Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Fukaya, Yuji; Ohashi, Hirofumi; Sato, Hiroyuki; Goto, Minoru; Kunitomi, Kazuhiko
Nihon Genshiryoku Gakkai Wabun Rombunshi (Internet), 21(2), p.116 - 126, 2022/06
An improvement electricity generation cost evaluation method for High Temperature Gas-cooled Reactors (HTGRs) has been performed. Japan Atomic Energy Agency (JAEA) had completed the commercial HTGR concept named Gas Turbine High Temperature Reactor (GTHTR300) and the electricity generation cost evaluation method approximately a decade ago. The cost evaluation was developed based on the method of Federation of Electric Power Companies (FEPC). The FEPC method was drastically revised after the Fukushima Daiichi nuclear disaster. Moreover, the escalation of material and labor cost for the decade should be consider to evaluate the latest cost. Therefore, we revised the cost evaluation method for GTHTR300 and the cost was compared with that of Light Water Reactor (LWR). As a result, it was found that the electricity generation cost of HTGR of 7.9 yen/kWh is cheaper than that of LWR of 11.7 yen/kWh by approximately 30% at the capacity factor of 70%.
Koyama, Shinichi; Nakagiri, Toshio; Osaka, Masahiko; Yoshida, Hiroyuki; Kurata, Masaki; Ikeuchi, Hirotomo; Maeda, Koji; Sasaki, Shinji; Onishi, Takashi; Takano, Masahide; et al.
Hairo, Osensui Taisaku jigyo jimukyoku Homu Peji (Internet), 144 Pages, 2021/08
JAEA performed 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 and Thermal Behavior Estimation of Fuel Debris))" in 2020JFY. 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 and Contaminated Water Management.
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.
Fleurence, A.*; Lee, C.-C.*; Friedlein, R.*; Fukaya, Yuki; Yoshimoto, Shinya*; Mukai, Kozo*; Yamane, Hiroyuki*; Kosugi, Nobuhiro*; Yoshinobu, Jun*; Ozaki, Taisuke*; et al.
Physical Review B, 102(20), p.201102_1 - 201102_6, 2020/11
Times Cited Count:2 Percentile:12.28(Materials Science, Multidisciplinary)no abstracts in English
Ohashi, Hirofumi; Goto, Minoru; Ueta, Shohei; Sato, Hiroyuki; Fukaya, Yuji; Kasahara, Seiji; Sasaki, Koei; Mizuta, Naoki; Yan, X.; Aoki, Takeshi*
Proceedings of 9th International Topical Meeting on High Temperature Reactor Technology (HTR 2018) (USB Flash Drive), 6 Pages, 2018/10
Conceptual design study of an experimental HTGR is performed to upgrade the plant system from Japanese High Temperature engineering Test Reactor (HTTR) to a commercial HTGR. Safety systems of HTTR are upgraded to demonstrate the commercial HTGR concept, such as a passive reactor cavity cooling system, a confinement, etc. An intermediate heat exchanger (IHX) is replaced by a steam generator (SG) for a process heat supply to demonstrate the technology for a commercial use. This paper describes the conceptual design study results of the plant system of the experimental HTGR.
Nakagawa, Shigeaki; Sato, Hiroyuki; Fukaya, Yuji; Tokuhara, Kazumi; Ohashi, Hirofumi
JAEA-Technology 2017-022, 32 Pages, 2017/09
As for the design of commercial HTGRs, the fuel design, core design, reactor coolant system design, secondary helium system design, decay heat removal system design and confinement system design are very important and quite different from those of LWRs. To contribute the establishment of the safety standards for commercial HTGRs, the evaluation items to attain safety requirements in fuel and core designs were studied. In this study, the excellence features of HTGRs based on passive safety or inherent safety were fully reflected. Additionally, concerning the core design, the stability to spatial power oscillation in reactor core of HTGR was studied. The evaluation items as the result of the study are applicable to the safety design of commercial HTGRs in the future.
Kobayashi, Fuyumi; Sumiya, Masato; Kida, Takashi; Kokusen, Junya; Uchida, Shoji; Kaminaga, Jota; Oki, Keiichi; Fukaya, Hiroyuki; Sono, Hiroki
JAEA-Technology 2016-025, 42 Pages, 2016/11
A preliminary test on MOX fuel dissolution for the STACY critical experiments had been conducted in 2000 through 2003 at Nuclear Science Research Institute of JAEA. Accordingly, the uranyl / plutonium nitrate solution should be reconverted into oxide powder to store the fuel for a long period. For this storage, the moisture content in the oxide powder should be controlled from the viewpoint of criticality safety. The stabilization of uranium / plutonium solution was carried out under a precipitation process using ammonia or oxalic acid solution, and a calcination process using a sintering furnace. As a result of the stabilization operation, recovery rate was 95.6% for uranium and 95.0% for plutonium. Further, the recovered oxide powder was calcined again in nitrogen atmosphere and sealed immediately with a plastic bag to keep its moisture content low and to prevent from reabsorbing atmospheric moisture.
