Development of numerical evaluation method for heat transportation with sodium mist in the cover gas region of sodium-cooled fast reactor

Hayakawa, Satoshi*; Hagiwara, Hiroyuki*; Imamura, Akira*; Onoda, Yuichi; Tanaka, Masaaki; Nakamura, Hironori*

Proceedings of 13th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS13) (Internet), 8 Pages, 2024/11

In a sodium-cooled fast reactor, a cover gas region filled with argon gas is located above the sodium pool in the main vessel to prevent the hot sodium from contacting the structures. This region involves heat transportation by natural convection of the cover gas, radiation among liquid surface and structures, and sodium phase change between mist and vapor. In this study, the numerical evaluation method has been developed with a commercial CFD code, Fluent, incorporating the sodium mist transport and growth models, and the radiation scattering model. Simulations of a laboratory scale test with a cylindrical cover gas region was carried out for the validation of the method and showed that the temperature distribution and sodium mist concentration in the cover gas region are in good agreements with the test results. A simulation of a pool-type sodium cooled fast reactor has also conducted and the basic aspect of physical phenomena taking place in the cover gas region were evaluated.

In-plane movement of isolated poly(methacrylate) chains on a hydrophilic solid surface

Kawano, Masayuki*; Morimitsu, Yuma*; Liu, Y.*; Miyata, Noboru*; Miyazaki, Tsukasa*; Aoki, Hiroyuki; Kawaguchi, Daisuke*; Yamamoto, Satoru*; Tanaka, Keiji*

Macromolecules, 57(14), p.6625 - 6633, 2024/07

https://doi.org/10.1021/acs.macromol.4c00724

Validation of the hybrid turbulence model in detailed thermal-hydraulic analysis code SPIRAL for fuel assembly using sodium experiments data of 37-pin bundles

Yoshikawa, Ryuji; Imai, Yasutomo*; Kikuchi, Norihiro; Tanaka, Masaaki; Ohshima, Hiroyuki

Nuclear Technology, 210(5), p.814 - 835, 2024/05

https://doi.org/10.1080/00295450.2023.2249707

In the study of safety enhancement on advanced sodium-cooled fast reactor, it is essential to clarify the thermal-hydraulics under various operation conditions in a fuel assembly (FA) with the wire-wrapped fuel pins to assess the structural integrity of the fuel pin. A finite element thermal-hydraulics analysis code named SPIRAL has been developed to analyze the detailed thermal-hydraulics phenomena in a FA. In this study, the numerical simulations of the 37-pin bundle sodium experiments at different Re number conditions, including a transitional condition between laminar and turbulent flows and turbulent flow conditions, were performed to validate the hybrid turbulence model equipped in SPIRAL. The temperature distributions predicted by SPIRAL was consistent with those measured in the experiments. Through the validation study, the applicability of the hybrid turbulence model in SPIRAL to thermal-hydraulic evaluation of sodium-cooled FA in the wide range of Re number was confirmed.

Neutron reflectometry analysis of condensed water layer formation at a solid interface of epoxy resins under high humidity

Liu, Y.*; Miyata, Noboru*; Miyazaki, Tsukasa*; Shundo, Atsuomi*; Kawaguchi, Daisuke*; Tanaka, Keiji*; Aoki, Hiroyuki

Langmuir, 39(29), p.10154 - 10162, 2023/06

https://doi.org/10.1021/acs.langmuir.3c01091

Development plan for coupling technology between high temperature gas-cooled reactor HTTR and Hydrogen Production Facility, 2; Development plan for coupling equipment between HTTR and Hydrogen Production Facility

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

Proceedings of 30th International Conference on Nuclear Engineering (ICONE30) (Internet), 6 Pages, 2023/05

Analysis of the activities of the website "Question and Answer about radiation in daily life" after the accident at the Fukushima Daiichi Nuclear Power Plant and some lessons learned from it; To pass on this experience to the future

Kono, Takahiko; Tanaka, Masato*; Tanaka, Hitomi*; Shimo, Michikuni*; Torii, Hiroyuki*; Uno, Kazuko*

