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Journal Articles

Establishment of numerical model to investigate heat transfer and flow characteristics by using scale model of vessel cooling system for HTTR

Takada, Shoji; Narayana, I. W.*; Nakatsuru, Yukihiro*; Terada, Atsuhiko; Murakami, Kenta*; Sawa, kazuhiko*

Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 13 Pages, 2019/05

In the loss of core cooling test using HTTR, a technical issue is to improve prediction accuracy of temperature distribution of components in vessel cooling system (VCS). An establishment of reasonable 2D model was started by using numerical code FLUENT, which was validated using the test data by 1/6 scale model of VCS for HTTR. The pressure vessel (PV) temperature was set around 200$$^{circ}$$C attributed to relatively high ratio of natural convection heat transfer around 20% in total heat removal, which is useful for code to experiment benchmark to improve prediction accuracy. It is necessary to confirm heat transfer flow characteristics around the top of PV which is heated up by natural convection flow which was considered to be affected by separation, re-adhesion and transition flow. The k-$$omega$$-SST model was selected for turbulent calculation attributed to predict the effects mentioned above adequately. The numerical results using the k-$$omega$$-SST model reproduced the temperature distribution of PV especially the top region which is considered to be affected by separation, re-adhesion and transition flow in contract to that using k-$$varepsilon$$ model which does not account the effects.

Journal Articles

Improvement of heat-removal capability using heat conduction on a novel reactor cavity cooling system (RCCS) design with passive safety features through radiation and natural convection

Takamatsu, Kuniyoshi; Matsumoto, Tatsuya*; Liu, W.*; Morita, Koji*

Annals of Nuclear Energy, 122, p.201 - 206, 2018/12

 Percentile:100(Nuclear Science & Technology)

A RCCS having passive safety features through radiation and natural convection was proposed. The RCCS design consists of two continuous closed regions: an ex-reactor pressure vessel region and a cooling region with a heat-transfer surface to ambient air. The RCCS uses a novel shape to remove efficiently the heat released from the RPV through as much radiation as possible. Employing air as the working fluid and ambient air as the ultimate heat sink, the RCCS design can strongly reduce the possibility of losing the working fluid and the heat sink for decay-heat-removal. This study addresses an improvement of heat-removal capability using heat conduction on the RCCS. As a result, a heat flux removed by the RCCS could be doubled; therefore, it is possible to halve the height of the RCCS or increase the thermal reactor power.

Journal Articles

Experimental study on heat removal performance of a new Reactor Cavity Cooling System (RCCS)

Hosomi, Seisuke*; Akashi, Tomoyasu*; Matsumoto, Tatsuya*; Liu, W.*; Morita, Koji*; Takamatsu, Kuniyoshi

Proceedings of 11th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-11) (Internet), 7 Pages, 2018/11

A new RCCS with passive safety features consists of two continuous closed regions. One is a region surrounding RPV. The other is a cooling region with heat transferred to the ambient air. The new RCCS needs no electrical or mechanical driving devices. We started experiment research with using a scaled-down test section. Three experimental cases under different emissivity conditions were performed. We used Monte Carlo method to evaluate the contribution of radiation to the total heat released from the heater. As a result, after the heater wall was painted black, the contribution of radiation to the total heat could be increased to about 60%. A high emissivity of RPV surface is very effective to remove more heat from the reactor. A high emissivity of the cooling part wall is also effective because it not only increases the radiation emitted to the ambient air, but also may increase the temperature difference among the walls and enhance the convection heat transfer in the RCCS.

Journal Articles

Development of numerical simulation method to evaluate heat transfer performance of air around fuel debris, 2; Validation of JUPITER for free convection heat transfer

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

Proceedings of 25th International Conference on Nuclear Engineering (ICONE-25) (CD-ROM), 7 Pages, 2017/07

Journal Articles

Study of the flow characteristics of coolant channel of fuel blocks for HTGR

Tsuji, Nobumasa*; Ohashi, Kazutaka*; Tazawa, Yujiro*; Tachibana, Yukio; Ohashi, Hirofumi; Takamatsu, Kuniyoshi

FAPIG, (190), p.20 - 24, 2015/07

In a loss of forced cooling accident, decay heat in HTGRs must be removed by radiation, thermal conduction and natural convection. Passive heat removal performance is of primary concern for enhancing inherent safety features of HTGRs. Therefore, the thermal hydraulic analyses for normal operation and a loss of forced cooling accident are conducted by using thermal hydraulic CFD code. And further, a multi-hole type fuel block of MHTGR is also modeled and the flow and heat transfer characteristics are compared with a pin-in-block type fuel block.

