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JAEA Reports

Stabilization treatment of Pu-bearing organic materials

Morishita, Kazuki; Sato, Takumi; Onishi, Takashi; Seki, Takayuki*; Sekine, Shinichi*; Okitsu, Yuichi*

JAEA-Technology 2021-024, 27 Pages, 2021/10

JAEA-Technology-2021-024.pdf:2.41MB

In the case of Plutonium (Pu)-bearing organic materials, organic materials are decomposed by alpha rays emitted mainly from Pu to generate hydrogen gas and other substances. Therefore, to safely store Pu-bearing organic materials for an extended period of time, organic materials must be eliminated. In addition, carbide and nitride fuels must be converted into oxides for safe storage in order to prevent the exothermal reaction of these fuels with oxygen/moisture in air. A survey of the literature on the stabilization treatment of Pu-bearing organic materials confirmed that organic materials can be decomposed and removed by heating at 950 $$^{circ}$$C (1223.15 K) or greater in air. Furthermore, based on the calculated thermodynamic parameters of oxidation reaction of carbide and nitride fuels in air, it was estimated that these fuels would be oxidized in air at 950 $$^{circ}$$C because the equilibrium oxygen partial pressure in the oxidation reaction at 950 $$^{circ}$$C was lower than 2.1$$times$$10$$^{4}$$ Pa (oxygen partial pressure in air). Therefore, it was decided to stabilize Pu-bearing organic materials by heating at 950 $$^{circ}$$C in air to remove the organic materials and oxidize the carbide and nitride fuels. As a mock-up test to remove the organic materials, thin sheets of epoxy resin were heated in air. The changes in appearance and weight before and after heating in air showed that organic materials can be removed. After the mock-up test, Pu-bearing organic materials were also stabilized by heating in the similar condition.

JAEA Reports

Study on the evaluation method to determine the radioactivity concentration in radioactive waste on Oarai Research and Development Institute (FY2020)

Asakura, Kazuki; Shimomura, Yusuke; Donomae, Yasushi; Abe, Kazuyuki; Kitamura, Ryoichi; Miyakoshi, Hiroyuki; Takamatsu, Misao; Sakamoto, Naoki; Isozaki, Ryosuke; Onishi, Takashi; et al.

JAEA-Review 2021-020, 42 Pages, 2021/10

JAEA-Review-2021-020.pdf:2.95MB

The disposal of radioactive waste from the research facility need to calculate from the radioactivity concentration that based on variously nuclear fuels and materials. In Japan Atomic Energy Agency Oarai Research and Development Institute, the study on considering disposal is being advanced among the facilities which generate radioactive waste as well as the facilities which process radioactive waste. This report summarizes a study result in FY2020 about the evaluation method to determine the radioactivity concentration in radioactive waste on Oarai Research and Development Institute.

Journal Articles

Summary results of subsidy program for the "Project of Decommissioning and Contaminated Water Management (Development of Analysis and Estimation Technology for Characterization of Fuel Debris (Development of Technologies for Enhanced Analysis Accuracy and Thermal Behavior Estimation of Fuel Debris))"

Koyama, Shinichi; Nakagiri, Toshio; Osaka, Masahiko; Yoshida, Hiroyuki; Kurata, Masaki; Ikeuchi, Hirotomo; Maeda, Koji; Sasaki, Shinji; Onishi, Takashi; Takano, Masahide; et al.

Hairo, Osensui Taisaku jigyo jimukyoku Homu Peji (Internet), 144 Pages, 2021/08

JAEA performed the subsidy program for the "Project of Decommissioning and Contaminated Water Management (Development of Analysis and Estimation Technology for Characterization of Fuel Debris (Development of Technologies for Enhanced Analysis Accuracy and Thermal Behavior Estimation of Fuel Debris))" in 2020JFY. This presentation summarized briefly the results of the project, which will be available shortly on the website of Management Office for the Project of Decommissioning and Contaminated Water Management.

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:39.17(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

Neutron scattering study of yttrium iron garnet

Shamoto, Shinichi; Ito, Takashi; Onishi, Hiroaki; Yamauchi, Hiroki; Inamura, Yasuhiro; Matsuura, Masato*; Akatsu, Mitsuhiro*; Kodama, Katsuaki; Nakao, Akiko*; Moyoshi, Taketo*; et al.

Physical Review B, 97(5), p.054429_1 - 054429_9, 2018/02

 Times Cited Count:10 Percentile:63.24(Materials Science, Multidisciplinary)

Nuclear and magnetic structure and full magnon dispersions of yttrium iron garnet Y$$_3$$Fe$$_5$$O$$_{12}$$ have been studied by neutron scattering. The lowest-energy dispersion below 14 meV exhibits a quadratic dispersion as expected from ferromagnetic magnons. The imaginary part of $$q$$-integrated dynamical spin susceptibility $$chi$$"($$E$$) exhibits a square-root energy-dependence in the low energies. The magnon density of state is estimated from the $$chi$$"($$E$$) obtained on an absolute scale. The value is consistent with a single chirality mode for the magnon branch expected theoretically.

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:97.85

no abstracts in English

Journal Articles

Effects of $$gamma$$ irradiation on the adsorption characteristics of xerogel microcapsules

Onishi, Takashi; Tanaka, Kosuke; Koyama, Shinichi; Ou, L. Y.*; Mimura, Hitoshi*

NEA/NSC/R(2017)3, p.463 - 469, 2017/11

no abstracts in English

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

Beam commissioning of the linac for iBNCT

Naito, Fujio*; Anami, Shozo*; Ikegami, Kiyoshi*; Uota, Masahiko*; Ouchi, Toshikatsu*; Onishi, Takahiro*; Oba, Toshiyuki*; Obina, Takashi*; Kawamura, Masato*; Kumada, Hiroaki*; et al.

