Fully ceramic microencapsulated fuel in prismatic high temperature gas-cooled reactors: Analysis of reactor performance and safety characteristics

Cihang Lu, Briana D. Hiscox, Kurt A. Terrani, Nicholas R. Brown

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    30 Scopus citations

    Abstract

    Advanced nuclear reactor technologies have the potential to expand the missions of nuclear energy while reducing carbon emissions. This paper presents scoping reactor physics and thermal hydraulics analysis of a high temperature gas-cooled reactor (HTGR) using the fully ceramic microencapsulated (FCM) fuel form, and demonstrates the feasibility of FCM fueled HTGRs. FCM fuel consists of tristructural isotropic (TRISO) coated fuel particles embedded in a matrix of silicon carbide (SiC). The potential advantages of FCM fuel, which uses a monolithic SiC matrix, over conventional HTGR fuels with a carbon-based matrix include: a long refueling interval; high stability of the SiC matrix under irradiation with limited swelling; high fission product retention of the fuel form, with the SiC matrix acting as an additional barrier to fission product release; and enhanced oxidation resistance during normal operation and air ingress accidents. In addition, the literature shows that the effective thermal conductivity of SiC fuel compacts and conventional HTGR compacts are expected to be similar. The key finding of this study is that FCM fuel, within the form factor of a typical General Atomics prismatic graphite block, exhibits similar fuel cycle performance to conventional HTGR fuel. The reactor cycle length, discharge burnup, and natural resource utilization are similar. However, the reduced moderation in the FCM designs considered here does marginally reduce the discharge burnup, and therefore natural resource utilization, versus the reference HTGR design. The hardened neutron flux spectrum resulting from the SiC matrix, which displaces carbon from the core, requires a slightly higher packing fraction of conventional uranium oxy-carbide (UCO) fuel kernels or the use of higher density uranium mononitride (UN)-based fuel kernels. These options will marginally increase the decay power, because they harden the neutron flux energy spectrum and increase the density of 238U in the fuel. In one case considered, this will increase the absorption of neutrons in 238U, and the resultant impact of 239Np isotope on the decay power. The Doppler coefficients normalized per total fuel heat capacity are weaker in the FCM-fueled designs than in the reference HTGR design. This impacts the energy deposition in a control rod ejection accident, and hence the design of potential transient tests of these fuel forms. In addition, analyses of loss-of-forced cooling accidents indicate that the fuel temperature during these design basis accidents are up to ∼30 °C higher with FCM fuel than with conventional HTGR fuels due to the increased decay power.

    Original languageEnglish (US)
    Pages (from-to)277-287
    Number of pages11
    JournalAnnals of Nuclear Energy
    Volume114
    DOIs
    StatePublished - Apr 2018

    All Science Journal Classification (ASJC) codes

    • Nuclear Energy and Engineering

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