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Thermal Predictions of the AGR-3/4 Experiment

[+] Author Affiliations
Grant L. Hawkes, James W. Sterbentz, John T. Maki

Idaho National Laboratory, Idaho Falls, ID

Paper No. IMECE2013-65155, pp. V06BT07A054; 8 pages
doi:10.1115/IMECE2013-65155
From:
  • ASME 2013 International Mechanical Engineering Congress and Exposition
  • Volume 6B: Energy
  • San Diego, California, USA, November 15–21, 2013
  • Conference Sponsors: ASME
  • ISBN: 978-0-7918-5629-1
  • Copyright © 2013 by ASME

abstract

A thermal analysis was performed for the Advanced Gas Reactor test experiment number three/four (AGR-3/4) irradiated at the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL). Several fuel irradiation experiments are planned for the AGR Fuel Development and Qualification Program which supports the development of the Very-High-Temperature gas-cooled Reactor (VHTR) under the Next-Generation Nuclear Plant (NGNP) project.

AGR-3/4 combines two tests in a series of planned AGR experiments to test tristructural-isotropic (TRISO)-coated, low-enriched uranium oxycarbide fuel. The AGR-3/4 test was designed primarily to assess fission product transport through various graphite materials. The AGR-3/4 test was inserted in the ATR beginning in 2011 and is currently still in the reactor. Forty-eight (48) TRISO fueled compacts were inserted into twelve separate capsules for the experiment.

The purpose of this analysis was to calculate the temperatures of each compact and graphite layer to obtain daily average temperatures and to compare with experimentally measured thermocouple data. Heat rates were input from a detailed physics analysis using the MCNP code for each day during the experiment. Individual heat rates for each non-fuel component were input as well. A steady-state thermal analysis was performed for each daily calculation. A finite element model was created for this analysis using the commercial finite element heat transfer and stress analysis package ABAQUS. The fission and neutron gamma heat rates were calculated with the nuclear physics code MCNP. ATR outer shim control cylinders and neck shim rods along with driver fuel power and fuel depletion were incorporated into the daily physics heat rate calculations. Compact and graphite thermal conductivity were input as a function of temperature and neutron fluence with the field variable option in ABAQUS. Surface-to-surface radiation heat transfer along with conduction heat transfer through the gas mixture of helium-neon (used for temperature control) was used in these models. The kinetic theory of gases was used to correlate the thermal conductivity of the gas mixture. Model results are compared to thermocouple data taken during the experiment. Future thermal analysis models will consider control temperature gas gaps and fuel compact–graphite holder gas gaps varying from the original fabrication dimensions as a function of fast neutron fluence.

Copyright © 2013 by ASME

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