Ohira, Hiroaki*; Tanaka, Masaaki; Yoshikawa, Ryuji; Ezure, Toshiki
Annals of Nuclear Energy, 172, p.109075_1 - 109075_10, 2022/07
In order to evaluate the mist behavior in the cover gas region of Sodium-cooled Fast Reactors (SFRs) in good accuracy, turbulent model for Rayleigh-Bnard convection (RBC) was selected, and the Reynolds-averaged number density and momentum equations for mist behavior were developed and incorporated into the OpenFOAM code. In the first stage, the RBC in a simple parallel channel was calculated using Favre-averaged k- SST model. The average temperature and flow characteristics agreed well with results from DNS, LES, and experiments. Then the basic heat transfer experiment simulating the cover gas region of SFRs was calculated using this turbulent model and new mist models. The calculated average temperature distribution in the height direction and the mist mass concentration agreed well with the experimental results. We developed a method that could simulate the mist behavior in turbulent RBC environments and the cover gas region of SFRs with high accuracy.
Aihara, Jun; Yasuda, Atsushi*; Ueta, Shohei; Ogawa, Hiroaki; Honda, Masaki*; Ohira, Koichi*; Tachibana, Yukio
Nihon Genshiryoku Gakkai Wabun Rombunshi, 18(4), p.237 - 245, 2019/12
Development of fabrication and inspection technologies of oxidation resistant fuel element for improvement of safety of high temperature gas-cooled reactors (HTGRs) in severe oxidation accident was carried out. Simulated coated fuel particles (CFPs), alumina particles, were over-coated with mixed powder of Si, C and small amount of resin to form over-coated particles, and over-coated particles were molded and hot-pressed to sinter simulated oxidation resistant fuel elements with SiC/C mixed matrix. Simulated oxidation resistant fuel elements with matrix whose Si/C mole ratio is 1.00 were fabricated. Failure fraction of CFPs in fuel elements is one of very important inspection subjects of HTGR fuel. It is essential that CFPs are extracted from fuel elements without additional failure. Development of method for extraction of CFPs was carried out. Desolation of SiC by KOH method or pressurized acidolysis method should be applied to extraction of CFPs.
Aihara, Jun; Honda, Masaki*; Ueta, Shohei; Ogawa, Hiroaki; Ohira, Koichi*; Tachibana, Yukio
Nihon Genshiryoku Gakkai Wabun Rombunshi, 18(1), p.29 - 36, 2019/03
Japan Atomic Energy Agency carried out development of fabrication technology of oxidation resistant fuel element for improvement of safety of high temperature gas-cooled reactors in serious oxidation accident, based on precursor research in former JAEA. Dummy coated fuel particles (alumina particles) were over-coated with mixed powder of Si, C and small amount of resin to form over-coated particles, and over-coated particles were molded and hot-pressed to sinter dummy oxidation resistant fuel elements with SiC/C mixed matrix. We fabricated dummy oxidation resistant fuel elements with matrix whose Si/C mole ratio (about 0.551) is three times as large as that in precursor research. Si peak was not detected by X-ray diffraction of matrix. Better oxidation resistant was confirmed with oxidation test in 20% O at 1673 K than that of ordinal fuel compact with ordinal graphite/carbon matrix. All dummy coated fuel particles were held in specimen after 10 h oxidation.
Mori, Takero; Ohira, Hiroaki; Sotsu, Masutake; Fukano, Yoshitaka
Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 9 Pages, 2017/04
Since safety measures against severe accidents (SAs) such as a long-term station blackout (SBO) are required for Japanese prototype fast breeder reactor Monju, a validation is necessary for the plant dynamics code during SBO. In order to take into account the phenomena in natural circulation: a heat transfer among subassemblies and a flow redistribution, a whole core model has been developed for the plant dynamics code, Super-COPD. This model has been validated by test results of natural circulation in actual facility. In this study, this whole core model was applied to Monju core to evaluate safety measures against SBO, and the pressure loss model of Monju was validated by comparing with results of the plant trip test from the power of 40%. In addition, an analysis was conducted for SBO to investigate the applicability of this model to Monju. The applicability of this model was confirmed by comparing with analytical results using the model without heat transfer between assemblies.
