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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.
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：11 Percentile：98.3(Energy & Fuels)no abstracts in English
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：6 Percentile：45.92(Nuclear Science & Technology)Penetration behavior of radionuclides such as 
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.
Maeda, Koji; Sasaki, Shinji; Kumai, Misaki; Sato, Isamu; Suto, Mitsuo; Osaka, Masahiko; Goto, Tetsuo*; Sakai, Hitoshi*; Chigira, Takayuki*; Murata, Hirotoshi*
Journal of Nuclear Science and Technology, 51(7-8), p.1006 - 1023, 2014/07
Times Cited Count：14 Percentile：72.55(Nuclear Science & Technology)Since the start of the severe accident at the Fukushima Daiichi Nuclear Power Plant in March 2011, concrete surfaces within the reactor buildings have been exposed to radioactive contaminants. Released radiation sources still remain too high to permit entry into some areas of the RBs to allow the damage to be assessed and to allow carrying out the restoration of lost safety functions, decommissioning activities, etc. In order to clarify the situation of this contamination in the RBs, 18 samples were subjected to analyses to determine the surface radionuclide concentrations and to characterize the radionuclide distributions in the samples. Decontamination tests on the sample of Unit 2 were conducted to reduce the levels of radioactivity present near the sample surface. As a result of the tests, the level of radioactivity of the sample was reduced with the removal of 97% of the contamination present near the sample surface.
Usuki, Toshiyuki; Sato, Isamu; Suto, Mitsuo; Maeda, Koji; Osaka, Masahiko; Koyama, Shinichi; Tokoro, Daishiro*; Sekioka, Ken*; Ishigamori, Toshio*
JAEA-Testing 2014-001, 29 Pages, 2014/05
JAEA-Testing-2014-001.pdf:5.33MB
The penetration tests with solution containing radioactive nuclides were experimented to understand basic data for floor and wall materials of Fukushima Daiichi reactor buildings. The solution prepared from irradiated fuels was used as solution containing radioactive nuclides. The solution was applied to surface of epoxy paint, dried concrete and mortar used as specimens. Dose-rate profiles of direction of depth were given by radiation measurement and grinding of the specimens. The penetrations of radioactive nuclides for epoxy paint specimens were not clearly observed and the penetration depths would be within 0.4 mm. The penetrations of radioactive nuclides for dried concrete specimens proceeded. The penetration rates were substantially decreased when 16 days have elapsed from start. The dose rates of penetrated dried concrete specimens were reduced to background by grinding-2.0 mm. 
-ray spectrometry measurement showed that penetration behavior of near surface concrete are different among nuclides and the penetration behavior of radioactive nuclides into dried concrete and mortar materials through solution is similar to migration behavior of ions into those water-saturated materials.
-1) towards the decommissioning of Fukushima Daiichi Nuclear Power plant; Results of Examinations of contaminated samples at JAEA hot laboratoriesMaeda, Koji; Sasaki, Shinji; Kumai, Misaki; Sato, Isamu; Suto, Mitsuo; Osaka, Masahiko
JAEA-Research 2013-025, 123 Pages, 2014/01
JAEA-Research-2013-025-01.pdf:50.58MBJAEA-Research-2013-025-02.pdf:61.94MBJAEA-Research-2013-025-03.pdf:52.86MBJAEA-Research-2013-025-04.pdf:61.52MBJAEA-Research-2013-025-05.pdf:44.49MB
In order to clarify the situation of the contamination in the Fukushima Daiichi reactor buildings of Units 1, 2 and 3, selected samples were transported to the Oarai Engineering Center of JAEA where they were subjected to analyses to determine the surface radionuclide concentrations and to characterize the radionuclide distributions in the samples. The analysis results indicate that the situation of contamination in the building of Unit 2 was different from others, and the protective surface coatings on the concrete floors provided significant protection against radionuclide penetration. contaminants.
