Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
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 950
C, 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.
Tachi, Yukio; Sato, Tomofumi*; Akagi, Yosuke*; Kawamura, Makoto*; Nakane, Hideji*; Terashima, Motoki; Fujiwara, Kenso; Iijima, Kazuki
Science of the Total Environment, 724, p.138098_1 - 138098_11, 2020/07
Times Cited Count：14 Percentile：56.04(Environmental Sciences)To understand and predict radiocesium transport behaviors in the environment, highly contaminated sediments from Ukedo and Odaka rivers around the Fukushima Daiichi Nuclear Power Plant were investigated systematically focusing on key factors controlling radiocesium sorption and fixation, including particle size, clay mineralogy and organic matter.
Tachi, Yukio; Sato, Tomofumi*; Takeda, Chizuko*; Ishidera, Takamitsu; Fujiwara, Kenso; Iijima, Kazuki
Science of the Total Environment, 724, p.138097_1 - 138097_10, 2020/07
Times Cited Count：9 Percentile：42.38(Environmental Sciences)To understand and predict radiocesium transport behaviors in the environment, sorption and fixation behaviors of radiocesium on river sediments from Ukedo and Odaka rivers around the Fukushima Daiichi Nuclear Power Plant were investigated systematically focusing on Cs sorption and fixation mechanisms and their relationship with Cs concentrations and sediment properties including clay mineralogy and organic matter.
Tachi, Yukio; Ito, Tsuyoshi*; Akagi, Yosuke*; Sato, Hisao*; Martin, A. J.*
Water Resources Research, 54(11), p.9287 - 9305, 2018/11
Times Cited Count：5 Percentile：26.11(Environmental Sciences)Effects of fine-scale surface alterations on radionuclide migration in fractured crystalline rocks were investigated by a comprehensive approach coupling a series of laboratory tests, microscopic observations and modelling, using a single fractured granodiorite sample from the Grimsel Test Site, Switzerland. Laboratory tests including through-diffusion, batch sorption and flow-through tests using five tracers indicated that tracer retention was consistently in the sequence of HDO, Se, Cs, Ni, Eu, and as well as showing the existence of a diffusion-resistance layer near the fracture surface, cation excess and anion exclusion effects for diffusion. Microscale heterogeneities in structural properties around the fracture were clarified quantitatively by coupling X-ray CT and EPMA. A three layer model including weathered vermiculite, foliated mica and undisturbed matrix layers, and their properties such as porosity, sorption and diffusion parameters, could provide a reasonable interpretation for breakthrough curves and concentration distributions near fracture surface of all tracers, measured in flow-through tests.
test plantYan, X.; Sato, Hiroyuki; Sumita, Junya; Nomoto, Yasunobu*; Horii, Shoichi*; Imai, Yoshiyuki; Kasahara, Seiji; Suzuki, Koichi*; Iwatsuki, Jin; Terada, Atsuhiko; et al.
Nuclear Engineering and Design, 329, p.223 - 233, 2018/04
Times Cited Count：20 Percentile：90.01(Nuclear Science & Technology)The pre-licensing design of an HTGR cogeneration test plant to be coupled to JAEA's existing test reactor HTTR is presented. The plant is designed to demonstrate the system of JAEA commercial plant design GTHTR300C. With construction planned to be completed around 2025, the test plant is expected to be the first-of-a-kind nuclear system operating on two of the advanced energy conversion systems attractive for the HTGR closed cycle helium gas turbine for power generation and thermochemical iodine-sulfur water-splitting process for hydrogen production.
Theis, C.*; Carbonez, P.*; Feldbaumer, E.*; Forkel-Wirth, D.*; Jaegerhofer, L.*; Pangallo, M.*; Perrin, D.*; Urscheler, C.*; Roesler, S.*; Vincke, H.*; et al.
