Takasaki, Koji; Yasumune, Takashi; Hashimoto, Makoto; Maeda, Koji; Kato, Masato; Yoshizawa, Michio; Momose, Takumaro
JAEA-Review 2019-003, 48 Pages, 2019/03
June 6, 2017, at Plutonium Fuel Research Facility in Oarai Research and Development Center of JAEA, when five workers were inspecting storage containers containing plutonium and uranium, resin bags in a storage container ruptured, and radioactive dust spread. Though they were wearing a half face mask respirator, they inhaled radioactive materials. In the evaluation of the internal exposure dose, the aerodynamic radioactive median diameter (AMAD) is an important parameter. We measured 14 smear samples and a dust filter paper with imaging plates, and estimated the AMAD by image analysis. As a result of estimating the AMAD, from the 14 smear samples, the AMADs are 4.3 to 11 m or more in the case of nitrate plutonium, and the AMADs are 5.6 to 14 m or more in the case of the oxidized plutonium. Also, from the dust filter paper, the AMAD is 3.0 m or more in the case of nitrate plutonium, and the AMAD is 3.9 m or more in the case of the oxidized plutonium.
Sasaki, Shinji; Maeda, Koji; Furuya, Hirotaka*
Journal of Nuclear Science and Technology, 55(3), p.276 - 282, 2018/03
Ishimi, Akihiro; Katsuyama, Kozo; Maeda, Koji; Furuya, Hirotaka*
Journal of Nuclear Science and Technology, 54(11), p.1274 - 1276, 2017/11
Two- and three-dimensional images were obtained in the reaction product between zircaloy and MOX fuel by X-ray CT. In addition, the -ray intensity distributions of two fission products (Cs-137 and Eu-154) and one neutron-activated nuclide (Co-60) were obtained in this specimen by -ray measurements. The average values of the fuel density (about 10.5 g/cm) and the cladding density (about 6.55 g/cm) were obtained in the metallic phase region by evaluation of the density distributions on two-dimensional X-ray CT images. In addition, the distributions of the roughly crushed fuel pellet and the pores in the specimen could be clearly observed on the three-dimensional X-ray CT images. From the -ray measurement, Cs-137 was observed on the unreacted fuel region and the region where pores exist in the metallic phase, and Eu-154 was widely distributed to all regions. On the other hand, Co-60 was confirmed only in the metallic phase region.
Ikusawa, Yoshihisa; Maeda, Koji; Kato, Masato; Uno, Masayoshi*
Nuclear Technology, 199(1), p.83 - 95, 2017/07
Based on thermal computation results obtained using an irradiation behavior analysis code, we have evaluated the effect of O/M ratio on fuel restructuring from the results of PIEs for the B14 irradiation test fuel, which was a mixed oxide fuel and was irradiated in the experimental reactor Joyo. The thermal computation results showed that fuel restructuring in the stoichiometric oxide fuel was accelerated, though the fuel temperature in the stoichiometric oxide fuel was evaluated as lower than that of the hypo-stoichiometric one. We explained this behavior as follows: first, the fuel temperature decreased due to the high thermal conductivity at stoichiometry; second, the pore migration velocity increased due to the increase in vapor pressure caused by the high vapor pressure of UO, which was derived from the high oxygen potential at stoichiometry. In addition, our results indicated that the central void diameter strongly depended on not only fuel temperature, but also vapor pressure.
Sasaki, Shinji; Tanno, Takashi; Maeda, Koji
Proceedings of 54th Annual Meeting of Hot Laboratories and Remote Handling (HOTLAB 2017) (Internet), 6 Pages, 2017/00
During irradiation in a fast reactor, the microstructure change of the mixed oxide fuels and the changes of element distributions occur because of a radial temperature gradient. Therefore, it is important to study the irradiation behavior of MA-MOX for advancement of fast reactor fuels. In order to make detailed observations of microstructure and elemental analyses of MA-MOX, irradiated MA-MOX specimens were carried out PIE by using a FE-SEM equipped with WDX. Because fuel samples have high radio activities and emit alpha-particles, the instrument was modified. the instrument was installed in a lead shield box and the control unit was separately located outside the box. The microstructure changes were observed in irradiated MA-MOX specimen. The characteristic X-rays peaks were detected successfully. By measuring the intensities of characteristic X-rays, it was tried quantitative analysis of U, Pu, Am along radial direction of irradiated specimen.
