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JAEA Reports

Flow separation at inlet causing transition and intermittency in circular pipe flow

Ogawa, Masuro*

JAEA-Technology 2019-010, 22 Pages, 2019/07

JAEA-Technology-2019-010.pdf:1.5MB

Transition phenomena from laminar to turbulent flow are roughly classified into three categories. Circular pipe flow of the third category is linearly stable against any small disturbance, despite that flow actually transitions and transitional flow exhibits intermittency. These are among major challenges that are yet to be resolved in fluid dynamics. Thus, author proposes hypothesis as follows; "Flow in a circular pipe transitions from laminar flow because of vortices released from separation bubble forming in vicinity of inlet of pipe, and transitional flow becomes intermittent because vortex-shedding is intermittent." Present hypothesis can easily explain why linear stability theory has not been able to predict transition in circular pipe flow, why circular pipe flow actually transitions, why transitional flow actually exhibits intermittency even due to small disturbance, and why numerical analysis has not been able to predict intermittency of transitional flow in circular pipe.

Journal Articles

Free convective heat transfer experiment to validate air-cooling performance analysis of fuel debris

Uesawa, Shinichiro; Yamashita, Susumu; Shibata, Mitsuhiko; Yoshida, Hiroyuki

Proceedings of 11th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-11) (Internet), 6 Pages, 2018/11

Journal Articles

Particle-based simulation of heat transfer behavior in EAGLE ID1 in-pile test

Morita, Koji*; Ogawa, Ryusei*; Tokioka, Hiromi*; Liu, X.*; Liu, W.*; Kamiyama, Kenji

Proceedings of 12th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, Operation and Safety (NUTHOS-12) (USB Flash Drive), 11 Pages, 2018/10

The EAGLE in-pile ID1 test has been performed by Japan Atomic Energy Agency to demonstrate early fuel discharge from a fuel subassembly with an inner duct structure, which is named FAIDUS. It was deduced that early duct wall failure observed in the test was initiated by high heat flux from the molten pool of fuel and steel mixture. The posttest analyses suggest that molten pool-to-duct wall heat transfer might be enhanced effectively by the molten steel with large thermal conductivity in the pool without the presence of fuel crust on the duct wall. In this study, mechanisms of heat transfer from the molten pool to the duct wall was analyzed using a fully Lagrangian approach based on the finite volume particle method for multi-component, multi-phase flows. A series of pin disruption, molten pool formation and duct wall failure behaviors was simulated to investigate mixing and separation behavior of molten steel and fuel in the pool, and their effect on molten pool-to-duct wall heat transfer. The present 2D particle-based simulations demonstrated that large thermal load beyond 10 MW/m$$^{2}$$ on the duct wall was caused by effective heat transfer due to direct contact of liquid fuel with nuclear heat to the duct wall.

Journal Articles

Evaluation of heat removal during the failure of the core cooling for new critical assembly

Eguchi, Yuta; Sugawara, Takanori; Nishihara, Kenji; Tazawa, Yujiro; Tsujimoto, Kazufumi

Proceedings of 26th International Conference on Nuclear Engineering (ICONE-26) (Internet), 8 Pages, 2018/07

In order to investigate the basic neutronics characteristics of the accelerator-driven subcritical system (ADS), JAEA has a plan to construct a new critical assembly in the J-PARC project, Transmutation Physics Experimental Facility (TEF-P). This study aims to evaluate the natural cooling characteristics of TEF-P core which has large decay heat by minor actinide (MA) fuel, and to achieve a design that does not damage the core and the fuels during the failure of the core cooling system. In the evaluation of the TEF-P core temperature, empty rectangular lattice tube outer of the core has a significant effect on the heat transfer characteristics. The experiments by using the mockup device were performed to validate the heat transfer coefficient and experimental results were obtained. By using the obtained experimental results, the three-dimensional heat transfer analysis of TEF-P core were performed, and the maximum core temperature was obtained, 294$$^{circ}$$C. This result shows TEF-P core temperature would be less than 327$$^{circ}$$C that the design criterion of temperature.

