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Kurihara, Akikazu; Umeda, Ryota; Shimoyama, Kazuhito; Abe, Yuta; Kikuchi, Shin; Ohshima, Hiroyuki
Nihon 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
Nihon 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*
Nihon 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
Nihon 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
Nihon 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*
Nihon 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
Nihon 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
Nihon 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
Nihon 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.
Kino, Chiaki; Watanabe, Tadashi*; Nishida, Akemi; Takemiya, Hiroshi
Nihon 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
Nihon 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
Nihon 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
Nihon 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*
Nihon 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
Nihon 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.
Yada, Hiroki; Kanagawa, Akihiro*; Hattori, Shuji*
Nihon Kikai Gakkai Rombunshu, B, 78(788), p.811 - 820, 2012/04
no abstracts in English
Kikuchi, Shin; Kurihara, Akikazu; Ohshima, Hiroyuki
Nihon Kikai Gakkai Rombunshu, B, 78(787), p.475 - 479, 2012/03
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. The parameters, including melting points of Na and NaOH, transition temperature of NaOH, Na-NaOH reaction temperature, and the decomposition temperature of sodium hydride (NaH) have been identified from DTA curves. Based on the measured reaction temperature, rate constant of sodium monoxide (NaO) generation was obtained. Thermal analysis results indicated that NaO generation at the secondary overall reaction would be considered during the sodium-water reaction.
Hiyama, Tomoyuki; Doda, Norihiro; Ohshima, Hiroyuki; Iwasaki, Takashi*
Nihon Kikai Gakkai Rombunshu, B, 78(787), p.468 - 470, 2012/03
An analysis of JOYO MK-II natural circulation test has been performed to evaluate the applicability of the fast reactor plant dynamics code super-COPD to natural circulation decay heat removal phenomena. The analysis domain is set from the reactor core to the intermediate heat exchanger (IHX) so as to focus on the simulation accuracy of natural circulation behavior in the primary system and we compared the numerical results with experimental measurement. As a result, it was found that natural circulation behavior is much influenced by the coolant mixing regime in the upper plenum of the reactor vessel. The predicted transient changes of the core outlet coolant temperature and the primary flow rate showed good agreement with the test results by using a variable mesh partitioning method and by setting appropriate mixing volume in the upper plenum region which can include the effect of thermal stratification phenomena on the natural circulation behavior.
Doda, Norihiro; Ohshima, Hiroyuki; Kamide, Hideki; Watanabe, Osamu*
Nihon Kikai Gakkai Rombunshu, B, 78(787), p.465 - 467, 2012/03
In the design study for Japan Sodium-cooled Fast Reactor (JSFR), fully natural circulation system is adopted as the decay heat removal system. We have been developing a new evaluation method of core hot spot in transition from rated operation to natural circulation decay heat removal conditions. Since the method is currently based on conservative assumptions and data, there is room for further rationalization of the safety margin which can be achieved by conducting best estimate analyses with confidence and with quantified uncertainty of results. This paper describes a development of PIRT (Phenomena Identification and Ranking Table) for JSFR under natural circulation decay heat removal operations and the sensitivity analyses of the uncertainties in the event of loss of external power as the first step to improve the evaluation method.
Ito, Kei; Kunugi, Tomoaki*; Ohshima, Hiroyuki
Nihon Kikai Gakkai Rombunshu, B, 78(786), p.254 - 262, 2012/02
The cavitation due to sub-surface vortex, which can damage structural surfaces of a large-scale sodium-cooled fast reactor in Japan, have been studied. The authors are developing a CFD code to simulate the cavitation accurately. In this paper, as the first part of the development, a high-precision simulation scheme for vortical flows is presented. In this scheme, a vortical flow is identified in each cell and a vortical velocity distribution is determined locally to be consistent with the local velocity distribution. Then, the calculations of momentum transport through cell faces are performed in consideration with the vortical velocity distribution. As a fundamental verification of the developed scheme, inviscid vortex attenuation in two-dimensional domain is simulated. As a result, it is confirmed that the developed scheme can simulate vortical flows more accurately than conventional schemes.
