Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Mineo, Hideaki
Nihon Genshiryoku Gakkai-Shi ATOMO, 64(11), p.617 - 621, 2022/11
In December 2016 Decisions were made by the Government on the Fast Breeder Prototype Reactor "Monju", which were decommissioning of the reactor and installation of a new research reactor at the Monju site. After the decisions, MEXT started research to list reactor candidates suitable for the site. Among the candidates, medium power reactor type of which thermal output less than 10,000 kW was chosen to utilize neutron beams. Then, from 2020, MEXT launched an entrusted business and adopted JAEA, Kyoto University and University of Fukui as the core institutions of the business to carry out the conceptual design. This paper describes the system to proceed the conceptual design and to examine the utilization management of the new research reactor and also shows their status.
Takeda, Tetsuaki*; Inagaki, Yoshiyuki; Aihara, Jun; Aoki, Takeshi; Fujiwara, Yusuke; Fukaya, Yuji; Goto, Minoru; Ho, H. Q.; Iigaki, Kazuhiko; Imai, Yoshiyuki; et al.
High Temperature Gas-Cooled Reactors; JSME Series in Thermal and Nuclear Power Generation, Vol.5, 464 Pages, 2021/02
As a general overview of the research and development of a High Temperature Gas-cooled Reactor (HTGR) in JAEA, this book describes the achievements by the High Temperature Engineering Test Reactor (HTTR) on the designs, key component technologies such as fuel, reactor internals, high temperature components, etc., and operational experience such as rise-to-power tests, high temperature operation at 950C, safety demonstration tests, etc. In addition, based on the knowledge of the HTTR, the development of designs and component technologies such as high performance fuel, helium gas turbine and hydrogen production by IS process for commercial HTGRs are described. These results are very useful for the future development of HTGRs. This book is published as one of a series of technical books on fossil fuel and nuclear energy systems by the Power Energy Systems Division of the Japan Society of Mechanical Engineers.
Yonomoto, Taisuke; Mineo, Hideaki; Murayama, Yoji; Hohara, Shinya*; Nakajima, Ken*; Nakatsuka, Toru; Uesaka, Mitsuru*
Nihon Genshiryoku Gakkai-Shi ATOMO, 63(1), p.73 - 77, 2021/01
no abstracts in English
Mineo, Hideaki; Nishihara, Tetsuo; Ohashi, Hirofumi; Goto, Minoru; Sato, Hiroyuki; Takegami, Hiroaki
Nihon Genshiryoku Gakkai-Shi ATOMO, 62(9), p.504 - 508, 2020/09
High-Temperature Gas-cooled Reactor (HTGR) is one of thermal neutron reactor-type that employs helium gas coolant and graphite moderator. It has excellent inherent safety and can supply high-temperature heat which can be used not only for electric power generation but also for a wide range of application such as hydrogen production. Therefore, HTGR is expected to be an effective technology for reducing greenhouse gases in Japan as well as overseas. In this paper, we will introduce the forefront of technological development that JAEA is working toward the realization of an HTGR system consisting of a high temperature gas reactor and heat utilization facilities such as gas-turbine power generation and hydrogen production.
Uesaka, Mitsuru*; Mineo, Hideaki
Nihon Genshiryoku Gakkai-Shi ATOMO, 58(8), p.468 - 473, 2016/08
All the research reactors and critical assemblies (hereinafter RRCAs) in Japan are stopped in order to fulfil the new regulatory requirements, which were reinforced after the accident at the Tokyo Electric Power Company's Fukushima Daiichi Nuclear Power Station. These RRCAs have played important roles in the areas of human resource development, academic research, medical and industrial application of nuclear technology. Prolonged stoppage of RRCAs affects adversely those activities. Atomic Energy Society of Japan set up a group to discuss this issue. The group has shown a proposal that the roles of the RRCAs, which are indispensable facilities to nuclear human resource development, should be placed positively in the energy policy and the science and technology policy of the country.
