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Yamano, Hidemasa; Tobita, Yoshiharu
Proceedings of 5th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-5), p.716 - 723, 2006/11
An event-sequence analysis of a core disruptive accident (CDA) has been performed for a carbon-dioxide-gas-cooled fast reactor in order to evaluate the safety design in a feasibility study on commercialized fast reactor cycle systems in Japan. An unprotected loss-of-flow accident has been chosen as a representative CDA initiator. The analysis has showed that molten cladding relocation out of the core caused a power burst, leading to the molten fuel dispersion mainly toward a downward direction, then the reactivity settled down under the sub-critical condition. In the present design, the measure for avoiding the energetic recriticality is not necessary because of the short lower axial blanket (LAB). For extending LAB length aiming at a higher breeding ratio, the fuel penetration length into the LAB region just after core disruption has been estimated by a validated fuel-freezing model. A design measure enhancing early fuel discharge has been proposed for the core adopting the longer LAB length.
Morita, Koji*; Matsumoto, Tatsuya*; Fukuda, Kenji*; Tobita, Yoshiharu; Sato, Ikken; Yamano, Hidemasa
Proceedings of 5th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-5), p.211 - 218, 2006/11
In the present study, a series of experiments was performed for transient behaviors of large-scale bubble with condensation. Characteristics of the bubble behaviors observed in the experiments were estimated through the experimental analyses using the reactor safety analysis code SIMMER-III. SIMMER-III simulations suggest that the noncondensable gas has less inhibiting effect on condensation of large-scale bubbles, in which the gas and liquid phases are dispersively mixed without a buildup of the noncondensable gas. The present study indicates that SIMMER-III can simulate the condensation processes of large-scale bubbles under the effect of noncondensable gas reasonably in sufficient physical details.
Konishi, Kensuke; Kubo, Shigenobu*; Sato, Ikken; Koyama, Kazuya*; Toyooka, Junichi; Kamiyama, Kenji; Kotake, Shoji*; Vurim, A. D.*; Gaidaichuk, V. A.*; Pakhnits, A. V.*; et al.
Proceedings of 5th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-5), p.465 - 471, 2006/11
no abstracts in English
Tobita, Yoshiharu; Yamano, Hidemasa
Proceedings of 5th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-5), p.724 - 729, 2006/11
The present study is intended to draw a rough sketch of CDA event sequences, to understand the characteristics of CDA, and to judge whether the measures to eliminate energetic recriticality are necessary or not for a LBE-cooled fast reactor. The present analysis has analyzed a LBE-cooled mixed-nitride (MN) fuel reactor with 1875 MWth. In this study, an unprotected loss-of-flow (ULOF) with flow having time of 10.0 s has been chosen as a representative CDA initiator. In the analysis, the loss-of-flow event has been simulated by the reduction of inlet pressure, following a quasi steady-state calculation with nominal full power and flow rate. This analysis has showed that the swelling of fuel pellets is the key phenomenon during the CDA of LBE cooled fast reactor. The further accumulation of experimental knowledge on the swelling behavior of fuel, especially for nitride fuel, is essential in improving the accuracy and reliability of CDA analysis.
Shibamoto, Yasuteru; Yonomoto, Taisuke; Nakamura, Hideo
Proceedings of 5th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-5), p.226 - 233, 2006/11
no abstracts in English
Koshizuka, Seiichi*; Liu, J.*; Morita, Koji*; Arima, Tatsumi*; Tobita, Yoshiharu; Yamano, Hidemasa; Ito, Takahiro*; Shirakawa, Noriyuki*; Hosoda, Seigo*; Araki, Kazuhiro*; et al.
Proceedings of 5th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-5), p.472 - 479, 2006/11
A 5-year research project started in FY2005 in the framework of Innovative Nuclear Research and Development Program funded by the Ministry of Education, Culture, Sports, Science and Technology in Japan. A computer code, named COMPASS (Computer Code with Moving Particle Semi-implicit for Reactor Safety Analysis), is being developed using the Moving Particle Semi-implicit (MPS) method for various complex phenomena of severe accidents in fast breeder reactors. Both MOX and metal fuels are considered. Eutectic reactions between the metal fuel and the cladding material are being investigated by molecular dynamics and molecular orbital methods. The molten metal flow with solidification was analyzed by MPS. The elastic analysis of a hexagonal wrapper tube was analyzed by the MPS method as well. The results were compared with an experiment and an calculation using an commercial code. Eutectic reactions were calculated by molecular dynamics and compared with the references. We found that the combination of the above numerical methods was useful for multi-physics and multi-scale phenomena of core disruptive accidents in fast breeder reactors.
