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Papers by Heather Chichester

Research paper thumbnail of Summary of FUTURIX-FTA Non-Destructive Examination

Research paper thumbnail of Development of Metallic Fuels for Actinide Transmutation

Research paper thumbnail of Light Water Reactor Accident Tolerant Fuels Irradiation Testing

The purpose of Accident Tolerant Fuels (ATF) experiments is to test novel fuel and cladding conce... more The purpose of Accident Tolerant Fuels (ATF) experiments is to test novel fuel and cladding concepts designed to replace the current zirconium alloy uranium dioxide (UO2) fuel system. The objective of this Research and Development (R&D) is to develop novel ATF concepts that will be able to withstand loss of active cooling in the reactor core for a considerably longer time period than the current fuel system while maintaining or improving the fuel performance during normal operations, operational transients, design basis, and beyond design basis events. It was necessary to design, analyze, and fabricate drop-in capsules to meet the requirements for testing under prototypic LWR temperatures in Idaho National Laboratory's Advanced Test Reactor (ATR). Three industry led teams and one DOE team from Oak Ridge National Laboratory provided fuel rodlet samples for their new concepts for ATR insertion in 2015. As-built projected temperature calculations were performed on the ATF capsules ...

Research paper thumbnail of Microstructural effect on neutron irradiation response of alloy 800H

Research paper thumbnail of Metallography and fuel cladding chemical interaction in fast flux test facility irradiated metallic U-10Zr MFF-3 and MFF-5 fuel pins

Journal of Nuclear Materials, 2016

Abstract The Mechanistic Fuel Failure (MFF) series of metal fuel irradiations conducted in the Fa... more Abstract The Mechanistic Fuel Failure (MFF) series of metal fuel irradiations conducted in the Fast Flux Test Facility (FFTF) provides an important comparison between data generated in the Experimental Breeder Reactor (EBR-II) and that expected in a larger-scale fast reactor. The MFF fuel operated with a peak cladding temperature at the top of the fuel column, but developed peak burnup at the centerline of the core. This places the peak fuel temperature midway between the core center and the top of fuel, lower in the fuel column than in EBR-II experiments. Data from the MFF-3 and MFF-5 assemblies are most comparable to the data obtained from the EBR-II X447 experiment. The two X447 pin breaches were strongly influenced by fuel/cladding chemical interaction (FCCI) at the top of the fuel column. Post irradiation examination data from MFF-3 and MFF-5 are presented and compared to historical EBR-II data.

Research paper thumbnail of Performance of low smeared density sodium-cooled fast reactor metal fuel

Journal of Nuclear Materials

An experiment was performed in the Experimental Breeder Rector-II (EBR-II) in the 1990s to show t... more An experiment was performed in the Experimental Breeder Rector-II (EBR-II) in the 1990s to show that metallic fast reactor fuel could be used in reactors with a single, once-through core. To prove the long duration, high burnup, high neutron exposure capability an experiment where the fuel pin was designed with a very large fission gas plenum and very low fuel smeared density (SD). The experiment, X496, operated to only 8.3 at. % burnup because the EBR-II reactor was scheduled for shutdown at that time. Many of the examinations of the fuel pins only funded recently with the resurgence of reactor designs using very high-burnup fuel. The results showed that, despite the low smeared density of 59% the fuel swelled radially to contact the cladding, fission gas release appeared to be slightly higher than demonstrated in conventional 75%SD fuel tests and axial growth was about the same as 75% SD fuel. There were axial positions in some of the fuel pins which showed evidence of fuel restructuring and an absence of fission products with low melting points and gaseous precursors (Cs and Rb). A model to investigate whether these areas may have overheated due to a loss of bond sodium indicates that it is a possible explanation for the fuel restructuring and something to be considered for fuel performance modeling of low SD fuel.

Research paper thumbnail of Baseline Postirradiation Examination of the FUTURIX-FTA Experiments

Research paper thumbnail of Examination of Legacy Metallic Fuel Pins (U-10Zr) Tested in FFTF

Research paper thumbnail of Postirradiation examination results of several metallic fuel alloys and forms from low burnup AFC irradiations

Journal of Nuclear Materials

Research paper thumbnail of Postirradiation examination on metallic fuel in the AFC-2 irradiation test series

Journal of Nuclear Materials

Research paper thumbnail of Postirradiation Examination of FUTURIX-FTA metallic alloy experiments

Journal of Nuclear Materials

Research paper thumbnail of Advanced Reactor Fuels Irradiation Experiment Design Objectives

Research paper thumbnail of Microstructural effect on neutron irradiation response of alloy 800H

Research paper thumbnail of Use of Silicon Carbide Monitors in ATR Irradiation Testing

Research paper thumbnail of Microstructural effect on neutron irradiation response of alloy 800H

Research paper thumbnail of Hot Cell Sample Preparation of Metallic Nuclear Fuels

