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.

Heat transfer characteristic modelling and the effect of operating conditions on re-cooled cycle for a scramjet

2011 ◽  
Vol 115 (1164) ◽  
pp. 83-90 ◽  
Author(s):  
W. Bao ◽  
J. Qin ◽  
W. X. Zhou
Keyword(s):  
Heat Transfer ◽  
Heat Sink ◽  
Transfer Characteristic ◽  
Cooling Capacity ◽  
Sensitivity Study ◽  
Performance Model ◽  
Operating Conditions ◽  
Inlet Temperature ◽  
Fuel Flow ◽  
Temperature And Pressure

Abstract A re-cooled cycle has been proposed for a regeneratively cooled scramjet to reduce the hydrogen fuel flow for cooling. Upon the completion of the first cooling, fuel can be used for secondary cooling by transferring the enthalpy from fuel to work. Fuel heat sink (cooling capacity) is thus repeatedly used and fuel heat sink is indirectly increased. Instead of carrying excess fuel for cooling or seeking for any new coolant, the cooling fuel flow is reduced, and fuel onboard is adequate to satisfy the cooling requirement for the whole hypersonic vehicle. A performance model considering flow and heat transfer is build. A model sensitivity study of inlet temperature and pressure reveals that, for given exterior heating condition and cooling panel size, fuel heat sink can be obviously increased at moderate inlet temperature and pressure. Simultaneously the low-temperature heat transfer deterioration and Mach number constrains can also be avoided.

scholarly journals Fluid Choice Based on Thermal Model and Performance Testing for Direct Cooled Electric Drive

Energies ◽  
2020 ◽  
Vol 13 (22) ◽  
pp. 5867
Author(s):  
Robert Lehmann ◽  
Arthur Petuchow ◽  
Matthias Moullion ◽  
Moritz Künzler ◽  
Christian Windel ◽  
...  
Keyword(s):  
Heat Transfer ◽  
Thermal Model ◽  
Performance Testing ◽  
Sensitivity Study ◽  
Maximum Heat ◽  
Cooling Fluid ◽  
Two Fluids ◽  
Fluid Properties ◽  
And Performance ◽  
Electric Traction

In this publication, the cooling fluid for direct oil-cooled electric traction drive is investigated. A dedicated thermal resistance model was developed in order to show the influence of the fluid properties on the continuous performance. For this purpose, the heat transfer parameters are adjusted in the simulation using an exponential approach in order to evaluate the cooling fluid. In a sensitivity study, density, heat capacity, thermal conductivity, and viscosity are investigated. Because viscosity, within the range investigated, shows the largest percentage deviation from the reference fluid, the greatest effect on performance can be seen here. In order to check the plausibility of the calculated results of the thermal simulation, two fluids were chosen for performance testing on a dedicated electro motor cooling (EMC) test. Beyond the investigation of heat transfer, aging of the defined fluid at maximum heat input over several hours is also evaluated. Only slight changes of the fluid properties are detected. This publication presents a thermal model for direct oil-cooled drive trains, which consider fluid properties. Furthermore, the model was tested for plausibility on real hardware.

Assessing Advanced Test Reactor Process States Using Acoustic Measurement Infrastructure

2017 ◽  
Author(s):  
Vivek Agarwal ◽  
James A. Smith
Keyword(s):  
Nuclear Reactor ◽  
Signal Transmission ◽  
National Laboratory ◽  
Reactor Core ◽  
Operating Conditions ◽  
Acoustic Measurement ◽  
Primary Coolant ◽  
Test Reactor ◽  
Operation Cycle ◽  
Measurement Infrastructure

The core of any nuclear reactor presents a particularly harsh environment for sensors and instrumentations. The reactor core also imposes challenging constraints on signal transmission from inside the reactor core to outside of the reactor vessel. In this paper, an acoustic measurement infrastructure installed at the Advanced Test Reactor (ATR), located at Idaho National Laboratory, is presented. The measurement infrastructure consists of ATR in-pile structural components, coolant, acoustic receivers, primary coolant pumps, a data-acquisition system, and signal processing algorithms. Intrinsic and cyclic acoustic signals generated by the operation of the primary coolant pumps are collected and processed. The characteristics of the intrinsic signal can indicate the process state of the ATR (such as reactor startup, reactor criticality, reactor attaining maximum power, and reactor shutdown) during operation (i.e., real-time measurement). This paper demonstrated different in acoustic signature of the ATR under different operating conditions. In particular, ATR acoustic baseline is captured during typical operation cycle and during power axial locator mechanism operation cycle. The difference in two acoustic baseline is significant and highlights salient difference that are critical in the design and development of acoustically telemetered sensors.

