Page 219 - Thermal Hydraulics Aspects of Liquid Metal Cooled Nuclear Reactors

190 Thermal Hydraulics Aspects of Liquid Metal Cooled Nuclear Reactors

In water-cooled reactors, boiling crisis is the limiting phenomenon leading to an unex-
pected high temperature of cladding and fuel. Therefore, the ratio of the critical heat
flux to the real heat flux is the main design criterion. In LMRs, the coolant is in single-
phase flow (liquid-phase) conditions, and no boiling is allowed at both the normal
operating conditions and the accident conditions. Therefore, single-phase heat transfer
determines the cladding temperature:
q
T clad ¼ T coolant + (5.1)
α

where q stands for heat flux and α for heat-transfer coefficient.
The maximum cladding surface temperature affects directly the mechanical prop-
erties (stress limit) and the corrosion behavior of cladding. Therefore, one of the key
thermal-hydraulic tasks for the LMR design is the reliable prediction of the heat-
transfer coefficient. More discussion about this issue will be given in the next
subchapter.
The main tasks of reactor thermal-hydraulic analysis consist of
– behavior of flow and heat transfer in coolant, heat transport in fuel pin, and interaction
between coolant and fuel pin at normal and accident conditions;
– design analysis and design optimization of fuel assemblies and reactor cores at normal oper-
ating conditions and determination of operating parameters and safety margins;
– assessment of the transient response of the reactor core to postulated accident conditions,
evaluation of safety features, and optimization of safety systems.
Three different types of numerical approaches can be applied to analyze the thermal-
hydraulic behavior in reactor cores and/or in fuel assemblies. According to the indi-
vidual requirements, the domain of the thermal-hydraulic analysis may have different
scales, that is, entire reactor core, fuel assembly, and subchannel, as indicated in
Fig. 5.5.
Accordingly, the three different kinds of numerical approaches are system thermal
hydraulics (STH), SCTH, and three-dimensional computational fluid dynamics
(CFD). The system thermal hydraulic (STH) programs are one-dimensional numerical

D

P

(A) (B) (C)

Fig. 5.5 Geometric scales and numerical analysis approaches. (A) Reactor core (Huber, 2017);
(B) fuel assembly; (C) subchannel.