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Shibata, Motoki*; Nakanishi, Yohei*; Abe, Jun*; Arima, Hiroshi*; Iwase, Hiroki*; Shibayama, Mitsuhiro*; Motokawa, Ryuhei; Kumada, Takayuki; Takata, Shinichi; Yamamoto, Katsuhiro*; et al.
Polymer Journal, 55(11), p.1165 - 1170, 2023/11
Times Cited Count：1 Percentile：51.7(Polymer Science)
)O
( = Th, Np) prepared by supercritical hydrothermal synthesisShirasaki, Kenji*; Tabata, Chihiro*; Sunaga, Ayaki*; Sakai, Hironori; Li, D.*; Konaka, Mariko*; Yamamura, Tomoo*
Journal of Nuclear Materials, 563, p.153608_1 - 153608_11, 2022/05
Times Cited Count：2 Percentile：53.91(Materials Science, Multidisciplinary)We focused on the direct synthesis of (U, )O solid solution (=Th, Np) by extending our recent progress in hydrothermal synthesis with additives. The homogeneity of the (U, )O ( = Th, Np) systems prepared by supercritical hydrothermal reactions was investigated through crystallographic analysis based on Vegard's law, and the 
Na nuclear magnetic resonance (NMR) measurement of (U, Np, Na)O solid solutions. Our experimental and analytical results revealed that (i) an optimal additive is ammonium carbonate and starting uranium valence is IV in the case of (U, Th)O
, and (ii) an optimal additive is ethanol and starting uranium valence is VI in the case of (U, Np)O, for producing the homogeneous solid solutions by hydrothermal synthesis.
; Stoichiometry, crystal shape and size, and homogeneity observed using Na-NMR spectroscopy of (U, Na)OTabata, Chihiro*; Shirasaki, Kenji*; Sunaga, Ayaki*; Sakai, Hironori; Li, D.*; Konaka, Mariko*; Yamamura, Tomoo*
CrystEngComm (Internet), 23(48), p.8660 - 8672, 2021/12
Times Cited Count：5 Percentile：64.74(Chemistry, Multidisciplinary)The hydrothermal synthesis of pure uranium dioxide under supercritical water (SCW) conditions was investigated. The nonstoichiometry, crystallite size and morphology of the UO particles were investigated. The SCW hydrothermal synthesis may be a promising method for producing homogeneous UO and its solid solutions with well-defined nonstoichiometries (
), shapes, and sizes.
Sagayama, Yutaka; Ando, Masato
Nihon Genshiryoku Gakkai-Shi ATOMO
, 60(3), p.162 - 167, 2018/03
The Generation IV international Forum (GIF) has led international collaborative efforts to develop six next generation nuclear energy systems, such as Sodium-cooled Fast Reactor (SFR), Lead-cooled Fast Reactor (LFR), Gas-cooled Fast Reactor (GFR), Molten Salt Reactor (MSR), Supercritical Water-cooled Reactor (SCWR), and Very High Temperature Reactor (VHTR), which have superior characteristics for the Safety and Reliability, Economics, Sustainability, Proliferation Resistance and Physical Protection. Some systems are already in the Demonstration Phase and the commercialization of the system in 2030s, which is the target of GIF, comes into sight.
Tatsumoto, Hideki; Kato, Takashi; Aso, Tomokazu; Hasegawa, Shoichi; Ushijima, Isamu*; Otsu, Kiichi*; Ikeda, Yujiro
LA-UR-06-3904, Vol.2, p.426 - 434, 2006/06
In JSNS, Cadmium has been selected as a poison material in a hydrogen moderator to obtain narrow neutron pulse. The concern to adopt to Cd is how to bond Cd and Al alloy plate. R&Ds for bonding have been performed. But good bonding has not been obtained. Consequently, heat transfer between Cd poison and cryogenic hydrogen was studied for the case of insufficient bonding. The heat transfers for various bonding ratios were analyzed by CFD code (STAR-CD) without any turbulence model. The temperature rise in Cd poison for insufficient bonding was estimated. As a result, even the case of the bonding ratio of only 5 %, the maximum temperature of Cd is around 75K. Therefore, the expected heat transfer between the Cd poison and the hydrogen should be sufficient for insufficient bonding. Then, it is found that the any bonding method should be available for manufacturing method of Cd poison.
