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Journal Articles

Fundamental experiments of jet impingement and fragmentation simulating the fuel relocation in the core disruptive accident of sodium-cooled fast reactors

Imaizumi, Yuya; Kamiyama, Kenji; Matsuba, Kenichi; Isozaki, Mikio; Suzuki, Toru; Emura, Yuki

Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 5 Pages, 2017/04

In order to simulate the typical accident conditions of the fuel relocation phase in SFRs, the molten alloy of low melting point was discharged into a shallow water pool. The distance between the nozzle exit and the bottom plate was set to a value which was indicated to be insufficient to fragment. As the experimental result, the melt jet reached the bottom plate, and dispersed in all directions along the plate, together with the progress of fragmentation. In addition, the melt temperature on the bottom plate decreased rapidly along the radius direction. These results suggest that the fragmentation which would accompany this rapid cooling would be enhanced by the plate. This enhancement would be caused by the extension of the melt-water interface when the melt was dispersed forcibly by the plate. The solidified debris remained after the discharge showed remarkable fragmentation which was assumed to be caused by the formations of small vapor bubbles in the interface.

Journal Articles

An Empirical correlation to predict the distance for fragmentation of simulated Molten-Core materials discharged into a sodium pool

Matsuba, Kenichi; Isozaki, Mikio; Kamiyama, Kenji; Suzuki, Toru; Tobita, Yoshiharu

Proceedings of 11th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, Operation and Safety (NUTHOS-11) (USB Flash Drive), 8 Pages, 2016/10

In order to evaluate the distance for fragmentation of molten core material discharged into the lower sodium plenum during core disruptive accidents in sodium-cooled fast reactors, experiments with simulated molten materials and coolants (water, sodium) was carried out, where an empirical correlation of the distance for fragmentation was developed. The empirical correlation developed by this study showed a good agreement with the measurement results obtained by the present experiments. It was found that in order to well-predict the distance for fragmentation in sodium, thermal phenomena, such as sodium boiling and resultant vapor expansion, needed to be considered.

Journal Articles

Distance for fragmentation of a simulated molten-core material discharged into a sodium pool

Matsuba, Kenichi; Isozaki, Mikio; Kamiyama, Kenji; Tobita, Yoshiharu

Journal of Nuclear Science and Technology, 53(5), p.707 - 712, 2016/05

 Times Cited Count:9 Percentile:19.91(Nuclear Science & Technology)

In order to develop an evaluation method of the distance for fragmentation of molten core material discharged into the sodium plenum, a sodium experiment with visual observation was conducted using an X-ray imaging system. In the current experiments, 0.9 kg of molten aluminum (initial temperature: around 1473 K) was discharged into a sodium pool (initial temperature: 673 K) through a nozzle (inner diameter: 20 mm). Based on the experimental results, the distance for fragmentation of the liquid column was estimated to be 100 mm in the experiments. Through the sodium experiment, useful knowledge was obtained for the future development of an evaluation method of the distance for fragmentation of molten core material. As a next step, sodium experiments using higher-density molten materials will be conducted to enrich the experimental knowledge. Besides, a new semi-empirical correlation will be developed to evaluate more appropriately the distance for fragmentation under CDA conditions.

Journal Articles

Development of the evaluation methodology for the material relocation behavior in the core disruptive accident of sodium-cooled fast reactors

Tobita, Yoshiharu; Kamiyama, Kenji; Tagami, Hirotaka; Matsuba, Kenichi; Suzuki, Toru; Isozaki, Mikio; Yamano, Hidemasa; Morita, Koji*; Guo, L.*; Zhang, B.*

Journal of Nuclear Science and Technology, 53(5), p.698 - 706, 2016/05

AA2015-0794.pdf:2.46MB

 Times Cited Count:7 Percentile:27.98(Nuclear Science & Technology)

The in-vessel retention (IVR) of core disruptive accident (CDA) is of prime importance in enhancing safety characteristics of sodium-cooled fast reactors (SFRs). In the CDA of SFRs, molten core material relocates to the lower plenum of reactor vessel and may impose significant thermal load on the structures, resulting in the melt through of the reactor vessel. In order to enable the assessment of this relocation process and prove that IVR of core material is the most probable consequence of the CDA in SFRs, a research program to develop the evaluation methodology for the material relocation behavior in the CDA of SFRs has been conducted. This program consists of three developmental studies, namely the development of the analysis method of molten material discharge from the core region, the development of evaluation methodology of molten material penetration into sodium pool, and the development of the simulation tool of debris bed behavior.

