Jian Gan - Academia.edu (original) (raw)

Papers by Jian Gan

Microscopy and Microanalysis, 2020

Microscopy and Microanalysis, 2019

Journal of Nuclear Materials, 2017

Herein we report on the formation of defects in response to neutron irradiation of polycrystallin... more Herein we report on the formation of defects in response to neutron irradiation of polycrystalline Ti 3 SiC 2 and Ti 3 AlC 2 samples exposed to total fluences of ≈ 6×10 20 n/m 2 , 5×10 21 n/m 2 and 1.7×10 22 n/m 2 at irradiation temperatures of 121(12), 735(6) and 1085(68) °C. These fluences correspond to 0.14, 1.6 and 3.4 dpa, respectively. After irradiation to 0.14 dpa at 121 °C and 735 °C, black spots are observed via transmission electron microscopy in both Ti 3 SiC 2 and Ti 3 AlC 2. After irradiation to 1.6 and 3.4 dpa at 735 °C, basal dislocation loops, with a Burgers vector of b = ½ [0001] are observed in Ti 3 SiC 2 , with loop diameters of 21(6) and 30(8) nm after 1.6 dpa and 3.4 dpa, respectively. In Ti 3 AlC 2 , larger dislocation loops, 75(34) nm in diameter are observed after 3.4 dpa at 735 °C, in addition to stacking faults. Impurity particles of TiC, as well as stacking fault TiC platelets in the MAX phases, are seen to form extensive dislocation loops under all conditions. Cavities were observed at grain boundaries and within stacking faults after 3.4 dpa irradiation, with extensive cavity formation in the TiC regions at 1085 °C. Remarkably, denuded zones on the order of 1 μm are observed in Ti 3 SiC 2 after irradiation to 3.4 dpa at 735 °C. Small grains, 3-5 μm in diameter, are damage free after irradiation at 1085 °C at this dose. The results shown herein confirm once again that the presence of the A-layers in the MAX phases considerably enhance their irradiation tolerance. Based on these results, and up to 3.4 dpa, Ti 3 SiC 2 remains a promising candidate for high temperature nuclear applications as long as the temperature remains > 700 °C.

JOM, 2017

A low-enriched uranium U-10Mo monolithic nuclear fuel is being developed by the Material Manageme... more A low-enriched uranium U-10Mo monolithic nuclear fuel is being developed by the Material Management and Minimization Program, earlier known as the Reduced Enrichment for Research and Test Reactors Program, for utilization in research and test reactors around the world that currently use high-enriched uranium fuels. As part of this program, reactor experiments are being performed in the Advanced Test Reactor. It must be demonstrated that this fuel type exhibits mechanical integrity, geometric stability, and predictable behavior to high powers and high fission densities in order for it to be a viable fuel for qualification. This paper provides an overview of the microstructures observed at different regions of interest in fuel plates before and after irradiation for fuel samples that have been tested. These fuel plates were fabricated using laboratory-scale fabrication methods. Observations regarding how microstructural changes during irradiation may impact fuel performance are discussed.

Scripta Materialia, 2016

Here synchrotron scattering analysis results on U-7wt%Mo fuel specimens irradiated in the Advance... more Here synchrotron scattering analysis results on U-7wt%Mo fuel specimens irradiated in the Advanced Test Reactor to three burnup levels (3.0, 5.2, and 6.3×10 21 fission/cm 3) are reported. Mature fission gas bubble superlattice was observed to form at intermediate burnup. The superlattice constant was determined to be 11.7 and 12.0 nm by wide-angle and small-angle scattering respectively. Grain subdivision takes place throughout the irradiation and causes the collapse of the superlattice at high burnup. The bubble superlattice expands the U-Mo lattice and acts as strong sink for radiation-induced defects. The evolution of dislocation loops was, therefore, suppressed until the bubble superlattice collapsed.

