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Journal Articles

Feasibility study of new applications at the high-temperature gas-cooled reactor

Ho, H. Q.; Honda, Yuki*; Hamamoto, Shimpei; Ishii, Toshiaki; Takada, Shoji; Fujimoto, Nozomu*; Ishitsuka, Etsuo

Proceedings of 9th International Topical Meeting on High Temperature Reactor Technology (HTR 2018) (USB Flash Drive), 6 Pages, 2018/10

Journal Articles

Current R&D status of thermochemical water splitting hydrogen production iodine-sulfur process in Japan Atomic Energy Agency, 1; Hydrogen production test and component development

Takegami, Hiroaki; Noguchi, Hiroki; Tanaka, Nobuyuki; Iwatsuki, Jin; Kamiji, Yu; Kasahara, Seiji; Imai, Yoshiyuki; Terada, Atsuhiko; Kubo, Shinji

Proceedings of 9th International Topical Meeting on High Temperature Reactor Technology (HTR 2018) (USB Flash Drive), 7 Pages, 2018/10

Japan Atomic Energy Agency (JAEA) has been conducting R&D on the thermochemical iodine-sulfur (IS) process for nuclear-powered hydrogen production. The IS process is one of the promising candidates of heat application of the high-temperature gas-cooled reactors. JAEA fabricated main chemical reactors made of industrial structural materials and confirmed their integrity in practical corrosive environments in the IS process. Based on the results of these confirmation tests, JAEA have constructed a 100 NL/h-H$$_{2}$$-scale test facility made of industrial structural materials. This report will present an outline and results of hydrogen production tests and reliability improvements of operation stability and components, such as development of a strength estimation method for heat-resistant and corrosion-resistant ceramics components made of silicon carbide.

Journal Articles

Current R&D status of thermochemical water splitting hydrogen production iodine-sulfur process in Japan Atomic Energy Agency, 2; Reliability improvements of corrosion-resistant equipment

Kamiji, Yu; Noguchi, Hiroki; Takegami, Hiroaki; Tanaka, Nobuyuki; Iwatsuki, Jin; Kasahara, Seiji; Kubo, Shinji

Proceedings of 9th International Topical Meeting on High Temperature Reactor Technology (HTR 2018) (USB Flash Drive), 7 Pages, 2018/10

Japan Atomic Energy Agency (JAEA) has been conducting R&D on the thermochemical iodine-sulfur (IS) process for nuclear-powered hydrogen production. The IS process is one of the promising candidates of heat application of the HTGR. JAEA achieved continuous hydrogen production for one week with a hydrogen production rate of 30 NL/h by using a test apparatus made of glass and fluororesin material. Subsequently, JAEA fabricated main chemical reactors made of industrial materials and confirmed their integrity in corrosive environments in the IS process. Based on the results, JAEA has constructed a 100 NL/h-H$$_{2}$$-scale test facility made of industrial materials; one of the important materials is the glass-lined steel for corrosion resistant components such as vessels, pipes and protective sheaths of sensors. This report will present technical matters to improve reliability of the glass-lined protective sheaths of thermocouple. In addition, results of quality confirmation will be presented, which are stress analyses for the glass layer by FEM, tests for heat cycle, bending load and corrosion.

Journal Articles

Conceptual design of the steam reforming system for hydrogen production connected to HTTR

Iwatsuki, Jin; Ohashi, Hirofumi; Yan, X.

Proceedings of 9th International Topical Meeting on High Temperature Reactor Technology (HTR 2018) (USB Flash Drive), 7 Pages, 2018/10

The HTTR is a 30MW, 950$$^{circ}$$C high temperature engineering test reactor built and operated on the site of the Oarai Research & Development Center of Japan Atomic Energy Agency (JAEA). In the framework of the HTTR project, JAEA has been conducting a research and development on the steam reforming system (CH$$_{4}$$ + H$$_{2}$$O = 3H$$_{2}$$ + CO). JAEA had constructed a mock-up test facility in 2002, and investigated transient behavior of the hydrogen production system and established system controllability. Based on the results and experience of above, the conceptual design of steam reforming system for hydrogen production connected to HTTR has been studied. The system condition was optimized considering the HTTR specification and the experience on the construction and the operation of the mock-up test facility. The hydrogen production system is heated with about 0.2MW transported from the HTTR to the hydrogen system via a helium loop. The system produces about 70 Nm$$^{3}$$/h hydrogen.

