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Journal Articles

Numerical simulation of sodium mist behavior in turbulent Rayleigh-B$'e$nard convection using new developed mist models

Ohira, Hiroaki*; Tanaka, Masaaki; Yoshikawa, Ryuji; Ezure, Toshiki

Annals of Nuclear Energy, 172, p.109075_1 - 109075_10, 2022/07

 Times Cited Count:0 Percentile:0.02(Nuclear Science & Technology)

In order to evaluate the mist behavior in the cover gas region of Sodium-cooled Fast Reactors (SFRs) in good accuracy, turbulent model for Rayleigh-B$'e$nard convection (RBC) was selected, and the Reynolds-averaged number density and momentum equations for mist behavior were developed and incorporated into the OpenFOAM code. In the first stage, the RBC in a simple parallel channel was calculated using Favre-averaged k-$$omega$$ SST model. The average temperature and flow characteristics agreed well with results from DNS, LES, and experiments. Then the basic heat transfer experiment simulating the cover gas region of SFRs was calculated using this turbulent model and new mist models. The calculated average temperature distribution in the height direction and the mist mass concentration agreed well with the experimental results. We developed a method that could simulate the mist behavior in turbulent RBC environments and the cover gas region of SFRs with high accuracy.

Journal Articles

Investigation of applicability of subchannel analysis code ASFRE on thermal hydraulics analysis in fuel assembly with inner duct structure in sodium cooled fast reactor

Kikuchi, Norihiro; Imai, Yasutomo*; Yoshikawa, Ryuji; Doda, Norihiro; Tanaka, Masaaki

Proceedings of 28th International Conference on Nuclear Engineering (ICONE 28) (Internet), 8 Pages, 2021/08

In the design study of an advanced sodium-cooled fast reactor (Advanced-SFR) in JAEA, the use of a specific fuel assembly (FA) with an inner duct structure called FAIDUS has been investigated to enhance safety of Advanced-SFR. Due to the asymmetric layout of fuel rods by the inner duct, it is necessary to estimate the temperature distribution to confirm feasibility of FAIDUS. For the FAIDUS, confirmation of validity of the numerical results by a subchannel analysis code named ASFRE was not enough because the reference data on the thermal hydraulics in FAIDUS have not been obtained by the mock-up experiment, yet. Therefore, the code-to-code comparisons with numerical results of ASFRE and those of a CFD code named SPIRAL was conducted. The applicability of ASFRE was indicated through the confirmation of the consistency of mechanism of the specific temperature and velocity distributions appearing around the inner duct between the numerical results by ASFRE and those by SPIRAL.

Journal Articles

Validation study of finite element thermal-hydraulics analysis code SPIRAL to a large-scale wire-wrapped fuel assembly at low flow rate condition

Yoshikawa, Ryuji; Imai, Yasutomo*; Kikuchi, Norihiro; Tanaka, Masaaki; Gerschenfeld, A.*

Proceedings of Joint International Conference on Supercomputing in Nuclear Applications + Monte Carlo 2020 (SNA + MC 2020), p.73 - 80, 2020/10

A finite element thermal-hydraulics simulation code SPIRAL has been developed in Japan Atomic Energy Agency (JAEA) to analyze the detailed thermal-hydraulics phenomena in a fuel assembly (FA) of Sodium-cooled Fast Reactors (SFRs). The numerical simulation of a large-scale sodium test for 91-pin bundle (GR91) at low flow rate condition was performed for the validation of SPIRAL with the hybrid k-e turbulence model to take into account the low Re number effect near the wall in the flow and temperature fields. Through the numerical simulation, specific velocity distribution affected by the buoyancy force was shown on the top of the heated region and the temperature distribution predicted by SPIRAL agreed with that measured in the experiment and the applicability of the SPIRAL to thermal-hydraulic evaluation of large-scale fuel assembly at low flow rate condition was indicated.