Shizukawa, Yuta; Sekio, Yoshihiro; Sato, Isamu*; Maeda, Koji
Proceedings of 54th Annual Meeting of Hot Laboratories and Remote Handling (HOTLAB 2017) (Internet), 5 Pages, 2017/00
Electrochemical corrosion behavior under salt water in a type 304L stainless steel used to a part of BWR core materials was investigated to evaluate the possibility of crevice corrosion occurrence for the fuel assemblies which experienced seawater exposure in Fukushima Daiichi Nuclear Power Plant (1F) accident. Especially, focusing on the upper end plug part having the 304L SS crevice structure, measurement of repassivation potential for crevice corrosion () were carried out using the crevice test pieces fabricated by 304L SS plates. From the results, was lower than the spontaneous potential () when the conditions of 2500 ppm chloride ion concentration at over 50 C or that of 2500 ppm at over 80 C, which are included in the SFP water quality conditions. Therefore, in the 304L SS parts of the 1F fuel assemblies that experienced seawater exposure, there is a possibility of crevice corrosion occurrence.
Isozaki, Misaki; Sasaki, Shinji; Maeda, Koji; Katsuyama, Kozo
JAEA-Technology 2015-058, 28 Pages, 2016/03
During irradiation in the fast reactor "JOYO", the changes of fuel structures with the formation of central void occur in the uranium-plutonium mixed oxide fuels (MOX fuels) because of radial temperature gradient. The changes of element (U, Pu, and so on) distributions along radial direction proceed from these changes. Therefore, it is important to study the changes of fuel structures of the minute area in fuel pellet and the changes of element distribution behavior for development of fast reactor fuels. In order to make detailed observations of microstructure and elemental analyses of fuel samples, a field emission scanning electron microscope (FE-SEM) equipped with a wavelength-dispersive X-ray spectrometer (WDS) and an energy-dispersive X-ray spectrometer (EDS) were installed in Fuel Monitoring Facility (FMF). The samples of this FE-SEM are very high radioactivity because the samples contain the nuclear fuel elements (U, Pu, etc.), the fission products (Cs, Rh, etc.) and activation product (Co, Mn etc.). Owing to this, it is necessary to prevent leakage of radioactive materials (particularly, U, Pu is need tight accountancy in law) and to protect operators from radiation. In this installation of FE-SEM, it is selected JSM-7001F (made by JEOL) for base model. The notable modified points were as follows. (1) To protect operators from radiation, lead shields was installed around FE-SEM. (2) To prevent leakage of radioactive materials, the instrument was attached to a remote control air-tight sample transfer unit between a shielded hot cell and the FE-SEM and the instrument was fixing rigid structure without vibration damper. (3) The design and manufacture the lead shields with consideration of instrument maintainability. This paper was described the summary of FE-SEM, the notable modified points, the ways of FE-SEM installation, the result of performance test.
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
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.
Sekio, Yoshihiro; Yamagata, Ichiro; Yamashita, Shinichiro; Inoue, Masaki; Maeda, Koji
Proceedings of 2014 Nuclear Plant Chemistry Conference (NPC 2014) (USB Flash Drive), 10 Pages, 2014/10
In the Fukushima Dai-ichi Nuclear Power Plant accident, seawater was temporarily injected into the spent fuel pools since water cooling and feeding functions were lost. For fuel assemblies which experienced seawater immersion, surface corrosion due to seawater constituents and the resultant degradation of mechanical property are of concern. Therefore, in order to assess the integrity of fuel assemblies (especially cladding tubes), the effects of seawater immersion on corrosion behavior and mechanical properties for as-recieved and irradiated Zircaloy-2 cladding tubes were investigated in the present study. As a result, no obvious surface corrosion and no significant degradation in the tensile strength property were observed after both artificial and natural seawater immersion tests for both steels. This suggests that the effects of seawater immersions on corrosion behavior and mechanical property (especially tensile property) for Zircaloy-2 cladding tubes are probably negligible.
