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Kikuchi, Norihiro; Mori, Takero; Okajima, Satoshi; Tanaka, Masaaki; Miyazaki, Masashi
Dai-26-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (Internet), 5 Pages, 2022/07
In JAEA, the design optimization method for plant structure has been developed on the process to output optimal solution of the thickness of reactor vessel wall against thermal transient and seismic loads in a SFR as a representative problem. Resistance characteristic of the wall on the load derived from thermal transient is one of the most important factors for safety estimation on the structural integrity. Failure probability of components against thermal transient was set to one of variables in the objective function for a common scale to compare with other variables in different failure mechanisms. In the iterative process to achieve the optimal solution, a number of evaluations to measure the influence on the load derived from thermal transient was necessarily conducted. More reduction of required time for evaluations is desired. To perform the iterative evaluation process efficiently, the automatization of parametric analyses was implemented in the optimization process.
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.
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.
Tanaka, Masaaki; Ono, Ayako; Hamase, Erina; Ezure, Toshiki; Miyake, Yasuhiro*
Nihon 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).
Tanaka, Masaaki; Ohno, Shuji; Ohshima, Hiroyuki
Nuclear Engineering and Design, 299, p.174 - 183, 2016/04
Times Cited Count:4 Percentile:36.27(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.
Sato, Hiroshi; Aso, Tomokazu; Kogawa, Hiroyuki; Teshigawara, Makoto; Hino, Ryutaro
JAERI-Tech 2004-018, 23 Pages, 2004/03
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.
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:43 Percentile:91.32(Nuclear Science & Technology)no abstracts in English
Aso, Tomokazu; Kaminaga, Masanori; Terada, Atsuhiko*; Hino, Ryutaro
Nihon Genshiryoku Gakkai-Shi, 43(11), p.1149 - 1158, 2001/11
Times Cited Count:0 Percentile:0.01(Nuclear Science & Technology)no abstracts in English
Aso, Tomokazu; Kaminaga, Masanori; Terada, Atsuhiko*; Hino, Ryutaro
JAERI-Tech 2001-051, 22 Pages, 2001/08
no abstracts in English
Kaminaga, Masanori; Terada, Atsuhiko*; Haga, Katsuhiro; Kinoshita, Hidetaka; Ishikura, Shuichi*; Hino, Ryutaro
JAERI-Tech 2000-076, 70 Pages, 2001/01
no abstracts in English
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
no abstracts in English
Koshizuka, Seiichi*; *; Okano, Yasushi; *; Yamaguchi, Akira
JNC TY9400 2000-012, 91 Pages, 2000/03
no abstracts in English
Shirakawa, Noriyuki*; *; *; *
JNC TJ9440 2000-008, 47 Pages, 2000/03
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.
Aso, Tomokazu; Kaminaga, Masanori; Terada, Atsuhiko*; Hino, Ryutaro
JAERI-Tech 2000-018, p.49 - 0, 2000/03
no abstracts in English
Yang Zumao*;
JNC TN9400 2000-009, 81 Pages, 2000/02
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
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
Miyake, Yasuhiro*; *; ; Kimura, Nobuyuki
JNC TN9400 2000-016, 40 Pages, 1999/12
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.
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:0.01(Nuclear Science & Technology)no abstracts in English
Aso, Tomokazu; Kaminaga, Masanori; Terada, Atsuhiko; Hino, Ryutaro
JAERI-Tech 99-049, 45 Pages, 1999/06
no abstracts in English
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