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 Advanced Test Reactor: A National Scientific User Facility

2008 ◽  
Author(s):  
Clifford J. Stanley ◽  
Frances M. Marshall
Keyword(s):  
Neutron Flux ◽  
National Laboratory ◽  
Reactor Core ◽  
Nuclear Research ◽  
Reactor Power ◽  
Department Of Energy ◽  
Life Extension ◽  
Test Environment ◽  
User Facility ◽  
Test Reactor

This presentation and associated paper provides an overview of the research and irradiation capabilities of the Advanced Test Reactor (ATR) located at the U.S. Department of Energy Idaho National Laboratory (INL). The ATR which has been designated by DOE as a National Scientific User Facility (NSUF) is operated by Battelle Energy Alliance, LLC. This paper will describe the ATR and discuss the research opportunities for university (faculty and students) and industry researchers to use this unique facility for nuclear fuels and materials experiments in support of advanced reactor development and life extension issues for currently operating nuclear reactors. The ATR is a pressurized, light-water moderated and cooled, beryllium-reflected nuclear research reactor with a maximum operating power of 250 MWth. The unique serpentine configuration (Fig. 1) of the fuel elements creates five main reactor power lobes (regions) and nine flux traps. In addition to these nine flux traps there are 68 additional irradiation positions in the reactor core reflector tank. There are also 34 low-flux irradiation positions in the irradiation tanks outside the core reflector tank. The ATR is designed to provide a test environment for the evaluation of the effects of intense radiation (neutron and gamma). Due to the unique serpentine core design each of the five lobes can be operated at different powers and controlled independently. Options exist for the individual test trains and assemblies to be either cooled by the ATR coolant (i.e., exposed to ATR coolant flow rates, pressures, temperatures, and neutron flux) or to be installed in their own independent test loops where such parameters as temperature, pressure, flow rate, neutron flux, and chemistry can be controlled per experimenter specifications. The full-power maximum thermal neutron flux is ∼1.0 x1015 n/cm2-sec with a maximum fast flux of ∼5.0 x1014 n/cm2-sec. The Advanced Test Reactor, now a National Scientific User Facility, is a versatile tool in which a variety of nuclear reactor, nuclear physics, reactor fuel, and structural material irradiation experiments can be conducted. The cumulative effects of years of irradiation in a normal power reactor can be duplicated in a few weeks or months in the ATR due to its unique design, power density, and operating flexibility.

Validation of the Combustion Space Simulation of a Glass Furnace Model

2001 ◽  
Author(s):  
B. Golchert ◽  
S. L. Chang ◽  
M. Petrick ◽  
C. Q. Zhou
Keyword(s):  
Heat Transfer ◽  
Computer Simulations ◽  
Glass Furnace ◽  
Validation Studies ◽  
Radiative Heat ◽  
Operating Conditions ◽  
Department Of Energy ◽  
Gas Temperature ◽  
Gas Velocity ◽  
Actual Operating

Abstract A completely coupled glass furnace simulation that combines calculations for the combustion space, glass melt, and radiative heat transfer throughout the furnace has been developed as part of a Department of Energy sponsored program. A major component of this program entailed the collection of an extensive set of data from an operating glass furnace for the purpose of validation. The data collected were compared with the results from computer simulations of the furnace based on the actual operating conditions. These comparisons indicated that the computated results were valid. This paper will present and discuss the gas velocity and gas temperature measurements/validation studies.

scholarly journals Thermal-hydraulic modeling of the steady-state operating conditions of a fire-tube boiler

2009 ◽  
Vol 24 (1) ◽  
pp. 29-37 ◽  
Author(s):  
Ahmed Rahmani ◽  
Ahmed Dahia
Keyword(s):  
Heat Transfer ◽  
Steady State ◽  
Hydraulic Modeling ◽  
Sensitivity Study ◽  
Operating Conditions ◽  
Saturated Steam ◽  
Operating Pressure ◽  
Operating Data ◽  
Boiler Operation ◽  
Calculation Results

In this work, we are interested to simulate the thermal-hydraulic behavior of three-pass type fire-tube boiler. The plant is designed to produce 4.5 tons per hour of saturated steam at 8 bar destined principally for heating applications. A calculation program is developed in order to simulate the boiler operation under several steady-state operating conditions. This program is based upon heat transfer laws between hot gases and the fire-tube internal walls. In the boiler combustion chamber, the heat transfer has been simulated using the well-stirred furnace model. In the convection section, heat balance has been carried out to estimate the heat exchanges between the hot gases and the tube banks. The obtained results are compared to the steady-state operating data of the considered plant. A comparative analysis shows that the calculation results are in good agreement with the boiler operating data. Furthermore, a sensitivity study has been carried out to assess the effects of input parameters, namely the fuel flow rate, air excess, ambient temperature, and operating pressure, upon the boiler thermal performances.

scholarly journals Heat transfer performance of an oil jet impinging on a downward-facing stainless steel plate

