Fuel Performance Design Basis for the Versatile Test Reactor

Porter, D. L.; Crawford, D. C.

doi:10.1080/00295639.2021.2009983

Title: Fuel Performance Design Basis for the Versatile Test Reactor

Journal Article · 12 January 2022 · Nuclear Science and Engineering

DOI: https://doi.org/10.1080/00295639.2021.2009983 · OSTI ID:1839804

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Idaho National Laboratory, Nuclear Science and Technology, Idaho Falls, Idaho
Idaho National Laboratory, Materials and Fuels Complex, Idaho Falls, Idaho

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Research Organization:: Idaho National Laboratory (INL), Idaho Falls, ID (United States)

Sponsoring Organization:: USDOE; USDOE Office of Nuclear Energy (NE)

Grant/Contract Number:: AC07-05ID14517

OSTI ID:: 1839804

Report Number(s):: INL/JOU-21-63463; 11

Journal Information:: Nuclear Science and Engineering, Journal Name: Nuclear Science and Engineering Journal Issue: sup1 Vol. 196; ISSN 0029-5639

Publisher:: Informa UK LimitedCopyright Statement

Country of Publication:: United States

Language:: English

References (22)

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Mechanism-based modeling of thermal and irradiation creep behavior: An application to ferritic/martensitic HT9 steel Wen, W.; Kohnert, A.; Arul Kumar, M. International Journal of Plasticity, Vol. 126 https://doi.org/10.1016/j.ijplas.2019.11.012	journal	March 2020
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Metallic fuels for advanced reactors Carmack, W. J.; Porter, D. L.; Chang, Y. I. Journal of Nuclear Materials, Vol. 392, Issue 2 https://doi.org/10.1016/j.jnucmat.2009.03.007	journal	July 2009
Thermal annealing recovery of fracture toughness in HT9 steel after irradiation to high doses Byun, Thak Sang; Baek, Jong-Hyuk; Anderoglu, Osman Journal of Nuclear Materials, Vol. 449, Issue 1-3 https://doi.org/10.1016/j.jnucmat.2013.07.064	journal	June 2014
Metallography and fuel cladding chemical interaction in fast flux test facility irradiated metallic U-10Zr MFF-3 and MFF-5 fuel pins Carmack, W. J.; Chichester, H. M.; Porter, D. L. Journal of Nuclear Materials, Vol. 473 https://doi.org/10.1016/j.jnucmat.2016.02.019	journal	May 2016
Scanning electron microscopy examination of a Fast Flux Test Facility irradiated U-10Zr fuel cross section clad with HT-9 Harp, Jason M.; Porter, Douglas L.; Miller, Brandon D. Journal of Nuclear Materials, Vol. 494 https://doi.org/10.1016/j.jnucmat.2017.07.040	journal	October 2017
HT9 swelling in high burnup fast reactor fuel pin components Porter, D. L.; Miller, B. D.; Hilton, B. A. Journal of Nuclear Materials, Vol. 519 https://doi.org/10.1016/j.jnucmat.2019.03.053	journal	June 2019
A CALPHAD-informed approach to modeling constituent redistribution in Zr-based metallic fuels using BISON Hirschhorn, Jacob; Tonks, Michael; Matthews, Christopher Journal of Nuclear Materials, Vol. 544 https://doi.org/10.1016/j.jnucmat.2020.152657	journal	February 2021
Modeling constituent redistribution in U–Pu–Zr metallic fuel using the advanced fuel performance code BISON Galloway, J.; Unal, C.; Carlson, N. Nuclear Engineering and Design, Vol. 286 https://doi.org/10.1016/j.nucengdes.2015.01.014	journal	May 2015
Fuel-Cladding Chemical Interaction in U-Pu-Zr Metallic Fuels: A Critical Review Matthews, Christopher; Unal, Cetin; Galloway, Jack Nuclear Technology, Vol. 198, Issue 3 https://doi.org/10.1080/00295450.2017.1323535	journal	June 2017
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