Journal Articles

Development of experimental technology for simulated fuel-assembly heating to address core-material-relocation behavior during severe accident

Abe, Yuta; Yamashita, Takuya; Sato, Ikken; Nakagiri, Toshio; Ishimi, Akihiro; Nagae, Yuji

Proceedings of 26th International Conference on Nuclear Engineering (ICONE-26) (Internet), 9 Pages, 2018/07

Journal Articles

Evaluation and demonstration of cutting the fuel assembly heating examination by AWJ

Maruyama, Shinichiro*; Watatani, Satoshi*

Mitsui Sumitomo Kensetsu Gijutsu Kenkyu Kaihatsu Hokoku, (15), p.107 - 112, 2017/10

It is essential to estimate characteristics and forms of fuel debris for safe and reliable removing at the decommissioning of the Fukushima Daiichi Nuclear Power Plant (1F). For the estimation, melting behavior of fuel assembly in the accident is being researched. To proceed the research, the fuel debris were need to cut, and the abrasive water jet (AWJ) which had enough results for cutting ceramic material or mixed material of zirconium alloy and stainless. The test results demonstrated that AWJ could cut the fuel assembly and accumulated the cutting data which will be subservient when removing the fuel debris in future.

Journal Articles

Development of non-transfer type plasma heating technology to address CMR behavior during severe accident with BWR design conditions

Abe, Yuta; Sato, Ikken; Nakagiri, Toshio; Ishimi, Akihiro; Nagae, Yuji

Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 7 Pages, 2017/04

JAEA Reports

Thermal design study of lead-bismuth cooled accelerator driven system, 1; Study on thermal hydraulic behavior under normal operation condition

Akimoto, Hajime; Sugawara, Takanori

JAEA-Data/Code 2016-008, 87 Pages, 2016/09

JAEA-Data-Code-2016-008.pdf:15.62MB

Thermal hydraulic behavior in a lead-bismuth cooled accelerator driven system (ADS) is analyzed under normal operation condition. Input data for the ADS version of J-TRAC code have been constructed to integrate the conceptual design. The core part of the ADS is modeled in detail to evaluate the core radial power profile effect on the core cooling. As the result of the analyses, the followings are found; (1) Both maximum clad temperature and fuel temperature are below the design limits. (2) The radial power profile has little effect on the coolant flow distribution among fuel assemblies. (3) The radial power profile has little effect on the heat transfer coefficients along fuel rods. (4) The thermal hydraulic behaviors along four steam generators are identical. The thermal hydraulic behaviors along two pumps are also identical. A fast running input data is developed by the simplification of the detailed input data based on the findings mentioned above.

Journal Articles

Preparation for a new experimental program addressing core-material-relocation behavior during severe accident with BWR design conditions; Conduction of preparatory tests applying non-transfer-type plasma heating technology

Abe, Yuta; Sato, Ikken; Ishimi, Akihiro; Nakagiri, Toshio; Nagae, Yuji

Proceedings of 24th International Conference on Nuclear Engineering (ICONE-24) (DVD-ROM), 7 Pages, 2016/06

A new experimental program using non-transfer type plasma heating is under consideration in JAEA to clarify the uncertainty on core-material relocation (CMR) behavior of BWR. In order to confirm the applicability of this new technology, authors performed preparatory plasma heating tests using small-scale test pieces (107 mm $$times$$ 107 mm $$times$$ 222 mmh). Based on these preliminary results, an excellent perspective in terms of applicability of the non-transfer plasma heating technology to the SA (Severe Accident) experimental study was obtained. Furthermore, JAEA is preparing for the next step intermediate-scale preparatory tests in 2016 using ca. 50 rods and a control blade that would not only confirm its technical applicability, but also some insights relevant to the issue on CMR itself.

Journal Articles

Pool nucleate boiling for seawater containing minerals

Uesawa, Shinichiro; Koizumi, Yasuo; Yoshida, Hiroyuki

Proceedings of 9th International Conference on Multiphase Flow (ICMF 2016) (CD-ROM), 6 Pages, 2016/05

JAEA Reports

Fabrication and test results of testing equipment for remote-handling of MA fuel, 2; Evaluation of heat transfer parameter crossing rectangular lattice matrix

Eguchi, Yuta; Sugawara, Takanori; Nishihara, Kenji; Tazawa, Yujiro; Inoue, Akira; Tsujimoto, Kazufumi

