Finding
This demonstrates that SMRs can serve as a near-term deployment fit to SiC cladding with a sound management of its statistical brittle fracture.
Paper
Failure probabilities of SiC clad fuel during a LOCA in public acceptable simple SMR (PASS)
Citations: 23
Full textAbstract
Abstract Structural integrity of SiC clad fuels in reference Small Modular Reactors (SMRs) (NuScale, SMART, IRIS) and a commercial pressurized water reactor (PWR) are assessed with a multi-layered SiC cladding structural analysis code. Featured with low fuel pin power and temperature, SMRs demonstrate markedly reduced incore-residence fracture probabilities below ∼10 −7 , compared to those of commercial PWRs ∼10 −6 –10 −1 . This demonstrates that SMRs can serve as a near-term deployment fit to SiC cladding with a sound management of its statistical brittle fracture. We proposed a novel SMR named Public Acceptable Simple SMR (PASS), which is featured with 14 × 14 assemblies of SiC clad fuels arranged in a square ring layout. PASS aims to rely on radiative cooling of fuel rods during a loss of coolant accident (LOCA) by fully leveraging high temperature tolerance of SiC cladding. An overarching assessment of SiC clad fuel performance in PASS was conducted with a combined methodology—(1) FRAPCON-SiC for steady-state performance analysis of PASS fuel rods, (2) computational fluid dynamics code FLUENT for radiative cooling rate of fuel rods during a LOCA, and (3) multi-layered SiC cladding structural analysis code with previously developed SiC recession correlations under steam environments for both steady-state and LOCA. The results show that PASS simultaneously maintains desirable fuel cooling rate with the sole radiation and sound structural integrity of fuel rods for over 36 days of a LOCA without water supply. The stress level of PASS fuel rods during LOCA is predominantly governed by the initial cold gap pressure; a decrease of the cold gas gap pressure is a handy design direction to assure structural safety margin for SiC cladding.
Authors
Youho Lee, Ho Kim, H. No
Journal
Nuclear Engineering and Design