Ezure, Toshiki; Kimura, Nobuyuki; Hayashi, Kenji; Kamide, Hideki
Proceedings of 5th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-5), p.94 - 99, 2006/11
A compact sodium cooled reactor is one of important candidates as FBR and has been investigated in the feasibility study of FBR cycle. According to the compact sizing of reactor vessel, gas entrainment at the free surface of sodium coolant becomes one of the significant issues for reactor design, and it is required to clarify the criterion of gas entrainment at free surface and the tolerance. In the present study, some visualization experiments were performed in a water-air system focusing on the gas entrainment due to surface vortex and its transient phenomena. Influences of horizontal velocity were clarified by the visualization. Trends of circulation and length of gas core during transient phenomena of the gas entrainment were also measured by the particle image velocimetry and visualization. It was found that the gas core length extends with time delay to the increase of circulation around the vortex.
Kimura, Nobuyuki; Ezure, Toshiki; Tobita, Akira; Kamide, Hideki
Proceedings of 5th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-5), p.391 - 397, 2006/11
For innovative sodium cooled fast reactor, a compact reactor vessel (R/V) is designed to reduce the construction cost, where sodium flow velocity increases. One of the thermal hydraulic issues in this design is gas entrainment (GE) at a free surface in the R/V. Horizontal plates are set below the free surface to prevent the GE. A water experiment was performed using a partial model with large scale. The objective is to investigate occurrence conditions and the mechanism of the GE. It was found that there were two kinds of the GE and the occurrence conditions were far from the reactor condition. One of the GEs occurred at the wake region of the cold leg pipe due to larger horizontal velocity. Other one broke out at the region between the hot leg pipe and the R/V wall when the coolant level was low and the downward velocity was high. The mechanisms of the GE at the two regions were clarified by the detailed measurement of flow velocity field.
Kamide, Hideki; Ogawa, Hiroshi; Kimura, Nobuyuki; Hayashi, Kenji; Tobita, Akira
Proceedings of 5th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-5), p.421 - 428, 2006/11
Thermal stratification after a scram is one of main thermal loads of a reactor vessel in sodium cooled reactor. Water experiments using an 1/10th scaled model were carried out for an advanced loop type sodium cooled reactor, which was designed by FBR Feasibility Study in Japan. The reactor vessel is highly compact and has an upper inner structure (UIS), which consists of perforated baffle plates and has a slit in radial direction for fuel handling. This slit makes high velocity flow in the UIS. Steep temperature distribution across the stratification interface and temperature fluctuations were found near the UIS slit in the experiment. It was revealed that they were resulted from the impingement of the jet through the slit at the interface. Dominant period of temperature fluctuations at stratification interface was influenced by the jet velocity through the UIS slit and also temperature difference across the interface.
Ito, Kei; Sakai, Takaaki; Ohshima, Hiroyuki; Tanaka, Nobuatsu*
Proceedings of 5th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-5), p.398 - 405, 2006/11
Gas entrainment (GE) phenomena in a basic unsteady GE experiments were numerically surveyed to make clear an applicability of numerical simulations to a GE evaluation. To establish realistic evaluation methods, the numerical simulations were conducted on rather coarse mesh partition to reduce computational costs and the GE behaviors were evaluated by GE related parameters. As the results, it was confirmed that the numerical methods could reproduce the velocity fluctuation data and the GE related parameters even with the coarse mesh partition. In addition, the sensitivities of a turbulent modeling and a mesh partition on the simulation accuracy were discussed. These two factors could not lead significant improvements of the GE related parameters for the GE evaluation in this study. These results showed that the proposed numerical methods were appropriate to evaluate the GE phenomena.
Uchibori, Akihiro; Sakai, Takaaki; Ohshima, Hiroyuki
Proceedings of 5th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-5), p.414 - 420, 2006/11
Numerical analysis of the gas entrainment experiment using the 1/1.8 scaled partial model of the designed upper plenum of a sodium cooled fast breeder reactor was performed. The objective of this work was to investigate capability to simulate the flow field in the upper plenum and to confirm applicability of the newly developed assessment method of gas entrainment. The results of the numerical analysis showed that a vortex periodically appeared around the pipes which exist in the flow passage. The appearance position of the vortex, the calculated velocity profile around the vortex and the calculated circulation agreed with the experimental results. Possibility of gas entrainment occurrence was assessed from the numerical results by using our assessment method. For all the analysis cases, occurrence or non-occurrence of gas entrainment was correctly predicted by our method.
