COOLING ALFRED | SMRs AND ADVANCED REACTORS
150
100 50 0
-50
-100 -150
-150 -100 -50 0 50
the mathematical analysis and presentation of its thermal- hydraulics model difficult considering that there are no thermal-hydraulics properties and features of lead metal as a coolant in the available nuclear thermal-hydraulics codes such as COBRA-EN or RELAP5. A thermal-hydraulics program was thus developed for this class of reactors. The process of implementing the plan began with the
detailed specifications of the core geometry, thermal structures, material characteristics, geometry and type of fuel rod support networks (grid spacer) and finally the boundary conditions and characteristics of the flow entering the core are defined and determined. Then, according to the flow pattern entering the core, by using nuclear engineering codes and by writing a thermal- hydraulics program the analysis of the cooling behaviour in the reactor core was conducted and finally the thermal- hydraulics parameters are explored. Using the MCNPX code, neutron parameters such as the
effective multiplication factor, fast and thermal neutron flux distribution and changes in the thermal-hydraulics parameters of the reactor core, such as pressure, enthalpy, and mass quality of the coolant in the advanced ALFRED reactor, are calculated during stable working conditions. To combine the neutron codes with the thermal-
hydraulics model, at first each of the fuel assemblies in the core of the reactor are modelled using the MCNPX code. Each fuel assembly was divided into 10 parts in the axial direction and after neutronic calculations in MCNPX code radial and axial power distribution values in different parts of each fuel assembly are found. Also, the value of the effective multiplication factor of the core is also determined at this stage. These outputs should include the average power of the fuel assembly and the axial power distribution. After obtaining the power values in each part of the reactor core, these values are used as the input of the thermal-hydraulics code and used to calculate the temperatures of the fuel, coolant, cladding and also the density of the coolant fluid on the surface of the reactor core.
Thermal-hydraulics calculations are performed only for one fuel assembly at each stage, so separate axial power distribution files must be created for all assemblies using the neutronic module.
Coupling thermal hydraulics with neutronics By calculating the neutron parameters according to the characteristics of the Alfred reactor core and calculating the output parameters according to the thermal-hydraulics parameters and the characteristics of the reactor fuel assembly, it is possible to obtain the coupling for the fuel assembly. In this research, the integration of neutron codes and Thermal-Hydraulics model in the hot fuel assembly of ALFRED reactor was explored. The results show that thermal-hydraulics parameters such as fuel, coolant temperatures, and coolant density affect neutron parameters such as power distribution and reactor critical conditions. By changing each of the thermal-hydraulics parameters, the governing principles of neutron reactions also undergo changes. The opposite is also true and by changing each of the neutronic parameters, the thermal- hydraulics parameters also change. For example, by changing the temperature of the coolant, its density also changes, and this process, in turn, causes a change in the fluid's deceleration, thus changing the flux and power distribution in the reactor core. ■
100 150
x10-4
1.2 1
0.8 0.6 0.4 0.2 0
Above: The core geometry of the ALFRED reactor and a view of the radial flux distribution
Table 1: Main characteristics of ALFRED reactor Values 300 60 2
Unit MW
cm 4.5
0.15 0.6
10.5 4 5
~1.4
13.86 127 165
mm mm mm mm mm mm mm
1-ms mm -
cm
Parameter
Thermal power Active height
Pellet hollow diameter Pellet radius Gap thickness
Cladding thickness Pin diameter
Wrapper thickness
Distance between 2 wrappers Coolant velocity
Lattice pitch (hexagonal) Pins per FA
Inner vessel radius
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