2011 ◽  
Vol 15 (2) ◽  
pp. 397-408 ◽  
Author(s):  
Roy Issa
Keyword(s):  
Heat Transfer ◽  
Stainless Steel ◽  
Steel Plate ◽  
Considerable Effect ◽  
Operating Conditions ◽  
Bottom Surface ◽  
Plate Surface ◽  
Stainless Steel Plate ◽  
Plate Temperature ◽  
Oil Jet

An experimental study is carried out for the quenching of a stainless steel plate using a single oil jet impinging on the bottom surface of the plate. The objective of this study is to investigate the effect of the oil jet flow operating conditions onto the heat transfer effectiveness when the plate is heated to temperatures ranging from around 115 to 630?C, and the oil is heated to temperatures ranging from 60 to 75?C. Tests are conducted on the oil at various temperatures to determine its viscosity. Experiments are conducted for nozzle exit flow rates ranging from 113 to 381 ml/min, oil jet pressures from 3.1 to 12 psi, and nozzle-to-plate surface distances of 0.6 and 1 cm. The variation of the oil heat flux and heat transfer coefficient with the surface temperature for the different quenching parameters is calculated from the acquired temperature data. Tests results show the oil heat transfer effectiveness keeps increasing for increasing plate temperature. Oil jet pressure is shown to have a considerable effect on the oil heat transfer, while the nozzle-to-plate surface distance is shown to have a lesser effect. The results of this study shall lead to a better understanding of the parameters that play an important role in oil quenching for applications that are of interest to the metal process industry.

scholarly journals Status of the Combined Third and Fourth NGNP Fuel Irradiations in the Advanced Test Reactor

2013 ◽  
Author(s):  
S. Blaine Grover ◽  
David A. Petti ◽  
Michael E. Davenport
Keyword(s):  
United States ◽  
Fission Product ◽  
National Laboratory ◽  
The United States ◽  
Operating Conditions ◽  
United States Department ◽  
Next Generation ◽  
Monitoring And Control ◽  
On Line ◽  
Test Reactor

The United States Department of Energy’s Next Generation Nuclear Plant (NGNP) Advanced Gas Reactor (AGR) Fuel Development and Qualification Program is irradiating up to seven low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States. The experiments will be irradiated over the next several years to demonstrate and qualify new TRISO coated particle fuel for use in high temperature gas reactors. The goals of the experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which will each consist of several independent capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gas will also have on-line fission product monitoring on its effluent to track performance of the fuel in each individual capsule during irradiation. The first experiment (designated AGR-1) started irradiation in December 2006 and was completed in November 2009. The second experiment (AGR-2) started irradiation in June 2010 and is currently scheduled to be completed in September 2013. The third and fourth experiments have been combined into a single experiment designated (AGR-3/4), which started its irradiation in December 2011 and is currently scheduled to be completed in April 2014. Since the purpose of this combined experiment is to provide data on fission product migration and retention in the NGNP reactor, the design of this experiment is significantly different from the first two experiments, though the control and monitoring systems are extremely similar. The design of the experiment will be discussed followed by its progress and status to date.

Overview of Creep Strength and Oxidation of Heat-Resistant Alloy Sheets and Foils for Compact Heat-Exchangers

2005 ◽  
Author(s):  
Philip J. Maziasz ◽  
John P. Shingledecker ◽  
Bruce A. Pint ◽  
Neal D. Evans ◽  
Yukinori Yamamoto ◽  
...  
Keyword(s):  
Stainless Steel ◽  
Heat Exchangers ◽  
Gas Turbines ◽  
High Efficiency ◽  
National Laboratory ◽  
Department Of Energy ◽  
Oak Ridge ◽  
Heat Resistant Alloy ◽  
Heat Resistant ◽  
Industrial Gas Turbines

The Oak Ridge National Laboratory (ORNL) has been involved in research and development related to improved performance of recuperators for industrial gas turbines since about 1996, and in improving recuperators for advanced microturbines since 2000. Recuperators are compact, high efficiency heat-exchangers that improve the efficiency of smaller gas turbines and microturbines. Recuperators were traditionally made from 347 stainless steel and operated below or close to 650°C, but today are being designed for reliable operation above 700°C. The Department of Energy (DOE) sponsored programs at ORNL have helped defined the failure mechanisms in stainless steel foils, including creep due to fine grain size, accelerated oxidation due to moisture in the hot exhaust gas, and loss of ductility due to aging. ORNL has also been involved in selecting and characterizing commercial heat-resistant stainless alloys, like HR120 or the new AL20-25+Nb, that should offer dramatically improved recuperator capability and performance at a reasonable cost. This paper summarizes research on sheets and foils of such alloys over the last few years, and suggests the next likely stages for manufacturing recuperators with upgraded performance for the next generation of larger 200–250 kW advanced microturbines.

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