James W. Sterbentz
Idaho National Laboratory
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Featured researches published by James W. Sterbentz.
international conference on fuel cell science engineering and technology fuelcell collocated with asme international conference on energy sustainability | 2015
Grant L. Hawkes; James W. Sterbentz; Binh T. Pham
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.Copyright
Nuclear Technology | 2015
Grant L. Hawkes; James W. Sterbentz; Binh T. Pham
Abstract A new daily as-run thermal analysis was performed at the Idaho National Laboratory for the advanced gas cooled reactor (AGR) test experiment number two (AGR-2) in the Advanced Test Reactor (ATR). This thermal analysis incorporates gas gaps changing with time during the irradiation experiment due to graphite shrinkage resulting from neutron damage. The purpose of this analysis was to calculate the daily average temperatures of each TRISO (tristructural isotropic)–particle fuel compact. A steady-state thermal analysis was performed daily for each capsule with the commercial finite element heat transfer code ABAQUS. These new thermal predictions show the compact fuel temperature dependence on the variable gas gap method. Comparison between measured and calculated temperatures is discussed.
ASME 2015 International Mechanical Engineering Congress and Exposition | 2015
Grant L. Hawkes; James W. Sterbentz; Binh T. Pham
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.Copyright
ASME 2014 International Mechanical Engineering Congress and Exposition | 2014
Grant L. Hawkes; James W. Sterbentz; John T. Maki
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
10th International Conference on Applications of Nuclear Techniques,Crete, Greece,06/13/2009,06/20/2009 | 2009
James L. Jones; James W. Sterbentz; Woo Y. Yoon; Daren R. Norman
Energetic photon sources with energies greater than 6 MeV continue to be recognized as viable source for various types of inspection applications, especially those related to nuclear and/or explosive material detection. These energetic photons can be produced as a continuum of energies (i.e., bremsstrahlung) or as a set of one or more discrete photon energies (i.e., monoenergetic). This paper will provide a follow‐on extension of the photon dose comparison presented at the 9th International Conference on Applications of Nuclear Techniques (June 2008). Our previous paper showed the comparative advantages and disadvantages of the photon doses provided by these two energetic interrogation sources and highlighted the higher energy advantage of the bremsstrahlung source, especially at long standoff distances (i.e., distance from source to the inspected object). This paper will pursue higher energy photon inspection advantage (up to 100 MeV) by providing dose and stimulated photonuclear interaction predictions ...
Archive | 2018
James W. Sterbentz; Joshua J. Cogliati
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ASME 2016 International Mechanical Engineering Congress and Exposition | 2016
Grant L. Hawkes; James W. Sterbentz; John T. Maki; Binh T. Pham
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.Copyright
ASME 2013 International Mechanical Engineering Congress and Exposition | 2013
Grant L. Hawkes; James W. Sterbentz; John T. Maki
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
Nuclear Instruments & Methods in Physics Research Section A-accelerators Spectrometers Detectors and Associated Equipment | 2006
James L. Jones; Daren R. Norman; Kevin J. Haskell; James W. Sterbentz; Woo Y. Yoon; Scott M. Watson; James T. Johnson; John Zabriskie; Brion D. Bennett; Richard W. Watson; Cavin E. Moss; J. Frank Harmon
Nuclear Instruments & Methods in Physics Research Section B-beam Interactions With Materials and Atoms | 2005
James L. Jones; Woo Y. Yoon; Daren R. Norman; Kevin J. Haskell; John Zabriskie; Scott M. Watson; James W. Sterbentz