Study on the discharge behavior of the molten-core materials through the control rod guide tube; Investigations of the effect of an internal structure in the control rod guide tube on the discharge behavior

Kato, Shinya; Matsuba, Kenichi; Kamiyama, Kenji; Akaev, A.*; Vurim, A.*; Baklanov, V.*

Proceedings of 13th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, Operation and Safety (NUTHOS-13) (Internet), 12 Pages, 2022/09

The In-Vessel Retention (IVR) of molten-core in Core Disruptive Accidents (CDAs) is of prime importance in enhancing the safety of sodium-cooled fast reactors. One of the main subjects in ensuring IVR is to design the Control Rod Guide Tube (CRGT) which allows effective discharge of molten core materials from the core region. The effectiveness of the CRGT design is assessed through CDA analyses, and it is reasonable for these analyses to develop a computer code collaborated with experimental researches. Thus, experiments addressing the discharge behavior of the molten-core materials through the CRGT have proceeded as one of the subjects in the collaboration research named the EAGLE-3 project, and the obtained experimental results are reflected in the development of the SIMMER code. In this project, a series of out-of-pile tests using molten-alumina as the fuel simulant was conducted to understand the discharge behavior of molten-core materials through the CRGT. In this study, in order to investigate the effect of an internal structure in the CRGT on the discharge behavior of the molten-core materials, the data of an out-of-pile test in which the molten-alumina penetrated to a duct with the internal structure were analyzed. In addition, the post-test analysis using the SIMMER code was conducted and the results were compared with the test results.

Study on the discharge behavior of molten-core through the control rod guide tube in the core disruptive accident of SFR

Kato, Shinya; Matsuba, Kenichi; Kamiyama, Kenji; Ganovichev, D. A.*; Baklanov, V. V.*

Proceedings of 2019 International Congress on Advances in Nuclear Power Plants (ICAPP 2019) (Internet), 9 Pages, 2019/05

In order to ensure In-Vessel Retention (IVR) of molten-core in Core Disruptive Accident (CDA), we are investigating the possibility of the molten-core discharge through the control rod guide tube (CRGT) to prevent energetics due to exceeding the prompt criticality. Internal structures of the CRGT, such as a sodium-flow regulator when the CRGT is connected to the high-pressure plenum, may disturb the discharge of molten-core from the core region. Based on above background, an experimental program to clarify characteristics of molten-core discharge through the CRGT has been commenced as one of subjects under a joint study with National Nuclear Center of the Republic of Kazakhstan (NNC-RK) named EAGLE-3 project. An experiment using molten-alumina as fuel simulant and sodium was conducted at the out-of-pile test facility owned by NNC-RK to investigate sodium cooling effect around the sodium flow regulator on its destruction. The experimental result represented that void development at the initiation of molten-alumina discharge eliminated liquid-phase sodium from the discharge path and this also eliminated sodium cooling effect around the sodium flow regulator. As a result, early destruction of the sodium flow regulator and massive discharge of molten alumina occurred in turn.

The Research cooperation with the National Nuclear Center of the Republic of Kazakhstan to enhance the safety of sodium-cooled fast reactors

Kamiyama, Kenji

no journal, , 

Japan Atomic Energy Agency (JAEA) has agreed to the research cooperation on the core safety of sodium-cooled fast reactors (SFRs) with the National Nuclear Center of the Republic of Kazakhstan (NNC-RK), and it has been going on for over 20 years. This research cooperation is called the EAGLE project, which is an advanced and challenging research program utilizing the facilities of NNC-RK. The background and outline of this EAGLE program, as well as the implementation status and major achievements so far, are introduced here.

France-Japan collaboration on severe accident studies in sodium-cooled fast reactors, 5; IGR programs interpretation and the SIMMER code validation

Kato, Shinya; Imaizumi, Yuya; Matsuba, Kenichi; Kamiyama, Kenji; Yamano, Hidemasa; Kubo, Shigenobu; Dufour, E.*; Trotignon, L.*; Payot, F.*

no journal, , 

JAEA and CEA are going on the collaboration study to validate the SIMMER code since 2020. In this framework, JAEA and CEA are performing mutually the SIMMER calculations based on the joint interpretation on in-pile tests of the EAGLE program conducted in the IGR research reactor in Kazakhstan. The comparison between the calculation results and the experimental results for the model validation of the SIMMER code are conducted. In this presentation, the progress and the achievement of the SIMMER validations under this collaborative study are shown.

Experimental studies in the EAGLE-3 project for controlled material relocation in severe accidents of sodium-cooled fast reactors

Kato, Shinya; Matsuba, Kenichi; Kamiyama, Kenji; Mukhamedov, N.*; Akayev, A.*; Pakhnits, A.*; Vurim, A.*; Baklanov, V.*

no journal, ,