Refine your search:     
Report No.
 - 
Search Results: Records 1-20 displayed on this page of 50

Presentation/Publication Type

Initialising ...

Refine

Journal/Book Title

Initialising ...

Meeting title

Initialising ...

First Author

Initialising ...

Keyword

Initialising ...

Language

Initialising ...

Publication Year

Initialising ...

Held year of conference

Initialising ...

Save select records

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

Study on evaluation method for entrained gas flow rate by free surface vortex

Ito, Kei*; Ito, Daisuke*; Saito, Yasushi*; Ezure, Toshiki; Matsushita, Kentaro; Tanaka, Masaaki; Imai, Yasutomo*

Proceedings of 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-18) (USB Flash Drive), p.6632 - 6642, 2019/08

In this paper, a mechanistic model is proposed to calculate the entrained gas flow rate by a free surface vortex. The model contains the theoretical equation of transient gas core elongation and the empirical equation of critical gas core length for gas bubble detachment. Based on those two equations, the entrained gas flow rate is calculated as the portion of the gas core elongated beyond the critical gas core length per unit time. Then, the mechanistic model was applied to the calculation of the entrained gas flow rate in a simple water experiment. As a result, it is confirmed that the entrained gas flow rate grows rapidly when the liquid (water) flow rate, which determine the strength of a free surface vortex, exceeds a certain threshold value.

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

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

Nippon 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.

Journal Articles

Development of numerical simulation system for thermal-hydraulic analysis in fuel assembly of sodium-cooled fast reactor

Ohshima, Hiroyuki; Uwaba, Tomoyuki; Hashimoto, Akihiko*; Imai, Yasutomo*; Ito, Masahiro*

AIP Conference Proceedings 1702, p.040011_1 - 040011_4, 2015/12

A numerical simulation system, which consists of a deformation analysis program and three kinds of thermal-hydraulics analysis programs, is being developed in Japan Atomic Energy Agency in order to offer methodologies to clarify thermal-hydraulic phenomena in fuel assemblies of sodium-cooled fast reactors under various operating conditions including fuel deformation. This paper gives a summary of numerical methods of component programs of the system and their validation studies.

Journal Articles

Numerical analysis of flow field around simulated wire-wrapped fuel pins of fast reactor

Kikuchi, Norihiro; Ohshima, Hiroyuki; Imai, Yasutomo*; Hiyama, Tomoyuki; Nishimura, Masahiro; Tanaka, Masaaki

Nippon Kikai Gakkai Kanto Shibu Ibaraki Koenkai 2015 Koen Rombunshu, p.179 - 180, 2015/08

In an economically improved sodium-cooled fast reactor, a narrower gap is considered among the fuel pins so as to achieve a high burn-up. Therefore, it is needed to evaluate thermal-hydraulic characteristics in case of a change of the gap geometry due to deformation of fuel pin caused by such as a swelling and a thermal bowing. For this purpose, a FEM analysis code, SPIRAL has been being developed in JAEA and the code validations using water or sodium experimental results have also being performed. In this study, a numerical analysis of a flow field around wire-wrapped fuel pins based on a 3-pin bundle water experiment was carried out as a validation study of SPIRAL. As a result, it was demonstrated that the hybrid-type turbulent model incorporated in SPIRAL has a high applicability to investigate the flow structure of the narrow gap in the fuel assembly.

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

Study on flow in the subchannels of pin bundle with wrapping wire

Nishimura, Masahiro; Hiyama, Tomoyuki; Kamide, Hideki; Ohshima, Hiroyuki; Nagasawa, Kazuyoshi*; Imai, Yasutomo*

Proceedings of 9th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-9) (CD-ROM), 7 Pages, 2014/11

Journal Articles

Study on simulation code for dilute bubbly flow

Ito, Kei; Takata, Takashi*; Ohno, Shuji; Kogawa, Hiroyuki; Kamide, Hideki; Imai, Yasutomo*; Kawamura, Takumi*

Nippon Kikai Gakkai Rombunshu, B, 79(808), p.2630 - 2634, 2013/12

In a sodium-cooled fast reactor, inert gas exists in the primary coolant system as bubbles or dissolved gas. Similarly, small bubbles exist also in the mercury target loop in J-PARC to suppress cavitation erosion. To simulate these inert gas behaviors in liquid metal flows, the Japan Atomic Energy Agency (JAEA) has developed a plant dynamics code VIBUL. In this study, new models, i.e. the bubble release and bubble carry under models, are introduced to simulate the bubble behaviors in the fast reactor and mercury target system. Then, the small bubble behavior in the mercury target system is simulated to check the validity of the new models.