Journal Articles

Infrared thermography for analyzing heat transfer and fluid flow of benard-cell convection in a rectangular container with free surface

Inagaki, Terumi*; Hatori, Masakazu*; Suzuki, Tomohiro*; Shiina, Yasuaki

Proceedings of International Conference on Advanced Optical Diagnostics in Fluids, Solids and Combustion (VSJ-SPIE '04) (CD-ROM), 9 Pages, 2004/12

no abstracts in English

Journal Articles

Heat transfer phenomenon of natural convection in an open vessel and its infrared sensing

Inagaki, Terumi*; Kaneko, Toshinobu*; Hatori, Masakazu*; Shiina, Yasuaki

Nippon Kikai Gakkai Rombunshu, B, 70(699), p.279 - 286, 2004/11

no abstracts in English

Journal Articles

Research and development on passive cooling system

Takada, Shoji

Nuclear Engineering and Design, 233(1-3), p.185 - 195, 2004/10

 Times Cited Count:5 Percentile:59.73(Nuclear Science & Technology)

Experiments are carried out to investigate the effects of the natural convection of superheated gas as well as of the stand pipes on the temperature distributions of the components and the heat removal performance in the water-cooling panel system for the MHTGR for decay heat removal, and to verify the design and evaluation methods. The numerical results of the code THANPACST2 are compared with the experimental data to verify the numerical methods and axi-symmetric model proposed, which can simulate the three-dimensional configuration of the stand pipes on the upper head of the pressure vessel by using porous body cells. The experiments revealed that temperatures increased with elevation on the upper head, because the stand pipes restrict radiation heat transfer to the upper cooling panel and reduce the heat transfer area on the upper head which was superheated by natural convection of helium gas in the pressure vessel. The numerical methods were able to closely duplicate the pattern of the rising temperature profile with elevation around the top of the upper head.

Journal Articles

An Experimental study of heat transfer in an open thermosyphon

Imai, Etsuya*; Shiina, Yasuaki; Hishida, Makoto*

Heat Transfer-Asian Research, 30(4), p.301 - 312, 2001/06

no abstracts in English

Journal Articles

Three-dimensional numerical simulations of dust mobilization and air ingress characteristics in a fusion reactor during a LOVA event

Takase, Kazuyuki

Fusion Engineering and Design, 54(3-4), p.605 - 615, 2001/04

 Times Cited Count:10 Percentile:33.22

no abstracts in English

JAEA Reports

CompalisonoFnlermohydraulicCharacteristicsintheuseofvariousCoolants

; ; *; Yamaguchi, Akira

JNC-TN9400 2000-109, 96 Pages, 2000/11

JNC-TN9400-2000-109.pdf:9.56MB

Numerical calculations were carried out for a free surface sloshing, a thermal stratification, a thermal striping, and a natural convection as key phenomena of in-vessel thermohydraulics in future fast reactor systems with various fluids as coolants. This numerical work was initiaied based on a recognition that the fundamental characteristics of the phenomena have been unsolved quantitatively in the use of various coolants. From the analysis for the phenomena, the following results were obtained. [Free Surface Sloshing phenomena] (1)Ther is no remarkable difference betweeen liquid sodium and luquid Pb-Bi in characteristics of internal flows and free surface charatristics based on Fr number. (2)the AQUA-VOF code has a potentiall enough to evaluate gas entrainment behavior from the free surface including the internal flow characteristics. [thermal Stratification Phenomena] (1)On-set position of thermal entainment process due to dynamic vortex flows was moved to downstream direction with decreasing of Ri number. 0n the other hand, the position in the case of C0$$_{2}$$ gas was shifted to upstream side with decreasing of Ri number. (2)Destruction speed of the thermal stratyification interface was dependent on thermal diffusivity as fluid properties. therefor it was concluded that an elimination method is necessary for the interface generated in C0$$_{2}$$ gas. [thermal Striping Phenomena] (1)Large amplitudes of fluid temperature fluctuations was reached to down stream area in the use of CO$$_{2}$$ gas, due to larger fluid viscosity and smaller thermal diffusivity, compared with liquid sodium and liquid Pb-Bi cases. (2)To simulate thermal striping conditions such as amplitude and frequency of the fluid temperature fluctuations, it isnecessary for coincidences of Re number for the amplitude and of velocity value for the frequency, in various coolants. [Natural Convection Phenomena] (1)Fundamental behavior of the natural convection in various coolant follows buoyant jet ....

