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Mori, Tetsuya; Oki, Shigeo
Nuclear Technology, 20 Pages, 2025/00
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)This study investigates the characteristics of the Doppler coefficient and sodium void reactivity of a burning fast reactor core concept, which was constructed in a previous study. This concept allows for multiple recyclings of plutonium and minor actinides (transuraniums (TRU)). TRU degradation due to multiple recycling deteriorates the reactivity coefficients through indirect effects, such as by hardening the neutron spectrum and steepening the energy gradient of neutron importance. Using silicon carbide (SiC) structural material improves the reactivity coefficient by causing an opposite indirect effect of TRU degradation. This improvement results not only from neutron spectrum softening due to the neutron moderation effect from 
C but also from the neutron leakage effect resulting from the low structural material density. The disadvantage of increased calculation uncertainty by using SiC structural material can be practically ignored. Furthermore, the burning core has Doppler coefficient enhancement characteristics by the moderated neutron reflection effect from outside the core. This characteristic has the potential to provide a new measure for reactivity coefficient deterioration due to TRU degradation. The reactivity coefficient characteristics clarified in this study can provide valuable knowledge for future detailed designs and design improvements of a TRU burning core.
Hatakeyama, Nozomu*; Miura, Ryuji*; Miyamoto, Naoto*; Miyamoto, Akira*; Ara, Kuniaki; Shimoyama, Kazuhito; Kato, Atsushi; Yamamoto, Tomohiko
Journal of Computer Chemistry, Japan, 21(2), p.61 - 62, 2022/00
no abstracts in English
Yamamoto, Tomohiko; Matsubara, Shinichiro*; Harada, Hidenori*; Saunier, P.*; Martin, L.*; Gentet, D.*; Dirat, J.-F.*; Collignon, C.*
Nuclear Engineering and Design, 383, p.111406_1 - 111406_14, 2021/11
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)Japan-France collaboration on ASTRID (Advanced Sodium Technological Reactor for Industrial Demonstration) project is launched in 2014. In this project, Japan-France evaluates core assemblies with interferences on seismic event. The object of this study is to verify the seismic evaluation method on core assemblies between Japan and France by comparing the results. The analysis of this benchmark calculation shows a satisfactory agreement between the Japanese and French tools and the figures show a good behavior of the core in horizontal direction under French seismic condition.
Yamamoto, Tomohiko; Matsubara, Shinichiro*; Harada, Hidenori*; Saunier, P.*; Martin, L.*; Gentet, D.*; Dirat, J.-F.*; Collignon, C.*
Proceedings of 2019 International Congress on Advances in Nuclear Power Plants (ICAPP 2019) (Internet), 8 Pages, 2019/05
Japan-France collaboration on ASTRID (Advanced Sodium Technological Reactor for Industrial Demonstration) project is launched in 2014. In this project, Japan-France evaluates core assemblies with interferences on seismic event. The object of this study is to verify the seismic evaluation method on core assemblies between Japan and France by comparing the results. The analysis of this benchmark calculation shows a satisfactory agreement between the Japanese and French tools and the figures show a good behavior of the core in horizontal direction under French seismic condition.
Yamano, Hidemasa; Xu, Y.*; Sago, Hiromi*; Hirota, Kazuo*; Baba, Takeo*
Proceedings of 2016 International Congress on Advances in Nuclear Power Plants (ICAPP 2016) (CD-ROM), p.1029 - 1038, 2016/04
This study conducted the flow-induced vibration evaluation of the primary hot-leg piping in the demonstration reactor design of advanced loop-type sodium-cooled fast reactor in order to confirm the integrity of the piping. Following the description of the primary hot-leg piping design and a design guideline of the flow-induced vibration evaluation, this paper describes mainly the flow-induced vibration evaluation and thereby the integrity assessment. In the fatigue evaluation for the flow-induced vibration, the pipe stresses considering the stress concentration factor and so on, at representative locations were less than the design fatigue limit. Therefore, this evaluation confirmed the integrity of the primary hot-leg piping in the demonstration reactor.
Kojima, Saori*; Uchibori, Akihiro; Takata, Takashi; Ohno, Shuji; Fukuda, Takeshi*; Yamaguchi, Akira*
no journal, ,
Analytical evaluation on a self-wastage phenomenon at heat transfer tubes in the steam generator of sodium cooled fast reactors has been performed by using the sodium-water reaction analysis code SERAPHIM. In this study, a fluid-structure thermal coupling model was developed and incorporated in the SERAPHIM code to evaluate heat transfer between the sodium-side reacting flow and the outer surface of the heat transfer tube. The effect of the fluid-structure thermal coupling model on the temperature field was demonstrated through the numerical analyses.
