Thermal-Hydraulics Operation Parameters Modeling and Analysis of KLT-40S Reactor at Steady-State and Transient Condition using RELAP5-3D

Abstract

One of Floating Nuclear Power Plant (FNPP) designs in the world is currently being built by Russian Federation, named “Academic Lomonosov,” which uses two PWR types, KLT-40S as its power unit. However, too little information regarding its detailed technical specification is available, including its thermal-hydraulics parameters. The objective of this research is to create a thermal-hydraulic model of KLT-40S reactor core use RELAP5-3D and to predict fuel and cladding temperature value at the steady-state condition, and transient condition with a variety of primary coolant mass flow rate and pressure to simulate abnormal event within the reactor. The reactor thermal-hydraulic model is created by dividing 121 coolant channels in the actual nuclear fuel assemblies into two channels: one channel to simulate coolant flow in 120 fuel assemblies with average heat generation, and the other channel to simulate coolant flow in one fuel assembly with highest heat generation in the core. The fuel structure had solid cylinder geometry and made from ceramic-metal UO2 dispersed in the inert silumin matrix. The fuel cladding is made from zirconium alloy. These fuel heat structures generate heat from fission reaction and are modelled as a heat source according to the reactor power technical data, i.e., 150 MWt. The reactor axial power distribution is approximated by cosine distribution. Operation parameter variation that represents the real reactor normal operation condition in this research is a variation that has flow loss coefficient value 8,000, radial power peaking factor 1.1, and axial power peaking factor 1.1 with axial power peaking located in the middle of the fuel rod. The fuel and cladding temperature value at the steady-state condition and several transient conditions are predicted in this research, and there is no temperature value that goes beyond the safety limit.