Morita, Yasuji; Kawata, Yoshihisa*; Mineo, Hideaki; Koshino, Nobuyoshi*; Asanuma, Noriko*; Ikeda, Yasuhisa*; Yamasaki, Kazuhiko*; Chikazawa, Takahiro*; Tamaki, Yoshihisa*; Kikuchi, Toshiaki*
Journal of Nuclear Science and Technology, 44(3), p.354 - 360, 2007/03
Times Cited Count:16 Percentile:71.35(Nuclear Science & Technology)Precipitation behavior of Pu and other transuranium elements with N-cyclohexyl-2-pyrrolidone (NCP) has been examined to develop a simple reprocessing based only on precipitation method. From HNO solutions containing only Pu, both Pu(VI) and Pu(IV) were precipitated with NCP, but they required more NCP than in the U(VI) precipitation. Selective U(VI) precipitation from HNO solution containing U(VI) and Pu(IV) was achieved by stirring the solution for sufficient time after addition of NCP with ratio of [NCP]/[U]=1.4. Addition of an enough amount of NCP to U(VI)-Pu(VI) or U(VI)-Pu(IV) solutions gave a quantitative precipitation of both U and Pu. Neither Am(III) nor Np(V) was precipitated in the selective U precipitation and the simultaneous U-Pu precipitation. These results demonstrate the feasibility of the reprocessing by precipitation with NCP.
Morita, Yasuji; Kawata, Yoshihisa*; Mineo, Hideaki; Koshino, Nobuyoshi*; Asanuma, Noriko*; Ikeda, Yasuhisa*; Yamasaki, Kazuhiko*; Chikazawa, Takahiro*; Tamaki, Yoshihisa*; Kikuchi, Toshiaki*
Proceedings of International Conference on Nuclear Energy System for Future Generation and Global Sustainability (GLOBAL 2005) (CD-ROM), 6 Pages, 2005/10
N-cyclohexyl-2-pyrrolidone (NCP) can selectively precipitate U(VI) ions in aqueous nitric acid solutions. Utilizing this property, we have been developing a simple reprocessing process of spent nuclear fuel based only on precipitation method. In the first precipitation step, only U is separated by precipitation in a yield of about 70%, and in the second precipitation step both U and Pu are recovered and separated from fission products (FP) and other transuranium elements (TRU). In JAERI, precipitation behaviors of Pu and other TRU were examined experimentally, and the results showed the feasibility of the process establishement.
Yamasaki, Kazuhiko*; Chikazawa, Takahiro*; Tamaki, Yoshihisa*; Kikuchi, Toshiaki*; Morita, Yasuji; Kawata, Yoshihisa*; Mineo, Hideaki; Koshino, Nobuyoshi*; Asanuma, Noriko*; Harada, Masayuki*; et al.
Proceedings of International Conference on Nuclear Energy System for Future Generation and Global Sustainability (GLOBAL 2005) (CD-ROM), 4 Pages, 2005/10
N-cyclohexyl-2-pyrrolidone (NCP), can selectively precipitate U(VI) ions in aqueous nitric acid solutions. Utilizing this property, we have been developing a simple reprocessing process of spent nuclear fuel based only on precipitation method. In the first precipitation step, only U is separated by precipitation in a yield of about 70%, and in the second precipitation step both U and Pu are recovered and separated from fission products (FP) and other transuranium elements (TRU). In the present study, a precipitator and a precipitate separator were designed and built up, and were tested with aspets of operationability and system performance.
Morita, Yasuji; Asakura, Toshihide; Mineo, Hideaki; Hotoku, Shinobu; Uchiyama, Gunzo
JAERI-Conf 2005-007, p.25 - 30, 2005/08
Researches on process safety of reprocessing, development of an advanced reprocessing and partitioning of high-level liquid waste(HLLW) have been conducted in NUCEF - BECKY (Back-end Fuel Cycle Key Elements Research Facility), which has alpha-gamma concrete cells and many glove-boxes. This paper presents 10 year accomplishment of the above researches and future activities to be conducted in the field of separation process development.
