Kurihara, Akikazu; Umeda, Ryota; Shimoyama, Kazuhito; Abe, Yuta; Kikuchi, Shin; Ohshima, Hiroyuki
Nippon Kikai Gakkai Rombunshu, B, 79(808), p.2640 - 2644, 2013/12
Overheating tube rupture of adjacent tubes arises from water/steam leak in steam generators of sodium-cooled fast reactors. It is very important to predict the tube wall stress (tube wall temperature) with a high degree of accuracy on evaluation of overheating tube rupture, and is crucial to estimate quantitatively the heat transfer coefficient between reaction jet and adjacent tubes which is one of the major influencing factor. The authors carried out the sodium-water reaction test (SWAT-1R) under the simulated operation condition of a real plant, and measured the correlation between heat transfer coefficient and void fraction around an adjacent tube. The authors confirmed that thermal environment around an adjacent tube was inferable from measured data, and heat transfer correlation equation proposed by Hamada et al. was applicable to the operation condition at elevated pressure and temperature.
Uchibori, Akihiro; Kikuchi, Shin; Kurihara, Akikazu; Hamada, Hirotsugu; Ohshima, Hiroyuki
Nippon Kikai Gakkai Rombunshu, B, 79(808), p.2635 - 2639, 2013/12
Multi-physics analysis system for a heat transfer tube failure event in a steam generator of sodium-cooled fast reactors has been developed. In this study, applicability of the newly constructed numerical models in the analysis system was investigated. The droplet entrainment / transport model which was incorporated into the SERAPHIM code was verified through the analysis of the related experiment. The experimental data about the pressure variation when the droplet entrainment occurs was reproduced by our model successfully. The TACT code is integrated by the numerical models of fluid-structure thermal coupling, stress evaluation and failure judgment of the structure. The fluid-structure thermal coupling model could predict the temperature distribution formed by the flow around the circular cylinder. About the failure judgment model, the predicted time of failure occurrence showed good agreement with the results of the tube rupture simulation experiment.
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.
Kikuchi, Shin; Seino, Hiroshi; Ohno, Shuji
Nippon Kikai Gakkai Rombunshu, B, 79(808), p.2650 - 2654, 2013/12
For countermeasure against sodium leak, structural concrete is protected by steel liner in a sodium-cooled fast reactor (SFR). However, if considering severe and unexpected accidental condition such as breach of steel liner by intensive sodium leak, the reaction with liquid sodium and concrete potentially may occur. For the purpose of elucidating the mechanism of the sodium-concrete reaction in SFRs, kinetic study of the sodium (Na)-silica (SiO) reaction has been carried out by Differential Scanning Calorimetry (DSC) technique. The Na-SiO reaction temperaturewas identified from DSC curves. It was found that reactivity of Na-SiO reaction is similar with the reaction between Na and aggregate of practical used concrete. Based on the measured reaction temperature, rate constant of Na-SiO reaction was obtained. Thermal analysis results indicated that Na-SiO reaction could occur under the elevated temperature in the timeframe of sodium-concrete reaction.
Hayakawa, Satoshi*; Watanabe, Osamu*; Ito, Kei; Yamamoto, Tomohiko
Nippon Kikai Gakkai Rombunshu, B, 79(808), p.2645 - 2649, 2013/12
As the practical evaluation method of the effect of tsunami on buildings, the formula of tsunami force has been used. However, it cannot be applied to complex geometry of buildings. In this study, to analyze the effect of tsunami on the buildings of sodium-cooled fast reactor plant more accurately, three-dimensional tsunami analysis was performed. In the analysis, VOF (Volume of Fluid) method was used to capture free surface of tsunami. At the beginning, it was confirmed that the tsunami experiment results was reproduced by VOF method accurately. Next, the three-dimensional tsunami analysis was performed with VOF method to evaluate the flow field around the buildings of the plant from the beginning of the tsunami until the backwash of that.