Journal Articles

Study on Pu-burner high temperature gas-cooled reactor in Japan; Introduction scenario

Fukaya, Yuji; Goto, Minoru; Ueta, Shohei; Tachibana, Yukio; Okamoto, Koji*

Proceedings of 9th International Topical Meeting on High Temperature Reactor Technology (HTR 2018) (USB Flash Drive), 9 Pages, 2018/10

The research on introduction scenarios of Pu-burner High Temperature Gas-cooled Reactor (HTGR) of Japan has been performed based on the "Long-term Energy Supply and Demand Outlook" released by the Ministry of Economy, Trade and Industry (METI) of Japan in 2015. In the perspective, the electricity generation capacity of nuclear power generation reduces from 50 GWe (peak around 2010) to 30 GWe in 2030. To maintain the capacity, light water reactors (LWRs) should be introduced from 2025 to 2030. After 2030, HTGRs, which are superior to LWRs from the viewpoint of safety and economy, will be introduced to fill the capacity and incinerate plutonium. We assumed introduction of U fueled HTGR as well. The Pu-burner reactor will be introduced with the priority to incinerate separated plutonium by reprocessing. Moreover, we also evaluated hydrogen generation and its effect on CO$$_{2}$$ reduction. As a result, effective plutonium incineration and CO$$_{2}$$ reduction effect are confirmed.

Journal Articles

Conceptual design study of a high performance commercial HTGR

Fukaya, Yuji; Mizuta, Naoki; Goto, Minoru; Ohashi, Hirofumi; Yan, X.

Proceedings of 9th International Topical Meeting on High Temperature Reactor Technology (HTR 2018) (USB Flash Drive), 7 Pages, 2018/10

Conceptual design study of a high performance commercial HTGR has been performed at target output of 165MWt. Requirements for the HTGR are small-sized vessel for transportation, durability of vessel to irradiation damage, fuel reloading scheme to shorten the duration of reloading, low pressure drop fuel element, a small number of fuel enrichments, and so on. To satisfy the requirement, we investigated the core configuration, shielding and reflector configuration, fuel reloading scheme. As a result, we completed the design with the vessel diameter of 4.5m, which can be transported by any means, such as, by load, rail, ship, and air plane, and high load factor over 90%.

Journal Articles

Conceptual plant system design study of an experimental HTGR upgraded from HTTR

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.

Journal Articles

Conceptual study of an experimental HTGR upgraded from HTTR

Goto, Minoru; Fukaya, Yuji; Mizuta, Naoki; Inaba, Yoshitomo; Ohashi, Hirofumi; Yan, X.

Proceedings of 9th International Topical Meeting on High Temperature Reactor Technology (HTR 2018) (USB Flash Drive), 6 Pages, 2018/10

The HTTR (High Temperature engineering Test Reactor) constructed at JAEA-Oarai R&D center is a block-type experimental HTGR (High Temperature Gas-cooled Reactor) with 30 MW thermal power. It attained the first criticality at 1998 and has yielded very useful data for future HTGR design. Although the HTTR was designed very conservatively because the HTTR is the first HTGR for Japan, future HTGRs can be designed with a reasonable conservativeness based on the HTTR data. Additionally, it is possible to enhance the performance of the reactor core by improving the design and introducing new technologies. This paper describes a concept of an experimental HTGR that is upgraded from the HTTR by the reasonable conservativeness, the design improvement and the new technology introduction.