Ban, Yasutoshi; Mineo, Hideaki; Asakura, Toshihide; Hotoku, Shinobu; Matsumura, Masakazu; Kim, S.-Y.; Morita, Yasuji
JAERI-Conf 2004-011, p.101 - 102, 2004/07
Experimental studies and numerical analysis on physical and chemical behavior of actinide elements in the solutions of aqueous reprocessing process have been done for compiling technical data that are necessary for evaluating processes such as reprocessing facilities. The objective of the present study is obtaining technical data that are conductive to the construction of reprocessing process that cope with high burn-up nuclear fuel, the evaluation of nuclear fuel cycle, and the drawing up policy on reprocessing.
Iwamura, Takamichi; Okubo, Tsutomu; Akie, Hiroshi; Kugo, Teruhiko; Yonomoto, Taisuke; Kureta, Masatoshi; Ishikawa, Nobuyuki; Nagaya, Yasunobu; Araya, Fumimasa; Okajima, Shigeaki; et al.
JAERI-Research 2004-008, 383 Pages, 2004/06
The present report contains the achievement of "Research and Development on Reduced-Moderation Light Water Reactor with Passive Safety Features", which was performed by Japan Atomic Energy Research Institute (JAERI), Hitachi Ltd., Japan Atomic Power Company and Tokyo Institute of Technology in FY2000-2002 as the innovative and viable nuclear energy technology (IVNET) development project operated by the Institute of Applied Energy (IAE). In the present project, the reduced-moderation water reactor (RMWR) has been developed to ensure sustainable energy supply and to solve the recent problems of nuclear power and nuclear fuel cycle, such as economical competitiveness, effective use of plutonium and reduction of spent fuel storage. The RMWR can attain the favorable characteristics such as high burnup, long operation cycle, multiple recycling of plutonium (Pu) and effective utilization of uranium resources based on accumulated LWR technologies.
Asakura, Toshihide; Hotoku, Shinobu; Ban, Yasutoshi; Matsumura, Masakazu*; Kim, S.-Y.; Mineo, Hideaki; Morita, Yasuji
Proceedings of International Conference ATALANTE 2004 Advances for Future Nuclear Fuel Cycles, 5 Pages, 2004/06
In JAERI, PARC process based on PUREX technique has been studied to as the basis of future reprocessing. The key of concept is to obtain the products, U and Pu, within only a single extraction cycle by separating Np and Tc from U and Pu before U/Pu partition. Two flow-sheet tests on the process were performed with 44 GWd/t PWR spent-fuel solutions. It was demonstrated that remaining Np in raffinate from co-extraction could be decreased to 13 % compared to the dissolver solution with increased solvent flow rate and with increased nitric acid concentration of FP scrubbing solution. It was demonstrated that Np separation (selective reduction by n-butyraldehyde) efficiency could be improved from 36 % to 78 % by flow-sheet modification; increasing reductant concentration and scrubbing solution flow rate. The feasibility of the Tc separation technique by high acid scrubbing was demonstrated.
Mineo, Hideaki; Isogai, Hikaru; Morita, Yasuji; Uchiyama, Gunzo*
Journal of Nuclear Science and Technology, 41(2), p.126 - 134, 2004/02
Times Cited Count:7 Percentile:44.60(Nuclear Science & Technology)A simple equation was proposed for the dissolution rate of spent LWR fuel, of which the change in the dissolution area was estimated by taking into account of the area of the cracks occurring due to thermal shrinkage of the pellets during irradiation. The applicability of proposed equation was examined using LWR fuel dissolution test results in the present study as well as the results obtained by other workers. The equation showed good agreements with the dissolution test results obtained from spent fuel pellets and pulverized spent fuel. It was indicated that the proposed equation was simple and would be useful for the prediction of dissolution of spent LWR fuels. However, the initial effective dissolution area, the parameter of the equation, was found to depend on the temperature, which could not be explained by the proposed equation. Further studies on the role of other factors affecting dissolution rate, such as nitrous acid, in the dissolution of spent fuel was required.