JAEA Reports

Study on numerical simulation of bubble and dissolved gas behavior in liquid metal flow

Ito, Kei; Ohno, Shuji; Kamide, Hideki; Kogawa, Hiroyuki; Futakawa, Masatoshi; Kawamura, Takumi*; Imai, Yasutomo*

JAEA-Research 2013-008, 117 Pages, 2013/10

JAEA-Research-2013-008.pdf:6.55MB

The Japan Atomic Energy Agency has been developed a plant dynamics code VIBUL to simulate the concentration distributions of the dissolved gas and the bubbles in a fast reactor. In this study, the VIBUL code is improved to achieve accurate simulations, e.g. rigorous mole conservation of inert gas. Moreover, new modles are introduced to simulate the small bubble behaviors in the J-PARC mercury target system. To validate the improved models and the newly developed models, the inert gas behaviors in the large-scale sodium-cooled reactor and the small bubble behaviors are simulated. As a result, it is confirmed that the complicated bubble dynamics in each component, e.g. core, IHX or surge tank, can be simulated appropriately by the VIBUL code.

JAEA Reports

Study on evaluation method for gas entrainment and vortex cavitation phenomena; Evaluation of gas entrainment behavior in large-scale test and development of proto-type evaluation method for vortex cavitation

Ito, Kei; Ezure, Toshiki; Ohno, Shuji; Kamide, Hideki; Nakamine, Yoshiaki*; Imai, Yasutomo*

JAEA-Research 2013-007, 75 Pages, 2013/10

JAEA-Research-2013-007.pdf:5.21MB

In Japan Atomic Energy Agency (JAEA), various thermal hydraulics phenomena in an upper plenum region are evaluated in the study on the safety design criteria of the sodium-cooled fast reactor in Japan (JSFR). The gas entrainment (GE) from a free surface of coolant and the vortex cavitation (VC) at the H/L intake are important phenomena to be evaluated. Since these phenomena occur by the significant pressure drop in the vicinity of a vortex center, a technique to evaluate a vortex behavior is indispensable. The authors are developing a GE evaluation method using a numerical analysis and a vortex model. In this study, the evaluations are performed on the GE behavior in the 1/1.8 scaled water model test. In addition, a VC evaluation method is proposed on the basis of the GE evaluation method. As a basic validation of the VC evaluation method, the basic sub-surface vortex test in JAEA is evaluated.

Journal Articles

Investigation on velocity distribution around the wrapping wire in an inner subchannel of fuel pin bundle

Nishimura, Masahiro; Sato, Hiroyuki; Kamide, Hideki; Ohshima, Hiroyuki; Nagasawa, Kazuyoshi*; Imai, Yasutomo*

Proceedings of 20th International Conference on Nuclear Engineering and the ASME 2012 Power Conference (ICONE-20 & POWER 2012) (DVD-ROM), 10 Pages, 2012/07

Feature of stream regime in the subchannel existing wrapping wire was visualized in vertical and horizontal plane by the PIV method. And the time averaged velocity field in the horizontal plane was reconstructed from the two vertical plane data in different directions. A detailed simulation code based on FEM was applied to the experimental analysis. The calculated velocity distributions were consistent with the experimental data.

Journal Articles

Study on turbulent modeling in gas entrainment evaluation method

Ito, Kei; Ohshima, Hiroyuki; Nakamine, Yoshiaki*; Imai, Yasutomo*

Journal of Power and Energy Systems (Internet), 6(2), p.151 - 164, 2012/06

Suppression of gas entrainment (GE) phenomena caused by free surface vortices are very important to establish an economically superior design of the sodium-cooled fast reactor in Japan (JSFR). Therefore, the authors are developing a CFD-based evaluation method in which the non-linearity and locality of the GE phenomena can be considered. In this study, the authors develop a turbulent vortex model to evaluate the GE phenomena more accurately. Then, the improved GE evaluation method with the turbulent viscosity model is validated by analyzing the GC lengths observed in a simple experiment. The evaluation results show that the GC lengths analyzed by the improved method are shorter in comparison to the original method, and give better agreement with the experimental data.

