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

2015 ◽  
Author(s):  
Grant L. Hawkes ◽  
James W. Sterbentz ◽  
Binh T. Pham
Keyword(s):  
Heat Transfer ◽  
Thermal Conductivity ◽  
Fast Neutron ◽  
Temperature Sensitivity ◽  
Thermal Model ◽  
Neutron Fluence ◽  
Heat Rate ◽  
Sensitivity Evaluation ◽  
Input Parameters ◽  
Test Reactor

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.

scholarly journals Sensitivity Evaluation of the Daily Thermal Predictions of the AGR-2 Experiment in the Advanced Test Reactor

2015 ◽  
Author(s):  
Grant L. Hawkes ◽  
James W. Sterbentz ◽  
Binh T. Pham
Keyword(s):  
Thermal Conductivity ◽  
Temperature Sensitivity ◽  
National Laboratory ◽  
Heat Rate ◽  
Base Case ◽  
Sensitivity Evaluation ◽  
Input Parameters ◽  
Test Reactor ◽  
Fuel Particles ◽  
Sensitive Parameters

A temperature sensitivity evaluation has been performed for an individual test capsule in the AGR-2 TRISO particle fuel experiment. The AGR-2 experiment is the second in a series of fueled test experiments for TRISO coated fuel particles run in the Advanced Test Reactor at the Idaho National Laboratory. A series of cases were compared to a base case by varying different input parameters in an ABAQUS finite element thermal model. Most input parameters were varied by ±10%, with one parameter ±20%, to show the temperature sensitivity to each parameter. The most sensitive parameters were the outer control gap distance, heat rate in the fuel compacts, and neon gas fraction. The thermal conductivity of the fuel compacts and thermal conductivity of the graphite holder were of 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. Sensitivity calculations were also performed for the fast neutron fluence, which showed a general, but minimal, temperature rise with increasing fluence.

scholarly journals Thermal Conductivities of Unidirectional Materials

1967 ◽  
Vol 1 (2) ◽  
pp. 166-173 ◽  
Author(s):  
George S. Springer ◽  
Stephen W. Tsai
Keyword(s):  
Heat Transfer ◽  
Experimental Data ◽  
Thermal Conductivity ◽  
Thermal Model ◽  
Longitudinal Shear ◽  
Unidirectional Composite ◽  
Shear Loading ◽  
Square Packing ◽  
Good Agreement ◽  
Thermal Conductivities

In this paper the composite thermal conductivities of unidirec tional composites are studied and expressions are obtained for pre dicting these conductivities in the directions along and normal to the filaments. In the direction along the filament an expression is presented based on the assumption that the filaments and matrix are connected in parallel. In the direction normal to the filaments composite thermal conductivity values are obtained first by utiliz ing the analogy between the response of a unidirectional composite to longitudinal shear loading and to transverse heat transfer; second by replacing the filament-matrix composite with an idealized ther mal model. The results of the shear loading analogy agree reason ably well with the results of the thermal model particularly at filament contents below about 60%. These results were also com pared to experimental data reported in the literature and good agreement was found between the data and those theoretical re sults that were derived for circular filaments arranged in a square packing array.

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

2014 ◽  
Author(s):  
Grant L. Hawkes ◽  
James W. Sterbentz ◽  
John T. Maki
Keyword(s):  
Heat Transfer ◽  
Thermal Analysis ◽  
Finite Element ◽  
Fast Neutron ◽  
Neutron Fluence ◽  
National Laboratory ◽  
Surface Radiation ◽  
Time Varying ◽  
Fuel Component ◽  
Enriched Uranium

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.

Cylindricity Sensitivity Thermal Model of the AGR-5/6/7 Experiment in the Advanced Test Reactor

2020 ◽  
Author(s):  
Grant L. Hawkes
Keyword(s):  
Heat Transfer ◽  
Stainless Steel ◽  
Thermal Model ◽  
National Laboratory ◽  
Sensitivity Study ◽  
Operating Conditions ◽  
Department Of Energy ◽  
Easy Method ◽  
Gas Gap ◽  
Test Reactor