Journal Articles

A Numerical study on local fuel-coolant interactions in a simulated molten fuel pool using the SIMMER-III code

Cheng, S.; Matsuba, Kenichi; Isozaki, Mikio; Kamiyama, Kenji; Suzuki, Toru; Tobita, Yoshiharu

Annals of Nuclear Energy, 85, p.740 - 752, 2015/11

 Times Cited Count:19 Percentile:7.5(Nuclear Science & Technology)

Journal Articles

First analysis of local fuel-coolant interactions in a molten pool by SIMMER-III using reactor materials

Cheng, S.; Matsuba, Kenichi; Isozaki, Mikio; Kamiyama, Kenji; Suzuki, Toru; Tobita, Yoshiharu

Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 9 Pages, 2015/05

Journal Articles

The Effect of coolant quantity on local fuel-coolant interactions in a molten pool

Cheng, S.; Matsuba, Kenichi; Isozaki, Mikio; Kamiyama, Kenji; Suzuki, Toru; Tobita, Yoshiharu

Annals of Nuclear Energy, 75, p.20 - 25, 2015/01

 Times Cited Count:6 Percentile:41.19(Nuclear Science & Technology)

Journal Articles

SIMMER-III analyses of local fuel-coolant interactions in a simulated molten fuel pool; Effect of coolant quantity

Cheng, S.; Matsuba, Kenichi; Isozaki, Mikio; Kamiyama, Kenji; Suzuki, Toru; Tobita, Yoshiharu

Science and Technology of Nuclear Installations, 2015, p.964327_1 - 964327_14, 2015/00

 Times Cited Count:2 Percentile:75.47(Nuclear Science & Technology)

Journal Articles

Characteristics of pressure buildup from local fuel-coolant interactions in a simulated molten fuel pool, 2; Numerical analyses using SIMMER-III

Cheng, S.; Matsuba, Kenichi; Isozaki, Mikio; Kamiyama, Kenji; Suzuki, Toru; Tobita, Yoshiharu

Proceedings of 10th International Topical Meeting on Nuclear Thermal Hydraulics, Operation and Safety (NUTHOS-10) (USB Flash Drive), 12 Pages, 2014/12

Journal Articles

Distance for fragmentation of a simulated molten-core material discharged into a sodium pool

Matsuba, Kenichi; Isozaki, Mikio; Kamiyama, Kenji; Suzuki, Toru; Tobita, Yoshiharu

Proceedings of 10th International Topical Meeting on Nuclear Thermal Hydraulics, Operation and Safety (NUTHOS-10) (USB Flash Drive), 7 Pages, 2014/12

In order to develop an evaluation method of the distance for fragmentation of molten core material discharged into the sodium plenum, a sodium experiment with visual observation was conducted using an X-ray imaging system. In the current experiments, 0.9 kg of molten aluminum (initial temperature: around 1473 K) was discharged into a sodium pool (initial temperature: 673 K) through a nozzle (inner diameter: 20 mm). Based on the experimental results, the distance for fragmentation of the liquid column was estimated to be 100 mm in the experiments. Through the sodium experiment, useful knowledge was obtained for the future development of an evaluation method of the distance for fragmentation of molten core material. As a next step, sodium experiments using higher-density molten materials will be conducted to enrich the experimental knowledge. Besides, a new semi-empirical correlation will be developed to evaluate more appropriately the distance for fragmentation under CDA conditions.

