Core materials development for the fuel cycle R&D program
Author:
Publisher
Elsevier BV
Subject
Nuclear Energy and Engineering,General Materials Science,Nuclear and High Energy Physics
Reference2 articles.
1. Microstructural analysis of an HT9 fuel assembly duct irradiated in FFTF to 155dpa at 443°C
2. The effects of fast reactor irradiation conditions on the tensile properties of two ferritic/martensitic steels
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1. Comparison of hardening and microstructures of ferritic/martensitic steels irradiated with fast neutrons and dual ions;Journal of Nuclear Materials;2024-10
2. Thermomechanical Processing for Improved Mechanical Properties of HT9 Steels;Materials;2024-08-01
3. Small-scale mechanical testing and characterization of fuel cladding chemical interaction between HT9 cladding and advanced U-based metallic fuel alloy;Journal of Nuclear Materials;2022-08
4. The Role of Co-injected Helium on Swelling and Cavity Evolution at High Damage Levels in Ferritic-Martensitic Steels;Journal of Nuclear Materials;2021-07
5. The effects of microstructures and radiation damage on the deformation behavior of a HT-9 alloy using microtensile testing;Materials Characterization;2021-04
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