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JAEA Reports

Construction of the sodium test loop in advanced technology experiment sodium facility (AtheNa)

Imamura, Hiroaki; Suzuki, Masashi*; Shimoyama, Kazuhito; Miyakoshi, Hiroyuki

JAEA-Technology 2019-005, 163 Pages, 2019/06

JAEA-Technology-2019-005.pdf:25.24MB

For the R&D of safety enhance in future fast reactor development, the constructed the large sodium test loop (mother loop) in advanced technology experiment sodium facility (AtheNa) was completed. The sodium test loop possesses the largest capacity of about 240 tons of the world's largest sodium and can supply impurity-controlled high temperature sodium to large structural and technology demonstration test sections. It is greatly expected as R&D such as future international cooperation. For the purpose of future R&D tests, this report compiled the design specifications, fabrication and performance confirmation results of sodium test loop.

Journal Articles

Experimental investigation on bubble characteristics entrained by surface vortex

Ezure, Toshiki; Kimura, Nobuyuki; Miyakoshi, Hiroyuki; Kamide, Hideki

Nuclear Engineering and Design, 241(11), p.4575 - 4584, 2011/11

 Times Cited Count:11 Percentile:29.24(Nuclear Science & Technology)

The cover gas entrainment at the free surface of sodium coolant becomes one of the significant issues according to the compact sizing of reactor vessel in the latest design of sodium cooled fast reactor. In the present study, some basic water experiments for the gas entrainment due to the surface vortex were performed in order to obtain the fundamental knowledge about the entrained bubble size. Distributions of entrained bubble diameters in several experimental conditions were obtained from bubble images using an image processing technique. Velocity fields around vortices and surface dimple shapes (gas cores) due to surface vortices were measured to grasp those influences on bubble shapes. The result showed that mean equivalent diameters of bubbles were varied from 1.3 to 2.1 mm in the range of present experimental conditions. The bubble sizes were influenced by the thickness of gas core.

Journal Articles

Sodium experiments on decay heat removal system of Japan sodium cooled fast reactor; Start-up transient of decay heat removal system

Kamide, Hideki; Kobayashi, Jun; Ono, Ayako; Kimura, Nobuyuki; Miyakoshi, Hiroyuki; Watanabe, Osamu*

Proceedings of 7th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-7) (CD-ROM), 8 Pages, 2010/11

Fully natural circulation system is adopted in a decay heat removal system (DHRS) of Japan Sodium Cooled Fast Reactor (JSFR). The JSFR has two loops of the main heat transport system in order to reduce number of components and the construction cost. The DHRS of JSFR consists of one DRACS and two PRACS, which have a heat exchanger in a primary-side inlet plenum of IHX in each loop. Sodium experiments were carried out for heat transfer characteristics of a sodium-sodium heat exchanger of PRACS and also start-up transient of the DHRS loop. Heat transfer coefficient on the tube outer surface was in good agreement with a conventional correlation under operation condition in the reactor. The transient experiments for the start-up of DHRS loop showed that smooth increase of natural draft in the air duct followed by the sodium flow rate in the DHRS loop. Some delay of the flow rate increase was recognized in the DHRS loop as compared with that of the natural draft in the air cooler.

Journal Articles

Experimental study on gas entrainment due to nonstationary vortex in a sodium cooled fast reactor; Comparison of onset conditions between sodium and water

Kimura, Nobuyuki; Ezure, Toshiki; Miyakoshi, Hiroyuki; Kamide, Hideki; Fukuda, Takeshi*

Journal of Engineering for Gas Turbines and Power, 132(10), p.102908_1 - 102908_6, 2010/10

 Times Cited Count:7 Percentile:56.11(Engineering, Mechanical)

In sodium-cooled fast reactor, a compact reactor vessel (R/V) with increased sodium flow velocity was designed to reduce the construction cost. One of the thermal hydraulic problems in this design is gas entrainment at the free surface in the R/V. In most of past studies, water experiments were performed to investigate the gas entrainment in the reactor. It is necessary to evaluate an influence of fluid physical property on the gas entrainment phenomena. In this study, sodium experiments were carried out to clarify the onset criteria of the gas entrainment due to a free surface vortex. Water experiments using a test section in which geometry is the same as that in the sodium tests were also performed. The gas entrainment in water slightly tended to take place in comparison with that in sodium under low velocity conditions. Overall onset condition map on lateral and downward flow velocities in the sodium and water experiments were in good agreement.

