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.

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.

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.

Numerical Assessment on Time-Varying Temperature Field of Bridge Tower under Solar Radiation

2012 ◽  
Vol 204-208 ◽  
pp. 2236-2239 ◽  
Author(s):  
Bo Chen ◽  
Wei Hua Guo ◽  
Chun Fang Song ◽  
Kai Kai Lu
Keyword(s):  
Heat Transfer ◽  
Finite Element ◽  
Temperature Field ◽  
Temperature Distribution ◽  
Solar Radiation ◽  
Ambient Air ◽  
Heat Transfer Analysis ◽  
Time Varying ◽  
Long Span ◽  
Bridge Tower

Bridge tower, time-varying temperature field, heat transfer analysis, finite element model. Abstract. Long span suspension bridges are subjected to daily, seasonal and yearly environmental thermal effects induced by solar radiation and ambient air temperature. This paper aims to investigate the temperature distribution of a tower of a long span suspension bridge. Two-dimensional heat transfer models are utilized to determine the time-dependent temperature distribution of the bridge tower of the bridge. The solar radiation model is utilized to examine the time-varying temperature distribution. Finite element models are constructed for the bridge tower to compute the temperature distribution. The numerical models can successfully predict the structural temperature field at different time. The methodology employed in the paper can be applied to other long-span bridges as well.

Squeal and Thermal Analysis of Automobile Disc Brake Rotors

2015 ◽  
Vol 764-765 ◽  
pp. 369-373
Author(s):  
Wei Hsin Gau ◽  
Kun Nan Chen ◽  
Chin Yuan Hung
Keyword(s):  
Heat Transfer ◽  
Thermal Analysis ◽  
Finite Element ◽  
Aluminum Alloy ◽  
Cast Iron ◽  
Cfd Simulation ◽  
Transfer Coefficients ◽  
Element Analysis ◽  
Heat Transfer Coefficients ◽  
Brake Rotors

The brakes of an automobile are among the most critical components regarding the safety features, and disc brakes are the most common type used in passenger vehicles. In this research, the squeal phenomena of a swirl-vent brake rotor and the thermal analysis of two straight-vent brake rotors, made of cast-iron and aluminum-alloy, are investigated. For the squeal analysis, finite element models are created and analyzed using a prestressed modal analysis with complex eigen-solutions. For the thermal analysis, heat transfer coefficients on the surfaces of a rotor as functions of time are first estimated by CFD simulation, and then imported to a thermal analysis program as the boundary condition. Finally, the temperature distribution of the rotor can be calculated by finite element analysis. The simulation results show that vortices will arise in the vented passages of straight-vent rotors, which means less heat carried away and lower heat transfer coefficients. The swirl-vent brake design is clearly better for thermal ventilation. Furthermore, under the same condition, aluminum-alloy rotors exhibit more uniform temperature distributions with smaller temperature gradients than cast-iron rotors do.

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.

Heat Transfer Through Small Moveable Gas Gaps in a Multi-Body System Using the ANSYS Finite Element Software

2013 ◽  
Author(s):  
Joel L. McDuffee
Keyword(s):  
Heat Transfer ◽  
Finite Element ◽  
Thermal Contact ◽  
National Laboratory ◽  
Oak Ridge ◽  
Finite Element Software ◽  
Advanced Fuels ◽  
Accurate Temperature ◽  
Multi Body ◽  
Helium Neon

The Thermal Hydraulics and Irradiation Engineering (THIE) Group at Oak Ridge National Laboratory (ORNL) designs and builds capsules in which to irradiate advanced fuels and materials that are typically inserted into ORNL’s High Flux Isotope Reactor. Experiments are designed to achieve a target temperature that ranges from 250°C to 1200°C. Most capsules do not have active temperature measurement or control, which puts an imperative on accurate temperature simulation. Temperature control in these capsules is accomplished by designing specific gaps between adjacent parts and filling the capsules with an inert gas: helium, neon, or argon. Most any finite element solver will do an excellent job estimating temperatures within individual parts, but the simulation challenge for these complex, multi-body systems is to accurately predict the heat transfer through contact surfaces or interstitial gas gaps. The gas gaps are on the order of 150 μm, so accurate simulation must include thermal expansion of the adjacent parts, the thermal jump effect on the part surfaces, and the possibility the parts will touch or break contact during expansion. This paper will discuss the limitations in thermal contact modeling in finite element modelers and the algorithms the THIE Group uses to overcome these limitations.

Thermal Analysis Safety Margins Using ABAQUS for the MP-2 Experiment in the Advanced Test Reactor

2018 ◽  
Author(s):  
Grant L. Hawkes ◽  
Douglas S. Crawford ◽  
Gregory K. Housley
Keyword(s):  
Thermal Analysis ◽  
Finite Element ◽  
Flow Instability ◽  
Water Saturation ◽  
National Laboratory ◽  
Saturation Temperature ◽  
Nucleate Boiling ◽  
New Technique ◽  
Safety Margins ◽  
Test Reactor

The Mini-Plate 2 (MP-2) irradiation test is a fueled experiment designed for irradiation in multiple test locations in the Advanced Test Reactor (ATR). MP-2 is considered a non-instrumented drop-in test where small aluminum-clad fuel plate samples are cooled directly by the ATR Primary Coolant System (PCS) water. The MP-2 fuel plate experiment will be irradiated in several different irradiation locations of the ATR. This fueled experiment contains aluminum-clad fuel mini plates consisting of monolithic U-Mo. Four different types of fuel plates were analyzed. A thermal analysis has been performed on the MP-2 experiment to be irradiated in the ATR at Idaho National Laboratory (INL). A new technique for calculating Departure from Nucleate Boiling Ratio (DNBR) and Flow Instability Ratio (FIR) using the commercial finite element and heat transfer code ABAQUS is demonstrated. This new technique calculates DNBR for the fuel plate surfaces and FIR for all water components for each finite element surface and node. Pressure drop data is fed into the calculations in order to geometrically calculate the water saturation temperature. Results from the DNBR and FIR calculations are displayed with the ABAQUS post processor named Viewer. By calculating these parameters at each location in the finite element model, conservatism is replaced with accuracy. This allows for a greater margin for the thermal hydraulic safety parameters.

The Finite Element Thermal Analysis for Lubricating Oil Transfer Pump in Different Convection Heat Transfer Coefficient with Special Material Properties

2012 ◽  
Vol 625 ◽  
pp. 167-170
Author(s):  
Jian Ning Xu ◽  
Wei Lv ◽  
Wen Jie Lv ◽  
Duan Yin Zhu
Keyword(s):  
Heat Transfer ◽  
Thermal Analysis ◽  
Finite Element ◽  
Temperature Field ◽  
Thermal Deformation ◽  
Heat Transfer Analysis ◽  
Lubricating Oil ◽  
Analysis Model ◽  
Screw Pump ◽  
Pump System

Lubricating oil transfer pump is important functional subsystem in two-screw pump system, and all moving parts works very fast, it is easy to generate heat, that not only affect the lubricating oil transfer pump’s temperature field, but also make transmission failure by the thermal deformation which caused by high temperature of parts in contact. This paper established a finite element thermal analysis model and boundary conditions of lubricating oil transfer pump, and calculated the temperature field and thermal deformation for it during the process of oil extraction in different coefficient of convective heat transfer, analysis the change rule of the steady state temperature field and thermal deformation, proved that lubricating oil transfer pump can normal work in certain conditions.

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