JAEA Reports

Study on natural convection heat transfer in a vertical enclosure of double coaxial cylinder; Cooling by natural circulation of air

Zhang, Y.*; Takeda, Tetsuaki; Inaba, Yoshitomo

JAERI-Tech 2000-065, 109 Pages, 2000/11

JAERI-Tech-2000-065.pdf:6.17MB

no abstracts in English

Journal Articles

Deposition of cesium iodide particles in bends and sections of vertical pipe under severe accident conditions

Hidaka, Akihide; Shibazaki, Hiroaki*; Yoshino, T.*; Sugimoto, Jun

Journal of Aerosol Science, 31(9), p.1045 - 1059, 2000/09

 Times Cited Count:1 Percentile:80.66

no abstracts in English

Journal Articles

Natural convection heat transfer with micro-encapsulated phase-change-material slurry

Kubo, Shinji; Akino, Norio

Transactions of the American Nuclear Society, 81, p.352 - 353, 1999/11

no abstracts in English

Journal Articles

Melting of phase change fluid in heat storage capsules by convection

; Shiina, Yasuaki; Inagaki, Terumi*

Kashika Joho Gakkai-Shi, 19(75), p.41 - 45, 1999/10

no abstracts in English

JAEA Reports

A Note on the representation of rate-of-rise of the thermal stratification interface in reactor plenum

Tokuhiro, Akira; Kimura, Nobuyuki

JNC-TN9400 2000-015, 26 Pages, 1999/09

JNC-TN9400-2000-015.pdf:1.43MB

The quantification of the rate-of-rise of the thermal stratification interface, a "thin" vertical zone where the temperature gradient is the steepest, is important in assessing the potential implications of thermally-induced stress problems in liquid-metal cooled reactors. Thermal stratification can likewise occur in confined volumes containing ordinary fluids (Pr$$geq$$1), where there is an input of thermal convective energy. In the prominent case of liquid metal reactors, there have been many studies on quantifying the rate-of-rise of a defined stratification interface, in terms of one or more of the following dimensionless groups, mainly: Richardson (Ri), Reynolds (Re), Grashof (Gr), Rayleigh (Ra) and/or Froude (Fr) numbers. Stratification is also a transient process in the volume in question. In the present work the anthors presents a derivation based on order-of-magnitude analysis (OMA), including an sensible energy balance, that produces a new representation more consistent than p

Journal Articles

Analysis for performance of passive heat removal by a water cooling panel from a high-temperature gas-cooled reactor

Takada, Shoji; *; Inagaki, Yoshiyuki; Sudo, Yukio

Nippon Kikai Gakkai Rombunshu, B, 65(635), p.303 - 311, 1999/07

no abstracts in English

Journal Articles

Experimental study of heat transfer in an open thermosyphon

Imai, Etsuya*; Shiina, Yasuaki; Hishida, Makoto*

Nippon Kikai Gakkai Rombunshu, B, 65(634), p.227 - 233, 1999/06

no abstracts in English

Journal Articles

Experimental and numerical studies on performance of passive decay heat removal by a water cooling panel from a high-temperature gas-cooled reactor

Takada, Shoji; Suzuki, Kunihiro; Inagaki, Yoshiyuki; Sudo, Yukio

Journal of Nuclear Science and Technology, 36(5), p.413 - 423, 1999/05

 Times Cited Count:2 Percentile:76.23(Nuclear Science & Technology)

no abstracts in English

Journal Articles

Performance of passive heat removal by an air cooling panel from a high-temperature gas-cooled reactor

Takada, Shoji; Suzuki, Kunihiro; Inagaki, Yoshiyuki; Sudo, Yukio

Nippon Kikai Gakkai Rombunshu, B, 65(633), p.248 - 254, 1999/05

no abstracts in English

96 (Records 1-20 displayed on this page)