Tanaka, Kosuke; Suto, Mitsuo; Onishi, Takashi; Akutsu, Yoko; Yoshitake, Tsunemitsu; Yamashita, Shinichiro; Sekioka, Ken*; Ishigamori, Toshio*; Obayashi, Hiroshi; Koyama, Shinichi
JAEA-Research 2013-036, 31 Pages, 2013/12
JAEA-Research-2013-036.pdf:3.31MB
In the accident of Fukushima Daiichi NPPs, the water ingress was performed in order to decrease the reactor temperature. At that time, sea water was temporarily used as a coolant and the water contacted with nuclear fuel directly. It can be supposed that fission products (FP) were easily migrated from the fuel to sea water in this situation and that affect the water quality. The knowledge of leaching behavior, therefore, is necessary for evaluating the integrity of reactor component materials such as steels for pressure containment vessel and for reactor vessel. In order to obtain the fundamental knowledge for leaching behavior of FP in the hot sea water, the leaching tests of irradiated fuel were performed and the leachates were subjected to chemical analysis. It is found that he leaching rate of each nuclides obtained in this study were similar to that of the leaching results simulating the underground water.
Inoue, Masaki; Suto, Mitsuo; Koyama, Shinichi; Otsuka, Satoshi; Kaito, Takeji
JAEA-Research 2013-009, 78 Pages, 2013/10
JAEA-Research-2013-009.pdf:3.75MB
In order to exammine the applicability for advanced aqueous reprocessing system, the martensitic oxide dispersion strengthened ferritic steel (9Cr-ODS steel), which is the primary candidate material for high burnup fuel pin cladding tube in fast reactor cycle, was evaluated for the amount of corrosion products in the dissolution process. The quantity of corrosion products was calculated to investigate the influence of both various chemical processes and waste glass (vitrified high level radioactive wastes) by use of the results of a maximum cladding temperature fuel subassembly and the sum of all fuel subassemblies, respectively. The experimental results of immersion tests in flowing liquid sodium loops and fuel pin irradiation tests in fast reactors were reviewed to consider the effect of outer and inner corrosions in high burnup fuel pins on corrosion products. This work revealed that the sum of corrosion products depends largely on the mass transfer behavior in flowing liquid sodium.
Koyama, Shinichi; Suto, Mitsuo; Obayashi, Hiroshi; Oaki, Hiroshi*; Takeshita, Kenji*
Proceedings of International Nuclear Fuel Cycle Conference; Nuclear Energy at a Crossroads (GLOBAL 2013) (CD-ROM), p.549 - 553, 2013/09
A series of separation experiments was performed in order to study the recovery process for minor actinides (MAs) from the actual spent fuel by using an extraction chromatographic technique. The 10 mol% TPPEN-gel was used to improve adsorption coefficient of Am and a condition of eluent temperature was changed in order to confirm the temperature swing effect on TPEN-gel for MA. More than 90% Eu was detected in the eluent after washing with 0.01M NaNO
(pH 3.5) at 5
C. Americium was backwardly detected and eluted continuously during in the same condition. After removal of Eu, the eluent temperature was changed to 32C, then Am was detected (pH 3.0). Finally remained Am could be stripped from TPPEN-gel by changing the pH of the eluent to 2.0.
Sato, Isamu; Suto, Mitsuo; Miwa, Shuhei; Hirosawa, Takashi; Koyama, Shinichi
Journal of Nuclear Materials, 437(1-3), p.275 - 281, 2013/06
Times Cited Count：5 Percentile：38.62(Materials Science, Multidisciplinary)To obtain the source term data in severe accidents for advanced reactors, americium and plutonium release behaviors were evaluated with thermochemical consideration for release kinetics and adhere mechanism.
