Progress in conceptual design of a pool-type sodium-cooled fast reactor in Japan

Kato, Atsushi; Kubo, Shigenobu; Chikazawa, Yoshitaka; Miyagawa, Takayuki*; Uchita, Masato*; Suzuno, Tetsuji*; Endo, Junji*; Kubo, Koji*; Murakami, Hisatomo*; Uzawa, Masayuki*; et al.

Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Sustainable Clean Energy for the Future (FR22) (Internet), 11 Pages, 2022/04

The authors are carrying out conceptual design studies for a pool-type sodium-cooled fast reactor. There are main challenges such as measures against severe earthquake in Japan, thermal hydraulic in a reactor vessel (RV), a decay heat removal system design. When the JP-pool SFR of 650 MWe is installed in Japan, it shall be designed against the severe seismic conditions. Additionally, a newly three-dimensional seismic isolation system is under development.

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.

Generation of He clusters via neutron-He absorption reaction toward visualization of full velocity field in quantum turbulence

Matsushita, Taku*; Sonnenschein, V.*; Guo, W.*; Hayashida, Hirotoshi*; Hiroi, Kosuke; Hirota, Katsuya*; Iguchi, Tetsuo*; Ito, Daisuke*; Kitaguchi, Masaaki*; Kiyanagi, Yoshiaki*; et al.

Journal of Low Temperature Physics, 196(1-2), p.275 - 282, 2019/07

https://doi.org/10.1007/s10909-018-02112-3

Negative muon capture ratios for nitrogen oxide molecules

Ninomiya, Kazuhiko*; Ito, Takashi; Higemoto, Wataru; Kawamura, Naritoshi*; Strasser, P.*; Nagatomo, Takashi*; Shimomura, Koichiro*; Miyake, Yasuhiro*; Kita, Makoto*; Shinohara, Atsushi*; et al.

Journal of Radioanalytical and Nuclear Chemistry, 319(3), p.767 - 773, 2019/03

https://doi.org/10.1007/s10967-018-6366-3

Oxidation characteristics of lead-alloy coolants in air ingress accident

Kondo, Masatoshi*; Okubo, Nariaki; Irisawa, Eriko; Komatsu, Atsushi; Ishikawa, Norito; Tanaka, Teruya*

Energy Procedia, 131, p.386 - 394, 2017/12

https://doi.org/10.1016/j.egypro.2017.09.473

The chemical behaviors of lead (Pb) based coolants in the air ingress accident of fast reactors were investigated by means of the thermodynamic considerations and the static oxidation experiments for Pb alloys at various chemical compositions. The results of the static oxidation tests for lead-bismuth (Pb-Bi) alloys indicated that Pb was depleted from the alloy due to the preferential formation of PbO in air at 773K. Pb-Bi oxide and BiO were formed after the enrichment of Bi in the alloys due to the Pb depletion. The oxidation rates of the alloys were much larger than that of the steels, and became larger with higher Pb concentration in the alloys. The compatibility of Pb-Bi alloys with stainless steel was worse when the Pb concentration in the alloys became low, since the dissolution type corrosion was promoted by the Bi composition in the alloy. The Pb-Li alloys were oxidized as they formed LiPbO and LiCO. Then, Li was depleted from the alloy.

Proposal of maintenance management of nuclear power plants at R&D stage by taking account of their features

Takaya, Shigeru; Chikazawa, Yoshitaka; Hayashida, Kiichi; Tagawa, Akihiro; Kubo, Shigenobu; Yamashita, Atsushi

Hozengaku, 15(4), p.71 - 78, 2017/01

A maintenance management suitable to nuclear power plants (NPP) at R&D stage was discussed. Objectives of maintenance management of NPP at R&D stage was first clarified. Next, applicability of codes for maintenance management of commercial NPP to NPP at R&D stage was discussed. Then, requirements and consideration for maintenance management of NPP at R&D stage was proposed. Finally, an example that the proposal was applied to setting maintenance program of sodium-cooled fast reactor was presented.

