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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

Methodology development for transient flow distribution analysis in high temperature gas-cooled reactor

Aoki, Takeshi; Sato, Hiroyuki; Ohashi, Hirofumi

Proceedings of 2020 International Conference on Nuclear Engineering (ICONE 2020) (Internet), 6 Pages, 2020/08

The flow distribution analysis, which is a part of thermal hydraulic design of the prismatic-type of the high temperature gas cooled reactor (HTGR) considering unintended flows between graphite blocks, has been performed for steady and conservative conditions. On the other hand, the transient analysis for satisfactorily realistic conditions will be helpful for the design improvement of prismatic-type HTGR. The present study aims to develop the transient flow distribution analysis code and confirm its applicability for the transient flow distribution analysis for prismatic-type HTGRs during anticipated operational occurrences and accidents utilizing experiences on high temperature engineering test reactor (HTTR) design. The calculation model and code were developed and validated for analysis of the unintended flows in the core and the molecular diffusion dominant in beginning air ingress behavior in an air ingress accident.

Journal Articles

Development of numerical estimation method for thermal hydraulics in reactor vessel of sodium-cooled fast reactor under decay heat removal system operation conditions; Preliminary thermal hydraulics simulation for simulated reactor vessel in sodium experimental apparatus PLANDTL-2

Tanaka, Masaaki; Ono, Ayako; Hamase, Erina; Ezure, Toshiki; Miyake, Yasuhiro*

Nippon Kikai Gakkai Kanto Shibu Ibaraki Koenkai 2018 Koen Rombunshu (CD-ROM), 4 Pages, 2018/08

Decay heat removal system (DHRS) by using the natural circulation without depending on the pump as the mechanical equipment is recognized as one of the most effective methodologies for the sodium-cooled fast reactor from the viewpoint of the safety enhancement. The numerical estimation method which can predict thermal hydraulic phenomena in the natural circulation under the plant cooling process by operating the various DHRSs including the severe accident is necessarily required. In this paper, the numerical results of the preliminary analysis for the sodium experiment condition with the apparatus of PLANDTL-2, in which the core and the upper plenum with a dipped-type direct heat exchanger (DHX) were modeled, were discussed, in order to establish an appropriate numerical models for the direct heat exchanger (DHX).

Journal Articles

Development of V2UP (V&V plus uncertainty quantification and prediction) procedure for high cycle thermal fatigue in fast reactor; Framework for V&V and numerical prediction

Tanaka, Masaaki; Ohno, Shuji; Ohshima, Hiroyuki

Nuclear Engineering and Design, 299, p.174 - 183, 2016/04

 Times Cited Count:3 Percentile:59.43(Nuclear Science & Technology)

A procedure called as V2UP (Verification and Validation plus Uncertainty quantification and Prediction) was made by referring to the existing guidelines on V&V and the methodologies of the safety assessment (CSAU, ISTIR, EMDAP). The V2UP consisted of five components as follows: (1) phenomena analysis with the Phenomena Identification and Ranking Table (PIRT) method, (2) implementation of the V&V, (3) design and rearrangement of experiments for the V&V, (4) uncertainty quantification in each problem and integration of uncertainties and (5) numerical prediction (estimation) for the target issue. Although the complete application of the procedure has not been performed at this moment, a flow chart of the V2UP procedure was described in this paper with recent results of the examinations.

JAEA Reports

Thermal-hydraulic analyses of poisoned cold moderator vessel, 1; Study on Poison Plate Layout

Sato, Hiroshi; Aso, Tomokazu; Kogawa, Hiroyuki; Teshigawara, Makoto; Hino, Ryutaro

JAERI-Tech 2004-018, 23 Pages, 2004/03

JAERI-Tech-2004-018.pdf:2.42MB

The Japan Atomic Energy Research Institute is constructing a mega-watt class spallation neutron source in cooperation with the High Energy Accelerator Research Organization. A cold moderator using liquid hydrogen is one of the key components in the system, which directly affects the neutronic performance both in intensity and pulse time structure. Since a hydrogen temperature rise in the moderator vessel affects the neutronic performance, it is necessary to suppress the recirculation and stagnant regions which would cause hot spots. A cold moderator with a poison plate (poisoned decoupled moderator) has a high possibility to generate the stagnant region on and near the poison plate. Thermal-hydraulic analyses were carried out with proposed inner structure of the poisoned cold moderator. The stagnant and recirculation regions could be reduced by making a gap between the poison plate end and the vessel bottom surface, and the local temperature rise also could be kept under the required design value.

