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ANFTF System

01anftf.gifThe Advanced Nuclear Fuel Test Facility (ANFTF) is a Westinghouse-funded facility located at ARL's Test Site, approximately eight miles from Penn State's University Park campus. It was constructed to economically test and evaluate a wide variety of fuel component designs of Westinghouse's commercial and advanced power reactors. The facility was completed in 1996 and has since been successfully testing the AP600 (Advanced Passive 600-MW Reactors) fuel components.

Simulated Fuel Bundle

02anftf.gif The main test objective of ANFTF is to characterize the effect of different fuel grid spacers on the onset of the departure from nucleate boiling (DNB) phenomenon and to quantify the critical heat flux (CHF) as the DNB event occurs. DNB, sometimes known as boiling crisis, is an important multiphase phenomenon that could occur in a reactor. It limits the power output of a reactor for safe operation. Thus the DNB phenomenon needs to be thoroughly characterized for all grid spacer designs before the ones with favorable thermal performance may be selected for further water testing, and subsequent Nuclear Regulatory Commission licensing.

System Characteristics

  • ASME-coded and -stamped pressure vessel rated at 500 psi and 400°F
  • New Allegheny Power electrical service: available up to 900 kW at 60 Vdc
  • Designed maximum core mass velocity: 5.0x106 lbm/ft2-hr
  • Currently testing 5x5 fuel bundles with AP600 grids; can test 7x7 bundles in the future
  • Currently use a safe refrigerant as test fluid which offers substantial economical savings compared to the water tests
  • Currently testing 120-inch-long fuel rods: 40-inch unheated upstream, 40-inch heated, and 40-inch unheated downstream lengths
  • System can deliver approximately 30 DNB points in a working day


  • Critical heat flux (or DNB) tests for advanced reactor fuel designs
  • Qualitative and quantitative tests of fuel components of commercial reactors currently in service
  • Study of flow-induced fuel bundle and core vibration
  • Study of passive cooling of reactor core by natural convection