Journal Articles

Subchannel analysis of thermal-hydraulics in a fuel assembly with inner duct structure of a sodium-cooled fast reactor

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

Journal of Nuclear Engineering and Radiation Science, 5(2), p.021001_1 - 021001_12, 2019/04

In the design study of an advanced loop-type sodium-cooled fast reactor in Japan, a specific fuel assembly (FA) named FAIDUS (Fuel Assembly with Inner DUct Structure) has been considered as one of the measures to enhance safety of the reactor during the core disruptive accident. In this study, thermal-hydraulics in FAIDUS was investigated with the in-house subchannel analysis code named ASFRE. Before the application to FAIDUS, applicability of ASFRE to FAs was confirmed through the numerical simulations for the experiments of simulated FA. Through the comparisons between the numerical results of thermal-hydraulic analyses of FAIDUS and a typical FA without the inner duct, it was indicated that significant asymmetric temperature distribution would not occur in FAIDUS at both high and low flow rate conditions.

Journal Articles

Development of numerical analysis method for core thermal-hydraulics during natural circulation decay heat removal in SFR, 1; Validation of ASFRE code in estimation of radial heat transfer phenomena

Kikuchi, Norihiro; Doda, Norihiro; Hashimoto, Akihiko*; Yoshikawa, Ryuji; Tanaka, Masaaki; Ohshima, Hiroyuki

Dai-23-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (USB Flash Drive), 5 Pages, 2018/06

For the thermal-hydraulic design regarding a fuel assembly of sodium-cooled fast reactors, a subchannel analysis code ASFRE has been developed by JAEA. ASFRE was applied to numerical simulations of several kinds of water and sodium experiments as its validation studies and it was confirmed that pressure drops and temperature distributions measured in the experiments can be well reproduced. To enhance safety of sodium-cooled fast reactor, it is required to evaluate thermal-hydraulics in a core during decay heat removal by natural circulation. It is necessary to estimate radial heat transfer phenomena between fuel assemblies. In this study, a numerical simulation of a 37-pin bundle sodium experiment with radial heat flux was carried out and it was confirmed that ASFRE can be qualitatively reproduced temperature distributions in a fuel assembly affected by radial heat transfer.

Journal Articles

Thermal-hydraulics analysis of fuel assembly with inner duct structure of a sodium-cooled fast reactor

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

Nihon Kikai Gakkai Kanto Shibu Ibaraki Koenkai 2017 Koen Rombunshu (CD-ROM), 4 Pages, 2017/08

A specific fuel assembly named FAIDUS (Fuel Assembly with Inner Duct Structure) has been developed as one of the measures to enhance safety of the reactor in the core disruptive accident (CDA) in JAEA. Thermal-hydraulics evaluations in FAIDUS under various operation conditions including the CDA are required to confirm its design feasibility. Therefore, numerical simulations by using thermal-hydraulics analysis program named SPIRAL developed in JAEA are conducted to analyze the thermal-hydraulics in the FAIDUS. Through the numerical simulation in the FAIDUS under tentative rated operation condition of an Advanced SFR, it was indicated that the flow and temperature distribution in the FAIDUS showed the same tendency as that in ordinary FA and specific characteristics was not observed.

Journal Articles

Thermal-hydraulic analysis of fuel assembly with inner duct structure of an advanced loop-type sodium-cooled fast reactor using ASFRE code

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

Proceedings of 25th International Conference on Nuclear Engineering (ICONE-25) (CD-ROM), 12 Pages, 2017/07

In the design study of an advanced loop-type SFR in JAEA, a specific fuel assembly (FA) named FAIDUS (Fuel Assembly with Inner DUct Structure) has been adopted as one of the measures to enhance safety of the reactor. Thermal-hydraulics evaluations of FAIDUS under various operation conditions are required to confirm its design feasibility. In this study, after the applicability of ASFRE to FAs was confirmed through the numerical analysis using simulated FA tests, thermal-hydraulic analyses of a FA without an inner duct and a FAIDUS were conducted. Through the numerical analyses, it was indicated that asymmetric temperature distribution in a FAIDUS would not be occurred and characteristics of the temperature distribution was almost the same as that in a FA without an inner duct. Under the low flow rate condition, it was expected that the local flow acceleration caused by the buoyancy force in a FAIDUS could bring the flow redistribution and make the temperature distribution flat.

