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Koyama, Shinichi; Ikeuchi, Hirotomo; Mitsugi, Takeshi; Maeda, Koji; Sasaki, Shinji; Onishi, Takashi; Tsai, T.-H.; Takano, Masahide; Fukaya, Hiroyuki; Nakamura, Satoshi; et al.
Hairo, Osensui, Shorisui Taisaku Jigyo Jimukyoku Homu Peji (Internet), 216 Pages, 2023/11
In FY 2021 and 2022, JAEA perfomed 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, Thermal Bahavior Estimation, and Simplified Analysis of Fuel Debris)" started in FY 2021. This presentation material summarized the results of the project, which will be available shortly on the website of Management Office for the Project of Decommissioning, Contaminated Water and Treated Water Management.
Yamashita, Takuya; Shimomura, Kenta; Nagae, Yuji; Yamaji, Akifumi*; Mizokami, Shinya; Mitsugi, Takeshi; Koyama, Shinichi
Hairo, Osensui, Shorisui Taisaku Jigyo Jimukyoku Homu Peji (Internet), 53 Pages, 2023/10
JAEA performed the subsidy program for the "Project of Decommissioning, Contaminated Water and Treated Water Management (Development of Analysis and Estimation Technologies for Characterization of Fuel Debris (Development of Estimation Technologies of RPV Damaged Condition, etc.) in 2022JFY. 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, Contaminated Water and Treated Water Management.
Tsubota, Yoichi; Porcheron, E.*; Journeau, C.*; Delacroix, J.*; Suteau, C.*; Lallot, Y.*; Bouland, A.*; Roulet, D.*; Mitsugi, Takeshi
Proceedings of International Conference on Environmental Remediation and Radioactive Waste Management (ICEM2023) (Internet), 6 Pages, 2023/10
In order to safely remove fuel debris from the Fukushima Daiichi Nuclear Power Station (1F), it is necessary to quantitatively evaluate radioactive airborne particulate generated by the cutting of nuclear fuel debris. We fabricated Uranium-bearing simulated fuel debris (SFD) with In/Ex-Vessel compositions and evaluated the physical and chemical properties of aerosols generated by heating the SFDs. Based on these results, we estimated the isotopic composition and radioactivity of aerosols produced when 1F-Unit2 fuel debris is laser cut, which is a typical example of a heating method. Plutonium, mainly 
Pu,
Am, and 
Cm were found to be the alpha nuclide, and Pu, 
Cs-Ba, and 
Sr-Y were found to be the beta nuclide of interest.
Porcheron, E.*; Journeau, C.*; Delacroix, J.*; Berlemont, R.*; Bouland, A.*; Lallot, Y.*; Tsubota, Yoichi; Ikeda, Atsushi; Mitsugi, Takeshi
Proceedings of International Conference on Environmental Remediation and Radioactive Waste Management (ICEM2023) (Internet), 5 Pages, 2023/10
Results of the URASOL project aimed at evaluating the generation and dispersion of radioactive aerosols during the cutting of fuel debris, a key issue in the decommissioning of the damaged reactors at the Fukushima Daiichi Nuclear Power Station (1F), are presented in this report. Characterization of aerosols generated during heating and mechanical cutting of simulated fuel debris in terms of mass concentration, real-time number density, mass-based particle size distribution, morphology, and chemical properties is reported. In the heating tests, an increase in particle size with increasing temperature was observed, and in terms of particle number density, the case using depleted uranium simulated fuel debris had a smaller number density than the test using Hf-containing simulated fuel debris. In mechanical cleavage, the aerodynamic median mass diameter of the aerosol was almost the same for the radioactive and non-radioactive samples (about 3.7
4.4 
m).
Nakayoshi, Akira; Mitsugi, Takeshi; Sasaki, Shinji; Maeda, Koji
JAEA-Data/Code 2021-011, 279 Pages, 2022/03
JAEA-Data-Code-2021-011.pdf:37.76MB
At the TEPCO's Fukushima Daiichi Nuclear Power Station (1F), an investigation inside the reactors has been carried out, and R&D has been made on methods of fuel debris retrieval and storage after retrieval. In order to carry out the decommissioning work safely and steadily, understanding characteristics of fuel debris in the reactors is required. Therefore, in the development of technologies for grasping and analyzing properties of fuel debris project, the characteristics of simulated fuel debris, such as hardness, drying behavior, etc., of fuel debris for design of removal and storage, have been investigated and estimated, and provided to other projects conducting the decommissioning work. As part of this project, U-containing particles in samples (e.g., deposit on the investigation equipment, sediment in the reactors, etc.) obtained during the internal investigation of the reactors of 1F units 1 to 3 were analyzed. This report summarized the results of FE-SEM/WDX, FE-SEM/EDS, STEM/EDS, and TEM analysis, which were extracted from all analysis results obtained, as a database for the evaluation of the generation mechanism of U-containing particles. The analyses were performed at the JAEA Oarai Research and Development Institute and Nippon Nuclear Fuel Development Co., LTD.
