Validation of dynamic analysis code for high temperature gas-cooled reactor connecting hydrogen production system with a heat removal fluctuation mitigation system

Hasegawa, Takeshi; Aoki, Takeshi; Shimizu, Atsushi; Sato, Hiroyuki

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

Japan Atomic Energy Agency (JAEA) is planning a nuclear hydrogen production demonstration test by connecting a hydrogen production to the HTTR (High Temperature engineering Test Reactor), a 30MWt High Temperature Gas-cooled Reactor (HTGR) test reactor located at JAEA's Oarai Research Institute. The objective of the test is to establish coupling technologies between HTGR and a hydrogen production plant. One of key technologies to be developed is a thermal absorber consists of a steam generator and a radiator that can mitigate rapid fluctuations in helium temperatures during loss of heat load in the hydrogen production plant. The present study aims to develop an evaluation model for the absorber system that can be used for safety analyses of the HTTR for coupling a hydrogen production plant. A set of closed loop model is newly introduced to a RELAP5-based system analysis code to enable evaluation of steam-water natural circulation behavior between the steam generator heated by helium and the radiator cooled by ambient air. Experimental data obtained from a mock-up test facility of thermal absorber system is compared with calculated results of the developed model. As a results of the comparative study, it was confirmed that numerical analysis results were agreed well with experimental results. In conclusion, the proposed evaluation model is applicable for the safety analysis of a thermal absorber system to be deployed in HTGR hydrogen production systems.

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 (NUTHOS-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.

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 (NUTHOS-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 (NUTHOS-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 (NUTHOS-14) (Internet), 12 Pages, 2024/08