Journal of Radiation Protection and Research, 47(3), p.167 - 179, 2022/09

https://doi.org/10.14407/jrpr.2021.00346

After the accident at the Fukushima Daiichi Nuclear Power Plant, artificial radionuclides such as radioactive cesium and iodine were released into the environment. It caused great anxiety not only in the vicinity of the Fukushima Daiichi Nuclear Power Plant but also in other regions of the world. Some members of the Japan Health Physics Society (JHPS), a leading academic society in Japan in the field of radiation protection, volunteered to establish a website called "Question and Answer about radiation in Daily Life" shortly after the accident to help reduce the residents anxiety about the health effects of radiation. In August 2011, Committee for "Question and Answer about radiation in Daily Life" was established in JHPS, making the website-related activities a responsibility of JHPS. The Q&A website continued to respond to the questions from the general public with expertise and sincerity until February 2013 when the Committee members decided to end the activities because the number of questions received had gradually decreased with the passage of time. This paper aims to introduce the following: the activities of the Q&A website during the two years (2011-2013), the stance chosen for the activities, the information related to the website activities and the analysis of Twitter data. Building on the experience and the knowledge obtained from the activities, it also discusses issues and experiences that can be utilized in the initial response to emergencies for radiation protection experts as well as other fields.

Development of plant lifecycle optimization method, ARKADIA for advanced reactors

Uchibori, Akihiro; Sogabe, Joji; Okano, Yasushi; Takata, Takashi*; Doda, Norihiro; Tanaka, Masaaki; Enuma, Yasuhiro; Wakai, Takashi; Asayama, Tai; Ohshima, Hiroyuki

Proceedings of Technical Meeting on State-of-the-art Thermal Hydraulics of Fast Reactors (Internet), 10 Pages, 2022/09

The ARKAIDA has been developed to realize automatic optimization of plant design from safety evaluation for the advanced reactors represented by a sodium-cooled fast reactor. ARKADIA-Design offers functions to support design optimization both in normal operating conditions and design basis events. The multi-level simulation approach by the coupled analysis such as neutronics, core deformation, core thermal hydraulics was developed as one of the main technologies of the ARKADIA-Design. On the other hand, ARKAIDA-Safety aims for safety evaluation considering severe accidents. As a key technology, the numerical methods for in- and ex-vessel coupled phenomena during severe accidents in sodium-cooled fast reactors were tested through a hypothetical severe accident event.

Uranium chelating ability of decorporation agents in serum evaluated by X-ray absorption spectroscopy

Uehara, Akihiro*; Matsumura, Daiju; Tsuji, Takuya; Yakumaru, Haruko*; Tanaka, Izumi*; Shiro, Ayumi*; Saito, Hiroyuki*; Ishihara, Hiroshi*; Homma-Takeda, Shino*

Analytical Methods, 14(24), p.2439 - 2445, 2022/06

https://doi.org/10.1039/D2AY00565D

Angular momentum transfer in multinucleon transfer channels of O + Np

Tanaka, Shoya*; Hirose, Kentaro; Nishio, Katsuhisa; Kean, K. R.*; Makii, Hiroyuki; Orlandi, R.; Tsukada, Kazuaki; Aritomo, Yoshihiro*

Physical Review C, 105(2), p.L021602_1 - L021602_5, 2022/02

https://doi.org/10.1103/PhysRevC.105.L021602

Rabi-oscillation spectroscopy of the hyperfine structure of muonium atoms

Nishimura, Shoichiro*; Torii, Hiroyuki*; Fukao, Yoshinori*; Ito, Takashi; Iwasaki, Masahiko*; Kanda, Sotaro*; Kawagoe, Kiyotomo*; Kawall, D.*; Kawamura, Naritoshi*; Kurosawa, Noriyuki*; et al.

Physical Review A, 104(2), p.L020801_1 - L020801_6, 2021/08

https://doi.org/10.1103/PhysRevA.104.L020801