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Hamdani, A.; Soma, Shu; Abe, Satoshi; Shibamoto, Yasuteru
Progress in Nuclear Energy, 185, p.105771_1 - 105771_13, 2025/07
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)Takeda, Takeshi
JAEA-Data/Code 2023-007, 72 Pages, 2023/07
JAEA-Data-Code-2023-007.pdf:3.24MB
An experiment denoted as IB-HL-01 was conducted on November 19, 2009 using the Large Scale Test Facility (LSTF) in the Rig of Safety Assessment-V (ROSA-V) Program. The ROSA/LSTF experiment IB-HL-01 simulated a 17% hot leg intermediate break loss-of-coolant accident due to a double-ended guillotine break of pressurizer surge line in a pressurized water reactor (PWR). The break was simulated by a long nozzle upwardly mounted flush with a hot leg inner surface. The test assumptions included total failure of both high pressure injection system of emergency core cooling system (ECCS) and auxiliary feedwater system. In the experiment, relatively large size of break led to a fast transient of phenomena. The primary pressure steeply dropped after the break, and became lower than steam generator (SG) secondary-side pressure. Break flow turned from single-phase flow to two-phase flow soon after the break. Core uncovery started simultaneously with liquid level drop in downflow-side of crossover leg before loop seal clearing (LSC). The LSC was induced in both loops by steam condensation on accumulator (ACC) coolant of ECCS injected into cold legs. The whole core was quenched owing to the rapid recovery in the core liquid level after the LSC. Peak cladding temperature of simulated fuel rods was detected almost concurrently with the LSC. During the ACC coolant injection, liquid levels recovered in the hot legs and SG inlet plena because of liquid entrainment from the hot leg into the SG inlet plenum by high-velocity steam flow. 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 IB-HL-01.
Watanabe, So; Ogi, Hiromichi*; Shibata, Atsuhiro; Nomura, Kazunori
International Journal of Nuclear and Quantum Engineering (Internet), 13(4), p.169 - 174, 2019/04
As a part of STRAD project conducted by JAEA, condensation of radioactive liquid waste containing various chemical compounds using reverse osmosis (RO) membrane filter was examined for efficient and safety treatment of the liquid wastes accumulated inside hot laboratories. NH
ion in the feed solution was successfully concentrated, and NH ion involved in the effluents became lower than target value; 100 ppm. Solidification of simulated aqueous and organic liquid wastes was also tested. Those liquids were successfully solidified by adding cement or coagulants. Nevertheless, optimization in materials for confinement of chemicals is required for long time storage of the final solidified wastes.
Takeda, Takeshi; Otsu, Iwao
Mechanical Engineering Journal (Internet), 5(4), p.18-00077_1 - 18-00077_14, 2018/08
Sakaki, Toshihiro*; Komatsu, Mitsuru*; Takeuchi, Ryuji
Vadose Zone Journal (Internet), 15(8), 7 Pages, 2016/08
Times Cited Count：2 Percentile：8.62(Environmental Sciences)In the Groundwater Recovery Experiment in Tunnel which is underway at the MIU, water saturation is one of the key parameters to monitor in order to understand the recovery processes. In this study, the process of saturation increase in the quasi-saturated zone where trapped discrete air phase affects saturation due to further increase in water pressure was studied. The relationship between the positive water pressure and saturation was measured by laboratory test using sands with trapped air bubbles and the effect of compression of bubbles on the changes in saturation was analyzed. The experimental results showed that the water pressure-saturation curves closely followed the relationship estimated solely based on the air bubble compression using Boyle's law. Based on this observation, a mathematical model was established to define the water pressure - saturation curve for the region where the water pressure is positive.
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.
Takeda, Takeshi; Otsu, Iwao
Journal of Energy and Power Sources, 2(7), p.274 - 290, 2015/07
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.
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 
2 mm 2 mm, 3 mm 3 mm 3 mm and 2 mm 3 mm 2 mm (depth top width 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 2 mm 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.
