Study on heat transfer behavior of a cylindrical particle bed with volumetric heating

Wen, J.*; Kamada, Yuto*; Yokoyama, Kosei*; Matsumoto, Tatsuya*; Liu, W.*; Morita, Koji*; Imaizumi, Yuya; Tagami, Hirotaka; Matsuba, Kenichi; Kamiyama, Kenji

Proceedings of 13th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS13) (Internet), 8 Pages, 2024/11

Experimental study on the coolability of molten core materials discharged into a depth- and volume-limited sodium plenum

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

Proceedings of 31st International Conference on Nuclear Engineering (ICONE31) (Internet), 7 Pages, 2024/11

During a severe accident in sodium-cooled fast reactors, molten core materials could be discharged from the core region toward the lower sodium region of the reactor vessel through coolant channels, such as control rod guide tubes. Typical SFRs have a sodium plenum with limited depth and volume, such as the core inlet plenum located under the core region. Therefore, it is important to evaluate the coolability of molten core materials discharged into a depth- and volume-limited sodium plenum. In the present study, to deepen the understanding on the coolability of molten core materials discharged into such a sodium plenum, conditions under which molten core materials form solidified fragments were discussed based on an experiment discharging a molten fuel simulant (molten Al2O3) into a test vessel filled with liquid sodium.

SIMMER-III and SIMMER-IV; Computer codes for LMFR core disruptive accident analysis

Kondo, Satoru; Tobita, Yoshiharu*; Morita, Koji*; Kamiyama, Kenji; Yamano, Hidemasa; Suzuki, Toru*; Tagami, Hirotaka; Sogabe, Joji; Ishida, Shinya

JAEA-Research 2024-008, 235 Pages, 2024/10

JAEA-Research-2024-008.pdf:4.77MB

The SIMMER-III and SIMMER-IV computer codes, developed at the Japan Atomic Energy Agency are the codes with two- and three-dimensional, multi-field, multi-component fluid-dynamics models, coupled with a space- and time-dependent neutron kinetics model. The codes have been used widely for simulating complex phenomena during core-disruptive accidents in liquid-metal fast reactors. Advanced features of the codes in comparison with the former codes include: stable and robust fluid-dynamics algorithm with up to 8 velocity fields, improved representation of structures and multi-phase flow topology, comprehensive treatment of complex heat and mass transfer processes, accurate analytic equations of state, a stable and efficient neutron flux shape solution method and decay heat model. This report describes the models and methods of SIMMER-III and SIMMER-IV. For those individual models, the details of which have been reported elsewhere, only the outlines of the models are presented. The reports of code verification and validation have been already published.

France-Japan collaboration on severe accident studies in sodium-cooled fast reactors, 1; Severe accident scenarios assessment

Onoda, Yuichi; Ishida, Shinya; Fukano, Yoshitaka; Kamiyama, Kenji; Yamano, Hidemasa; Kubo, Shigenobu; Shibata, Akihiro*; Bertrand, F.*; Seiler, N.*

Proceedings of International Conference on Nuclear Fuel Cycle (GLOBAL2024) (Internet), 4 Pages, 2024/10

Thinning behavior of solid boron carbide immersed in molten stainless steel for core disruptive accident of sodium-cooled fast reactor

Emura, Yuki; Takai, Toshihide; Kikuchi, Shin; Kamiyama, Kenji; Yamano, Hidemasa; Yokoyama, Hiroki*; Sakamoto, Kan*

Journal of Nuclear Science and Technology, 61(7), p.911 - 920, 2024/07

https://doi.org/10.1080/00223131.2023.2287109

Experiment and new analysis model simulating in-place cooling of a degraded core in severe accidents of sodium-cooled fast reactors

Imaizumi, Yuya; Aoyagi, Mitsuhiro; Kamiyama, Kenji; Matsuba, Kenichi; Akaev, A.*; Mikisha, A.*; Baklanov, V.*; Vurim, A.*

Annals of Nuclear Energy, 194, p.110107_1 - 110107_11, 2023/12

https://doi.org/10.1016/j.anucene.2023.110107

A Series of molten stainless steel-sodium interaction experiments to develop an evaluation methodology for jet breakup during core disruptive accidents in sodium-cooled fast reactors

Matsuba, Kenichi; Emura, Yuki; Kamiyama, Kenji

Proceedings of 2023 International Congress on Advanced in Nuclear Power Plants (ICAPP 2023) (Internet), 8 Pages, 2023/04

A series of Fuel-Coolant Interaction (FCI) experiments discharging about 1 kg of molten stainless steel (SS), one of prototypic materials, into a sodium pool (about 30 cm in diameter, about 1 m in depth) has been conducted at the MELT facility of Japan Atomic Energy Agency. X-ray visualization program confirmed that molten SS jets with diameters of about 1 to 2 cm were broken up by FCIs which occurred with sodium vapor expansion at depths near the initial sodium level. Responses of thermocouples suggested that the jets were rapidly cooled near the initial sodium level and became SS fragments before reaching the bottom of the sodium pool. Results of experiments with molten SS jets confirmed that an evaluation methodology for jet breakup under sodium-cooled conditions needs to be developed by taking into account effects of thermal parameters related to sodium vaporization as well as conventional parameters, such as density ratio of a jet and coolant, and jet diameter.

A 3D particle-based simulation of heat and mass transfer behavior in the EAGLE ID1 in-pile test

Zhang, T.*; Morita, Koji*; Liu, X.*; Liu, W.*; Kamiyama, Kenji

Annals of Nuclear Energy, 179, p.109389_1 - 109389_10, 2022/12

https://doi.org/10.1016/j.anucene.2022.109389

Measurements of pressure drop and void fraction of air-water two-phase flow in a sphere-packed bed

Yamamoto, Seishiro*; Odaira, Naoya*; Ito, Daisuke*; Ito, Kei*; Saito, Yasushi*; Imaizumi, Yuya; Matsuba, Kenichi; Kamiyama, Kenji

Proceedings of 12th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS12) (Internet), 4 Pages, 2022/10

Analysis on cooling behavior for simulated molten core material impinging to a horizontal plate in a sodium pool

Matsushita, Hatsuki*; Kobayashi, Ren*; Sakai, Takaaki*; Kato, Shinya; Matsuba, Kenichi; Kamiyama, Kenji

Proceedings of 13th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, Operation and Safety (NUTHOS-13) (Internet), 9 Pages, 2022/09

During core disruptive accidents in sodium-cooled fast reactors, the molten core material flows through flow channels, such as the control rod guide tubes, into the core inlet plenum under the core region. The molten core material can be cooled and solidified while impinging on a horizontal plate of the inlet plenum in a sodium coolant. However, the solidification and cooling behaviors of molten core materials impinged on a horizontal structure have not been sufficiently studied thus far. Notably, this is an important phenomenon that needs to be elucidated from the perspective of improving the safety of sodium-cooled fast reactors. Accordingly, a series of experiments on discharging a simulated molten core material (alumina: AlO) into a sodium coolant on a horizontal structure was conducted at the experimental facility of the National Nuclear Center of the Republic of Kazakhstan. In this study, analyses on the sodium experiments using SIMMER-III as the fast reactor safety evaluation code were performed. The analysis methods were validated by comparing the results and experiment data. In addition, the cooling and solidification behaviors during jet impingement were evaluated. The results indicated that the molten core material exhibited fragmentation owing to the impingement on the horizontal plate and was, therefore, scattered toward the periphery. Furthermore, the simulated molten core material was evaluated to be cooled by sodium and subsequently solidified.