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Uchibori, Akihiro; Yanagisawa, Hideki*; Takata, Takashi; Ohshima, Hiroyuki
Nippon Kikai Gakkai Rombunshu (Internet), 86(883), p.19-00353_1 - 19-00353_6, 2020/03
Evaluation of occurrence possibility of tube failure propagation under sodium-water reaction accident is an important issue. In this study, a numerical analysis method to predict occurrence of failure propagation by overheating rupture was constructed to expand application range of an existing computer code. Applicability of the method was constructed through the numerical analysis of the experiment on water vapor discharging in liquid sodium.
Muramatsu, Toshiharu; Sato, Yuji; Kamei, Naomitsu; Aoyagi, Yuji*; Shobu, Takahisa
Nippon Kikai Gakkai Dai-13-Kai Seisan Kako, Kosaku Kikai Bumon Koenkai Koen Rombunshu (No.19-307) (Internet), p.157 - 160, 2019/10
no abstracts in English
Tanaka, Masaaki; Kudo, Yoshiro*; Nakada, Kotaro*; Koshizuka, Seiichi*
Proceedings of 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-18) (USB Flash Drive), p.1473 - 1484, 2019/08
Verification and validation (V&V) including uncertainty quantification on modeling and simulation activities has been very much focused on. Due to increase of requirement for standardization of the procedures on the V&V and prediction process to enhance the simulation credibility, "Guideline for Credibility Assessment of Nuclear Simulations (AESJ-SC-A008: 2015)" was published on July 2016 from the AESJ through ten-year discussion. The paper describes brief history of discussion in the AESJ to the publication and introductory explanation of the procedures in the five major elements and one scheme described in the Guideline. And also, a practical experience of the V&V activity according to the fundamental concept indicated in the Guideline is introduced.
Uesawa, Shinichiro; Suzuki, Takayuki*; Yoshida, Hiroyuki
Konsoryu Shimpojiumu 2018 Koen Rombunshu (Internet), 2 Pages, 2018/08
no abstracts in English
Hirakawa, Moe*; Kikuchi, Yuichiro*; Sakai, Takaaki*; Tanaka, Masaaki; Ohshima, Hiroyuki
Proceedings of 26th International Conference on Nuclear Engineering (ICONE-26) (Internet), 8 Pages, 2018/07
Gas entrainment (GE) from cover gas is one of key issue for Sodium-cooled fast reactors to prevent unexpected effects to core reactivity. By using a computational fluid dynamics (CFD) code, analyses have been conducted to estimate the drifting vortexes on water experiments which were generated as wake vortexes behind a plate obstacle in the circulating water channel. In this paper, the results of comparison between experiments and analyses were discussed and the gas core lengths from the surface vortexes were evaluated by using the evaluation tool named StreamViewer developed by Japan Atomic Energy Agency.
Yoshida, Hiroyuki; Uesawa, Shinichiro; Horiguchi, Naoki; Miyahara, Naoya; Ose, Yasuo*
Dai-23-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (USB Flash Drive), 5 Pages, 2018/06
no abstracts in English
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.
Muramatsu, Toshiharu
Hikari Araiansu, 28(12), p.31 - 35, 2017/12
no abstracts in English
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.
Tanaka, Masaaki; Kobayashi, Jun; Nagasawa, Kazuyoshi*
Dai-22-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (USB Flash Drive), 4 Pages, 2017/06
In JAEA, a numerical simulation code named MUGTHES which can deal with conjugate heat transfer between the fluid and the structure parts has been developed for estimation of the thermal fatigue issue. In fundamental validation, the benchmark analysis was considered using the experiment of planar triple parallel jet sodium test (PLAJEST). Three specific experimental conditions at Vr=1, 1.56, and 5.56 were employed for the benchmark analyses according to the knowledge in the literatures. Through the benchmarks, applicability of the large eddy simulation (LES) approach with the standard Smagorinsky model in MUGTHES to simulate thermal striping phenomena was potentially confirmed and issues to be modified in the future works were indicated.
Muramatsu, Toshiharu
Reza Kenkyu, 44(12), p.799 - 803, 2016/12
no abstracts in English
Horiguchi, Naoki; Yoshida, Hiroyuki; Nakao, Yasuhiro*; Kaneko, Akiko*; Abe, Yutaka*
Proceedings of 10th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-10) (USB Flash Drive), 7 Pages, 2016/11
Muramatsu, Toshiharu
Dai-84-KaiReza Kako Gakkai Koen Rombunshu, p.113 - 116, 2016/01
A general-purpose three-dimensional thermohydraulics numerical simulation code SPLICE was developed at Japan Atomic Energy Agency and designed to deal with gas-liquid-solid consolidated incompressible viscous flows with a phase change process in various laser applications, such as welding, coating, cutting, etc. The result obtained from laser coating simulations is very encouraging in the sense that the SPLICE code would be used as one of efficient front-loading tools for related to the laser coating processes.
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.
