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Doda, Norihiro; Hamase, Erina; Yokoyama, Kenji; Tanaka, Masaaki
Dai-25-Kai Nippon Keisan Kogaku Koenkai Rombunshu (CD-ROM), 4 Pages, 2020/06
With the aim of advancing the design optimization in fast reactors, neutronics and thermal-hydraulics coupled analysis method which can consider the temporal change of neutron flux distribution in the core has been developed. A three-dimensional neutronics analysis code and a plant dynamics analysis code are coupled on a platform using Python programing. In this report, outlines of the coupling method of analysis codes, the results of its application to the actual plant under a virtual accidental condition, and the future development is described.
Kawaguchi, Munemichi; Miyahara, Shinya*; Uno, Masayoshi*
Journal of Nuclear Engineering and Radiation Science, 6(2), p.021305_1 - 021305_9, 2020/04
Sodium-concrete reaction (SCR) is one of the important phenomena during severe accidents in sodium-cooled fast reactors (SFRs) owing to the generation of large sources of hydrogen and aerosols in the containment vessel. In this study, SCR experiments with an internal heater were performed to investigate the chemical reaction beneath the internal heater (800
C), which was used to simulate the obstacle and heating effect on SCR. Furthermore, the effects of the internal heater on the self-termination mechanism were discussed. The internal heater on the concrete hindered the transport of Na into the concrete. Therefore, Na could start to react with the concrete at the periphery of the internal heater, and the concrete ablation depth at the periphery was larger than under the internal heater. The high Na pool temperature of 800C increased largely the Na aerosol release rate, which was explained by Na evaporation and hydrogen bubbling, and formed the porous reaction product layer, whose porosity was 0.54-0.59 from the mass balance of Si and image analyzing EPMA mapping. They had good agreement with each other. The porous reaction products decreased the amount of Na transport into the reaction front. The Na concentration around the reaction front became about 30wt.% despite the position of the internal heater. It was found that the Na concentration condition was one of the dominant parameters for the self-termination of SCR, even in the presence of the internal heater.
Yamano, Hidemasa; Takai, Toshihide; Furukawa, Tomohiro
Nippon Kikai Gakkai Rombunshu (Internet), 86(883), p.19-00360_1 - 19-00360_13, 2020/03
It is necessary to simulate a eutectic melting reaction and relocation behavior of boron carbide (B
C) as a control rod material and stainless steel (SS) during a core disruptive accident in an advanced sodium-cooled fast reactor designed in Japan because the BC-SS eutectic relocation behavior has a large uncertainty in the reactivity history based on a simple calculation. A physical model simulating the eutectic melting reaction and relocation was developed and implemented into a severe accident simulation code. The developed model must be validated by using test data. To validate the physical model, therefore, the visualization tests of SS-BC eutectic melting reaction was carried out by contacting SS melts of several kg with a BC pellet heated up to about 1500 C. The tests have shown the eutectic reaction visualization as well as freezing and relocation of the BC-SS eutectic in upper part of the solidified test piece due to the density separation. Post-test material analyses by using X-ray diffraction and transmission electron microscope techniques have indicated that FeB appeared at the BC-SS contact interface and (Fe,Cr)
B at the top surface of the test piece. Glow discharge optical emission spectrometry has been applied to quantitative analysis of boron concentration distributions. The boron concentration was high at the upper surface and near the original position of the BC pellet.
Uchibori, Akihiro; Yanagisawa, Hideki*; Takata, Takashi; Ohshima, Hiroyuki
Nippon Kikai Gakkai Rombunshu (Internet), 86(883), p.19-00353_1 - 19-00353_6, 2020/03
Evaluation of occurrence possibility of tube failure propagation under sodium-water reaction accident is an important issue. In this study, a numerical analysis method to predict occurrence of failure propagation by overheating rupture was constructed to expand application range of an existing computer code. Applicability of the method was constructed through the numerical analysis of the experiment on water vapor discharging in liquid sodium.
Yamano, Hidemasa; Takai, Toshihide; Furukawa, Tomohiro; Kikuchi, Shin; Emura, Yuki; Kamiyama, Kenji; Fukuyama, Hiroyuki*; Higashi, Hideo*; Nishi, Tsuyoshi*; Ota, Hiromichi*; et al.
