Network


Latest external collaboration on country level. Dive into details by clicking on the dots.

Hotspot


Dive into the research topics where Ken-ichi Matsuba is active.

Publication


Featured researches published by Ken-ichi Matsuba.


Journal of Nuclear Science and Technology | 2013

Experimental study on fuel-discharge behaviour through in-core coolant channels

Kenji Kamiyama; Masaki Saito; Ken-ichi Matsuba; Mikio Isozaki; Ikken Sato; Kensuke Konishi; Vradimir A. Zuyev; Alexander A. Kolodeshnikov; Yuri S. Vassiliev

In core-disruptive accidents of sodium-cooled fast reactors, fuel discharge from the core region reduces the possibility of severe re-criticality events. In-core coolant channels with large hydraulic diameters, such as the control-rod guide tube and a concept of the Fuel Assembly with Inner Duct Structure have a potential to provide effective fuel-discharge paths if effects of sodium in these paths on molten fuel discharge are limited. Two series of experiments were conducted to investigate fuel-discharge behaviour through the sodium-filled channels. In the first series of experiments, an alloy with low melting temperature was ejected into a water channel to clarify dominant phenomena for melt discharge through the coolant-filled channel and to develop methodologies for evaluating the effects of coolant on melt discharge. In the second series of experiments, a molten alumina was discharged through the sodium-filled channel in order to verify the applicability of the knowledge and evaluation methodologies obtained in the first series of experiments to the sodium-filled channel. These series of experiments showed that the discharge path can be entirely voided by the vaporisation of a part of the coolant at the initial melt discharge phase that this is followed by coolant vapour expansion and that melt penetrates significantly into the voided channel. Preliminary extrapolation of the present results to the in-core coolant channel suggests that the effects of the sodium on fuel discharge are limited and, therefore, in-core coolant channels will provide effective fuel-discharge paths for reducing neutronic activity.


Journal of Nuclear Science and Technology | 2016

Distance for fragmentation of a simulated molten-core material discharged into a sodium pool

Ken-ichi Matsuba; Mikio Isozaki; Kenji Kamiyama; Yoshiharu Tobita

A series of experiments has been carried out to obtain experimental knowledge on the distance for fragmentation of a molten core material discharged into the sodium plenum during postulated core disruptive accidents of sodium-cooled fast reactors. In the current experiments, 0.9 kg of molten aluminum (initial temperature: around 1473 K) was discharged into a sodium pool (diameter: 0.11 m, depth: 1 m, initial temperature: 673 K) through a nozzle (inner diameter: 20 mm). Visual observation of the fragmentation behavior was performed using an X-ray imaging system. The following experimental results were obtained. (1) Liquid column of molten aluminum was intensively fragmented almost simultaneously with a rapid expansion of sodium vapor in the vicinity of the column. (2) Due to the intensive fragmentation, penetration of the liquid column was limited to approximately 100 mm or so from the sodium level. (3) The molten aluminum was rapidly cooled after the intensive fragmentation. Based on these results, the distance for fragmentation of the liquid column was estimated to be 100 mm in the experiments. Through the current experiment, useful knowledge was obtained for the future development of an evaluation method of the distance for fragmentation of the molten core material.


Volume 3: Next Generation Reactors and Advanced Reactors; Nuclear Safety and Security | 2014

Development of an Evaluation Methodology for the Fuel-Relocation Into the Coolant Plenum in the Core Disruptive Accident of Sodium-Cooled Fast Reactors

Kenji Kamiyama; Yoshiharu Tobita; Tohru Suzuki; Ken-ichi Matsuba

In the case of postulated core disruptive accidents (CDAs) of sodium-cooled fast reactors (SFRs), fuel relocation out of the core region decrease the potential for severe power excursions caused by recriticality and control-rod guide tubes (CRGTs) provide an effective path for fuel-relocation. Therefore, a methodology for evaluating molten-fuel relocation through CRGTs is required in order to realistically evaluate sequences and consequences of CDAs. Since the liquid sodium exists at the coolant plenum, pressurizations and coolant void development associated with fuel-coolant interactions (FCIs) are considered to affect fuel-relocation. Therefore, the objective of the present study is to develop the methodology for evaluation of molten-fuel relocation into the coolant plenum with FCIs. In the present study, the SIMMER code which has been developed for CDA analyses was utilized as a technical basis since this code can treat multi-phase, multi-component fluid dynamics with phase-changes supposed to take place in the coolant plenum during fuel-relocation. The evaluation methodology was developed through validations of the SIMMER code using experimental data. A series of fundamental experiments were selected for model validations in which an alloy with low melting temperature and water were used as simulant materials for the fuel and the coolant respectively since the experiments were performed under a simulated CDA condition of SFRs in which a liquid-liquid direct contact was maintained between the melt and water contact surface, and the visual observation on FCI process was effective to validate models based on phenomenological considerations. The code was validated by two steps: In the first step, fundamental validations of melt-discharge into the coolant were performed, namely, momentum exchange functions of flowing-melt both to the wall of relocation-path and to the coolant were validated based on experimental data in which effects of FCIs on melt-discharge into the coolant were eliminated or negligible. In the second step, comprehensive validations of melt-ejection into the coolant were performed, namely, models which affect heat-exchange between the melt and the coolant were validated based on experimental data in which the melt was relocated into the coolant with FCIs. The second step validation required model improvements for suppression of melt-coolant interfacial area based on the results of visual observation in the experiments in order to reproduce the experimental results appropriately. Through the present model validations, the methodology to evaluate molten-fuel relocation into the coolant plenum with FCIs was successfully developed.Copyright