Doda, Norihiro; Ohira, Hiroaki; Kamide, Hideki
Proceedings of 2016 International Congress on Advances in Nuclear Power Plants (ICAPP 2016) (CD-ROM), p.1618 - 1625, 2016/04
Sodium-cooled fast reactors have been developed aiming at introducing natural circulation decay heat removal systems by utilizing the characteristic of having a large coolant temperature difference between at the inlet and at the outlet of reactor vessel. In this study, as part of validation for core hot spot evaluation method which is required for adoption of natural circulation decay heat removal systems, an analysis of EBR-II (Experimental Breeder Reactor II) shutdown heat removal test using the method was performed. The results demonstrated that the evaluation method sufficiently predicts the whole plant thermal hydraulic behaviors and the maximum coolant temperature in a fuel subassembly during natural circulation decay heat removal operations.
Briggs, L.*; Monti, S.*; Hu, W.*; Sui, D.*; Su, G. H.*; Maas, L.*; Vezzoni, B.*; Partha Sarathy, U.*; Del Nevo, A.*; Petruzzi, A.*; et al.
Proceedings of 16th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-16) (USB Flash Drive), p.3030 - 3043, 2015/08
The International Atomic Energy Agency Coordinated Research Project, "Benchmark Analyses of an EBR-II Shutdown Heat Removal Test" is in the third year of its four-year term. Nineteen participants representing eleven countries have simulated two of the most severe transients performed during the Shutdown Heat Removal Tests program conducted at Argonne's Experimental Breeder Reactor II. Benchmark specifications were created for these two transients, enabling project participants to develop computer models of the core and primary heat transport system, and simulate both transients. In phase 1 of the project, blind simulations were performed and then evaluated against recorded data. During phase 2, participants have refined their models to address areas where the phase 1 simulations did not predict as well as desired the experimental data. This paper describes the progress that has been made to date in phase 2 in improving on the earlier simulations and presents the direction of planned work for the remainder of the project.
Nabeshima, Kunihiko; Doda, Norihiro; Ohshima, Hiroyuki; Mori, Takero; Ohira, Hiroaki; Iwasaki, Takashi*
Proceedings of 16th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-16) (USB Flash Drive), p.1041 - 1049, 2015/08
Natural circulation is one of the most important mechanisms to remove decay heat in the sodium cooled fast reactors from the viewpoint of passive safety. On the other hand, it is difficult to evaluate plant dynamics accurately under low flow natural circulation condition. In this study, Super-COPD has been validated through the application to the analysis of natural circulation tests in the experimental fast reactor JOYO. Almost all plant components in JOYO including four air-coolers were modeled in Super COPD. Furthermore, the full scale modeling of fuel subassembly was also adopted in this analysis. The natural circulation test after reactor scram from 100 MW full power at JOYO was selected and simulated by Super-COPD. The transient behaviors predicted by Super-COPD showed good agreement with the experimental data.
Ohira, Hiroaki; Doda, Norihiro; Kamide, Hideki; Iwasaki, Takashi*; Minami, Masaki*
Proceedings of 2015 International Congress on Advances in Nuclear Power Plants (ICAPP 2015) (CD-ROM), p.2585 - 2592, 2015/05
IAEA's Coordinated Research Project on Benchmark Analyses of Shutdown Heat Removal Test (SHRT) performed at the Experimental Breeder Reactor-II (EBR-II) has been carried out since 2012. The benchmark specifications were provided by the Argonne National Laboratory (ANL) and the model development for thermal-hydraulics codes and/or plant dynamics codes has been conducted by participating organizations. The experimental data were also provided by the ANL after the calculations have been performed as the blind simulation. JAEA participated in this benchmark analyses, and the plant dynamics analysis code; Super-COPD was applied to the SHRT-17 simulation. The calculated inlet temperature of the high pressure plenum agreed well with the test data in all simulation time. Although the Z-pipe inlet temperature and the IHX intermediate outlet temperature had some discrepancy in the first 400 sec. caused by larger mass flow rate of the primary pump and the perfect mixing model of upper plenum, these temperatures and the flow rate agreed well with the measured data after 400 sec. Hence it was concluded the present analytical model could predict the natural circulation in good accuracy.
Passerini, S.*; Carardi, C.*; Grandy, C.*; Azpitarte, O. E.*; Chocron, M.*; Japas, M. L.*; Bubelis, E.*; Perez-Martin, S.*; Jayaraj, S.*; Roelofs, F.*; et al.
