NDLI: Sensitivity Evaluation of the AGR 3-4 Experiment Thermal Model Irradiated in the Advanced Test Reactor

Content Provider	The American Society of Mechanical Engineers (ASME) Digital Collection
Author	Grant, L. Hawkes James, W. Sterbentz Binh, T. Pham
Copyright Year	2015
Abstract	A temperature sensitivity evaluation has been performed on a thermal model for the AGR-3/4 fuel experiment on an individual capsule. The experiment was irradiated in the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL). Four TRISO fuel irradiation experiments are planned for the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program which supports the development of the Very High Temperature Gas-cooled Reactor under the Next-Generation Nuclear Plant project. AGR-3/4 is the third TRISO-particle fuel test of the four planned and is intended to test tri-structural-isotropic (TRISO)-coated, low-enriched uranium oxy-carbide fuel. The AGR-3/4 test was specifically designed to assess fission product transport through various graphite materials. The AGR-3/4 irradiation test in the ATR started in December 2011 and finished in April 2014. Forty-eight (48) TRISO-particle fueled compacts were inserted into 12 separate capsules for the experiment (four compacts per capsule). The purpose of this analysis was to assess the sensitivity of input variables for the capsule thermal model. A series of cases were compared to a base case by varying different input parameters into the ABAQUS finite element thermal model. These input parameters were varied by ±10% to show the temperature sensitivity to each parameter. The most sensitive parameter was the compact heat rates, followed by the outer control gap distance and neon gas fraction. Thermal conductivity of the compacts and thermal conductivity of the various graphite layers vary with fast neutron fluence and exhibited moderate sensitivity. The least sensitive parameters were the emissivities of the stainless steel and graphite, along with gamma heat rate in the non-fueled components. Separate sensitivity calculations were performed varying with fast neutron fluence, showing a general temperature rise with an increase in fast neutron fluence. This is a result of the control gas gap becoming larger due to the graphite shrinkage with neutron damage. A smaller sensitivity is due to the thermal conductivity of the fuel compacts with fast neutron fluence. Heat rates and fast neutron fluence were input from a detailed physics analysis using the Monte Carlo N-Particle (MCNP) code. Individual heat rates for each non-fuel component were input as well. A steady-state thermal analysis was performed for each sensitivity calculation. ATR outer shim control cylinders and neck shim rods along with driver fuel power and fuel depletion were incorporated into the physics heat rate calculations. 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 the sensitivity calculations.
File Format	PDF
ISBN	9780791857441
DOI	10.1115/IMECE2015-53544
Volume Number	Volume 6B: Energy
Conference Proceedings	ASME 2015 International Mechanical Engineering Congress and Exposition
Language	English
Publisher Date	2015-11-13
Publisher Place	Houston, Texas, USA
Access Restriction	Subscribed
Subject Keyword	Temperature Neutrons Rods Fuels Fluence (radiation measurement) Heat conduction Uranium Thermal conductivity Damage Nuclear fission Cylinders Irradiation (radiation exposure) Shrinkage (materials) Helium Steady state Physics Heat Very high temperature reactors Radiation (physics) Particulate matter Stainless steel Graphite Finite element analysis Temperature control Thermal analysis Nuclear power stations Heat transfer
Content Type	Text
Resource Type	Article

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