Ishiyama, Shintaro; Mamitani, Masataka*; Kondo, Mitsunori*; Hiki, Akinori*
Nippon Kikai Gakkai Rombunshu, A, 79(806), p.1504 - 1516, 2013/10
For the purpose of removal of Cs from highly densified contaminate soil produced after contaminated soil washing process, reclamation soil techniques of contaminated soil through use of structural instability of mordenite contained in contaminated soil under high temperature heating was testified and the following results were obtained; (1) Structural instability of Cs-typed typical native mordinite yielding in Tohoku area in Japan was observed and dissociation of Cs from this mordenite was found after heat annealing at 1173 K. (2) Remarkable dissociation rate in the early stages was observed and then was slowed down afterword. Dissociation rate of Cs reached up to 65% after 1173 K 1 hr annealing. (3) 100% recovery of Cs from vaporized Cs gas was achieved by technology combination of electric precipitation FP filter newly developed and cold trap.
Minagawa, Keisuke*; Fujita, Satoshi*; Yamaguchi, Akira*; Takata, Takashi*; Kurisaka, Kenichi
Nippon Kikai Gakkai Rombunshu, C, 79(804), p.2684 - 2693, 2013/08
Yamaguchi, Yoshihito; Li, Y.*; Katsuyama, Jinya; Onizawa, Kunio
Nippon Kikai Gakkai Rombunshu, A, 79(802), p.730 - 734, 2013/06
In this study, an effect of excessive tensile and compressive loading on the crack growth behavior of piping has been evaluated through cracked plate testing. It was observed that excessive loading had changed crack growth rate. Effect on the crack growth behavior of excessive load was evaluated focusing on a crack blunting and a stress distribution near the crack tip. The crack growth evaluation method under seismic loading was proposed. Four-point bending seismic loading tests were performed using piping specimen in order to confirm the applicability of the proposed method to the piping. It was indicated that the crack growth behavior on piping could be evaluated conservatively using the proposed method, while the crack growth behavior due to seismic loading is evaluated non-conservatively by the existing method.
Ishiyama, Shintaro; Mamitani, Masataka*; Kondo, Mitsunori*
Nippon Kikai Gakkai Rombunshu, B, 79(802), p.1106 - 1121, 2013/06
Contaminated soil washing test was implemented in Fukushima for the purpose of performance verification and set-up of engineering database for construction of contaminated soil washing plant. Chemical interface controlled dispersion approach and high-speed shearing washing techniques were applied to wash the contaminated soil over 10000 Bq/kg and the following results were obtained in this first field test (1) Remarkable washing performance was achieved under the combination of dispersion and high-speed shearing washing conditions with bead cracking technique in the particle grain size group over 0.5 mm and enhanced by ceramic bead addition. (2) Well-ordered structural refinement and cumulation of the contaminated soil particle groups with strong FP confinement were observed at the particle size of 0.5 to 10 m after washing the contaminated soil particle group under 0.5 mm, and sieve classification technique can be applied for elective elimination of this soil particle groups.
Ito, Kei; Ezure, Toshiki; Ohno, Shuji; Kamide, Hideki
Nippon Kikai Gakkai Rombunshu, B, 79(801), p.838 - 847, 2013/05
The prevention of vortex cavitation is one of key issues in the study on a large-scale sodium-cooled fast reactor in Japan (JSFR). In this paper, the theoretical modeling of an unsteady vortex cavitation is proposed. In this model, the cavitation behavior in a strong sub-surface vortex is calculated based on the axisymmetric unsteady Navier-Stokes (N-S) equation. Then, the cavitating radius is determined by comparison of the calculated pressure with vapor pressure. As a basic test of this model, vortex development and resulting cavitation are calculated. The results show that the growth of the cavitating region with the development of circumferential can be calculated.
Kikuchi, Shin; Seino, Hiroshi; Kurihara, Akikazu; Ohshima, Hiroyuki
Nippon Kikai Gakkai Rombunshu, B, 79(799), p.271 - 275, 2013/03
For the purpose of elucidating the mechanism of the sodium-water surface reaction in a steam generator of sodium-cooled fast reactors, kinetic study of the sodium (Na)-sodium hydroxide (NaOH) reaction has been carried out by using Differential Thermal Analysis (DTA) technique. It was reconfirmed that sodium monoxide (NaO) generation should be considered during the sodium-water reaction in spite of variation of volume fraction (Na:NaOH). Na, NaOH and NaO as major chemical species were identified from the X-ray diffraction (XRD) analysis of the residues after the DTA experiment. From XRD analysis, it seems that NaO is reaction product and reaction ratio is less than 100 percent.