Journal Articles

Study on Pu-burner high temperature gas-cooled reactor in Japan; Design study of fuel and reactor core

Goto, Minoru; Aihara, Jun; Inaba, Yoshitomo; Ueta, Shohei; Fukaya, Yuji; Okamoto, Koji*

Proceedings of 9th International Topical Meeting on High Temperature Reactor Technology (HTR 2018) (USB Flash Drive), 6 Pages, 2018/10

JAEA has conducted design studies of a Pu-burner HTGR. The Pu-burner HTGR incinerates Pu by fission, and hence a high burn-up is required for the efficient incineration. In the fuel design, a thin ZrC layer, which acts as an oxygen getter and suppresses the internal pressure, was coated on the fuel kernel to prevent the CFP failure at the high burn-up. A stress analysis of the SiC layer, which acts as a pressure vessel for the CFP, was performed for with consideration of the depression effect due to the ZrC layer. As a result, the CFP failure fraction at high burn-up of 500 GWd/t satisfied the target value. In the reactor core design, an axial fuel shuffling was employed to attain the high burn-up, and the nuclear burn-up calculations with the whole core model and the fuel temperature calculations were performed. As a result, the nuclear characteristics, which are the shutdown margin and the temperature coefficient of reactivity, and the fuel temperature satisfied their target values.

Journal Articles

Study on source of radioactive material in primary coolant of HTTR

Ishii, Toshiaki; Shimazaki, Yosuke; Ono, Masato; Fujiwara, Yusuke; Ishitsuka, Etsuo; Hamamoto, Shimpei

Proceedings of 9th International Topical Meeting on High Temperature Reactor Technology (HTR 2018) (USB Flash Drive), 3 Pages, 2018/10

Journal Articles

Study on Pu-burner high temperature gas-cooled reactor in Japan; Test and characterization for ZrC coating

Ueta, Shohei; Aihara, Jun; Mizuta, Naoki; Goto, Minoru; Fukaya, Yuji; Tachibana, Yukio; Okamoto, Koji*

Proceedings of 9th International Topical Meeting on High Temperature Reactor Technology (HTR 2018) (USB Flash Drive), 7 Pages, 2018/10

The security and safety fuel (3S-TRISO fuel) employs the coated fuel particle with a fuel kernel made of plutonium dioxide (PuO$$_{2}$$) and yttria stabilized zirconia (YSZ) as an inert matrix. Especially, a zirconium carbide (ZrC) coating is one of key technologies of the 3S-TRISO, which performs as an oxygen getter to reduce the fuel failure due to internal pressure during the irradiation. R&Ds on ZrC coating directly on the dummy CeO$$_{2}$$-YSZ kernel have been carried in the Japanese fiscal year 2017. As results of ZrC coating tests by the bromide chemical vapor deposition process, stoichiometric ZrC coatings with 3 - 18 microns of thicknesses were obtained with 0.1 kg of particle loading weight.

Journal Articles

Investigation of irradiated properties of extended burnup TRISO fuel

Shaimerdenov, A.*; Gizatulin, S.*; Kenzhin, Y.*; Dyussambayev, D.*; Ueta, Shohei; Aihara, Jun; Shibata, Taiju

Proceedings of 9th International Topical Meeting on High Temperature Reactor Technology (HTR 2018) (USB Flash Drive), 6 Pages, 2018/10

The Institute of Nuclear Physics of the Republic of Kazakhstan (INP) conducts an irradiation test and post-irradiation examinations (PIEs) of the high-temperature gas-cooled reactor (HTGR) fuel and materials to develop the extend burnup fuel up to 100 GWd/t-U collaboratively with the Japan Atomic Energy Agency (JAEA) under projects in a frame of the International Science and Technology Centre (ISTC). Cylindrical fuel compact specimens consisting of newly-designed TRISO (tri-structural isotropic) coated fuel particles and a matrix made of graphite material were manufactured in Japan. An irradiation test of the fuel specimens using a helium-gas swept capsule designed and constructed in the INP has been performed up to 100 GWd/t-U in the WWR-K research reactor by April 2015. In the next stage, PIEs with the irradiated fuel specimens have been started in February 2017 as a new ISTC project. Several PIE technologies by non-destructive and destructive techniques with irradiated fuel compacts were developed by the INP. This report presents the developed technologies and interim results of the PIE for high burning TRISO fuel.