Journal Articles

Study on turbulent modeling in gas entrainment evaluation method

Ito, Kei; Ohshima, Hiroyuki; Imai, Yasutomo*

Proceedings of 19th International Conference on Nuclear Engineering (ICONE-19) (CD-ROM), 8 Pages, 2011/10

In the design study on a large-scale sodium-cooled fast reactor in Japan (JSFR), the suppression of gas entrainment (GE) phenomena caused by free surface vortices are very important. However, due to the non-linearity and/or locality of the GE phenomena, it is not easy to evaluate the occurrences of the GE phenomena accurately. Therefore, the authors are developing a CFD-based evaluation method in which the non-linearity and locality of the GE phenomena can be considered. In this study, the authors develop a turbulent vortex model to evaluate the GE phenomena more accurately. Then, the improved GE evaluation method with the turbulent viscosity model is validated by analyzing the GC lengths observed in a simple experiment. The evaluation results show that the GC lengths analyzed by the improved method are shorter in comparison to the original method, and give better agreement with the experimental data.

JAEA Reports

Study on improvement of gas entrainment evaluation method

Ito, Kei; Ohshima, Hiroyuki; Xu, Y.*; Imai, Yasutomo*

JAEA-Research 2010-063, 58 Pages, 2011/03

JAEA-Research-2010-063.pdf:2.09MB

Japan Atomic Energy Agency has conducted the FaCT project to study a conceptional design of a large-scale sodium-cooled fast reactor in which the coolant in the vessel has relatively higher velocity than conventional designs and may causes cover gas entrainment (GE) in an upper plenum region. The authors has been studied a evaluation method of GE in fast reactors and the 1st proposal (prototype) of "Design Guideline for Gas Entrainment Prevention Using CFD Method" was published in 2006. In this study, the prototype evaluation method was improved by introducing the surface tension and turbulent effects. The improved GE evaluation method was validated by analyzing the gas core lengths observed in simple experiments. Results showed that the analytical gas core lengths calculated by the improved GE evaluation method were shorter in comparison to the prototype GE evaluation method, and gave better agreement with the experimental data.

Journal Articles

Numerical simulation of flow field in wire-wrapped fuel pin bundle of sodium cooled fast reactor using "SPIRAL"

Ohshima, Hiroyuki; Imai, Yasutomo*

Proceedings of 5th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-5), p.480 - 487, 2006/11

A numerical simulation of flow field in a wire-wrapped fuel pin bundle of fast breeder reactor was carried out using a finite element method code SPIRAL under typical operating conditions. In this simulation, the influence of computational mesh scheme on the result and the prediction characteristics of three kinds of high Reynolds number turbulence models, modified k-e model, RNG k-e model and algebraic stress model, were clarified. SPIRAL was also applied to simulate a flow field in a wire-wrapped 19-pin bundle and its result was compared with available experimental data. The predicted distributions of axial flow through the subchannels, cross flow between fuel pins and swirl flow along the wrapper tube wall caused by the existence of wire-spacer were in good agreement with measured ones.

Journal Articles

Validation study of thermal-hydraulic analysis program "SPRAL" for fuel pin bundle of sodium-cooled fast reactor

Ohshima, Hiroyuki; Imai, Yasutomo*

Proceedings of 11th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-11) (CD-ROM), 14 Pages, 2005/10

A numerical simulation system is being developed at JNC in order to offer methodologies to clarify thermal-hydraulic phenomena in a fuel subassembly of sodium-cooled fast reactors under various operating conditions. This paper describes the validation study of SPIRAL that is one component code of the numerical simulation system and that plays the role to simulate detailed local flow and temperature fields in a wire-wrapped fuel pin bundle. Fundamental validity related to solving mass, momentum and energy conservation equations and applicability of turbulence models were confirmed by simulating several basic problems. As the typical examples, two kinds of simulations using mainly high Re number models, backward facing step flow and flow in 4-fuel-pin bundle in a rectangular duct, are selected and the characteristics of flow field prediction by the models as well as the validity of the component code are mentioned.

JAEA Reports

Coupling analysis of deformation and thermal hydraulics in a FBR fuel pin bundle using BAMBOO and ASFRE-IV codes

Ito, Masahiro*; Imai, Yasutomo*; Uwaba, Tomoyuki; Ohshima, Hiroyuki

JNC-TN9400 2004-003, 40 Pages, 2004/03

JNC-TN9400-2004-003.pdf:0.83MB

JNC has been developing a bundle deformation analysis code BAMBOO, a thermal hydraulics analysis code ASFRE and their coupling method as a simulation system for the evaluation on the integrity of deformed FBR fuel pin bundles. In this study, the simulation system was applied to a coupled analysis of deformation and thermal-hydraulics in a pin-bundle under a steady state condition just after startup for the purpose of the verification of the simulation system.

50 (Records 1-20 displayed on this page)