Hasegawa, Shoichi; Kato, Takashi; Aso, Tomokazu; Ushijima, Isamu*; Tatsumoto, Hideki; Otsu, Kiichi*; Ikeda, Yujiro
LA-UR-06-3904, Vol.2, p.402 - 407, 2006/06
In JSNS, the hydrogen of super critical state is adopted as moderators. Therefore the cryogenic hydrogen system is prepared, which consists of hydrogen circulating unit and transfer lines to moderators. The hydrogen system will immediately discharge hydrogen when an off-normal event occurs. In case of emergency, helium gas will be inputted to an insulation vacuum of the transfer line in order to enhance the heat transfer and the hydrogen discharge time should be shortened. Then, it is impotant to estimate the behaviour of pressure and velocity of discharging hydrogen in the emergency. During hydrogen discharge, the pressure rise in the discharge piping should be kept below the design pressure of 0.1 MPa. The result of analysis shows that the pressure of helium gas injection is suitable less than 0.04Mpa, and that the maximum hydrogen discharge flow is evaluated to be 0.047 kg/s after around 150 seconds. After five minutes with this condition, the hydrogen of around 90% in the moderator piping is discharged. Safety hydrogen release in the case of emergency can be established.
Tatsumoto, Hideki; Aso, Tomokazu; Hasegawa, Shoichi; Ushijima, Isamu*; Kato, Takashi; Otsu, Kiichi*; Ikeda, Yujiro
AIP Conference Proceedings 823, p.753 - 760, 2006/05
As one of the main experimental facilities in J-PARC, an intense spallation neutron source (JSNS) driven by proton beam power of 1 MW is constructed. In JSNS, cryogenic hydrogen at supercritical pressure is selected as a moderator. The total nuclear heating at the moderators is estimated to be 3.7 kW. A cryogenic hydrogen system to cool the moderators has been designed. As the most severe off-normal event for cryogenic hydrogen system, it is considered that the cryogenic hydrogen leaks when the pipe is cracked. In such a case, the hydrogen must be discharged safely as soon as possible. An analytical code that simulated the pressure change during hydrogen leak was developed. The pressure rise analysis for various sized cracks was performed, and then the required size of safety equipment was determined. It was found from the analysis that a safety valve of -42.7 mm and a rupture disk with the diameter of 37.1 mm can discharge hydrogen safely.
Tatsumoto, Hideki; Kato, Takashi; Aso, Tomokazu; Ushijima, Isamu*; Hasegawa, Shoichi; Otsu, Kiichi*
JAERI-Tech 2005-019, 16 Pages, 2005/03
JAERI-Tech-2005-019.pdf:1.29MB
As one of the main experimental facilities in J-PARC, an intense spallation neutron source (JSNS) is constructed. In JSNS, cryogenic hydrogen with temperature of 20 K and pressure of 0.5 to 1.5 MPa was selected as the moderator. The total nuclear heating at the moderators is estimated to be 3.7 kW for proton beam power of 1 MW. A cryogenic hydrogen circulation system, which plays a role in cooling spallation neutron and moderators, has been designed. For a certain operation condition, it is possible to occur boiling in the moderators. The boiling phenomenon would have an influence on the neutronic performance and the safety of the moderators. The heat transfer mechanism of cryogenic hydrogen in the moderators needs to be estimated. However, the mechanism has not been clarified until now. In this paper, the heat transfer of cryogenic hydrogen was estimated by using properties of cryogenic hydrogen and the heat transfer correlations used in other fluids, and then the operation condition of the cryogenic hydrogen system has been considered.
Meguro, Yoshihiro; Tomioka, Osamu; Imai, Tomoki*; Fujimoto, Shigeyuki*; Nakashima, Mikio; Yoshida, Zenko; Honda, Tadashi*; Koya, Fumio*; Kitamura, Nobu*; Wada, Ryutaro*; et al.