Journal Articles

Numerical evaluation on fluctuation absorption characteristics based on nuclear heat supply fluctuation test using HTTR

Takada, Shoji; Honda, Uki*; Inaba, Yoshitomo; Sekita, Kenji; Nemoto, Takahiro; Tochio, Daisuke; Ishii, Toshiaki; Sato, Hiroyuki; Nakagawa, Shigeaki; Sawa, Kazuhiro*

Proceedings of 9th International Topical Meeting on High Temperature Reactor Technology (HTR 2018) (USB Flash Drive), 7 Pages, 2018/10

Nuclear heat utilization systems connected to HTGRs will be designed on the basis of non-nuclear grade standards for easy entry of chemical plant companies, requiring reactor operations to continue even if abnormal events occur in the systems. The inventory control is considered as one of candidate methods to control reactor power for load following operation for siting close to demand area, in which the primary gas pressure is varied while keeping the reactor inlet and outlet coolant temperatures constant. Numerical investigation was carried out based on the results of nuclear heat supply fluctuation tests using HTTR by non-nuclear heating operation to focus on the temperature transient of the reactor core bottom structure by imposing stepwise fluctuation on the reactor inlet temperature under different primary gas pressures below 120C. As a result, it was emerged that the fluctuation absorption characteristics are not deteriorated by lowering pressure. It was also emerged that the reactor outlet temperature did not reach the scram level by increasing the reactor inlet temperature 10 C stepwise at 80% of the rated power as same with the full power case.

Journal Articles

Research and development for safety and licensing of HTGR cogeneration system

Sato, Hiroyuki; Ohashi, Hirofumi; Yan, X.

Proceedings of 9th International Topical Meeting on High Temperature Reactor Technology (HTR 2018) (USB Flash Drive), 9 Pages, 2018/10

Japan Atomic Energy Agency has been conducting research and development with a central focus on the utilization of HTTR, the first HTGR in Japan, towards the realization of industrial use of nuclear heat. On the basis of licensing experience through the HTTR construction, JAEA initiated an activity to establish an international safety standard for licensing of commercial HTGR cogeneration systems fully taking into account safety features of HTGRs. This paper explains a roadmap towards licensing of commercial HTGR cogeneration systems. A test plan using the HTTR to support the establishment of safety standards and safety analysis methods is also presented.

Journal Articles

RELAP5 modeling of the HTTR-GT/H$$_{2}$$ secondary system and turbomachinery

Humrickhouse, P. W.*; Sato, Hiroyuki; Imai, Yoshiyuki; Sumita, Junya; Yan, X.

Proceedings of 9th International Topical Meeting on High Temperature Reactor Technology (HTR 2018) (USB Flash Drive), 9 Pages, 2018/10

This work describes the development of a RELAP5-3D model of the HTTR-GT/H$$_{2}$$ plant secondary system. The RELAP5-3D model presently includes detailed models of several of the heat exchangers in the secondary system as well as the turbomachinery, which includes two compressors and two gas turbines connected to a common shaft and motor. The predictions of the model agreed well to design parameters in both sole power generation and hydrogen co-generation modes in most instances. Both the turbomachinery and heat exchanger models rely on extensive customization via RELAP5-3D control variables, and these implementations are outlined in detail. Potential improvements to the RELAP5-3D turbine model are discussed.

Journal Articles

Recent advances in the GIF very high temperature reactor system

F$"u$tterer, M. A.*; Li, F.*; Gougar, H.*; Edwards, L.*; Pouchon, M. A.*; Kim, M. H.*; Carr$'e$, F.*; Sato, Hiroyuki

Proceedings of 9th International Topical Meeting on High Temperature Reactor Technology (HTR 2018) (USB Flash Drive), 12 Pages, 2018/10

This paper provides an update on the international effort in the development of the VHTR system pursued through international collaboration between 8 countries in the GIF and an outlook on future R&D. The versatility of the VHTR enables it to be designed with inherent safety characteristics and optimized for both electric and non-electric applications, in particular for cogeneration of heat and power. Recent highlights from the four currently active GIF VHTR R&D projects are provided and placed into the context of the related national programs. Based on VHTR's relatively high technology readiness level, orientations for future R&D are outlined and will contribute to further enhancing the system's market readiness level.

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