EPJ Web of Conferences, 153, p.08018_1 - 08018_5, 2017/09
Times Cited Count：0 Percentile：0.03(Nuclear Science & Technology)At CERN, gas-filled ionization chambers PTW-34031 (PMI) are commonly used in radiation fields including neutrons, protons and 
-rays. A response function for each particle is calculated by the radiation transport code FLUKA. To validate a response function to high energy neutrons, benchmark experiments with quasi mono-energetic neutrons have been carried out at RCNP, Osaka University. For neutron irradiation with energies below 200 MeV, very good agreement was found comparing the FLUKA simulations and the measurements. In addition it was found that at proton energies of 250 and 392 MeV, results calculated with neutron sources underestimate the experimental data due to a non-negligible gamma component originating from the target 
Li(p,n)Be reaction.
test plant; System designYan, X.; Sato, Hiroyuki; Sumita, Junya; Nomoto, Yasunobu; Horii, Shoichi; Imai, Yoshiyuki; Kasahara, Seiji; Suzuki, Koichi*; Iwatsuki, Jin; Terada, Atsuhiko; et al.
Proceedings of 8th International Topical Meeting on High Temperature Reactor Technology (HTR 2016) (CD-ROM), p.827 - 836, 2016/11
Pre-licensing basic design for a cogenerating HTGR test plant system is presented. The plant to be coupled to existing 30 MWt 950C test reactor HTTR is intended as a system technology demonstrator for GTHTR300C plant design. More specifically the test plant of HTTR-GT/H aims to (1)demonstrate the licensability of the GTHTR300C for electricity production by gas turbine and hydrogen cogeneration by thermochemical process and (2) confirm the operation control and safety of such cogeneration system. With construction and operation completion by 2025, the test plant is expected to be the first of a kind HTGR-powered cogeneration plant operating on the two advanced energy conversion systems of closed cycle helium gas turbine for power generation and thermochemical iodine-sulfur water-splitting process for hydrogen production.
Yan, X.; Sato, Hiroyuki; Kamiji, Yu; Imai, Yoshiyuki; Terada, Atsuhiko; Tachibana, Yukio; Kunitomi, Kazuhiko
Nuclear Engineering and Design, 306, p.215 - 220, 2016/09
Times Cited Count：4 Percentile：36.27(Nuclear Science & Technology)The latest design upgrade has incorporated several major technological advances made in the past ten years to both reactor and balance of plant in GTHTR300. As described in this paper, these advances have enabled raising the design basis reactor core outlet temperature to 950C and increasing power generating efficiency by nearly 5% point. Further implementation of seawater desalination cogeneration is made through employing a newly-proposed multi-stage flash process. Through efficient waste heat recovery of the reactor gas turbine power conversion cycle, a large cost credit is obtained against the conventionally produced water prices. Together, the design upgrade and the cogeneration are shown to reduce the GTHTR300 cost of electricity to under 2.7 cent/kW h.
Sato, Hiroyuki; Yan, X.; Sumita, Junya; Terada, Atsuhiko; Tachibana, Yukio
Journal of Nuclear Engineering and Radiation Science, 2(3), p.031010_1 - 031010_6, 2016/07
This paper explains the outline of HTTR demonstration program with a plant concept of the heat application system directed at establishing an HTGR cogeneration system with 950C reactor outlet temperature for production of power and hydrogen as recommended by the task force. Commercial deployment strategy including a development plan for the helium gas turbine is also presented.
Sato, Hiroyuki; Ohashi, Hirofumi; Nakagawa, Shigeaki; Tachibana, Yukio; Kunitomi, Kazuhiko
Progress in Nuclear Energy, 82, p.46 - 52, 2015/07
Times Cited Count：12 Percentile：70.16(Nuclear Science & Technology)Safety requirements and design considerations for a HTGR hydrogen production system by IS process are examined. Requirements in order to construct hydrogen production plants under conventional chemical plant regulation are identified. In addition, safety requirements for the collocation of the nuclear facility and hydrogen production plant utilizing IS process are investigated. Furthermore, design considerations to comply with the requirements are suggested and the technical feasibility of the design considerations is evaluated. The evaluation results clarified that design considerations suggested for coupling IS plant to HTGR are reasonably practicable.
Sato, Hiroyuki; Sumita, Junya; Terada, Atsuhiko; Ohashi, Hirofumi; Yan, X.; Nishihara, Tetsuo; Tachibana, Yukio; Inagaki, Yoshiyuki
Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 8 Pages, 2015/05
This paper explains the outline and schedule of HTTR demonstration program with a plant concept of the heat application system directed at establishing an HTGR cogeneration system with 950C reactor outlet temperature for production of power and hydrogen as recommended by the task force.