Liu, W.; Onuki, Akira; Kureta, Masatoshi; Takase, Kazuyuki; Akimoto, Hajime
Proceedings of 5th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-5), p.340 - 345, 2006/11
A new reactor concept of Innovative Water Reactor for Flexible Fuel Cycle (FLWR) is under development at Japan Atomic Energy Agency (JAEA). A design for high conversion type FLWR with 1350 MWe has been constructed, which has the conversion ratio of 1.04, a negative void reactivity coefficient, the high burnup of 65 GWd/t and the 15-month operational cycle length. This study investigates the thermal feasibility of the designed FLWR core using transient analysis code TRAC-BF1 which includes the critical power correlation and the pressure drop estimation method verified with experimental data. Core inlet coolant flow rate which ensures the thermal margin of MCPR = 1.3 was derived by the TRAC-BF1 code. Analysis to the postulated severest abnormal transient event-flow decrease event due to a failure of one recirculation pump, was performed using the core flow rate as the initial one. No boiling transition was confirmed to be occurred during the event.
Sakai, Takaaki; Eguchi, Yuzuru*; Monji, Hideaki*; Iwasaki, Takashi*; Ito, Kei; Ohshima, Hiroyuki
Proceedings of 5th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-5), p.406 - 413, 2006/11
Design criteria for gas entrainments (GE) from the liquid surface in a fast breeder reactor system were proposed in this paper for the two types of GE phenomena from a vortex dimple based on a computational fluid dynamics (CFD) method. The first gas entrainment phenomenon is a gas core extension directly to the outlet piping level, which induces large amount of GE to the flow system. The second is continuous bubble detachments from the tip of the vortex dimple. Based on CFD calculations for elemental experiments of the surface vortex, local CFD non-dimensional numbers were defined as the design criteria to prevent GE. In conclusion, it was found that the CFD non-dimensional numbers are useful for the design parameters of GE prevention.
Ohshima, Hiroyuki; Imai, Yasutomo*
Proceedings of 5th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-5), p.480 - 487, 2006/11
A numerical simulation of flow field in a wire-wrapped fuel pin bundle of fast breeder reactor was carried out using a finite element method code SPIRAL under typical operating conditions. In this simulation, the influence of computational mesh scheme on the result and the prediction characteristics of three kinds of high Reynolds number turbulence models, modified k-e model, RNG k-e model and algebraic stress model, were clarified. SPIRAL was also applied to simulate a flow field in a wire-wrapped 19-pin bundle and its result was compared with available experimental data. The predicted distributions of axial flow through the subchannels, cross flow between fuel pins and swirl flow along the wrapper tube wall caused by the existence of wire-spacer were in good agreement with measured ones.
Niwa, Hajime
Proceedings of 5th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-5), p.29 - 31, 2006/11
The progress of a feasibility study on commercialized fast reactor cycle systems in Japan is reported, followed by the introduction of innovative technologies in sodium-cooled MOX-fueled fast reactor selected as a principal concept in the phase 2 study. The roadmap toward commercialization are presented with highlighting the importance of international collaboration.
Sakai, Masayuki*; Tanaka, Nobuatsu*; Ohshima, Hiroyuki
Proceedings of 5th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-5), p.301 - 304, 2006/11
The flowing gas or liquid in the industrial machines, it is important that improving in quality. We analyzed about systems of unsteady gas entrainment with the analysis code for turbulence eddy flow, which based on local inter-scale equilibrium assumption (LISEA). And we speculated about adequacy of LES application and affect of mesh division on unsteady gas entrainment phenomenon. In the result, we obtain the following knowledge about unsteady eddy problems as object. (1) LES affected by mesh division more than DNS. So appropriate result needs to give consideration to mesh division. (2) DNS affected lightly by mesh mutation in the purview of this analysis. (3) In the case of using LES, it needs to set proper parameters carefully.
Kunugi, Tomoaki*; Kawara, Zensaku*; Ose, Yasuo*; Ito, Kei; Sakai, Takaaki
Proceedings of 5th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-5), p.385 - 390, 2006/11
In order to design the compact FBR with higher coolant velocity compared to the conventional reactor designs, it is necessary to clarify a criterion of a cover gas entrainment (GE) from the free-surface of the coolant pool in the reactor vessel to the heat exchanger through the hot leg. Three flow regimes are considered as the GE phenomena: a vortex dimple, a waterfall and a surface disturbance. In this study, to evaluate the GE vortex phenomena: especially for the vortex dimple, the DNS based on the MARS (Multi-interfaces Advection and Reconstruction Solver (Kunugi, 2001)) were performed for simulating the unsteady vortex-shedding experiment accompanied with the GE phenomena (Okamoto et al., 2004). The applicability of the present DNS to predict the onset of the GE vortex phenomena is discussed.