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Tobita, Minoru*; Goto, Katsunori*; Omori, Takeshi*; Osone, Osamu*; Haraga, Tomoko; Aono, Ryuji; Konda, Miki; Tsuchida, Daiki; Mitsukai, Akina; Ishimori, Kenichiro
JAEA-Data/Code 2023-011, 32 Pages, 2023/11
JAEA-Data-Code-2023-011.pdf:0.93MB
Radioactive wastes generated from nuclear research facilities in Japan Atomic Energy Agency are planning to be buried in the near surface disposal field as trench and pit. Therefore, it is required to establish the method to evaluate the radioactivity concentrations of radioactive wastes until the beginning of disposal. In order to contribute to the study of radioactivity concentration evaluation methods for radioactive wastes generated from nuclear research facilities, we collected and analyzed concrete samples generated from JRR-3, JRR-4 and JAERI Reprocessing Test Facility. In this report, we summarized the radioactivity concentrations of 23 radionuclides (
H, 
C, 
Cl, 
Ca, 
Co, 
Ni, 
Sr, 
Nb, 
Ag, 
Cs, 
Ba, 
Eu, 
Eu, 
Ho, 
U, 
U, 
U, Pu, 
Pu, 
Pu, 
Am, 
Am, 
Cm) which were obtained from radiochemical analysis of the samples in fiscal years 2021-2022.
Takabe, Yugo; Otsuka, Noriaki; Fuyushima, Takumi; Sayato, Natsuki; Inoue, Shuichi; Morita, Hisashi; Jaroszewicz, J.*; Migdal, M.*; Onuma, Yuichi; Tobita, Masahiro*; et al.
JAEA-Technology 2022-040, 45 Pages, 2023/03
JAEA-Technology-2022-040.pdf:6.61MB
Because of the decommission of the Japan Materials Testing Reactor (JMTR), the domestic neutron irradiation facility, which had played a central role in the development of innovative nuclear reactors and the development of technologies to further improve the safety, reliability, and efficiency of light water reactors, was lost. Therefore, it has become difficult to pass on the operation techniques of the irradiation test reactors and irradiation technologies, and to train human resources. In order to cope with these issues, we conducted a study on the implementation of irradiation tests using overseas reactors as neutron irradiation sites as an alternative method. Based on the "Arrangement between the National Centre for Nuclear Research and the Japan Atomic Energy Agency for Cooperation in Research and Development on Testing Reactor," the feasibility of conducting an irradiation test at the MARIA reactor (30 MW) owned by the National Centre for Nuclear Research (NCBJ) using the temperature control system, which is one of the JMTR irradiation technologies, was examined. As a result, it was found that the irradiation test was possible by modifying the ready-made capsule manufactured in accordance with the design and manufacturing standards of the JMTR. After the modification, a penetration test, an insulation continuity test, and an operation test in the range of room temperature to 300
C, which is the operating temperature of the capsule, were conducted and favorable results were obtained. We have completed the preparations prior to transport to the MARIA reactor.
Tobita, Minoru*; Konda, Miki; Omori, Takeshi*; Nabatame, Tsutomu*; Onizawa, Takashi*; Kurosawa, Katsuaki*; Haraga, Tomoko; Aono, Ryuji; Mitsukai, Akina; Tsuchida, Daiki; et al.
JAEA-Data/Code 2022-007, 40 Pages, 2022/11
JAEA-Data-Code-2022-007.pdf:1.99MB
Radioactive wastes generated from nuclear research facilities in Japan Atomic Energy Agency are planning to be buried in the near surface disposal field. Therefore, it is required to establish the method to evaluate the radioactivity concentrations of radioactive wastes until the beginning of disposal. In order to contribute to this work, we collected and analyzed concrete, ash, ceramic and brick samples generated from JRR-3, JRR4 and JRTF facilities. In this report, we summarized the radioactivity concentrations of 24 radionuclides (H, C, Cl, Ca, Co, Ni, Sr, Nb, 
Tc, Ag, 
I, Cs, Ba, Eu, Eu, Ho, U, U, Pu, Pu, Pu, Am, Am, Cm) which were obtained from radiochemical analysis of the samples in fiscal years 2020-2021.
