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Journal Articles

Water experiments on thermal striping phenomena at the core outlet of an advanced sodium-cooled fast reactor, 1; Proposal of countermeasures to mitigate temperature fluctuations around control rods

Kobayashi, Jun; Aizawa, Kosuke; Ezure, Toshiki; Kurihara, Akikazu; Tanaka, Masaaki

Hozengaku, 20(3), p.89 - 96, 2021/10

Hot sodium from the fuel assembly can mix with cold sodium from the control rod (CR) channel and the blanket assemblies at the bottom plate of the Upper Internal Structure (UIS) of Advanced-SFR. Temperature fluctuation due to mixing of the fluids at different temperature between the core outlet and cold channel may cause high cycle thermal fatigue on the structure around the bottom of UIS. A water experiment using a 1/3 scale 60 degree sector model simulating the upper plenum of the Advanced-SFR has been conducted to examine countermeasures for the significant temperature fluctuation generated around the bottom of UIS. We focused on the temperature fluctuations near the primary and backup control rod channels, and studied the countermeasure structure to mitigate the temperature fluctuation through temperature distribution and flow velocity distribution measurements. As a result, effectiveness of the countermeasure to mitigate the temperature fluctuation intensity was confirmed.

Journal Articles

Water experiments on thermal striping phenomena at the core outlet of an advanced sodium-cooled fast reactor, 2; Proposal of countermeasures to mitigate temperature fluctuations around radial blanket fuel assemblies

Kobayashi, Jun; Aizawa, Kosuke; Ezure, Toshiki; Kurihara, Akikazu; Tanaka, Masaaki

Hozengaku, 20(3), p.97 - 101, 2021/10

Focusing on the thermal striping phenomena that occurs at a bottom of the internal structure of an advanced sodium-cooled fast reactor (Advanced-SFR) that has been designed by the Japan Atomic Energy Agency, a water experiment using a 1/3 scale 60 degree sector model simulating the upper plenum of the Advanced-SFR has been conducted to examine countermeasures for the significant temperature fluctuation generated around the bottom of Upper Internal Structure (UIS). In the previous paper, we reported the effect of measures to mitigate temperature fluctuations around the control rod channels. In this paper, the same test section was used, and a water experiment was conducted to obtain the characteristics of temperature fluctuations around the radial blanket fuel assembly. And the shape of the Core Instrumentation Support Plate (CIP) was modified, and it was confirmed that it was highly effective in alleviating temperature fluctuations around the radial blanket fuel assembly.

Journal Articles

Analysis of gas entrainment phenomenon from free liquid surface for a sodium-cooled fast reactor design (Velocity profile and Strouhal number in a flow field)

Uchida, Mao*; Sakai, Takaaki*; Ezure, Toshiki; Tanaka, Masaaki

Mechanical Engineering Journal (Internet), 8(4), p.21-00161_1 - 21-00161_11, 2021/08

An evaluation method based on numerical analyses has been developed to predict occurrences gas entrainment phenomena at a free surface in a sodium-cooled fast reactor. In this study, experiments were conducted for gas entrainments due to drifting free surface vortexes observed in a circulating water tunnel geometry. Numerical analyses were also conducted in the same geometry using a computational fluid dynamics (CFD) code. Then, Strouhal numbers of vortex frequency and detailed flow velocity profiles were compared between experimental results and numerical results to clarify the evaluation accuracy of CFD calculation. As the results, the Strouhal numbers of the vortex frequency obtained from numerical analyses showed good agreement with the experimental data.

Journal Articles

The Examination of advanced analysis method on unsteady gas entrainment vortex applying AMR method

Matsushita, Kentaro; Fujisaki, Tatsuya*; Ezure, Toshiki; Tanaka, Masaaki; Uchida, Mao*; Sakai, Takaaki*

Keisan Kogaku Koenkai Rombunshu (CD-ROM), 26, 6 Pages, 2021/05

For the gas entrainment vortex at the free surface in sodium-cooled fast reactors, development of the numerical analysis method to evaluate amount of the gas entrainment from the free surface has been developing. In this paper, the automatic creation of analysis meshes which can suppress the calculation cost while maintaining the prediction accuracy of the vortex shape is investigated, and the adaptive mesh refinement (AMR) method is examined to the creation of analysis mesh applying to the unsteady vortex system. The refined mesh based on the criterion evaluated by vorticity, Q-value as second invariant of the velocity and the discriminant for the eigen equation of the velocity gradient tensor is considered, and it found that the AMR method based on Q-value can refine the analysis meshes most efficiently.