Journal Articles

Development of sodium-water coupled thermal-hydraulics simulation code for sodium-heated straight tube steam generator of fast reactors

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

Proceedings of 11th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, Operation and Safety (NUTHOS-11) (USB Flash Drive), 12 Pages, 2016/10

A sodium-water coupled thermal-hydraulics simulation code TSG has been developed for numerical estimation of three-dimensional thermal-hydraulic phenomena in the straight-tube steam generator. The water analysis module was developed by using the parallel channel model of heat transfer tubes, and the sodium analysis module was developed by using porous body approach. As the first step of validation, simulation results by TSG were compared with the measured data of 1MWt SG experiments under steady state conditions. Through the numerical simulation, the coupled simulation method used in TSG was validated and applicability of TSG to simulate thermal-hydraulics of the straight tube SG in the steady state was confirmed.

JAEA Reports

Straight tube steam generator three-dimensional thermal-hydraulic code TSG; User's manual of water side simulation

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

JAEA-Data/Code 2014-034, 84 Pages, 2015/03

JAEA-Data-Code-2014-034.pdf:2.35MB

TSG (Three-dimensional Thermal-hydraulics Analysis Code for Steam Generators) is being developed for analyses of thermal hydraulics in double wall straight tube steam generator of Fast Breeder Reactor. TSG code is a thermal hydraulics simulation system which couples sodium side three dimensional simulation with water side multi-channel simulation. The three dimensional flow field in the sodium side is simulated by a commercial code FLUENT with porous media model. The multi-channel two-phase flow is simulated by an in-house module with drift-flux model. The sodium side simulation is coupled with the water side simulation by the transmission of heat transfer rate through the heat transfer tube. This report presents a description of the computational models, input and output as the user's manual of TSG water side module.

Journal Articles

Irradiation history of Itokawa regolith material deduced from noble gases in the Hayabusa samples

Nagao, Keisuke*; Okazaki, Ryuji*; Nakamura, Tomoki*; Miura, Yayoi*; Osawa, Takahito; Bajo, Kenichi*; Matsuda, Shintaro*; Ebihara, Mitsuru*; Ireland, T.*; Kitajima, Fumio*; et al.

Science, 333(6046), p.1128 - 1131, 2011/08

 Times Cited Count:118 Percentile:95.65(Multidisciplinary Sciences)

A steroid surface materials record regolith processes and a history of cosmic-ray irradiation. Noble gas isotopes in three rocky grains from Itokawa have been determined. High concentrations of solar He, Ne, and Ar, as high as those in lunar soils, are released at variable temperatures from each sample. The isotopic compositions are essentially identical to those of solar wind but distinguishable in He relative abundance. These noble gas characteristics can be explained by repeated implantation and preferential loss of solar He by removal of weathered He-rich rim on the grain surface through friction among regolith grains on Itokawa. Residence time of regolith materials on Itokawa is alculated to be shorter than 10 Myr, suggesting that regolith materials of small asteroids would escape easily to space.

JAEA Reports

Horonobe Underground Research Laboratory Project; Synthesis of phase I investigation 2001 - 2005, Volume "Geological disposal research"

Fujita, Tomo; Taniguchi, Naoki; Matsui, Hiroya; Tanai, Kenji; Maekawa, Keisuke; Sawada, Atsushi; Makino, Hitoshi; Sasamoto, Hiroshi; Yoshikawa, Hideki; Shibata, Masahiro; et al.

JAEA-Research 2011-001, 193 Pages, 2011/03

JAEA-Research-2011-001.pdf:5.23MB

This report summarizes the progress of research and development on geological disposal during the surface-based investigation phase (2001-2005) in the Horonobe Underground Research Laboratory project, of which aims are to apply the design methods of geological disposal and mass transport analysis to actual geological conditions obtained from the surface-based investigations in the Horonobe Underground Research Laboratory project as an example of actual geological environment.