Ando, Masaki; Nakahara, Yukio; Tsuda, Shuichi; Yoshida, Tadayoshi; Matsuda, Norihiro; Takahashi, Fumiaki; Mikami, Satoshi; Kinouchi, Nobuyuki; Sato, Tetsuro*; Tanigaki, Minoru*; et al.
Journal of Environmental Radioactivity, 139, p.266 - 280, 2015/01
Times Cited Count：53 Percentile：83(Environmental Sciences)A series of car-borne surveys using the KURAMA and KURAMA-II systems was conducted in a wide area in eastern Japan from June 2011 to December 2012 to evaluate the distribution of air dose rates around the Fukushima Dai-ichi Nuclear Power Plant, and to determine the time-dependent trend of decrease in air dose rates. An automated data processing system was established, which enables analyses of large amounts of data obtained with the use of about 100 units of the measurement system in a short time. The initial data for studying the migration status of radioactive cesium was obtained in the first survey, followed by the other car-borne surveys having wider measurement ranges. Comparing the measured air dose rates obtained in each survey, it was found that the decreasing trend of air dose rates measured through car-borne surveys was larger than that expected from the physical decay of radioactive cesium and that measured using NaI (Tl) survey meters around the road.
Sato, Hiroyuki; Ohashi, Hirofumi; Tachibana, Yukio; Kunitomi, Kazuhiko; Ogawa, Masuro
Journal of Nuclear Science and Technology, 51(11-12), p.1317 - 1323, 2014/11
Times Cited Count：7 Percentile：48.18(Nuclear Science & Technology)Transient analyses are presented of temperature behavior of reactor during loss-of-coolant accident with scram. The influence of reactor thermal properties, operating power density, geometry of active core and selection of fuel type on the capability of decay heat removal against the accident are studied. It is shown that the reactor design envelope is fully determined by the key parameters. The range of the envelope is shown to enlarge considerably by selecting high refractory fuel. High temperature gas-cooled reactor (HTGR), a graphite-moderated reactor with TRISO coated fuel particle, is the primary candidate which can fulfill the requirement to the design concept of nuclear reactor independent of coolant for decay heat removal.
Sato, Hiroyuki; Ohashi, Hirofumi; Nakagawa, Shigeaki; Tachibana, Yukio; Kunitomi, Kazuhiko
Journal of Nuclear Science and Technology, 51(11-12), p.1324 - 1335, 2014/11
Times Cited Count：6 Percentile：42.7(Nuclear Science & Technology)A qualification of the thermal hydraulic system code RELAP5 code is conducted for the analysis of helically-coiled heat exchangers used in high temperature environment. The experimental data obtained from the HTTR are utilized to compare with calculated data by RELAP5-based model with built-in closure models. A set of closure model is also suggested considering the heat transfer enhancement by thermal radiation based on the past separate effect test data and validated against the measured data. In addition, the modified RELAP5 code is tested for the analysis of the HTTR-IS system. The comparison of calculated and measure data with steady state operation showed that the prediction temperature with the suggested model generally agreed well. As a conclusion of the present study, the use of thermal hydraulic system code with the suggested closure model is acceptable for the analysis of the IHX in HTGR nuclear hydrogen production systems in the safety evaluation.
Ohashi, Hirofumi; Sato, Hiroyuki; Nakagawa, Shigeaki; Tachibana, Yukio; Nishihara, Tetsuo; Yan, X.; Sakaba, Nariaki; Kunitomi, Kazuhiko
Proceedings of 7th International Topical Meeting on High Temperature Reactor Technology (HTR 2014) (USB Flash Drive), 10 Pages, 2014/10
The research committee on Safety requirements for HTGR design was established in 2013 under the Atomic Energy Society of Japan to develop the draft safety requirements for the design of commercial High Temperature Gas-cooled Reactor (HTGR), which incorporate the HTGR safety features demonstrated using the High Temperature Engineering Test Reactor (HTTR), lessons learned from the accident of Fukushima Daiichi Nuclear Power Station and requirements for the integration of the hydrogen production plants. The safety design approach for the commercial HTGR which is a basement of the safety requirements is determined prior to the development of the safety requirements. This paper describes the main topics of the research committee, the safety design approaches and the safety functions of the commercial HTGR determined in the research committee.
