NSUF 20-4201: Microstructural characterization of neutron irradiated HT-9 heats (ORNL, LANL and EBR II) at LWR and fast reactor relevant temperatures
Ferritic-martensitic (F-M) steels are being considered as candidate in-core structural materials for fast reactors and advanced LWR because of their excellent resistance to radiation-induced void swelling, microstructural stability, thermal conductivity and superior irradiation creep properties. HT-9 steel being the first-generation F-M steel has a relatively large mechanical properties and microstructural database under irradiated conditions. HT-9 was selected as the fuel clad and duct material in FFTF and EBR-II, and it is still the first-choice candidate core material for a number of advanced reactor concepts due to its service performance and the relatively large database on it. Currently, commercial nuclear power companies (e.g. TerraPower) have rejuvenated the manufacturing of HT-9.
In order to address the issue of low-temperature (~425°C) neutron irradiation hardening and embrittlement, it is necessary to conduct systematic investigations on the mechanical behavior and microstructure of HT-9 with slight variations in chemical composition and heat treatment over a wide range of doses and temperatures. Three HT-9 heats (ORNL, LANL and EBR II) with variations in manufacturing process, chemical composition and heat treatment were neutron irradiated in the ATR as a part of the UW-Madison NSUF pilot irradiation experiment.
PNNL completed a funded RTE project and obtained mechanical properties of higher dose (~8 dpa; actual) HT-9 as a function of processing conditions and irradiation temperatures (291°C, 360°C and 431°C; actual). Researchers at the University of Michigan performed TEM studies of three HT-9 variants irradiated to 8 dpa (431°C; actual) under a funded RTE project. In order to comprehensively characterize these HT-9 variants after neutron irradiation and to obtain processing-properties-structure-dose-temperature correlation, it is essential to perform detailed SEM/EBSD and APT studies as two smaller projects due to the RTE funding limit.
The proposed project aims to understand the effect of HT-9 processing conditions (heats: ORNL, LANL and EBR II; different austenitizing and tempering parameters, and chemical composition) and neutron irradiation (~8 dpa) as a function of irradiation temperatures (291°C, 360°C and 431°C; actual) on the microstructural features of HT-9. Failed/tested tensile specimen shoulders (control and irradiated) are available at PNNL for the proposed study. SEM and EBSD will be used to evaluate control and irradiated samples and determine the general grain structure, prior austenite grain size, martensite packet, lath structure and primary carbide structure as a function of irradiation temperatures, since these parameters can significantly affect the mechanical behavior and radiation resistance of HT-9. Thermocalc simulations will be utilized to obtain correlations between heat treatments and observed microstructures. Efforts will be made to comprehensively understand the effects of radiation damage on HT-9 heats at the LWR and fast reactor relevant temperatures, and to develop appropriate processing-property-structure-temperature-dose correlations. The results of the proposed work could be extended beyond HT-9, and it would be relevant to many F-M steels. Thus, the proposed project will have substantial implications for the deployment of next-generation advanced reactors.
The project performance (sample preparation, imaging and analysis) is expected to take place during July-September 2020 and will result in one conference presentation and one journal article publication.
Additional Info
| Field | Value |
|---|---|
| Abstract | Ferritic-martensitic (F-M) steels are being considered as candidate in-core structural materials for fast reactors and advanced LWR because of their excellent resistance to radiation-induced void swelling, microstructural stability, thermal conductivity and superior irradiation creep properties. HT-9 steel being the first-generation F-M steel has a relatively large mechanical properties and microstructural database under irradiated conditions. HT-9 was selected as the fuel clad and duct material in FFTF and EBR-II, and it is still the first-choice candidate core material for a number of advanced reactor concepts due to its service performance and the relatively large database on it. Currently, commercial nuclear power companies (e.g. TerraPower) have rejuvenated the manufacturing of HT-9. In order to address the issue of low-temperature (~425°C) neutron irradiation hardening and embrittlement, it is necessary to conduct systematic investigations on the mechanical behavior and microstructure of HT-9 with slight variations in chemical composition and heat treatment over a wide range of doses and temperatures. Three HT-9 heats (ORNL, LANL and EBR II) with variations in manufacturing process, chemical composition and heat treatment were neutron irradiated in the ATR as a part of the UW-Madison NSUF pilot irradiation experiment. PNNL completed a funded RTE project and obtained mechanical properties of higher dose (~8 dpa; actual) HT-9 as a function of processing conditions and irradiation temperatures (291°C, 360°C and 431°C; actual). Researchers at the University of Michigan performed TEM studies of three HT-9 variants irradiated to 8 dpa (431°C; actual) under a funded RTE project. In order to comprehensively characterize these HT-9 variants after neutron irradiation and to obtain processing-properties-structure-dose-temperature correlation, it is essential to perform detailed SEM/EBSD and APT studies as two smaller projects due to the RTE funding limit. The proposed project aims to understand the effect of HT-9 processing conditions (heats: ORNL, LANL and EBR II; different austenitizing and tempering parameters, and chemical composition) and neutron irradiation (~8 dpa) as a function of irradiation temperatures (291°C, 360°C and 431°C; actual) on the microstructural features of HT-9. Failed/tested tensile specimen shoulders (control and irradiated) are available at PNNL for the proposed study. SEM and EBSD will be used to evaluate control and irradiated samples and determine the general grain structure, prior austenite grain size, martensite packet, lath structure and primary carbide structure as a function of irradiation temperatures, since these parameters can significantly affect the mechanical behavior and radiation resistance of HT-9. Thermocalc simulations will be utilized to obtain correlations between heat treatments and observed microstructures. Efforts will be made to comprehensively understand the effects of radiation damage on HT-9 heats at the LWR and fast reactor relevant temperatures, and to develop appropriate processing-property-structure-temperature-dose correlations. The results of the proposed work could be extended beyond HT-9, and it would be relevant to many F-M steels. Thus, the proposed project will have substantial implications for the deployment of next-generation advanced reactors. The project performance (sample preparation, imaging and analysis) is expected to take place during July-September 2020 and will result in one conference presentation and one journal article publication. |
| Award Announced Date | 2020-07-14T14:15:33.813 |
| Awarded Institution | Idaho National Laboratory |
| Facility | Advanced Test Reactor |
| Facility Tech Lead | Alina Zackrone, Stuart Maloy |
| Irradiation Facility | None |
| PI | Indrajit Charit |
| PI Email | [email protected] |
| Project Type | RTE |
| RTE Number | 4201 |