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Showing 1 to 20 of 67 for “"Research Reactor"”.
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Neutron radiography with the VPI & SU research reactor
… of neutron radiography with the VPI & SU Research Reactor has been completed. Various experimental facilities of the reactor were tested to determine which gave the best beam parameters. The beam parameters considered were: 1) thermal neutron flux; 2) thermal to fast-flux ratio; 3) gamma …
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Multi-Purpose Research Reactor Designs Using MCNP and Origen
Made available in DSpace on 2017-07-06T18:48:39Z (GMT). No. of bitstreams: 3 Hu_Jianwei_MS.pdf: 12941059 bytes, checksum: bab58c978d20c697c63132361864a936 (MD5) Hu_Jianwei_MS_ABS.pdf: 35914 bytes, checksum: e11195f3129796aebd383242e916a02c (MD5) license.txt: 4813 bytes, checksum: …
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PREDICTING ACTIVATION OF EXPERIMENTS INSIDE THE ANNULAR CORE RESEARCH REACTOR
… experiments conducted inside of the Annular Core Research Reactor (ACRR) at Sandia National Laboratories and eliminate the need for users to write code. This estimation is achieved by using MCNP to model the neutron fluxes in the reactors central cavity where experiments are conducted using one of …
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The Determination of the Thermal Neutron Flux in a Research Reactor
Since a research reactor serves as a neutron source for a large number of experiments, knowledge of the neutron flux, and of its spatial and spectral energy distribution is of great importance, not only for purposes of experimental reactor physics (e.g., determination of specific data with regard …
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Design of central irradiation facilities for the MITR-II research reactor.
Thesis. 1977. M.S.--Massachusetts Institute of Technology. Dept. of Nuclear Engineering.
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Thermal hydraulic mixing transients in the MIT research reactor core tank
Thesis (Ph. D.)--Massachusetts Institute of Technology, Dept. of Nuclear Engineering, 1996.
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Estimate of radiation release for MIT research reactor during design basis accident
… a postulated design basis accident at the MIT Research Reactor (MITR), radioactive fission products may be released from melted fuel plates into the containment. To comply with regulations, the whole-body dose and thyroid dose at the boundary of the exclusion area as a result of this accident …
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Analysis of the MIT research reactor fission product and actinide radioactivity inventories
The current analysis of the MITR core radioactivity inventory eliminates unnecessary assumptions made in previous estimates of the inventory, and revises the list of contributory isotopes to include all actinide and fission product isotopes necessary for a proper accident source term calculation. …
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A temperature study of the V.P.I. training and research reactor (UTR-10)
… in the designing and construction of any nuclear reactor system is the removal of heat from the core of the reactor. In the V.P.I. UTR-10, the heat is removed from the fuel plates by the light water moderator-coolant which is pumped between the plates. The maximum temperature which the fuel will …
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The inverse kinetics method and PID compensation of the Annular Core Research Reactor
… of a model describing the Annular Core Research Reactor (ACRR), the application of the inverse kinetics method to calculate the current reactivity level within the ACRR model, and the development of a PID compensator to automatically control the reactivity level within the model and thus …
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Impact assessment for the MIT Research Reactor low enrichment uranium fuel fabrication tolerances
… the Massachusetts Institute of Technology Reactor (MITR) is planning to convert from highly enriched uranium (HEU) to low enriched uranium (LEU) fuel. A new type of high-density LEU fuel based on a monolithic U-10Mo alloy is being qualified to allow the conversion of all remaining U.S. high …
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Design and optimization of a high thermal flux research reactor via Kriging-based algorithm
… to increasing demands for the services of research reactors, a 5 MW LEU-fueled research reactor core is developed and optimized to provide high thermal flux within specified limits upon thermal hydraulic performance, cycle length, irradiation utilization, and manufacturability. A novel fuel …
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Space-Time Kinetics of the AGN-201M Research Reactor at the University of New Mexico
… measurements were performed on the AGN- 201M reactor (AGN) at the University of New Mexico. Steady-state measurements were made with the AGN in various delayed critical configurations. These delayed critical configurations are defined by moving the coarse and fine control rods to unique …
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Feasibility Assessment of Adopting the ADDER Computer Code for the MIT Research Reactor Fuel Management
The Massachusetts Institute of Technology Reactor (MITR) is a 6 MW research reactor currently operating with highly enriched uranium (HEU) plate-type fuel. Fuel management calculations for this reactor are performed using MCODE, which allows for the coupling of a neutron transport code and a …
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Analysis of in-core experiment activities for the MIT Research Reactor using the ORIGEN computer code
… in in-core experimental assemblies at the MIT Research Reactor (MITR-II). ORIGEN-S is a nuclear depletion and decay analysis code. It accounts for all types of nuclear reactions and eliminates the need for selection of the dominant reactions that will occur in a given experiment, as must be …
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Friction pressure drop measurements and flow distribution analysis for LEU conversion study of MIT Research Reactor
The MIT Nuclear Research Reactor (MITR) is the only research reactor in the United States that utilizes plate-type fuel elements with longitudinal fins to augment heat transfer. Recent studies on the conversion to low-enriched uranium (LEU) fuel at the MITR, together with the supporting thermal …
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Thermal analysis of the reflector system of the Massachusetts Institute of Technology Research Reactor (MITR-III)
Thesis (M.S.)--Massachusetts Institute of Technology, Dept. of Nuclear Engineering, 1996.
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Maximum hypothetical accident analysis for HEU to LEU fuel conversion at the University of Missouri Research Reactor
… Initiative (GTRI) and the Reduced Enrichment for Research and Test Reactors (RERTR) programs, the University of Missouri Research Reactor (MURR) intends to convert from 93% highly-enriched uranium (HEU) aluminide dispersion fuel to 19.75% low-enrichment uranium (LEU) U-Mo monolithic fuel. Under …
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Design of a low enrichment, enhanced fast flux core for the Massachusetts Institute of Technology Research Reactor
Worldwide, there is limited test reactor capacity to perform the required irradiation experiments on advanced fast reactor materials and fuel designs. This is particularly true in the U.S., which no longer has an operating fast reactor but depends upon two aging thermal reactors for testing the …
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