The silicide dispersion fuel of U3Si2/Al has been recognized as a reasonably good performance fue... more The silicide dispersion fuel of U3Si2/Al has been recognized as a reasonably good performance fuel for nuclear research and test reactors except that it requires the use of high enrichment uranium. An irradiated U3Si2/Al dispersion fuel (~75% enrichment) from the high flux side of a RERTR-8 (U0R040) plate was characterized using transmission electron microscopy (TEM). The fuel plate was irradiated

Metallurgical and Materials Transactions E, 2015

Low-enrichment (235 U < 20 pct) U-Mo monolithic fuel is being developed for use in research and t... more Low-enrichment (235 U < 20 pct) U-Mo monolithic fuel is being developed for use in research and test reactors. The earliest design for this fuel that was investigated via reactor testing consisted of a nominally U-10Mo fuel foil encased in AA6061 (Al-6061) cladding. For a fuel design to be deemed adequate for final use in a reactor, it must maintain dimensional stability and retain fission products throughout irradiation, which means that there must be good integrity at the fuel foil/cladding interface. To investigate the nature of the fuel/cladding interface for this fuel type after irradiation, fuel plates were fabricated using a friction bonding process, tested in INL's advanced test reactor (ATR), and then subsequently characterized using optical metallography, scanning electron microscopy, and transmission electron microscopy. Results of this characterization showed that the fuel/cladding interaction layers present at the U-Mo fuel/AA6061 cladding interface after fabrication became amorphous during irradiation. Up to two main interaction layers, based on composition, could be found at the fuel/cladding interface, depending on location. After irradiation, an Al-rich layer contained very few fission gas bubbles, but did exhibit Xe enrichment near the AA6061 cladding interface. Another layer, which contained more Si, had more observable fission gas bubbles. In the samples produced using a focused ion beam at the interaction zone/AA6061 cladding interface, possible indications of porosity/debonding were found, which suggested that the interface in this location is relatively weak.

Journal of Nuclear Materials, 2015

Atom probe tomography (APT) was performed to study the effects of Cr concentrations, irradiation ... more Atom probe tomography (APT) was performed to study the effects of Cr concentrations, irradiation doses and irradiation temperatures on α′ phase formation in Fe-Cr model alloys (10-16 at.%) irradiated at 300 and 450°C to 0.01, 0.1 and 1 dpa. For 1 dpa specimens, α′ precipitates with an average radius of 1.0-1.3 nm were observed. The precipitate density varied significantly from 1.1x10 23 to 2.7x10 24 1/m 3 , depending on Cr concentrations and irradiation temperatures. The volume fraction of α′ phase in 1 dpa specimens qualitatively agreed with the phase diagram prediction. For 0.01 dpa and 0.1 dpa, frequency distribution analysis detected slight Cr segregation in high-Cr specimens, but not in Fe-10Cr specimens. Proximity histogram analysis showed that the radial Cr concentration was highest at the center of α′ precipitates. For most precipitates, the Cr contents were significantly lower than that predicted by the phase diagram. The Cr concentration at precipitate center increased with increasing precipitate size.

Journal of Nuclear Materials, 2015

In situ and ex situ transmission electron microscopy observation of small Kr bubbles in both sing... more In situ and ex situ transmission electron microscopy observation of small Kr bubbles in both single-crystal and polycrystalline UO 2 were conducted to understand the inert gas bubble behavior in oxide nuclear fuel. The bubble size and volume swelling are shown as weak functions of ion dose but strongly depend on the temperature. The Kr bubble formation at room temperature was observed for the first time. The depth profiles of implanted Kr determined by atom probe tomography are in good agreement with the calculated profiles by SRIM, but the measured concentration of Kr is about 1/4 of the calculated concentration. This difference is mainly due to low solubility of Kr in UO 2 matrix and high release of Kr from sample surface under irradiation.

The need to provide more accurate property information on U-Mo fuel alloys to reactor operators, ... more The need to provide more accurate property information on U-Mo fuel alloys to reactor operators, modelers, researchers, fabricators, and regulators increases as success of the RERTR program continues. This presentation will provide an overview of fresh fuel U-Mo characterization activities on monolithic fuel occurring at the Idaho National Laboratory. The overview will particularly be focused on properties available through current