Kato, Masato; Hirooka, Shun; Ikusawa, Yoshihisa; Takeuchi, Kentaro; Akashi, Masatoshi; Maeda, Koji; Watanabe, Masashi; Komeno, Akira; Morimoto, Kyoichi
Proceedings of 19th Pacific Basin Nuclear Conference (PBNC 2014) (USB Flash Drive), 12 Pages, 2014/08
Uranium and plutonium mixed oxide (MOX) fuel has been developed for Japan sodium-cooled fast reactors. Science based fuel technologies have been developed to analyse behaviours of MOX pellets in the sintering process and irradiation conditions. The technologies can provide appropriate sintering conditions, irradiation behaviour analysis results and so on using mechanistic models which are derived based on theoretical equations to represent various properties.
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
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.
Ikusawa, Yoshihisa; Ozawa, Takayuki; Hirooka, Shun; Maeda, Koji; Kato, Masato; Maeda, Seiichiro
Proceedings of 22nd International Conference on Nuclear Engineering (ICONE-22) (DVD-ROM), 6 Pages, 2014/07
In order to develop MA contained MOX (MA-MOX) fuel design method, the analysis models to predict irradiation behavior of MA-MOX fuel have to be developed and the accuracy of irradiation behavior analysis code should be evaluated with the result of post-irradiation examinations (PIEs) for MA-MOX fuels. In this study, we developed the computer module "TRANSIT" to compute thermal properties of MA-MOX fuel. TRANSIT can give thermal conductivity, melting temperature and vapor pressures of MA-MOX. By using this module, we improved the thermal behavior analysis code "DIRAD" and developed DIRAD-TRANSIT code system to compute the irradiation behavior of MA-MOX fuel. This system was verified with the results of PIEs for the conventional MOX fuels and the MA-MOX fuels irradiated in the experimental fast reactor "JOYO". As the result of the verification, it can be mentioned that the DIRAD-TRANSIT system would precisely predict the fuel thermal behavior, i.e. fuel temperature and fuel restructuring, for oxide fuels containing several percent minor actinides.
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
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.
Sakaki, Hironao; Nishiuchi, Mamiko; Maeda, Shota; Sagisaka, Akito; Pirozhkov, A. S.; Pikuz, T.; Faenov, A.*; Ogura, Koichi; Fukami, Tomoyo; Matsukawa, Kenya*; et al.
Review of Scientific Instruments, 85(2), p.02A705_1 - 02A705_4, 2014/02
High intensity laser-plasma interaction has attracted considerable interest for a number of years. The laser-plasma interaction is accompanied by generation of various charged particle beams. Results of simultaneous novel measurements of electron-induced photonuclear neutrons (photoneutron), which are a diagnostic of the laser-plasma interaction, are proposed to use for optimization of the laser-plasma ion generation. The proposed method is demonstrated by the laser irradiation with the intensity os 110 W/cm on the metal foil target. The photoneutrons are measured by using NE213 liquid scintillation detectors. Heavy-ion signal is registered with the CR39 track detector simultaneously. The measured signals of the electron-induced photoneutrons are well reproduced by using the Particle and Heavy Ion Transport code System (PHITS). The results obtained provide useful approach for analyzing the various laser based ion beams.
Takasaki, Koji; Yasumune, Takashi; Onishi, Takashi; Nakamura, Keisuke; Ishimi, Akihiro; Ito, Chikara; Osaka, Masahiko; Ono, Masashi*; Hatakeyama, Shuichi*; Takahashi, Hiroyuki*; et al.