Kawaguchi, Munemichi; Miyahara, Shinya; Uno, Masayoshi*
no journal, ,
no abstracts in English
Otsuka, Satoshi; Tanno, Takashi; Oka, Hiroshi; Yano, Yasuhide; Uwaba, Tomoyuki; Kaito, Takeji; Furukawa, Tomohiro; Kato, Shoichi
no journal, ,
Maximum temperature of ODS steel cladding tube for long life fast reactor fuel is very high (approximately 700
C) in normal operation condition. It was reported that, in reactor operation, mass transfer phenomena (dissolution, deposition, penetration) took place as a result of increased solubility of steel constituent elements in liquid Na. The driving force of these phenomena is the chemical potential gap of solute elements in steel and liquid Na, which is dependent of not only temperature but also other factors such as impurity concentrations in Liquid Na. For appropriately evaluating experimental data and predicting the corrosion behavior in actual plant, it is required to list up the key factors including other factors than temperature and residence time and understand the effects of these factors. In this study, transfer behavior of Cr (main alloying element of ODS steel) is discussed; modelling and numerical calculation were carried out on Cr dissolution behavior from fast reactor fuel cladding tube into liquid Na.
Sugimoto, Taro*; Saito, Shimpei*; Kaneko, Akiko*; Abe, Yutaka*; Uchibori, Akihiro; Takata, Takashi; Ohshima, Hiroyuki
no journal, ,
A computational fluid dynamics code for a sodium-water reaction phenomenon in a steam generator of sodium-cooled fast reactors has been developed. In order to provide the data for validation of this code, the visualization experiment on liquid droplet entrainment in the high-pressure air jet submerged in the water pool was carried out. The experiment successfully elucidated the velocity of the entrained liquid droplet.
Yamamoto, Tomohiko; Watakabe, Tomoyoshi; Miyagawa, Takayuki*; Fukasawa, Tsuyoshi*; Okamura, Shigeki*; Fujita, Satoshi*
no journal, ,
This report describes that the concepts and specifications of "3-dimensional seismic isolation device" adopting in a Sodium-cooled fast reactor, summary of various tests and analyzes, and shows feasibility of 3-dimensional seismic isolation device for earthquake response reduction effect.
Xu, Y.*; Sago, Hiromi*; Hirota, Kazuo*; Baba, Takeo*; Yamano, Hidemasa
no journal, ,
Structural integrity evaluation has been carried out for a hot-leg pipe due to random vibration induced by turbulence of pipe flow using "proposed guideline of flow-induced vibration evaluation for the primary hot-leg piping in sodium-cooled fast reactor", which has reflected the R&D results of the flow-induced vibration for a large-diameter piping. This gave the prospect of integrity of the primary hot-leg piping in the demonstration fast reactor.
Hatakeyama, Nozomu*; Miura, Ryuji*; Suzuki, Ai*; Miyamoto, Naoto*; Miyamoto, Akira*; Ara, Kuniaki; Shimoyama, Kazuhito; Kato, Atsushi; Yamamoto, Tomohiko
no journal, ,
no abstracts in English
Kawaguchi, Munemichi*; Kai, Kotaro*; Sawa, Kazuhiro*; Sato, Rika; Seino, Hiroshi
no journal, ,
We have researched a stable aerosol generation and supply technology to reveal the trapping effect of radioactive substances e.g. caesium in a sodium pool. This study will show our investigation results on the aerosolization of the liquid suspension.
particlesKai, Kotaro*; Kawaguchi, Munemichi*; Sawa, Kazuhiro*; Sato, Rika; Seino, Hiroshi
no journal, ,
The wet dispersion method is a method to obtain test aerosol particles by spraying the suspension into the air and drying it. This study confirmed the performance of the wet dispersion method using TiO (1 micrometre in diameter) and water as test particles and a solvent.
Yamamoto, Tomohiko; Matsubara, Shinichiro*; Iwasaki, Akihisa*; Kawamura, Kazuki*; Harada, Hidenori*
no journal, ,
A fast reactor core consists of hundreds of core elements, which lengthen due to thermal expansion and swelling. So, the core elements are self-standing on the core support structure and not restrained in the axial direction. The authors carried out vibration tests and verification of analysis code (REVIAN-3) to evaluate 3D core vibration behavior. This report describes the summary of some experimental results and analysis.
Kawaguchi, Munemichi; Miyahara, Shinya; Uno, Masayoshi*
no journal, ,
To evaluate the chemical reaction process of Na-concrete reaction, the model which adopted the thermodynamic database and the latest chemical reaction kinetics in COMSOL Multiphysics has been developed. The chemical reaction processes at each temperature were simulated well by the numerical calculation with this model.
Fujimura, Koji*; Shirakura, Shota*; Oki, Shigeo; Takeda, Toshikazu*
no journal, ,
no abstracts in English
Kawaguchi, Munemichi; Miyahara, Shinya; Uno, Masayoshi*
no journal, ,
Sodium-concrete reaction (SCR) experiments in sodium-cooled fast reactors were performed to reveal the phenomena that the reaction gradually terminates by the sedimentation effect of the reaction products. In addition, some physical properties of the reaction products (density, specific heat, melting point) were measured after SCR experiment.
