Neutronics and thermal hydraulics analysis of a small modular reactor

Highlights: • Core physics analysis of a small modular reactor design using MCNPX code. • Neutronics (core cycle burn-up) and thermal hydraulics (single channel) analyses are the focus of the study. • Reactor safety coefficients analyzed. • Rod cluster control assembly analyses carried out with different neutron absorbing materials. - Abstract: The small modular reactor (SMR) offers many feasible pathways for the construction of more nuclear power plants. A physics model of a near term deployable SMR of the integral pressurized water reactor (IPWR) design is developed. Fuel depletion simulations are performed to optimize the active fuel length, fuel enrichment and core loading pattern in order to achieve a uniform core power distribution. The optimized core can produce 500 MW of thermal power with a four year core life-time at a capacity factor of 87%. The core consists of 69 uranium dioxide (UO₂) fuel assemblies; 5 assemblies at 4.4 at% ²³⁵U enrichment and 64 assemblies at 4.95 at% ²³⁵U enrichment. The active fuel length is 200 cm and the core diameter is 194.55 cm for an active core height-to-diameter ratio of 1.03. As part of the study the active fuel length is increased to 240 cm resulting in an increased capacity factor of 95% at 530 MW of thermal power output for an active core height-to-diameter ratio of 1.23. Rod cluster control assemblies (RCCAs) are placed strategically to reduce the overall core power peaking factor to 1.3. Estimated reactor kinetics parameters such as the delayed neutron fraction and mean neutron generation time are typical of existing larger pressurized water reactors (PWRs) from which much of the IPWR based SMR design is derived. This study showed that Doppler, moderator temperature, void and power reactivity coefficients are all negative over the core life-time of four years indicating the possibility of safe reactor operation. A semi-analytical thermal hydraulics analysis reveals acceptable radial and axial fuel element temperature profiles with significant safety margin from industry standards on peak fuel and clad surface temperature limits. The critical heat flux (CHF) is calculated and is not exceeded even in 10% overpower conditions. In addition the nucleate boiling ratio (DNBR) is calculated and found to be above 4.8 for the entirety of the active core region. These parameters further engender confidence in the safety of the SMR design.