Some thermal hydraulic challenges in sodium cooled fast reactors

In this work, the heat transfer intensification in shell and helically coiled tube heat exchangers via installing annular fins on the outer surface of the helical coil, has been numerically investigated. Thirteen heat exchangers were designed for this purpose. All geometrical parameters of the heat exchangers are same except, fin's number or height. All of the heat exchangers have been studied at three different shell side Reynolds number (Re sh  = 7500, 15,000 and 30,000). In order to validation of the numerical model two method has been used. In the first method, the calculated heat transfer has been compared with the approximate method which is based on the experimental correlations for both coil and shell side Nusselt numbers and consideration of the E-NTU relations of these types of heat exchangers. In the second method, the heat transfer coefficient of the shell side has been compared with the experimental heat transfer coefficients of the previous works. An acceptable agreement has been observed in these comparisons. Furthermore, the optimum cases and some correlations have been obtained for prediction of the heat transfer coefficient of the shell side. Results indicate that, in the range of 7500 ≤ Re sh  ≤ 30,000, the heat transfer rate can increase up to 44.11%.

Numerical investigation of heat transfer intensification in shell and helically coiled finned tube heat exchangers and design optimization

Status of CATHARE code for sodium cooled fast reactors

The lead-based reactor is one of the fourth-generation nuclear energy systems with excellent development prospects. It has attracted much attention because of its unique advantages in fuel transmutation and breeding, passive safety, and economic efficiency. In this study, the main features of typical lead-based fast reactors (LFRs) worldwide are reviewed and current challenges in their development are pointed out. Moreover, the research progress in four main thermal-hydraulic aspects is overviewed and discussed, including the flow and heat transfer characteristics of the coolant, thermal-hydraulic analysis of the reactor core, lead pool and reactor system. These aspects are further divided into specific issues according to the developing needs of LFR. For each issue, up-to-date research findings, existing difficulties, and future trends are presented. This study can serve as a reference for the design and thermal-hydraulic analysis of LFRs.

Review of Thermal-Hydraulic Issues and Studies of Lead-based fast reactors

A multi layer core catcher concept for future sodium cooled fast reactors

Heat transfer in heat exchangers of sodium cooled fast reactor systems

Development of the plant dynamics analysis code NETFLOW

Design of a passive residual heat removal system for the FUJI-233Um molten salt reactor system

Analysis of thermal stratification in the upper plenum of the “Monju” reactor

International benchmark on the natural convection test in Phenix reactor

Hiroyasu Mochizuki

Papers

Heat transfer in heat exchangers of sodium cooled fast reactor systems

Development of the plant dynamics analysis code NETFLOW

Design of a passive residual heat removal system for the FUJI-233Um molten salt reactor system

Analysis of thermal stratification in the upper plenum of the “Monju” reactor

International benchmark on the natural convection test in Phenix reactor