Washiya, Tadahiro; Koyama, Shinichi; Takano, Masahide; Mitsugi, Takeshi
Denki Hyoron, 105(9), p.64 - 71, 2020/09
For the retrieval of fuel debris in the 1F decommissioning, a retrieval tool and a retrieval method according to the characteristics of fuel debris are being studied. In addition, for stable storage, treatment, and disposal after retrieval, it is necessary to fully understand the characteristics and chemical stability of fuel debris and select appropriate measures. In this paper, we will introduce the characteristics of fuel debris that have been discovered in the previous studies and the problems in handling them.
Ito, Kazuhiro; Omura, Akiko; Hoshiya, Taiji; Mitsugi, Takeshi
Proceedings P.29-32, p.29 - 32, 2004/10
The Joyo MK-III rated power operation was started in 2003 to do various irradiation test more efficiently. Promotion of the outside use is being advanced. As the part and consideration is being advanced about formation of irradiation equipment of low flux neutron field near the space. By this report, this, JOYO as a low neutron irradiation field is introduced.
Takamatsu, Misao; Kono, Naomi; Amezawa, Takayuki; Mitsugi, Takeshi; Maeda, Yukimoto
Saikuru Kiko Giho, (22), p.80 - 82, 2004/00
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.
Mitsugi, Takeshi; Odo, Toshihiro; Abe, Tomoyuki; Isozaki, Kazunori; Maeda, Koji; Ito, Chikara
Genshiryoku eye, 49(9), 1 Pages, 2003/08
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.
Miyakawa, Shunichi; Yanagisawa, Tsutomu; ; ; Mitsugi, Takeshi
P247, 247 Pages, 2002/00
None
Soga, Tomonori; Tobita, Koichi; Mitsugi, Takeshi; Miyakawa, Shunichi
Saikuru Kiko Giho, (8), p.13 - 22, 2000/09
None
Okamoto, Kaoru; Miyakawa, Shunichi; Mitsugi, Takeshi; Kitamura, Ryoichi
JNC TN9410 99-010, 350 Pages, 1999/06
In the needs of the fuel irradiation test in "Joyo" MK-III core, there have been required that the irradiation of high performance fuel at high liner heat rate to high burn-up range, or the irradiation of advanced fuel such as MA fuel and Vipac fuel. In order to carry out these irradiation tests, newly designed irradiation subassembly is required with special features of; (1)Capability of the re-assembling after post-irradiation examination, even if the number of fuel in the identical irradiation condition decreases because of intermediate inspection. (2)Enhanced flexibility of the irradiation temperature setting ( in the present, UNIS-B's has 6 cases on the maximum). (3)Sufficient flexibility for the coolant flow distribution in the subassembly by extending variety of the flow rate setting. UNIS-D is a fuel irradiation subassembly which has been developed from above viewpoints. It is a compartment loading type irradiation subassembly that is able to load maximum of 18 compartments. Two types of compartments 
-type and 
-typc arc prepared for UNIS-D. Thc sufficient consideration has also been made on the rc-assembling. A -type is the same compartment as the existing UNIS-B's and a -type is the newly designed one for UNIS-D. Three to five fuel pins are loaded into a -type compartment and only one pin is loadcd into a -type compartment. It is possible to carry out the irradiation test in a maximum of 18 test temperature conditions within a subassembly, since it has the sufficient flexibility for the coolant flow distribution. As for the development of UNIS-D, we have finished the detailed structure design and the design verification by the water flow examination, which confirmed that the UNIS-D exceeded its required performances in use and that its structure design was satisfactory.