JAEA Reports

Development of computer code for water-steam flow in steam generator and simulation of flow instabilities

Yoshikawa, Ryuji; Ohshima, Hiroyuki

JAEA-Research 2010-007, 44 Pages, 2010/06

JAEA-Research-2010-007.pdf:1.33MB

The feasibility assessment of steam generator with straight heat transfer tube is being carried out. In this study, the computer code for water-steam flow in SG was developed. The drift-flux model and semi-implicit method were used. The total flow rate and temperature of flow into the inlet plenum, and the pressure of outlet plenum were given as boundary conditions. The flow instability experiments with two parallel channels were simulated. The capability of computer code on predictions of flow oscillation and stable boundary was confirmed. The sensitivity analysis was also carried out to quantify the impact of each parameter on oscillation period and the stable boundary. It can be expected to improve the accuracy of the computer code by using appropriate drift velocity and two-phase frictional multiplier correlations.

JAEA Reports

Numerical methods on flow instabilities in steam generator

Yoshikawa, Ryuji; Hamada, Hirotsugu; Ohshima, Hiroyuki; Yanagisawa, Hideki*

JAEA-Research 2008-058, 29 Pages, 2008/06

JAEA-Research-2008-058.pdf:1.31MB

In Japan Atomic Energy Agency, investigations on heat transfer characteristics of steam generator are being performed for the development of Sodium-cooled Fast Breeder Reactor. In this report, the numerical methods were studied for two-phase flow instability analysis in steam generator. For numerical simulation purpose, the flow instability analysis code was developed with homogeneous equilibrium model on single heat transfer tube. The special algorithm to calculate inlet flow rate with inlet pressure, outlet pressure and heat flux as boundary conditions for the density-wave instability analysis has been established. The flow instability in single tube was successfully simulated with homogeneous equilibrium model. Then the drift-flux model including the effects of subcooled boiling and two phase slip was adopted to improve the accuracy. The capability of drift-flux model for simulating density-wave instability in single tube was confirmed.

JAEA Reports

Construction of interfacial area concentration model for sodium-water reaction

Yoshikawa, Ryuji; Ohshima, Hiroyuki; Hamada, Hirotsugu; Kurihara, Akikazu; Uchibori, Akihiro

JAEA-Research 2008-055, 24 Pages, 2008/06

JAEA-Research-2008-055.pdf:3.19MB

In Japan Atomic Energy Agency, thermal hydraulic studies on sodium-water reaction are being performed with the multi-component and multi-phase code SERAPHIM. The interfacial area concentration of sodium droplets in the steam is important for the accurate analysis of sodium-water reaction. In this report, the theoretical analysis and numerical models for gas jets were reviewed to understand the mixing process of sodium and water. As for theoretical analysis, existing critical flow rate, depressurization and entrainment analysis for jet flows were summarized. The applicability of critical flow rate equations for subcooled water at 17MPa were confirmed after investigating its effect of compressibility. Based on the available knowledge on entrained droplet sizes in gas jets, a transport equation of sodium droplet interfacial area concentration was constructed for multiphase flow simulation.

Journal Articles

Development of thermal hydraulic computer code for steam-water flow in steam generator of fast breeder reactor

Yoshikawa, Ryuji; Hamada, Hirotsugu; Ohshima, Hiroyuki; Yanagisawa, Hideki*

Dai-13-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu, p.495 - 496, 2008/06

In Japan Atomic Energy Agency, investigations on heat transfer characteristics of steam generator are being performed for the development of commercialized sodium-cooled fast breeder reactor. In this study, the computer code for flow instability analysis in single heat transfer tube was developed with drift-flux model which included the effects of subcooled boiling and two phase slip. The special algorithm to calculate inlet flow rate with inlet pressure, outlet pressure and heat flux as boundary conditions for the density-wave instability analysis has been established. The subcooled model was verified by calculating the void fraction distribution of steady heat transfer flow. The capability of drift flux model for simulating density-wave instability in single tube was confirmed.