Journal of Nuclear Science and Technology | 2014

Experimental studies on the upward fuel discharge for elimination of severe recriticality during core-disruptive accidents in sodium-cooled fast reactors

Kenji Kamiyama; Kensuke Konishi; Ikken Sato; Jun-ichi Toyooka; Ken-ichi Matsuba; Vladimir A. Zuyev; Alexander V. Pakhnits; Vladimir A. Vityuk; Alexander D. Vurim; Valery A. Gaidaichuk; Alexander A. Kolodeshnikov; Yuri S. Vassiliev

In order to eliminate the energetic potential in the case of postulated core-disruptive accidents (CDAs) of sodium-cooled fast reactors, introduction of a fuel subassembly with an inner-duct structure (FAIDUS) has been considered. Recently, a design option of FAIDUS which leads molten fuel to upward discharge has been considered as the reference core design of the Japan Sodium-Cooled Fast Reactor (JSFR). In this study, a series of experiments which consisted of three out-of-pile tests and one in-pile test were conducted to obtain experimental knowledge of the upward discharge of molten fuel. Experimental data which showed a sequence of upward fuel discharge and effects of initial pressure conditions on upward discharge were obtained through the out-of-pile and in-pile test. Preliminary extrapolation of the present results to the supposed condition in the early phase of the CDA in the JSFR design suggests that the sufficient upward flow rate of molten fuel is expected to prevent the core melting from progressing beyond the fuel subassembly scale and that the upward discharge option will be effective in eliminating the energetic potential.


Journal of Nuclear Science and Technology | 2018

Sedimentation behavior of mixed solid particles

Abdur Rob Sheikh; Eikaku Son; Motoki Kamiyama; Tohru Morioka; Tatsuya Matsumoto; Koji Morita; Ken-ichi Matsuba; Kenji Kamiyama; Tohru Suzuki

ABSTRACT During the material relocation phase of core-disruptive accidents in sodium-cooled fast reactors, the sedimentation behavior of fragmented debris discharged from the reactor core into the lower plenum region leading to a debris-bed formation is crucial in regard to in-vessel retention and safety concerns. The height of the beds formed may influence both the cooling of the bed from the decay heat in the fuel and the neutronic characteristics. To develop an experimental database of bed formation behavior, a series of experiments using simulant materials, namely, Al2O3, ZrO2, 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 with critical factors, which were identified in this and previous experiments.


Science and Technology of Nuclear Installations | 2015

SIMMER-III Analyses of Local Fuel-Coolant Interactions in a Simulated Molten Fuel Pool: Effect of Coolant Quantity

Songbai Cheng; Ken-ichi Matsuba; Mikio Isozaki; Kenji Kamiyama; Tohru Suzuki; Yoshiharu Tobita

Studies on local fuel-coolant interactions (FCI) in a molten pool are important for the analyses of severe accidents that could occur for sodium-cooled fast reactors (SFRs). To clarify the mechanisms underlying this interaction, in recent years, several experimental tests, with comparatively larger difference in coolant volumes, were conducted at the Japan Atomic Energy Agency by delivering a given quantity of water into a molten pool formed with a low-melting-point alloy. In this study, to further understand this interaction, interaction characteristics including the pressure buildup as well as mechanical energy release and its conversion efficiency are investigated using the SIMMER-III, an advanced fast reactor safety analysis code. It is found that the SIMMER-III code not only reasonably simulates the transient pressure and temperature variations during local FCIs, but also supports the limited tendency of pressurization and resultant mechanical energy release as observed from experiments when the volume of water delivered into the pool increases. The performed analyses also suggest that the most probable reason leading to such limited tendency should be primarily due to an isolation effect of vapor bubbles generated at the water-melt interface.


