The economic expenditure required to generate electricity per kilowatt of newly constructed reactor capacity is notably higher than the marginal cost incurred by augmenting operational power and prolonging operational duration. Scholars have investigated the potential utilization of uranium zirconium alloy fuel, particularly U-50Zr, within PWR. The material properties of uranium zirconium alloy exhibit distinct advantages, including a higher solidus temperature, enhanced irradiation stability, U-50Zr has reduced swelling and a decreased fission gas release rate compared to U-10Zr. In addition, uranium zirconium alloy fuels can be processed into helical cruciform geometry, which can significantly enhance heat transfer efficiency. U-50Zr helical cruciform fuel (HCF) concepts exhibit a certain degree of innovation in both material composition and geometric configuration, and the ultimate application is expected to achieve significantly higher power density than traditional PWRs. The nuclear thermal coupling calculation program encompasses sophisticated reactor physics computations, intricate thermal conductivity assessments, and interactive evaluations among various physical parameters, necessitating extensive programming expertise, meticulous algorithm design, and a profound understanding of nuclear-related fundamentals. For scholars and researchers in the domains of nuclear science and nuclear engineering, the endeavor of constructing three-dimensional coupling calculation programs from scratch poses a formidable challenge. Consequently, this article adopts an efficient and practical solution: using cardinal code combined with open source MOOSE framework and OpenMC code to perform neutronic-thermal coupling analyses on HCF. Currently, research endeavors focusing on HCF primarily involve the analysis of its internal mechanical behavior and the examination of single-phase and two-phase flow issues associated with the external coolant. Notably, the thermal source term utilized in this study diverges from the conventional cosine distribution employed in standard thermal analyses. The source term of the thermal conductivity equation within the fuel is determined by the elemental composition of the fuel rod and the progression of the fission reaction, with these factors being prescribed by the OpenMC code. The radial temperature non-uniformity distribution calculated in this article will contribute to a more accurate evaluation of coolant flow heat transfer and fuel mechanical properties.