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Title: Development of LWR Fuels with Enhanced Accident Tolerance ATF Feasibility Analysis and Final Technical Report Deliverable for the Westinghouse Accident Tolerant Fuel Program

Abstract

The Westinghouse Electric Company LLC (Westinghouse) accident tolerant fuel (ATF) program utilizes chromium (Cr) coated zirconium alloy cladding with doped UO2 (ADOPT) and uranium silicide (U3Si2) high density/high thermal conductivity fuel for its lead test rod (LTR) program with irradiation beginning in 2019. ADOPT is a Cr2O3+Al2O3 doped UO2 pellet with increased oxidation resistance and density, and increased resistance to fission gas release. Together, this is the near term Westinghouse EnCore fuel product. The lead test assembly (LTA) program will use both SiGA SiC/SiC composites from General Atomics and Cr coated zirconium alloy claddings with the high density/high thermal conductivity/oxidation resistant U3Si2 and doped UO2 pellets which will begin in 2022. The SiGA SiC/SiC composites with U3Si2 fuel is the long term EnCore fuel product. Over the past several years, Westinghouse has tested the Cr coated zirconium (Zr) and silicon carbide (SiC) claddings in autoclaves and in the Massachusetts Institute of Technology (MIT) reactor and the U3Si2 pellets have been tested in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) and at Westinghouse. The manufacturing parameters for the SiC have been found to have a significant effect on the corrosion rate of the SiC in light water reactormore » (LWR) conditions. A Cr coating is being developed as a backup option for SiC cladding to further reduce LWR corrosion. High temperature tests at the state-of-the-art Westinghouse facilities in Churchill, PA and at Karlsruhe Institute of Technology (KIT) have been carried out to determine the time and temperature limits for the SiC and Cr coated zirconium claddings. These tests indicate that Cr coated Zr cladding can take temperatures up to about 1500°C for short periods of time without becoming totally oxidized. The main issue is the formation of the Cr-Zr eutectic at 1333°C resulting in the migration of this eutectic inward with the formation of ZrO2 on the outside of the tube. The Cr coated Zr cladding delays the bursting of the tube and reduces the burst area. SiC cladding can withstand temperatures up to about 1800°C to 1900°C before excessive corrosion reactions begin and do not balloon and burst. The Cr protective coating will oxidize and provide corrosion tolerance to >2000°C at least for a short time. Fuel rod and assembly design in preparation for the LTR and LTA programs is underway as well as licensing efforts with the Nuclear Regulatory Commission (NRC). Finally, accident analyses coupled with economic evaluations for both operating savings as well as fuel savings have been initiated. Cr coated zirconium alloy claddings showed low corrosion rates in both autoclave and in-reactor tests and will be fabricated using commercial processes for the first LTRs for insertion in 2019. Current autoclave results for SiC composite claddings indicate that a corrosion rate of fewer than 2 micrometers per year can be achieved which meets corrosion requirements under normal operating conditions. Modular Accident Analysis Program 5 (MAAP5) calculations indicate that solid fission products can be contained within SiC cladding for up to two hours longer than current Zr based cladding in a station blackout scenario. This additional time is due to the much higher temperature capability of SiC. These two hours can be used to implement additional response options instituted through the FLEX program by reactor operators. While the Cr coated Zr alloy option offers more modest ATF gains in similar situations, the coatings do delay the runaway oxidation encountered by uncoated zirconium cladding. Both ATF cladding options also reduce hydrogen production which dramatically reduces primary system and containment pressure and the risk of fission product release beyond containment in the unlikely event of an accident. The lower pressure in the system allows more time to feed cooling water to the core, resulting in the avoidance of fuel melting. The coping time is extended indefinitely as long as the modest water flow provided by FLEX continues. SiC exhibits the highest accident tolerance of any cladding and significant advances in understanding the design, manufacture and accident behavior of SiC cladding have been made. Department of Energy (DOE) support is needed to continue the on-going experimental work addressing the interaction between U3Si2 and SiC