Journal Articles

Experimental study on thermal stratification in a reactor vessel of innovative sodium-cooled fast reactor; Mitigation approach of temperature gradient across stratification interface

Kimura, Nobuyuki; Miyakoshi, Hiroyuki; Kamide, Hideki

Journal of Nuclear Science and Technology, 47(9), p.829 - 838, 2010/09

 Times Cited Count:4 Percentile:65.91(Nuclear Science & Technology)

Thermal stratification after a scram is one of main thermal loads of the reactor vessel (R/V) of sodium cooled fast reactor. A water experiment using an 1/10 scale model was carried out. Steep temperature gradient across the stratification interface was observed at the neighborhood of the UIS slit in the experiment. In order to mitigate the temperature gradient across the stratification interface, the height of a plug, which was installed in the upper plenum to infill a hole at a dipped plate for a fuel handling machine, was changed as the parameter in the experiment. The experimental result shows that the temperature gradient near the R/V wall in the case of the lower plug position was smaller than that in the case of the higher plug position.

Journal Articles

Experimental studies of natural circulation decay Heat removal in Japan Sodium Cooled Fast Reactor (JSFR)

Kamide, Hideki; Miyakoshi, Hiroyuki; Watanabe, Osamu*; Eguchi, Yuzuru*; Koga, Tomonari*

Nippon Kikai Gakkai Rombunshu, B, 76(763), p.460 - 462, 2010/03

Fully natural circulation system is adopted in a decay heat removal system (DHRS) of the designs of Japan Sodium Cooled Fast Reactor (JSFR). Several investigations of experiments and simulation methods on this DHRS are performed. Water experiments are carried out for the primary heat transportation system including a reactor vessel and heat exchangers of DHRS using a 1/10 model. As for the DHRS loop, sodium experiments were carried out, especially for a heat exchanger installed in an Intermediate Heat Exchanger (IHX). Here, several results of the sodium experiments were described. Transient characteristics during the start up in the air system of the air cooler, secondary loop of DHRS, and the primary loop were examined by the sodium experiments. Smooth increase of natural circulation flow rates in all systems of air and sodium was confirmed. Verifications of numerical simulation methods are planned based on the water and sodium experiments in this investigation plan.

Journal Articles

Experimental study of gas entrainment phenomena in sodium cooled fast reactor by 1/1.8 scale partial model

Ezure, Toshiki; Kimura, Nobuyuki; Miyakoshi, Hiroyuki; Kamide, Hideki

Nippon Kikai Gakkai Rombunshu, B, 76(763), p.457 - 459, 2010/03

Gas entrainment (GE) at free surface is one of the significant issues for the design of Japan Sodium Cooled Fast Reactor (JSFR). It is required to evaluate the occurrence of GE in the operating condition of JSFR. In the present study, a water experiment was performed in a large-scaled partial upper plenum model to simulate the startup operation of JSFR, where the sodium level was lower than in the rated condition. The onset condition of GE was clarified by the visualization of free surface. Vertical velocity distribution was also measured in order to grasp the mechanism of GE. As a result, it was confirmed that GE did not occur in the startup condition of JSFR.

JAEA Reports

Experimental investigation on gas entrainment in reactor vessel using 1/1.8th scale model; Evaluation of onset condition under low liquid level condition at startup operation

Ezure, Toshiki; Kimura, Nobuyuki; Tobita, Akira; Miyakoshi, Hiroyuki; Kamide, Hideki

JAEA-Research 2009-021, 44 Pages, 2009/09

JAEA-Research-2009-021.pdf:22.4MB

An compact sodium-cooled fast reactor has been investigated in the Fast Reactor Cycle Technology Development Project. The compact-sizing may cause the cover gas entrainment (GE) at the free surface of reactor vessel. The prevention of gas entrainment is one of the significant thermal hydraulic issues. In the previous study, a water experiment in the 1/1.8 scale partial model was performed to check the adequacy of reactor design. From the result, it has been assured GE did not occur in the rated condition of reactor. However, there are the specific operating conditions with the low sodium level due to the thermal contraction concerning the cold startup or shutdown of reactor, where the sodium temperature is low. The objective of this study is to evaluate the onset condition of gas entrainment in those condition. The onset condition of GE was clarified in low liquid level. From the result, it was assured gas entrainment did not occur with the doubled D/P geometry.

