Preconceptual design of a fluoride high temperature salt-cooled engineering demonstration reactor: Motivation and overview

A. Louis Qualls, Benjamin R. Betzler, Nicholas R. Brown, Juan J. Carbajo, M. Scott Greenwood, Richard Hale, Thomas J. Harrison, Jeffrey J. Powers, Kevin R. Robb, Jerry Terrell, Aaron J. Wysocki, Jess C. Gehin, Andrew Worrall

    Research output: Contribution to journalArticlepeer-review

    26 Scopus citations


    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 and the Experimental Breeder Reactor-II for sodium-cooled fast reactors. Engineering demonstrations have historically played a vital role in advancing the technology readiness level of reactor concepts. 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. It would use tristructural-isotropic (TRISO) particle fuel in compacts within prismatic graphite blocks. FLiBe (2 7LiF-BeF2) is the reference primary coolant. The FHR DR is designed to be small, simple, and affordable. Development of the FHR DR is an intermediate step to enable near-term commercial FHRs. The design philosophy of the FHR DR was focused on safety, near-term deployment, and flexibility. Lower risk technologies are purposely included in the initial FHR DR design to ensure that the reactor can be built, licensed, and operated as an engineering demonstration with minimal risk and cost. These technologies include TRISO particle fuel, replaceable core structures, and consistent structural material selection for core structures and the primary and intermediate loops, and tube-and-shell primary-to-intermediate heat exchangers. Important capabilities to be demonstrated by building and operating the FHR DR include: • core design methodologies,• heat exchanger performance (including passive decay heat removal),• pump performance,• reactivity control,• salt chemistry control to maximize plant life,• salt procurement, handling, maintenance and ultimate disposal, and• tritium management. Non-nuclear separate and integral test efforts (e.g., heated salt loops or loops using simulant fluids) are necessary to develop the technologies that will be demonstrated in the FHR DR.

    Original languageEnglish (US)
    Pages (from-to)144-155
    Number of pages12
    JournalAnnals of Nuclear Energy
    StatePublished - Sep 1 2017

    All Science Journal Classification (ASJC) codes

    • Nuclear Energy and Engineering


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