Neutronics Analysis of Pressurized Water Reactor Cores using IDEA Code [version 1; peer review: 2 approved] - Ubaid ur Rehman; Zahid Rasool; M. Jawad Hussain; Ali Mansoor; Farhana Kausar; Haseeb ur Rehman; Aman ur Rehman (Nuclear Science and Technology Open Research)
Background This paper discusses the development and benchmarking of IDEA (Indigenous Diffusion Equation Analyzer), an in-house transient, multi-group, 3-dimensional neutron diffusion equation (NDE) code. IDEA can perform both steady-state criticality calculations and reactivity transient analysis for 3-dimensional rectangular geometries, with potential extension to cylindrical geometries. Methods The IDEA code employs the finite difference method (FDM) for spatial and temporal discretization, enhanced by source extrapolation-based acceleration. Its performance was evaluated by comparing its results with those of established NDE-based codes, PRIDE and CITATION. Various 2D and 3D steady-state and transient reactivity benchmark problems were used for this comparison. Results The results obtained from IDEA were found to be in close agreement with reference values from publicly available literature. In addition, the IDEA code demonstrated superior accuracy and consistency compared to PRIDE and CITATION in both steady-state and transient reactivity analyses. A neutronic analysis of a typical small PWR core under clean and cold conditions also showed validation against reference results. Conclusions The IDEA code has shown promising figures of merit to be used for practical neutronic calculations. Moreover, a multitude of users can benefit from its further development due to ease of access and lack of proprietary restrictions.