Proceedings of International Nuclear Fuel Cycle Conference / Light Water Reactor Fuel Performance Conference (Global/Top Fuel 2019) (USB Flash Drive), p.418 - 427, 2019/09
Eutectic reactions between boron carbide (BC) and stainless steel (SS) as well as its relocation are one of the key issues in a core disruptive accident (CDA) evaluation in sodium-cooled fast reactors. Since such behaviors have never been simulated in CDA numerical analyses, it is necessary to develop a physical model and incorporate the model into the CDA analysis code. This study is focusing on BC-SS eutectic melting experiments, thermophysical property measurement of the eutectic melt, and physical model development for the eutectic melting reaction. The eutectic experiments involve the visualization experiments, eutectic reaction rate experiments and material analyses. The thermophysical properties are measured in the range from solid to liquid state. The physical model is developed for a severe accident computer code based on the measured data of the eutectic reaction rate and the physical properties. This paper describes the project overview and progress of experimental and analytical studies by 2017. Specific results in this paper is boron concentration distributions of solidified BC-SS eutectic sample in the eutectic melting experiments, which would be used for the validation of the eutectic physical model implemented into the computer code.
C addition on thermophysical properties of austenitic stainless steel in a liquid stateFukuyama, Hiroyuki*; Higashi, Hideo*; Yamano, Hidemasa
Proceedings of International Nuclear Fuel Cycle Conference / Light Water Reactor Fuel Performance Conference (Global/Top Fuel 2019) (USB Flash Drive), p.853 - 857, 2019/09
Thermophysical properties of molten mixture of 316L stainless steel (SS316L) and control-rod material (BC) are necessary for the development of computer simulation codes that describe core degradation mechanisms during severe accidents in nuclear power plants involving sodium-cooled fast reactors. The effect of BC addition to SS316L on the solidus and liquidus temperatures were first measured by differential scanning calorimetry. An electromagnetic levitation technique performed in a static magnetic field was used to measure the density, surface tension, normal spectral emissivity, specific heat capacity, and thermal conductivity of molten SS316L and SS316L containing BC. The effects of BC addition to SS316L on the thermophysical properties were studied up to 10 mass%.
CLiu, X.*; Morita, Koji*; Yamano, Hidemasa
Proceedings of International Nuclear Fuel Cycle Conference / Light Water Reactor Fuel Performance Conference (Global/Top Fuel 2019) (USB Flash Drive), p.47 - 51, 2019/09
Investigation of the eutectic reaction in a core disruptive accident of sodium cooled reactor is of importance since reactor criticality will be affected by the change in reactivity after eutectic reaction. In this study, we performed 1st step of validation analysis using a fast reactor safety analysis code, SIMMER-III, with the developed model based on a new series of experiments, where a BC pellet was immersed into a molten stainless steel (SS) pool. The simulation results showed the general behavior of eutectic material formation measured in the experiments reasonably. The eutectic reaction consumes solid BC and liquid SS, and then the liquid eutectic composition is produced at the early stage of reaction due to the high temperature of molten SS. Movement of the eutectic material in the molten pool leads to the redistribution of boron element. Molten SS pool then freezes to solid SS and movement of eutectic material is stopped by surrounding solid SS. Boron concentration in the pool was measured after molten SS freezes into a solid. Simulation results indicate that boron tends to accumulate in the upper part of the molten pool. This is attributed to the buoyancy force acting on lighter boron in the molten SS pool. A parametric study was also conducted by changing the initial temperature of BC pellet and SS to investigate the temperature sensitivity on the eutectic reaction behavior.
Tanaka, Masaaki; Kudo, Yoshiro*; Nakada, Kotaro*; Koshizuka, Seiichi*
Proceedings of 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-18) (USB Flash Drive), p.1473 - 1484, 2019/08
Verification and validation (V&V) including uncertainty quantification on modeling and simulation activities has been very much focused on. Due to increase of requirement for standardization of the procedures on the V&V and prediction process to enhance the simulation credibility, "Guideline for Credibility Assessment of Nuclear Simulations (AESJ-SC-A008: 2015)" was published on July 2016 from the AESJ through ten-year discussion. The paper describes brief history of discussion in the AESJ to the publication and introductory explanation of the procedures in the five major elements and one scheme described in the Guideline. And also, a practical experience of the V&V activity according to the fundamental concept indicated in the Guideline is introduced.