Journal Articles

Experimental study on high cycle thermal fatigue in T-junction; Effect of local flow velocity on transfer of temperature fluctuation from fluid to structure

Kimura, Nobuyuki; Ono, Ayako; Miyakoshi, Hiroyuki; Kamide, Hideki

Proceedings of 13th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-13) (CD-ROM), 16 Pages, 2009/09

A quantitative evaluation on high cycle thermal fatigue due to temperature fluctuation in fluid is of importance for structural integrity in the reactor. It is necessary for the quantitative evaluation to investigate occurrence and propagate processes of temperature fluctuation, e. g. decay of temperature fluctuation near structures and transfer of temperature fluctuation from fluid to structures. In this study, a water experiment using T-junction was performed to evaluate the transfer characteristics of temperature fluctuation from fluid to structure. In the experiment, temperatures and local velocities were measured simultaneously to evaluate the correlation between the temperature and velocity under the non-stationary fields. The large heat transfer coefficients were registered at the region where the local velocity was high.

Journal Articles

Experimental investigation on characteristic of entrained bubbles due to surface vortex

Ezure, Toshiki; Kimura, Nobuyuki; Miyakoshi, Hiroyuki; Kamide, Hideki

Proceedings of 13th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-13) (CD-ROM), 16 Pages, 2009/09

According to the compact sizing of reactor vessel, gas entrainment at the free surface of sodium coolant becomes one of the significant issues in the latest sodium-cooled fast reactor design. In the present study, some basic experiments for gas entrainment due to surface vortices were performed in order to obtain the fundamental knowledge about entrained bubbles. The distribution of bubble diameters was obtained from an image processing technique from the images. In addition to that, velocity field and surface shape of gas core were measured to grasp the influences of the local velocity field on the gas core shape and detached bubbles. The results showed that the mean equivalent diameters of bubbles were varied from 1.34 to 2.06 mm in the range of present experimental conditions.

Journal Articles

Study on velocity field in a wire wrapped fuel pin bundle of sodium cooled reactor; Detailed velocity distribution in a subchannel

Sato, Hiroyuki; Kobayashi, Jun; Miyakoshi, Hiroyuki; Kamide, Hideki

Proceedings of 13th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-13) (CD-ROM), 13 Pages, 2009/09

A sodium cooled fast reactor is designed to attain a high burn-up core in a feasibility study on commercialized fast reactor cycle systems. It is of importance to obtain the flow velocity distribution in a wire wrapped pin bundle for the high burn-up core. A 2.5 times enlarged 7-pin bundle water model was applied to investigate the detailed velocity distribution in an inner subchannel surrounded by 3 pins with wrapping wire. The velocity distribution in an inner subchannel with the wrapping wire was measured by Particle Image Velocimetry. In the vertical velocity distribution in a narrow space between the pins, the wrapping wire decreased the velocity downstream of the wire and asymmetrical flow distribution was formed between the pin and wire. In the horizontal velocity distribution, swirl flow around the wrapping wire was obviously observed.

Journal Articles

Experimental study of gas entrainment phenomena in sodium cooled fast reactor by 1/1.8 scale partial model

Ezure, Toshiki; Kimura, Nobuyuki; Miyakoshi, Hiroyuki; Kamide, Hideki

Dai-14-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu, p.423 - 424, 2009/06

Gas entrainment (GE) at free surface is one of the significant issues for the design of Japan Sodium Cooled Fast Reactor. In the previous study, authors confirmed that GE did not occur under the conditions of the rated operating mode of the reactor. In the present study, a water experiment was performed in a large-scaled partial model of the reactor upper plenum to simulate the startup operation of reactor where the sodium level was lower than in the rated condition. The onset condition of GE was observed by the visualization of free surface. Vertical velocity distribution was also measured in order to qualify the mechanism of GE. As a result, it was confirmed that GE did not occur in the startup condition of reactor.

