FY20 SAM Code Developments and Validations for Transient Safety Analysis of Advanced non-LWRs

A Limited-Scope Probabilistic Risk Assessment Study to Risk-Inform the Design of a Fuel Storage System for Spent Pebble-Filled Dry Casks

“…The VHTR is a type of proposed next-generation nuclear plant that is a helium-cooled, graphite-moderated, and graphite-reflected reactor with tri-structural isotropic (TRISO)-coated pebble fuel or prismatic block fuel [13]. The VHTR system research plan [14] was initiated to elaborate tasks for fuel and the fuel cycle [15]; materials [16,17]; hydrogen production; computational validation and methodology [18][19][20], and benchmark project [21]. In China, a high-temperature gas-cooled reactor-pebble bed module 600 (HTR-PM 600) was developed as a commercial version of the HTR-PM, which is a follow-up reactor of the high-temperature gas-cooled reactor-10 (HTR-10) demonstration project [22].…”

“…In China, a high-temperature gas-cooled reactor-pebble bed module 600 (HTR-PM 600) was developed as a commercial version of the HTR-PM, which is a follow-up reactor of the high-temperature gas-cooled reactor-10 (HTR-10) demonstration project [22]. In the United States, as well as the development of VHTR with a small and micro pebble, the advanced reactor demonstration program (ARDP) was launched by INL and Argonne National Laboratory (ANL) with U.S. universities to research and apply advanced computational techniques for system analysis [18], developing the application tool based on thermal-hydraulic simulation codes [19], and coupling codes to validate the experiments [20]. Additionally, from the material management perspective, the Oak Ridge National Laboratory (ORNL) established the material control and counting (MC&A) plan for pebble bed reactors (PBR) to resolve the safeguards and security-related issues [16,17].…”