Proceedings of International Waste Management Symposium 2004 (WM '04) (CD-ROM), 8 Pages, 2004/03
Supercritical CO
fluid leaching (SFL) method using supercritical CO fluid containing a complex of HNO
- tri-n-butyl phosphate (TBP) was applied to removal of uranium from radioactive solid wastes. Sea sands, incineration ashes and porous alumina bricks were employed as matrixes of simulated solid wastes. Real radioactive incineration ash wastes and firebrick wastes were also subjected to the SFL treatment. It was found that uranium could be efficiently removed from both of the simulated wastes and real wastes by the SFL method. The removal efficiency of uranium from the real waste was lower than that from the corresponding artificial waste. About 1 g and 35 mg of uranium were recovered from 10 g of the real ash waste and 37 g of the real firebrick waste, respectively.
Dairaku, Masayuki; Ezato, Koichiro; Suzuki, Satoshi; Akiba, Masato; Imai, Tsuyoshi
KEK Proceedings 2003-16 (CD-ROM), 4 Pages, 2004/02
no abstracts in English
Meguro, Yoshihiro; Ogiyanagi, Jin*; Tomioka, Osamu; Imura, Hisanori*; Ohashi, Kozaburo*; Yoshida, Zenko; Nakashima, Mikio
Proceedings of 2nd International Symposium on Supercritical Fluid Technology for Energy and Environment Applications (Super Green 2003), p.175 - 179, 2004/00
One of the most attractive properties of SFE is that changing solvent properties by tuning pressure can control distribution behavior of a metal ion. Distribution ratio (D) of uranium(VI) and plutonium(IV) with tributyl phosphate (TBP) from a nitric acid solution and palladium(II) with 2-methyl-8-qunolinol (HMQ) from a hydrochloric solution were determined in SFE at various pressures. In the extraction system using TBP, a linear relationship between the logarithmic distribution ratio (log D) and the solubility parameter of CO was observed. The solubility parameter is difined based on the regular solution theory and is one of the parameters depending on the pressure. On the other hand, a linear relationship with a positive slope between log D and the solubility parameter was observed in the extraction system using HMQ. Most of the extractant was dissolved in the aqueous phase as HMQ
under the extraction condition examined.
Watanabe, Takeshi*; Tsushima, Satoru*; Yamamoto, Ichiro*; Tomioka, Osamu; Meguro, Yoshihiro; Nakashima, Mikio; Wada, Ryutaro*; Nagase, Yoshiyuki*; Fukuzato, Ryuichi*
Proceedings of 2nd International Symposium on Supercritical Fluid Technology for Energy and Environment Applications (Super Green 2003), p.363 - 366, 2004/00
Recovery of salts by supercritical fluid leaching (SFL) method using carbon dioxide was experimentally studied. It was confirmed that LiCl was recovered with a mixed fluid of carbon dioxide and methanol, and KCl and SrCl were recovered with a mixed fluid of carbon dioxide, methanol and crown ether. The influence of crown ether for KCl and SrCl extraction were found to increase in the order of 15-crown-5 (15C5) 
18-crown-6 (18C6) dicychlohexyl-18-crown-6 (DC18C6). It is expected that other salts can be recovered selectively with a mixed fluid of carbon dioxide, methanol and suitable crown ether.
Nagase, Yoshiyuki*; Masuda, Kaoru*; Wada, Ryutaro*; Yamamoto, Ichiro*; Tomioka, Osamu; Meguro, Yoshihiro; Fukuzato, Ryuichi*
Proceedings of 2nd International Symposium on Supercritical Fluid Technology for Energy and Environment Applications (Super Green 2003), p.254 - 257, 2004/00
no abstracts in English
Enoeda, Mikio; Kosaku, Yasuo; Hatano, Toshihisa; Kuroda, Toshimasa*; Miki, Nobuharu*; Homma, Takashi; Akiba, Masato; Konishi, Satoshi; Nakamura, Hirofumi; Kawamura, Yoshinori; et al.