Journal Articles

An Experimental study on local fuel-coolant interactions by delivering water into a simulated molten fuel pool

Cheng, S.; Matsuba, Kenichi; Isozaki, Mikio; Kamiyama, Kenji; Suzuki, Toru; Tobita, Yoshiharu

Nuclear Engineering and Design, 275, p.133 - 141, 2014/08

 Times Cited Count:17 Percentile:12.91(Nuclear Science & Technology)

Journal Articles

First application of SIMMER-III to local fuel-coolant interactions in a simulated molten fuel pool

Cheng, S.; Matsuba, Kenichi; Isozaki, Mikio; Kamiyama, Kenji; Suzuki, Toru; Tobita, Yoshiharu

Proceedings of International Symposium on Future I&C for Nuclear Power Plants and International Symposium on Symbiotic Nuclear Power Systems (ISOFIC 2014/ISSNP 2014) (Internet), 10 Pages, 2014/08

Journal Articles

Characteristics of pressure buildup from local fuel-coolant interactions in a simulated molten fuel pool

Cheng, S.; Matsuba, Kenichi; Isozaki, Mikio; Kamiyama, Kenji; Suzuki, Toru; Tobita, Yoshiharu

Proceedings of 22nd International Conference on Nuclear Engineering (ICONE-22) (DVD-ROM), 7 Pages, 2014/07

Journal Articles

Fundamental experiment on the distance for fragmentation of molten core material during core disruptive accidents in sodium-cooled fast reactors

Matsuba, Kenichi; Isozaki, Mikio; Kamiyama, Kenji; Tobita, Yoshiharu; Suzuki, Toru

International Electronic Journal of Nuclear Safety and Simulation (Internet), 4(4), p.272 - 277, 2013/12

In order to develop an evaluation method of the distance for fragmentation of molten core material discharged into the lower sodium plenum during core disruptive accidents (CDAs) in sodium cooled fast reactors, fundamental experiments were conducted using a high-density melt and water as simulants for the molten fuel and coolant, respectively. The melt was discharged into a water pool through a nozzle (inner diameter: from 30 mm to 150 mm) under a simulated CDA condition where a liquid-liquid direct contact is maintained between the melt and water. The present results showed that measured distances for fragmentation were limited to approximately 10 percent of predictions by the existing representative correlation, and that vapor expansion with pressure buildup near the melt could facilitate the fragmentation and thus contribute to the reduction of the distance for fragmentation. Through the fundamental experiments, useful knowledge was obtained for the future development of an evaluation method.

Journal Articles

Preliminary results of a fuel-coolant interaction experiment in simulated molten fuel pool

Cheng, S.; Matsuba, Kenichi; Isozaki, Mikio; Kamiyama, Kenji; Suzuki, Toru; Tobita, Yoshiharu

Proceedings of International Symposium on Symbiotic Nuclear Power Systems for 21st Century (ISSNP 2013) (CD-ROM), 7 Pages, 2013/11

Journal Articles

Fundamental experiment on the distance for fragmentation of molten core material during core disruptive accidents in sodium-cooled fast reactors

Matsuba, Kenichi; Isozaki, Mikio; Kamiyama, Kenji; Tobita, Yoshiharu; Suzuki, Toru

Proceedings of International Symposium on Symbiotic Nuclear Power Systems for 21st Century (ISSNP 2013) (CD-ROM), 6 Pages, 2013/11

In order to develop an evaluation method of the distance for fragmentation of molten core material discharged into the lower sodium plenum during core disruptive accidents (CDAs) in sodium cooled fast reactors, fundamental experiments were conducted using a high-density melt and water as simulants for the molten fuel and coolant, respectively. The melt was discharged into a water pool through a nozzle (inner diameter: from 30 mm to 150 mm) under a simulated CDA condition where a liquid-liquid direct contact is maintained between the melt and water. The present results showed that measured distances for fragmentation were limited to approximately 10 percent of predictions by the existing representative correlation, and that vapor expansion near the melt could facilitate the fragmentation and thus contribute to the reduction of the distance for fragmentation. Through the fundamental experiments, useful knowledge was obtained for the future development of an evaluation method.

Journal Articles

Experimental study on fuel-discharge behavior through in-core coolant channels

Kamiyama, Kenji; Saito, Masaki*; Matsuba, Kenichi; Isozaki, Mikio; Sato, Ikken; Konishi, Kensuke; Zuyev, V. A.*; Kolodeshnikov, A. A.*; Vassiliev, Y. S.*

Journal of Nuclear Science and Technology, 50(6), p.629 - 644, 2013/06

 Times Cited Count:16 Percentile:15.87(Nuclear Science & Technology)

In core disruptive accidents of sodium cooled fast reactors, fuel discharge from the core region reduces the possibility of severe re-criticality events. In-core coolant channels such as the control-rod guide tube and a concept of the FAIDUS (Fuel Assembly with Inner Duct Structure) provide effective fuel discharge paths if effects of sodium in these paths on molten fuel discharge are limited. Two series of experiments conducted in the present study showed that the discharge path can be entirely voided by the vaporization of a part of the coolant at the initial melt discharge phase, that this is followed by coolant vapor expansion, and that melt penetrates significantly into the voided channel. In conclusion, the effects of the sodium on fuel discharge are limited and therefore in-core coolant channels provide effective fuel discharge paths for reducing neutronic activity.

