Estimation of coping time in pressurized water reactors for accident tolerant fuel claddings
Massachusetts Institute of Technology. Department of Nuclear Science and Engineering.
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The Fukushima Nuclear Power Plant (NPP) accident in Japan has motivated improving the safety of current light water reactors (LWRs). Accident tolerant fuels (ATF) are being developed to enhance the safety of LWRs by tolerating loss of active cooling in the core for a longer duration compared to standard UO₂ and Zirconium-based claddings. In this work, high-temperature steam oxidation characteristics of potential ATF claddings, monolayer iron-chromium-aluminum (FeCrAl) and Cr-coated Zircaloy, are experimentally investigated. Specifically, this work investigates the high-temperature oxidation characteristics of FeCrAl alloy after exposure to 1000-1400 °C steam for 1 hour. A model for oxidation of FeCrAl alloy was developed based on the measured weight gain. The severe degradation of the FeCrAl samples from the steam attack was observed at ~1400 °C. Experimental investigation of ATF claddings also included high-temperature oxidation of Cr-coated Zircaloy pressure tube. Post-test analysis showed that for some regions, the Cr-coating is still present after 90 minutes of exposure to 1200°C steam, protecting the Zircaloy substrate beneath the coated layer. The performances of the FeCrAl and Cr-coated ATF claddings under beyond design basis accidents (BDBA) are modeled with thermal-hydraulics design basis code TRACE. A 3-loop Pressurized Water Reactor (PWR) model is created and the following BDBAs are simulated for this study: large break loss of coolant accident (LOCA) without safety injection systems, short-term station blackout (SBO) without any mitigation actions from the beginning and long-term SBO with auxiliary feedwater flow for the first 24 hours and the no mitigation actions afterwards. Two models are used for high-temperature oxidation of FeCrAl: the MIT model based on the experimental results of this work, and the Oak Ridge National Laboratory (ORNL) model based on experimental results of ORNL's work. The results showed that ATF claddings increase the coping time and produce less hydrogen compared to Zircaloy cladding under the considered BDBAs scenarios.
Thesis: S.M., Massachusetts Institute of Technology, Department of Nuclear Science and Engineering, 2018.This electronic version was submitted by the student author. The certified thesis is available in the Institute Archives and Special Collections.Cataloged from student-submitted PDF version of thesis. Page 105 blank.Includes bibliographical references (pages 99-104).
DepartmentMassachusetts Institute of Technology. Department of Nuclear Science and Engineering.
Massachusetts Institute of Technology
Nuclear Science and Engineering.