Uchibori, Akihiro; Ohshima, Hiroyuki
Nippon Kikai Gakkai Rombunshu, B, 79(799), p.263 - 266, 2013/03
In order to establish a safety evaluation method of a steam generator of sodium-cooled fast reactors, a computer program called SERAPHIM calculating compressible multicomponent multiphase flow with sodium-water chemical reaction under tube failure accident has been developed. In this study, a numerical model for liquid droplet entrainment from an interface of the gaseous jet and its transport was newly constructed to evaluate the environment of the liquid droplet impingement erosion. The applicability of the SERAPHIM program which incorporates the droplet entrainment / transport model was demonstrated through the analysis of vertical discharging of water vapor in the liquid sodium pool under the actual condition of the steam generator.
Ishiyama, Shintaro; Yamaguchi, Katsuhiko*
Nippon Kikai Gakkai Rombunshu, C, 79(799), p.718 - 725, 2013/03
Prototype remotely operated vehicle equipped with high resolution TV monitor and underwater radiation measuring instrument was developed and actual condition of contaminated soil in the regulation pond in Fukushima was investigated. Remarkable contaminated point was observed beneath rainwater outfall of impounding reservoir (Length Width Depth, 112m 30m 3.5m) in the order of 3.3 to 5.24 Sv/h. It is found that contaminated small particles in the order of 1 to 3mm are deposited in the polluted mud at the bottom of the reservoir and suspended by disturbance of water flow and the contaminated zones in the order of 0.75 to 3.5 Sv/h are distributed in a patchy fashion caused by confluent of outfall in the demarcated region (10m 10m 10m).
Kino, Chiaki; Watanabe, Tadashi*; Nishida, Akemi; Takemiya, Hiroshi
Nippon Kikai Gakkai Rombunshu, B, 78(796), p.2113 - 2126, 2012/12
Flow around an in-line forced oscillating circular cylinder was simulated numerically by using OpenFOAM in order to clarify the mechanism of flow induced vibration. Immersed boundary Method is used to solve the moving boundary. Reynolds number is set to 1000 and the reduced velocity is set to the range from 0 to 10. In the first excitation region, it is shown that negative drag force which is a factor for an in-line oscillation of a circular cylinder comes from contacting between high pressure region and a circular cylinder. The present simulation shows that twin vortex has an important role on the contact phenomena. In the second excitation region, it is shown that time averaged lift drag doesn't become zero on some oscillating conditions. It is considered that a cross-flow oscillation comes from the phenomena.
Murakami, Takahiro*; Eguchi, Yuzuru*; Tanaka, Masaaki; Yamano, Hidemasa
Nippon Kikai Gakkai Rombunshu, B, 78(792), p.1388 - 1391, 2012/08
In a conceptual design of Japan Sodium-cooled Fast Reactor (JSFR), short-elbow pipings with large-diameter are adopted for cooling system, in which the coolant flow causes periodical pressure fluctuation at the short-elbow. However, the mechanism of the periodical pressure fluctuation has not been clarified yet. In this paper, unsteady three-dimensional analysis by a finite element large-eddy-simulation (LES) code is carried out in order to explore the mechanism of the periodical pressure fluctuation in short-elbow pipings, based on visualizations of dynamic flow structure in the numerical results.