Maintenance management of nuclear power reactors at the stage of research and development

Takaya, Shigeru; Chikazawa, Yoshitaka; Hayashida, Kiichi; Tagawa, Akihiro; Kubo, Shigenobu; Yamashita, Atsushi

JAEA-Research 2016-006, 66 Pages, 2016/07

JAEA-Research-2016-006.pdf:3.4MB

A maintenance management required to nuclear power reactors at the R&D stage was discussed. It is the most important to ensure safety of nuclear power plants by taking account of characteristics of nuclear power reactors at the R&D stage. In addition, it is needed to establish a system of maintenance management technologies suitable for reactor types. In this report, objectives of maintenance management of nuclear power reactors at the R&D stage was clarified. Next, requirements and consideration for maintenance management was discussed according to the objectives. "Codes for maintenance management of nuclear power plants" and "Guides for maintenance management of nuclear power plants" were refereed in the discussion. Then, a draft of codes for maintenance management of nuclear power plants at the R&D stage were newly proposed. Finally, an example that the draft codes were applied to components containing sodium, typical components of sodium-cooled fast reactor, was presented.

Dismantlement of large fusion experimental device JT-60U

Ikeda, Yoshitaka; Okano, Fuminori; Sakasai, Akira; Hanada, Masaya; Akino, Noboru; Ichige, Hisashi; Kaminaga, Atsushi; Kiyono, Kimihiro; Kubo, Hirotaka; Kobayashi, Kazuhiro; et al.

Nihon Genshiryoku Gakkai Wabun Rombunshi, 13(4), p.167 - 178, 2014/12

https://doi.org/10.3327/taesj.J14.007

The JT-60U torus was disassembled so as to newly install the superconducting tokamak JT-60SA torus. The JT-60U used the deuterium for 18 years, so the disassembly project of the JT-60U was the first disassembly experience of a fusion device with radioactivation in Japan. All disassembly components were stored with recording the data such as dose rate, weight and kind of material, so as to apply the clearance level regulation in future. The lessons learned from the disassembly project indicated that the cutting technologies and storage management of disassembly components were the key factors to conduct the disassembly project in an efficient way. After completing the disassembly project, efforts have been made to analyze the data for characterizing disassembly activities, so as to contribute the estimation of manpower needs and the radioactivation of the disassembly components on other fusion devices.

Status quo of the injector for the IFMIF/EVEDA prototype accelerator

Shinto, Katsuhiro; Ichikawa, Masahiro; Takahashi, Yasuyuki*; Kubo, Takashi*; Tsutsumi, Kazuyoshi; Kikuchi, Takayuki; Kasugai, Atsushi; Sugimoto, Masayoshi; Gobin, R.*; Girardot, P.*; et al.

Proceedings of 11th Annual Meeting of Particle Accelerator Society of Japan (Internet), p.1009 - 1012, 2014/10

http://www.pasj.jp/web_publish/pasj2014/proceedings/PDF/SUP0/SUP025.pdf

The prototype accelerator is being developed as an engineering validation for the International Fusion Materials Irradiation Facility (IFMIF) equipped with an accelerator-driven-type neutron source for developing fusion reactor materials. This prototype accelerator is a deuteron linear accelerator consisting of an injector, an RFQ, a superconducting linac and their auxiliaries. It aims to produce a CW D beam with the energy and current of 9 MeV/125 mA. The injector test was completed at CEA/Saclay in 2012 for producing a CW H beam and a CW D beam with the energy and current of 100 keV/140 mA. After the beam test at CEA/Saclay, the injector was transported to the International Fusion Energy Research Centre (IFERC) located in Rokkasho, Aomori, Japan. In the end of 2013, installation of the injector was started at IFERC for the injector beam test beginning from summer 2014 in order to obtain better beam qualities to be satisfied with the injection and acceleration of the following accelerators. In this paper, some results of the injector beam test performed at CEA/Saclay and the status quo of the installation of the injector at IFERC are presented.

Performance evaluation on secondary sodium fire measures in JSFR

Chikazawa, Yoshitaka; Kato, Atsushi; Yamamoto, Tomohiko; Kubo, Shigenobu; Iwasaki, Mikinori*; Hara, Hiroyuki*; Shimakawa, Yoshio*; Sakaba, Hiroshi*

Proceedings of 2014 International Congress on the Advances in Nuclear Power Plants (ICAPP 2014) (CD-ROM), p.523 - 530, 2014/04

JSFR adopts double boundary for all sodium components. However, design measures are investigated for the secondary sodium fire inside the reactor building, which might be assumed as design extension conditions (DECs). Candidates of sodium fire measures in the secondary sodium systems such as sodium drain, nitrogen injection, pressure release valve, catch pan and leak sodium drain system have been compared from the view point of safety. Wide range of sodium fires in the steam generator room have been analyzed evaluating performances of the candidate sodium fire measures.