Tobita, Minoru*; Haraga, Tomoko; Endo, Tsubasa*; Omori, Hiroyuki*; Mitsukai, Akina; Aono, Ryuji; Ueno, Takashi; Ishimori, Kenichiro; Kameo, Yutaka
JAEA-Data/Code 2021-013, 30 Pages, 2021/12
JAEA-Data-Code-2021-013.pdf:1.47MB
Radioactive wastes generated from nuclear research facilities in Japan Atomic Energy Agency are planning to be buried in the near surface disposal field. Therefore, it is required to establish the method to evaluate the radioactivity concentrations of radioactive wastes until the beginning of disposal. In order to contribute to this work, we collected and analyzed concrete samples generated from JPDR facility. In this report, we summarized the radioactivity concentrations of 21 radionuclides (H, C, Cl, Ca, Co, Ni, Sr, Nb, Ag, Cs, Eu, Eu, Ho, U, U, Pu, Pu, Pu, Am, Am, Cm) which were obtained from radiochemical analysis of the samples in fiscal year 2018-2019.
Hamamoto, Shimpei; Ishitsuka, Etsuo; Nakagawa, Shigeaki; Goto, Minoru; Matsuura, Hideaki*; Katayama, Kazunari*; Otsuka, Teppei*; Tobita, Kenji*
Proceedings of 2021 International Congress on Advances in Nuclear Power Plants (ICAPP 2021) (USB Flash Drive), 5 Pages, 2021/10
Impurity concentrations of hydrogen and hydride in the coolant were investigated in detail for the HTTR, a block type high-temperature gas reactor owned by Japan. As a result, it was found that CH
was 1/10 of H
concentration, which was under the conventional detection limit. If the ratio of H to CH in the coolant is the same as the ratio of HT to CH
T, the CHT has a larger dose conversion factor, and this compositional ratio is an important finding for the optimal dose evaluation. Further investigation of the origin of CH suggested that CH was produced as a result of a thermal equilibrium reaction rather than being released as an impurity from the core.
Tsuchida, Daiki; Haraga, Tomoko; Tobita, Minoru*; Omori, Hiroyuki*; Omori, Takeshi*; Murakami, Hideaki*; Mitsukai, Akina; Aono, Ryuji; Ishimori, Kenichiro; Kameo, Yutaka
JAEA-Data/Code 2020-022, 34 Pages, 2021/03
JAEA-Data-Code-2020-022.pdf:1.74MB
Radioactive wastes generated from nuclear research facilities in Japan Atomic Energy Agency are planning to be buried in the near surface disposal field. Therefore, it is required to establish the method to evaluate the radioactivity concentrations of radioactive wastes until the beginning of disposal. In order to contribute to this work, we collected and analyzed concrete samples generated from JRR-3 and JPDR. In this report, we summarized the radioactivity concentrations of 22 radionuclides(H, C, Cl, Ca, Co, Ni, Sr, Nb, Ag, Ba, Cs, Eu, Eu, Ho, U, U, Pu, 
Pu, Am, Am, Cm) which were obtained from radiochemical analysis of the samples.
Tobita, Minoru*; Haraga, Tomoko; Sasaki, Takayuki*; Seki, Kotaro*; Omori, Hiroyuki*; Kochiyama, Mami; Shimomura, Yusuke; Ishimori, Kenichiro; Kameo, Yutaka
JAEA-Data/Code 2019-016, 72 Pages, 2020/02
JAEA-Data-Code-2019-016.pdf:2.67MB
In the future, radioactive wastes which generated from research and testing reactors in Japan Atomic Energy Agency are planning to be buried for the near surface disposal. Therefore, it is required to establish the method to evaluate the radioactivity concentrations of radioactive wastes by the time it starts disposal. In order to contribute to this work, we collected and analyzed the samples generated from JRR-2, JRR-3 and Hot laboratory facilities. In this report, we summarized the radioactivity concentrations of 25 radionuclides (H, C, Cl, Co, Ni, Sr, Nb, 
Mo, Tc, 
Ag, 
Sn, I, Cs, Eu, Eu, 
U, U, U, Pu, Pu, Pu, Pu, Am, Am, Cm) which were obtained from radiochemical analysis of those samples.