2014 22nd International Conference on Nuclear Engineering | 2014

Characteristics of Pressure Buildup From Local Fuel-Coolant Interactions in a Simulated Molten Fuel Pool

Songbai Cheng; Ken-ichi Matsuba; Mikio Isozaki; Kenji Kamiyama; Tohru Suzuki; Yoshiharu Tobita

Studies on local fuel-coolant interactions (FCI) in a molten pool are important for severe accident analyses of sodium-cooled fast reactors (SFRs). To clarify the mechanisms underlying this interaction, in this study a series of experiments was conducted by delivering a given quantity of water into a simulated molten fuel pool (formed with a low-melting-point alloy). Based on the experimental data obtained from a variety of conditions, including difference in water volume, melt temperature and water subcooling, the characteristics of pressure-buildup during local FCIs was investigated. It is found that under our experimental conditions the water volume and melt temperature have remarkable impact on the interaction, while the role of water subcooling seems to be less prominent. The performed analyses also suggest that the pressurization from local FCIs should be intrinsically limited, due to a suppressing role caused by the increasing of coolant volume entrapped within the pool as well as the transition of boiling mode. Current work, which gives a palette of favorable data for a better understanding and an improved estimation of severe accidents in SFRs, is expected to benefit future analyses and verifications of computer models developed in advanced fast reactor safety analysis codes.© 2014 ASME


Volume 2: Plant Systems, Structures, and Components; Safety and Security; Next Generation Systems; Heat Exchangers and Cooling Systems | 2012

Mechanism of Upward Fuel Discharge During Core Disruptive Accident in Sodium-Cooled Fast Reactors

Ken-ichi Matsuba; Mikio Isozaki; Kenji Kamiyama; Yoshiharu Tobita

The elimination of severe power excursion with significant mechanical-energy release during Core Disruptive Accidents (CDAs) is a key issue for the enhanced safety of Sodium-cooled Fast Reactors (SFRs). In order to prevent the formation of a large-scale molten fuel pool within a reactor core, which is one of factors leading to the severe power excursion during CDAs, Japan Atomic Energy Agency (JAEA) is studying the introduction of Fuel Assembly with Inner Duct Structure (FAIDUS). In the current reference design for FAIDUS, the top end of the inner duct is open whereas the bottom end of the inner duct is closed, and therefore it is expected that the molten fuel will be discharged from a reactor core towards an upper sodium plenum through the inner duct. The objective of the present study is to clarify the fundamental mechanism for upward fuel discharge through the inner duct structure in FAIDUS, and thereby to confirm the effectiveness of FAIDUS.In the previous paper, the possibility of upward discharge of a high-density melt driven by coolant vapor was confirmed by visual observation in the JAEA’s out-of-pile experiment, in which molten Wood’s metal (density at the room temperature: 8700 kg/m3, melting point: 352 K) simulating the molten fuel was injected into a coolant channel (equivalent inner diameter: 30 mm, total height: 2 m, fluid content: water) simulating the inner duct structure. In this paper, the mechanism of upward discharge of a high-density melt driven by coolant vapor pressure and/or flow in this experiment is discussed in terms of the application to reactor conditions. Through this discussion, the following mechanisms were clarified.1) Coolant vapor pressure is built up within the coolant channel after the melt injection. The magnitude of the pressure buildup becomes larger with increase of melt-enthalpy-injection rate which is defined by the product of melt-mass-injection rate into the coolant channel and melt specific enthalpy.2) Following the pressure buildup, the melt is discharged upward being driven by the coolant vapor flow directing towards the top opening end of the coolant channel. The upward discharge mass rate becomes higher with the increase of the magnitude of the pressure buildup and therefore melt-enthalpy-injection rate.From these experimental knowledge, it was suggested that the coolant pressure buildup could act as one of the driving force for the upward discharge of a high-density melt through the inner duct structure in FAIDUS under reactor conditions with higher melt-enthalpy-injection rate than the current experimental condition.Copyright


Nuclear Engineering and Technology | 2015

A preliminary evaluation of unprotected loss-of-flow accident for a prototype fast-breeder reactor

Tohru Suzuki; Yoshiharu Tobita; Kenichi Kawada; Hirotaka Tagami; Joji Sogabe; Ken-ichi Matsuba; Kei Ito; Hiroyuki Ohshima


Annals of Nuclear Energy | 2015

A numerical study on local fuel–coolant interactions in a simulated molten fuel pool using the SIMMER-III code

Songbai Cheng; Ken-ichi Matsuba; Mikio Isozaki; Kenji Kamiyama; Tohru Suzuki; Yoshiharu Tobita

Collaboration


Dive into the Ken-ichi Matsuba's collaboration.

Top Co-Authors

Avatar

Kenji Kamiyama

Japan Atomic Energy Agency

View shared research outputs
Top Co-Authors

Avatar

Yoshiharu Tobita

Japan Nuclear Cycle Development Institute

View shared research outputs
Top Co-Authors

Avatar

Mikio Isozaki

Japan Atomic Energy Agency

View shared research outputs
Top Co-Authors

Avatar

Tohru Suzuki

Japan Atomic Energy Agency

View shared research outputs
Top Co-Authors

Avatar

Songbai Cheng

Japan Atomic Energy Agency

View shared research outputs
Top Co-Authors

Avatar

Jun-ichi Toyooka

Japan Atomic Energy Agency

View shared research outputs
Top Co-Authors

Avatar
Top Co-Authors

Avatar

Chikara Ito

Japan Atomic Energy Agency

View shared research outputs
Top Co-Authors

Avatar

Hirotaka Kawahara

Japan Atomic Energy Agency

View shared research outputs
Researchain Logo
Decentralizing Knowledge