Proceedings of 2015 International Congress on Advances in Nuclear Power Plants (ICAPP 2015) (CD-ROM), p.780 - 790, 2015/05
Honda, Kei; Ohira, Hiroaki; Mori, Takero
Proceedings of 10th International Topical Meeting on Nuclear Thermal Hydraulics, Operation and Safety (NUTHOS-10) (USB Flash Drive), 12 Pages, 2014/12
Thermal-hydraulics analyses of the reactor upper plenum had been performed in IAEA/Monju-CRP from 2008 to 2012. However, all of the participants got a temperature distribution which didn't agree the measured data on the thermocouple plug. In this study, we re-evaluated the inlet boundary conditions and performed another analysis. The calculated temperature distribution on the thermocouple plug had good agreement with the measured data. Thermocouples and flow guide tubes are attached over the subassembly outlets. The calculated temperature at the thermocouples agreed with the temperature of the boundary conditions. And the calculated temperature at the thermocouples had good agreement with the measured data. Therefore, the temperature at the thermocouples can be regarded as the temperature of the subassembly outlets. From these results, the inlet conditions are an appropriate ones.
Monti, S.*; Latge, C.*; Long, B.*; Azpitarte, O. E.*; Chellapandi, P.*; Stieglitz, R.*; Eckert, S.*; Ohira, Hiroaki; Lee, J.*; Roelofs, F.*; et al.
Proceedings of 2014 International Congress on the Advances in Nuclear Power Plants (ICAPP 2014) (CD-ROM), p.474 - 481, 2014/04
Ono, Jun; Mori, Takero; Sotsu, Masutake; Ohira, Hiroaki
Proceedings of ASME 2013 International Mechanical Engineering Congress and Exposition (IMECE 2013) (DVD-ROM), 9 Pages, 2013/11
The severe accident evaluation on the EVST of MONJU has ever been performed by one-dimensional flow network code "Super-COPD". However, it is difficult to predict thermal-hydraulics in the EVST accurately because the fluid in the EVST is driven by natural circulation. Thus we have performed multidimensional thermal-hydraulic analysis in order to clarify the thermal-hydraulic behavior and evaluate the appropriateness of the flow network model. As a result, it was noted that the multidimensionality on temperature and velocity in the EVST was small enough and the flow network model would be almost appropriate. It should be noted that flow resistance of the supporting plates or the heat transfer center of the cooling coils should be set conservatively for the safety analysis.
Mori, Takero; Sotsu, Masutake; Honda, Kei; Suzuki, Satoshi*; Ohira, Hiroaki
Journal of Energy and Power Engineering, 7(9), p.1644 - 1655, 2013/09
The prototype fast breeder reactor "MONJU" has an ex-vessel fuel storage system which consists mainly of an ex-vessel fuel storage tank (EVST) and an EVST sodium cooling system. EVST sodium cooling system consists of three independent loops. In this study, an analysis and evaluation of the plant dynamics for the spent fuel and the EVSS structural integrity during an station blackout (SBO) were performed. When the number of cooling loops was not changed and natural circulation occurred in only two loops, the sodium temperature in the EVST increased to approximately 450C. However, the structural integrity of the EVSS was maintained. The analytical results, therefore, help clarify the number of necessary cooling loops for efficient decay heat removal and sodium temperature behavior in an SBO.
Ohira, Hiroaki; Honda, Kei; Sotsu, Masutake
Journal of Energy and Power Engineering, 7(4), p.679 - 688, 2013/04
Ohira, Hiroaki; Honda, Kei; Sotsu, Masutake
Proceedings of 14th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-14) (CD-ROM), 12 Pages, 2011/09
In order to evaluate the upper plenum thermal-hydraulics of the Monju reactor vessel, we have performed detail calculations under the 40% rated power operational condition using high resolution mesh models by a commercial FVM code, FrontFlow/Red. In this study, we applied a high resolution meshes around the flow holes (FHs) on the inner barrel. We mainly made clear that the thermal-hydraulics did not change largely since the flow rates through the FHs were small enough to the total coolant flow rate but were affected largely incase without FHs on the honeycomb structure.