Haraga, Tomoko; Tobita, Minoru*; Takahashi, Shigemi*; Seki, Kotaro*; Izumo, Sari; Shimomura, Yusuke; Ishimori, Kenichiro; Kameo, Yutaka
JAEA-Data/Code 2016-017, 53 Pages, 2017/02
JAEA-Data-Code-2016-017.pdf:3.17MB
Fugen Nuclear Power Station was shut down and now is under decommissioning. Many radioactivity concentration data of dismantled materials have to be accumulated to calculate the scaling factors of radioactive wastes and to verify that the cleared dismantled materials conform to the clearance levels. A simple and rapid radioactivity determination method for radioactive waste samples was developed by Department of Decommissioning and Waste Management. For its demonstration, the simple and rapid radioactivity determination method was applied to metal samples, which were taken from dismantled pipes in contact with heavy water or carbon dioxide gas of Fugen. This report summarizes the radioactivity data obtained from the analysis of those samples.
Haraga, Tomoko; Tobita, Minoru*; Takahashi, Shigemi*; Ishimori, Kenichiro; Kameo, Yutaka
JAEA-Data/Code 2015-025, 52 Pages, 2016/03
JAEA-Data-Code-2015-025.pdf:1.92MB
Fugen Nuclear Power Station was shut down and now is under decommissioning. Many radioactivity concentration data of dismantled materials have to be accumulated to calculate the scaling factors of radioactive wastes and verify that the cleared dismantled materials conform to the clearance levels. A simple and rapid radioactivity determination method for radioactive waste samples was developed in Department of Decommissioning and Waste Management. For the demonstration, the simple and rapid radioactivity determination method was applied to metal samples, which were taken from dismantled pipes of Fugen. This report summarizes the radioactivity data obtained from the analysis of those samples.
Haraga, Tomoko; Tobita, Minoru*; Takahashi, Shigemi*; Sakatani, Keiichi; Ishimori, Kenichiro; Takahashi, Kuniaki
JAEA-Data/Code 2014-007, 52 Pages, 2014/06
JAEA-Data-Code-2014-007.pdf:28.47MB
Fugen Nuclear Power Station was shut down and now is under decommissioning. Many radioactivity concentration data of dismantled materials have to be accumulated to calculate the scaling factors of radioactive wastes and verify that the cleared dismantled materials conform to the clearance levels. A simple and rapid radioactivity determination method for radioactive waste samples was developed in Nuclear Cycle Backend Directorate. For the demonstration, the simple and rapid radioactivity determination method was applied to metal samples, which were taken from dismantled pipes of Fugen. This report summarizes the radioactivity data obtained from the analysis of those samples.
Toyooka, Junichi; Endo, Hiroshi*; Tobita, Yoshiharu; Takahashi, Minoru*
Nihon Genshiryoku Gakkai Wabun Rombunshi, 13(2), p.35 - 50, 2014/05
In the design of future sodium-cooled fast reactor, a design measure to prevent severe re-criticality events even in case of core disruptive accidents is considered. This design adopts inner duct within the fuel sub-assembly that should allow molten fuel ejection out of the core region. The effectiveness of this design is dependent on failure time of the duct and it depends significantly on heat transfer from the melting core materials to the duct. In the previous study by the authors, heat transfer from molten fuel/steel mixture to the inner duct was evaluated with a computer model simulation for an in-pile experiment performed in IGR (Impulse Graphite Reactor) focusing on demonstration of the design effectiveness. In the present study, possible uncertainties in the assumption and model parameters in the previous study were evaluated so that validity of the main conclusion of the previous study could be confirmed and re-enforced. This confirmation consisted of evaluation of necessary fuel-to-steel heat transfer area, effect of hydrodynamic fragmentation of steel droplets, steel-vapor condensation heat transfer onto the duct surface and fuel crust formation. Furthermore, possible effect of variation in fuel designs and transient scenarios to the heat transfer was evaluated changing steel volume fraction as the initial boundary conditions. It was concluded that the previous study was appropriate in representing the realistic situation and the conclusions in the previous study were enforced. An additional set of analysis showed that possible under-estimation of heat transfer from fuel/steel mixture to the duct could be enhanced with a condition where steel volume fraction is less. Future model improvement is preferable for this characteristic.
