Development of a coarse-mesh subchannel CFD model for prediction of core thermal-hydraulics in natural circulation conditions

Hamase, Erina; Miyake, Yasuhiro*; Imai, Yasutomo*; Doda, Norihiro; Ono, Ayako; Tanaka, Masaaki

Nuclear Engineering and Design, 432, p.113738_1 - 113738_12, 2025/02

https://doi.org/10.1016/j.nucengdes.2024.113738

To enhance the safety of sodium-cooled fast reactors, the natural circulation (NC) decay heat removal systems with a dipped-type direct heat exchanger (D-DHX) have been investigated. During the D-DHX operation, since the core-plenum interaction occurs, the reactor vessel model using a computational fluid dynamics code (RV-CFD) is required to be established. Previously, the CFD model based on the subchannel analysis was developed. In this study, to achieve lower computational cost maintaining the prediction accuracy, the coarse-mesh subchannel CFD (CMSC) model was developed, and was incorporated into the core of RV-CFD. As a result of PLANDTL-1 test analysis, the RV-CFD with the CMSC model can reproduce the core-plenum interaction under NC conditions.

Development of reactor vessel thermal-hydraulic analysis method in natural circulation conditions; Investigation of interwrapper Gap model

Hamase, Erina; Doda, Norihiro; Ono, Ayako; Tanaka, Masaaki; Miyake, Yasuhiro*; Imai, Yasutomo*

Proceedings of 14th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, Operation, and Safety (NTHOS-14) (Internet), 12 Pages, 2024/08

We have been developed a thermal-hydraulic analysis model in the reactor vessel using the computational fluid dynamics code with a low computational cost to evaluate core-plenum interactions during a natural circulation decay heat removal using a dipped-type direct heat exchanger in a design of sodium-cooled fast reactors. In this study, we investigate the coarse mesh modeling of interwrapper gap (IWG) using correlations for the purpose of the development of a practical model which can reduce the computational cost maintaining the prediction accuracy. An influence of combinations of the coarse mesh and the correlation for pressure loss in the IWG on the thermal-hydraulics and the core temperature distribution is revealed through the numerical analysis of a sodium experiment.

Magnetic and transport properties of the pseudobrookite AlTiO single crystals

Takahama, Ryusei*; Arizono, Mitsutoshi*; Indo, Daigo*; Yoshinaga, Taisei*; Terakura, Chieko*; Takeshita, Nao*; Shirasaki, Takumi*; Noda, Masaaki*; Kuwahara, Hideki*; Kajimoto, Ryoichi; et al.

JPS Conference Proceedings (Internet), 38, p.011114_1 - 011114_6, 2023/05

https://doi.org/10.7566/JPSCP.38.011114

Core thermal-hydraulics analysis during dipped-type direct heat exchanger operation in natural circulation conditions

Hamase, Erina; Miyake, Yasuhiro*; Imai, Yasutomo*; Doda, Norihiro; Ono, Ayako; Tanaka, Masaaki

Mechanical Engineering Journal (Internet), 9(4), p.21-00438_1 - 21-00438_15, 2022/08

https://doi.org/10.1299/mej.21-00438

To enhance the safety of sodium-cooled fast reactors, a dipped-type direct heat exchanger (D-DHX) has been investigated in a natural circulation decay heat removal system. During the D-DHX operation, the core-plenum interactions occurs and the thermal-hydraulics in the reactor vessel (RV) is complicated, the establishment of thermal-hydraulic analysis model in the RV for computational fluid dynamics code (RV-CFD) is required to simulate the thermal stratification in the upper plenum and thermal-hydraulics in the core. In this study, in terms of using RV-CFD for design study, the subchannel CFD model with low computational cost was adopted to the core of RV-CFD and the numerical simulation was carried out in comparison with the measured data in the sodium test facility named PLANDTL-1. As the result, the calculated sodium temperature in the core had good agreement with the experimental result and the applicability of the RV-CFD for the core-plenum interactions was confirmed.

Gas entrainment phenomenon from free liquid surface in a sodium-cooled fast reactor; Measurements and evaluation on a gas core growth form the liquid surface

Uchida, Mao*; Alzahrani, H.*; Shiono, Mikihito*; Sakai, Takaaki*; Matsushita, Kentaro; Ezure, Toshiki; Tanaka, Masaaki

Proceedings of 19th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-19) (Internet), 16 Pages, 2022/03

Gas entrainment from cover gas is one of key issues for sodium-cooled fast reactors design to prevent unexpected effects to core reactivity. A vortex model based evaluation method has been developed to evaluate the surface vortex gas core growth at the free surface in the reactor vessel. In this study, water experiments were performed to clarify the prediction accuracy for the vortex gas core growth during the vortex drift motion using a circulating water tunnel with an open flow channel test section. Gas core growth were predicted by applying the evaluation method to the numerical analyses performed in the same geometry of the experiments, and compared with the experimental results. It was observed the gas core growth became large at downstream region where downward velocity became large in experiment. However, the gas core length which were predicted from numerical result showed a discrepancy with the experimental result on the peak position and an overestimation of peak value.