Ozawa, Mayumi; Fukaya, Hiroyuki; Sato, Makoto; Kamohara, Keiko*; Suyama, Kenya; Tonoike, Kotaro; Oki, Keiichi; Umeda, Miki
Proceedings of 53rd Annual Meeting of Hot Laboratories and Remote Handling Working Group (HOTLAB 2016) (Internet), 9 Pages, 2016/11
Honda, Yuki; Fukaya, Yuji; Nakagawa, Shigeaki; Baker, R. I.*; Sato, Hiroyuki
Proceedings of 8th International Topical Meeting on High Temperature Reactor Technology (HTR 2016) (CD-ROM), p.704 - 713, 2016/11
A high-temperature gas-cooled reactor (HTGR) has superior safety characteristics. A loss of forced cooling (LOFC) test using a high-temperature engineering test reactor (HTTR) has been carried out to verify the inherent safety of an HTGR when forced cooling is diminished without reactor scram. In the test, an all-gas circulator was tripped with an initial reactor power of 9 MW and re-criticality was shown. This study focuses on developing a point kinetics method with RELAP5-3D code for an LOFC accident. There is a large temperature difference between the inlet and outlet of the core in an HTGR, and the temperature fluctuation range has been large in several accidents. We analyze the temperature dependency of xenon-135 reactivity and show that the temperature dependency of xenon-135 microscopic absorption cross-section affected the re-criticality time of the LOFC test.
Fukaya, Masaaki*; Hata, Koji*; Akiyoshi, Kenji*; Sato, Shin*; Takeda, Yoshinori*; Miura, Norihiko*; Uyama, Masao*; Kaneda, Tsutomu*; Ueda, Tadashi*; Toda, Akiko*; et al.
JAEA-Technology 2014-040, 199 Pages, 2015/03
The researches on engineering technology in the Mizunami Underground Research Laboratory (MIU) project consists of (1) development of design and construction planning technologies, (2) development of construction technology, (3) development of countermeasure technology, (4) development of technology for security, and (5) development of technologies for restoration and/or reduction of the excavation damage. The researches on engineering technology such as verification of the initial design were being conducted by using data measured during construction as a part of the second phase of the MIU plan. Examination about the plug for reflood test in the GL-500m Access/Research Gallery-North as part of the development of technologies for restoration and/or reduction of excavation damage were carried out. Specifically, Literature survey was carried out about the plug, based on the result of literature survey, examination of the design condition, design of the plug and rock stability using numerical simulation, selection of materials for major parts, and grouting for water inflow from between rock and plug, were carried out in this study.
Suyama, Kenya; Uchiyama, Gunzo; Fukaya, Hiroyuki; Umeda, Miki; Yamamoto, Toru*; Suzuki, Motomu*
Nuclear Back-end and Transmutation Technology for Waste Disposal, p.47 - 56, 2015/00
In fission products in used nuclear fuel, there are several stable isotopes which have large neutron absorption effect. It is known that there are several hardly measurable elements in such important fission products. JAEA had been developed the method to assess the amount of fission products which are hardly measurable and have large neutron capture cross section, under the auspices of the JNES. In this development, the measurement method was developed combining a simple and effective chemical separation scheme of fission products from used nuclear fuel and ICP-MS with high-sensitivity and high-precision. This method was applied to the measurement program for used BWR 99 fuel assembly. This method is applicable to the required measurement for the countermeasure to the accident of the Fukushima Dai-ichi Nuclear Power Plants of Tokyo Electric Power Company. This presentation describes the measurement method developed in the study as well as the future measurement plan in JAEA.
Fukaya, Hiroyuki; Suyama, Kenya; Sonoda, Takashi; Okubo, Kiyoshi; Umeda, Miki; Uchiyama, Gunzo
JAEA-Research 2013-020, 81 Pages, 2013/10
Japan Atomic Energy Agency conducted a project "Isotopic Composition measurement of Fission Products in Spent Fuel from FY2008 to FY2011" by the entrustment of Japan Nuclear Energy Safety Organization. In that project, we measured the isotopic composition of neodymium isotopes which are important to evaluate the burnup value of spent nuclear fuel by using two different methods and obtained different results. So that we carried out the follow-up measurement in order to investigate the reason of the difference between two neodymium measurements. It was found that we needed correction to the measurement results of neodymium for two samples and a part of other fission products for all samples in total five samples. This report summarizes the all works carried out in this follow-up measurement and obtained results.
