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Journal Articles

High temperature gas-cooled reactors

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$$^{circ}$$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.

Journal Articles

Design of HTTR-GT/H$$_{2}$$ test plant

Yan, 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:10 Percentile:87.34(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.

Journal Articles

Improvement of gross theory of beta-decay on single particle treatment

Koura, Hiroyuki; Yoshida, Tadashi*; Tachibana, Takahiro*; Chiba, Satoshi*

JAEA-Conf 2017-001, p.205 - 210, 2018/01

no abstracts in English

Journal Articles

Improvement of gross theory of beta-decay for application to nuclear data

Koura, Hiroyuki; Yoshida, Tadashi*; Tachibana, Takahiro*; Chiba, Satoshi*

EPJ Web of Conferences, 146, p.12003_1 - 12003_4, 2017/09

BB2016-0950.pdf:0.33MB

 Times Cited Count:1 Percentile:69.68

Journal Articles

HTTR-GT/H$$_{2}$$ test plant; System design

Yan, 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 950$$^{circ}$$C test reactor HTTR is intended as a system technology demonstrator for GTHTR300C plant design. More specifically the test plant of HTTR-GT/H$$_{2}$$ 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.

Journal Articles

GTHTR300 cost reduction through design upgrade and cogeneration

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:2 Percentile:25.63(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 950$$^{circ}$$C 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.

Journal Articles

HTTR demonstration program for nuclear cogeneration of hydrogen and electricity

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 950$$^{circ}$$C 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.

Journal Articles

Compilation for chart of the nuclides 2014; A Comprehensive decay data

Koura, Hiroyuki; Katakura, Junichi*; Tachibana, Takahiro*; Minato, Futoshi

JAEA-Conf 2015-003, p.147 - 152, 2016/03

A chart of the nuclides 2014 version is now preparing to be published from JAEA. This will be the latest successive version of the chart since 1977, and continues every (approximately) four years until 2010. These charts include decay data of isotopes as half-lives, decay modes, and some isomeric states. In addition, the periodic table of elements, fundamental physical constants, thermal neutron capture and fission cross sections are tabulated. The latest version is now compiled with recent experimental data until the end of June in 2014. In the compilation process, we improved in the following parts: (1) Neutron or proton-unbound nuclei in the lighter region. (2) Drawing the neutron and proton-drip lines, and a boundary line of $$beta$$-delayed neutron emission. (3) 1- or 2-proton-emission theoretical half-lives are added for unmeasured nuclei in addition to original three partial half-lives of $$alpha$$-decay, $$beta$$-decay and spontaneous fission. We compiled totally 3150 nuclides, which were experimental identified including 2914 life-measured nuclei. We will show overview of the chart with some statistics and examples.

Journal Articles

Safety design consideration for HTGR coupling with hydrogen production plant

Sato, Hiroyuki; Ohashi, Hirofumi; Nakagawa, Shigeaki; Tachibana, Yukio; Kunitomi, Kazuhiko

Progress in Nuclear Energy, 82, p.46 - 52, 2015/07

 Times Cited Count:8 Percentile:65.81(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.

Journal Articles

HTTR demonstration program for nuclear cogeneration of hydrogen and electricity

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 950$$^{circ}$$C reactor outlet temperature for production of power and hydrogen as recommended by the task force.

Journal Articles

Thermal analysis of heated cylinder simulating nuclear reactor during loss of coolant accident

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:55.77(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.

Journal Articles

Validation and application of thermal hydraulic system code for analysis of helically coiled heat exchanger in high-temperature environment

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:5 Percentile:43.74(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.

Journal Articles

Safety design approach for the development of safety requirements for design of commercial HTGR

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.

Journal Articles

GTHTR300 cost reduction through design upgrade and cogeneration

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 950$$^{circ}$$C 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.

JAEA Reports

HTTR demonstration test plan for industrial utilization of nuclear heat

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.

Journal Articles

GTHTR300; A Nuclear power plant design with 50% generating efficiency

Sato, Hiroyuki; Yan, X.; Tachibana, Yukio; Kunitomi, Kazuhiko

Nuclear Engineering and Design, 275, p.190 - 196, 2014/08

 Times Cited Count:18 Percentile:86.15(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 850$$^{circ}$$C to the level of 950$$^{circ}$$C 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 950$$^{circ}$$C 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.

Journal Articles

Feasibility study on Naturally Safe HTGR (NSHTR) for air ingress accident

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:28.59(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.

Journal Articles

Analysis of core heat removal capability under DLOFC accidents for HTGRs

Sato, Hiroyuki; Ohashi, Hirofumi; Tachibana, Yukio; Kunitomi, Kazuhiko; Ogawa, Masuro

Nuclear Engineering and Design, 271, p.530 - 536, 2014/05

 Times Cited Count:2 Percentile:19.81(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.

Journal Articles

Transient analysis of depressurized loss-of-forced circulation accident without scram in high temperature gas-cooled reactor

Sato, Hiroyuki; Yan, X.; Tachibana, Yukio; Kunitomi, Kazuhiko; Kato, Yukitaka*

Nuclear Technology, 185(3), p.227 - 238, 2014/03

 Times Cited Count:2 Percentile:19.81(Nuclear Science & Technology)

Transient response of HTGR to DLOFC combined with failure of all reactor trip systems is analyzed. The characteristic behavior of the reactor during the long-term conduction cooldown event is found to be shaped by several parameters that are usually not considered in the safety design of the HTGR. For example, the reactivity coefficient of temperature of the graphite moderator is found to be a critical parameter to determining the final settling temperature of the fuel. Furthermore, this study finds that the peak fuel temperature reached during this event is correlated strongly even to the initial core operating temperatures prior to the initiation of the transient event. These and other results of this study are expected to provide useful input to the development of enhanced safety design guidelines for commercial HTGR reactor in the aftermath of the Fukushima accident.

JAEA Reports

Conceptual design of small-sized HTGR system, 5; Safety design and preliminary safety analysis

Ohashi, Hirofumi; Sato, Hiroyuki; Tazawa, Yujiro; Aihara, Jun; Nomoto, Yasunobu; Imai, Yoshiyuki; Goto, Minoru; Isaka, Kazuyoshi; Tachibana, Yukio; Kunitomi, Kazuhiko

JAEA-Technology 2013-017, 71 Pages, 2014/02

JAEA-Technology-2013-017.pdf:3.64MB

Japan Atomic Energy Agency (JAEA) has started a conceptual design of a 50 MWt small-sized high temperature gas cooled reactor (HTGR) for steam supply and electricity generation (HTR50S). Though the safety design of HTR50S was determined based on that of the High Temperature Engineering Test Reactor (HTTR) for the early deployment of HTR50S, the shutdown cooling system, which is the forced cooling heat removal system, was categorized as non-safety class to optimize the protection to provide the highest level of safety that can reasonably be achieved, and the vessel cooling system, which is categorized as the safety class system, was designed as a passive safety features. The preliminary safety analysis of HTR50S for the rupture of co-axial hot gas duct in primary cooling system and the tube rupture of steam generator was conducted to confirm the adequacy of the safety design. It was confirmed that the analysis results satisfied the acceptance criteria.

126 (Records 1-20 displayed on this page)