ANTEO+: A subchannel code for thermal-hydraulic analysis of liquid metal cooled systems

Liquid metal cooled fast reactor thermal hydraulic research development: A review

The growing interest in fast reactors demands further innovative technologies to enhance their safety and reliability. Understanding thermal hydraulic activities required for advanced reactor technology in design and development is key. However, knowledge of Heavy Liquid Metal (HLM) coolants technology is not mature. The liquid metal-cooled facilities are required experimental platforms for studying HLM technology. As such, efficient thermal hydraulic experimental result is important in the accurate validation of numerical results. In this vein, there is a need to closely review existing thermo-hydraulic studies in HLM test facilities and the test sections. This review aims to assess existing Lead-cooled Fast Reactor (LFR) research facilities, numerical and validation works and Liquid Metal-cooled Fast Reactor (LMFR) databases around the world in the last two decades. Thus, recent thermal hydraulic research studies on experimental facilities and numerical research that support the design and development of LFRs are discussed. This review paper highlights thermal hydraulic issues and developmental objectives of HLM, briefly describes experimental facilities, experimental campaigns and numerical activities, and identifies research key findings, achievements and future research direction in HLM cooled reactors. This review will enhance knowledge and improve advanced nuclear reactor technology that ensures a sustainable, secure, clean and safe energy future.

Validation of the COTENP Code: A Steady-State Thermal-Hydraulic Analysis Code for Nuclear Reactors with Plate Type Fuel Assemblies

This article presents the validation of the Code for Thermal-hydraulic Evaluation of Nuclear Reactors with Plate Type Fuels (COTENP), a subchannel code which performs steady-state thermal-hydraulic analysis of nuclear reactors with plate type fuel assemblies operating with the coolant at low pressure levels. The code is suitable for design analysis of research, test, and multipurpose reactors. To solve the conservation equations for mass, momentum, and energy, we adopt the subchannel and control volume methods based on fuel assembly geometric data and thermal-hydraulic conditions. We consider the chain or cascade method in two steps to facilitate the analysis of whole core. In the first step, we divide the core into channels with dimensions equivalent to that of the fuel assembly and identify the assembly with largest enthalpy rise as the hot assembly. In the second step, we divide the hot fuel assembly into subchannels with size equivalent to one actual coolant channel and similarly identify the hot subchannel. The code utilizes the homogenous equilibrium model for two-phase flow treatment and the balanced drop pressure approach for the flow rate determination. The code results include detailed information such as core pressure drop, mass flow rate distribution, coolant, cladding and centerline fuel temperatures, coolant quality, local heat flux, and results regarding onset of nucleate boiling and departure of nucleate boiling. To validate the COTENP code, we considered experimental data from the Brazilian IEA-R1 research reactor and calculated data from the Chinese CARR multipurpose reactor. The mean relative discrepancies for the coolant distribution were below 5%, for the coolant velocity were 1.5%, and for the pressure drop were below 10.7%. The latter discrepancy can be partially justified due to lack of information to adequately model the IEA-R1 experiment and CARR reactor. The results show that the COTENP code is sufficiently accurate to perform steady-state thermal-hydraulic design analyses for reactors with plate type fuel assemblies.