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Kudo, Hideyuki*; Ota, Junki*; Sugiyama, Kenichiro*; Narabayashi, Tadashi*; Ohshima, Hiroyuki; Kurihara, Akikazu
Hozengaku, 11(4), p.90 - 97, 2013/01
When a heat transfer tube wall in a steam generator of a sodium-cooled fast reactor fails, high-pressure steam leaks into low-pressure liquid sodium side. Then the high-temperature and highly corrosive reaction jet causes secondary failures of neighboring heat transfer tubes. The objective of the present study is to develop an experimental method to obtain data that is necessary to validate and improve a safety evaluation code on the sodium-water reaction. In the present paper, a method of sodium-droplet erosion test at the position of the local structure in underexpanded inert-gas impinging jets using the visualization method, which was developed in our previous study, was reported. The erosion phenomena observed in the sodium-droplet entrained region, where intensive erosion is expected, were found to be discussed using the existing knowledge of liquid droplet impingement (LDI) obtained in water experiments.
Kudo, Hideyuki*; Sugiyama, Kenichiro*; Narabayashi, Tadashi*; Ohshima, Hiroyuki; Kurihara, Akikazu
Journal of Nuclear Science and Technology, 50(1), p.72 - 79, 2013/01
Times Cited Count：7 Percentile：46.40(Nuclear Science & Technology)In order to accurately model sodium-water reaction jets in steam generators of fast breeder reactors, knowledge of size distributions or mean diameters of liquid sodium droplets entrained into the reaction jets are prerequisite. In the present study, argon-gas jet behaviors, without chemical reaction, injected into liquid sodium were successfully visualized using an endoscope and a glass tube, and the size distributions and mean diameters of liquid sodium droplets entrained into the gas jet were also obtained in the bubbling regime.
Kudo, Hideyuki*; Zhao, D.*; Sugiyama, Kenichiro*; Narabayashi, Tadashi*; Ohshima, Hiroyuki; Kurihara, Akikazu
Journal of Nuclear Science and Technology, 49(12), p.1175 - 1185, 2012/12
Times Cited Count：3 Percentile：23.63(Nuclear Science & Technology)Little work on the void fraction behaviors along structural materials with poor-wettability for liquid metals has been performed. In the present study, void fraction behaviors around a single cylinder with non-wetting surface condition were quantitatively discussed by using a gas jet-cylinder system where the impinging jet flow, the boundary layer flow, the separation flow, and the wake flow appear. The characteristics in each flow field as well as the relationship between flow fields, which have not been quantitatively discussed so far, are obtained. The local void fraction around a single cylinder with wetting condition or non-wetting condition was measured by using resistivity probes.
Kudo, Hideyuki*; Sugiyama, Kenichiro*; Narabayashi, Tadashi*; Ohshima, Hiroyuki; Kurihara, Akikazu
Proceedings of 20th International Conference on Nuclear Engineering and the ASME 2012 Power Conference (ICONE-20 & POWER 2012) (DVD-ROM), 6 Pages, 2012/07
For the sodium-water reaction accident, it is important to grasp the structure of gas jets submerged in liquid sodium and associated droplet size. In this study, we successfully obtained visualized images of inert gas jets injected into liquid sodium. Formation processes of liquid sodium droplets entrained into the gas jets and drop-size distributions are discussed.
Yoshida, Atsuro*; Higashi, Yuma*; Narabayashi, Tadashi*; Khoo Chong Weng, W.*; Arae, Kunihiko*; Tsuji, Masashi*; Ohshima, Hiroyuki; Kurihara, Akikazu
Proceedings of 19th International Conference on Nuclear Engineering (ICONE-19) (CD-ROM), 8 Pages, 2011/10
One of the design basis accidents in sodium-cooled fast reactor is sodium-water reaction at steam generator (SG). In case of a defect occurred on a heat transfer tube, the high-pressure water/vapor will spout into the low-pressure sodium surrounding outside the tube. As sodium is ordinarily quite reactive with water, this will initiate sodium-water reactions accompanied by high chemical heat generation. The liquid droplet in the reaction steam outflow would impinge on neighboring tubes to cause erosion, while the chemical reaction will cause corrosion, eventually may lead to secondary tube failure. Focusing on the erosion part, this study is to evaluate the liquid droplet impingement erosion (LDIE) rate on neighboring tubes caused by SG heat transfer tube rupture. In this paper, as a basic study, the pressure and temperature distribution of high -pressure two-phase free jet into the air is measured.
Ohshima, Hiroyuki; Yamaguchi, Akira*; Narabayashi, Tadashi*; Deguchi, Yoshihiro*
Dai-16-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu, p.1 - 2, 2011/06
When a heat transfer tube is failed in a steam generator (SG) of a sodium-cooled fast reactor (SFR), pressurized water and/or water vapor leaks into liquid sodium surrounding the tube and forms a reacting jet with high temperature. This reacting jet might cause the secondary failure of adjacent heat transfer tubes due to wastage or over-heating tube rapture resulting in undesirable development of the accident. Therefore, the sodium-water reaction phenomenon (SWR) is one of most important issues for the design and safety assessment of SFRs. This paper describes the research and development plan of a new multi-physics numerical simulation system which is based on mechanistic and theoretical modeling of the SWR rather than empirical modeling and can contribute to detailed and quantitative evaluations of the SWR in any types of SGs including commercial SFRs.