Kawaguchi, Munemichi; Miyahara, Shinya; Uno, Masayoshi*
Journal of Nuclear Science and Technology, 56(6), p.513 - 520, 2019/06
Times Cited Count：1 Percentile：50.57(Nuclear Science & Technology)This study revealed melting points and thermal conductivities of four samples generated by sodium-concrete reaction (SCR). We prepared the samples using two methods such as firing mixtures of sodium and grinded concrete powder, and sampling depositions after the SCR experiments. In the former, the mixing ratios were determined from the past experiment. The latter simulated the more realistic conditions such as the temperature history and the distribution of Na and concrete. The thermogravimetry-differential thermal analyzer (TG-DTA) measurement showed the melting points were 865-942C, but those of the samples containing metallic Na couldn't be clarified. In the two more realistic samples, the compression moldings in a furnace were observed. The observation revealed the softening temperature was 800-840C and the melting point was 840-850C, which was 10-20C lower than the TG-DTA results. The thermodynamics calculation of FactSage 7.2 revealed the temperature of the onset of melting was caused by melting of the some components such as NaSiO
and/or NaSiO. Moreover, the thermal conductivity was 
=1-3W/m-K, which was comparable to xNaO-1-xSiO (x=0.5, 0.33, 0.25), and those at 700C were explained by the equation of 
.
Yamano, Hidemasa; Vasile, A.*; Kang, S.-H.*; Summer, T.*; Tsige-Tamirat, H.*; Wang, J.*; Ashurko, I.*
Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 7 Pages, 2019/05
The Generation IV (GEN-IV) international forum is a framework for international co-operation in research and development for the next generation of nuclear energy systems. Within the GEN-IV sodium-cooled fast reactor (SFR) system arrangement, the SFR Safety and Operation (SO) project addresses the areas of safety technology and reactor operation technology developments. The aims of the SO project include (1) analyses and experiments that support establishing safety approaches and validating performance of specific safety features, (2) development and verification of computational tools and validation of models employed in safety assessment and facility licensing, and (3) acquisition of reactor operation technology, as determined largely from experience and testing in operating SFR plants. The tasks in the SO topics are categorized into the following three work packages (WP): WP-SO-1 "Methods, Models and Codes", WP-SO-2 "Experimental Programs and Operational Experience", and WP-SO-3 "Studies of Innovative Design and Safety Systems". This paper reports recent activities within the SO project.
Kamide, Hideki
Genshiryoku No Ima To Ashita, p.265 - 268, 2019/03
no abstracts in English
Fukano, Yoshitaka
Journal of Nuclear Engineering and Radiation Science, 5(1), p.011001_1 - 011001_13, 2019/01
Local subassembly faults (LFs) have been considered to be of greater importance in safety evaluation in sodium-cooled fast reactors (SFRs) because fuel elements were generally densely arranged in the subassemblies (SAs) in this type of reactors, and because power densities were higher compared with those in light water reactors. A hypothetical total instantaneous flow blockage at the coolant inlet of an SA (HTIB) gives most severe consequences among a variety of LFs. Although an evaluation on the consequences of HTIB using SAS4A code was performed in the past study, SAS4A code was further developed by implementing analytical model of power control system in this study. An evaluation on the consequences of HTIB in an SFR by this developed SAS4A code clarified that the conclusion in the past study was almost same as that in this study. Furthermore SAS4A code was newly validated using four in-pile experiments which simulated HTIB events. The validity of SAS4A application to safety evaluation on the consequence of HTIB was further enhanced in this study. Thus the methodology of HTIB evaluation was established in this study together with the past study and is applicable to HTIB evaluations in other SFRs.
Okano, Yasushi
Nippon Genshiryoku Gakkai-Shi, 60(12), p.764 - 769, 2018/12
JAEA has developed the second safety design guidelines report, "Safety Design Guidelines on Structures, Systems and Components" for Generation-IV SFR system, following the previously published SFR Safety Design Criteria and the first SFR Safety Design Guidelines report and with the reviews by a technical committee under Atomic Energy Society of Japan. This article explains about 14 key points on reactor core system, coolant system, and containment system and also current consistency of international SFR designs to the safety design guidelines.