Proceedings of 13th Annual Meeting of Particle Accelerator Society of Japan (Internet), p.1244 - 1246, 2016/11

The proton linac installed in the Ibaraki Neutron Medical Research Center is used for production of the intense neutron flux for the Boron Neutron Capture Therapy (BNCT). The linac consists of the 3-MeV RFQ and the 8-MeV DTL. Design average beam current is 10mA. Target is made of Beryllium. First neutron production from the Beryllium target was observed at the end of 2015 with the low intensity beam as a demonstration. After the observation of neutron production, a lot of improvement s was carried out in order to increase the proton beam intensity for the real beam commissioning. The beam commissioning has been started on May 2016. The status of the commissioning is summarized in this report.

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:40.75(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:54.79(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

Applicability study of nuclear graphite material IG-430 to VHTR

Osaki, Hirotaka; Shimazaki, Yosuke; Sumita, Junya; Shibata, Taiju; Konishi, Takashi; Ishihara, Masahiro

Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 8 Pages, 2015/05

For the design on the VHTR graphite components, it is desirable to employ graphite material with higher strength. IG-430 graphite has been developed as an advanced candidate for VHTR. However, the new developed IG-430 does not have enough databases for the design of HTGR. In this paper, the compressive strength (Cs) of IG-430, one of important strengths for design data, is statistically evaluated. The component reliability is evaluated based on the safety factors defined by the graphite design code, and the applicability as the VHTR graphite material is discussed. It was found that IG-430 has higher strength (about 11%) and lower standard deviation (about 27%) than IG-110 which is one of traditional graphites used for HTGR, because the crack in IG-430 would not easy to propagate rather than IG-110. Since fracture probability for IG-430 is low, the higher reliability of core-component will be achieved using IG-430. It is expected that IG-430 is applicable for VHTR graphite material.

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

Journal Articles

Compatibility of Ni and F82H with liquid Pb-Li under rotating flow

Kanai, Akihiko*; Park, C.*; Noborio, Kazuyuki*; Kasada, Ryuta*; Konishi, Satoshi*; Hirose, Takanori; Nozawa, Takashi; Tanigawa, Hiroyasu

Fusion Engineering and Design, 89(7-8), p.1653 - 1657, 2014/10

 Times Cited Count:4 Percentile:36.81(Nuclear Science & Technology)

Journal Articles

Research program for the evaluation of fission product release and transport behavior focusing on FP chemistry

Sato, Isamu; Miwa, Shuhei; Tanaka, Kosuke; Nakajima, Kunihisa; Hirosawa, Takashi; Iwasaki, Maho; Onishi, Takashi; Osaka, Masahiko; Takai, Toshihide; Amaya, Masaki; et al.

Proceedings of 2014 Water Reactor Fuel Performance Meeting/ Top Fuel / LWR Fuel Performance Meeting (WRFPM 2014) (USB Flash Drive), 6 Pages, 2014/09

A new research program on severe accidents is lunched for the evaluation of FP release and transport behavior in BWR system. The purpose of the program is to improve the FP release and transport model using experimental database about FP chemistry focusing on Cs and I chemistry. In this program, effects of B including in control rod materials, B$$_{4}$$C for the Cs and I chemistry are paid attention. The experimental database used for the improvement will consist of results to obtain with newly-prepared test device under atmosphere with broad-ranging oxygen and/or steam partial pressure simulated those in BWR. The state of preparation for these experimental studies and analyses is introduced. In addition, the preliminary test was moved into action to show B chemical effect on Cs and I transport under one of the processes, which is deposited Cs compounds and B vapor and aerosol interaction. In this experiment, a "B stripping effect" to deposited CsI was observed.

Journal Articles

Observation of a $$p$$-wave one-neutron halo configuration on $$^{37}$$Mg

Kobayashi, Nobuyuki*; Nakamura, Takashi*; Kondo, Yosuke*; Tostevin, J. A.*; Utsuno, Yutaka; Aoi, Nori*; Baba, Hidetada*; Barthelemy, R.*; Famiano, M. A.*; Fukuda, Naoki*; et al.

Physical Review Letters, 112(24), p.242501_1 - 242501_5, 2014/06

 Times Cited Count:70 Percentile:94.21(Physics, Multidisciplinary)

no abstracts in English

Journal Articles

Development of evaluation method with X-ray tomography for material property of IG-430 graphite for VHTR/HTGR

Sumita, Junya; Shibata, Taiju; Fujita, Ichiro*; Kunimoto, Eiji*; Yamaji, Masatoshi*; Eto, Motokuni*; Konishi, Takashi*; Sawa, Kazuhiro

Nuclear Engineering and Design, 271, p.314 - 317, 2014/05

 Times Cited Count:7 Percentile:55.77(Nuclear Science & Technology)

In this study, in order to develop evaluation method for material properties and to evaluate the irradiation-induced property changes under higher neutron doses for IG-430, the oxidation and densification effects on elastic modulus of IG-430 were investigated. Moreover, the correlation of the microstructure based on the X-ray tomography images and the material properties was discussed. It was shown that the elastic modulus of the densified graphite depends on only the closed pores and it is possible to evaluate the material properties of graphite by using X-ray tomography method. However, it is necessary to take into account of the change in the number and shape of closed pores in the grain to simulate the elastic modulus of the highly oxidized and irradiated materials by the homogenization analysis.

152 (Records 1-20 displayed on this page)