Microstructural evolution of heavy-ion irradiated HT-9 ferritic steel
- National Tsing-Hua Univ. (Taiwan)
Ferritic and martensitic steel HT-9 is considered to be the leading candidate for the cladding and structural materials of fast breeder reactors (FBRs) and the first walls and blankets in conceptual fusion reactor designs for its superior resistance to void swelling and adequate mechanical properties at elevated temperatures. This study was designed to investigate the microstructural stability, including precipitation response, void swelling, and dislocation evolution, of HT-9 following heavy-ion irradiation at elevated temperatures. The most significant result obtained in this study is that without helium implantation, there is virtually no void swelling in all specimens following heavy ion irradiation to a peak damage level of 200 dpa in the temperature range between 300 and 600{degree}C. Even with 100 appm helium preimplantation, the void swelling rate in HT-9 specimens is <0.01% per dpa, which is much better swelling resistance than 1% per dpa, which is much better swelling resistance than 1% per dpa for austenitic stainless steels (e.g., Type 316 or 304 stainless steel). This superior void swelling resistance suggested that the void swelling may not be the critical problem for using the HT-9 alloy in FBRs or fusion reactors.
- OSTI ID:
- 6502773
- Report Number(s):
- CONF-891103-; CODEN: TANSA; TRN: 90-034527
- Journal Information:
- Transactions of the American Nuclear Society; (USA), Vol. 60; Conference: Winter meeting of the American Nuclear Society (ANS) and nuclear power and technology exhibit, San Francisco, CA (USA), 26-30 Nov 1989; ISSN 0003-018X
- Country of Publication:
- United States
- Language:
- English
Similar Records
Irradiation Damages of Structural Materials under Different Irradiation Environments
Martensitic/ferritic steels as container materials for liquid mercury target of ESS
Related Subjects
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
36 MATERIALS SCIENCE
ALLOY-HT-9
MICROSTRUCTURE
PHYSICAL RADIATION EFFECTS
LMFBR TYPE REACTORS
REACTOR MATERIALS
THERMONUCLEAR REACTORS
BREEDING BLANKETS
DISLOCATIONS
FIRST WALL
FUEL CANS
HELIUM
ION IMPLANTATION
IRRADIATION
MECHANICAL PROPERTIES
METALLOGRAPHY
NICKEL IONS
PRECIPITATION
STABILITY
SWELLING
TEMPERATURE DEPENDENCE
TRANSMISSION ELECTRON MICROSCOPY
ALLOYS
BREEDER REACTORS
CHARGED PARTICLES
CHROMIUM ALLOYS
CHROMIUM STEELS
CORROSION RESISTANT ALLOYS
CRYSTAL DEFECTS
CRYSTAL STRUCTURE
ELECTRON MICROSCOPY
ELEMENTS
EPITHERMAL REACTORS
FAST REACTORS
FBR TYPE REACTORS
FLUIDS
GASES
HEAT RESISTANT MATERIALS
HEAT RESISTING ALLOYS
HIGH ALLOY STEELS
IONS
IRON ALLOYS
IRON BASE ALLOYS
LINE DEFECTS
LIQUID METAL COOLED REACTORS
MARTENSITIC STEELS
MATERIALS
MICROSCOPY
MOLYBDEN
MOLYBDENUM ADDITIONS
NONMETALS
RADIATION EFFECTS
RARE GASES
REACTOR COMPONENTS
REACTORS
SEPARATION PROCESSES
STAINLESS STEELS
STEEL-CR12MOV
STEELS
THERMONUCLEAR REACTOR WALLS
700209* - Fusion Power Plant Technology- Component Development & Materials Testing
210500 - Power Reactors
Breeding
360106 - Metals & Alloys- Radiation Effects
360102 - Metals & Alloys- Structure & Phase Studies