Design of HTTR-GT/H 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

https://doi.org/10.1016/j.nucengdes.2017.10.009

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.

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

https://doi.org/10.1051/epjconf/201714612003

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

https://doi.org/10.1016/j.nucengdes.2016.02.023

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.

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

https://doi.org/10.11484/jaea-conf-2015-003

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 -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 -decay, -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.

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 950C reactor outlet temperature for production of power and hydrogen as recommended by the task force.

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

https://doi.org/10.1080/00223131.2014.928605

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.

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 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.

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

https://doi.org/10.1016/j.nucengdes.2014.05.004

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.

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

https://doi.org/10.1016/j.nucengdes.2013.12.028

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.

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.