Thermal Safety Margin Calculation of the MP-2 Experiment in the Advanced Test Reactor

2020 ◽  
Author(s):  
Grant L. Hawkes
Keyword(s):  
Finite Element ◽  
Flow Instability ◽  
National Laboratory ◽  
Thermal Analyses ◽  
Safety Margin ◽  
Heat Rate ◽  
Element Model ◽  
Nucleate Boiling ◽  
Primary Coolant ◽  
Test Reactor

Abstract 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) at the Idaho National Laboratory (INL). The experiment is a drop-in test where small aluminum-clad fuel plate samples (mini plates) 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 with fuel meat thickness and cladding are part of the MP-2 test. A thermal analysis has been performed on the MP-2 experiment. A method for calculating Departure from Nucleate Boiling Ratio (DNBR) and Flow Instability Ratio (FIR) during a reactivity transient using the commercial finite element and heat transfer code ABAQUS is discussed. At the start of an ATR cycle the heat generation rate of the fueled experiment is high and the heat rate multiplier from the outer shim control cylinders is low, while the reverse is true at the end of the ATR cycle. Thermal analyses at 10-day increments during the cycle calculate the DNBR and FIR during a reactivity transient. This technique calculates DNBR for the fuel plate surfaces and FIR for all water components for each finite element surface and node at various times during the ATR cycle. Heat rates vary with time during the transient calculations that are provided by a detailed physics analysis. Oxide growth on the fuel plates is also incorporated. Results from the transient calculations are displayed with the ABAQUS post processor. 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.

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.

Thermal Analysis of the FSP-1 Experiment in the Advanced Test Reactor

2016 ◽  
Author(s):  
Grant L. Hawkes ◽  
Nicolas E. Woolstenhulme
Keyword(s):  
Thermal Analysis ◽  
Flow Instability ◽  
National Laboratory ◽  
Safety Evaluation ◽  
Thermal Safety ◽  
Primary Coolant ◽  
Coolant Pump ◽  
Safety Margins ◽  
Enriched Uranium ◽  
Test Reactor

The U.S. High Performance Research Reactor Conversions fuel development team is focused on developing and qualifying the uranium-molybdenum (U-Mo) alloy monolithic fuel to support conversion of domestic research reactors to low enriched uranium. Several previous irradiations have demonstrated the favorable behavior of the monolithic fuel. The Full Scale Plate 1 (FSP-1) fuel plate experiment will be irradiated in the northeast (NE) flux trap of the Advanced Test Reactor (ATR). This fueled experiment contains six aluminum-clad fuel plates consisting of monolithic U-Mo fuel meat. Three different types of fuel plates with matching pairs for a total of six plates were analyzed. These three types of plates are: full burn, intermediate power, and thick meat. A thermal analysis has been performed on the FSP-1 experiment to be irradiated in the ATR at the Idaho National Laboratory (INL). A thermal safety evaluation was performed to demonstrate that the FSP-1 irradiation experiment complies with the thermal-hydraulic safety requirements of the ATR Safety Analysis Report (SAR). The ATR SAR requires that minimum safety margins to critical heat flux and flow instability be met in the case of a loss of commercial power with primary coolant pump coast-down to emergency flow. The thermal safety evaluation was performed at 26 MW NE lobe power to encompass the expected range of operating power during a standard cycle. Additional safety evaluations of reactivity insertion events, loss of coolant event, and free convection cooling in the reactor and in the canal are used to determine the response of the experiment to these events and conditions. This paper reports and shows that each safety evaluation complies with each safety requirement of the ATR SAR.

Dynamic – Thermal Analyses of a Structurally Reconfigured Electronics Package Onboard Mini Satellite

2014 ◽  
Vol 592-594 ◽  
pp. 2117-2121 ◽  
Author(s):  
P. Veeramuthuvel ◽  
S. Jayaraman ◽  
Shankar Krishnapillai ◽  
M. Annadurai ◽  
A.K. Sharma
Keyword(s):  
Finite Element ◽  
Printed Circuit Board ◽  
Thermal Environment ◽  
Thermal Analyses ◽  
Simulation Models ◽  
Element Model ◽  
Circuit Board ◽  
Element Analysis ◽  
Transfer Path ◽  
Vibration Responses

The electronics package in a spacecraft is subjected to a variety of dynamic loads during launch phase and suitable thermal environment for the mission life. The dynamic and thermal analyses performed for a structurally reconfigured electronics package. Two different simulation models are developed to carry out the analyses. This paper discusses in two parts, in part-1, the vibration responses are determined at various critical locations, including on the Printed Circuit Board (PCB) for the vibration loads specified by launch vehicle using Finite Element Analysis (FEA). The mechanical properties of PCB are determined from the test specimens, which are then incorporated in the finite element model. In part-2, the steady-state temperature distributions on the components and on the PCB are determined, to check the effectiveness of heat transfer path from the components to the base of the package and to verify the predicted values are within the acceptable temperature limits specified. The predicted temperature values are then compared with on-orbit observations.