Suzuki, Mitsuhiro; Takeda, Takeshi; Asaka, Hideaki; Nakamura, Hideo
Journal of Nuclear Science and Technology, 43(1), p.55 - 64, 2006/01
Times Cited Count：10 Percentile：55.89(Nuclear Science & Technology)Effects of non-condensable gas from the accumulator tanks on secondary depressurization, as one of accident management (AM) measures in case of high pressure injection system failure, are studied at the ROSA-V/LSTF experiments simulating a ten instrument-tube break LOCA at the PWR vessel bottom. In an experiment with no gas inflow, the secondary depressurization at -55 K/h initiated by SI signal with 10 minutes delay succeeded in the LPI actuation. On the other hand, the gas inflow in another experiment degraded the primary depressurization and resulted in core uncovery before the LPI start. The third experiment with rapid secondary depressurization and continuous auxiliary feedwater supply, however, showed a possibility of long-term core cooling by the LPI actuation. RELAP5/MOD3 code analyses well predicted these experiment results and clarified that condensation heat transfer was largely degraded by the gas in the U-tubes. In addition, a primary pressure - coolant mass map was found to be useful for indication of key plant parameters of AM measures.
Suzuki, Mitsuhiro; Takeda, Takeshi; Asaka, Hideaki; Nakamura, Hideo
JAERI-Research 2005-014, 170 Pages, 2005/06
JAERI-Research-2005-014.pdf:7.64MB
A small break LOCA (SBLOCA) experiment was conducted at the LSTF of ROSA-V program to study effects of accident management (AM) on core cooling, which is important in case of high pressure injection (HPI) system failure during an SBLOCA at a PWR. The experiment, SB-PV-03, simulated ten instrument-tube break LOCA at the PWR vessel bottom equivalent to 0.2% cold leg break, total HPI failure, non-condensable gas inflow from accumulator injection system (AIS) and AM actions on secondary depressurization at -55 K/h and auxiliary feedwater (AFW) supply for 30 minutes. It was clarified that the AM actions were effective on primary depressurization until AIS injection end at 1.6 MPa, but thereafter became less effective by the gas inflow, resulting in low pressure injection (LPI) delay and whole core heatup under continuous water discharge at the break. The report describes these phenomena including core heatup related with primary coolant mass and AM actions, primary-to-secondary heat transfer analysis and estimation of gas in the primary loops.
Kansai Research Establishment
JAERI-Review 2004-030, 150 Pages, 2005/02
JAERI-Review-2004-030.pdf:11.32MB
This report is the fifth issue of the annual report of Kansai Research Establishment, Japan Atomic Energy Research Institute. It covers status reports of R&D and results of experiments conducted at the Advanced Photon Research Center and the Synchrotron Radiation Research Center during the period from April 1, 2003 to March 31, 2004.
Suzuki, Mitsuhiro; Takeda, Takeshi; Asaka, Hideaki; Nakamura, Hideo
Proceedings of 6th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, Operations and Safety (NUTHOS-6) (CD-ROM), 14 Pages, 2004/10
no abstracts in English
Urata, Kazuhiro*; Shinohara, Koji; Suzuki, Masanobu*; Kamata, Isao*
JAERI-Data/Code 2004-007, 45 Pages, 2004/03
JAERI-Data-Code-2004-007.pdf:5.63MB
As the toroidal magnetic field generated by discrete TF coils involves magnetic field ripple, the fast ion loss is induced to damage vacuum vessel in tokamaks. An idea of ripple compensation using ferromagnetic is proposed. Since low activation ferritic steel have low activation and thermal conduction properties, the ferritic steel is planned to install in tokamak reactors. Installation of ferritic steel plates with toroidal symmetry is effective to compensate ripple, however in the actual devices it is difficult for interference with other components. Besides the first wall shapes are often asymmetric. So it is better to treat toroidal asymmetry to evaluate the ripple induced loss in the actual devices. For the purpose, magnetic field calculation code considering ferritic steel; FEMAG(FErrite generating MAGnetic field)has been speeded up. On the basis of this magnetic field data, OFMC (Orbit Following Monte Carlo) has been upgraded to treat toroidal asymmetry. The use of FEMAG/OFMC, applications to the JFT-2M experiments, and the national centralized tokamak facility are reported.
