Assembly-Level Analyses of Accident-Tolerant Cladding Concepts for a Long-Life Civil Marine SMR Core using Micro-Heterogeneous Duplex Fuel

Abstract

In this reactor physics study, we examine the neutronic performance of accident-tolerant fuel (ATF) claddings -- austenitic type 310 stainless steel (310SS), ferritic Fe-20Cr-5Al (FeCrAl), advanced powder metallurgic ferritic (APMT), and silicon carbide (SiC)-based materials — as alternative cladding materials compared with Zircaloy-4 (Zr) cladding. The cores considered use 18% 235U enriched micro-heterogeneous ThO2-UO2 duplex fuel and, for purposes of comparison, 15% 235U enriched homogeneously mixed all-UO2 fuel, loaded into 1313 pin arrays. A constant cladding coating thickness of 655 μm is assumed. We use the WIMS reactor physics code to analyse the associated reactivity, achievable discharge burnup, spectral variations, rim effect and reactivity feedback parameters for the candidate cladding materials at the assembly level. The results show that candidate fuels with 310SS cladding exhibit a ∼13% discharge burnup penalty compared to Zr due to the presence of a very high nickel (Ni) concentration. The high neutron absorption cross-sections of iron (Fe) in the FeCrAl and APMT claddings also lead to a ∼10% discharge burnup penalty. The fuels with SiC cladding can achieve a ∼1% higher discharge burnup compared to Zr due to the low thermal neutron absorption cross-sections of its constituents and the softer neutron spectrum. The claddings with lower capture cross-sections (SiC and Zr) exhibit higher relative fission power at the pellet periphery. For both candidate fuels, the end-of-life 239Pu (for UO2 fuel) and 233U (for duplex fuel) inventories are higher for the claddings (Fe-based: FeCrAl, APMT and steel-based: 310SS) with higher thermal capture cross-sections, unlike for SiC and Zr, where SiC provides higher end-of-life 239Pu and 233U inventories despite having lower capture cross-section than that of the Zr. Reactivity feedback parameter values (moderator and fuel temperature coefficients) are more negative for the duplex fuel than the UO2 fuel for all the candidate claddings, with claddings with harder spectra exhibiting more negative values. The duplex fuel yields a softer spectrum than the UO2 fuel with the candidate claddings, which improves neutron economy and thus discharge burnup.

Department(s)

Nuclear Engineering and Radiation Science

Keywords and Phrases

Absorption spectroscopy; Accidents; Aluminum alloys; Austenitic stainless steel; Chromium alloys; Cladding (coating); Ferrite; Fuel economy; Neutron absorption; Neutrons; Reactivity (nuclear); Silicon carbide; Small nuclear reactors; Ternary alloys; Thickness measurement; Thoria; Uranium dioxide; Zirconium; Zirconium alloys; Absorption cross sections; Accident tolerant fuels; Capture cross sections; Discharge burn-up; Micro-heterogeneous; Reactivity feedback; Silicon carbides (SiC); Thermal capture cross section; Fuels; Accident-tolerant cladding; Achievable discharge burnup; Micro-heterogeneous duplex fuel; Reactivity feedback parameters; Rim effect; Soluble-boron-free design; Spectral hardening

International Standard Serial Number (ISSN)

0149-1970

Document Type

Article - Journal

Document Version

Citation

File Type

text

Language(s)

English

Rights

© 2019 Elsevier, All rights reserved.

Publication Date

01 Mar 2019

Link to Full Text

Share

 
COinS