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Title: Development of LWR Fuels with Enhanced Accident Tolerance

Significant progress was made on the technical, licensing, and business aspects of the Westinghouse Electric Company’s Enhanced Accident Tolerant Fuel (ATF) by the Westinghouse ATF team. The fuel pellet options included waterproofed U15N and U3Si2 and the cladding options SiC composites and zirconium alloys with surface treatments. Technology was developed that resulted in U3Si2 pellets with densities of >94% being achieved at the Idaho National Laboratory (INL). The use of U3Si2 will represent a 15% increase in U235 loadings over those in UO₂ fuel pellets. This technology was then applied to manufacture pellets for 6 test rodlets which were inserted in the Advanced Test Reactor (ATR) in early 2015 in zirconium alloy cladding. The first of these rodlets are expected to be removed in about 2017. Key characteristics to be determined include verification of the centerline temperature calculations, thermal conductivity, fission gas release, swelling and degree of amorphization. Waterproofed UN pellets have achieved >94% density for a 32% U3Si2/68% UN composite pellet at Texas A&M University. This represents a U235 increase of about 31% over current UO2 pellets. Pellets and powders of UO2, UN, and U3Si2the were tested by Westinghouse and Los Alamos National Laboratory (LANL) using differential scanning calorimetrymore » to determine what their steam and 20% oxygen corrosion temperatures were as compared to UO2. Cold spray application of either the amorphous steel or the Ti2AlC was successful in forming an adherent ~20 micron coating that remained after testing at 420°C in a steam autoclave. Tests at 1200°C in 100% steam on coatings for Zr alloy have not been successful, possibly due to the low density of the coatings which allowed steam transport to the base zirconium metal. Significant modeling and testing has been carried out for the SiC/SiC composite/SiC monolith structures. A structure with the monolith on the outside and composite on the inside was developed which is the current baseline structure and a SiC to SiC tube closure approach. Permeability tests and mechanical tests were developed to verify the operation of the SiC cladding. Steam autoclave (420°C), high temperature (1200°C) flowing steam tests and quench tests were carried out with minimal corrosion, mechanical or hermeticity degradation effect on the SiC cladding or end plug closure. However, in-reactor loop tests carried out in the MIT reactor indicated an unacceptable degree of corrosion, likely due to the corrosive effect of radiolysis products which attacked the SiC.« less
Authors:
 [1] ;  [1]
  1. Westinghouse Electric Company, LLC, Cranberry Woods, PA (United States)
Publication Date:
OSTI Identifier:
1233713
Report Number(s):
DOE/NE--0000566
RT-TR-15-34; TRN: US1600203
DOE Contract Number:
NE0000566
Resource Type:
Technical Report
Research Org:
Westinghouse Electric Company, LLC, Cranberry Woods, PA (United States)
Sponsoring Org:
USDOE Office of Nuclear Energy (NE)
Country of Publication:
United States
Language:
English
Subject:
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS; ACCIDENT-TOLERANT NUCLEAR FUELS; SILICON CARBIDES; FUEL PELLETS; MECHANICAL TESTS; ZIRCONIUM ALLOYS; STEELS; URANIUM DIOXIDE; URANIUM SILICIDES; COMPOSITE MATERIALS; FUEL CANS; URANIUM NITRIDES; URANIUM 235; OXYGEN; CORROSION; DENSITY; STEAM; ATR REACTOR; THERMAL CONDUCTIVITY; TEMPERATURE RANGE 0400-1000 K; TEMPERATURE RANGE 1000-4000 K; RADIOLYSIS; COMPARATIVE EVALUATIONS; TITANIUM CARBIDES; ALUMINIUM CARBIDES; SPRAYED COATINGS; PERMEABILITY; SWELLING; FISSION PRODUCT RELEASE; AMORPHOUS STATE; SURFACE TREATMENTS; PERFORMANCE; POWDERS; SIMULATION; WATER COOLED REACTORS; WATER MODERATED REACTORS; TEMPERATURE DISTRIBUTION; IN PILE LOOPS; FUEL RODS