Ohashi, Hirofumi; Sato, Hiroyuki; Goto, Minoru; Yan, X.; Sumita, Junya; Tazawa, Yujiro*; Nomoto, Yasunobu; Aihara, Jun; Inaba, Yoshitomo; Fukaya, Yuji; et al.
International Journal of Nuclear Energy, 2013, p.918567_1 - 918567_18, 2013/00
Japan Atomic Energy Agency (JAEA) has conducted a conceptual design of a 50 MWt small-sized high temperature gas cooled reactor (HTGR) for multiple heat applications, named HTR50S, with the reactor outlet coolant temperature of 750 C and 900
C. It is first-of-a-kind of the commercial plant or a demonstration plant of a small-sized HTGR system to deploy it in developing countries in the 2020s. The design concept of HTR50S is to satisfy the user requirements for multipurpose heat application, to upgrade its performance compared to that of HTTR without significant R&D utilizing the knowledge obtained by the HTTR design and operation, and to fulfill the high level of safety by utilizing the inherent features of HTGR and a passive decay heat removal system.
Goto, Minoru; Seki, Yasuyoshi; Fukaya, Yuji; Inaba, Yoshitomo; Ohashi, Hirofumi; Sato, Hiroyuki; Tachibana, Yukio
Proceedings of 6th International Topical Meeting on High Temperature Reactor Technology (HTR 2012) (USB Flash Drive), 10 Pages, 2012/10
Japan Atomic Energy Agency (JAEA) has started a conceptual design study of a small-sized High Temperature Gas-cooled Reactor (HTGR) with 50 MW thermal power (HTR50S) to be deployed in developing countries in the 2020s. The nuclear design of the HTR50S is performed by upgrading that of a High Temperature Engineering Test Reactor (HTTR), which is the Japanese HTGR with 30 MW thermal power. In the HTTR design, 12 kinds of fuel enrichment was used to optimize the power distribution. In the previous study of the HTR50S, we succeeded in reducing the number of the fuel enrichment to 3. The present study challenges the nuclear design for effective use of uranium by utilizing high burn-up fuel and axial fuel shuffling, in which a half of the loaded fuel elements is discharged from the core every 2 years and the remains are reloaded. The core burn-up calculations were performed and the nuclear characteristics were confirmed to satisfy the design requirement.
Goto, Minoru; Seki, Yasuyoshi; Inaba, Yoshitomo; Ohashi, Hirofumi; Sato, Hiroyuki; Fukaya, Yuji; Tachibana, Yukio
JAEA-Technology 2012-017, 29 Pages, 2012/06
Japan Atomic Energy Agency has started a conceptual design of a small-sized HTGR with 50 MW thermal power (HTR50S), which is a first-of-a-kind commercial or demonstration plant of a small-sized HTGR to be deployed in developing countries in the 2020s. The nuclear of the HTR50S was performed by upgrading the proven technology of High Temperature Engineering Test Reactor (HTTR) to reduce cost for the construction. In the nuclear design, reduce the number of fuel enrichment comparing with the HTTR is one of the important subject to be upgraded. The optimization of the power distribution in the core, which is required to suppress the maximum fuel temperature below the limitation, was completed successfully by using only three fuel enrichment and the number of fuel enrichment was reduced significantly compared with the HTTR.
Goto, Minoru; Seki, Yasuyoshi; Inaba, Yoshitomo; Ohashi, Hirofumi; Sato, Hiroyuki; Fukaya, Yuji; Tachibana, Yukio
Proceedings of 2012 International Congress on Advances in Nuclear Power Plants (ICAPP '12) (CD-ROM), p.341 - 348, 2012/06
Japan Atomic Energy Agency has started a conceptual design of a small-sized HTGR with 50 MW thermal power (HTR50S), which is a first-of-a-kind commercial or demonstration plant of a small-sized HTGR to be deployed in developing countries in the 2020s. The nuclear of the HTR50S was performed by upgrading the proven technology of High Temperature Engineering Test Reactor (HTTR) to reduce cost for the construction. In the nuclear design, reduce the number of fuel enrichment comparing with the HTTR is one of the important subject to be upgraded. The optimization of the power distribution in the core, which is required to suppress the maximum fuel temperature below the limitation, was completed successfully by using only three fuel enrichment and the number of fuel enrichment was reduced significantly compared with the HTTR.