Journal Articles

Preliminary analysis of sodium experimental apparatus PLANDTL-2 for development of evaluation method for thermal-hydraulics in reactor vessel of sodium fast reactor under decay heat removal system operation condition

Ono, Ayako; Tanaka, Masaaki; Miyake, Yasuhiro*; Hamase, Erina; Ezure, Toshiki

Mechanical Engineering Journal (Internet), 7(3), p.19-00546_1 - 19-00546_11, 2020/06

Fully natural circulation decay heat removal systems (DHRSs) are to be adopted for sodium fast reactors, which is a passive safety feature without any electrical pumps. It is required to grasp the thermal-hydraulic phenomena in the reactor vessel and evaluate the coolability of the core under the natural circulation not only for the normal operating condition but also for severe accident conditions. In this paper, the numerical results of the preliminary analysis for the sodium experimental condition with the PLANDTL-2 are discussed to establish an appropriate numerical models for the reactor core including the gap region among the subassemblies and the DHX. From these preliminary analyses, the characteristics of the thermal-hydraulics behavior in the PLANDTL-2 to be focused are extracted.

Journal Articles

Experiments on gas entrainment phenomena due to free surface vortex induced by flow passing beside stagnation region

Ezure, Toshiki; Ito, Kei; Tanaka, Masaaki; Ohshima, Hiroyuki; Kameyama, Yuri*

Nuclear Engineering and Design, 350, p.90 - 97, 2019/08

 Times Cited Count:3 Percentile:57.07(Nuclear Science & Technology)

This paper reports the results of an experiment on surface vortex-type gas entrainment, which occurs in a shear flow area where flow passes besides the stagnation region. The relationship between the free surface dimple shape and the velocity distribution around the free surface vortex was simultaneously grasped under several horizontal and suction velocity conditions by a combination of visualization and particle image velocimetry measurements. The circulation and the vertical velocity gradient were also evaluated from the velocity distributions at a plane just below the free surface and the middle plane between the free surface and suction nozzle. Quantitative relationships between the circulation, vertical velocity gradient, and gas core length were obtained in time-trends as fundamental data to develop the evaluation method of gas entrainment. Furthermore, it was confirmed that the evaluation method based on a vortex model was an effective way to evaluate gas entrainment.

Journal Articles

Study on multi-dimensional core cooling behavior of sodium-cooled fast reactors under DRACS operating conditions

Ezure, Toshiki; Onojima, Takamitsu; Tanaka, Masaaki; Kobayashi, Jun; Kurihara, Akikazu; Kameyama, Yuri*

Proceedings of 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-18) (USB Flash Drive), p.3355 - 3363, 2019/08

Steady-state sodium experiments under the operating conditions of a decay heat removal system (DHRS) were carried out as part of the safety enhancement of sodium-cooled fast reactors using the PLANDTL 2 facility, which has 30 heated channels with electric heaters and 25 no-heated channels as the simulated core. In the experiments, a direct reactor auxiliary cooling system (DRACS) with a dipped type direct heat exchanger (DHX) in the upper plenum was used as the DHRS. This paper reports on the preliminary experimental results of the PLANDTL 2 experiments under the DRACS operating conditions without flow in the primary circuit, including the thermal hydraulic interactions between the upper plenum and the core under the DHX operating conditions and the resulting core cooling behavior from the outside of the multiple rows of the fuel assemblies

Journal Articles

Study on evaluation method for entrained gas flow rate by free surface vortex

Ito, Kei*; Ito, Daisuke*; Saito, Yasushi*; Ezure, Toshiki; Matsushita, Kentaro; Tanaka, Masaaki; Imai, Yasutomo*

Proceedings of 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-18) (USB Flash Drive), p.6632 - 6642, 2019/08

In this paper, a mechanistic model is proposed to calculate the entrained gas flow rate by a free surface vortex. The model contains the theoretical equation of transient gas core elongation and the empirical equation of critical gas core length for gas bubble detachment. Based on those two equations, the entrained gas flow rate is calculated as the portion of the gas core elongated beyond the critical gas core length per unit time. Then, the mechanistic model was applied to the calculation of the entrained gas flow rate in a simple water experiment. As a result, it is confirmed that the entrained gas flow rate grows rapidly when the liquid (water) flow rate, which determine the strength of a free surface vortex, exceeds a certain threshold value.

