TEM investigation of irradiation, irradiation creep and thermal annealing effects in nuclear graphite
Principal Investigator
- Name:
- Timothy Burchell
- Email:
- [email protected]
- Phone:
- (208) 526-6918
Team Members:
| Name: | Institution: | Expertise: | Status: |
|---|---|---|---|
| Jose D Arregui-Mena | Oak Ridge National Laboratory | J. David Arregui Mena is a post-doctoral researcher in Nuclear Materials Science and Technology group at Oak Ridge National Laboratory where he is developing new techniques to measure the effects of irradiation in the microstructure of graphite. He is also investigating other carbon base materials such as SiC and carbon fiber composites. | Post Doc |
| Philip D Edmondson | Oak Ridge National Laboratory | Philip Edmondson is Radiation Effects Characterization Team Leader in the Nuclear Materials Science and Technology Group at ORNL in which he uses techniques such as advanced electron microscopy and atom probe tomography to characterize radiation effects in metals, ceramics and composites. Prior to this, he was an Engineering and Physical Sciences Research Council (EPSRC) Career Acceleration Fellow at Oxford University (UK) specializing in the characterization of radiation effects in materials. He has published approximately 60 papers on radiation effects in materials and is an Associate Editor of the journal Nuclear Materials and Energy. | Other |
Experiment Details:
- Experiment Title:
- TEM investigation of irradiation, irradiation creep and thermal annealing effects in nuclear graphite)
- Work Description:
- The PCEA graphite specimens are part of the Advanced Graphite Creep-1 (AGC-1), the irradiation conditions are: An unirradiated sample, 2 crept irradiated samples (4.36 and 5.66 dpa irradiated at a temperature of 606°C and 669 °C respectively) and 2 uncrept irradiated samples (3.91 and 5.02 dpa at a nominal temperature of 533°C and 585 °C respectively). High magnification and atomic resolution TEM images will be used to characterize microstructure evolution of graphite under irradiation as well as crystal lattice defects. Moreover, the post-annealing imaging will be performed to identify the temperatures at which the recovery process starts. Diffraction patterns will be used to quantify the crystallinity of the samples at each stage of irradiation and annealing condition.
Project Summary
The Advanced Graphite Creep-1 (AGC-1) program was developed to qualify six types (NBG-18, NBG-17, H-451, PCEA, IG110, IG-430) of graphite grades for their use in Gen IV nuclear reactor designs or to produce a reference database. The irradiation and creep conditions were performed at a similar range of temperatures and dose as a Gen IV graphite reactor core. The atomic-level and microstructure data of graphite under this irradiation conditions are scarce, limiting the understanding between the irradiation defects and bulk material properties. This research selected samples irradiated at different stages of irradiation and creep to understand the mechanisms that alter the microstructure and crystal lattice of graphite.
PCEA graphite samples will be used to determine the atomic-level structure, microstructural and irradiation defects of 7 samples. A unirradiated sample, 2 crept irradiated samples (4.36 and 5.66 dpa at a nominal temperature of 700°C) and 2 uncrept irradiated samples (3.9 and 5.02 dpa at a nominal temperature of 700°C) were selected for this study. SEM and TEM imaging will be used to characterize the different microstructures and types of defects found in each sample. Moreover, an in-situ annealing of the graphite specimens will be performed (1200°C) to understand the annealing effects on each specimen.
PCEA graphite samples will be used to determine the atomic-level structure, microstructural and irradiation defects of 7 samples. A unirradiated sample, 2 crept irradiated samples (4.36 and 5.66 dpa at a nominal temperature of 700°C) and 2 uncrept irradiated samples (3.9 and 5.02 dpa at a nominal temperature of 700°C) were selected for this study. SEM and TEM imaging will be used to characterize the different microstructures and types of defects found in each sample. Moreover, an in-situ annealing of the graphite specimens will be performed (1200°C) to understand the annealing effects on each specimen.
Relevance
This work is a microstructural study of nuclear graphite which is the moderator of High Temperature Reactors (HTR) such as Gas-cooled Reactor (GCR) and Molten Salt Reactors (MSR). In both systems the moderator will be subjected to irradiation damage and creep this work will enhance our understating on the interaction between irradiation damage-creep and pore structure. The irradiated samples used in this research were provided by DOE-NE Advanced Reactors Program (ARP) by Oak Ridge National Lab.
The proposed study will characterize the microstructural changes and defects produced by irradiation and creep on graphite. TEM and SEM images will provide atomic and microstructure information that will improve the understanding of irradiated graphite. Furthermore, a novel study of annealing of irradiated graphite will improve the investigations of irradiation damage on graphite. Graphite components in Very-high Temperature Reactors are expected to withheld high temperature gradients that are not fully understand at an atomistic and microstructural level. This research will support the DOE funded ongoing irradiation campaigns of nuclear graphite at Oak Ridge National Laboratory and Idaho National Laboratory.
Improve understanding of creep is important for graphite moderated reactor operation and safety. A gap in our knowledge base exists because due to the lack of publications and data on the microstructural changes that occur in irradiated and crept graphite. The results of this research would be of considerable interest to the DOE (GCR) Program.
The proposed study will characterize the microstructural changes and defects produced by irradiation and creep on graphite. TEM and SEM images will provide atomic and microstructure information that will improve the understanding of irradiated graphite. Furthermore, a novel study of annealing of irradiated graphite will improve the investigations of irradiation damage on graphite. Graphite components in Very-high Temperature Reactors are expected to withheld high temperature gradients that are not fully understand at an atomistic and microstructural level. This research will support the DOE funded ongoing irradiation campaigns of nuclear graphite at Oak Ridge National Laboratory and Idaho National Laboratory.
Improve understanding of creep is important for graphite moderated reactor operation and safety. A gap in our knowledge base exists because due to the lack of publications and data on the microstructural changes that occur in irradiated and crept graphite. The results of this research would be of considerable interest to the DOE (GCR) Program.
Book / Journal Publications
"Characterisation of the spatial variability of material properties of Gilsocarbon and NBG-18 using random fields" Jose Arregui-Mena, Philip Edmondson, Lee Margetts, DV Griffiths, William Windes, Mark Carroll, Paul Mummery, Journal of Nuclear Materials 511 2018 91-108 Link
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About Us
The Nuclear Science User Facilities (NSUF) is the U.S. Department of Energy Office of Nuclear Energy's only designated nuclear energy user facility. Through peer-reviewed proposal processes, the NSUF provides researchers access to neutron, ion, and gamma irradiations, post-irradiation examination and beamline capabilities at Idaho National Laboratory and a diverse mix of university, national laboratory and industry partner institutions.
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