Soga, Tomonori; Miyakawa, Shunichi; Mitsugi, Takeshi
JNC TN9400 99-052, 355 Pages, 1999/06
Currently, the lifetime of control rods in JOYO is limited by Absorber-Cladding Mechanical Interaction (ACMI) due to swelling of B
C(boron carbide) pellets accelerated by relocation of pellet fragments. A sodium bonded type control rod was developed which improves the thermal conductivity by means of charging sodium into the gap between BC and cladding and by utilizing a shroud which wraps the pellet fragments in a thin tube. This new design will be able to enlarge the gap between BC and cladding, without heating BC or fragment relocation, thus extending the life of the control rod. The sodium bonded type will be fabricated as the ninth reload control rods in JOYO. (1)The specification of a sodium bonded type control rod was determined with the wide gap between BC and cladding. In the design simulation, main component temperature were below the maximum limit. And the local heating by helium bubble generated from BC in the sodium gap, was not a serious problem in the analysis which was considered. (2)A structural design for the sodium entrance into the pin was determined. A formula was developed which the limit for sodium charging given physical dimension of the structure and sodium property. Result from sodium out-pile experiments validated the theoretical formula. (3)The analysis of ACMI indicated a lifetime extension of the sodium bonded type by 4.6% in comparison with lifetime of the helium bonded type of 1.6%. This is due to the boron10 burn-up rate being three times higher in the sodium bonded type than in the helium bonded type. To achieve a target burn-up 10% in the future, it will be necessary to modify design based on irradiation data which will be obtained by practical use of the sodium bonded control rods in JOYO. (4)The effects due to Absorber-Cladding Chemical Interaction (ACCI) were reduced by controlling the cladding temperature and chromium coating to the cladding's inner surface. It was confirmed that ...
Maeda, Koji; Nagamine, Tsuyoshi; ; Mitsugi, Takeshi; Matsumoto, Shinichiro
Dai-3-Kai Shoshago Shiken Ni Kansuru Nikkan Semina, 0 Pages, 1999/00
None
Katsuyama, Kozo; Mitsugi, Takeshi; Asaka, Takeo
PNC TN9410 98-047, 152 Pages, 1998/04
Internal rod pressure of a FBR fuel pin is one of the important factors which restrict the lifetime of a fuel pin, because fission gases (krypton, xenon and helium) are released from fuel pellets to increase internal rod pressure with increasing burn-up. Due to their relatively higher fission yield, the release behavior of krypton and xenon have already been studied well. However, there are few studies on helium release behavior. It is supposed that substantial amount of helium gas is produced in high burn-up (150Gwd/t) MOX fuels and the fuels containing minor actinides. Therefore we established the quantitative measurement method of helium gas released from FBR fuels, and the helium gas release behavior was evaluated. The results are shown as follows. (1) Quantitative Measurement of helium gas. In order to measure the volume of helium gas, the carrier gas of gas-chromatograph was changed from helium gas to argon gas. Consequently, a direct measurement of helium gas in the irradiated fuel pins became possible. The comparison between present method and the previous one made clear that the data obtained in the previous method were reliable. (2) Calculation of helium generation. Helium generation was calculated by considering 
decay, ternary fission of heavy nuclides elements, and (n, ) reaction of light elements in the MOX fuel. The calculation showed that the amount of helium production from (n, ) reaction is influenced by the neutron spectrum, and increased with increasing burn-up. The amount of helium from decay is influenced by the amount of trans-uranium elements such as Am-241, and by the length of cooling time after reactor shut down. (3) Helium gas release behavior. The amount of helium gas released from fuel pellets is increased with burn-up, and the helium gas release fraction varies from 50% to 100%. As compared to the case of fission gas release, the fraction of helium gas release is higher. In the case of ...
Uematsu, Shinichi; Mitsugi, Takeshi; ; Kobayashi, Tetsuro*; Yokoya, Jun*
Nihon Genshiryoku Gakkai-Shi, 39(10), p.870 - 880, 1997/10
Times Cited Count：1 Percentile：14.31(Nuclear Science & Technology)None
Mitsugi, Takeshi; ;
Proceedings of International Topical Meeting on Light Water Reactor Fuel Performance, p.54 - 61, 1997/00
None
Ichikawa, Michio*; ; ; ; ; Mitsugi, Takeshi; ; Ito, Kunio*; ; Doi, Soichi*; et al.
Nihon Genshiryoku Gakkai-Shi, 39(2), p.93 - 111, 1996/00
None
