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Fast reactor fuel analysis methods and designs

A new code, the Fuel Engineering and Structural analysis Tool (FEAST), was developed to predict the performance of metal and oxide fuel in sodium-cooled fast reactors. FEAST describes key irradiation phenomena and can analyze the fuel and clad thermo-mechanical behavior at both steady-state and design-basis transients. FEAST has been coupled to the coolant subchannel analysis code COBRA, to improve the accuracy of temperature predictions in a fuel assembly. t was used to support fuel design in the Terrapower reactor.

Innovative fuel concepts were compared to the traditional cylindrical geometry. The annular fuel concept, with internal and external cooling, was found best suited for low-conversion metal-fuelled cores, where it can enable a power uprate of ~20%; limited by the fuel clad chemical interaction upon a hypothetical blockage of the inner-annular channel. The bottle-shaped fuel concept was found best suited for tight high-conversion ratio cores, where it can reduce the overall core pressure drop by >30%, with a corresponding increase in core height between 15 and 18%.

Publications

  1. A. Karahan and N. C. Andrews “ Extended Fuel Swelling Models and Ultra High Burnup Fuel Behavior of U-Pu-Zr Metallic Fuel Using FEAST-METAL”, J. of Nuclear Materials, accepted 2013.
  2. J. Fricano, J. Buongiorno, “Development and Application of an Integrated Fuel Performance and Sub-Channel Model for Analysis of Sodium Fast Reactors” Nuclear Technology, accepted 2013.
  3. ┬áM. Memmott, J. Buongiorno, P. Hejzlar, “An Evaluation of the Annular Fuel and Bottle-shaped Fuel Concepts for Sodium Fast Reactors”, Nuclear Technology, Vol. 173, 2011.
  4. A. Karahan, J. Buongiorno, “A New Code for Predicting the Thermo-Mechanical and Irradiation Behavior of Metallic Fuels in Sodium Fast Reactors”, J. Nuclear Materials, Vol. 396 (2–3), 283–293, 2010.
  5. A. Karahan, J. Buongiorno, “Modeling of Thermo-Mechanical and Irradiation Behavior of Mixed Oxide Fuel for Sodium Fast Reactors”, J. Nuclear Materials, Vol. 396 (2–3), 272–282, 2010.

 

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