Microscopy and Microanalysis, 2012

Research paper thumbnail of Relationship between lath boundary structure and radiation induced segregation in a neutron irradiated 9wt.% Cr model ferritic/martensitic steel

Journal of Nuclear Materials, 2014

Ferritic/Martensitic (F/M) steels with high Cr content posses the high temperature strength and l... more Ferritic/Martensitic (F/M) steels with high Cr content posses the high temperature strength and low swelling rates required for advanced nuclear reactor designs. Radiation induced segregation (RIS) occurs in F/M steels due to solute atoms preferentially coupling to point defect fluxes which migrate to defect sinks, such as grain boundaries (GBs). The RIS response of F/M steels and austenitic steels has been shown to be dependent on the local structure of GBs where low energy structures have suppressed RIS responses. This relationship between local GB structure and RIS has been demonstrated primarily in ion-irradiated specimens. A 9 wt.% Cr model alloy steel was irradiated to 3 dpa using neutrons at the Advanced Test Reactor (ATR) to determine the effect of a neutron radiation environment on the RIS response at different GB structures. This investigation found the relationship between GB structure and RIS is also active for F/M steels irradiated using neutrons. The data generated from the neutron irradiation is also compared to RIS data generated using proton irradiations on the same heat of model alloy.

Research paper thumbnail of Thermomechanical treatment for improved neutron irradiation resistance of austenitic alloy (Fe–21Cr–32Ni)

Journal of Nuclear Materials, 2013

An optimized thermomecha nical treatment (TMT) applied to austenitic alloy 800H (Fe-21Cr-32Ni) ha... more An optimized thermomecha nical treatment (TMT) applied to austenitic alloy 800H (Fe-21Cr-32Ni) had shown significant improvements in corrosion resistance and basic mechanical properties. This study examine d its effect on radiation resistance by irradiating both the solution-annealed (SA) and TMT samples at 500 °C for 3 dpa. Microstructural characterization using transmission electron microscopy revealed that the radiation-induced Frank loops, voids, and c 0-Ni 3 (Ti,Al) precipitates had similar sizes between the SA and TMT samples. The amounts of radiation-induced defects and more significantly c 0 precipita tes, however, were reduced in the TMT samples. These reductions would approximately reduce by 40.9% the radiation hardening compared to the SA samples. This study indicates that optimized-TMT is an economical approach for effective overall property improvements.

Research paper thumbnail of Baseline Postirradiation Examination of Fuel Rodlets from the AFC-2C Experiment

Research paper thumbnail of Accident Tolerant Fuel Analysis

Research paper thumbnail of Summary of FUTURIX-FTA Non-Destructive Examination

Research paper thumbnail of Development of Metallic Fuels for Actinide Transmutation

Research paper thumbnail of Light Water Reactor Accident Tolerant Fuels Irradiation Testing

The purpose of Accident Tolerant Fuels (ATF) experiments is to test novel fuel and cladding conce... more The purpose of Accident Tolerant Fuels (ATF) experiments is to test novel fuel and cladding concepts designed to replace the current zirconium alloy uranium dioxide (UO2) fuel system. The objective of this Research and Development (R&D) is to develop novel ATF concepts that will be able to withstand loss of active cooling in the reactor core for a considerably longer time period than the current fuel system while maintaining or improving the fuel performance during normal operations, operational transients, design basis, and beyond design basis events. It was necessary to design, analyze, and fabricate drop-in capsules to meet the requirements for testing under prototypic LWR temperatures in Idaho National Laboratory's Advanced Test Reactor (ATR). Three industry led teams and one DOE team from Oak Ridge National Laboratory provided fuel rodlet samples for their new concepts for ATR insertion in 2015. As-built projected temperature calculations were performed on the ATF capsules ...

Research paper thumbnail of Microstructural effect on neutron irradiation response of alloy 800H

Research paper thumbnail of Metallography and fuel cladding chemical interaction in fast flux test facility irradiated metallic U-10Zr MFF-3 and MFF-5 fuel pins

Journal of Nuclear Materials, 2016

Abstract The Mechanistic Fuel Failure (MFF) series of metal fuel irradiations conducted in the Fa... more Abstract The Mechanistic Fuel Failure (MFF) series of metal fuel irradiations conducted in the Fast Flux Test Facility (FFTF) provides an important comparison between data generated in the Experimental Breeder Reactor (EBR-II) and that expected in a larger-scale fast reactor. The MFF fuel operated with a peak cladding temperature at the top of the fuel column, but developed peak burnup at the centerline of the core. This places the peak fuel temperature midway between the core center and the top of fuel, lower in the fuel column than in EBR-II experiments. Data from the MFF-3 and MFF-5 assemblies are most comparable to the data obtained from the EBR-II X447 experiment. The two X447 pin breaches were strongly influenced by fuel/cladding chemical interaction (FCCI) at the top of the fuel column. Post irradiation examination data from MFF-3 and MFF-5 are presented and compared to historical EBR-II data.