Honda, Kei; Ohira, Hiroaki; Sotsu, Masutake; Yoshikawa, Shinji
Proceedings of 8th International Topical Meeting on Nuclear Thermal-Hydraulics, Operation and Safety (NUTHOS-8) (CD-ROM), 12 Pages, 2010/10
In this study, we calculated the thermal hydraulics of the upper plenum of MONJU by the detailed analysis model using commercial FVM code, FrontFlow/Red. The present analysis model simulates all structures with high resolution meshes. The 1st order upwind and 2nd order central difference scheme were applied to the advection and diffusion terms, respectively. And RNG - model was applied to turbulence modeling. These calculation results indicated that the structures installed in the plenum except for UIS did not affect largely to the temperature and velocity, the flow characteristics in the present results had similar tendencies with porous media approached applied to the UCS region and that the difference between the temperature measured in the UCS region and that of SA outlets is relatively small.
Yamada, Fumiaki; Ohira, Hiroaki
Proceedings of 3rd Joint US-European Fluids Engineering Summer Meeting and 8th International Conference on Nanochannels, Microchannels, and Minichannels (ASME 2010) (CD-ROM), 8 Pages, 2010/08
The advanced flow network models of the RV upper plenum and the IHX inlet plenum of MONJU were explained and validated by the previous SSTs. The whole plant dynamics of MONJU were also predicted using the validated flow networks. The natural circulation experiments both in the PHTS and the SHTS were conducted applying the previous SST conditions. The whole plant dynamics model with the advanced IHX model was also validated by these test results. Through these validations, we concluded that the present plant dynamics model of Super-COPD could simulate the whole plant dynamics in good accuracy, which was applicable to the next SSTs.
Ohira, Hiroaki; Takamatsu, Misao
Proceedings of 17th International Conference on Nuclear Engineering (ICONE-17) (CD-ROM), 8 Pages, 2009/06
In the incident of the experimental fast reactor Joyo on June 2007, little amount of metal powder was estimated to be produced by the contact between the bottom of the upper core structure and the test section of the material testing rig with temperature control. Up to now, no foreign materials were detected by the in-vessel observations. However, a small amount of metal powder which could be placed into the fuel pin bundles was assumed in the present study. Preliminary safety evaluation of the wire-wrapped fuel pin bundles of Joyo with local blockage were performed in the rated power operational condition, an anticipated transient during operation and an accident. A single-phase transient subchannel analysis code ASFRE, which was verified by various local blockage experiments in France and in Japan, was applied to the present evaluations. From these results, it was concluded a small amount of metal powder assumed to be produced by the incident would not affect to the safety operations in rated power conditions, in anticipated transients and in accidents.
Tanahashi, Takahiko*; Miyoshi, Ichiro*; Ara, Kuniaki; Ohira, Hiroaki
JNC TY9400 2004-016, 121 Pages, 2004/08
Investigation of Magnetohydrodynamic (MHD) turbulent model applying Large Eddy Simulation (LES) method was started in FY15 in order to evaluate MHD turbulent behavior on the conditions of high Reynolds numbers and high magnetic Reynolds numbers. In FY15, a proposed Subgrid Scale (SGS) model for direct magnetic fields was applied to Generalized Simplified Marker and Cell (GSMAC) method to Finite Element Method (FEM). A Direct Numerical Simulation (DNS) method was also developed to verify the theory and create a new SGS model.
Ohira, Hiroaki; Ara, Kuniaki
JNC TN9400 2004-045, 75 Pages, 2004/07
In order to investigate the dispersion behavior of metallic nano-particles into liquid sodium and to evaluate the chemical reactivity with water, the dispersion experiments and the water-sodium reaction experiments were carried out by changing the production method of the metallic particles and the metal elements. The following results were obtained from these experiments:1) Sodium D lines were observed from all the nano-fluid vapor and the metals of the nano-particles were detected in the vapor. From these results, it was estimated that the small-sized nano-particles combined with sodium atoms to form complexes and dispersed stably in the sodium.2) The oxide surface layers of the metallic nano-particles produced by condensing the vapor were observed to be negligibly thin compared with the particles produced by a chemical synthesis technique. It was also estimated that the smaller and the uniform sized particles were also produced by advansing the condensing technique.3) Although the temperature changes by the water-nanofluid reactions were in the same levels with those by the water-sodium reactions, the pressure gradients by the water-nanofluid reactions showed smaller values than those by the water-sodium reactions: These results indicated the possibility to reduce the chemical reactivity of the nano-fluids.From above results, we are going to evaluate the chemical reaction by applying the smaller-sized nano-particles in the next experiments.