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Title: Development of LWR Fuels with Enhanced Accident Tolerance ATF Feasibility Analysis 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 uranium silicide (U 3Si 2) high density/high thermal conductivity fuel for its lead test rod (LTR) program with irradiation beginning in 2019. The lead test assembly (LTA) program will use both SiC/SiC composites and Cr coated zirconium alloy claddings with the high density/high thermal conductivity U 3Si 2 pellet which will begin in 2022. 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 U 3Si 2 pellets have been tested in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL). High temperature tests at the state-of-the-art Westinghouse facilities in Churchill, PA have been carried out to determine the time and temperature limits for the SiC and Cr coated zirconium claddings. 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 zirconiummore » 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 which meets corrosion requirements may be achieved 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 (~2000°C). 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 (~200°C higher than bare Zr) 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 water flow 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 support is needed to continue the on-going experimental work addressing the interaction between U 3Si 2 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 U 3Si 2 and subsequent fission gas release has been demonstrated up to 20 MWd/kgU. Irradiation experiments with U 3Si 2 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 U 3Si 2 from UF 6 without going through the U metal step. Oxidation tests of fuel pellets and powder in steam and synthetic air indicate that U 3Si 2 has a lower oxidation reaction initiation temperature than UO 2. Options to reduce the U 3Si 2 oxidation rate are being explored. A hybrid fuel design using pellets with the weight ratio of 30%/70% U 3Si 2/UN is currently undergoing testing in the ATR. This approach is being explored to further increase the positive economics of the U 3Si 2 fuel. In addition to the work supported by the USDOE, 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 are 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/ UO 2 fuel. There are no showstoppers for accident tolerant fuel technologies when considering supply chain, fabrication, enrichment suppliers, production, and licensing. 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 and LTAs. Planning is underway to implement this schedule.« less

Authors:
ORCiD logo [1];  [1]
  1. Westinghouse Electric Company LLC, Cranberry Woods, PA (United States)
Publication Date:
Research Org.:
Westinghouse Electric Company LLC, Cranberry Woods, PA (United States)
Sponsoring Org.:
USDOE Office of Nuclear Energy (NE), Fuel Cycle Technologies (NE-5)
Contributing Org.:
Westinghouse Electric Company LLC Argonne National Laboratory Ceramic Tubular Products Exelon Nuclear General Atomics Idaho National Laboratory Institute for Energy Technology (Norway) Los Alamos National Laboratory Massachusetts Institute of Technology National Nuclear Laboratory (United Kingdom) Paul Scherrer Institute (Switzerland) Pennsylvania State University Southern Nuclear Company United Technologies Research Center University of Wisconsin
OSTI Identifier:
1464441
Report Number(s):
GATFT-18-30
GATFT-18-030
DOE Contract Number:  
NE0008222
Resource Type:
Technical Report
Country of Publication:
United States
Language:
English
Subject:
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS; accident tolerant; nuclear fuel; SiC; coated; cladding; chromium; coated zirconium; uranium silicide

Citation Formats

Lahoda, Edward J., and Boyan, Frank A. Development of LWR Fuels with Enhanced Accident Tolerance ATF Feasibility Analysis Report Deliverable for the Westinghouse Accident Tolerant Fuel Program. United States: N. p., 2018. Web.
Lahoda, Edward J., & Boyan, Frank A. Development of LWR Fuels with Enhanced Accident Tolerance ATF Feasibility Analysis Report Deliverable for the Westinghouse Accident Tolerant Fuel Program. United States.
Lahoda, Edward J., and Boyan, Frank A. Wed . "Development of LWR Fuels with Enhanced Accident Tolerance ATF Feasibility Analysis Report Deliverable for the Westinghouse Accident Tolerant Fuel Program". United States.
@article{osti_1464441,
title = {Development of LWR Fuels with Enhanced Accident Tolerance ATF Feasibility Analysis Report Deliverable for the Westinghouse Accident Tolerant Fuel Program},
author = {Lahoda, Edward J. and Boyan, Frank A.},
abstractNote = {The Westinghouse Electric Company LLC (Westinghouse) accident tolerant fuel (ATF) program utilizes chromium (Cr) coated zirconium alloy cladding with uranium silicide (U3Si2) high density/high thermal conductivity fuel for its lead test rod (LTR) program with irradiation beginning in 2019. The lead test assembly (LTA) program will use both SiC/SiC composites and Cr coated zirconium alloy claddings with the high density/high thermal conductivity U3Si2 pellet which will begin in 2022. 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). High temperature tests at the state-of-the-art Westinghouse facilities in Churchill, PA have been carried out to determine the time and temperature limits for the SiC and Cr coated zirconium claddings. 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 which meets corrosion requirements may be achieved 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 (~2000°C). 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 (~200°C higher than bare Zr) 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 water flow 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 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 being explored. A hybrid fuel design using pellets with the weight ratio of 30%/70% U3Si2/UN is currently undergoing testing in the ATR. This approach is being explored to further increase the positive economics of the U3Si2 fuel. In addition to the work supported by the USDOE, 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 are 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. There are no showstoppers for accident tolerant fuel technologies when considering supply chain, fabrication, enrichment suppliers, production, and licensing. 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 and LTAs. Planning is underway to implement this schedule.},
doi = {},
journal = {},
number = ,
volume = ,
place = {United States},
year = {2018},
month = {6}
}

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