JAEA Reports

Horonobe Underground Research Laboratory project synthesis of phase I investigation 2001-2005 volume "Geological Disposal Research"

Fujita, Tomo; Taniguchi, Naoki; Matsui, Hiroya; Tanai, Kenji; Nishimura, Mayuka; Kobayashi, Yasushi; Hiramoto, Masayuki; Maekawa, Keisuke; Sawada, Atsushi; Makino, Hitoshi; et al.

JAEA-Research 2007-045, 140 Pages, 2007/03

JAEA-Research-2007-045-1.pdf:38.33MB
JAEA-Research-2007-045-2.pdf:44.62MB

This report summarizes the progress of research and development on geological disposal during the surface-based investigation phase (2001-2005) in the Horonobe Underground Research Laboratory project (HOR), of which aims are to apply the design methods of geological disposal and mass transport analysis to actual geological environments and derive future subjects based on the results. The design methods of geological disposal were reviewed based on the recent knowledge and the advertences were identified. Then the parameters for virtual design of engineered barrier system, backfill deposition hole and tunnel were set up based on actual geological conditions obtained from the surface-based investigations in HOR as an example of actual geological environment. The conceptual structure from site investigation and evaluation to mass transport analysis was represented as a work flow at first. Then following this work flow a series of procedures for mass transport analysis were carried out based on actual geological conditions obtained from the surface-based investigations in HOR to illustrate the functioning of the work flow and the applicability of this methodology. Consequently, knowledge that will help follow-up or future execution and planning activities was obtained.

JAEA Reports

None

Makino, Hitoshi; Sawada, Atsushi; Maekawa, Keisuke; Shibata, Masahiro; Sasamoto, Hiroshi; Yoshikawa, Hideki; Wakasugi, Keiichiro; Koo, Shigeru*; Hama, Katsuhiro; Kurikami, Hiroshi; et al.

JNC TN1400 2005-021, 148 Pages, 2005/09

JNC-TN1400-2005-021.pdf:32.85MB

None

Oral presentation

Development of simulation code for flow instability in steam generator of fast breeder reactor

Yoshikawa, Ryuji; Ohshima, Hiroyuki

no journal, , 

The computer code for flow instability analysis in the water side of steam generator of FBR was developed with drift-flux model and semi-implicit method. The code was verified by the simulation of density-wave instabilities in parallel channels. The sensitivity analysis was also carried out. The capability of drift flux model and simi-implicit method for simulating density-wave instability in parallel channels was confirmed.

Oral presentation

Validation of simulation code for flow instability in steam generator of fast breeder reactor

Yoshikawa, Ryuji; Ohshima, Hiroyuki

no journal, , 

A computer code to analyze flow stability in water-side of steam generator was developed with drift-flux model. The semi-implicit numerical scheme was used, and the flow equations were differenced over the staggered mesh. The total flow rate and the temperature of water flowing into the inlet plenum, the pressure of outlet plenum, and the heat flux through the heat transfer tube were given as boundary conditions for the simulation of water-side flow. The computer code was validated by the simulation of flow instabilities in two parallel channels and comparison with experimental results. The capability of the computer code on prediction of flow oscillation and stable boundary in two parallel channels was confirmed. The sensitivity analysis was also carried out to quantify the impacts of some important parameters on oscillation period and the stable boundary.

Oral presentation

Development of evaluation method for flow stability in steam generator of fast breeder reactor

Yoshikawa, Ryuji; Ohshima, Hiroyuki

no journal, , 

The feasibility assessment of steam generator with straight heat transfer tube is being carried out within the research and development activities for the practical application of Fast Breeder Reactor. So the computer code to analyze flow stability of the steam generator is being developed. A single-channel simulation code which couples the sodium flow with water flow was developed by adding the temperature calculation modules of sodium and heat transfer tube to the water side flow simulation. The flow stability experiments of 70 MW steam generator were simulated, and the capability of computer code on predictions of flow oscillation was confirmed.

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