Above: Schematics of the ALFRED reactor

The plan to use advanced fast reactors with molten lead cooling in the form of fourth generation reactors was proposed in 2000.

In 2010, for the first time a plan to develop and build fast reactors with molten lead reactor was proposed and in 2012 the conceptual design of the ALFRED reactor was investigated based on the LEADER project. The ALFRED reactor and the ELFR design and different aspects of the ALFRED reactor design and its safety aspects were introduced in 2013 and in 2014 the concept design of rotary steam generators for the ALFRED reactor was reviewed. In the same year the dynamic behavior of the ALFRED reactor was simulated and investigated.

However, even though it has been several years since conceptual plans were proposed in the field of advanced ALFRED reactors, published calculations about this type of reactor are limited, especially simultaneous neutronic and thermal-hydraulic calculations.

In this research, the neutronic and thermal-hydraulic behaviour of an advanced reactor core with ALFRED lead coolant and neutron code couple and its thermal-hydraulics model are investigated and analysed. The purpose of performing neutron calculations in the core of a reactor is to calculate the neutron flux distribution in the core and calculate the effective multiplication factor. Due to the necessity of performing accurate neutron calculations, first of all, the real geometry of the core, the composition and richness of the fuel, the grid pitch, the radius and height of the fuel rods, the composition and location of the combustible absorbers, the types and location of the control rods, and the arrangement of the assembly are needed. The fuel in the heart of the reactor, the radial and axial reflectors, should be determined and specified.

Modelling ALFRED

MCNPX code is used for the neutron calculations using the Monte Carlo statistical method.

In thermal-hydraulics calculations, due to the heat transfer mechanism affecting the displacement between the fuel and the surrounding lead coolant, it actually makes 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.

Authors: Korosh Rahbari, Darush Masti, Kamran Sepanloo, and Ehsan Zarifi, Department of Nuclear Engineering, Bushehr Branch, Islamic Azad University, Bushehr, Iran, Reactor and nuclear safety school, Nuclear Science and Technology Research Institute (NSTRI) Iran and the Department of Nuclear Engineering, Science and Research Tehran Branch, Islamic Azad University