Abe, Hiroyoshi; Haga, Takahisa; Fukaya, Hiroyuki; Sonoda, Takashi; Sakazume, Yoshinori; Ito, Mitsuo; Shirahashi, Koichi
JAEA-Technology 2009-008, 24 Pages, 2009/03
Analysis of the uranyl nitrate solution fuel is carried out at the analytical laboratory of NUCEF (Nuclear Fuel Cycle Engineering Research Facility), which provides essential data for operation of STACY (Static Experiment Critical Facility), TRACY (Transient Experiment Critical Facility) and the fuel treatment system. Analyzed in FY 2007 were uranyl nitrate solution fuel samples taken before and after experiments of STACY and TRACY, samples for the preparation and pulification of uranyl nitrate solution fuel in the fuel treatment system and samples for nuclear material accountancy purpose. The total number of the samples analyzed in FY 2007 was 143. This report summarizes work related to the analysis and management of the analytical laboratory in the FY 2007.
Sakazume, Yoshinori; Aoki, Hiromichi; Haga, Takahisa; Fukaya, Hiroyuki; Sonoda, Takashi; Shimizu, Kaori; Niitsuma, Yasushi*; Ito, Mitsuo; Inoue, Takeshi
JAEA-Technology 2007-069, 44 Pages, 2008/02
Analysis of the uranyl nitrate solution fuel is carried out at the analytical laboratory of NUCEF (Nuclear Fuel Cycle Engineering Research Facility), which provides essential data for operation of STACY (Static Experiment Critical Facility), TRACY (Transient Experiment Critical Facility) and the fuel treatment system. Analyzed in FY 2006 were uranyl nitrate solution fuel samples taken before and after experiments of STACY and TRACY, samples for the preparation of uranyl nitrate solution fuel, and samples for nuclear material accountancy purpose. The total number of the samples analyzed in FY 2006 was 254. This report summarizes work related to the analysis and management of the analytical laboratory in the FY 2006.
Fukaya, Hiroyuki; Aoki, Hiromichi; Haga, Takahisa; Nishizawa, Hidetoshi; Sonoda, Takashi; Sakazume, Yoshinori; Shimizu, Kaori; Niitsuma, Yasushi*; Shirahashi, Koichi; Inoue, Takeshi
JAEA-Technology 2007-005, 27 Pages, 2007/03
Analysis of the uranyl nitrate solution fuel is carried out at the analytical laboratory of NUCEF (Nuclear Fuel Cycle Engineering Research Facility), which provides essential data for operation of STACY (Static Experiment Critical Facility), TRACY (Transient Experiment Critical Facility) and the fuel treatment system. Analyzed in FY 2005 were uranyl nitrate solution fuel samples taken before and after experiments of STACY and TRACY, samples for the preparation of uranyl nitrate solution fuel, and samples for nuclear material accountancy purpose. Also analyzed were the samples from raffinate treatment and its preliminary tests. The raffinate was generated, since FY 2000, during preliminary experiments on U/Pu extraction-pulification to fix the operation condition to prepare plutonium solution fuel to be used for criticality experiments at STACY. This report summarizes work related to the analysis and management of the analytical laboratory in the FY 2005.
Nishizawa, Hidetoshi; Fukaya, Hiroyuki; Sonoda, Takashi; Sakazume, Yoshinori; Shimizu, Kaori; Haga, Takahisa; Sakai, Yutaka*; Akutsu, Hideyuki*; Niitsuma, Yasushi; Inoue, Takeshi; et al.
JAEA-Technology 2006-007, 24 Pages, 2006/03
Analysis of the uranyl nitrate solution fuel is carried out at the analytical laboratory of NUCEF(Nuclear Fuel Cycle Engineering Research Facility), which provides essential data for operation of STACY(Static Experiment Critical Facility), TRACY(Transient Experiment Critical Facility)and the fuel treatment system. Analyzed in FY 2004 were uranyl nitrate solution fuel samples taker before and after experiments of STACY and TRACY, samples for the preparation of uranyl nitrate solution fuel, and samples for nuclear material accountancy purpose. Also analyzed were the samples from raffinate treatment and its preliminary tests. The raffinate was generated, since FY 2000, during preliminary experiments on U/Pu extraction-pulification to fix the operation condition to prepare plutonium solution fuel to be used for criticality experiments at STACY. The total number of the samples analyzed in FY 2004 was 160. This report summarizes works related to the analysis and management of the analytical laboratory in the FY 2004.