Iwamoto, Yukiharu*; Kondo, Manabu*; Ogawa, Shota*; Tanaka, Masaaki; Yamano, Hidemasa
Nippon Kikai Gakkai Rombunshu, B, 78(792), p.1383 - 1387, 2012/08
LDV measurements in a 90 degrees elbow which curvature radius coincides with the diameter have been conducted. This paper especially focuses on a result of the deflected inflow, comparing with a result of the short pipe. The result shows that the deflected inflow reinforced a convex velocity distribution occurring near the curvature inside in the downstream region, concluding that the deflected inflow promotes the secondary flow of Prandtl's first kind in the elbow. Its Strouhal number increases to 0.6 from 0.5, compared with the short pipe case. Results of frequency analyses are also shown for other cases that we have been examined. Dominant Strouhal numbers in most of the cases become 0.5, except for 0.6 in cases of the inflow from the long pipe and deflector. This frequency shift might be related with the boundary layer size and the local flow velocity, since the corresponding fluctuation is caused by vortex shedding from the boundary layer at the elbow inside.
Sago, Hiromi*; Shiraishi, Tadashi*; Watakabe, Hisato*; Xu, Y.*; Aizawa, Kosuke; Yamano, Hidemasa
Nippon Kikai Gakkai Rombunshu, B, 78(792), p.1378 - 1382, 2012/08
A conceptual design study of the Japan Sodium-cooled Fast Reactor (JSFR) is in progress in "the Fast Reactor Cycle Technology Development (FaCT) project", and a two-loop primary system is adopted in order to economize the plant construction cost. In the JSFR the pipe thickness is designed to be considerably thinner and the mean sodium velocity increases. To understand the behavior of flow-induced vibration that is derived from the hydraulic characteristics under high Reynolds number conditions experiments were performed to evaluate and confirm the integrity.
Tanaka, Masaaki; Sago, Hiromi*; Iwamoto, Yukiharu*; Ebara, Shinji*; Ono, Ayako; Murakami, Takahiro*; Hayakawa, Satoshi*
Nippon Kikai Gakkai Rombunshu, B, 78(792), p.1392 - 1396, 2012/08
A study on flow induced vibration in the primary cooling system of Japan Sodium cooled Fast Reactor (JSFR) consisting of large diameter pipe and pipe elbow with short curvature radius ("short-elbow") has been conducted. Flow-induced vibration in the short-elbow is an important issue in design study of JSFR, because it may affect to structural integrity of the pipe. In this paper, unsteady flow characteristics in the JSFR short-elbow pipe related to the large-scale eddy motion were estimated based on knowledge from existing studies for curved pipes and scaled water experiments and numerical simulations for the JSFR hot-leg piping.
Tanaka, Masaaki; Takita, Hiroki*; Monji, Hideaki*; Ohshima, Hiroyuki
Nippon Kikai Gakkai Rombunshu, B, 78(792), p.1462 - 1465, 2012/08
Water experiment and numerical simulation under thermal interaction conditions between fluid and structure were conducted for a T-junction piping system. In the experiment, temperatures at 2 mm from the wall in fluid, on the wall and at 3 mm inside from the wall in structure along a vertical trace line were measured simultaneously. Numerical results indicated that the fluid temperature distribution near the wall was much affected by the thermal interaction with the structure and that fluid-structure thermal interaction was necessarily considered for thermal fatigue estimation in the thermal striping phenomena.
Yamaguchi, Yoshihito; Li, Y.*; Sugino, Hideharu*; Katsuyama, Jinya; Onizawa, Kunio
Nippon Kikai Gakkai Rombunshu, A, 78(789), p.613 - 617, 2012/05
In recent years, Japanese nuclear power plants experienced multiple large earthquakes, such as Niigata-ken Chuetsu-Oki Earthquake in 2007 and the Tohoku District - off the Pacific Ocean Earthquake in 2011. Therefore, it is very important to assess the structural integrity of reactor piping under such a large earthquake when a crack exists in the piping. In this study, an effect of excessive compressive loading on the crack growth behavior of piping materials has been evaluated through cracked plate testing. It was observed that excessive compressive loading had an acceleration effect on crack growth rate. The evaluation method using J-integral to evaluate the crack growth behavior under the condition beyond small-scale yielding has been proposed for the acceleration effect of excessive compressive loading on crack growth rate. It was indicated that the acceleration effect by excessive compressive loading simulating a large earthquake could be evaluated using the proposed method.