Journal of Nuclear Materials, 2014

The microstructural changes and associated effects on thermal conductivity were examined in UO 2 ... more The microstructural changes and associated effects on thermal conductivity were examined in UO 2 after irradiation using 3.9 MeV He 2+ ions. Lattice expansion of UO 2 was observed in x-ray diffraction after ion irradiation up to 5×10 16 He 2+ /cm 2 at low-temperature (< 200 °C). Transmission electron microscopy (TEM) showed homogenous irradiation damage across an 8 µm thick plateau region, which consisted of small dislocation loops accompanied by dislocation segments. Dome-shaped blisters were observed at the peak damage region (depth around 8.5 µm) in the sample subjected to 5×10 16 He 2+ /cm 2 , the highest fluence reached, while similar features were not detected at 9×10 15 He 2+ /cm 2. Laser-based thermo-reflectance measurements showed that the thermal conductivity for the irradiated layer decreased about 55 % for the high fluence sample and 35% for the low fluence sample as compared to an un-irradiated reference sample. Detailed analysis for the thermal conductivity indicated that the conductivity reduction was caused by the irradiation induced point defects.

This project has elucidated approaches for managing damage evolution through control of matrix so... more This project has elucidated approaches for managing damage evolution through control of matrix solute and multiphase structures. Radiation microstructures, microchemistries and properties affected by irradiation were evaluated with the goal of developing damage-resistant alloys for next generation nuclear power systems. Three complementary experimental approaches identified key factors for the development of alloys: Ni ++ irradiation, proton irradiation and manipulation of simulated irradiation characteristics in non-irradiated alloys. Specifically, alloys processed with oversized solute, i.e., Hf, and multiphase alloys show promise for retarding damage evolution.

Journal of Nuclear Materials, 2004

The effect of the oversized hafnium or platinum (0.3 at.%) solutes on the evolution of irradiated... more The effect of the oversized hafnium or platinum (0.3 at.%) solutes on the evolution of irradiated microstructure in 316SS was investigated. Irradiations were conducted with 5 MeV Ni-ions at 500 °C to doses up to 50 dpa or with 3.2 MeV protons at 400 °C to a dose of 2.5 dpa, and previous studies demonstrated that these irradiations are capable of producing similar irradiated microstructure and microchemistry relevant to LWR cores. Microstructures of 316SS, 316SS+0.3 at.% Pt and 316SS+0.3 at.% Hf were characterized using transmission electron microscopy. The addition of Hf showed a strong effect in suppressing radiation-induced microstructure evolution; no voids were observed at doses up to 50 dpa for Ni-ion irradiation and 2.5 dpa for proton irradiation. The mean diameter of the Frank loops in the Hf-doped samples is about 40% smaller than loops in 316SS. The microstructural examinations from both types of particle irradiation revealed that for 0.3 at.% Pt addition there is no beneficial effect on irradiated microstructure. The mechanisms for the role of oversize solute additions on the microstructure evolution are discussed.

Journal of Nuclear Materials, 2001

ABSTRACT

Journal of Nuclear Materials, 2011

This is a preprint of a paper intended for publication in a journal or proceedings. Since changes... more This is a preprint of a paper intended for publication in a journal or proceedings. Since changes may be made before publication, this preprint should not be cited or reproduced without permission of the author. This document was prepared as an account of work sponsored by an agency of the United States Government. Neither the United States Government nor any agency thereof, or any of their employees, makes any warranty, expressed or implied, or assumes any legal liability or responsibility for any third party's use, or the results of such use, of any information, apparatus, product or process disclosed in this report, or represents that its use by such third party would not infringe privately owned rights. The views expressed in this paper are not necessarily those of the United States Government or the sponsoring agency.

Journal of Nuclear Materials, 2007

This is a preprint of a paper intended for publication in a journal or proceedings. Since changes... more This is a preprint of a paper intended for publication in a journal or proceedings. Since changes may be made before publication, this preprint should not be cited or reproduced without permission of the author. This document was prepared as an account of work sponsored by an agency of the United States Government. Neither the United States Government nor any agency thereof, or any of their employees, makes any warranty, expressed or implied, or assumes any legal liability or responsibility for any third party's use, or the results of such use, of any information, apparatus, product or process disclosed in this report, or represents that its use by such third party would not infringe privately owned rights. The views expressed in this paper are not necessarily those of the United States Government or the sponsoring agency.

