Three-Dimensional Modeling of Thermal-Mechanical Behavior of Accident Tolerant Fuels

Zitao Zeng,Cenxi Yuan,Yongyu Pan,Jie Zhang,Shixin Gao,Xiujie He,Xi Chen,Chunyu Yin,Chunyu Zhang,Yi Zhou

doi:10.3389/fenrg.2021.636502

Abstract

Considering the safety issues of the traditional UO2-Zr fuel, a variety of accident-tolerant fuel (ATF) candidates have been proposed in recent years. Among the several ATFs, U3Si2, and UN are the two promising candidates for fuel materials owing to their high thermal conductivity and high uranium density. The FeCrAl alloy and the SiC/SiC composite material are the two promising candidates for cladding owing to their high oxidation resistance and high strength. In order to quantitatively evaluate the performance of ATFs, this study summarizes the physical models of typical ATF cladding materials (FeCrAl and SiC) and pellet materials (UN and U3Si2). Then a three-dimensional non-linear finite element method is applied to simulate the thermal-mechanical behavior of several typical fuel-cladding combinations, including UO2-FeCrAl, UN-FeCrAl, U3Si2-FeCrAl, U3Si2-Zr, and U3Si2-SiC. The important physical quantities, such as the fuel centerline temperature, the deformation of the pellet and the cladding as well as the pellet-cladding mechanical interaction (PCMI) were studied. The fission gas release model was also verified and improved.

Highlights

The performance of fuel elements is a key factor ensuring the safety and economy of nuclear reactors
In order to quantitatively evaluate the performance of accident tolerant fuels (ATFs), this study summarizes the physical models of typical ATF cladding materials (FeCrAl and SiC) and pellet materials (UN and U3Si2)
After the Fukushima accident in 2011, great efforts have been put on investigating various accident tolerant fuels (ATFs) to improve the safety of fuels

Summary

INTRODUCTION

The performance of fuel elements is a key factor ensuring the safety and economy of nuclear reactors. A comprehensive study is still absent for the thermal-mechanical performance and fission gas release behavior of ATF materials under LWR conditions. The irradiation swelling of SiC is expressed by the model of Katoh et al (2018): FIGURE 2 | Thermal conductivities of fresh U3Si2 (White, 2018), UN (Hayes et al, 1990a), and UO2 (Williamson, 2011). Where εc,irr is the strain rate in s−1, σ is the effective stress in MPa and is the neutron flux in n/m2s It is proposed by Koyanagi et al (2017) that the thermal creep rates of SiC-based materials are very low at the temperature below ∼1,000◦C. For normal operation temperatures, the thermal creep can be neglected for thermo-mechanical modeling of SiC cladding.

MPa He

RESULTS AND DISCUSSION

DATA AVAILABILITY STATEMENT