Abstract The AGR-5/6/7 experiment is currently being irradiated in the Advanced Test Reactor (ATR) at the Idaho National Laboratory and is approximately 70% complete. Several fuel and material irradiation experiments have been planned for the U.S. Department of Energy Advanced Gas Reactor Fuel Development and Qualification Program, which supports the development and qualification of tristructural isotropic (TRISO)-coated particle fuel for use in high-temperature gas-cooled reactors. The goals of these experiments are to provide irradiation performance data to support fuel process development, qualify fuel for normal operating conditions, support development of fuel performance models and codes, and provide irradiated fuel and materials for post-irradiation examination and safety testing. Originally planned and named as separate fuel experiments, but subsequently combined into a single test train, AGR-5/6/7 is testing low-enriched uranium oxycarbide TRISO fuel. The AGR-5/6/7 test train has five capsules with thermocouples and independent gas control mixtures. Unique to this paper is a sensitivity study concerning the cylindricity of the graphite holders containing the fuel compacts and their eccentricity in relation to the stainless-steel capsule walls. Each capsule has small nubs on the outside used for centering the graphite holder inside the stainless-steel capsule with a small gas gap used to control temperature. Due to machining tolerances of these nubs, and vibration wearing the nubs down when the experiment is running in the reactor, the possibility exists that the holder may move around radially. Each capsule is equipped with several thermocouples placed at various radii and depths within each graphite holder. This paper will show the sensitivity of offsetting the graphite holder for various radii in 45-degree increments around the circle with the objective of minimizing the difference between the measured thermocouples and the modeled thermocouple temperatures. Separate gas mixtures of helium/neon are introduced into this gas gap between the holder and capsule wall and changed as necessary to maintain the desired thermocouple temperatures to keep the fuel compacts at a constant temperature as the nuclear reactor conditions change. The goal of the sensitivity study is to find a radius and an angle to offset the holder from perfectly centered for each of the five capsules separately. The complex thermal model includes fission heating, gamma heating, radiation heat transfer, and heat transfer via conduction and radiation across the control gaps. Subroutines linked to the thermal model offer an easy method to offset the graphite holder from the capsule walls without remeshing the entire model.

Thermal Predictions of the AGR-3/4 Experiment

2013 ◽  
Author(s):  
Grant L. Hawkes ◽  
James W. Sterbentz ◽  
John T. Maki
Keyword(s):  
Heat Transfer ◽  
Thermal Conductivity ◽  
Thermal Analysis ◽  
Finite Element ◽  
Neutron Fluence ◽  
National Laboratory ◽  
Surface Radiation ◽  
Gas Mixture ◽  
Fuel Component ◽  
Enriched Uranium

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.

scholarly journals Thermal Predictions of the AGR-3/4 Experiment With Post Irradiation Examination Measured Time-Varying Gas Gaps

2017 ◽  
Vol 3 (4) ◽  
Author(s):  
Grant L. Hawkes ◽  
James W. Sterbentz ◽  
John T. Maki ◽  
Binh T. Pham
Keyword(s):  
Thermal Analysis ◽  
Fast Neutron ◽  
Neutron Fluence ◽  
National Laboratory ◽  
Element Model ◽  
Post Irradiation ◽  
Measured Time ◽  
Dimensional Measurements ◽  
Test Reactor ◽  
Experiment Individual

A thermal analysis was performed for the advanced gas reactor test experiment (AGR-3/4) with post irradiation examination (PIE) measured time (fast neutron fluence) 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 under the advanced reactor technologies project. The AGR-3/4 test was designed primarily to assess fission product transport through various graphite materials. Irradiation in the ATR started in December 2011 and finished in April 2014. Forty-eight (48) tristructural-isotropic-fueled compacts were inserted into 12 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 using PIE-measured time (fast neutron fluence) varying gas gaps and compare with experimentally measured thermocouple (TC) data. PIE-measured experimental data were used for the graphite shrinkage versus fast neutron fluence. PIE dimensional measurements were taken on all the fuel compacts, graphite holders, and all of the graphite rings used. Heat rates were input from a detailed physics analysis for each day during the experiment. Individual heat rates for each nonfuel 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.

Thermal Predictions of the AGR-3/4 Experiment Using PIE-Measured Time Varying Gas Gaps

2016 ◽  
Author(s):  
Grant L. Hawkes ◽  
James W. Sterbentz ◽  
John T. Maki ◽  
Binh T. Pham
Keyword(s):  
Heat Transfer ◽  
Thermal Analysis ◽  
Finite Element ◽  
Fast Neutron ◽  
Neutron Fluence ◽  
National Laboratory ◽  
Surface Radiation ◽  
Time Varying ◽  
Fuel Component ◽  
Measured Time

A thermal analysis was performed for the Advanced Gas Reactor test experiment (AGR-3/4) with Post Irradiation Examination (PIE) measured 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 PIE-measured time (fast neutron fluence) varying gas gaps and to compare with experimentally measured thermocouple data. PIE-measured 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 fast 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.

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