Haraga, Tomoko; Kameo, Yutaka; Ishimori, Kenichiro; Shimada, Asako; Tobita, Minoru*; Takahashi, Shigemi*; Takahashi, Kuniaki
JAEA-Data/Code 2012-031, 39 Pages, 2013/02
JAEA-Data-Code-2012-031.pdf:9.28MB
The Fugen Nuclear Power Station was shut down and decommissioning of the Fugen has been implemented. To calculate the scaling factor of radioactive waste or advance the clearance of dismantled materials, a large number of radioactivity concentration data of dismantled materials have to be accumulated. For these reasons, the simple and rapid radioactivity determination method was applied for metal samples, which were taken from pipes of the Fugen. The present report is summarized analytical procedures and obtained radioactivity data of the Fugen pipe samples.
Kameo, Yutaka; Haraga, Tomoko; Ishimori, Kenichiro; Shimada, Asako; Tobita, Minoru*; Takahashi, Shigemi*; Takahashi, Kuniaki
JAEA-Data/Code 2010-028, 32 Pages, 2011/02
JAEA-Data-Code-2010-028.pdf:1.62MB
The Fugen Nuclear Power Station was shut down and decommissioning of the Fugen has been implemented. To calculate the scaling factor of radioactive waste or advance the clearance of dismantled materials, a large number of radioactivity concentration data of dismantled materials have to be accumulated. For these reasons, the simple and rapid radioactivity determination method was applied for metal samples, which were taken from pipes of the Fugen. The present report is summarized analytical procedures and obtained radioactivity data of the Fugen pipe samples.
Iijima, Kazuki; Tomura, Tsutomu*; Tobita, Minoru*; Suzuki, Yasuyuki*
Radiochimica Acta, 98(9-11), p.729 - 736, 2010/11
Times Cited Count：3 Percentile：23.92(Chemistry, Inorganic & Nuclear)Distribution behavior of Cs and Am in the synthetic groundwater-bentonite colloids-granite ternary system was investigated. Radionuclide sorbed onto the bentonite colloids is desorbed by addition of granite, indicating that the sorption of Cs and Am onto the bentonite colloids are reversible. The sorption model based on cation exchange and surface complexation reaction considering high edge site density for bentonite colloids is applicable to explain the sorption behavior of Am and Cs in the ternary system.
Nakamura, Hirofumi; Sakurai, Shinji; Suzuki, Satoshi; Hayashi, Takumi; Enoeda, Mikio; Tobita, Kenji; DEMO Plant Design Team
Fusion Engineering and Design, 81(8-14), p.1339 - 1345, 2006/02
Times Cited Count：51 Percentile：94.72(Nuclear Science & Technology)no abstracts in English
Inoue, Takashi; Hanada, Masaya; Kashiwagi, Mieko; Nishio, Satoshi; Sakamoto, Keishi; Sato, Masayasu; Taniguchi, Masaki; Tobita, Kenji; Watanabe, Kazuhiro; DEMO Plant Design Team
Fusion Engineering and Design, 81(8-14), p.1291 - 1297, 2006/02
Times Cited Count：11 Percentile：60.11(Nuclear Science & Technology)Requirement and technical issues of the neutral beam inejctor (NBI) is discussed for fusion DEMO plant. The NBI for the fusion DEMO plant should be high efficiency, high energy and high reliability with long life. From the view point of high efficiency, use of conventional electrostatic accelerator is realistic. Due to operation under radiation environment, vacuum insulation is essential in the accelerator. According to the insulation design guideline, it was clarified that the beam energy of 1.5
2 MeV is possible in the accelerator. Development of filamentless, and cesium free ion source is required, based on the existing high current/high current density negative ion production technology. The gas neutralization is not applicable due to its low efficiency (60%). R&D on an advanced neutralization scheme such as plasma neutralization (efficiency: 
80%) is required. Recently, development of cw high power semiconductor laser is in progress. The paper shows a conceptual design of a high efficiency laser neutralizer utilizing the new semiconductor laser array.
