Study on evaluation method of decay heat removal of pool-type sodium-cooled fast reactor

Ezure, Toshiki; Kobayashi, Jun; Matsuno, Keiko*; Tanaka, Masaaki

Dai-30-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (Internet), 5 Pages, 2026/06

In the design of the demonstration tank-type sodium-cooled fast reactor, an evaluation method for decay heat removal including under natural circulation conditions have developed as one of the key R&D issues in the conceptual design phase. For the demonstration reactor, installation of Intermediate Reactor Auxiliary Cooling System (IRACS) as one of decay removal system is considered as one of the diversified decay heat removal systems to enhance the safety. This paper focuses on thermal stratification in the hot plenum during the decay heat removal using IRACS and presents the results of PLANDTL-2 experiments conducted with consideration of similarity laws to the demonstration reactor.

Development of numerical analysis technologies in an integrated evaluation system, ARKADIA, for plant system thermal-hydraulics related issues in advanced sodium-cooled fast reactors

Tanaka, Masaaki; Doda, Norihiro; Hamase, Erina; Mori, Takero; Yoshimura, Kazuo; Yoshikawa, Ryuji; Kikuchi, Norihiro; Matsushita, Kentaro; Mochinaga, Shota; Ezure, Toshiki

Proceedings of International Conference on Fast Reactors and Related Fuel Cycles (FR26) (Internet), 9 Pages, 2026/05

This paper presents the outline of the evaluation support program on the evaluations of the thermal-hydraulics related issues in the conceptual design study of the sodium-cooled fast reactor for demonstration in Japan, utilizing the technical basis on the simulations and evaluations configured in the integrated evaluation system named ARKADIA that has been developed with experiences through the development of the SFRs and related R&Ds in Japan Atomic Energy Agency. Moreover, as the extension of the function of the ARKADIA, the possible framework for SFR lifecycle development introducing the Digital-Twin for the MBD with ARKADIA is presented with an example of a whole plant coupled analysis model developed in ARKADIA.

Stability threshold for stratification-induced flowrate oscillations in PLANDTL-2 sodium natural circulation experiment

Renaudie de Vaux, S.*; Li, S.*; Marrel, A.*; Ezure, Toshiki; Tanaka, Masaaki

Nuclear Engineering and Design, 444, p.114382_1 - 114381_15, 2025/12

https://doi.org/10.1016/j.nucengdes.2025.114382

In this study, experimental results on sodium natural circulation in PLANDTL-2 under eight transient scenarios were reported. The experimental parameters were the initial state of the primary circuit, the shutdown inertia of the secondary circuit, and the operating mode of the Decay Heat Removal system (DHRS). As the results, it was clarified that flow reversals can be observed in the primary circuit under rapid cooling cases by DHRS linked to an enhancement of thermal stratification in the plenum. A novel method was introduced to estimate the delay between DHR system activation and increase in stratification. In addition, some guidelines for future investigations were given in order to avoid the presently described instabilities.

Evaluation of vortex gas entrainment phenomena

Ito, Kei*; Matsushita, Kentaro; Ezure, Toshiki; Tanaka, Masaaki; Odaira, Naoya*; Ito, Daisuke*; Saito, Yasushi*

Nihon Kikai Gakkai 2025-Nendo Nenji Taikai Koen Rombunshu (Internet), 5 Pages, 2025/09

The estimation of entrained gas flow rate by a bathtub vortex is important in terms of a possibility to causes the performance degradation when the entrained bubbles are mixed into fluid machineries, e.g. pumps. In this study, to confirm the applicability of a model based on circulating annular flow model proposed by the authors, entrained gas flow rate is evaluated using the liquid velocity distribution around free surface dent of vortex (gas core), obtained by CFD data. As a result, it was indicated that it would be possible to evaluate the gas entrainment flow rate by setting an appropriate evaluation region.

