@inproceedings{3aba5b9d4d1a45b09fadbcd1dbb6d6e7,
title = "Investigation of Cladding Thermal Behavior Under Simulated Reflood Conditions",
abstract = "Cr-coated Zr-alloy and FeCrAl alloy are two of the most promising LWR Accident Tolerant Fuel (ATF) cladding concepts by virtue of their excellent corrosion resistance at high burnup fuel cycle and oxidation resistance under high temperature accident conditions.A single rod reflood facility is designed and constructed to investigate thermal-hydraulic behavior of these ATF cladding concepts during a simulated loss-of-coolant accident (LOCA).The facility is capable of producing cladding temperatures up to 1200 °C, water subcooling in the range of 0-75 K, and reflood velocities up to 10 cm/s.Moreover, cladding test samples can be oxidized using high temperature steam prior to water reflooding.In this work, stainless steel tubes and surrogate metallic pellets are used to investigate key thermal-hydraulic parameters impacting the reflooding behavior for determining the test matrix of actual ATF claddings in the future.Cladding wall thermal boundary conditions are obtained using inverse heat conduction analysis informed by thermocouple data within the surrogate fuel pellets inside the cladding tube.Reproducibility testing with stainless steel cladding tubes demonstrates the capability of determining quench temperatures and quench front velocities.",
keywords = "Accident Tolerant Fuels, LOCA, Quench Temperature",
author = "Cole Dunbar and Jung, {Woo Hyun} and Thomas Demo and Michael Corradini and Kumar Sridharan and Hwasung Yeom and Robert Armstrong and David Kamerman and Benjamin Maier and Andrew Hoffman and Raul Rebak",
note = "Publisher Copyright: {\textcopyright} 2023 Proceedings of the 20th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2023. All rights reserved.; 20th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2023 ; Conference date: 20-08-2023 Through 25-08-2023",
year = "2023",
doi = "10.13182/NURETH20-40257",
language = "English",
series = "Proceedings of the 20th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2023",
publisher = "American Nuclear Society",
pages = "3346--3359",
booktitle = "Proceedings of the 20th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2023",
}