Neutronic study of a new generation of the small modular pressurized water reactor using Monte-Carlo simulation

Highlights: • Neutronic study of a typical small modular pressurized water reactor is done. • Monte-Carlo simulation is conducted by using MCNPX code. • The neutron flux distribution, radial and axial power peaking are investigated. • The influence of the control rods on the thermal flux is studied. • The geometry and neutronic parameters are visualized via qualified 3-D figures. - Abstract: Small Modular Reactors (SMRs) are the innovative design of nuclear reactors making remarkable interest during recent years. Since there is not enough available operating experience on SMRs, it might be possible to initiate extensive investigations on these types of reactors for the purpose of improving the current performance level of these systems, significantly. The main purpose of this present study is neutronic study of a typical small modular pressurized water reactor via Monte-Carlo method using the MCNPX code. The CAREM25 is chosen as the reference SMR. The reactor core geometry is simulated and neutronic parameters are visualized and analyzed via high qualified 3-D figures. They are figure out neutronic nature of the chosen case study. They capture the simulation of the reactor core geometry, and skim the neutron flux and power distribution, radial and axial power peaking factors and the influence of the control rods on the thermal flux. Central fuel assembly is determined as the hottest fuel assembly with power peaking factor of 1.778. The hottest fuel rod power peaking factor is calculated as 1.846 in the hottest assembly. The maximum calculated axial power peaking factor of the hot rod is 2.85. Results show that the maximum axial power along the fuel rods occurred below the mid-plane of the rod. The ratio of the hot to average rod axial power peaking factor as a safety parameter used to calculate the maximum heat flux in the hottest channel, is calculated close to 2 in almost 70% of the core height. The core reactivity at cold and hot shut down without safety injections of boron acid is calculated as $\text{0}\text{.}09712\left(\frac{\Delta k}{k}\right)$ and $-\text{0}\text{.}00103\left(\frac{\Delta k}{k}\right)$, respectively. Results show suitable neutronic behavior and responses during virtual tests and analyses.