Nishizaki, Masanori*; Tsuruoka, Hokuto*; Sugiyama, Kenichiro*; Narabayashi, Tadashi*; Ohshima, Hiroyuki
Proceedings of 17th International Conference on Nuclear Engineering (ICONE-17) (CD-ROM), 6 Pages, 2009/07
The secondary tube failure may occur due to overheating by sodium-water reaction in LMFBR steam generator. It is very important to understand the void fraction distribution in sodium pool to evaluate the overheating tube rupture. In the present study, the Ar jet of 17.3 m/s to 129.8 m/s was injected from nozzle of 3.5 mm diameter in sodium pool with 443 K and 293 K. The authors measured the void fraction without chemical reaction along the jet-center axis. As the result, the void fraction increased when the distance from the nozzle decreased. The void fraction did not change when the distance from the nozzle was blow or equal to about 1.0 mm. The void fraction in sodium was lower the that in water, it is suggested that this trend reflects the fact the surface tension of sodium is higher than that of water.
Tsuruoka, Hokuto*; Tamura, Takeshi*; Sugiyama, Kenichiro*; Narabayashi, Tadashi*; Ohshima, Hiroyuki
Proceedings of 16th International Conference on Nuclear Engineering (ICONE-16) (CD-ROM), 6 Pages, 2008/05
The occurrence of secondary heat transfer tube failure due to overheating by sodium-water reaction in LMFBR steam generators has been concerned from the viewpoint of public acceptance. To evaluate the phenomena, a sophisticated computer code SERAPHIM has been developed by JAEA. For the purpose of obtaining fundamental data for the validation of the code, a sodium experiment was carried out, where the void fraction around a single rod set in a sodium pool without sodium-water reaction was measured. The void fraction was observed to somewhat increase with increasing the gas jet velocity. The increase rate was clearly smaller compared with that in the water experiment. The void fraction also showed more monotonous distribution from the stagnation point to the rear point than that in water pool. These results reflect the difference of surface tension between water and sodium. It is concluded that the entrainment of ambient sodium is easily caused and this leads monotonous distribution of void fraction in the sodium pool.
Tamura, Takeshi*; Soga, Kazuo*; Sugiyama, Kenichiro*; Narabayashi, Tadashi*; Ohshima, Hiroyuki; Suda, Kazunori
Proceedings of 15th International Conference on Nuclear Engineering (ICONE-15) (CD-ROM), 6 Pages, 2007/04
To evaluate the phenomena for the secondary heat transfer tube failure due to overheating by sodium-water reaction in steam generator of liquid sodium cooled fast breeder reactor, a sophisticated computer code SERAPHIM has been developed by JAEA. As the first step to verify the adequacy of SERAPHIM code, a visualization experiment of Ar gas jet impinging a single rod with 20mm in diameter immersed in a water pool was performed in our previous study. In this paper, we measured the void fraction around a single rod in the water pool as a basic experiment using the apparatus capable of doing sodium pool experiment to investigate the flow pattern and the water entrainment around a single rod. The result of the void fraction reflected the result of the heat transfer experiment that had been reported before was obtained, and a certain prospect of the measurement of the void fraction in sodium pool was obtained.
Soga, Kazuo*; Niikura, Hideto*; Sugiyama, Kenichiro*; Narabayashi, Tadashi*; Ohshima, Hiroyuki; Suda, Kazunori
Proceedings of 14th International Conference on Nuclear Engineering (ICONE-14) (CD-ROM), 6 Pages, 2006/07
A series of experiments that investigate the entrainment process of ambient liquid toward jet interior are carried out by using a laser-sheet visualization and a void meter in water pool in the present work. It was observed that the entrainment of water into Ar gas jet is constantly caused in two regions just above the nozzle and just below the single rod. In the region just above the nozzle, negative pressure causes the entrainment of water. In the region below the rod, the entrainment of water is caused because the preceding Ar gas jet is caught up by the succeeding gas jet. The basic behavior of Ar gas jet causing the entrainment of water was confirmed to be almost same over the Reynolds number range of Ar gas jet, 2.17
103 to 2.17104, in the present study.
Kato, Keisuke*; Yoshida, Atsuro*; Narabayashi, Tadashi*; Ohshima, Hiroyuki; Kurihara, Akikazu
no journal, ,
In order to quantitatively evaluate the effect of liquid droplet impacting erosion on tube wastage in case of sodium-water reaction in steam generator of sodium-cooled fast reactor, the authors have measured the behavior of water/vapor two-phase jet in water as a fundamental research using the simulated specimen of tube in steam genarator.