Journal Articles

Experimental studies of natural circulation decay heat removal in Japan Sodium Cooled Fast Reactor (JSFR)

Kamide, Hideki; Miyakoshi, Hiroyuki; Watanabe, Osamu*; Eguchi, Yuzuru*; Koga, Tomonari*

Dai-14-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu, p.427 - 428, 2009/06

Fully natural circulation system is adopted in a decay heat removal system (DHRS) of the designs of Japan Sodium Cooled Fast Reactor (JSFR). Several investigations of experiments and simulation methods on this DHRS are performed. Water experiments were carried out for the primary heat transportation system including a reactor vessel and heat exchangers of DHRS using a 1/10 model. As for the DHRS loop, sodium experiments were carried out, especially for a heat exchanger installed in an Intermediate Heat Exchanger (IHX). Here, several results of the sodium experiments were described. Transient characteristics during the start up in the air system of the air cooler, secondary loop of DHRS, and the primary loop were examined by the sodium experiments. Smooth increase of natural circulation flow rates in all systems of air and sodium were confirmed. Verifications of numerical simulation methods are planned based on the water and sodium experiments in this investigation plan.

Journal Articles

Experimental study on gas entrainment due to non-stationary vortex in a sodium cooled fast reactor; Comparison of onset conditions between sodium and water

Kimura, Nobuyuki; Ezure, Toshiki; Miyakoshi, Hiroyuki; Kamide, Hideki; Fukuda, Takeshi*

Proceedings of 17th International Conference on Nuclear Engineering (ICONE-17) (CD-ROM), 8 Pages, 2009/06

In an innovative sodium cooled fast reactor, a compact reactor vessel (R/V) with increased sodium flow velocity was designed to reduce the construction cost. One of the thermal hydraulic problems in this design is gas entrainment at the free surface in the R/V. Sodium experiments were carried out to clarify the onset criteria of the gas entrainment due to a free surface vortex. Water experiments using a test section in which geometry is the same as that in the sodium tests were also performed. The gas entrainment in water slightly tended to take place in comparison with that in sodium under low velocity conditions.

Journal Articles

Experimental study on onset of gas entrainment by 1/1.8 scale model of sodium cooled fast reactor

Ezure, Toshiki; Kimura, Nobuyuki; Miyakoshi, Hiroyuki; Kamide, Hideki

Proceedings of 6th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-6) (USB Flash Drive), 6 Pages, 2008/11

Gas entrainment (GE) at free surface is one of the significant issues for the design of Japanese Sodium Cooled Fast Reactor. In the previous study, authors confirmed that GE did not occur at the rated operating mode in the reactor. In the present study, a water experiment to simulate the startup operation of reactor was performed in a large-scaled partial model of the reactor upper plenum in the reactor. The onset condition of GE was observed by the visualization of free surface. Vertical velocity distribution was also measured in order to qualify the mechanism of GE. As a result, it was confirmed that GE did not occur in the startup condition of reactor.

Journal Articles

Experimental study on thermal stratification in a reactor vessel of innovative sodium cooled fast reactor; Mitigation approach of temperature gradient across stratification interface

Kimura, Nobuyuki; Miyakoshi, Hiroyuki; Kamide, Hideki

Proceedings of 6th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-6) (USB Flash Drive), 7 Pages, 2008/11

An innovative sodium cooled fast reactor has been investigated as part of the fast reactor cycle technology development (FaCT) project. Thermal stratification after a scram is one of main thermal loads of the reactor vessel (R/V). An upper inner structure (UIS) has a slit in radial direction for fuel handling. A water experiment using an 1/10 scale model was carried out. Steep temperature gradient across the stratification interface was observed at the neighborhood of the UIS slit in the experiment. In order to mitigate the temperature gradient across the stratification interface, the height of a plug, which was installed in the upper plenum to infill a hole at a dipped plate for a fuel handling machine, was changed as the parameter in the experiment. The experimental result shows that the temperature gradient near the R/V wall in the case of the lower plug position was 13% smaller than that in the case of the higher plug position.