Journal Articles

Thermal-hydraulic characteristics of IFMIF liquid lithium target

Ida, Mizuho*; Nakamura, Hideo; Nakamura, Hiroshi*; Nakamura, Hiroo; Ezato, Koichiro; Takeuchi, Hiroshi

Fusion Engineering and Design, 63-64, p.333 - 342, 2002/12

 Times Cited Count:40 Percentile:8.08(Nuclear Science & Technology)

no abstracts in English

Journal Articles

Development of cold moderator vessel for the spallation neutron source; Flow field measurements and thermal hydraulic analyses in cold moderator vessel

Aso, Tomokazu; Kaminaga, Masanori; Terada, Atsuhiko*; Hino, Ryutaro

Nippon Genshiryoku Gakkai-Shi, 43(11), p.1149 - 1158, 2001/11

 Times Cited Count:0 Percentile:100(Nuclear Science & Technology)

no abstracts in English

JAEA Reports

Preliminary thermal-hydraulic and structural strength analyses of pre-moderator of cold moderator

Aso, Tomokazu; Kaminaga, Masanori; Terada, Atsuhiko*; Hino, Ryutaro

JAERI-Tech 2001-051, 22 Pages, 2001/08

JAERI-Tech-2001-051.pdf:4.51MB

no abstracts in English

JAEA Reports

Study of integrated structure of mercury target container with safety hull

Kaminaga, Masanori; Terada, Atsuhiko*; Haga, Katsuhiro; Kinoshita, Hidetaka; Ishikura, Shuichi*; Hino, Ryutaro

JAERI-Tech 2000-076, 70 Pages, 2001/01

JAERI-Tech-2000-076.pdf:4.01MB

no abstracts in English

JAEA Reports

Study on reduced-moderation water reactor (RMWR) core design; Joint research report, FY1998-1999 (Joint research)

Research Group for Advanced Reactor System; Research Group for Reactor Physics; Research Group for Thermal and Fluid Engineering

JAERI-Research 2000-035, 316 Pages, 2000/09

JAERI-Research-2000-035.pdf:19.81MB

no abstracts in English

JAEA Reports

None

Koshizuka, Seiichi*; *; Okano, Yasushi; *; Yamaguchi, Akira

JNC-TY9400 2000-012, 91 Pages, 2000/03

JNC-TY9400-2000-012.pdf:2.82MB

no abstracts in English

JAEA Reports

A feasibility study of the particle interaction method for the flow regimes with the chemical reaction; (Report under the contract between JNC and Toshiba Corporation)

Shirakawa, Noriyuki*; *; *; *

JNC-TJ9440 2000-008, 47 Pages, 2000/03

JNC-TJ9440-2000-008.pdf:1.96MB

The numerical thermohydraulic analysis of a LMFR component should involve its whole boundaly in order to evaluate the effect of chemical reaction within it. Therefore, it becomes difficult mainly due to computing time to adopt microscopic approach for the chemical reaction directly. Thus, the thermohydraulic code is required to model the chemically reactive fluid dynamics with constitutive correlations. The reaction rate denpends on the binary contact areas between components such as continuous liquids, droplets, solid particles, and bubbles. The contact areas change sharply according to the interface state between components. Since no experiments to study the jet flow with sodium-water chemical reaction have been done, the goal of this study is to obtain the knowledge of flow regimes and contact areas by analyzing the fluid dynamics of multi-pahse and reactive components mechanistically with the particle interaction method. For the first stage of the study, the applicability of this method to the nalysis of a liquid jet into the other liquid pool was investigated. Based on the literatures, we investigated the jet flow mechanisms and analyzed the experiment of a water jet into a gasoline pool. We also analyzed SWAT3/Run19 test, the jet flow in a rod bundle, to study the applicability of the method to a complicated boundary without a chemical reaction model. The calculated fluid dynamics was in good agreement with the experiment. Furthermore, we studied and formulated the paths of phase change and chemical reaction, and conceptually designed the adopting the heat-transfer-limited phase change model and the synthesizd reaction model with a water-hydrogen conversion ratio.

JAEA Reports

Study of flow characteristics in a cold moderator, 3; Flow Pattern measurement and analysis with flat model

Aso, Tomokazu; Kaminaga, Masanori; Terada, Atsuhiko*; Hino, Ryutaro

JAERI-Tech 2000-018, p.49 - 0, 2000/03

JAERI-Tech-2000-018.pdf:3.24MB

no abstracts in English

JAEA Reports

Numerical Investigation on Thermal Stratification and Striping Phenomena in Various Coolants

Yang Zumao*;

JNC-TN9400 2000-009, 81 Pages, 2000/02

JNC-TN9400-2000-009.pdf:47.3MB

It is important to study thermal stratification and striping phenomena for they can induce thermal fatigue failure of structures. This presentation uses the AQUA code, which has been developed in Japan Nuclear Cycle Development Institute (JNC), to investigate the characteristics of these thermal phenomena in water, liquid sodium, liquid lead and carbon dioxide gas.There are altogether eight calculated cases with same Richardson number and initial inlet hot velocity in thermal stratification calculations, in which four cases have same velocity difference between inlet hot and cold fluid, the other four cases with same temperature difference. The calculated results show : (1) The fluid's properties and initial conditions have considerable effects on thermal stratification, which is decided by the combination of such as thermal conduction, viscous dissipation and buoyant force, etc., and (2) The gas has distinctive thermal stratification characteristics from those of liquid because for