Kansai Research Establishment
JAERI-Review 2003-042, 140 Pages, 2004/02
JAERI-Review-2003-042.pdf:11.75MB
This report is the fourth issue of the annual report of Kansai Research Establishment, Japan Atomic Energy Research Institute. It covers status reports of R&D and results of experiments conducted at the Advanced Photon Research Center and the Synchrotron Radiation Research Center during the period from April 1, 2002 to March 31, 2003.
Advanced Photon Research Center
JAERI-Conf 2003-008, 283 Pages, 2003/07
JAERI-Conf-2003-008.pdf:35.08MB
no abstracts in English
Research Evaluation Committee
JAERI-Review 2003-017, 34 Pages, 2003/06
JAERI-Review-2003-017.pdf:2.83MB
The Research Evaluation Committee, which consisted of 13 members from outside of the Japan Atomic Energy Research Institute (JAERI), set up an Ad Hoc Review Committee on Advanced Photon and Synchrotron Radiation Research in accordance with the "Fundamental Guideline for the Evaluation of Research and Development (R&D) at JAERI" and its subsidiary regulations in order to evaluate the adequacy of the R&D accomplishments achieved for five years until Fiscal Year 2001 at Advanced Photon Research Center and at Synchrotron Radiation Research Center in Kansai Research Establishment of JAERI. The Ad Hoc Review Committee consisted of ten specialists from outside of JAERI.The Ad Hoc Review Committee conducted its activities from November 2002 to January 2003. The evaluation was performed on the basis of the materials submitted in advance and of the oral presentations made at the Ad Hoc Review Committee meeting which was held on December 26, 2002, in line with the items, viewpoints, and criteria for the evaluation specified by the Research Evaluation Committee.The result of the evaluation by the Ad Hoc Review Committee was submitted to the Research Evaluation Committee, and was judged to be appropriate at its meeting held on March 7, 2003.This report describes the result of the evaluation by the Ad Hoc Review Committee on Advanced Photon and Synchrotron Radiation Research.
Yonomoto, Taisuke; Okubo, Tsutomu; Iwamura, Takamichi; Ishida, Toshihisa
Proceedings of 11th International Conference on Nuclear Engineering (ICONE-11) (CD-ROM), 8 Pages, 2003/04
An innovative concept of the pressure suppression system having functions of water injection and non-condensable gas confinement is developed for the next generation light water reactors (LWRs). The use of the system is advantageous for the mitigation of effects of the loss-of-coolant accidents (LOCAs) in (1) keeping the containment pressure as low as for the conventional LWRs, (2) injecting water to the containment for cooling the reactor pressure vessel (RPV) and/or flooding a break, and (3) confining the non-condensable gas in the drywell. The gas confinement function makes the system considerably suitable for reactor designs with passive cooling systems utilizing heat exchangers, such as the steam generator (SG) secondary side cooling system for an integral reactor, and the passive containment cooling system (PCCS), because it avoids adverse effects of non-condensable gas on the heat transfer performance during LOCAs. The usefulness of the developed concept is confirmed in the RELAP5/MOD3 code calculation.
Kansai Research Establishment
JAERI-Review 2002-042, 135 Pages, 2003/02
JAERI-Review-2002-042.pdf:11.88MB
This report is the third issue of the annual report of Kansai Research Establishment, Japan Atomic Energy Research Institute. It covers status reports of R&D and results of experiments conducted at the Advanced Photon Research Center and the Synchrotron Radiation Research Center during the period from April 1, 2001 to March 31, 2002.
Suzuki, Mitsuhiro
JAERI-Tech 2002-071, 171 Pages, 2002/10
JAERI-Tech-2002-071.pdf:11.26MB
no abstracts in English