Yamano, Hidemasa; Tobita, Yoshiharu
Proceedings of 11th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-11) (Internet), 3 Pages, 2018/11
Based on the event tree analysis, the present numerical analyses investigated the capability of fuel discharge through the one-dimensional single fuel assembly geometry and the two-dimensional geometry of a CRGT channel with neighboring fuel assemblies. The single fuel assembly analyses showed that the fuel blockage formed in the lower shielding region because fuel solidified by contacting with cold sodium in case of no fission gas release. On the assumption that fission gas was released, the molten fuel successfully relocated below the core. The next analyses using the CRGT channel indicated a significant fuel discharge through the CRGT channel. This is because the fuel temperature was still high just after the CRGT wall failure and sodium in the CRGT channel was quickly voided just after the ingress of a small amount of molten fuel.
Ohgama, Kazuya; Oki, Shigeo; Kitada, Takanori*; Takeda, Toshikazu*
Proceedings of 21st Pacific Basin Nuclear Conference (PBNC 2018) (USB Flash Drive), p.942 - 947, 2018/09
Tanaka, Masaaki; Ono, Ayako; Hamase, Erina; Ezure, Toshiki; Miyake, Yasuhiro*
Nippon 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).
Ito, Daisuke*; Ito, Kei*; Saito, Yasushi*; Aoyagi, Mitsuhiro; Matsuba, Kenichi; Kamiyama, Kenji
Nuclear Engineering and Design, 334, p.90 - 95, 2018/08
Times Cited Count：1 Percentile：72.34(Nuclear Science & Technology)Two-phase flow through porous media must be well understood to develop a severe accident analysis code not only for light water reactor but also sodium-cooled fast reactor. When a core disruptive accident occurs in sodium-cooled fast reactor, the fuel inside the core become melted and interacts with the coolant. As a result, gas-liquid two-phase flow will be formed in the debris bed, which may have porous nature depending on the cooling process. In such condition, the local porosity and its distribution are very important to characterize two-phase flow field in the porous media. In this study, X-ray radiography was applied to measure the local porosity in the packed bed of spheres. The radial profiles were estimated from the chordal profiles measured by the X-ray method and compared with the previous porosity model. In addition, the void fraction radial profiles were also obtained in air-water two-phase flow.
Hirakawa, Moe*; Kikuchi, Yuichiro*; Sakai, Takaaki*; Tanaka, Masaaki; Ohshima, Hiroyuki
Proceedings of 26th International Conference on Nuclear Engineering (ICONE-26) (Internet), 8 Pages, 2018/07
Gas entrainment (GE) from cover gas is one of key issue for Sodium-cooled fast reactors to prevent unexpected effects to core reactivity. By using a computational fluid dynamics (CFD) code, analyses have been conducted to estimate the drifting vortexes on water experiments which were generated as wake vortexes behind a plate obstacle in the circulating water channel. In this paper, the results of comparison between experiments and analyses were discussed and the gas core lengths from the surface vortexes were evaluated by using the evaluation tool named StreamViewer developed by Japan Atomic Energy Agency.
Sheikh, Md. A. R.*; Son, E.*; Kamiyama, Motoki*; Morioka, Toru*; Matsumoto, Tatsuya*; Morita, Koji*; Matsuba, Kenichi; Kamiyama, Kenji; Suzuki, Toru*
Journal of Nuclear Science and Technology, 55(6), p.623 - 633, 2018/06
Times Cited Count：4 Percentile：24.89(Nuclear Science & Technology)During core-disruptive accidents in sodium-cooled fast reactors, the sedimentation behavior of fragmented debris is crucial for in-vessel retention. The height of the beds formed may influence both the cooling of the bed and the neutronic characteristics. To develop an experimental database of bed formation behavior, a series of experiments using simulant materials, namely, AlO, ZrO, and stainless steel, were performed under gravity-driven discharge of solid particles from a nozzle into a quiescent cylindrical water pool. The bed height was measured for particles of different size, density, and sphericity, and an injection nozzle with varying diameter, injection velocity, and injection height. From these experiments, an empirical correlation was established to predict the bed height for both homogeneous and mixed particles for the different properties. This correlation reproduces reasonably well the experimental trend in bed height.
CFukuyama, Hiroyuki*; Higashi, Hideo*; Yamano, Hidemasa
Proceedings of 2018 International Congress on Advances in Nuclear Power Plants (ICAPP 2018) (CD-ROM), p.1014 - 1019, 2018/04
In this study, densities, surface tensions, normal spectral emissivities, heat capacities and thermal conductivities of molten SUS316L and SUS316L containing 5mass%-BC were measured by the electromagnetic levitation technique in a static magnetic field.