Research paper thumbnail of Performance of low smeared density sodium-cooled fast reactor metal fuel

Journal of Nuclear Materials

An experiment was performed in the Experimental Breeder Rector-II (EBR-II) in the 1990s to show t... more An experiment was performed in the Experimental Breeder Rector-II (EBR-II) in the 1990s to show that metallic fast reactor fuel could be used in reactors with a single, once-through core. To prove the long duration, high burnup, high neutron exposure capability an experiment where the fuel pin was designed with a very large fission gas plenum and very low fuel smeared density (SD). The experiment, X496, operated to only 8.3 at. % burnup because the EBR-II reactor was scheduled for shutdown at that time. Many of the examinations of the fuel pins only funded recently with the resurgence of reactor designs using very high-burnup fuel. The results showed that, despite the low smeared density of 59% the fuel swelled radially to contact the cladding, fission gas release appeared to be slightly higher than demonstrated in conventional 75%SD fuel tests and axial growth was about the same as 75% SD fuel. There were axial positions in some of the fuel pins which showed evidence of fuel restructuring and an absence of fission products with low melting points and gaseous precursors (Cs and Rb). A model to investigate whether these areas may have overheated due to a loss of bond sodium indicates that it is a possible explanation for the fuel restructuring and something to be considered for fuel performance modeling of low SD fuel.

Research paper thumbnail of Baseline Postirradiation Examination of the FUTURIX-FTA Experiments

Research paper thumbnail of Examination of Legacy Metallic Fuel Pins (U-10Zr) Tested in FFTF

Research paper thumbnail of Postirradiation examination results of several metallic fuel alloys and forms from low burnup AFC irradiations

Journal of Nuclear Materials

Research paper thumbnail of Postirradiation examination on metallic fuel in the AFC-2 irradiation test series

Journal of Nuclear Materials

Research paper thumbnail of Postirradiation Examination of FUTURIX-FTA metallic alloy experiments

Journal of Nuclear Materials

Research paper thumbnail of Advanced Reactor Fuels Irradiation Experiment Design Objectives

Research paper thumbnail of Microstructural effect on neutron irradiation response of alloy 800H

Research paper thumbnail of Use of Silicon Carbide Monitors in ATR Irradiation Testing

Research paper thumbnail of Microstructural effect on neutron irradiation response of alloy 800H

Research paper thumbnail of Hot Cell Sample Preparation of Metallic Nuclear Fuels

Microscopy and Microanalysis, 2012

Research paper thumbnail of Relationship between lath boundary structure and radiation induced segregation in a neutron irradiated 9wt.% Cr model ferritic/martensitic steel

Journal of Nuclear Materials, 2014

Ferritic/Martensitic (F/M) steels with high Cr content posses the high temperature strength and l... more Ferritic/Martensitic (F/M) steels with high Cr content posses the high temperature strength and low swelling rates required for advanced nuclear reactor designs. Radiation induced segregation (RIS) occurs in F/M steels due to solute atoms preferentially coupling to point defect fluxes which migrate to defect sinks, such as grain boundaries (GBs). The RIS response of F/M steels and austenitic steels has been shown to be dependent on the local structure of GBs where low energy structures have suppressed RIS responses. This relationship between local GB structure and RIS has been demonstrated primarily in ion-irradiated specimens. A 9 wt.% Cr model alloy steel was irradiated to 3 dpa using neutrons at the Advanced Test Reactor (ATR) to determine the effect of a neutron radiation environment on the RIS response at different GB structures. This investigation found the relationship between GB structure and RIS is also active for F/M steels irradiated using neutrons. The data generated from the neutron irradiation is also compared to RIS data generated using proton irradiations on the same heat of model alloy.

Research paper thumbnail of Thermomechanical treatment for improved neutron irradiation resistance of austenitic alloy (Fe–21Cr–32Ni)

Journal of Nuclear Materials, 2013

An optimized thermomecha nical treatment (TMT) applied to austenitic alloy 800H (Fe-21Cr-32Ni) ha... more An optimized thermomecha nical treatment (TMT) applied to austenitic alloy 800H (Fe-21Cr-32Ni) had shown significant improvements in corrosion resistance and basic mechanical properties. This study examine d its effect on radiation resistance by irradiating both the solution-annealed (SA) and TMT samples at 500 °C for 3 dpa. Microstructural characterization using transmission electron microscopy revealed that the radiation-induced Frank loops, voids, and c 0-Ni 3 (Ti,Al) precipitates had similar sizes between the SA and TMT samples. The amounts of radiation-induced defects and more significantly c 0 precipita tes, however, were reduced in the TMT samples. These reductions would approximately reduce by 40.9% the radiation hardening compared to the SA samples. This study indicates that optimized-TMT is an economical approach for effective overall property improvements.

Research paper thumbnail of Baseline Postirradiation Examination of Fuel Rodlets from the AFC-2C Experiment

Research paper thumbnail of Accident Tolerant Fuel Analysis