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Journal Articles

Evaluation of gas entrainment flow rate by free surface vortex

Torikawa, Tomoaki*; Odaira, Naoya*; Ito, Daisuke*; Ito, Kei*; Saito, Yasushi*; Matsushita, Kentaro; Ezure, Toshiki; Tanaka, Masaaki

Konsoryu, 36(1), p.63 - 69, 2022/03

On free surface of a sodium cooled fast reactor, gas entrainment can be caused by free surface vortices, which may result in disturbance in core power. It is important to develop an evaluation model to predict accurately entrained gas flow rate. In this study, entrained gas flow rate a simple gas entrainment experiment is conducted with focusing on effect of pressure difference between upper and lower tanks. Pressure difference between upper and lower tanks are controlled by changing gas pressure in lower tank. As a result, it is confirmed that the entrained gas flow rate increases with increasing pressure difference between upper and lower tanks. By visualization of swirling annular flow in suction pipe, it is also observed that pressure drop in suction pipe increases with increase in entrained gas flow rate, which implies that entrained gas flow rate can be predicted by evaluation model based on pressure drop in swirling annular flow region.

Journal Articles

Flow regime and void fraction predictions in vertical rod bundle flow channels

Han, X.*; Shen, X.*; Yamamoto, Toshihiro*; Nakajima, Ken*; Sun, Haomin; Hibiki, Takashi*

International Journal of Heat and Mass Transfer, 178, p.121637_1 - 121637_24, 2021/10

 Times Cited Count:1 Percentile:38.69(Thermodynamics)

JAEA Reports

Data report of ROSA/LSTF experiment SB-PV-09; 1.9% pressure vessel top small break LOCA with SG depressurization and gas inflow

Takeda, Takeshi

JAEA-Data/Code 2021-006, 61 Pages, 2021/04

JAEA-Data-Code-2021-006.pdf:2.78MB

An experiment denoted as SB-PV-09 was conducted on November 17, 2005 using the Large Scale Test Facility (LSTF) in the Rig of Safety Assessment-V (ROSA-V) Program. The ROSA/LSTF experiment SB-PV-09 simulated a 1.9% pressure vessel top small-break loss-of-coolant accident in a pressurized water reactor (PWR). The test assumptions included total failure of high pressure injection system and non-condensable gas (nitrogen gas) inflow to the primary system from accumulator (ACC) tanks of emergency core cooling system (ECCS). In the experiment, liquid level in the upper-head was found to control break flow rate. When maximum core exit temperature reached 623 K, steam generator (SG) secondary-side depressurization was initiated by fully opening the relief valves in both SGs as an accident management (AM) action. The AM action, however, was ineffective on the primary depressurization until the SG secondary-side pressure decreased to the primary pressure. Meanwhile, the core power was automatically reduced when maximum cladding surface temperature of simulated fuel rods exceeded the pre-determined value of 958 K to protect the LSTF core due to late and slow response of core exit temperature. After the automatic core power reduction, loop seal clearing (LSC) was induced in both loops by steam condensation on the ACC coolant injected into cold legs. The whole core was quenched because of core recovery after the LSC. After the ACC tanks started to discharge nitrogen gas, the pressure difference between the primary and SG secondary sides became larger. After the continuous core cooling was confirmed through the actuation of low pressure injection system of ECCS, the experiment was terminated. This report summarizes the test procedures, conditions, and major observations in the ROSA/LSTF experiment SB-PV-09.

Journal Articles

Experimental study on local interfacial parameters in upward air-water bubbly flow in a vertical 6$$times$$6 rod bundle

Han, X.*; Shen, X.*; Yamamoto, Toshihiro*; Nakajima, Ken*; Sun, Haomin; Hibiki, Takashi*

International Journal of Heat and Mass Transfer, 144, p.118696_1 - 118696_19, 2019/12

 Times Cited Count:4 Percentile:37.36(Thermodynamics)

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

Local gas-liquid two-phase flow characteristics in rod bundle geometry

Xiao, Y.*; Shen, X.*; Miwa, Shuichiro*; Sun, Haomin; Hibiki, Takashi*

Konsoryu Shimpojiumu 2018 Koen Rombunshu (Internet), 2 Pages, 2018/08

In order to develop constitutive equations of two-fluid model in rod bundle flow channels, experiments of adiabatic air-water upward two-phase flow in 6$$times$$6 rod bundle flow channel were performed. Local flow parameters such as void fraction, interfacial area concentration (IAC) and so on were measured by a double-sensor optical probe. The area-averaged void fraction and IAC data were compared with the predictions from a drift-flux model and an IAC correlation.