JAEA-Technology 2015-052, 34 Pages, 2016/03

JAEA-Technology-2015-052.pdf:5.02MB

Transmutation Physics Experimental Facility (TEF-P) planned in the J-PARC project uses minor actinide (MA) fuel which has large decay heat. So it is necessary to consider the increase of the core temperature when the core cooling system is stopped. This change of the core temperature was evaluated by thermal conduction analysis. It was found that the impact of thermal insulation in the empty rectangular lattice matrix area was large. So it is necessary to verify reliability and accuracy of heat transfer effect used in this area. Testing equipment was fabricated to verify the accuracy of calculation model for the empty lattice matrix which was the free convection model of sealed fluid. By using this equipment, thermal distribution and one dimensional heat flow through the lattice were measured. It was observed that the actual equivalent thermal conductivity in the lattice was larger than the free convection model. It was also confirmed that the insertion of the aluminum block into the empty lattice could achieve the higher equivalent thermal conductivity. These results could be the useful data for the thermal conduction analysis.

Journal Articles

Evaluation of seawater effects on thermal-hydraulic behavior for severe accident conditions, 2; Heat transfer and flow visualization experiment by using internally heated annulus

Uesawa, Shinichiro; Nagatake, Taku; Jiao, L.; Liu, W.; Takase, Kazuyuki; Yoshida, Hiroyuki

Proceedings of International Conference on Power Engineering 2015 (ICOPE 2015) (CD-ROM), 11 Pages, 2015/11

Journal Articles

Evaluation of seawater effects on thermal-hydraulic behavior for severe accident conditions, 1; Outline of the research project

Yoshida, Hiroyuki; Uesawa, Shinichiro; Nagatake, Taku; Jiao, L.; Liu, W.; Takase, Kazuyuki

Proceedings of International Conference on Power Engineering 2015 (ICOPE 2015) (CD-ROM), 9 Pages, 2015/11

Journal Articles

Prediction of heater surface temperature change at subcooled flow boiling DNB

Liu, W.; Podowski, M. Z.*

Nippon Kikai Gakkai Netsu Kogaku Konfarensu 2015 Koen Rombunshu (CD-ROM), 2 Pages, 2015/10

This paper gives prediction to the transient heat transfer at Departure of Nucleate Boiling (DNB) point for subcooled flow boiling. The prediction is carried out by solving the heat conduction equations in cylindrical coordinates with convective boundary condition, which changes with the change of the heat transfer mode on the heated surface. DNB is assumed to happen at the complete dryout of liquid sublayer trapped between the heated wall and an elongated vapor clot, during the passing time of the vapor clot. Important parameters including initial thickness of the liquid sublayer, vapor clot length, vapor clot velocity and void fraction etc., are calculated from the Liu - Nariai model. The initial heater surface temperature is derived from the Jens-Lottes correlation. The transient changes of liquid sublayer thickness, surface temperature at DNB are reported. No obvious temperature jumping is observed at DNB. To predict temperate excursion at Critical Heat Flux (CHF), more simulations to the transient boiling and film boiling processes are needed.

JAEA Reports

Development of thermal-hydraulic design code for transmutation system with lead-bismuth cooled accelerator driven reactor

Akimoto, Hajime

JAEA-Data/Code 2014-031, 75 Pages, 2015/03

JAEA-Data-Code-2014-031.pdf:37.23MB

A thermal-hydraulic analysis code for transmutation system with lead-bismuth cooled accelerator-driven system (ADS) has been developed using the Japanese-version of Transient Reactor Analysis Code (J-TRAC) as the framework to apply the design studies of ADS. To identify the required capabilities of the thermal-hydraulic analysis code for ADS, previous thermal-hydraulic analyses of light water reactors, sodium-cooled fast reactor and ADS have been surveyed. To make up for insufficient capabilities of the J-TRAC code as a thermal-hydraulic analysis code of ADS, physical properties of lead-bismuth eutectic (LBE), argon gas and nitride nuclear fuel were implemented to the J-TRAC code. It was confirmed that the implemented capabilities worked as expected through verification calculations on (1) single-phase LBE flow, (2) heat transfer in a fuel assembly, and (3) heat transfer in a steam generator.