Yan, X.; Sato, Hiroyuki; Kamiji, Yu; Imai, Yoshiyuki; Terada, Atsuhiko; Tachibana, Yukio; Kunitomi, Kazuhiko
Proceedings of 7th International Topical Meeting on High Temperature Reactor Technology (HTR 2014) (USB Flash Drive), 7 Pages, 2014/10
The latest design upgrade has incorporated several major technological advances made in the past 10 years to GTHTR300. These advances have enabled raising the design basis reactor outlet temperature to 950C and increasing power generating efficiency by nearly 5% point. Further implementation of desalination cogeneration is made through employing a newly-proposed multi-stage flash process. Through efficient waste heat recovery of the reactor gas turbine cycle, a large cost credit is obtained against the conventionally produced water prices. Together, the design upgrade and the cogeneration result in reducing the GTHTR300 cost of electricity to under 2.7 US cent per KWh.
Sato, Hiroyuki; Ohashi, Hirofumi; Yan, X.; Kubo, Shinji; Nishihara, Tetsuo; Tachibana, Yukio; Inagaki, Yoshiyuki
JAEA-Technology 2014-031, 30 Pages, 2014/09
JAEA-Technology-2014-031.pdf:17.95MB
In the present study, identification of test items to be validated in the HTTR demonstration test to accomplish a formulation of safety requirement and design consideration for coupling a hydrogen production plant to a nuclear facility as well as confirmation of overall performance of helium gas turbine system. In addition, a plant concept for the heat utilization system to be connected with the HTTR is clarified.
Sato, Hiroyuki; Yan, X.; Tachibana, Yukio; Kunitomi, Kazuhiko
Nuclear Engineering and Design, 275, p.190 - 196, 2014/08
Times Cited Count：22 Percentile：85.07(Nuclear Science & Technology)Three major improvements have since been made to further increase efficiency for the GTHTR300. First, the cycle parameters are upgraded by utilizing the newly-acquired design data including those from component tests. Next, the core design is optimized to raise the reactor outlet coolant temperature from the baseline of 850C to the level of 950C demonstrated on the long-term test reactor operation. Finally, an advanced type of turbine blade material that has only recently entered in commercial service in aircraft engine is found to be useable for this design to realize a turbine inlet temperature of 950C without requiring blade cooling. These design improvements result in a nearly 5% gain in overall plant efficiency and enable the GTHTR300 to break the 50% efficiency barrier of nuclear plant while using only the existing technologies.
Ohashi, Hirofumi; Sato, Hiroyuki; Tachibana, Yukio; Kunitomi, Kazuhiko; Ogawa, Masuro
Nuclear Engineering and Design, 271, p.537 - 544, 2014/05
Times Cited Count：3 Percentile：23.79(Nuclear Science & Technology)The concept of the Naturally-safe HTGR is that the release of radioactive materials is kept at very low level and no harmful effect on people and the environment is ensured by only physical phenomena without any engineered safety features. In this study, the CO concentration and the heat generated by graphite oxidation inside the circular tube were evaluated parametrically using a steady-state one-dimensional model to confirm the feasibility of the Naturally-safe HTGR at a severe condition of the air ingress accident (i.e., a massive air ingress by simultaneous rupture of two primary pipes). It was confirmed that the CO concentration at the outlet of coolant channel can be maintained below the explosion limit due to the reaction with oxygen in the air, and the reaction heat can be removed with the decay heat by physical phenomena under certain conditions of the coolant channel geometry without any engineered safety features.
Sato, Hiroyuki; Ohashi, Hirofumi; Tachibana, Yukio; Kunitomi, Kazuhiko; Ogawa, Masuro
Nuclear Engineering and Design, 271, p.530 - 536, 2014/05
Times Cited Count：3 Percentile：16.35(Nuclear Science & Technology)Design envelope of prismatic High Temperature Gas-cooled Reactors in terms of core heat removal capability under depressurized loss-of-forced-circulation accidents without operating active or passive decay heat removal systems are investigated. Lumped element models consist of core, reactor pressure vessel and cavity wall are presented in order to evaluate transient response of core temperature. Parametric calculations changing the core height, initial core temperature, thickness of side reflector, cavity size and peaking factor are performed. A series of calculation provides relationships of core radius to average power density and reactor thermal power which can remove the heat in core without reliance on specific design features. The results clarified the design envelope for the Naturally Safe HTGR in terms of core decay heat removal.