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.

scholarly journals ANALYSIS ON THE PERFORMANCE OF THE BANDUNG CONVERSION FUEL-PLATE TRIGA REACTOR IN STEADY STATE WITH CONSTANT COOLANT FLOW RATE

2020 ◽  
Vol 22 (2) ◽  
pp. 41
Author(s):  
Endiah Puji Hastuti ◽  
Sudjatmi K. Alfa ◽  
Sudarmono Sudarmono
Keyword(s):  
Steady State ◽  
Fuel Element ◽  
Flow Rate ◽  
Flow Instability ◽  
Safety Margin ◽  
Nucleate Boiling ◽  
Coolant Flow ◽  
Coolant Flow Rate ◽  
Steady State Condition ◽  
Triga Reactor

Bandung TRIGA2000 Reactor, a General Atomic (GA)-made research reactor used for training, research andiIsotope production, has been upgraded to operate at power of 2000 kW using TRIGA fuel rod type. Recently, the TRIGA reactor fuel element producers are going to discontinue the production of TRIGA fuel element. To overcome the unavailability of TRIGA fuel element, BATAN planned to modify TRIGA2000 fuel type from rod-type to U3Si2-Al plate-type fuel with 19.75% enrichment, similar to the domestically fabricated one used in RSG-GAS. The carried out design emphasized on the determination of operation condition limits for setting the reactor protection system in accordance to the reactor safety calculation results. The conceptual design of the innovative fuel plate TRIGA reactor cooling system is expected to remove heat generated by fuels with nominal power of 1 MW up to 2 MW. The design is developed through modelling and safety analysis using COOLOD-N2 validated code. The safety margin is set to its flow instability at transient condition of the fuel plate, which is ≥ 2.38; departure from nucleate boiling ratio ≥1.50; and no onset of nucleate boiling, ΔTONB ≥ 0oC. The primary coolant flow rate accommodating the existing Bandung TRIGA reactor capability is as high as 50 kg/s. The analysis results show that at power of 1 MW, the reactor can safely operate, while at power of 2 MW the safety margin is exceeded. In other words, the plate TRIGA reactor that employs forced convection mode operates safely at 1 MW with excess power 120% of its nominal power.Keywords: 1 MW, Thermalhydraulic design, Steady state condition, TRIGA plate, Constant flowrate

Structural Analysis of Light Aircraft Wing Components

2012 ◽  
Vol 225 ◽  
pp. 201-206
Author(s):  
Abdelmunem Bushra ◽  
Mohammed Mahdi ◽  
Mohammed A. Elhadi
Keyword(s):  
Finite Element ◽  
Finite Element Model ◽  
Safety Margin ◽  
Element Model ◽  
Geometrical Model ◽  
Aerodynamic Load ◽  
Element Mesh ◽  
Root Edge ◽  
Safety Margins ◽  
The Finite Element Model

This paper aims to demonstrate the structure analysis of strut-braced wing of a typical manufactured Light-Aircraft by using FEM software (MSC PATRAN/NASTRAN) and determine the safety margin in all of its components, which are useful to determine the structure strength requirements. The geometrical model of the wing was created in CATIA and then exported to PATRAN, which is the modeler to build the finite element model. PATRAN model geometry was modified and prepared to create the mesh. The structural components have various functions and shapes, thus different element mesh was created. After creating the finite element model for all parts, the elements and material properties were assigned and the model was fixed at the spar root edge and strut-braced end, and loaded by distributing the inertia load and aerodynamic load, calculated using (CFD), acting on the rib edge. Then the model was submitted to NASTRAN for linear static analysis. The obtained Stress Results and Safety Margins of each part were calculated and found to be acceptable.

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 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 Analysis of a Spur Gearbox Part 1: Steady State Finite Element Analysis

1988 ◽  
Vol 202 (4) ◽  
pp. 245-256 ◽  
Author(s):  
D Joule ◽  
S Hinduja ◽  
J N Ashton
Keyword(s):  
Finite Element ◽  
Steady State ◽  
Boundary Conditions ◽  
Experimental Work ◽  
Thermal Analyses ◽  
Element Model ◽  
Transfer Coefficients ◽  
Element Analysis ◽  
Thermal Boundary Conditions ◽  
Spur Gearbox

The accuracy of previously published thermal analyses of machine tools has been severely limited by imprecise thermal boundary conditions, in particular the values and distribution of heat inputs and heat-transfer coefficients. This paper describes an attempt to overcome some of the difficulties by determining the thermal boundary conditions for calculating the temperatures in a gearbox using the finite element method. A spur gearbox test rig has been designed and constructed, and a finite element model of the test gearbox developed. Temperature measurements and lubricant flow observations from experimental work have been combined with relevant theory to derive the boundary conditions. In the first part of this paper the experimental work and finite element steady state results are described. Sufficient agreement is evident between the two sets of results to indicate that the approach adopted here could be usefully extended to the analysis of other similar problems.

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