Journal Articles

Mechanism of upward fuel discharge during core disruptive accident in sodium-cooled fast reactors

Matsuba, Kenichi; Isozaki, Mikio; Kamiyama, Kenji; Tobita, Yoshiharu

Journal of Engineering for Gas Turbines and Power, 135(3), p.032901_1 - 032901_9, 2013/03

 Times Cited Count:3 Percentile:76.85(Engineering, Mechanical)

The introduction of Fuel Assembly with Inner Duct Structure (FAIDUS) is being considered to prevent the formation of a large-scale molten fuel pool within a reactor core, which is one of factors leading to the severe power excursion during CDA of SFRs. In the current reference design for FAIDUS, the top end of the inner duct is open whereas the bottom end is closed. In order to clarify the fundamental mechanism for upward fuel discharge through the inner duct, JAEA conducted an out-of-pile experiment in which a high-density melt simulating the molten fuel was injected into a simulated inner duct. In this paper, the mechanism of upward discharge observed in this experiment is discussed in terms of the application to reactor conditions. It was suggested that the coolant pressure buildup could act as one of the driving force for the upward discharge under reactor conditions with higher melt-enthalpy-injection rate than the current simulant experimental condition.

Journal Articles

Mechanism of upward fuel discharge during core disruptive accident in sodium-cooled fast reactors

Matsuba, Kenichi; Isozaki, Mikio; Kamiyama, Kenji; Tobita, Yoshiharu

Proceedings of 20th International Conference on Nuclear Engineering and the ASME 2012 Power Conference (ICONE-20 & POWER 2012) (DVD-ROM), 11 Pages, 2012/07

The introduction of Fuel Assembly with Inner Duct Structure (FAIDUS) is being studied to prevent the formation of a large-scale molten fuel pool within a reactor core, which is one of factors leading to the severe power excursion during CDA of SFRs. In the current reference design for FAIDUS, the top end of the inner duct is open whereas the bottom end is closed. In order to clarify the fundamental mechanism for upward fuel discharge through the inner duct, JAEA conducted an out-of-pile experiment in which a high-density melt simulating the molten fuel was injected into a simulated inner duct. In this paper, the mechanism of upward discharge observed in this experiment is discussed in terms of the application to reactor conditions. It was suggested that the coolant pressure buildup could act as one of the driving force for the upward discharge under reactor conditions with higher melt-enthalpy-injection rate than the current experimental condition.

Journal Articles

Experimental study on upward fuel discharge during core disruptive accident in JSFR; Results of an out-of-pile experiment with visual observation

Matsuba, Kenichi; Isozaki, Mikio; Kamiyama, Kenji; Tobita, Yoshiharu

Proceedings of 19th International Conference on Nuclear Engineering (ICONE-19) (CD-ROM), 6 Pages, 2011/10

The introduction of Fuel Assembly with Inner Duct Structure (FAIDUS) is being studied to prevent the formation of a large-scale molten fuel pool within a reactor core, which is one of factors leading to the severe power excursion during Core Disruptive Accidents of Sodium-cooled Fast Reactors. In the current reference design for FAIDUS, the top end of the inner duct is open whereas the bottom end is closed, and therefore it is expected that the molten fuel will be discharged from a reactor core toward an upper sodium plenum through the inner duct. An out-of-pile experiment, in which a high-density melt simulating the molten fuel was injected into a simulated inner duct structure, was carried out in order to clarify the fundamental mechanism for upward discharge of a high-density melt. Through the experiment, upward melt discharge driven by coolant vapor flow was visually observed, and the fundamental mechanism for upward discharge of a high-density melt was clarified.

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