DOE Scientific and Technical Information. DOE Scientific and Technical Information. ...

Microscopy and Microanalysis, 2020

Microscopy and Microanalysis, 2019

Journal of Nuclear Materials, 2017

Herein we report on the formation of defects in response to neutron irradiation of polycrystallin... more Herein we report on the formation of defects in response to neutron irradiation of polycrystalline Ti 3 SiC 2 and Ti 3 AlC 2 samples exposed to total fluences of ≈ 6×10 20 n/m 2 , 5×10 21 n/m 2 and 1.7×10 22 n/m 2 at irradiation temperatures of 121(12), 735(6) and 1085(68) °C. These fluences correspond to 0.14, 1.6 and 3.4 dpa, respectively. After irradiation to 0.14 dpa at 121 °C and 735 °C, black spots are observed via transmission electron microscopy in both Ti 3 SiC 2 and Ti 3 AlC 2. After irradiation to 1.6 and 3.4 dpa at 735 °C, basal dislocation loops, with a Burgers vector of b = ½ [0001] are observed in Ti 3 SiC 2 , with loop diameters of 21(6) and 30(8) nm after 1.6 dpa and 3.4 dpa, respectively. In Ti 3 AlC 2 , larger dislocation loops, 75(34) nm in diameter are observed after 3.4 dpa at 735 °C, in addition to stacking faults. Impurity particles of TiC, as well as stacking fault TiC platelets in the MAX phases, are seen to form extensive dislocation loops under all conditions. Cavities were observed at grain boundaries and within stacking faults after 3.4 dpa irradiation, with extensive cavity formation in the TiC regions at 1085 °C. Remarkably, denuded zones on the order of 1 μm are observed in Ti 3 SiC 2 after irradiation to 3.4 dpa at 735 °C. Small grains, 3-5 μm in diameter, are damage free after irradiation at 1085 °C at this dose. The results shown herein confirm once again that the presence of the A-layers in the MAX phases considerably enhance their irradiation tolerance. Based on these results, and up to 3.4 dpa, Ti 3 SiC 2 remains a promising candidate for high temperature nuclear applications as long as the temperature remains > 700 °C.

JOM, 2017

A low-enriched uranium U-10Mo monolithic nuclear fuel is being developed by the Material Manageme... more A low-enriched uranium U-10Mo monolithic nuclear fuel is being developed by the Material Management and Minimization Program, earlier known as the Reduced Enrichment for Research and Test Reactors Program, for utilization in research and test reactors around the world that currently use high-enriched uranium fuels. As part of this program, reactor experiments are being performed in the Advanced Test Reactor. It must be demonstrated that this fuel type exhibits mechanical integrity, geometric stability, and predictable behavior to high powers and high fission densities in order for it to be a viable fuel for qualification. This paper provides an overview of the microstructures observed at different regions of interest in fuel plates before and after irradiation for fuel samples that have been tested. These fuel plates were fabricated using laboratory-scale fabrication methods. Observations regarding how microstructural changes during irradiation may impact fuel performance are discussed.

Scripta Materialia, 2016

Here synchrotron scattering analysis results on U-7wt%Mo fuel specimens irradiated in the Advance... more Here synchrotron scattering analysis results on U-7wt%Mo fuel specimens irradiated in the Advanced Test Reactor to three burnup levels (3.0, 5.2, and 6.3×10 21 fission/cm 3) are reported. Mature fission gas bubble superlattice was observed to form at intermediate burnup. The superlattice constant was determined to be 11.7 and 12.0 nm by wide-angle and small-angle scattering respectively. Grain subdivision takes place throughout the irradiation and causes the collapse of the superlattice at high burnup. The bubble superlattice expands the U-Mo lattice and acts as strong sink for radiation-induced defects. The evolution of dislocation loops was, therefore, suppressed until the bubble superlattice collapsed.