Nuclear Fusion, 43(12), p.1837 - 1844, 2003/12
Times Cited Count：101 Percentile：93.45(Physics, Fluids & Plasmas)no abstracts in English
Kurihara, Ryoichi; Watanabe, Kenichi*; Konishi, Satoshi
JAERI-Review 2003-020, 37 Pages, 2003/07
JAERI-Review-2003-020.pdf:2.08MB
no abstracts in English
Konishi, Satoshi; Nishio, Satoshi; Tobita, Kenji; DEMO Design Team
Fusion Engineering and Design, 63-64, p.11 - 17, 2002/12
Times Cited Count：50 Percentile：93.5(Nuclear Science & Technology)The first fusion power plant DEMO must have some reality that ITER and other facilities in the same period are expected to prove its feasibility. The DEMO should also be so attractive and advanced that the future society would be interested in constructing based on its concept. The present DEMO plant concept intends to satisfy these antagonistic requirements assuming construction in 2030s immediately after successful completion of fundamental ITER mission. A steady tokamak is minimized to have 5.8m of major radius with 2.3GW with Q exceeds 30. Modestly ambitious plasma parameters are chosen. Technology improvement is assumed to make maximum 20 T magnet, metal first wall and super critical water cooled ITER-like blanket modules feasible. Tritium inventory is reduced to 1kg with improved safety system concept. This conceptual design identifies various technical issues that are expected to be solved by intensive R&D efforts during ITER period, and indicates a possible step immediately after ITER.
Inui, Masanori*; Hong, X.*; Matsusaka, Tetsuya*; Ishikawa, Daisuke*; Kazi, M. H.*; Tamura, Kozaburo*; Funakoshi, Kenichi*; Utsumi, Wataru
Journal of Non-Crystalline Solids, 312-314, p.274 - 278, 2002/10
Times Cited Count：2 Percentile：30.84(Materials Science, Ceramics)no abstracts in English
Kakuta, Toshiya*; Hirata, Shingo*; Mori, Seiji*; Konishi, Satoshi; Kawamura, Yoshinori; Nishi, Masataka; Ohara, Yoshihiro
Fusion Science and Technology, 41(3), p.1069 - 1073, 2002/05
Research-and-development of the supercritical water-cooled prototype fusion reactor which has cost competitiveness has been performed in Japan Atomic Energy Research Institute (JAERI). It is necessary to establish immediately the design concept of the blanket tritium recovery system which collects tritium continuously and safely from the supercritical water-cooled blanket because fuel self-sufficiency is inevitable in the prototype reactor. The candidate systems are; 1) batch-processing cryogenic molecular sheave bed recovery system with cryogenic temperature operation, 2) continuous processing Pd membrane penetration recovery system with high vacuum operation. In the present study, however, the third candidate system, the hydrogen pump system with protonic conductors, was investigated. As a result of the study, it was made clear that the system with minimized energy consumption and minimized accidental tritium release could be realized by using the hydrogen pump for the blanket tritium recovery system of the prototype fusion reactor.
Kosaku, Yasuo; Yanagi, Yoshihiko*; Enoeda, Mikio; Akiba, Masato
Fusion Science and Technology, 41(3), p.958 - 961, 2002/05
As a candidate DEMO blanket, the design of solid breeder blanket cooled by supercritical water has been performed. The candidate structural material is F82H. The coolant is supercritical water (pressure; 25 MPa, temperature; 550-780K) to achieve high generation efficiency. The temperature of cooling tubes in tritium breeder zone has been evaluated at 650-800K. In this temperature range, tritium permeation must be investigated from the view point of safety management, because high temperature coolant is directly supplied to the power generation system. In the present work, the tritium permeation into the first wall cooling water by the implantation and that through cooling tubes in tritium breeder zone have been evaluated. Assuming tritium injection energy and flux are same as SSTR, the calculated value of the tritium permeation rate into the first wall cooling water is 68.3 g/day. On the other hand, that of the permeation rate through cooling tubes is 75.3 g/day (20% of generated tritium) when helium gas flows so that tritium partial pressure becomes 1 Pa at the outlet.
Nakajima, Ken; Yamane, Yuichi; Ogawa, Kazuhiko; Aizawa, Eiju; Yanagisawa, Hiroshi; Miyoshi, Yoshinori
JAERI-Data/Code 2002-007, 123 Pages, 2002/03
JAERI-Data-Code-2002-007.pdf:4.34MB
no abstracts in English
