Study on eutectic melting behavior of control rod materials in severe accidents of sodium-cooled fast reactors, 2; Modeling of multi-phase eutectic reaction behavior

Morita, Koji*; Yamano, Hidemasa

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

This paper describes the generalized model developed for these eutectic reactions between boron carbide (BC) and stainless steel (SS) as well as for the reactions that occur between eutectic reaction products in the solid and liquid states and SS or BC. We also describe the thermophysical property model based on thermophysical property data.

Evaluation of interface capturing schemes of VOF method through application to bubble flows with single orifice

Fukuda, Takanari; Yamashita, Susumu; Yoshida, Hiroyuki

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

The VOF method is a type of CFDs that is most widely applied to multiphase flow analysis involving advective interfaces, and several interface-capturing schemes have been developed for an accurate advection of VOF values. However, the performance of these schemes has typically been evaluated only for limited numerical problems where velocity fields are spatially orderly and fixed in time. Few studies have been conducted to evaluate the performance of these schemes for more realistic and complex conditions, such as gas-liquid two-phase flows in nuclear reactors. Therefore, in this study, three-dimensional analysis of bubble flows has been conducted using the interface-capturing schemes of THINC and THINC/WLIC, which have been developed relatively recently. Evaluation is performed using more engineering indicators such as the number, volume, and trajectory of bubbles, which can influence the void fraction distribution in reactor cores. The results of these comparisons showed that the VOF value could be significantly diffused, leading to numerical brake-up and dissipation of the bubbles, with the influence of interface-capturing scheme.

Application of the GIF safety design criteria and safety design guidelines on natural circulation capability to next generation sodium-cooled fast reactor in Japan

Yamano, Hidemasa; Futagami, Satoshi; Higurashi, Koichi*

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

This paper describes the application of safety design criteria (SDC) and safety design guidelines (SDG) developed in the Generation-IV International Forum on the natural circulation of sodium to sodium-cooled fast reactors (SFRs) recently designed in Japan.

Fukushima Daiichi Nuclear Power Plant accident analysis considering the thermal stratification and containment leakage

Nakamura, Yuki*; Kojima, Yoshihiro*; Yamashita, Takuya; Shimomura, Kenta; Mizokami, Shinya

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

Development of reactor vessel thermal-hydraulic analysis method in natural circulation conditions with coarse-mesh subchannel CFD model

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

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

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, development of the reactor vessel model including the more model by using a computational fluid dynamics code (RV-CFD) is required. Previously, the CFD model based on the subchannel analysis was developed. In this study, to achieve much lower computational cost maintaining the prediction accuracy, the coarse-mesh subchannel CFD (CMSC) model has been 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 radial heat transfer under NC conditions.

Study on eutectic melting behavior of control rod materials in core disruptive accidents of sodium-cooled fast reactors, 1; Project overview and progress until 2020

Yamano, Hidemasa; Takai, Toshihide; Emura, Yuki; Fukuyama, Hiroyuki*; Higashi, Hideo*; Nishi, Tsuyoshi*; Ota, Hiromichi*; Morita, Koji*; Nakamura, Kinya*; Fukai, Hirofumi*; et al.

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

This paper describes the project overview and progress of experimental and analytical studies conducted until 2020. Specific results in this paper are the measurement of the eutectic reaction rates and the validation of physical model describing the eutectic reaction in the analysis code through the numerical analysis of the BC-SS eutectic reaction rate experiments in which a BC pellet was placed in a SS crucible.

Development of the simplified boiling model applied to the large-scale detailed simulation

Ono, Ayako; Yamashita, Susumu; Sakashita, Hiroto*; Suzuki, Takayuki*; Yoshida, Hiroyuki

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

Japan Atomic Energy Agency is developing the computational fluid dynamics code, JUPITER, based on the volume of fluid (VOF) method to analyze detailed thermal-hydraulics in a reactor. The detailed numerical simulation of boiling from a heating surface needs a substantial computational cost to resolve the microscale thermal-hydraulic phenomena such as the bubble generation from a cavity and evaporation of a micro-layer. This study developed the simplified boiling model from the heating surface to reduce the computational cost, which will apply to the detailed simulation code based on the surface tracking method such as JUPITER. We applied the simplified boiling model to JUPITER, and compared the simulation results with the experimental data of the vertical heating surface in the forced convection. We confirmed the degree of their reproducibility, and the issues to be modified were extracted.