The silicide dispersion fuel of U3Si2/Al has been recognized as a reasonably good performance fue... more The silicide dispersion fuel of U3Si2/Al has been recognized as a reasonably good performance fuel for nuclear research and test reactors except that it requires the use of high enrichment uranium. An irradiated U3Si2/Al dispersion fuel (~75% enrichment) from the high flux side of a RERTR-8 (U0R040) plate was characterized using transmission electron microscopy (TEM). The fuel plate was irradiated

Metallurgical and Materials Transactions E, 2015

Low-enrichment (235 U < 20 pct) U-Mo monolithic fuel is being developed for use in research and t... more Low-enrichment (235 U < 20 pct) U-Mo monolithic fuel is being developed for use in research and test reactors. The earliest design for this fuel that was investigated via reactor testing consisted of a nominally U-10Mo fuel foil encased in AA6061 (Al-6061) cladding. For a fuel design to be deemed adequate for final use in a reactor, it must maintain dimensional stability and retain fission products throughout irradiation, which means that there must be good integrity at the fuel foil/cladding interface. To investigate the nature of the fuel/cladding interface for this fuel type after irradiation, fuel plates were fabricated using a friction bonding process, tested in INL's advanced test reactor (ATR), and then subsequently characterized using optical metallography, scanning electron microscopy, and transmission electron microscopy. Results of this characterization showed that the fuel/cladding interaction layers present at the U-Mo fuel/AA6061 cladding interface after fabrication became amorphous during irradiation. Up to two main interaction layers, based on composition, could be found at the fuel/cladding interface, depending on location. After irradiation, an Al-rich layer contained very few fission gas bubbles, but did exhibit Xe enrichment near the AA6061 cladding interface. Another layer, which contained more Si, had more observable fission gas bubbles. In the samples produced using a focused ion beam at the interaction zone/AA6061 cladding interface, possible indications of porosity/debonding were found, which suggested that the interface in this location is relatively weak.

Journal of Nuclear Materials, 2015

Atom probe tomography (APT) was performed to study the effects of Cr concentrations, irradiation ... more Atom probe tomography (APT) was performed to study the effects of Cr concentrations, irradiation doses and irradiation temperatures on α′ phase formation in Fe-Cr model alloys (10-16 at.%) irradiated at 300 and 450°C to 0.01, 0.1 and 1 dpa. For 1 dpa specimens, α′ precipitates with an average radius of 1.0-1.3 nm were observed. The precipitate density varied significantly from 1.1x10 23 to 2.7x10 24 1/m 3 , depending on Cr concentrations and irradiation temperatures. The volume fraction of α′ phase in 1 dpa specimens qualitatively agreed with the phase diagram prediction. For 0.01 dpa and 0.1 dpa, frequency distribution analysis detected slight Cr segregation in high-Cr specimens, but not in Fe-10Cr specimens. Proximity histogram analysis showed that the radial Cr concentration was highest at the center of α′ precipitates. For most precipitates, the Cr contents were significantly lower than that predicted by the phase diagram. The Cr concentration at precipitate center increased with increasing precipitate size.

Journal of Nuclear Materials, 2015

In situ and ex situ transmission electron microscopy observation of small Kr bubbles in both sing... more In situ and ex situ transmission electron microscopy observation of small Kr bubbles in both single-crystal and polycrystalline UO 2 were conducted to understand the inert gas bubble behavior in oxide nuclear fuel. The bubble size and volume swelling are shown as weak functions of ion dose but strongly depend on the temperature. The Kr bubble formation at room temperature was observed for the first time. The depth profiles of implanted Kr determined by atom probe tomography are in good agreement with the calculated profiles by SRIM, but the measured concentration of Kr is about 1/4 of the calculated concentration. This difference is mainly due to low solubility of Kr in UO 2 matrix and high release of Kr from sample surface under irradiation.

The need to provide more accurate property information on U-Mo fuel alloys to reactor operators, ... more The need to provide more accurate property information on U-Mo fuel alloys to reactor operators, modelers, researchers, fabricators, and regulators increases as success of the RERTR program continues. This presentation will provide an overview of fresh fuel U-Mo characterization activities on monolithic fuel occurring at the Idaho National Laboratory. The overview will particularly be focused on properties available through current