Journal Articles

ROSA/LSTF test on nitrogen gas behavior during reflux condensation in PWR and RELAP5 code analyses

Takeda, Takeshi; Otsu, Iwao

Mechanical Engineering Journal (Internet), 5(4), p.18-00077_1 - 18-00077_14, 2018/08

Journal Articles

Off-gas processing system operations for mercury target vessel replacement at J-PARC

Kai, Tetsuya; Uchida, Toshitsugu; Kinoshita, Hidetaka; Seki, Masakazu; Oi, Motoki; Wakui, Takashi; Haga, Katsuhiro; Kasugai, Yoshimi; Takada, Hiroshi

Journal of Physics; Conference Series, 1021(1), p.012042_1 - 012042_4, 2018/06

 Times Cited Count:0 Percentile:0.11

Journal Articles

Conceptual design of the iodine-sulfur process flowsheet with more than 50% thermal efficiency for hydrogen production

Kasahara, Seiji; Imai, Yoshiyuki; Suzuki, Koichi*; Iwatsuki, Jin; Terada, Atsuhiko; Yan, X.

Nuclear Engineering and Design, 329, p.213 - 222, 2018/04

 Times Cited Count:15 Percentile:89.72(Nuclear Science & Technology)

A conceptual design of a practical large scale plant of the thermochemical water splitting iodine-sulfur (IS) process flowsheet was carried out as a heat application of JAEA's commercial high temperature gas cooled reactor GTHTR300C plant design. Innovative techniques proposed by JAEA were applied for improvement of hydrogen production thermal efficiency; depressurized flash concentration H$$_{2}$$SO$$_{4}$$ using waste heat from Bunsen reaction, prevention of H$$_{2}$$SO$$_{4}$$ vaporization from a distillation column by introduction of H$$_{2}$$SO$$_{4}$$ solution from a flash bottom, and I$$_{2}$$ condensation heat recovery in an HI distillation column. Hydrogen of about 31,900 Nm$$^{3}$$/h would be produced by 170 MW heat from the GTHTR300C. A thermal efficiency of 50.2% would be achievable with incorporation of the innovative techniques and high performance HI concentration and decomposition components and heat exchangers expected in future R&D.

JAEA Reports

Data report of ROSA/LSTF experiment SB-PV-07; 1% Pressure vessel top break LOCA with accident management actions and gas inflow

Takeda, Takeshi

JAEA-Data/Code 2018-003, 60 Pages, 2018/03

JAEA-Data-Code-2018-003.pdf:3.68MB

Experiment SB-PV-07 was conducted on June 9, 2005 using LSTF. Experiment simulated 1% pressure vessel top small-break LOCA in PWR under total failure of HPI system and nitrogen gas inflow to primary system from ACC tanks. Liquid level in upper-head was found to control break flow rate. Coolant was started to manually inject from HPI system into cold legs as first accident management (AM) action when maximum core exit temperature reached 623 K. Fuel rod surface temperature largely increased because of late and slow response of core exit temperature. SG secondary-side depressurization was initiated by fully opening relief valves as second AM action when primary pressure decreased to 4 MPa. However, second AM action was not effective on primary depressurization until SG secondary-side pressure decreased to primary pressure. Pressure difference became larger between primary and SG secondary sides after ACC tanks started to discharge nitrogen gas.