Journal Articles

Study on analysis method for inert gas behavior in liquid metal flow with considering dissolution and entrainment at free surface

Matsushita, Kentaro; Ito, Kei*; Ezure, Toshiki; Tanaka, Masaaki

Dai-24-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (USB Flash Drive), 5 Pages, 2019/06

In the design study on a sodium-cooled fast reactor (SFR), a numerical simulation code named SYRENA has been developed in Japan Atomic Energy Agency to analyze the behavior of gas bubbles and/or dissolved gas in the primary coolant system. In the present study, the effect of the non-condensable gas entrainment at the free surface on the bubble and the dissolved gas behavior in the primary coolant system were investigated for a typical pool type reactor, and also effect of a dipped-plate (D/P) installed below the free surface in the reactor vessel to suppress the gas bubble entrainment into the primary coolant system was especially investigated. It was clarified that the D/P was influential to the non-condensable gas behavior and the molar flow rate of gas bubbles in the primary coolant system varies depending on the relationship between the gas entrainment rate at the free surface and the exchange flow rate through the D/P.

Journal Articles

Preliminary analysis of sodium experimental apparatus PLANDTL-2 for development of evaluation method for thermal hydraulics in reactor vessel of sodium fast reactor under decay heat removal system operation condition

Ono, Ayako; Tanaka, Masaaki; Miyake, Yasuhiro*; Hamase, Erina; Ezure, Toshiki

Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 7 Pages, 2019/05

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. In this paper, the numerical simulation results of the preliminary analysis for the sodium experiment 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 appropriate numerical models for the reactor core including the gap region among the subassemblies and the DHX.

Journal Articles

Parametric analysis of bubble and dissolved gas behavior in primary coolant system of sodium-cooled fast reactors

Matsushita, Kentaro; Ito, Kei*; Ezure, Toshiki; Tanaka, Masaaki

Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 9 Pages, 2019/05

A numerical simulation code named SYRENA has been developed in JAEA to analyze the behavior of entrained bubbles and dissolved gas in the primary coolant of sodium-cooled fast reactor (SFR). In the present study, a flow network model of SYRENA to a hypothetical pool type reactor was developed and the non-condensable gas behavior was investigated through the comparison with that in the loop type reactor. The effect of the dipped-plate (D/P) tentatively introduced into the pool-type reactor on the gas behavior was investigated through the parametric analyses about the sodium exchange flow rate through the D/P and the gas entrainment rate at the free surface. It was suggested that the increase in the exchange flow rate through the D/P doesn't always work to decrease the bubble volume in the primary coolant system.

Journal Articles

Preliminary calculation on thermal stratification phenomena in the fundamental sodium experiment "SuperCAVNA"

Ezure, Toshiki; Nagasawa, Kazuyoshi*; Tanaka, Masaaki

Proceedings of 11th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-11) (Internet), 5 Pages, 2018/11

To establish an evaluation method of thermal stratification in sodium-cooled fast reactors (SFRs), a benchmark exercise was performed for a sodium experiment (SuperCAVNA) with a rectangular test section and heated wall. This paper presents a preliminary result using three-dimensional finite differential code AQUA. The influences of mesh size for heat exchange and turbulence model are studied, and the calculation results were also compared to the experimental results in the literature. Then, the calculation results reproduced the thermal stratification in SuperCAVNA experiment. The position and the temperature gradient of the stratified surface also showed good agreement with the experimental result. The applicability of the numerical approach employed in this study for the evaluation of thermal stratification problem in SFRs was confirmed.

Journal Articles

Development of numerical estimation method for thermal hydraulics in reactor vessel of sodium-cooled fast reactor under decay heat removal system operation conditions; Preliminary thermal hydraulics simulation for simulated reactor vessel in sodium experimental apparatus PLANDTL-2

Tanaka, Masaaki; Ono, Ayako; Hamase, Erina; Ezure, Toshiki; Miyake, Yasuhiro*

Nihon 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).

Journal Articles

Experiments on gas entrainment phenomena due to free surface vortex induced by flow passing beside stagnation region

Ezure, Toshiki; Ito, Kei; Tanaka, Masaaki; Ohshima, Hiroyuki; Kameyama, Yuri*

Proceedings of 17th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-17) (USB Flash Drive), 9 Pages, 2017/09

In the design of sodium cooled fast reactors, cover gas entrainment into sodium coolant (gas entrainment) is one of significant thermal hydraulic issues. This paper describes experimental results on surface vortex type gas entrainment which occurs in a share flow area where flow passes beside the stagnation region. In the experiment, the relationship between the free surface dimple shape and the velocity distribution around the free surface vortex was simultaneously grasped under several horizontal and suction velocity conditions by means of visualization measurement and Particle Image Velocimetry measurement. As the results, quantitative relationships among circulation, vertical velocity gradient and the gas core length were obtained in time-trends as fundamental data to develop the evaluation method of gas entrainment. Furthermore, it was confirmed that the evaluation method based on a vortex model, was an effective way to evaluate gas entrainment.