Specifications for benchmark analyses of transient thermal-hydraulics in reactor vessel and primary heat transport system during decay heat removal operation

Kobayashi, Jun; Tanaka, Masaaki; Hamase, Erina; Ezure, Toshiki

JAEA-Data/Code 2025-009, 74 Pages, 2025/08

JAEA-Data-Code-2025-009.pdf:4.7MB

In a sodium-cooled fast reactor, a diversified auxiliary cooling system to remove decay heat from the core is required to enhance its safety. The decay heat removal systems (DHRSs) include a direct reactor auxiliary cooling system (DRACS) with a heat exchanger in the upper plenum (UP) of the reactor vessel (RV), a primary reactor auxiliary cooling system (PRACS) with a heat exchanger in the primary heat transport system (PHTS), an intermediate reactor auxiliary cooling system (IRACS) with a heat exchanger in the secondary heat transport system (SHTS), a heat removal system which employs a steam generator, and a reactor vessel auxiliary cooling system (RVACS) that effects cooling from outside the RV. In the operation of the DRACS with a dipped-type direct heat exchanger (D-DHX) installed in the UP of the RV (UP-RV), the thermal interaction, called core-plenum interaction (CPI), regarding the thermal-hydraulic phenomena in the UP-RV and the core is observed. The CPI includes the penetration flow of the sodium at a low temperature from the D-DHX into the core assemblies, the flow in the gap between assemblies, and the radial heat transfer through sodium in the gap. On the other hand, in the operation of the PRACS or IRACS, where the flowrate in the PHTS is maintained, the core coolability is affected by plant operating conditions. Two transient tests conducted at the PLANDTL-DHX sodium test facility in Japan Atomic Energy Agency were provided to develop an appropriate numerical analysis model for prediction of transient thermal-hydraulics in the DHRSs, the core, and the PHTS. In this document, the geometry information of the RV and the PHTS, including the heat exchangers for the DHRS, and the measured flowrate and temperature transients at each inlet of the intermediate heat exchanger (IHX) on the SHTS side and DHRS were specified as the boundary conditions for the benchmark analyses.

Development of gas entrainment evaluation model based on distribution of pressure along vortex center line; Application to a gas entrainment experiment with traveling vortices in an open water channel flow?

Matsushita, Kentaro; Ezure, Toshiki; Tanaka, Masaaki; Imai, Yasutomo*; Fujisaki, Tatsuya*; Sakai, Takaaki*

Nuclear Engineering and Design, 432, p.113785_1 - 113785_16, 2025/02

https://doi.org/10.1016/j.nucengdes.2024.113785

Establishing an evaluation method for the gas entrainment (GE) of argon cover gas due to surface vortices is required in terms of safety design of sodium-cooled fast reactors. To modify the evaluation model in an in-house evaluation tool for GE, StreamViewer, a modified evaluation model on the pressure distribution along the vortex center line (PVL model) was proposed to identify the vortex center lines by connecting continuous vortex center points from the suction port to the surface and evaluate gas core length based on the balance between the hydrostatic pressure and the pressure decrease distribution along the vortex center line. PVL model was applied the three-dimensional numerical analysis results for the experiments where a plate induced unsteady traveling vortices in the open channel flow. Consequently, the GE evaluation using StreamViewer with PVL model could reproduce the relation between the inlet flow velocity and the gas core length in the unsteady vortex flow experiments.

Application of AMR method for numerical analysis of water experiment involving advective vortices

Matsushita, Kentaro; Ezure, Toshiki; Fujisaki, Tatsuya*; Imai, Yasutomo*; Tanaka, Masaaki

Nihon Kikai Gakkai 2024-Nendo Nenji Taikai Koen Rombunshu (Internet), 5 Pages, 2024/09

An evaluation method of gas entrainment phenomena due to free surface vortices has been developed for the design of a reactor vessel of sodium-cooled fast reactor. The method predicts vortex dimple using the vortex model to the flow field obtained from three dimensional hydraulic analyses of an evaluation area. In this study, the application of adaptive mesh refinement (AMR) method to a water flow experiment in a rectangular channel with advection vortices was examined to create analysis meshes automatically. Transient analyses were conducted using refined meshes obtained by AMR under different initial grid size conditions. Then, the quantities related to vortex formation and the computation cost were compared with the result in a reference mesh with uniformly fine grids. As the result, it was confirmed that the variation of the grid number is possible to use as a criterion to judge the refinement termination in AMR, and the calculated cost of transient analysis can be reduced by AMR.