Mineo, Hideaki; Asakura, Toshihide; Hotoku, Shinobu; Ban, Yasutoshi; Morita, Yasuji
Proceedings of GLOBAL2003 Atoms for Prosperity; Updating Eisenhower's Global Vision for Nuclear Energy (CD-ROM), p.1250 - 1255, 2003/11
An advanced aqueous reprocessing process has been proposed for the next generation fuel cycle. Key technologies applied to the process are: removal of I-129, separation of Np and FP(Tc) separation by selective reduction of Np(VI) and high acid scrubbing of Tc within a single cycle process, MA separation by extraction chromatography and Cs/Sr separation. U separation just after dissolution was supposed to be effective to reduce the required capacity of the following extraction step. Among them Np reduction rate in TBP solution was measured, which was found to be lower than that in aqueous solution. Using an improved flow sheet spent fuel test, based on the Np reduction test, was carried out and about 90% of Np was separated before U and Pu partitioning step.
Mineo, Hideaki; Goto, Minoru; Iizuka, Masaru*; Fujisaki, Susumu; Hagiya, Hiromichi*; Uchiyama, Gunzo
Separation Science and Technology, 38(9), p.1981 - 2001, 2003/05
Times Cited Count:24 Percentile:65.31(Chemistry, Multidisciplinary)no abstracts in English
Mineo, Hideaki; Suzuki, Tadashi; Morita, Yasuji
Proceedings of 2nd International Symposium on Supercritical Fluid Technology for Energy and Environment Applications (Super Green 2003), p.334 - 338, 2003/00
Behavior of spent nuclear fuel in subcritical water was investigated to look at the feasibility of fission-products (FPs) separation without organic solvent. The study employed unirradiated UO particles simulating spent fuel burned up to 45,000MWdt, which includes FP elements in oxide form: Sr, Zr, Mo, Ru, Rh, Pd, Ag, Ba, La, Ce, Pr, Nd and Sm. Also, alloy particles consisted of Mo, Ru, Rh and Pd were prepared to simulate the metallic phase of FP. 12.728 g of the fuel and 52 mg of the alloy were placed in a 10 ml pressure vessel, where subcritical water was fed. The temperature was 523, 573, 623 and 663K, while the pressure was kept at 29MPa. Dissolved fraction decreased with elevating temperature. It was found that more than 5% of Ba, Mo and Pr were respectively dissolved. The dissolved fraction of Sr and Rh were about 1%, and about 0.3% for Zr. La, Ce, Nd and Sm, indicated almost the same result as U, which was about 0.1%. It was suggested that the subcritical water could separate portion of FP. Further study would be carried out with smaller-sized fuel.
Uchiyama, Gunzo; Mineo, Hideaki; Asakura, Toshihide; Hotoku, Shinobu; Iizuka, Masaru*; Fujisaki, Susumu; Isogai, Hikaru; Ito, Yoshinori*; Sato, Makoto; Hosoya, Noriaki
Journal of Nuclear Science and Technology, 39(Suppl.3), p.925 - 928, 2002/11
no abstracts in English
Hotoku, Shinobu; Asakura, Toshihide; Mineo, Hideaki; Uchiyama, Gunzo
Journal of Nuclear Science and Technology, 39(Suppl.3), p.313 - 316, 2002/11
no abstracts in English
Mineo, Hideaki; Goto, Minoru; Iizuka, Masaru*; Fujisaki, Susumu; Uchiyama, Gunzo
Journal of Nuclear Science and Technology, 39(3), p.241 - 247, 2002/03
Times Cited Count:28 Percentile:82.74(Nuclear Science & Technology)no abstracts in English
Uchiyama, Gunzo; Mineo, Hideaki; Asakura, Toshihide; Hotoku, Shinobu
Nihon Kikai Gakkai Dai-8-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu, p.483 - 488, 2002/00
no abstracts in English