During a period of escalating energy crises, the development of nuclear energy is highly significant. In the reactor core, fluid flow is extremely complex, involving single-phase flow, subcooled boiling, and critical heat flux (CHF). Accurately predicting the heat transfer phenomena within fuel assemblies is crucial to understanding the core behavior of fuel assemblies, which ultimately improves CHF and enhances reactor safety. Current CHF prediction mainly focus on uniformly heated pipes; however, in reality, fuel assemblies have a non-uniform heat flux distribution. The differences in heating methods can impact research outcomes. Computational Fluid Dynamics (CFD) can be used to extract three-dimensional characteristics such as pressure, velocity, and temperature, providing theoretical guidance for studying the heat transfer characteristics of fuel assemblies. In this study the RPI boiling model combined with the Euler-Euler two-fluid model is employed to simulate subcooled boiling and CHF phenomena for a helical cruciform single-rod fuel under non-uniform heating conditions and compares it to a round single-rod fuel under the same conditions. The research results provide valuable insights for further understanding the heat transfer characteristics of helical cruciform fuel assemblies.