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Title: Investigations on the Thermal-hydraulic Behavior of Accident Tolerant Fuel Cladding Materials

Abstract

Since the 2011 Fukushima accident, significant research has been devoted to developing accident-tolerant fuel (ATF) cladding materials. These investigations have mostly focused on the ability to resist runaway steam oxidation and retain mechanical strength and structural integrity under thermal shocks. However, much remains unknown about the materials’ thermal-hydraulic behavior, particularly under light-water reactor (LWR) operating conditions. Two phenomena that determine safety margins in both normal and off-normal operating conditions—i.e., the critical heat flux (CHF) and Leidenfrost point (LFP) temperature—have not been investigated thoroughly for ATF materials. Surface properties, such as wettability and roughness, are known to influence pool and flow boiling CHF, as well as the LFP temperature. However, little is known about the surface wettability of ATF materials, particularly at LWR pressure and temperature. Little has been done on flow-boiling CHF and quenching heat transfer, especially for droplet quenching. Nonetheless, it is known that CHF in transient conditions, e.g., an exponentially escalating power transient, can be significantly different from that expected in steady-state operation. Very few studies have investigated transient CHF, either on ATF materials or under LWR pressure and temperature. The thorough understanding of transient CHF under prototypical reactor conditions will benefit not only the deployment of ATFmore » but also the upcoming national efforts to study transient ATF behavior in the Transient Reactor Test Facility (TREAT) at Idaho National Laboratory. In summary, as a part of ATF development efforts, there is an urgent need to understand how and how much these materials may affect two-phase heat transfer phenomena in nuclear reactor conditions.« less

Authors:
 [1];  [1]; ORCiD logo [2]
  1. Massachusetts Institute of Technology
  2. Idaho National Laboratory
Publication Date:
Research Org.:
Idaho National Lab. (INL), Idaho Falls, ID (United States)
Sponsoring Org.:
USDOE Office of Nuclear Energy (NE)
OSTI Identifier:
1594995
Report Number(s):
INL/EXT-19-56455-Rev000
DOE Contract Number:  
DE-AC07-05ID14517
Resource Type:
Technical Report
Country of Publication:
United States
Language:
English
Subject:
22 - GENERAL STUDIES OF NUCLEAR REACTORS; ATF cladding materials

Citation Formats

Su, G., Bucci, M., and Sabharwall, Piyush. Investigations on the Thermal-hydraulic Behavior of Accident Tolerant Fuel Cladding Materials. United States: N. p., 2020. Web. doi:10.2172/1594995.
Su, G., Bucci, M., & Sabharwall, Piyush. Investigations on the Thermal-hydraulic Behavior of Accident Tolerant Fuel Cladding Materials. United States. doi:10.2172/1594995.
Su, G., Bucci, M., and Sabharwall, Piyush. Wed . "Investigations on the Thermal-hydraulic Behavior of Accident Tolerant Fuel Cladding Materials". United States. doi:10.2172/1594995. https://www.osti.gov/servlets/purl/1594995.
@article{osti_1594995,
title = {Investigations on the Thermal-hydraulic Behavior of Accident Tolerant Fuel Cladding Materials},
author = {Su, G. and Bucci, M. and Sabharwall, Piyush},
abstractNote = {Since the 2011 Fukushima accident, significant research has been devoted to developing accident-tolerant fuel (ATF) cladding materials. These investigations have mostly focused on the ability to resist runaway steam oxidation and retain mechanical strength and structural integrity under thermal shocks. However, much remains unknown about the materials’ thermal-hydraulic behavior, particularly under light-water reactor (LWR) operating conditions. Two phenomena that determine safety margins in both normal and off-normal operating conditions—i.e., the critical heat flux (CHF) and Leidenfrost point (LFP) temperature—have not been investigated thoroughly for ATF materials. Surface properties, such as wettability and roughness, are known to influence pool and flow boiling CHF, as well as the LFP temperature. However, little is known about the surface wettability of ATF materials, particularly at LWR pressure and temperature. Little has been done on flow-boiling CHF and quenching heat transfer, especially for droplet quenching. Nonetheless, it is known that CHF in transient conditions, e.g., an exponentially escalating power transient, can be significantly different from that expected in steady-state operation. Very few studies have investigated transient CHF, either on ATF materials or under LWR pressure and temperature. The thorough understanding of transient CHF under prototypical reactor conditions will benefit not only the deployment of ATF but also the upcoming national efforts to study transient ATF behavior in the Transient Reactor Test Facility (TREAT) at Idaho National Laboratory. In summary, as a part of ATF development efforts, there is an urgent need to understand how and how much these materials may affect two-phase heat transfer phenomena in nuclear reactor conditions.},
doi = {10.2172/1594995},
journal = {},
number = ,
volume = ,
place = {United States},
year = {2020},
month = {1}
}

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