Comparison of Neutron-Physical Parameters of EPR and WWER-1500 using MCNP Computer Code. Issue 1 (1st January 2023)

• Record Type: Journal Article
• Authors: Dimitrov, Dobromir; Belousov, Sergey
• Is Part Of: IOP conference series. Volume 1128:Issue 1(2023)
• Publication Date: 2023-01-01
• Subjects: Earth sciences -- Periodicals; Environmental sciences -- Congresses; Environmental sciences -- Periodicals; 550.5
• DOI: 10.1088/1755-1315/1128/1/012023
• Languages: English
• ISSNs: 1755-1307

Abstract

Abstract: Three-dimensional calculations of the neutron-physical parameters were performed with the MCNP computer code for two models of a reactor core. The purpose was to determine the coefficient of reproduction of neutrons keff in each of the defined cases, as well as to qualitatively illustrate the neutron flux in axial and radial directions of the reactor core. A generalized analysis of the obtained results is performed, and it can be said that the prepared models, being relatively simple, meet the goal of this paper, namely presenting the basic principles of performing criticality calculations with the MCNP computer code. All calculations for this paper were performed with MCNP5, ver. 1.4. In the present paper, two active core models are considered – one for the EPR reactor and the other for the WWER-1500. Both projects are categorized as Generation 3 nuclear reactors. The following components of the core are included in both models: fuel elements, control rods and coolant. The fuel elements are presented as solid uranium dioxide cylinders coated with zirconium alloy. The fuel elements are grouped in fuel assemblies, in the case of the EPR reactor the fuel elements are arranged in a square grid, and in the case of the WWER-1500 reactor – in a triangular grid (hexagonal geometry). The limits of both models in the radial-azimuthal direction coincide with the inner diameter of the reactor vessel, and in the axial direction the limit is 56 cm in both directions from the end of the fuel elements. The calculations performed in this paper do not aim to quantify the distribution of neutron flux in the core, nor to find with great accuracy the coefficient of reproduction of neutrons keff at different stages of the reactor campaign. The calculations are made in order to verify the accuracy of the compiled input files. A criterion for this could be the demonstration of basic neutron-physical parameters, such as an increase in the value of keff with increasing the fuel enrichment and qualitative illustration of the change in the distribution of the neutron flux when introducing control rods in the core.