Study of fabrication of SiC-matrixed fuel compact for HTGR

Kawano, Takahiro*; Mizuta, Naoki; Ueta, Shohei; Tachibana, Yukio; Yoshida, Katsumi*

JAEA-Technology 2023-014, 37 Pages, 2023/08

JAEA-Technology-2023-014.pdf:2.35MB

Fuel compact for High Temperature Gas-cooled Reactor (HTGR) is fabricated by calcinating a matrix consisting of graphite and binder with the coated fuel particle. The SiC-matrixed fuel compact uses a new matrix made of silicon carbide (SiC) replacing the conventional graphite. Applying the SiC-matrixed fuel compact for HTGRs is expected to improve their performance such as power densities. In this study, the sintering conditions for applying SiC as the matrix of fuel compacts for HTGR are selected, and the density and thermal conductivity of the prototype SiC are measured.

Development plan for coupling technology between high temperature gas-cooled reactor HTTR and hydrogen production facility, 1; Overview of the HTTR heat application test plan to establish high safety coupling technology

Nomoto, Yasunobu; Mizuta, Naoki; Morita, Keisuke; Aoki, Takeshi; Okita, Shoichiro; Ishii, Katsunori; Kurahayashi, Kaoru; Yasuda, Takanori; Tanaka, Masato; Isaka, Kazuyoshi; et al.

Proceedings of 30th International Conference on Nuclear Engineering (ICONE30) (Internet), 7 Pages, 2023/05

Document collection of the Special Committee on HTTR Heat Application Test

Aoki, Takeshi; Shimizu, Atsushi; Iigaki, Kazuhiko; Okita, Shoichiro; Hasegawa, Takeshi; Mizuta, Naoki; Sato, Hiroyuki; Sakaba, Nariaki

JAEA-Review 2022-016, 193 Pages, 2022/08

JAEA-Review-2022-016.pdf:42.06MB

Aiming to realize a massive, cost-effective and carbon-free hydrogen production technology utilizing a high temperature gas cooled reactor (HTGR), Japan Atomic Energy Agency (JAEA) is planning a HTTR heat application test producing hydrogen with High Temperature Engineering Test Reactor (HTTR) achieved 950C of the highest reactor outlet coolant temperature in the world. In the HTTR heat application test, it is required to establish its safety design realizing highly safe connection of a HTGR and a hydrogen production plant by the Nuclear Regulation Authority to obtain the permission of changes to reactor installation. However, installation of a system connecting the hydrogen production plant and a nuclear reactor, and its safety design has not been conducted so far in conventional nuclear power plant including HTTR in the world. A special committee on the HTTR heat application test, established under the HTGR Research and Development Center, considered a safety design philosophy for the HTTR heat application test based on an authorized safety design of HTTR in terms of conformity to the New Regulatory Requirements taking into account new considerable events as a result of the plant modification and connection of the hydrogen production plant. This report provides materials of the special committee such as technical reports, comments provided from committee members, response from JAEA for the comments and minutes of the committee.

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

https://doi.org/10.1016/C2018-0-05383-9

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.

Research and development on high burnup HTGR fuels in JAEA

Ueta, Shohei; Mizuta, Naoki; Sasaki, Koei; Sakaba, Nariaki; Ohashi, Hirofumi; Yan, X.

Mechanical Engineering Journal (Internet), 7(3), p.19-00571_1 - 19-00571_12, 2020/06

https://doi.org/10.1299/mej.19-00571

JAEA has been progressing to design HTGR fuels for not only small-type practical HTGRs but also VHTR proposed in GIF which can be utilized for various purposes with high-temperature heat at 750 to 950 C. To increase economy of these HTGRs, JAEA has been upgrading the design method for the HTGR fuel, which can maintain their integrities at the burnup of three to four times higher than that of the conventional HTTR fuel. Design principles and specifications of various concepts of the high burnup HTGR fuels designed by JAEA are reported. As the latest results on post-irradiation examinations of the high burnup HTGR fuel progressing in a framework of international collaboration with Kazakhstan, irradiation shrinkage rate of the fuel compact as a function of fast neutron fluence was obtained at around 100 GWd/t. Furthermore, the future R&Ds needed for the high burnup HTGR fuel are described based on these experimental results.

Study on plutonium burner high temperature gas-cooled reactor in Japan; Introduction scenario, reactor safety and fabrication tests of the 3S-TRISO fuel

Ueta, Shohei; Mizuta, Naoki; Fukaya, Yuji; Goto, Minoru; Tachibana, Yukio; Honda, Masaki*; Saiki, Yohei*; Takahashi, Masashi*; Ohira, Koichi*; Nakano, Masaaki*; et al.

Nuclear Engineering and Design, 357, p.110419_1 - 110419_10, 2020/02

https://doi.org/10.1016/j.nucengdes.2019.110419

The concept of a plutonium (Pu) burner HTGR is proposed to incarnate highly-effective Pu utilization by its inherent safety features. The security and safety fuel (3S-TRISO fuel) employs the coated fuel particle with a fuel kernel made of plutonium dioxide (PuO) and yttria stabilized zirconia (YSZ) as an inert matrix. This paper presents feasibility study of Pu burner HTGR and R&D on the 3S-TRISO fuel.

Research and development on high burnup HTGR fuels in JAEA

Ueta, Shohei; Mizuta, Naoki; Sasaki, Koei; Sakaba, Nariaki; Ohashi, Hirofumi; Yan, X.

Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 8 Pages, 2019/05

JAEA has been progressing to design HTGR fuels for not only small-type practical HTGRs but also VHTR proposed in GIF which can be utilized for various purposes with high-temperature heat at 750 to 950 C. To increase economy of these HTGRs, JAEA has been upgrading the design method for the HTGR fuel, which can maintain their integrities at the burnup of three to four times higher than that of the conventional HTTR fuel. Design principles and specifications of various concepts of the high burnup HTGR fuels designed by JAEA are reported. As the latest results on post-irradiation examinations of the high burnup HTGR fuel progressing in a framework of international collaboration with Kazakhstan, irradiation shrinkage rate of the fuel compact as a function of fast neutron fluence was obtained at around 100 GWd/thm. Furthermore, the future R&Ds needed for the high burnup HTGR fuel are described based on these experimental results.

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%.

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.

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) 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-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.