at temperatures >1200°C, the effect of SiC manufacturing methods and manufacturing variability on physical properties, and the ability to predict the mechanical properties of SiC composites based on the composite design. Additionally, experimental work on methods to rapidly fabricate SiC composite structures with high density and reduce the fabrication price of SiC fibers while maintaining a high level of performance is needed. Minimal (<1%) swelling of U3Si2 and subsequent fission gas release has been demonstrated up to 20 MWd/kgU. Irradiation experiments with U3Si2 fuel in ATR to determine these properties at 40 to 50 MWd/kgU are underway. These experiments will provide the data needed to license fuel rod design codes. Research is continuing on production methods for making U3Si2 from UF6 without going through the U metal step. Oxidation tests of fuel pellets and powder in steam and synthetic air indicate that U3Si2 has a lower oxidation reaction initiation temperature than UO2. Options to reduce the U3Si2 oxidation rate are also being explored. In addition to the work supported by the DOE, research and testing activities are being carried on in a world-wide effort funded by many countries such as Sweden, United Kingdom, Belgium, Netherlands, Spain, Germany, Japan and France. This work is being facilitated through the Westinghouse led Collaboration for Advanced Research on Accident Tolerant Fuel (CARAT) program. Annual meetings have been organized by Westinghouse as a venue for presentation of this work and to provide for the cross-fertilization of ideas among the many outstanding researchers in the ATF area. Since SiC, coated cladding, and high density fuel options are not currently used in the nuclear industry; support from the government and industry members is needed to further the significant effort of setting new standards. The same is true for the Nuclear Regulatory Commission (NRC) which must review and approve the commercial use of these new fuels since all current regulations are oriented toward Zr/ UO2 fuel. Several meetings have been held with the NRC to generate a fast-track approach to licensing ATF using a combination of atomic modeling, in-rod sensors, and in-reactor testing. This approach is being memorialized in a Qualification Plan currently being generated by Westinghouse. There are two outstanding technology items to be solved for U3Si2. The first is increasing the oxidation resistance so that there is not excessive reaction up to ~1400°C. This issue is being addressed by looking at coating the pellets, coating the U3Si2 grains, and by changing the composition of the U3Si2. The second is developing a method to manufacture that does not require U metal production. Although not a make-or-break technical issue, use of U metal would lower the economic potential of the final product. This issue is being addressed by looking at methods of productions starting with UF4 or UC (both of which can be readily obtained from UF6), and UF6 with SiH4 and other Si containing compounds. Thermodynamic studies indicate that these are potentially fruitful routes to study. Due to the cost savings from higher density fuel and the accident tolerance of coated claddings, Exelon has identified potential commercial plants to host the LTRs. Planning is underway to implement this schedule.« less

Authors:
ORCiD logo [1];
  1. WESTINGHOUSE ELECTRIC COMPANY
Publication Date:
Research Org.:
WESTINGHOUSE ELECTRIC COMPANY
Sponsoring Org.:
USDOE Office of Nuclear Energy (NE), Fuel Cycle Technologies (NE-5)
Contributing Org.:
General Atomics Idaho National Laboratory Karlsruhe Institute of Technology Los Alamos National Laboratory Massachusetts Institute of Technology National Nuclear Laboratory (United Kingdom) Oak Ridge National Laboratory Rennselaer Polytechnic Institute University of South Carolina University of Texas at San Antonio University of Virginia University of Wisconsin
OSTI Identifier:
1511013
Report Number(s):
GATFT-19-004, Revision 2
GATFT-19-004, Revision 2
DOE Contract Number:  
NE0008222
Resource Type:
Technical Report
Country of Publication:
United States
Language:
English
Subject:
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS; nuclear fuel accident tolerant SiC U3Si2 uranium silicide silican carbide coated cladding Cr chromium test corrosion in-reactor reactor design

Citation Formats

Lahoda, Edward Jean, and Boylan, Frank A. Development of LWR Fuels with Enhanced Accident Tolerance ATF Feasibility Analysis and Final Technical Report Deliverable for the Westinghouse Accident Tolerant Fuel Program. United States: N. p., 2019. Web. doi:10.2172/1511013.