JAEA-Research 2013-043, 33 Pages, 2014/01
In the Fukushima Daiichi Nuclear Power Plant, it is assumed that the core fuels melted partially or wholly, and the normal technique of accounting for a fuel assembly is not applicable. Therefore, it is necessary to develop the transparent and rational technique of accounting in the process of collection and storage of fuel debris. In this research, an application of the superconducting phase Transition Edge Sensor microcalorimeter (TES microcalorimeter) is studied for the accounting of nuclear materials in the fuel debris. It is expected that the detailed information of nuclear materials and fission products in fuel debris is obtained by using a high-resolution characteristic of TES microcalorimeter. In this report, the principle of TES microcalorimeter, the measurement experiment using TES in JAEA, and the simulated calculation using the EGS5 code system are summarized.
Maeda, Koji; Sasaki, Shinji; Kumai, Misaki; Sato, Isamu; Suto, Mitsuo; Osaka, Masahiko
JAEA-Research 2013-025, 123 Pages, 2014/01
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.
Yamagata, Ichiro; Hayashi, Takehiro; Mashiko, Shinichi*; Sasaki, Shinji; Inoue, Masaki; Yamashita, Shinichiro; Maeda, Koji
JAEA-Testing 2013-004, 23 Pages, 2013/11
In the accident of the Fukushima Daiichi Nuclear Power Plant of Tokyo Electric Power Co. accompanying the Great East Japan Earthquake, fuel assemblies kept in the spent fuel pool of reactor units 1-4, were exposed to the inconceivable environment such as falling and mixing of rubble, especially seawater were injected into unit 2-4. In order to evaluate the integrity of the fuel assemblies in spent fuel pools, and in the long-term storage after transported to the common storage pool, the immersion tests were performed using zircaloy-2 fuel cladding tubes irradiated in the advanced thermal reactor Fugen. The immersion liquid was prepared with doubling dilution of artificial seawater, which temperature was 80 C and immersion time was about 336 hours, as assuming the situation of the pool. The results indicated zircaloy-2 cladding tubes had no significant corrosion and no influence on mechanical property by immersion tests with artificial seawater conditions of this work.
Maeda, Koji; Sasaki, Shinji; Kumai, Misaki; Sato, Isamu; Osaka, Masahiko; Fukushima, Mineo; Kawatsuma, Shinji; Goto, Tetsuo*; Sakai, Hitoshi*; Chigira, Takayuki*; et al.
Proceedings of International Nuclear Fuel Cycle Conference; Nuclear Energy at a Crossroads (GLOBAL 2013) (CD-ROM), p.272 - 277, 2013/09
Ishimi, Akihiro; Katsuyama, Kozo; Maeda, Koji; Nagamine, Tsuyoshi; Asaka, Takeo; Furuya, Hirotaka
Journal of Nuclear Science and Technology, 49(12), p.1144 - 1155, 2012/12
no abstracts in English
Satoh, Daiki; Maeda, Yoshikazu*; Tameshige, Yuji*; Nakashima, Hiroshi; Shibata, Tokushi*; Endo, Akira; Tsuda, Shuichi; Sasaki, Makoto*; Maekawa, Motokazu*; Shimizu, Yasuhiro*; et al.
Journal of Nuclear Science and Technology, 49(11), p.1097 - 1109, 2012/11
At the Fukui Prefectural Hospital Proton Therapy Center, neutron doses behind concrete shields and at maze have been measured by using radiation monitors, DARWIN, Wendi-2, a rem meter, and solid state nuclear track detectors. The measured data were compared with the estimations by analytical models and Monte Carlo code PHITS. The analytical model with the parameters employed in shielding design of the facility gave considerably larger estimates than the measured data. This means that the facility was designed with an enough safety margin. The calculation results of PHITS were less than those of the analytical model, and were about 3 times larger than the measured data. From the view point of a safety policy with conservative estimation for shielding design, Monte Carlo simulation is a better tool for estimating radiation safety at accelerator-based proton treatment facilities.