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Emura, Yuki; Kamiyama, Kenji; Yamano, Hidemasa
Dai-26-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (Internet), 4 Pages, 2022/07
no abstracts in English
Kondo, Satoru; Tobita, Yoshiharu
JAEA-Research 2019-009, 382 Pages, 2020/03
JAEA-Research-2019-009.pdf:22.82MB
The SIMMER-III computer code, developed at the Japan Atomic Energy Agency (JAEA, the former Power Reactor and Nuclear Fuel Development Corporation), is a two-dimensional, multi-velocity-field, multi-component fluid-dynamics code, coupled with a space- and time-dependent neutron kinetics model. The code is being used widely for simulating complex phenomena during core-disruptive accidents (CDAs) in liquid-metal fast reactors (LMFRs). In parallel to the code development, a comprehensive assessment program was performed in two phases: Phase 1 for verifying individual fluid-dynamics models; and Phase 2 for validating its applicability to integral phenomena important to evaluating LMFR CDAs. The SIMMERIII assessment program was participated by European research and development organizations, and the achievement of Phase 1 was compiled and synthesized in 1996. This report has been edited by revising and reproducing the original 1996 informal report, which compiled the achievement of Phase 1 assessment. A total of 34 test problems were studied in the areas: fluid convection, interfacial area and momentum exchange, heat transfer, melting and freezing, and vaporization and condensation. The problems identified have been reflected to the Phase 2 assessment and later model development and improvement. Although the revisions were made in the light of knowledge base obtained later, the original individual contributions by the participants, both positive and negative, are retained except for editorial changes.
Matsuba, Kenichi; Kamiyama, Kenji; Toyooka, Junichi; Zuev, V. A.*; Kolodeshnikov, A. A.*
Proceedings of 17th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-17) (USB Flash Drive), 11 Pages, 2017/09
To clarify jet fragmentation and accompanying cooling behaviors of molten core materials discharged into sodium, results of the out-of-pile experiments with a simulant material (Al
O
) were analyzed. The results clarified that while AlO jets were entirely fragmented into smaller particles during their penetration to several tenths of a meter in depth of sodium, an additional significant distance was needed to cool down the particles to the degree that thermal loading on the steel structures could be neglected. Based on the results, it is concluded that in terms of the reduction of thermal load on the lower structures in the reactor vessels, the cooling distance after fragmentation should be treated.
JNC TN9400 2000-038, 98 Pages, 2000/04
JNC-TN9400-2000-038.pdf:7.49MB
As an effort in the feasibility study on commercialized Fast Breeder Reactor cycle systems, an evaluation of the measures to prevent the energetic re-criticality in sodium-cooled large MOX core, which is one of the candidates for the commercialized reactor, has been performed. The core disruptive accident analysis of Demonstration FBR showed that the fuel compaction of the molten fuel by radial motion in a large molten core pool had a potential to drive the severe super-prompt re-criticality phenomena in ULOF sequence. ln order to prevent occurrence of the energetic re-criticality, a subassembly with an inner duct and the removal of a part of LAB are suggested based on CMR (Controlled Material Relocation) concept. The objective of this study is the comparison of the effectiveness of CMR among these measures by the analysis using SIMMER-III. The molten fuel in the subassembly with inner duct flows out faster than that from other measures. The subassembly with inner duct will work effectively in preventing energetic re-criticality. Though the molten fuel in the subassembly without a part of LAB flows out a little slower, it is still one of the promising measures. However, the UAB should be also removed from the same pin to prevent the fuel re-entries into the core region due to the pressurization by FCl below the core, unless it disturbs the core performance. The effect of the axial fuel length of the center pin to CMR behavior is small, compared to the effect of the existence of UAB.