Journal Articles

CFD analysis

Takase, Kazuyuki; Misawa, Takeharu*

Supercritical-Pressure Light Water Cooled Reactors, p.301 - 319, 2014/12

no abstracts in English

Journal Articles

Study on drop wise condensation by using functionalized heat transfer surface

Koizumi, Yasuo; Yoshizawa, Shota*

Proceedings of the ASME 2014 International Mechanical Engineering Congress and Exposition (IMECE 2014) (DVD-ROM), 7 Pages, 2014/11

The enhancement of drop wise condensation heat transfer by functionalizing a heat transfer surface was examined for 0.1 MPa steam. A gold-plated surface was used to produce the drop wise condensation. Rectangular-grooved heat transfer surfaces were adopted to functionalize the heat transfer surface. The size of the grooves were 2 mm $$times$$ 2 mm $$times$$ 2 mm, 3 mm $$times$$ 3 mm $$times$$ 3 mm and 2 mm $$times$$ 3 mm $$times$$ 2 mm (depth $$times$$ top width $$times$$ bottom width), respectively. The heat flux of the grooved surface was larger than that of the plain gold-plated surface. When the groove size was 2 mm $$times$$ 2 mm $$times$$ 2 mm and the top parts and the walls of grooves were plated with gold, the heat transfer rate augmentation was highest; the augmentation rate was 1.53. Since to increase the width of the top part of the grooves tended to bring the quality of the surface structure close to the plain surface, it was not right direction. It was also implied that to make summits and troughs on the surface to collect condensate tended to expose the summit part to steam more, which might result in the heat transfer augmentation.

Journal Articles

Study on nucleate boiling heat transfer by measuring instantaneous surface temperature distribution by infrared radiation camera

Koizumi, Yasuo; Takahashi, Kazuki*

Proceedings of 15th International Heat Transfer Conference (IHTC 2014) (USB Flash Drive), 13 Pages, 2014/08

Pool nucleate boiling heat transfer experiments were performed for water at 0.101 MPa to examine the elementary process of the nucleate boiling. Heat transfer surface was made from a copper printed circuit board. Direct current was supplied to heat it up. The Bakelite plate of the backside of a copper layer was taken off at the center portion of the heat transfer surface to expose the back side of the copper layer. The instantaneous variation of the backside temperature of the heat transfer surface was measured with an infrared radiation camera. Bubble behavior was recorded with a high speed video camera. In the isolated bubble region, surface temperature was uniform during waiting time. When boiling bubble generation started, a large dip in the surface temperature distribution was formed under the bubble. After the bubble left from the heat transfer surface, the surface temperature distribution returned to former uniform distribution. Surface temperature was not affected by the bubble generation beyond 1.8 mm from the center of the bubble. In the intermediate and high heat flux region, the variation of surface temperature and heat flux did not become large even though the heat flux increased. The heat flux variation was close to that at the isolated boiling region.

Journal Articles

Measurement of LBE flow velocity profile by UDVP

Kikuchi, Kenji; Takeda, Yasushi*; Obayashi, Hiroo*; Tezuka, Masao*; Sato, Hiroshi

Journal of Nuclear Materials, 356(1-3), p.273 - 279, 2006/09

 Times Cited Count:8 Percentile:43.49(Materials Science, Multidisciplinary)

Measurements of LBE flow velocity profile were realized in the spallation target model by the ultrasonic Doppler velocity profile technique. Hitherto, it has not yet been done well because both of poor wetting property of LBE with stainless steels and poor performance of supersonic probes at high temperatures. Measurement was made for a return flow in the target model, which has coaxially arranged annular and tube channels. The electromagnetic pump generates LBE flow and the flow rate was measured by the electromagnetic flow meter. Measurement results show that re-circulation occurred near the surface of beam window, which might affect a heat transfer of target container.

Journal Articles

Experimental examination of heat removal limitation of screw cooling tube at high pressure and temperature conditions

Ezato, Koichiro; Suzuki, Satoshi; Dairaku, Masayuki; Akiba, Masato

Fusion Engineering and Design, 81(1-7), p.347 - 354, 2006/02

 Times Cited Count:7 Percentile:47.89(Nuclear Science & Technology)

no abstracts in English

307 (Records 1-20 displayed on this page)