Development of gas entrainment evaluation method considering three-dimensional pressure decrease distribution along the center of free surface vortex

Matsushita, Kentaro; Ezure, Toshiki; Imai, Yasutomo*; Fujisaki, Tatsuya*; Tanaka, Masaaki

Dai-27-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (Internet), 5 Pages, 2023/09

https://doi.org/10.1299/jsmepes.2023.27.A123

In design of sodium-cooled fast reactors (SFRs), cover gas entrainment phenomena induced by vortex dimple at free surface in upper plena is an important thermal-hydraulic issue. Authors have developed an evaluation method of gas entrainment with an evaluation tool named "StreamViewer". In this study, modification of evaluation model to improve quantitatively prediction accuracy of gas core length was investigated. In this model, vortex center lines which elongated from suction port where entrance of gas to heat transport system, for instance, IHX inlet in pool type SFRs, to free surface in plenum were to be identified, and distribution of pressure decrease along vortex center line was calculated to judge possibility of gas entrainment in comparisons with hydraulic head. This evaluation model was applied to results of water experiment with a rectangular open channel, where unsteady vortices are generated. It was confirmed that this model can identify occurrence of gas entrainment.

Experiment on gas entrainment evaluation method from free liquid surface in a sodium-cooled fast reactor, 2; Measurement of gas core length by dynamic image processing

Endo, Kazuki*; Kobayashi, Shunsuke*; Jasmine, H.*; Sakai, Takaaki*; Matsushita, Kentaro; Ezure, Toshiki; Tanaka, Masaaki

Dai-27-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (Internet), 5 Pages, 2023/09

https://doi.org/10.1299/jsmepes.2023.27.A124

Assuming gas entrainment (GE) to the main coolant circulation system from cover gas which is an inert gas to cover sodium coolant in a reactor vessel of the sodium cooled fast reactor, there is a concern that reactivity disturbance will occur when bubbles pass through the reactor core. Conventionally, an evaluation method based on static vortex extension theory has been employed for the GE prediction. However, it is known that the method gives rather overestimation for the GE occurrence from the unsteady traveling vortex dimple at the wide liquid surface. In order to contribute to understand the phenomena, experimental data have been accumulated by the basic water experiment. In this study, measurement was performed for the length of a gas cores that grew while moving on the free liquid surface by dynamic image processing, and the types of the GEs and the occurrence conditions were evaluated.

Transient behavior of multi-dimensional core cooling by D-DHX in sodium-cooled fast reactors

Ezure, Toshiki; Akimoto, Yuta; Onojima, Takamitsu; Kurihara, Akikazu; Tanaka, Masaaki

Proceedings of 20th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-20) (Internet), p.3652 - 3662, 2023/08

In order to grasp the thermal-hydraulic behaviors during decay heat removal by dipped-direct heat exchangers (D-DHXs) in a sodium-cooled fast reactor, an experimental study was performed using a sodium experimental facility. The simulated core of PLANDTL-2 was formed by 55 hexagonal-shaped flow channel tubes, which allows to examine the cooling behavior of the simulated core region having multiple rows of fuel assemblies including the thermal hydraulic behavior to the radial direction. In this study, transient core cooling behavior in the situation after the scram with the decay heat removal using a D-DHX was examined. The time evolution of the temperature was measured in the whole system including the simulated core region. As the results, it was confirmed there was not excessively skewed temperature distribution in the radial direction in the core region.