condition

Sato, Isamu; Onishi, Takashi; Tanaka, Kosuke; Iwasaki, Maho; Koyama, Shinichi

Journal of Nuclear Materials, 461, p.22 - 28, 2015/06

 Times Cited Count:6 Percentile:55.25(Materials Science, Multidisciplinary)

In order to evaluate B influence on the release and transport of Cs and I during severe accidents, basic experiments have been performed on the interaction between deposited Cs/I compounds and vapor/aerosol B compounds. CsI and B$$_{2}$$O$$_{3}$$ were utilized as a Cs/I compound and a B compound, respectively. Deposited CsI on the thermal gradient tube (TGT), which is exposed to temperatures ranging from 423 K to 1023 K was reacted with vapor/aerosol B$$_{2}$$O$$_{3}$$, and then observed to determine how it changed Cs/I decomposition profiles. As a result, vapor/aerosol B$$_{2}$$O$$_{3}$$ stripped a portion of deposited CsI within a temperature range from 830 K to 920 K to make gaseous CsBO$$_{2}$$ and I$$_{2}$$. In addition, gaseous I$$_{2}$$ was re-deposited at a temperature range from 530 K to 740 K, while CsBO$$_{2}$$ travelled through the sampling tubes and filters without deposition. It is implied that B influences Cs carriers such as CsBO$$_{2}$$ to transport Cs to the colder regions.

Journal Articles

Oxidation behavior of Am-containing MOX fuel pellets in air

Tanaka, Kosuke; Yoshimochi, Hiroshi; Obayashi, Hiroshi; Koyama, Shinichi

Energy Procedia, 71, p.282 - 292, 2015/05

 Times Cited Count:2 Percentile:88.75

Americium-containing MOX (Am-MOX) fuels were subjected to heating tests using thermogravimetric and differential thermal analysis (TG-DTA) measurements in a flowing gas atmosphere of dry air to investigate the effect of Am addition on oxidation behavior of MOX fuel.

Journal Articles

Penetration behavior of water solution containing radioactive species into dried concrete/mortar and epoxy resin materials

Sato, Isamu; Maeda, Koji; Suto, Mitsuo; Osaka, Masahiko; Usuki, Toshiyuki; Koyama, Shinichi

Journal of Nuclear Science and Technology, 52(4), p.580 - 587, 2015/04

 Times Cited Count:5 Percentile:48.62(Nuclear Science & Technology)

Penetration behavior of radionuclides such as $$^{137}$$Cs into dried concrete material, dried mortar material and epoxy paint for a few dozen days was observed using a solution containing fission products extracted from irradiated fuels to obtain fundamental information on the radionuclide penetration rate and depth. Hardly any radionuclides could penetrate into the epoxy paint. The radionuclide solution penetrated into concrete and mortar materials to a depth of a few millimeters for a few dozen days. The penetration behavior observed near the surface of concrete and mortar materials was similar to the diffusion of nuclides in media such as water-saturated concrete, bentonite and cement materials.

Journal Articles

Chromatographic separation of nuclear rare metals by highly functional xerogels

Onishi, Takashi; Koyama, Shinichi; Masud, R. S.*; Kawamura, Takuya*; Mimura, Hitoshi*; Niibori, Yuichi*

Nihon Ion Kokan Gakkai-Shi, 25(4), p.220 - 227, 2014/11

no abstracts in English

257 (Records 1-20 displayed on this page)