Journal of Nuclear Materials, 2014

The microstructural changes and associated effects on thermal conductivity were examined in UO 2 ... more The microstructural changes and associated effects on thermal conductivity were examined in UO 2 after irradiation using 3.9 MeV He 2+ ions. Lattice expansion of UO 2 was observed in x-ray diffraction after ion irradiation up to 5×10 16 He 2+ /cm 2 at low-temperature (< 200 °C). Transmission electron microscopy (TEM) showed homogenous irradiation damage across an 8 µm thick plateau region, which consisted of small dislocation loops accompanied by dislocation segments. Dome-shaped blisters were observed at the peak damage region (depth around 8.5 µm) in the sample subjected to 5×10 16 He 2+ /cm 2 , the highest fluence reached, while similar features were not detected at 9×10 15 He 2+ /cm 2. Laser-based thermo-reflectance measurements showed that the thermal conductivity for the irradiated layer decreased about 55 % for the high fluence sample and 35% for the low fluence sample as compared to an un-irradiated reference sample. Detailed analysis for the thermal conductivity indicated that the conductivity reduction was caused by the irradiation induced point defects.

This project has elucidated approaches for managing damage evolution through control of matrix so... more This project has elucidated approaches for managing damage evolution through control of matrix solute and multiphase structures. Radiation microstructures, microchemistries and properties affected by irradiation were evaluated with the goal of developing damage-resistant alloys for next generation nuclear power systems. Three complementary experimental approaches identified key factors for the development of alloys: Ni ++ irradiation, proton irradiation and manipulation of simulated irradiation characteristics in non-irradiated alloys. Specifically, alloys processed with oversized solute, i.e., Hf, and multiphase alloys show promise for retarding damage evolution.

Journal of Nuclear Materials, 2004

The effect of the oversized hafnium or platinum (0.3 at.%) solutes on the evolution of irradiated... more The effect of the oversized hafnium or platinum (0.3 at.%) solutes on the evolution of irradiated microstructure in 316SS was investigated. Irradiations were conducted with 5 MeV Ni-ions at 500 °C to doses up to 50 dpa or with 3.2 MeV protons at 400 °C to a dose of 2.5 dpa, and previous studies demonstrated that these irradiations are capable of producing similar irradiated microstructure and microchemistry relevant to LWR cores. Microstructures of 316SS, 316SS+0.3 at.% Pt and 316SS+0.3 at.% Hf were characterized using transmission electron microscopy. The addition of Hf showed a strong effect in suppressing radiation-induced microstructure evolution; no voids were observed at doses up to 50 dpa for Ni-ion irradiation and 2.5 dpa for proton irradiation. The mean diameter of the Frank loops in the Hf-doped samples is about 40% smaller than loops in 316SS. The microstructural examinations from both types of particle irradiation revealed that for 0.3 at.% Pt addition there is no beneficial effect on irradiated microstructure. The mechanisms for the role of oversize solute additions on the microstructure evolution are discussed.

Journal of Nuclear Materials, 2001

ABSTRACT

Journal of Nuclear Materials, 2011

This is a preprint of a paper intended for publication in a journal or proceedings. Since changes... more This is a preprint of a paper intended for publication in a journal or proceedings. Since changes may be made before publication, this preprint should not be cited or reproduced without permission of the author. This document was prepared as an account of work sponsored by an agency of the United States Government. Neither the United States Government nor any agency thereof, or any of their employees, makes any warranty, expressed or implied, or assumes any legal liability or responsibility for any third party's use, or the results of such use, of any information, apparatus, product or process disclosed in this report, or represents that its use by such third party would not infringe privately owned rights. The views expressed in this paper are not necessarily those of the United States Government or the sponsoring agency.

Journal of Nuclear Materials, 2007

This is a preprint of a paper intended for publication in a journal or proceedings. Since changes... more This is a preprint of a paper intended for publication in a journal or proceedings. Since changes may be made before publication, this preprint should not be cited or reproduced without permission of the author. This document was prepared as an account of work sponsored by an agency of the United States Government. Neither the United States Government nor any agency thereof, or any of their employees, makes any warranty, expressed or implied, or assumes any legal liability or responsibility for any third party's use, or the results of such use, of any information, apparatus, product or process disclosed in this report, or represents that its use by such third party would not infringe privately owned rights. The views expressed in this paper are not necessarily those of the United States Government or the sponsoring agency.

DOE Scientific and Technical Information. DOE Scientific and Technical Information. ...