Lahoda, Edward Jean, & Boylan, Frank A. Development of LWR Fuels with Enhanced Accident Tolerance ATF Feasibility Analysis and Final Technical Report Deliverable for the Westinghouse Accident Tolerant Fuel Program. United States. doi:10.2172/1511013.
Lahoda, Edward Jean, and Boylan, Frank A. Tue . "Development of LWR Fuels with Enhanced Accident Tolerance ATF Feasibility Analysis and Final Technical Report Deliverable for the Westinghouse Accident Tolerant Fuel Program". United States. doi:10.2172/1511013. https://www.osti.gov/servlets/purl/1511013.
@article{osti_1511013,
title = {Development of LWR Fuels with Enhanced Accident Tolerance ATF Feasibility Analysis and Final Technical Report Deliverable for the Westinghouse Accident Tolerant Fuel Program},
author = {Lahoda, Edward Jean and Boylan, Frank A},
abstractNote = {The Westinghouse Electric Company LLC (Westinghouse) accident tolerant fuel (ATF) program utilizes chromium (Cr) coated zirconium alloy cladding with doped UO2 (ADOPT) and uranium silicide (U3Si2) high density/high thermal conductivity fuel for its lead test rod (LTR) program with irradiation beginning in 2019. ADOPT is a Cr2O3+Al2O3 doped UO2 pellet with increased oxidation resistance and density, and increased resistance to fission gas release. Together, this is the near term Westinghouse EnCore fuel product. The lead test assembly (LTA) program will use both SiGA SiC/SiC composites from General Atomics and Cr coated zirconium alloy claddings with the high density/high thermal conductivity/oxidation resistant U3Si2 and doped UO2 pellets which will begin in 2022. The SiGA SiC/SiC composites with U3Si2 fuel is the long term EnCore fuel product. Over the past several years, Westinghouse has tested the Cr coated zirconium (Zr) and silicon carbide (SiC) claddings in autoclaves and in the Massachusetts Institute of Technology (MIT) reactor and the U3Si2 pellets have been tested in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) and at Westinghouse. The manufacturing parameters for the SiC have been found to have a significant effect on the corrosion rate of the SiC in light water reactor (LWR) conditions. A Cr coating is being developed as a backup option for SiC cladding to further reduce LWR corrosion. High temperature tests at the state-of-the-art Westinghouse facilities in Churchill, PA and at Karlsruhe Institute of Technology (KIT) have been carried out to determine the time and temperature limits for the SiC and Cr coated zirconium claddings. These tests indicate that Cr coated Zr cladding can take temperatures up to about 1500°C for short periods of time without becoming totally oxidized. The main issue is the formation of the Cr-Zr eutectic at 1333°C resulting in the migration of this eutectic inward with the formation of ZrO2 on the outside of the tube. The Cr coated Zr cladding delays the bursting of the tube and reduces the burst area. SiC cladding can withstand temperatures up to about 1800°C to 1900°C before excessive corrosion reactions begin and do not balloon and burst. The Cr protective coating will oxidize and provide corrosion tolerance to >2000°C at least for a short time. Fuel rod and assembly design in preparation for the LTR and LTA programs is underway as well as licensing efforts with the Nuclear Regulatory Commission (NRC). Finally, accident analyses coupled with economic evaluations for both operating savings as well as fuel savings have been initiated. Cr coated zirconium alloy claddings showed low corrosion rates in both autoclave and in-reactor tests and will be fabricated using commercial processes for the first LTRs for insertion in 2019. Current autoclave results for SiC composite claddings indicate that a corrosion rate of fewer than 2 micrometers per year can be achieved which meets corrosion requirements under normal operating conditions. Modular Accident Analysis Program 5 (MAAP5) calculations indicate that solid fission products can be contained within SiC cladding for up to two hours longer than current Zr based cladding in a station blackout scenario. This additional time is due to the much higher temperature capability of SiC. These two hours can be used to implement additional response options instituted through the FLEX program by reactor operators. While the Cr coated Zr alloy option offers more modest ATF