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Matsushita, Kentaro; Ezure, Toshiki; Imai, Yasutomo*; Fujisaki, Tatsuya*; Tanaka, Masaaki
Nihon Kikai Gakkai Kanto Shibu Ibaraki Koenkai 2021-Nendo Koen Rombunshu (Internet), 4 Pages, 2021/08
For evaluation of gas entrainment phenomenon at free surface in reactor vessel of sodium-cooled fast reactor, the gas entrainment evaluation tool named "Stream Viewer" has been developed. In Stream Viewer, depth of surface vortex dimple is predicted by calculating pressure decrease at the vortex center using velocity distribution around the vortex and Burgers vortex model. In this report, a method to identify continuous vortex center lines from a velocity distribution is newly developed. It becomes possible to evaluate three-dimensional distribution of pressure decrease along vortex center line. Then, the method is validated by applying Stream Viewer to an open channel experiment. As the result, it was confirmed that vortex center lines were successfully identified by the improved Stream Viewer. Moreover, it was also shown that the evaluation accuracy of gas entrainment was expected to be improved by considering distribution of pressure decrease along vortex center line.
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).
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.
Kikuchi, Norihiro; Ohshima, Hiroyuki; Imai, Yasutomo*; Hiyama, Tomoyuki; Nishimura, Masahiro; Tanaka, Masaaki
Nihon 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.
Tanaka, Masaaki; Takaya, Shigeru; Fujisaki, Tatsuya*
Nihon 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.
Jiao, L.; Yoshida, Hiroyuki; Takase, Kazuyuki
Nihon Kikai Gakkai Kanto Shibu Ibaraki Koenkai 2013 Koen Rombunshu, p.107 - 108, 2013/09
In Japan Atomic Energy Agency, the detailed two-phase flow analysis code TPFIT has been developed to simulate and evaluate two-phase flow characteristics in nuclear systems. In this study, a numerical simulation of bubbly flow in a vertical circular pipe was performed to validate the applicability of TPFIT code on bubbly flow simulation. By checking bubble distribution development in the flow direction, the calculation of the forces acting on bubbles was validated through comparing simulation results and experimental results from Lucas et al. (2005), who used wire-mesh sensor to obtain a high resolution of the gas fraction data in space as well as in time.
short elbow piping systems of JSFRTanaka, Masaaki; Fujisaki, Tatsuya*; Oki, Hiroshi*
Nihon Kikai Gakkai Kanto Shibu Ibaraki Koenkai 2013 Koen Rombunshu, p.141 - 142, 2013/09
Numerical simulation method to analyze unsteady flow phenomena in the upper plenum and the hot-leg piping system in Japan Sodium cooled Fast Reactor (JSFR) has been developed. A flow-induced vibration in the hot-leg piping is mainly targeted in this study. A numerical model of the hot-leg piping integrated into a full domain of the upper plenum was constructed. By comparing with the existing experimental results, applicability of numerical model was confirmed.
Kamiji, Yu; Terada, Atsuhiko; Hino, Ryutaro
Nihon Kikai Gakkai Kanto Shibu Ibaraki Koenkai 2012 Koen Rombunshu, p.63 - 64, 2012/08
Low-speed up-flow is generated due to the water radiolysis in the spent vessel used by contaminated water treatment facility of Fukushima Dai-ichi Nuclear Power Plant. Though it is important to understand the inside flow of the vessel for storage, it is not clear. Thus flow test was performed with a simulated acrylic cylindrical vessel. This paper reports results of visualization of low-speed upflow in a cylindrical vessel by using LASER light sheet and high speed camera. It was confirmed that stagnant flow was generated around the stepped section, and the stagnant flow repeated appearance and dissipation. It was also found that the behavior of the stagnation depended on gas flow rate.
Noguchi, Hiroki; Kubo, Shinji; Onuki, Kaoru
Nihon Kikai Gakkai Kanto Shibu Ibaraki Koenkai 2010 Koen Rombunshu, p.77 - 78, 2010/08
The Japan Atomic Energy Agency (JAEA) has been conducting R&D on thermo-chemical Iodine-Sulfur (IS) process, which is one of most attractive water-splitting hydrogen production methods using nuclear heat of a high temperature gas-cooled reactor (HTGR). Direct contact heat exchanger (DCHX) is considered to apply for sulfuric acid concentration. DCHX is a contactor between low temperature fluid and high temperature fluid. It have several advantages over surface-type heat exchanger in terms of its simple structure and efficient heat exchange due to lager heat transfer surface area and no surface fouling to deteriorate the heat transfer. This paper discusses an experimental method for the measurement of mass transfer coefficient of H
SO
and HO, which is an important factor for the optimum design of DCHX.
