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Showing 1 to 20 of 22 for “"Thermal Hydraulic Analysis"”.
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Thermal/hydraulic analysis methods for PWR's
Thesis. 1976. Nucl.E.--Massachusetts Institute of Technology. Dept. of Nuclear Engineering.
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Thermal hydraulic analysis of hydride fuels in BWR's
… More specifically, this thesis deals with the thermal hydraulic analysis of BWR reactors. Several papers and theses have already been written for this project, mainly focusing on PWR reactors. The primary goal of this thesis is to find the optimal fuel rod lattice pitch and diameter such that a …
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Steady state thermal hydraulic analysis of hydride fueled BWRs
… the results obtained in the main body of the analysis account only for thermal-hydraulic constraints, an estimate of the power reduction due to the application of neutronic constraints is also performed. This investigation, focused only on the "New Core" cases, is coupled with an increase of …
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Thermal-hydraulic analysis of tight lattice light water reactors.
Thesis. 1977. M.S.--Massachusetts Institute of Technology. Dept. of Nuclear Engineering.
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Thermal-hydraulic analysis of a spectral shift reactor core
… program has been developed which performs a thermal-hydraulic analysis of a mechanical spectral shift reactor fuel assembly. In this reactor type, the fuel to water volume ratio is changed to shift the neutron energy spectrum and control reactivity. The ability to handle different fuel to …
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Two dimensional transport coefficients for the PWR's thermal/hydraulic analysis.
Thesis. 1976. M.S.--Massachusetts Institute of Technology. Dept. of Nuclear Engineering.
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Thermal-hydraulic analysis of innovative fuel configurations for the sodium fast reactor
… interaction (FCCI) for metallic fuel, cladding thermal and irradiation strain, the fuel melting point, sodium boiling, and to a lesser extent the sodium pressure drop in the fuel channels. Therefore, innovative fuel configurations that reduce clad stresses, sodium pressure drops, and fuel/clad …
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Multiscale approach for the thermal-hydraulic analysis of heavy liquid metal pool systems
L'abstract è presente nell'allegato / the abstract is in the attachment
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Thermal-hydraulic analysis of cross-shaped spiral fuel in high power density BWRs
Preliminary analysis of the cross-shaped spiral (CSS) fuel assembly suggests great thermal-hydraulic upside. According to computational models, the increase in rod surface area, combined with an increase in coolant turbulence and inter-channel mixing will allow for a greater than 25% uprate in …
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Reactor thermal-hydraulic analysis improvement and application of the code COBRA-IIIC/MIT
Thesis (Nucl.E.)--Massachusetts Institute of Technology, Dept. of Nuclear Engineering, 1981.
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Thermal-Hydraulic Analysis for the Proposed Upgrade of the Univeristy of Illinois Advanced TRIGA Reactor to 3 MW
Made available in DSpace on 2017-07-06T18:48:15Z (GMT). No. of bitstreams: 3 Golchert_Lucia_MS.pdf: 56434124 bytes, checksum: 34880e25e1902be6c97ff7f2373b2345 (MD5) Golchert_Lucia_MS_ABS.pdf: 591212 bytes, checksum: b4b3dbdb931bc4400e29804f1b1cd526 (MD5) license.txt: 4813 bytes, checksum: …
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LEU-HEU Mixed Core Conversion Thermal-hydraulic Analysis and Coolant System Upgrade Assessment for the MIT Research Reactor
… reactor power increases from 6.0 MW to 7.0 MW thermal, and primary coolant flow rate increases from 2000 gpm to 2400 gpm. Detailed analyses were completed for initial LEU core with 22 fuel elements, and demonstrated both neutronic and thermal hydraulic safety requirements are met throughout …
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A thermal-hydraulic analysis of the cooling system for the 500 KW Virginia Polytechnic Institute and State University reactor
… the feasibility of operation of 500 KW. From a thermal-hydraulic point of view, the following limiting conditions should not be exceeded for a 500 KW operation: 1) bulk boiling of the coolant should not occur, 2) surface boiling should not occur on any fuel plate, 3) fuel meltdown must not occur …
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Improved multidimensional numerical methods for the steady state and transient thermal-hydraulic analysis of fuel pin bundles and nuclear reactor cores.
Thesis: Sc. D., Massachusetts Institute of Technology, Department of Nuclear Engineering, 1977
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Neutronic Performance and Thermal Hydraulic Analysis of the MIT Reactor Fission Converter Experimental Facility Using High-Density U-10Mo Low-Enriched Uranium Fuel Elements
… graphite reflector. The FC serves to boost the thermal flux from the core and send a hardened neutron spectrum to an irradiation target, providing a fission energy flux spectrum without the need to put a sample inside the core tank. It was previously used for boron-neutron capture therapy …
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Thermal hydraulic performance analysis of a small integral pressurized water reactor core
A thermal hydraulic analysis of the International Reactor Innovative and Secure (IRIS) core has been performed. Thermal margins for steady state and a selection of Loss Of Flow Accidents have been assessed using three methodologies to account for uncertainty. The thermal hydraulic analysis has …
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Thermal hydraulic performance analysis of a small integral pressurized water reactor core
A thermal hydraulic analysis of the International Reactor Innovative and Secure (IRIS) core has been performed. Thermal margins for steady state and a selection of Loss Of Flow Accidents have been assessed using three methodologies to account for uncertainty. thermal margin for steady state …
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Reactor configurations to support advanced material research
… N-particle (MCNP) transport code, version 6. Thermal-hydraulic analysis was performed with ANSYS Fluent. The HPC was able to support high flux levels in the order of 1E14 <em>n cm</em><sup>-2</sup> <em>s</em><sup>-1</sup> with continuous operation over 2 months before refueling. The LPC …
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On-line refueling for the advanced high temperature reactor
… refueling operation. A systematic neutronic and thermal-hydraulic analysis approach has been developed and performed to assess the viability and safety of the refueling operations, followed by the evaluation of the core design requirements and a quantification of the economic advantages deriving …
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Optimization of a seed and blanket thorium-uranium fuel cycle for pressurized water reactors
… tools in the nuclear industry. The neutronic analysis was performed using the Studsvik Core Management System (CMS), which consists of three codes: CASMO-4, TABLES-3 and SIMULATE-3. Thermal-hydraulic analysis was performed using EPRI's VIPRE-01.
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