Suzuki, Takayuki; Yoshida, Hiroyuki; Abe, Yutaka*; Kaneko, Akiko*
Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 7 Pages, 2015/05
In order to improve the safety of Boiling Water Reactor (BWR), it is required to know the behavior of the plant when an accident occurred. Especially, it is important to estimate the behavior of molten core jet in the lower part of the reactor pressure vessel at a severe accident. In the BWR lower plenum, the flow characteristics of molten core jet are affected by many complicated structures, such as control rod guide tubes, instrument guide tubes and core support plate. The objective of this study is to develop the simulation method for the flow characteristic of molten core jet including the effects of the complicated structures in the lower plenum based on interface tracking method code TPFIT (Two Phase Flow simulation code with Interface Tracking). To verify and validate the applicability of the developed method in detail, it is necessary to obtain the experimental data that can be compared with detailed numerical results by the TPFIT. Therefore, experimental works by use of multi-phase flow visualization technique were also carried out. In the experiments, time series of interface shapes are observed by high speed camera and velocity profiles in/out of the jet were measured by the PIV method. In this paper, we carried out a numerical simulation of the jet breakup phenomena in the multi-channels with various simulant molten materials to evaluate the influence of properties on the jet breakup phenomena. As a result, it was confirmed that density and surface tension affected on the falling down velocity of the simulant materials and the interface behavior of the molten jet. However, viscosities of the simulant materials have small effects on jet breakup phenomena, including the interface shape and size of fragments.
Uchibori, Akihiro; Ohshima, Hiroyuki
Proceedings of 9th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-9) (CD-ROM), 6 Pages, 2014/11
A computer code called SERAPHIM calculating compressible multicomponent multiphase flow with sodium-water chemical reaction has been developed to evaluate wastage environment under tube failure accident in a steam generator of sodium-cooled fast reactors. In this study, the numerical model for liquid droplet entrainment and its transport was developed. The applicability of the model was investigated through the analysis of the basic experiment. It was demonstrated that our numerical model could reproduce the time to end of entrainment and the pressure variation during the occurrence of entrainment.
Jang, S.*; Takata, Takashi; Yamaguchi, Akira*; Uchibori, Akihiro; Kurihara, Akikazu; Ohshima, Hiroyuki
Proceedings of 9th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-9) (CD-ROM), 8 Pages, 2014/11
Numerical quantification of the self-wastage phenomenon has been carried out using a multi-dimensional computational code: SERAPHIM. The width of the completely enlarged crack was investigated in this study. Several steps of numerical calculations were devised to reproduce transient self-wastage phenomenon caused by Sodium Water Reaction (SWR). In the analyses, 2-dimensional calculation was carried out to obtained thermal hydraulic properties in the reaction zone. The wastage amount was evaluated based on hypothetical Arrhenius equation by using the temperature and molar concentration of Sodium hydroxide. New analytical grid was created by exchanging the solid cells to fluid cells in the reaction based on the wastage amount evaluation. These series of procedure have been repeated. The width and the shape of the enlarged crack showed good agreement with the experimental results.
Muramatsu, Toshiharu; Yamada, Tomonori; Hanari, Toshihide; Takebe, Toshihiko; Nguyen, P. L.; Matsunaga, Yukihiro
JAEA-Research 2014-018, 41 Pages, 2014/09
JAEA-Research-2014-018.pdf:42.21MB
In decommissioning works of the Fukushima Daiichi Nuclear Power Plants, it is required that fuel debris solidifying mixed materials of fuels and in-vessel structures should be removed. The fuel debris is considered to have characteristics, such as indefinite shapes, porous bodies, multi-compositions, higher hardness, etc. from the knowledge in the U.S. and the Three Mile Island nuclear power plant. Laser lights are characterized by higher power density, local processability, remote controllabilitiy, etc. and can be performed thermal cutting and crushing-up for various materials which does not depend on fracture toughness. This report describes a research program and research activities in FY2013 aiming at developing removal system of fuel debris by the use of laser lights.
Tanaka, Masaaki; Takaya, Shigeru; Fujisaki, Tatsuya*
Nippon Kikai Gakkai Kanto Shibu Ibaraki Koenkai 2014 Koen Rombunshu, p.203 - 204, 2014/09
As flow-induced vibration (FIV) in the hot-leg piping with the short elbow (curvature radius corresponds to diameter) is one of the targeted issues in this study, numerical estimation method for dynamic analysis of mechanical stress on piping has been developed. As the preliminary step in development of the fluid-structure mechanical interaction simulation method, time history response analysis of piping by using the time history data of fluid pressure obtained by the unsteady hydraulics simulation as its boundary condition was attempted. Through the numerical results, potential capability of the dynamic analysis of piping were confirmed and unsteady behavior of pipe due to the unsteady flow phenomena was analyzed.
Takase, Kazuyuki; Ose, Yasuo*; Yoshida, Hiroyuki; Akimoto, Hajime; Aoki, Takayuki*
Dai-24-Kai Nippon Shimyureshon Gakkai Taikai Happyo Rombunshu, p.161 - 164, 2005/07
no abstracts in English