Kamiyama, Kenji; Matsuba, Kenichi; Tobita, Yoshiharu; Toyooka, Junichi; Pakhnits, A. V.*; Vityuk, V. A.*; Kukushkin, I.*; Vurim, A. D.*; Baklanov, V. V.*; Kolodeshnikov, A. A.*
no journal, ,
no abstracts in English
Kamiyama, Kenji; Matsuba, Kenichi; Isozaki, Mikio; Imaizumi, Yuya; Suzuki, Toru; Emura, Yuki
no journal, ,
no abstracts in English
Emura, Yuki; Matsuba, Kenichi; Isozaki, Mikio; Imaizumi, Yuya; Kamiyama, Kenji
no journal, ,
A series of simulation tests discharging a molten alloy with low-melting point into a shallow water pool is being conducted in order to investigate fuel-coolant interaction and sedimentation behavior of molten core materials on the structure of core inlet coolant plenum in the core disruptive accident of sodium-cooled fast reactors. The present study shows effects of the initial temperature of the molten alloy, the confined coolant volume, etc. on the test results.
Matsuba, Kenichi; Isozaki, Mikio; Toyooka, Junichi; Kamiyama, Kenji; Zuev, V.*; Kolodeshnikov, A.*
no journal, ,
In order to assess the fragmentation behavior of molten core material discharged into the lower sodium plenums during core disruptive accidents in sodium-cooled fast reactors, applicability of empirical correlations of the distance for fragmentation was investigated by comparing the values predicted using a correlation with the results obtained in the experiments where a simulated core material (molten alumina) was discharged into a sodium pool.
Kato, Shinya; Matsuba, Kenichi; Kamiyama, Kenji
no journal, ,
In order to clarify the event progression and its influential factors in the core disruptive accident of the sodium-cooled fast reactor, JAEA has carried out EAGLE(Experimental Acquisition of Generalized Logic to Eliminate re-criticalities)-3 project with the cooperation of National Nuclear Center of the Republic of Kazakhstan. The result of the out-of-pile test focused on the outflow of the molten-core material through the control rod guide tube will be shown.
Emura, Yuki; Isozaki, Mikio; Matsuba, Kenichi; Kamiyama, Kenji
no journal, ,
In core-disruptive accidents of sodium-cooled fast reactors, molten-core materials are discharged into a lower part of reactor vessel and are mixed with sodium there. In this study, molten stainless steel was poured into sodium and its behavior in the sodium was observed by a visualization system using X-ray.
Emura, Yuki; Isozaki, Mikio; Matsuba, Kenichi; Kamiyama, Kenji
no journal, ,
In core-disruptive accidents of sodium-cooled fast reactors, molten core materials flow down and penetrate into sodium in a lower part of reactor vessel. In this study, molten stainless steel which is one of components of molten core materials was poured into sodium pool and its penetration behavior into sodium was observed using X-ray system.
Igarashi, Kai*; Sakai, Takaaki*; Kato, Shinya; Matsuba, Kenichi; Kamiyama, Kenji
no journal, ,
In order to clarify the accumulation behavior on the core disruptive accident in sodium cooled fast reactors, an analysis for the experiment in which the low-melting point alloy drop into a shallow water pool was conducted by using SIMMER code. In this presentation, the validation result for the analysis geometry model conducted by the comparison of the calculation result and experimental value is presented.
Kato, Shinya; Matsuba, Kenichi; Kamiyama, Kenji; Mikisha, A.*; Akayev, A.*; Vurim, A.*; Baklanov, V.*
no journal, ,
In order to enrich experimental knowledge of discharge behavior of molten core materials through a sodium-filled channel in core disruptive accidents of sodium cooled fast reactors, an out-of-pile experiment was conducted, in which molten alumina was used as a molten-fuel simulant and it penetrated into a sodium-filled channel with an internal structure reducing a flow area. In this presentation, consideration on effects of the internal structure on melt discharge-behavior will be presented based on experimental results.
Kato, Shinya; Matsuba, Kenichi; Kamiyama, Kenji
no journal, ,
no abstracts in English
Matsushita, Hatsuki*; Kobayashi, Ren*; Sakai, Takaaki*; Kato, Shinya; Matsuba, Kenichi; Kamiyama, Kenji
no journal, ,
no abstracts in English
Eto, Masao*; Yokoi, Shinobu*; Yamashita, Masato*; Miura, Kazuhiro*; Okafuji, Takashi*; Onoda, Yuichi; Yamamoto, Tomohiko; Uchita, Masato*
no journal, ,
no abstracts in English