Journal Articles

Thermal-hydraulic analysis of liquid lithium target for international fusion materials irradiatioin facility (IFMIF)

Ida, Mizuho*; Nakamura, Hideo; Nakamura, Hiroo; Takeuchi, Hiroshi

Proceedings of Japan-US Workshop on High Heat Flux Components and Plasma Surface Interactions for Next Fusion Devices, p.59 - 68, 2000/00

no abstracts in English

JAEA Reports

Post processing system for multi-dimensionaI thermal-hydraulic analyses

Miyake, Yasuhiro*; *; ; Kimura, Nobuyuki

JNC-TN9400 2000-016, 40 Pages, 1999/12

JNC-TN9400-2000-016.pdf:3.71MB

ln the conventional visualization system for the computational results, only Japanese (Nihongo) Line Printer (NLP) was available to print two dimensional cross sectional plots of vector and scalar fields. To evaluate the phenomena, an analyst had to print many plots on the NLP. This task makes difficult to check the computational results immediately after the calculation. Recently, as the visualization tools, we introduced Micro AVS and Field View which are utilized widely in the scientific and the industrial fields. ln order to show the numerical results on the visualization software, we constructed a post processing system which convert the results of the numerical code to "lntermediate files" which can be read by the visualization tools. As using this system, the examination of the numerical results can be executed on the display of the personal computer. Furthermore, the persuasive report and paper with high quality can be produced due to the color printing. As for the transient calculation, the change of the phenomena can be visually evaluated by using the animation function.

Journal Articles

Generation of smoke and clogging of ventilation filter under burning of bitumen/salt mixture

Abe, Hitoshi; Takada, Junichi; Tsukamoto, Michio; *; Murata, Mikio

Journal of Nuclear Science and Technology, 36(7), p.619 - 625, 1999/07

 Times Cited Count:0 Percentile:100(Nuclear Science & Technology)

no abstracts in English

JAEA Reports

Study on flow characteristics in cold source moderator, 2; Flow pattern measurement and analysis, thermal-hydraulic analysis in cold source moderator vessel

Aso, Tomokazu; Kaminaga, Masanori; Terada, Atsuhiko*; Hino, Ryutaro

JAERI-Tech 99-049, 45 Pages, 1999/06

JAERI-Tech-99-049.pdf:2.62MB

no abstracts in English

Journal Articles

Preliminary design of mercury target; Return flow type

Hino, Ryutaro; Kaminaga, Masanori; Ishikura, Shuichi*; *; *; *; *; *

Proc. of 14th Meeting of the Int. Collaboration on Advanced Neutron Sources (ICANS-14), 1, p.278 - 287, 1998/00

no abstracts in English

JAEA Reports

Installation of aerosol behavior model into multi-dimensional thermaI hydraulic analysis code AQUA

; Yamaguchi, Akira

PNC-TN9410 98-028, 33 Pages, 1997/12

PNC-TN9410-98-028.pdf:0.93MB

The safety analysis of FBR plant system for sodium leak phenomena needs to evaluate the deposition of the aerosol particle to the components in the plant, the chemical reaction of aerosol to humidity in the air and the effect of the combustion heat through aerosol to the structural component. For this purpose, ABC-INTG (Aerosol Behavior in Containment-INTeGrated Version) code has been developed and used until now. This code calculates aerosol behavior in the gas area of uniform temperature and pressure by 1 cell-model. Later, however, more detailed calculation of aerosol behavior requires the installation of aerosol model into multi-cell thermal hydraulic analysis code AQUA. AQUA can calculate the carrier gas flow, temperature and the distribution of the aerosol spatial concentration. On the other hand, ABC-INTG can calculate the generation, deposition to the wall and flower, agglomeration of aerosol particle and figure out the distribution of the aerosol particle size. Thus, the combination of these two codes enables to deal with aerosol model coupling the distribution of the aerosol spatial concentration and that of the aerosol particle size. AQUA and ABC-INTG were developed separately, therefore, several subroutine were modified and composed. Especially, the interface program which exchanges data between these two codes is important to execute transient calculation. This report describes aerosol behavior model, how to install the aerosol model to AQUA and new subroutine equipped to the code. Furthermore, the test calculations of the simple structural model were executed by this code, appropriate results were obtained. Thus, this code has prospect to predict aerosol behavior by the introduction of coupling analysis with multi-dimensional gas thermo-dynamics for sodium combustion evaluation.

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