In Nuclear Power Plants the Design Extension Conditions are more complex and severe than those postulated as Design Basis Accidents, therefore, they must be taken into account in the safety analyses. In this study, many hypothetical investigated transients are applied on KONVOI pressurized water reactor during a 6-in. (182 cm2) cold leg Small Break Loss-of-Coolant-Accident to revise the effects of all safety systems ways through their availability/ nonavailability on the thermal hydraulic behaviour of the reactor. The investigated transients are represented through three cases of Small Break Loss-of-Coolant-Accident as, case-1, without scram and all of the safety systems are failure, case-2, the normal scram actuation with failure of all safety systems (nonavailability), and finally case 3, with normal actuation scram sequence and normal sequential actuation of all safety systems (availability). These three investigated transient cases are simulated by creation a model using Analysis of Thermal-Hydraulics of LEaks and Transient code. In all transient cases, all types of reactivity feedbacks, boron, moderator density, moderator temperature and fuel temperature are considered. The steady-state results are nearly in agreement with the plant parameters available in previous literatures. The results show the importance effects of the feedbacks reactivity at Loss-of-Coolant-Accident on the fallouts power, since they are considered the key parameters for controlling the clad and fuel temperatures to maintain them below their melting point. Moreover, the calculated results in all cases show that the thermal hydraulic parameters are in acceptable ranges and encounter the safety criterion during Loss-of-Coolant-Accident the Design Extension Conditions accidents processes. Furthermore, the results show that the core uncovers and fuel heat up do not occur in KONVOI pressurized water reactor in theses the Design Extension Conditions simulations since, all safety systems provide adequate core cooling by sufficient water inventory into the core to cover it.

A Super Critical Water-cooled Nuclear Reactor (SCWR) is a Generation IV concept currently being developed worldwide. Unique to this reactor type is the use of light-water coolant above its critical point. The number of SCWR components is reduced since steam separators and dryers are not required. This advantage drives down capital and maintenance costs. Safety is also increased, because a dry out phenomenon does not occur in SCW conditions; SCW remains in a single phase. A computer program by Engineering Equation Solver, (EES) has been produced for inquiry of the fuel, clad and coolant temperatures under supercritical conditions for supercritical water reactor powered by ThO 2 -UO 2 mixture as a fuel. In the calculation, uniform axial heat flux and average channel were considered. The bulk fluid, clad and fuel temperatures along fuel length were obtained for supercritical pressures 26, 30 and 40 MPa. Also, the UO 2 percentage added to ThO 2 was varied as, 4% and 10%. It was found that the maximum fuel temperature reached 1917 o C for a pressure of 26 MPa and 1896 o C for a pressure of 30 MPa in case of 4% UO 2 . However, the maximum temperature of the fuel was 1915 o C for a pressure of 26 MPa and 1894 o C for a pressure of 30 MPa in case of 10% UO 2 , which is surpasses the industry limit of 1850 o C.

The thermal-hydraulic behavior of the spent fuel in dry storage casks under forced convection mode is experimentally and numerically investigated. For this purpose, a test rig is designed and constructed to simulate the cooling loop cask. This test rig contains 21 spent fuel discharged from a pressurized water reactor (PWR). A numerical simulation is performed by ANSYS-CFX fluid dynamic code. The effect of decay heat generation and inlet air velocity are investigated. The results show that the increase in the inlet air velocity improves the coolability of the fuel, while the increase in decay heat leads to a decrease in the coolability of the fuel. Within the range (1.1< V < 2.8 m/s) for inlet air velocity and heaters power (630 < Q < 1260 watt), a new empirical correlation has been obtained for Nusselt number, Nu as a function in Reynolds number, Re. The comparisons between experimental and numerical results show a good agreement.