Core thermal-hydraulic analysis during dipped-type direct heat exchanger operation in natural circulation conditions

Hamase, Erina; Doda, Norihiro; Ono, Ayako; Tanaka, Masaaki; Miyake, Yasuhiro*; Imai, Yasutomo*

Proceedings of 28th International Conference on Nuclear Engineering (ICONE 28) (Internet), 10 Pages, 2021/08

https://doi.org/10.1115/ICONE28-63380

To enhance the safety of sodium-cooled fast reactors, a dipped-type direct heat exchanger (D-DHX) has been investigated in a natural circulation decay heat removal system. During the D-DHX operation, the core-plenum interactions occurs, therefore, a thermal-hydraulic analysis model in the reactor vessel for computational fluid dynamics code (RV-CFD model) is necessarily required. In this study, the application of the subchannel analysis method for subassemblies to the RV-CFD model was attempted to reduce the calculation costs. Analysis results were compared to the experimental data obtained in the sodium experimental apparatus PLANDTL-1. As the result, the behavior of cold sodium into the simulated core was well grasped and the calculated sodium temperature in the core had good agreement with the experimental result. The applicability of the RV-CFD model was confirmed.

Quasifree neutron knockout reaction reveals a small -Orbital component in the Borromean nucleus B

Yang, Z. H.*; Kubota, Yuki*; Corsi, A.*; Yoshida, Kazuki; Sun, X.-X.*; Li, J. G.*; Kimura, Masaaki*; Michel, N.*; Ogata, Kazuyuki*; Yuan, C. X.*; et al.

Physical Review Letters, 126(8), p.082501_1 - 082501_8, 2021/02

AA2020-0819.pdf:1.29MB

https://doi.org/10.1103/PhysRevLett.126.082501

A quasifree (,) experiment was performed to study the structure of the Borromean nucleus B, which had long been considered to have a neutron halo. By analyzing the momentum distributions and exclusive cross sections, we obtained the spectroscopic factors for and orbitals, and a surprisingly small percentage of 9(2)% was determined for . Our finding of such a small component and the halo features reported in prior experiments can be explained by the deformed relativistic Hartree-Bogoliubov theory in continuum, revealing a definite but not dominant neutron halo in B. The present work gives the smallest - or -orbital component among known nuclei exhibiting halo features and implies that the dominant occupation of or orbitals is not a prerequisite for the occurrence of a neutron halo.

Status of study of long-term assessment of transport of radioactive contaminants in the environment of Fukushima (FY2018) (Translated document)

Nagao, Fumiya; Niizato, Tadafumi; Sasaki, Yoshito; Ito, Satomi; Watanabe, Takayoshi; Dohi, Terumi; Nakanishi, Takahiro; Sakuma, Kazuyuki; Hagiwara, Hiroki; Funaki, Hironori; et al.

JAEA-Research 2020-007, 249 Pages, 2020/10

JAEA-Research-2020-007.pdf:15.83MB

The accident of the Fukushima Daiichi Nuclear Power Station, Tokyo Electric Power Company Holdings, Inc. occurred due to the Great East Japan Earthquake, Sanriku offshore earthquake, of 9.0 magnitude and the accompanying tsunami. As a result, large amount of radioactive materials was released into the environment. Under these circumstances, Japan Atomic Energy Agency (JAEA) has been conducting "Long-term Assessment of Transport of Radioactive Contaminants in the Environment of Fukushima" concerning radioactive materials released in environment, especially migration behavior of radioactive cesium since November 2012. This report is a summary of the research results that have been obtained in environmental dynamics research conducted by JAEA in Fukushima Prefecture.

Unveiling spin-dependent unoccupied electronic states of CoMnGe (Ga) film via Ge (Ga) absorption spectroscopy

Yoshikawa, Tomoki*; Antonov, V. N.*; Kono, Takashi*; Kakoki, Masaaki*; Sumida, Kazuki; Miyamoto, Koji*; Takeda, Yukiharu; Saito, Yuji; Goto, Kazuki*; Sakuraba, Yuya*; et al.

Physical Review B, 102(6), p.064428_1 - 064428_7, 2020/08

https://doi.org/10.1103/PhysRevB.102.064428

Preliminary analysis of sodium experimental apparatus PLANDTL-2 for development of evaluation method for thermal-hydraulics in reactor vessel of sodium fast reactor under decay heat removal system operation condition

Ono, Ayako; Tanaka, Masaaki; Miyake, Yasuhiro*; Hamase, Erina; Ezure, Toshiki

Mechanical Engineering Journal (Internet), 7(3), p.19-00546_1 - 19-00546_11, 2020/06

https://doi.org/10.1299/mej.19-00546

Fully natural circulation decay heat removal systems (DHRSs) are to be adopted for sodium fast reactors, which is a passive safety feature without any electrical pumps. It is required to grasp the thermal-hydraulic phenomena in the reactor vessel and evaluate the coolability of the core under the natural circulation not only for the normal operating condition but also for severe accident conditions. In this paper, the numerical results of the preliminary analysis for the sodium experimental condition with the PLANDTL-2 are discussed to establish an appropriate numerical models for the reactor core including the gap region among the subassemblies and the DHX. From these preliminary analyses, the characteristics of the thermal-hydraulics behavior in the PLANDTL-2 to be focused are extracted.