Preconceptual design of a fluoride high temperature salt-cooled engineering demonstration reactor: Core design and safety analysis. (May 2017) - Brown, Nicholas R.; Betzler, Benjamin R.; Carbajo, Juan J.; Wysocki, Aaron J.; Greenwood, M. Scott; Gentry, Cole; Qualls, A. Louis
Highlights: Engineering demonstration reactors are built as proof of concept for technology options. This paper details a preconcept for a fluoride salt-cooled engineering demonstration reactor (FHR DR). Core design, fuel cycle, and safety of the FHR DR preconcept have been evaluated. Preliminary safety analysis indicates that the reactor could be used to demonstrate inherent safety potential of FHRs. Abstract: Engineering demonstration reactors are nuclear reactors built to establish proof of concept for technology options that have never been built. Examples of engineering demonstration reactors include Peach Bottom 1 for high temperature gas-cooled reactors (HTGRs) and Experimental Breeder Reactor II (EBR-II) for sodium-cooled fast reactors. Engineering demonstrations have historically played a vital role in advancing the technology readiness level of reactor technologies. This paper details a preconceptual design for a fluoride salt-cooled engineering demonstration reactor. The fluoride salt-cooled high-temperature reactor (FHR) demonstration reactor (DR) is a concept for a salt-cooled reactor with 100 megawatts of thermal output (MWt). It would use tristructural-isotropic (TRISO) particle fuel within prismatic graphite blocks. FLiBe (2 7 LiF-BeF2 ) is the reference primary coolant. The FHR DR is designed to be small, simple, and affordable. Core design characteristics, fuel cycle performance, and safety analysis of the FHR DR preconcept have been evaluated. The FHR DR core design features a negative or negligible void coefficient throughout a reactor operating cycle. Both single-batch (cartridge) and multiple-batch fuel cycles can be demonstrated in the FHR DR. The single-batch cycle length of the FHR DR core is estimated at between 12 and 18 months, assuming the successful qualification of composite carbon (C/C) or silicon carbide (SiC/SiC) structural fuel block tie rod material. Fuel cycle performance of the FHR DR is similar to a high temperature gas-cooled reactor. Preliminary safety analysis of the FHR DR indicates that the reactor could be used to demonstrate the inherent safety characteristics of FHR designs.