Ando, Masanori; Tezuka, Taiji*; Sato, Keisuke*; Kawasaki, Nobuchika; Enuma, Yasuhiro*; Tsukimori, Kazuyuki
Nihon Kikai Gakkai Kanto Shibu Ibaraki Koenkai 2009 Koen Rombunshu, p.171 - 172, 2009/08
no abstracts in English
Baba, Shinichi; Shibata, Taiju; Yamaji, Masatoshi*; Sumita, Junya; Ishihara, Masahiro; Motohashi, Yoshinobu*; Sawa, Kazuhiro
Nihon Kikai Gakkai Kanto Shibu Ibaraki Koenkai (2004) Koen Rombunshu (No.040-3), p.61 - 62, 2004/09
no abstracts in English
Shibata, Taiju; Motohashi, Yoshinobu*; Ishihara, Masahiro; Baba, Shinichi; Ito, Tsutomu*; Sawa, Kazuhiro
Nihon Kikai Gakkai Kanto Shibu Ibaraki Koenkai (2004) Koen Rombunshu (No.040-3), p.63 - 64, 2004/09
no abstracts in English
O
joints bonded using superplastic ceramics as interlayerSato, Takashi*; Motohashi, Yoshinobu*; Sakuma, Takaaki*; Waseda, Kazuyoshi*; Shibata, Taiju; Ishihara, Masahiro; Sawa, Kazuhiro
Nihon Kikai Gakkai Kanto Shibu Ibaraki Koenkai (2004) Koen Rombunshu (No.040-3), p.55 - 56, 2004/09
no abstracts in English
Kikuchi, Makoto*; Motohashi, Yoshinobu*; Ito, Tsutomu*; Sakuma, Takaaki*; Shibata, Taiju; Baba, Shinichi; Ishihara, Masahiro; Sawa, Kazuhiro; Hojo, Tomohiro*; Tsuji, Nobumasa*
Nihon Kikai Gakkai Kanto Shibu Ibaraki Koenkai (2004) Koen Rombunshu (No.040-3), p.57 - 58, 2004/09
no abstracts in English
Aso, Tomokazu; Sato, Hiroshi; Kaminaga, Masanori; Hino, Ryutaro; Monde, Masanori*
Nihon Kikai Gakkai Kanto Shibu Ibaraki Koenkai (2003) Koen Rombunshu (No.030-3), p.45 - 46, 2003/09
no abstracts in English
Sato, Hiroshi; Aso, Tomokazu; Kogawa, Hiroyuki; Hino, Ryutaro
Nihon Kikai Gakkai Kanto Shibu Ibaraki Koenkai (2003) Koen Rombunshu (No.030-3), p.47 - 48, 2003/00
no abstracts in English
Koyama, Tomofumi*; Futakawa, Masatoshi; Kogawa, Hiroyuki; Ishikura, Shuichi*
Nihon Kikai Gakkai Kanto Shibu Ibaraki Koenkai (2002) Koen Rombunshu, p.5 - 6, 2002/09
no abstracts in English
Naoe, Takashi*; Futakawa, Masatoshi; Wakui, Takashi*; Kogawa, Hiroyuki
Nihon Kikai Gakkai Kanto Shibu Ibaraki Koenkai (2002) Koen Rombunshu, p.3 - 4, 2002/09
no abstracts in English
Yamamoto, Junichiro*; Futakawa, Masatoshi; Kurata, Yuji; Naoe, Takashi*
Nihon Kikai Gakkai Kanto Shibu Ibaraki Koenkai (2002) Koen Rombunshu, p.1 - 2, 2002/09
no abstracts in English
Shibata, Taiju; Ishihara, Masahiro; Motohashi, Yoshinobu*; Baba, Shinichi; Hoshiya, Taiji; Sakuma, Takaaki*
Nihon Kikai Gakkai Kanto Shibu Ibaraki Koenkai (2002) Koen Rombunshu (No.020-3), p.123 - 124, 2002/09
Applications of superplastic ceramics to nuclear energy fields are very promising because of their superior mechanical and thermal properties. In this study, Zr ion irradiation effects on superplastic deformation behavior of a typical superplastic ceramic, 3mol% yttria containing tetragonal zirconia polycrystals (3Y-TZP), were studied. Zr
ions with 130MeV were irradiated on the 3Y-TZP specimens by TANDEM accelerator at fluence levels of 3.5x10
and 2.1x10
ions/m
. Their mechanical properties were examined by three-point bending test at elevated temperatures after the irradiation. As a result, at the temperature range from the room temperature to 1473K, any obvious irradiation effects were not found by the bending test. On the other hand, at the range from 1623K to 1773K, it was found that the superplastic deformation behavior was changed by the Zr ion irradiation. The activation energy of superplastic deformation was increased with increasing the irradiation fluence.