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Thermal Predictions of the AGR-3/4 Experiment With Time Varying Gas Gaps

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

Idaho National Laboratory, Idaho Falls, ID

Paper No. IMECE2014-36943, pp. V06BT07A023; 10 pages
  • ASME 2014 International Mechanical Engineering Congress and Exposition
  • Volume 6B: Energy
  • Montreal, Quebec, Canada, November 14–20, 2014
  • Conference Sponsors: ASME
  • ISBN: 978-0-7918-4952-1
  • Copyright © 2014 by ASME


A thermal analysis was performed for the Advanced Gas Reactor test experiment (AGR-3/4) with time varying gas gaps. The experiment was 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 tri-structural-isotropic (TRISO)-coated, low-enriched uranium oxy-carbide fuel. The AGR-3/4 test was designed primarily to assess fission product transport through various graphite materials. The AGR-3/4 test irradiation in the ATR started in December 2011 and finished in April 2014. Forty-eight (48) TRISO fueled compacts were inserted into twelve separate capsules for the experiment (four compacts per capsule).

The purpose of this analysis was to calculate the temperatures of each compact and graphite layer to obtain daily average temperatures using time (fast neutron fluence) varying gas gaps and to compare with experimentally measured thermocouple data. Previous experimental data was used for the graphite shrinkage versus fast neutron fluence. Heat rates were input from a detailed physics analysis using the Monte Carlo N-Particle (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 each capsule 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. Model results are compared to thermocouple data taken during the experiment.

Copyright © 2014 by ASME



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