Kato, Keisuke*; Narabayashi, Tadashi*; Yoshida, Atsuro*; Arae, Kunihiko*; Ohshima, Hiroyuki; Kurihara, Akikazu
no journal, ,
There is a need for quantitative evaluation of wastage phenomena in steam generator for FBR. We focused attention on liquid droplet impingement erosion (LDIE) in wastage phenomena and performed basic study with piping group mock up specimen for quantitative evaluation of LDIE. First, we did visualization test of high pressure and high speed jet into the water. Test section mock up the crack of heat exchanger tube and neighboring heat exchanger tubes. We did the test under the following test conditions. Upstream pressure is 0.3 MPa, vapor temperature is 300 K, crack width is 0.1 mm, and crack length is 40 mm. (crack diameter is 0.2 mm) Second, we did pressure and temperature measurement test in the same test conditions as before. We evaluated jet behavior at test section by those two tests. In addition, we did two phase flow analysis of the jet with TRAC code.
Tamura, Takeshi*; Soga, Kazuo*; Sugiyama, Kenichiro*; Narabayashi, Tadashi*; Ohshima, Hiroyuki; Suda, Kazunori
no journal, ,
no abstracts in English
Tamura, Takeshi*; Soga, Kazuo*; Sugiyama, Kenichiro*; Narabayashi, Tadashi*; Ohshima, Hiroyuki; Suda, Kazunori
no journal, ,
no abstracts in English
Yoshida, Atsuro*; Khoo Chong Weng, W.*; Narabayashi, Tadashi*; Tsuji, Masashi*; Ohshima, Hiroyuki; Kurihara, Akikazu
no journal, ,
Basic research has been carried out to clarify the wastage phenomena due to tube failure of steam generator in sodium-cooled fast reactor. The authors report the pressure distribution in two-phase free jet under high temperature and high pressure conditions measured by traverse survey.
Khoo Chong Weng, W.*; Higashi, Yuma*; Narabayashi, Tadashi*; Tsuji, Masashi*; Ohshima, Hiroyuki; Kurihara, Akikazu
no journal, ,
In a steam generator of sodium-cooled fast reactor, in case of small reacting jet development, a deterioration of neighboring tube is caused mainly by erosion-corrosion. This study is carried out to evaluate the thinning rate of SG heat transfer tube material caused by liquid droplet impingement erosion using high speed rotating disc test rig.
Ohshima, Hiroyuki; Kurihara, Akikazu; Yamaguchi, Akira*; Takata, Takashi*; Narabayashi, Tadashi*; Deguchi, Yoshihiro*
no journal, ,
A new multi-physics numerical simulation system is being developed based on a 4-year R&D plan for the sake of the evaluation of sodium-water reaction phenomena, which are caused when a heat transfer tube is failed in a steam generator of a sodium-cooled fast reactor. 2-year study results are summarized as an interim report.
Kato, Keisuke*; Narabayashi, Tadashi*; Tsuji, Masashi*; Chiba, Go*; Ohshima, Hiroyuki; Kurihara, Akikazu; Uchibori, Akihiro
no journal, ,
The authors measured the two-phase impinging jet behavior using tube bundle visualization test rig, and analyzed this expetriment by use of SERAPHIM code.
Kurihara, Akikazu; Umeda, Ryota; Shimoyama, Kazuhito; Kikuchi, Shin; Ohshima, Hiroyuki; Narabayashi, Tadashi*
no journal, ,
Wastage phenomena on adjacent tubes (target-wastage) arise from water/steam leak in steam generators of sodium-cooled fast reactors. Target-wastage is likely to be caused by liquid droplet impingement erosion (LDI) and flow-accelerated corrosion (FAC) in an environment marked by high-temperature and high-alkali (reaction jet) due to sodium-water reaction. The authors carried out flow-accelerated corrosion experiments as a part of phenomena clarification experiments for target-wastage by using tube material under high-temperature sodium-hydroxide and sodium monoxide conditions which are mainly generated by sodium-water reaction. New wastage correlations were derived from LDI and FAC data based on influencing factors which were formed on the periphery of an adjacent tube, and were confirmed those applicability to water leak event in this report.
Ohshima, Hiroyuki; Kurihara, Akikazu; Yamaguchi, Akira*; Takata, Takashi*; Narabayashi, Tadashi*; Deguchi, Yoshihiro*
no journal, ,
When a heat transfer tube is failed in a steam generator (SG) of a sodium-cooled fast reactor (SFR), pressurized water and/or water vapor leaks into liquid sodium surrounding the tube and forms a reacting jet with high temperature and high alkali. This reacting jet might cause the secondary failure of adjacent heat transfer tubes due to wastage or over-heating tube rapture resulting in undesirable failure propagation. Therefore, the sodium-water reaction phenomenon (SWR) is one of the most important issues for the design and safety assessment of SFRs. The authors have carried out systematic experiments for the elucidation of SWR and developed a new multi-physics numerical simulation system which is based on mechanistic and theoretical modeling of SWR rather than empirical modeling and can contribute to detailed and quantitative evaluations of SWR in any types of SGs. This paper summarizes the results of four years' R&D activities.
