Circular Arc Fuel Plate Stability Experiments and Analyses for the Advanced Neutron Source PDF Download
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Languages : en
Pages :
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Book Description
The thin fuel plates planned for the Advanced Neutron Source are to be cooled by forcing heavy water at high velocity, 25 m/s, through thin cooling channels on each side of each plate. Because the potential for structural failure of the plates is a design concern, considerable effort has been expended in assessing this potential. As part of this effort, experimental flow tests and analyses to evaluate the structural response of circular arc plates have been conducted, and the results are given in this report.
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Category :
Languages : en
Pages :
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Book Description
The thin fuel plates planned for the Advanced Neutron Source are to be cooled by forcing heavy water at high velocity, 25 m/s, through thin cooling channels on each side of each plate. Because the potential for structural failure of the plates is a design concern, considerable effort has been expended in assessing this potential. As part of this effort, experimental flow tests and analyses to evaluate the structural response of circular arc plates have been conducted, and the results are given in this report.
Author:
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ISBN:
Category :
Languages : en
Pages : 67
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Book Description
The reactor for the planned Advanced Neutron Source uses closely spaced plates cooled by heavy water flowing through narrow channels. Two sets of tests were performed on the upper and lower fuel plates for the structural response of the fuel plates to the required high coolant flow velocities. This report contains the data from the second round of tests. Results and conclusions from all of the tests are also included in this report. The tests were done using light water on full-scale epoxy models, and through model theory, the results were related to the prototype plates, which are aluminum-clad aluminum/uranium silicide involute-shaped plates.
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Category :
Languages : en
Pages : 48
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Book Description
The planned reactor for the Advanced Neutron Source (ANS) will use closely spaced arrays of involute-shaped fuel plates that will be cooled by water flowing through the channels between the plates. There is concern that at certain coolant flow velocities, adjacent plates may deflect and touch, with resulting failure of the plates. Experiments have been conducted at the Oak Ridge National Laboratory to examine this potential phenomenon. Results of the experiments and comparison with analytical predictions are reported. The tests were conducted using full-scale epoxy plate models of the aluminum/uranium silicide ANS involute-shaped fuel plates. Use of epoxy plates and model theory allowed lower flow velocities and pressures to explore the potential failure mechanism. Plate deflections and channel pressures as functions of the flow velocity are examined. Comparisons with mathematical models are noted.
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Category :
Languages : en
Pages : 43
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Book Description
The reactor for the planned Advanced Neutron Source will use closely spaced arrays of fuel plates. The plates are thin and will have a core containing enriched uranium silicide fuel clad in aluminum. The heat load caused by the nuclear reactions within the fuel plates will be removed by flowing high-velocity heavy water through narrow channels between the plates. However, the plates will still be at elevated temperatures while in service, and the potential for excessive plate deformation because of creep must be considered. An analysis to include creep for deformation and stresses because of temperature over a given time span has been performed and is reported herein.
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Category :
Languages : en
Pages : 43
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Book Description
The Advanced Neutron Source (ANS) reactor was designed to provide a research tool with capabilities beyond those of any existing reactors. One portion of its state-of-the-art design required high-speed fluid flow through narrow channels between the fuel plates in the core. Experience with previous reactors has shown that fuel plate damage can occur when debris becomes lodged at the entrance to these channels. Such debris disrupts the fluid flow to the plate surfaces and can prevent adequate cooling of the fuel. Preliminary ANS designs addressed this issue by providing an unheated entrance length for each fuel plate so that any flow disruption would recover, thus providing adequate heat removal from the downstream, heated portions of the fuel plates. As part of the safety analysis, the adequacy of this unheated entrance length was assessed using both analytical models and experimental measurements. The Flow Blockage Test Facility (FBTF) was designed and built to conduct experiments in an environment closely matching the ANS channel geometry. The FBTF permitted careful measurements of both heat transfer and hydraulic parameters. In addition to these experimental efforts, a thin, rectangular channel was modeled using the Fluent computational fluid dynamics computer code. The numerical results were compared with the experimental data to benchmark the hydrodynamics of the model. After this comparison, the model was extended to include those elements of the safety analysis that were difficult to measure experimentally. These elements included the high wall heat flux pattern and variable fluid properties. The results were used to determine the relationship between potential blockage sizes and the unheated entrance length required.
Author: Edward R. Gotsky
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ISBN:
Category : Diffusion
Languages : en
Pages : 28
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Book Description
Effects of fuel plates successively withdrawn from the center fuel element of a seven-by-three core loading at the Oak Ridge Bulk Shielding Facility were evaluated by two-dimensional two-group diffusion calculations performed on the NASA reactor simulator. Two calculation methods were used: (1) The slowing-down properties of the experimental fuel element were represented by infinite media parameters; and (2) the finite size of the experimental fuel element was recognized, and the slowing-down properties of the surrounding core were attributed to this small region. Reasonable agreement existed between experimental and calculated effects.
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Category :
Languages : en
Pages :
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Book Description
The thin aluminum-clad fuel plates proposed for the Advanced Neutron Source reactor are stressed by the high-velocity coolant flowing on each side of the plates and by the thermal gradients in the plates. The total stress, composed of the sum of the flow stress and the thermal stress at a point, could be reduced if the thermal loads tend to relax when the stress magnitude approaches the yield stress of the material. The potential of this occurring would be very significant in assessing the structural reliability of the fuel plates and has been investigated through experiment. The results of this investigation are given in this report.
Author:
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ISBN:
Category : Government publications
Languages : en
Pages : 600
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Author:
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ISBN:
Category : Nuclear energy
Languages : en
Pages : 730
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Author:
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ISBN:
Category :
Languages : en
Pages : 99
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Book Description
This report discusses the following about the Advanced Neutron Source: Project Management; Research and Development; Fuel Development; Corrosion Loop Tests and Analyses; Thermal-Hydraulic Loop Tests; Reactor Control and Shutdown Concepts; Critical and Subcritical Experiments; Material Data, Structural Tests, and Analysis; Cold-Source Development; Beam Tube, Guide, and Instrument Development; Hot-Source Development; Neutron Transport and Shielding; I C Research and Development; Design; and Safety.
