Critical heat flux on zircaloy and accident tolerant fuel cladding under prototypical conditions of pressurized and boiling water reactors

Moreira, Tiago Augusto; Lee, DongHwi; Anderson, Mark H.

doi:10.1016/j.applthermaleng.2022.118740

Critical heat flux on zircaloy and accident tolerant fuel cladding under prototypical conditions of pressurized and boiling water reactors

Journal Article · Sat May 28 00:00:00 EDT 2022 · Applied Thermal Engineering

DOI:https://doi.org/10.1016/j.applthermaleng.2022.118740· OSTI ID:1976862

Moreira, Tiago Augusto ^[1]; Lee, DongHwi ^[2]; Anderson, Mark H. ^[3]

University of Wisconsin, Madison, WI (United States); OSTI
Chonbuk National University, Jeonju (Korea, Republic of)
University of Wisconsin, Madison, WI (United States)

First of a kind high spatial and temporal resolution temperature measurements of Pressurized Water Reactor and Boiling Water Reactor simulated fuel pins (high heat flux cosine profile heaters) in annular rod type geometry during nucleate boiling, departure from nucleate boiling and re-wetting has been performed. Distributed fiber optic measurements of the cladding axial temperature with spatial resolutions down to 2.5 mm axially at up to three equally spaced azimuthal locations per rod have demonstrated the similarities and differences of the different accident tolerant fuel cladding materials. These experiments show the performance of accident tolerant cladding with a precise measurement of the location and evolution of dryout followed by the re-wetting phenomena (see supplementary time history of a critical heat flux event). Experimental parameters varied coolant mass fluxes from 1695 to 2712 kg/m²s, pressures from 10 to 20 MPa, and inlet subcooling from 10 to 55 °C. It was found that the coatings and different material have little effect on the value of the critical heat flux. Any discrepancies between the local critical heat flux verified among the cladding material are attributed to different departure from nucleate boiling phenomena captured during the experiment with the high-resolution optical fiber temperature data. Substantial differences in the performance of the different cladding material to survive a critical heat flux event were however observed by post-test X-ray and optical inspection. External evident damage and internal cracks to the simulated fuel pin having the bare zircaloy cladding were observed in regions associated with the occurrence of the critical heat flux for longer periods. Such damages were associated with zirconium oxidation, and not noticed on the Cr coated zircaloy or FeCrAl claddings. These results indicate that the use of FeCrAl or a Cr coated zircaloy as cladding in Pressurized Water Reactor and Boiling Water Reactor may increase the survivability of the cladding material for short transients.

View Accepted Manuscript (DOE)

Research Organization:: University of Wisconsin, Madison, WI (United States)

Sponsoring Organization:: Nuclear Regulatory Commission (NRC); USDOE Office of Nuclear Energy (NE), Nuclear Energy University Program (NEUP); University of Wisconsin

Grant/Contract Number:: NE0008714

OSTI ID:: 1976862

Journal Information:: Applied Thermal Engineering, Journal Name: Applied Thermal Engineering Journal Issue: C Vol. 213; ISSN 1359-4311

Publisher:: ElsevierCopyright Statement

Country of Publication:: United States

Language:: English

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