Journal Articles

Study on velocity field in a deformed fuel pin bundle; Influence of pin deformation and wrapping wire on velocity distribution

Sato, Hiroyuki; Kobayashi, Jun; Miyakoshi, Hiroyuki; Kamide, Hideki

Proceedings of 16th International Conference on Nuclear Engineering (ICONE-16) (CD-ROM), 9 Pages, 2008/05

A sodium cooled fast reactor is designed to attain a high burn-up core in a feasibility study on commercialized fast reactor cycle systems. In high burn-up fuel subassemblies, deformation of fuel pin due to the swelling and thermal bowing may decrease flow rate via change of flow area in the sub-assembly and influence the heat removal capability. A 2.5 times enlarged 7-pin bundle water model was applied to investigate the influence of pin bowing and wrapping wire. The test section consisted of a hexagonal acrylic duct tube and fluorinated resin pins which had the refractive index nearly the same with water and high light transmission rate. This enabled to visualize around the central pin through the outer pins. Velocity distribution was measured by using PIV in sub-channels around the central pin in reference and deformation condition. Velocity distribution around the wrapping wire was measured and the wire influenced to the velocity and RMS in the wide region near the pin surface.

Journal Articles

Experimental investigation on transfer characteristics of temperature fluctuation from liquid sodium to wall in parallel triple-jet

Kimura, Nobuyuki; Miyakoshi, Hiroyuki; Kamide, Hideki

International Journal of Heat and Mass Transfer, 50(9-10), p.2024 - 2036, 2007/05

 Times Cited Count:43 Percentile:15.32(Thermodynamics)

A quantitative evaluation on thermal striping, in which temperature fluctuation due to convective mixing causes thermal fatigue in structural components, is of importance for integrity of nuclear reactors and also general plants. Sodium cooled reactor had also several incidents of coolant leakage due to the high cycle thermal fatigue. A sodium experiment of parallel triple-jet configuration was performed to evaluate transfer characteristics of temperature fluctuation from fluid to structure. The non-stationary heat transfer characteristics could be represented by a heat transfer coefficient, which was constant in time and independent of the frequency of the temperature fluctuation.

JAEA Reports

Study on fluid mixing in a fuel subassembly of a fast reactor; Mixing phenomena with cross flow and applicability of subchannel analysis

Kawamata, Nobuhiro; Miyakoshi, Hiroyuki; Kamide, Hideki

JNC-TN9400 2004-047, 99 Pages, 2004/08

JNC-TN9400-2004-047.pdf:8.02MB

Sodium experiments were carried out for mixing phenomena in a subassembly of fast reactor. A blockage was installed in a wire wrapped 37-pin subassembly in order to form cross flow around the blockage. Temperature distributions in the subassembly were measured under a condition that only one pin was heated. Investigation on mixing phenomena with the cross flow is useful to predict thermohydraulics in a deformed pin bundle of high burnup core. The subchannel analysis code, ASFRE, was applied to the experimental analyses. The ASFRE solves momentum equations in multi dimensions and has the distributed resistance model for a wire spacer.Following remarks were obtained:1).Measured axial temperature distributions showed that the sodium temperature decreased due to the core flow coming from the cold blockage side. Local peaks were found at the subchannels which have the wire spacer. 2).ASFRE simulated the temperature decreases due to the cross flow and axial temperature profiles were in good agreement with the measured data.3).The local peaks at wire positions were also simulated by ASFRE, however, the peak value was underestimated. ASFRE has high applicability to the mixing phenomena in a wire wrapped subassembly where the cross flow has significant role due to the change of flow area.4).Experimental database for the mixing phenomena with the cross flow were obtained for a subassembly of sodium cooled fast reactor.

JAEA Reports

Study on Thermalhydraulics in Thermal Striping Phenomena; Evaluation on Convective mixing in Sodium and Water based on Sodium/Water Experiments and Quasi-DNS

Kimura, Nobuyuki; Nagasawa, Kazuyoshi*; Miyake, Yasuhiro*; Miyakoshi, Hiroyuki; Kamide, Hideki

JNC-TN9400 2004-064, 216 Pages, 2004/05

JNC-TN9400-2004-064.pdf:27.81MB

Sodium and watrer experiments on thermal striping phenomena were carried out. Triple-parallel jet along a wall was investigated. Further, quasi-DNS was also performed and comparisons between experiments and analyses were carried out. Temperature fluctuation intensity near the wall was larger in water than in sodium. This means that water experiment can predict sodium temperature fluctuation intensity with certain conservative factor.

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