Hayashi, Takao; Tobita, Kenji; Nishio, Satoshi; Ikeda, Kazuki*; Nakamori, Yuko*; Orimo, Shinichi*; DEMO Plant Design Team
Fusion Engineering and Design, 81(8-14), p.1285 - 1290, 2006/02
Times Cited Count：22 Percentile：79.98(Nuclear Science & Technology)Neutron transport calculations were carried out to evaluate the capability of metal hydrides and borohydrides as an advanced shielding material. Some hydrides indicated considerably higher hydrogen content than polyethylene and solid hydrogen. The hydrogen-rich hydrides show superior neutron shielding capability to the conventional materials. From the temperature dependence of dissociation pressure, ZrH and TiH can be used without releasing hydrogen at the temperature of less than 640 C at 1 atm. ZrH and Mg(BH) can reduce the thickness of the shield by 30% and 20% compared to a combination of steel and water, respectively. Mixing some hydrides with F82H produces considerable effects in 
-ray shielding. The neutron and -ray shielding capabilities decrease in order of ZrH Mg(BH) and F82H TiH and F82H water and F82H.
Sato, Masayasu; Sakurai, Shinji; Nishio, Satoshi; Tobita, Kenji; Inoue, Takashi; Nakamura, Yukiharu; Shinya, Kichiro*; Fujieda, Hirobumi*; DEMO Plant Design Team
Fusion Engineering and Design, 81(8-14), p.1277 - 1284, 2006/02
Times Cited Count：14 Percentile：68.04(Nuclear Science & Technology)no abstracts in English
Nishio, Satoshi; Omori, Junji*; Kuroda, Toshimasa*; Tobita, Kenji; Enoeda, Mikio; Tsuru, Daigo; Hirose, Takanori; Sato, Satoshi; Kawamura, Yoshinori; Nakamura, Hirofumi; et al.
Fusion Engineering and Design, 81(8-14), p.1271 - 1276, 2006/02
Times Cited Count：20 Percentile：78.77(Nuclear Science & Technology)The design guideline for the blanket is decided to meet the mission of the DEMO plant which is expected to use technologies to be proven by 2020 and present an economical prospect of fusion energy in the operational time of the reactor. To moderate the technological extrapolation, the structural material of reduced activation ferritic steel (F82H), ceramic tritium breeder of LiTiO and neutron multiplier of Be are introduced. To improve the economical aspect, the coolant material of the supercritical water with inlet/outlet temperatures of 280/510C, coolant pressure of 25 MPa is chosen. Resultantly the thermal efficiency of 41% is achieved. To obtain higher plasma performance, MHD instabilities suppressing shell structure is adopted with structural compatibility to the blanket structure. To meet higher plant availability requirement (more than 75%), the hot cell maintenance approach is selected for the replaceable power core components.
Tanaka, Masaaki; Kawashima, Shigeyo*; Igarashi, Minoru; Hayashi, Kenji; Tobita, Akira; Kamide, Hideki
JNC TN9400 2003-117, 65 Pages, 2004/03
JNC-TN9400-2003-117.pdf:3.49MB
Temperature fluctuation due to mixing of hot and cold fluids gives thermal fatigue to the structure (thermal striping phenomena).Investigation of this phenomenon is significant for the safety of a fast breeder reactor, which uses liquid metal as a coolant. In Japan Nuclear Cycle Development Institute, experiments and numerical analyses have been carried out to understand this phenomenon and also to construct the evaluation rule, which can be applied to the design. A water experiment of fluid mixing in T-pipe with long cycle fluctuation (WATLON),which notices thermal striping phenomena in the T-pipe junction, is performed to investigate the key factor of mixing phenomena by reason of long cycle fluctuation observed in a plant. By the former visualization test, it was showed that the flow pattern of branch pipe jet could be classified into (A) impinging jet, (B) deflecting jet (C) re-attachment jet and (D) wall jet according to the inflow condition. It was confirmed that the each jet pattern could be predicted by the momentum ratio of the each piping fluid. In this study, a thermo-chromic liquid crystal sheet was put on the inner wall in the main pipe, and temperature field on the wall surface was visualized. We established a new method to convert the color image data to temperature data. And measurement uncertainty of this method was evaluated + and - about 2.0 [deg-C], using by the typical picture in the temperature calibration test. From the temperature fluctuation visualization test by liquid crystal sheet, the cold spot was formed in just downstream region from the outlet of the branch pipe in the cases of the wall jet and impinging jet. Since this cold spot moved in time, high value of temperature fluctuation intensity was shown around the cold spot. And the validity of this method was shown from the comparison of the thermocouple data installed in a wall surface with the temperature conversion result.
