Engineering Papers⌕ Search

Engineering topics

Kim, Daejong

Publications and source records attributed to Kim, Daejong.

Progress in Fast Modular Reactor Conceptual Design

The Fast Modular Reactor (FMR) is a 100-MW(thermal) gas-cooled fast reactor being developed by General Atomics Electromagnetic System with the goal of developing a FMR for flexible and dispatchable power to the U.S. electricity market in the mid-2030s. The conceptual design aims to develop and verify simplified design features. These include an inert helium gas coolant, pellet-loaded fuel rods, installations with air cooling as ultimate heat sink, and small and passive heat removal systems. The goal is to ensure the development of a safe, maintainable, cost-effective, and distributed nuclear energy-generating station. The baseline technologies selected to achieve this goal are a helium coolant that is an inert gas with no chemical reaction with structural components, not activated, single phase, enabling high-temperature operation and a high thermal efficiency Brayton cycle; conventional uranium dioxide (UO 2 ) fuel, which is the most widely used and well-known fuel material, capable of high burnup (100 MWd/kg) and a long fuel life; and silicon carbide composite (SiGA®) cladding and internal structures that are chemically inert in the helium environment, exceptionally radiation tolerant, and being derisked by accident tolerant fuel technology development. Further, the reactor was specifically designed with passive safety features, including high-temperature in-core materials and a reactor vessel cooling system consisting of cooling panels of naturally circulating water. The passive safety of the core was confirmed for the depressurized loss-of–forced cooling accident, which showed the peak cladding temperature at ~1600°C during the transient, which is below the current design limit of 1800°C. The conceptual design of the FMR has been conducted for the reactor system, vessel system, generator and turbomachine, instrumentation and control, residual heat removal system, plant service system, and containment, as well as pre-application licensing documents.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Hermeticity of SiC/SiC composite and monolithic SiC tubes irradiated under radial high-heat flux

Demonstration of hermetic SiC fiber–reinforced SiC matrix composite cladding under normal operating environments has been identified as one of the most critical feasibility issues for accident-tolerant fuel cladding in light-water reactors. This study provides critical experimental data needed for understanding the effects of irradiation on hermeticity. SiC composite and monolithic tubes were neutron-irradiated to 2 displacements per atom with and without a nominal radial heat flux of 0.6 MW/m2 to produce a simulated in-pile stress state for the normal operation of a light-water reactor. The through-thickness temperature gradient under irradiation results in a gradient in swelling, which causes a significant stress buildup. Such irradiation-induced stress was modeled using a commercial finite element analysis code. The radial heat flux–irradiation synergism was experimentally investigated by constructing a special irradiation capsule and evaluating the helium hermeticity of the specimens. The simulated stress state exhibited a near equi-biaxial tensile axial and hoop stress of ~150 MPa at the inner surface of the SiC composite tube. This stress level is potentially beyond the matrix cracking stress. Degradation of hermeticity of the SiC composite tubes was observed after irradiation, indicating irradiation-induced cracking, whereas the irradiated monolithic SiC tubes remained hermetic. In conclusion, the results indicate that loss of hermeticity caused by radiation-induced microcracking is a potential issue for SiC composite cladding, depending on the magnitude of the temperature gradients. Coating the outer surface of the cladding was identified as a mitigation strategy that might overcome the cracking issue.

36 MATERIALS SCIENCE↗