Development of numerical simulation method to evaluate molten material behaviors in nuclear reactors

Yamashita, Susumu; Yoshida, Hiroyuki

Proceedings of 12th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, Operation and Safety (NUTHOS-12) (USB Flash Drive), 12 Pages, 2018/10

In order to simulate the relocation phenomena phenomenologically around the reactor core and inside pedestal without assumptions, inputted information, empirical knowledge and given scenarios, a numerical simulation code based on computational fluid dynamics, JUPITER, that can phenomenologically evaluate the melting phenomena has been developed in JAEA. We confirm the applicability of JUPITER to the corium spreading process inside the pedestal by simulating corium spreading behaviors and its distributions under several parameters such as the corium inflow condition. And we investigated detailed fuel debris distribution inside the pedestal and the cavity named sump pit located on the lower part of PCVs, that is distribution for each component of core materials such as stainless steel (SUS), bron carbide (BC), zircaloy (Zry) and uranium dioxide (UO). As a result, since JUPITER uses interface capturing scheme which can treat complicated behavior of interfaces such as a large deformation and a complicated separation among interfaces, the corium was spread with complicated mixing behavior and the distribution of fuel debris tend to accumulate complicatedly inside sump pits. Since existing SA analysis codes difficult to treat such complicated phenomena and the complicated fuel distribution, those result might be contributed to an understanding of circumstances inside PCV and, ultimately, also contributed to reactor decommissioning process.

Numerical simulation of natural circulation experiment under asymmetric cooldown using LSTF

Ishigaki, Masahiro; Watanabe, Tadashi*

Proceedings of 12th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, Operation and Safety (NUTHOS-12) (USB Flash Drive), 10 Pages, 2018/10

When coolant in one of the secondary side of steam generator (SG) is lost under some accident condition, the NC in the loop with the affected SG may terminate. Hence, the experiment was done in order to discuss the behavior of the natural circulation flow when the secondary side of the intact SG was depressurized stepwisely and that of the affected SG was empty of coolant. In this paper, we analyzed this NC experiment using the LSTF by the TRACE code. The objective of this analysis is to clarify the sensitivity of the code to the NC behavior. The calculated mass flow rate in the intact loop was slightly underestimated compared with the experimental result. On the other hand, the calculated mass flow rate in the affected loop was overestimated compared with the experimental result. In addition, we did the sensitivity analysis of the NC behavior in the case that the cooldown rate was changed.

Analyses of LSTF experiment and PWR plant for 5% cold-leg break loss of coolant accident

Watanabe, Tadashi*; Ishigaki, Masahiro*; Katsuyama, Jinya

Proceedings of 12th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, Operation and Safety (NUTHOS-12) (USB Flash Drive), 9 Pages, 2018/10

The analyses of LSTF experiment and PWR plant for 5% cold-leg break LOCA are performed using the RELAP5/MOD3.3 code. The discharge coefficient of critical flow model is determined so as to obtain the agreement of pressure transient between the LSTF experiment and the experimental analysis, and used for the PWR analysis. The characteristics of thermal-hydraulic phenomena in the experiment are shown to be simulated well by the two analyses. The decrease in core differential pressure during the loop-seal clearing is, however, underestimated by the two analyses, and the core heat up is not predicted. The loop flow rates are also underestimated by the two analyses. Although the duration of core heat up during the boil-off period is longer in the experimental analysis, the results of two analyses agree well, and the effect of scaling is found to be small between the experimental analysis and the PWR analysis.