Journal Articles

Water experiments on thermal striping in reactor vessel of advanced sodium-cooled fast reactor; Influence of flow collector of backup CR guide tube

Kobayashi, Jun; Ezure, Toshiki; Tanaka, Masaaki; Kamide, Hideki

Proceedings of 10th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-10) (USB Flash Drive), 5 Pages, 2016/11

JAEA has been conducting a design study for an advanced large-scale sodium-cooled fast reactor (SFR). Hot sodium from the fuel subassembly can mix with the cold sodium from the control rod (CR) channel at the bottom of Upper Internal Structure (UIS). Temperature fluctuation due to the fluid mixing at the core outlet may cause high cycle thermal fatigue at the bottom of UIS. JAEA had performed a water experiment to examine countermeasures for the significant temperature fluctuation generated at the bottom of SFRs UIS. Meanwhile, a self-actuated shutdown system (SASS) is equipped in a backup control rod (BCR) channel to ensure reactor shutdown. The BCR guide tubes have a flow guide structure "flow-collector" to provide reliable operation of SASS. Flow-collector may affect the thermal mixing behavior at the bottom of the UIS. This study has investigated the influence of the flow- collector on characteristics of the temperature fluctuation around the BCR channels.

Journal Articles

Influence of fluid viscosity on vortex cavitation at a suction pipe inlet

Ezure, Toshiki; Ito, Kei; Kameyama, Yuri*; Kamide, Hideki; Kunugi, Tomoaki*

Nihon Genshiryoku Gakkai Wabun Rombunshi, 15(3), p.151 - 158, 2016/09

no abstracts in English

Journal Articles

Study on behavior of vortex cavitation around suction pipes in sodium-cooled fast reactor geometry

Ezure, Toshiki; Ito, Kei; Kamide, Hideki; Kunugi, Tomoaki*

Thermal Science and Engineering, 24(3), p.31 - 38, 2016/07

Journal Articles

Experimental measurement of vortex cavitation around a suction pipe inlet

Ezure, Toshiki; Ito, Kei; Kameyama, Yuri*; Kurihara, Akikazu; Kunugi, Tomoaki*

Konsoryu, 30(2), p.189 - 196, 2016/06

Journal Articles

Visualization of distribution of shear stress due to water vortex flow with SSLCC

Okazaki, Soichiro*; Ezure, Toshiki; Ohshima, Hiroyuki; Kawara, Zensaku*; Yokomine, Takehiko*; Kunugi, Tomoaki*

Proceedings of 10th Pacific Symposium on Flow Visualization and Image Processing (PSFVIP-10), 8 Pages, 2015/06

A visualization study is performed under suction vortex geometry in water. In the experiment, the shear-sensitive liquid crystal coating (SSLCC) is applied to grasp the distribution of wall shear stress under the suction votex flow. As the result, it was found that the peak value of wall shear stress is appeared at the center and edge of the projected area of suction pipe. The non-dimensional profile of wall shear stress obtained by suction vortex flow experiment agrees well with that of numerical simulation.

Journal Articles

Water experiments on thermal striping in reactor vessel of Japan Sodium-cooled Fast Reactor; Countermeasures for significant temperature fluctuation generation

Kobayashi, Jun; Ezure, Toshiki; Kamide, Hideki; Oyama, Kazuhiro*; Watanabe, Osamu*

Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 6 Pages, 2015/05

A column type upper internal structure (UIS) is installed in the upper plenum of reactor vessel in JSFR. High cycle thermal fatigue may occur at the bottom plate (CIP) of the UIS where the hot sodium from the fuel subassembly can mix with the cold sodium from the control rod channel and the blanket fuel subassembly. We have been conducted a water experiment using a reactor upper plenum model to grasp the thermal-hydraulic phenomena around control rod (CR) channels, and to obtain countermeasures for significant temperature fluctuation on the CIP. The experimental apparatus has 1/3 scale and 60$$^{circ}$$ sector model of the reactor upper plenum. By the experiment, characteristics of fluid temperature fluctuation between the handling head of the assemblies and the CIP are measured and countermeasure for the significant temperature fluctuation generation will be discussed on the influence of the distance from the handling head outlet to the lower surface of the CIP.

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