Journal Articles

ROSA/LSTF tests and posttest analyses by RELAP5 code for accident management measures during PWR station blackout transient with loss of primary coolant and gas inflow

Takeda, Takeshi; Otsu, Iwao

Science and Technology of Nuclear Installations, 2018, p.7635878_1 - 7635878_19, 2018/00

 Times Cited Count:2 Percentile:27.95(Nuclear Science & Technology)

Journal Articles

Multi-dimensional gas-liquid two-phase flow in vertical large-diameter channels

Shen, X.*; Schlegel, J. P.*; Hibiki, Takashi*; Nakamura, Hideo

Proceedings of 2017 Japan-US Seminar on Two-Phase Flow Dynamics (JUS 2017), 6 Pages, 2017/06

Journal Articles

Numerical simulations of gas-liquid-particle three-phase flows using a hybrid method

Guo, L.*; Morita, Koji*; Tobita, Yoshiharu

Journal of Nuclear Science and Technology, 53(2), p.271 - 280, 2016/02

 Times Cited Count:6 Percentile:56.63(Nuclear Science & Technology)

JAEA Reports

Data report of ROSA/LSTF experiment SB-HL-12; 1% Hot leg break LOCA with SG depressurization and gas inflow

Takeda, Takeshi

JAEA-Data/Code 2015-022, 58 Pages, 2016/01

JAEA-Data-Code-2015-022.pdf:3.31MB

The SB-HL-12 test simulated PWR 1% hot leg SBLOCA under assumptions of total failure of HPI system and non-condensable gas (nitrogen gas) inflow. SG depressurization by fully opening relief valves in both SGs as AM action was initiated immediately after maximum fuel rod surface temperature reached 600 K. After AM action due to first core uncovery by core boil-off, the primary pressure decreased, causing core mixture level swell. The fuel rod surface temperature then increased up to 635 K. Second core uncovery by core boil-off took place before LSC induced by steam condensation on ACC coolant injected into cold legs. The core liquid level recovered rapidly after LSC. The fuel rod surface temperature then increased up to 696 K. The pressure difference became larger between the primary and SG secondary sides after nitrogen gas inflow. Third core uncovery by core boil-off occurred during reflux condensation. The maximum fuel rod surface temperature exceeded 908 K.

JAEA Reports

Study on numerical simulation of bubble and dissolved gas behavior in liquid metal flow

Ito, Kei; Tanaka, Masaaki; Ohno, Shuji; Ohshima, Hiroyuki

JAEA-Research 2014-023, 34 Pages, 2014/11

JAEA-Research-2014-023.pdf:4.48MB

In a sodium-cooled fast reactor, inert gas (bubbles or dissolved gas) exists in the primary coolant system. Such inert gas may cause disturbance in reactivity and/or degradation of IHX performance, and therefore, the inert gas behaviors have to be investigated to ensure the stable operation of a fast reactor. The authors have developed a plant dynamics code SYRENA to simulate the concentration distributions of the dissolved gas and the bubbles in a fast reactor. In this study, the models in SYRENA code are improved to achieve accurate simulations. Moreover, new models are introduced to simulate the various bubble behaviors in liquid metal flows. To validate the improved models and the newly developed models, the inert gas behaviors in the large-scale sodium-cooled reactor are simulated. As a result, it is confirmed that the complicated bubble dynamics in each component can be simulated appropriately by SYRENA code.

JAEA Reports

Data report of ROSA/LSTF experiment SB-CL-32; 1% cold leg break LOCA with SG depressurization and no gas inflow

Takeda, Takeshi

JAEA-Data/Code 2014-021, 59 Pages, 2014/11

JAEA-Data-Code-2014-021.pdf:5.16MB

Experiment SB-CL-32 was conducted on May 28, 1996 using the LSTF. The experiment SB-CL-32 simulated 1% cold leg small-break LOCA in PWR under assumptions of total failure of HPI system and no inflow of non-condensable gas from ACC tanks. Secondary-side depressurization of both SGs as AM action to achieve the depressurization rate of 200 K/h in the primary system was initiated 10 min after break. Core uncovery started with liquid level drop in crossover leg downflow-side. The core liquid level recovered rapidly after first LSC. The surface temperature of simulated fuel rod then increased up to 669 K. Core uncovery took place before second LSC induced by steam condensation on ACC coolant. The core liquid level recovered rapidly after second LSC. The maximum fuel rod surface temperature was 772 K. The continuous core cooling was confirmed because of coolant injection by LPI system. This report summarizes the test procedures, conditions and major observation.

