The TerraPower Traveling Wave Reactor (TWR) is a sodium-cooled fast reactor design that utilizes a high-burnup metallic uranium fuel cycle. The fuel system depends on a cladding material with demonstrated swelling resistance to high doses as well as adequate thermal creep strength. HT9 steel is a leading cladding candidate for the first TWR, having demonstrated excellent swelling and strain performance to doses > 200 dpa. A strain model was developed as a design tool to predict fuel pin deformation as a function of irradiation dose, stress, and temperature. The sources of strain deformation will be described along with the uncertainties in utilizing existing data to build a mechanistic model. The strain model is then incorporated into a fuel performance code to provide new insight in deformation behavior of HT9 fuel pins.
Skip Nav Destination
2014 22nd International Conference on Nuclear Engineering
July 7–11, 2014
Prague, Czech Republic
Conference Sponsors:
- Nuclear Engineering Division
ISBN:
978-0-7918-4589-9
PROCEEDINGS PAPER
HT9 Strain Modeling for Fuel Pin Deformation Available to Purchase
Micah Hackett,
Micah Hackett
TerraPower, LLC, Bellevue, WA
Search for other works by this author on:
Cheng Xu
Cheng Xu
TerraPower, LLC, Bellevue, WA
Search for other works by this author on:
Micah Hackett
TerraPower, LLC, Bellevue, WA
Ryan Latta
TerraPower, LLC, Bellevue, WA
Sam Miller
TerraPower, LLC, Bellevue, WA
Cheng Xu
TerraPower, LLC, Bellevue, WA
Paper No:
ICONE22-30414, V001T02A009; 6 pages
Published Online:
November 17, 2014
Citation
Hackett, M, Latta, R, Miller, S, & Xu, C. "HT9 Strain Modeling for Fuel Pin Deformation." Proceedings of the 2014 22nd International Conference on Nuclear Engineering. Volume 1: Plant Operations, Maintenance, Engineering, Modifications, Life Cycle and Balance of Plant; Nuclear Fuel and Materials; Plant Systems, Structures and Components; Codes, Standards, Licensing and Regulatory Issues. Prague, Czech Republic. July 7–11, 2014. V001T02A009. ASME. https://doi.org/10.1115/ICONE22-30414
Download citation file:
40
Views
Related Proceedings Papers
Related Articles
EBR-II Metallic Driver Fuel—A Live Option
J. Eng. Power (October,1981)
Thermomechanical Behavior and Modeling Between 350°C and 400°C of Zircaloy-4 Cladding Tubes From an Unirradiated State to High Fluence (0 to 85 s ˙ 10 24 nm − 2 , E > 1 MeV )
J. Eng. Mater. Technol (April,2000)
A Once-Through Fuel Cycle for Fast Reactors
J. Eng. Gas Turbines Power (October,2010)
Related Chapters
Polycrystalline Simulations of In-Reactor Deformation of Zircaloy-4 Cladding Tubes during Nominal Operating Conditions
Zirconium in the Nuclear Industry: 20th International Symposium
Nuclear Fuel Cycle
Non-Proliferation Nuclear Forensics: Canadian Perspective
Nuclear Fuel Materials and Basic Properties
Fundamentals of Nuclear Fuel