gains in similar situations, the coatings do delay the runaway oxidation encountered by uncoated zirconium cladding. Both ATF cladding options also reduce hydrogen production which dramatically reduces primary system and containment pressure and the risk of fission product release beyond containment in the unlikely event of an accident. The lower pressure in the system allows more time to feed cooling water to the core, resulting in the avoidance of fuel melting. The coping time is extended indefinitely as long as the modest water flow provided by FLEX continues. SiC exhibits the highest accident tolerance of any cladding and significant advances in understanding the design, manufacture and accident behavior of SiC cladding have been made. Department of Energy (DOE) support is needed to continue the on-going experimental work addressing the interaction between U3Si2 and SiC at temperatures >1200°C, the effect of SiC manufacturing methods and manufacturing variability on physical properties, and the ability to predict the mechanical properties of SiC composites based on the composite design. Additionally, experimental work on methods to rapidly fabricate SiC composite structures with high density and reduce the fabrication price of SiC fibers while maintaining a high level of performance is needed. Minimal (<1%) swelling of U3Si2 and subsequent fission gas release has been demonstrated up to 20 MWd/kgU. Irradiation experiments with U3Si2 fuel in ATR to determine these properties at 40 to 50 MWd/kgU are underway. These experiments will provide the data needed to license fuel rod design codes. Research is continuing on production methods for making U3Si2 from UF6 without going through the U metal step. Oxidation tests of fuel pellets and powder in steam and synthetic air indicate that U3Si2 has a lower oxidation reaction initiation temperature than UO2. Options to reduce the U3Si2 oxidation rate are also being explored. In addition to the work supported by the DOE, research and testing activities are being carried on in a world-wide effort funded by many countries such as Sweden, United Kingdom, Belgium, Netherlands, Spain, Germany, Japan and France. This work is being facilitated through the Westinghouse led Collaboration for Advanced Research on Accident Tolerant Fuel (CARAT) program. Annual meetings have been organized by Westinghouse as a venue for presentation of this work and to provide for the cross-fertilization of ideas among the many outstanding researchers in the ATF area. Since SiC, coated cladding, and high density fuel options are not currently used in the nuclear industry; support from the government and industry members is needed to further the significant effort of setting new standards. The same is true for the Nuclear Regulatory Commission (NRC) which must review and approve the commercial use of these new fuels since all current regulations are oriented toward Zr/ UO2 fuel. Several meetings have been held with the NRC to generate a fast-track approach to licensing ATF using a combination of atomic modeling, in-rod sensors, and in-reactor testing. This approach is being memorialized in a Qualification Plan currently being generated by Westinghouse. There are two outstanding technology items to be solved for U3Si2. The first is increasing the oxidation resistance so that there is not excessive reaction up to ~1400°C. This issue is being addressed by looking at coating the pellets, coating the U3Si2 grains, and by changing the composition of the U3Si2. The second is developing a method to manufacture that does not require U metal production. Although not a make-or-break technical issue, use of U metal would lower the economic potential of the final product. This issue is being addressed by looking at methods of productions starting with UF4 or UC (both of which can be readily obtained from UF6), and UF6 with SiH4 and other Si containing compounds. Thermodynamic studies indicate that these are potentially fruitful routes to study. Due to the cost savings from higher density fuel and the accident tolerance of coated claddings, Exelon has identified potential commercial plants to host the LTRs. Planning is underway to implement this schedule.},
doi = {10.2172/1511013},
journal = {},
number = ,
volume = ,
place = {United States},
year = {2019},
month = {4}
}