Journal Articles

Simulation of tritium spreading in controlled areas after a tritium release

Cristescu, I. R.*; Travis, J.*; Iwai, Yasunori; Kobayashi, Kazuhiro; Murdoch, D.*

Fusion Science and Technology, 48(1), p.464 - 467, 2005/07

 Times Cited Count:1 Percentile:10.86(Nuclear Science & Technology)

A model to simulate tritium behaviour after a release into a confined ventilated volume has been developed. The model assumes that for the investigated cases, tritium behaviour can be characterized by solving the dynamic equations of motion (the compressible Navier-Stokes equations) coupled with the classical k-$$varepsilon$$ turbulence model to simulate the ventilation in the room and mass diffusion for tritium spreading. The GASFLOW-II fluid dynamics field code, developed through a Los Alamos National Laboratory (LANL) - Forschungszentrum Karlsruhe co-operation, was used as the computational tool to solve the equations that describe the processes. The numerical results have been validated with experimental data collected on the experimental facility (Caisson) at the Tritium Process Laboratory (TPL) Japan. Additionally an investigation of the influence of the obstacles to the tritium distribution inside the Caisson is presented.

Journal Articles

Helium-air counter flow in rectangular channels

Fumizawa, Motoo; Tanaka, Gaku*; Zhao, H.*; Hishida, Makoto*; Shiina, Yasuaki

Nihon Genshiryoku Gakkai Wabun Rombunshi, 3(4), p.313 - 322, 2004/12

This paper deals with a computer simulation of a helium-air counter flow in a rectangular channel. The inclination angle is varied from 0$$^{circ}$$(horizontal) to 90$$^{circ}$$(vertical). Velocity profiles and concentration profiles are calculated with a computer program VSOP sub-module. Following main features of the counter flow are discussed. (1) Time required to establish a quasi-steady state counter flow. (2) The relationship between the inclination angle and the flow patterns of the counter flow (3) The developing process of velocity profiles and concentration profiles (4) The relationship between the inclination angle of the channel and the velocity profiles of upwards flow and the downwards flow (5) The relationship between the concentration profile and the inclination angle (6) The relationship between the net in-flow rate and the inclination angle We compared the computed velocity profile and the net in-flow rate with experimental data. A good agreement is obtained between the calculation and the experiment.

JAEA Reports

Structural integrity assessment of helium component during safety demonstration test using HTTR, 1 (Contract Research)

Sakaba, Nariaki; Nakagawa, Shigeaki; Furusawa, Takayuki; Tachibana, Yukio

JAERI-Tech 2004-045, 67 Pages, 2004/04

JAERI-Tech-2004-045.pdf:4.74MB

Safety demonstration tests using the HTTR are now underway in order to verify the inherent safety features and to improve the safety design and evaluation technologies for HTGRs, as well as to contribute to research and development for the VHTR, which is one of the Generation IV reactors. In the safety demonstration tests, the coolant flow reduction test by tripping one or two out of three gas circulators is being performed between FY2002 and FY 2005 and by tripping all the three gas circulators will be conducted after FY2006. This paper describes the structural integrity assessment of the primary pressurised water cooler after one and two gas circulators run down. Also, the possibility of natural convection in the primary coolant after all the three gas circulator stopped was evaluated by the operation data of the reactor-scram test performed during the rise-to-power tests.

Journal Articles

Two-dimensional particle simulation of the flow control in SOL and divertor plasmas

Takizuka, Tomonori; Hosokawa, Masanari*; Shimizu, Katsuhiro

Journal of Nuclear Materials, 313-316(1-3), p.1331 - 1334, 2003/03

 Times Cited Count:13 Percentile:65.83(Materials Science, Multidisciplinary)

In course of tokamak fusion research, particle and heat control is one of the most crucial issues. Helium ash exhaust and impurity retention in the divertor region owe to the plasma flow towards divertor plate. The localization of heat load on the plate depends on the flow pattern. Accordingly, particle and heat control can be achieved by the proper control of the flow in SOL and divertor plasmas. In this paper, the flow control is studied with two-dimensional particle simulations by PARASOL (PARticle Advanced simulation for SOL and divertor plasmas) code. Magnetic field configuration with separatrix like a tokamak divertor configuration is given. Hot particle source is put in the core plasma. Recycling cold particle source is located near the divertor plate. Particle source of gas puff in the SOL plasma is given for the flow control. Divertor biasing is available by changing the electrostatic potential on the plates. Effects of gas puff and biasing on the flow are studied. Controllability is evaluated from simulation results.

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