Koyama, Shinichi; Suto, Mitsuo; Obayashi, Hiroshi; Takeshita, Kenji*; Ogata, Takeshi*; Oaki, Hiroshi*; Inaba, Yusuke*
Separation Science and Technology, 47(14-15), p.2024 - 2028, 2012/11
Times Cited Count：3 Percentile：15.89(Chemistry, Multidisciplinary)A series of separation experiments was performed to study the recovery process for minor actinides (MAs) from the actual spent fuel by using an extraction chromatographic technique. TPPEN gel was used as a new extraction chromatographic agent. Mixed oxide fuel was used as a reference spent fuel to demonstrate the recovery of the MAs. The MOX fuel, including 29.9 wt% plutonium (Pu), was irradiated up to 119 GWd/MTM, and the fuel was then prepared for the extraction experiment. A Mixed solution of MAs and lanthanides (Lns) was prepared. The TPPEN gel was immersed in a 0.01 M NaNO solution, and the pH was adjusted to 4.0. Next, an extraction column was prepared using the TPPEN gel, and the mixed solution of MAs and Lns was passed through the column. The Lns were detected in the eluent after washing with 0.01 M NaNO (pH 4.0). For detecting the MAs, the pH of the eluent was changed to 2.0.
Inoue, Masaki; Ikeuchi, Hirotomo; Takeuchi, Masayuki; Koyama, Shinichi; Suto, Mitsuo
JAEA-Research 2011-057, 100 Pages, 2012/03
JAEA-Research-2011-057.pdf:3.23MB
Corrosion resistance of fuel pin cladding tube materials is one of the most important properties to design aqueous reprocessing process. The martensitic oxide dispersion strengthened ferritic steel, names as "9Cr-ODS" steel, is the primary candidate of high burnup fuel pin cladding tube for fast reactor cycle. Because 9Cr-ODS steel contains lower chromium than stainless steels, oxidizing species in nitric acid medium needs to reduce its corrosion rate. In spent fuel dissolvers, although both nitric acid and metallic ions concentrations change, corrosion potential of 9Cr-ODS steel tends to increase gradually and stabilize protective passive layer effectively.
Am sample irradiated in Japan Material Testing ReactorKoyama, Shinichi; Tanaka, Kenya; Mitsugashira, Toshiaki; Sato, Isamu*; Hara, Mitsuo*; Suto, Mitsuo*; Hanami, Akira*
no journal, ,
Radiochemical analysis of Am sample irradiated in JMTR was performed. After mutual separation of Am, Cm and Pu from the irradiated Am sample, the isotopic composition was analyzed by mass spectroscopy.
Am samplesKoyama, Shinichi; Sudo, Mitsuo; Tanaka, Kenya; Mitsugashira, Toshiaki*; Hanami, Akira*
no journal, ,
Analysis of Am samples irradiated at core center and reflector region in the experimental fast reactor JOYO were carried out. After removal of Am, Cm and Pu from irradiated Am samples, transmutation behavior was evaluated based on the measurement of isotpic composition and the ratio of these elements.
Sudo, Mitsuo; Obayashi, Hiroshi; Koyama, Shinichi; Sekine, Takashi; Tanaka, Kenya
no journal, ,
no abstracts in English
Sudo, Mitsuo; Koyama, Shinichi; Obayashi, Hiroshi; Meiliza, Y.*; Sagara, Hiroshi*; Saito, Masaki*
no journal, ,
Transmutation behavior of plutonium in uranium sample irradiated in experimental fast reactor Joyo was evaluated based on radiochemical analysis in order to develop a blanket fuel with high proliferatin resistance for fast breeder reactor.
Onishi, Takashi; Sudo, Mitsuo; Obayashi, Hiroshi; Koyama, Shinichi; Tanaka, Kenya; Meiliza, Y.*; Yamamoto, Tetsuro*; Sagara, Hiroshi*; Saito, Masaki*
no journal, ,
no abstracts in English
Sudo, Mitsuo; Obayashi, Hiroshi; Koyama, Shinichi; Maeda, Shigetaka; Tanaka, Kenya
no journal, ,
Based on a combustion burnup measurement (a Nd monitor method) result by the chemical analysis, I evaluated an irradiation condition to evaluate an irradiation condition of Am - plutonium-uranium mixed oxide fuel irradiated (24h) in "Joyo".
