Neutronic evaluation of a PWR with fully ceramic microencapsulated fuel. Part II: Nodal core calculations and preliminary study of thermal hydraulic feedback

Abstract The fully ceramic microencapsulated (FCM) fuel is based on the tri-isotropic (TRISO) carbon coated fuel particles. These particles were developed and demonstrated for use in high temperature gas reactors. It has been proposed to use these particles in light water reactors to provide potential operational and safety benefits. The reference fuel in this case assumes TRISO-like particles with a ∼20%-enriched uranium-nitride kernel embedded in a silicon carbide (SiC) matrix. The fuel particles are contained in a “compact” which is then inserted into a cladding. The fuel assembly features the same dimensions as a standard 17 × 17 Westinghouse fuel assembly. FCM fuel requires fission products to traverse several barriers in the proposed fuel design before reaching the cladding. FCM fuel may also reduce fuel-cladding interaction and fuel pellet swelling while enabling higher fuel burn-up. This study is a neutronic evaluation of the use of FCM fuel in an advanced pressurized water reactor (PWR). On the lattice level, the SERPENT Monte Carlo and TRITON deterministic tools were used, while the whole core simulation was based on the three-dimensional PARCS nodal code. The present paper focuses on two of the issues associated with this proposed implementation: specifically the development of a reasonable reference full-core model of an advanced PWR with FCM fuel and the response of the PWR to a reactivity insertion accident (RIA). This work addresses the issues of the increased power density and transients that occur on short time-scales in a PWR. In this case, the RIA takes the form of a control rod ejection for a typical PWR reactor. This results in a sudden increase in power and a corresponding increase in fuel kernel temperature. In the case of a PWR, this response is more demanding than in the case of a gas-cooled reactor, because the kinetic parameters and feedback coefficients of the two reactors are quite different. The parameters for the fuel and matrix material in the PARCS thermal–hydraulic module were modified to reflect the different geometry and materials. Preliminary data for both un-irradiated and irradiated SiC were obtained from the literature and included in the analyses. A super prompt critical RIA produces an average energy deposition (