Conceptual Design of Molten Salt Loop Experiment for MIT Research Reactor by Bean, Malcolm K., Dewitt, Gregory Lee, Cabeche, Dion T., Gerrity, Thomas P., Haratyk, Geoffrey, Kersting, Alyssa R., Lee, Youho, Virgen, Matthew M., Lenci, Giancarlo, Lin, Christie, Metzler, Florian, Ochoukov, Roman Igorevitch, Reed, Mark, Sobes, Vladimir, Sugrue, Rosemary M., Shwageraus, Eugene, Wisniowska, Agata Elzbieta, Youchak, Paul M.

“…In addition, preliminary cost estimates of constructing the loop experiment at MITR were performed. The main constraints on the choice of the loop’s individual components and diagnostics were: 1) the ability to function at the designed operating temperature, pressure, and flow rates; 2) the ability to function in a nuclear radiation environment; and 3) the necessity to meet MITR safety requirements. …”
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Research and educational activities at the MIT Research Reactor : Fiscal year 1968

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Feasibility Assessment of Adopting the ADDER Computer Code for the MIT Research Reactor Fuel Management by Garanzini, Maurane

“…The Massachusetts Institute of Technology Reactor (MITR) is a 6 MW research reactor currently operating with highly enriched uranium (HEU) plate-type fuel. …”
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Analysis of in-core experiment activities for the MIT Research Reactor using the ORIGEN computer code by Helvenston, Edward M. (Edward March)

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Estimate of radiation release for MIT research reactor during design basis accident by Li, Qing

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Thermal hydraulics analysis of the MIT research reactor in support of a low enrichment uranium (LEU) core conversion by Ko, Yu-Chih, Ph. D. Massachusetts Institute of Technology

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Friction pressure drop measurements and flow distribution analysis for LEU conversion study of MIT Research Reactor by Wong, Susanna Yuen-Ting

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Engineering design of a fission converter-based epithermal beam for neutron capture therapy by Sutharshan, Balendra, 1966-

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Thermal hydraulic limits analysis for the MIT Research Reactor low enrichment uranium core conversion using statistical propagation of parametric uncertainties by Chiang, Keng-Yen

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Quantifying measurement uncertainties of MIT Research Reactor's new digital nuclear safety system by Parus, Rachel (Rachel Elizabeth)

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In vivo boron-10 analysis for the pre-screening of compounds for BNCS by Zhu, Xuping, 1970-

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Development of a Graphical User Interface for In-Core Fuel Management Using MCODE by Romano, Paul Kollath, Newton, Thomas H., Jr., Forget, Benoit Robert Yves

“…Together, these tools will help the staff at the MITR perform in-core fuel management calculations quickly and with a higher level of detail than that previously possible.…”
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LEU-HEU Mixed Core Conversion Thermal-hydraulic Analysis and Coolant System Upgrade Assessment for the MIT Research Reactor by Zhao, Yinjie

“…The MIT Research Reactor (MITR) is in the process of converting from the current 93%-enriched U-235 highly-enriched uranium (HEU) fuel to the low enriched uranium (LEU, <20%-enriched U-235) fuel, as part of the global non-proliferation initiatives. …”
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Simulation of Irradiation of a Molten Salt Loop at the MIT Reactor by Carayannopoulos, Loukas

“…The Nuclear Energy University Program approved a project to design and build a molten salt test bed that will be irradiated by the MIT Reactor (MITR), which will provide testing capabilities that have not existed since the shutdown of the Molten Salt Reactor Experiment (MSRE) in 1969. …”
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Evaluation of the Thermal-Hydraulic Operating Limits of HEU-LEU Transition Cores for the MIT Research Reactor by Wan, Yunzhi, Hu, Lin-Wen

“…The MIT Research Reactor (MITR) is in the process of conducting a design study to convert from High Enrichment Uranium (HEU) fuel to Low Enrichment Uranium (LEU) fuel. …”
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Evaluation of the thermal-hydraulic operating limits of the HEU-LEU transition cores for the MIT Research Reactor by Wang, Yunzhi (Yunzhi Diana)

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Orientation of gamma-ray telescope by Moore, Charles H

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Tritium retention in nuclear graphite, system-level transport, and management strategies for the fluoride-salt-cooled high-temperature reactor by Dolan, Kieran Patrick.

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Behavior of triplex silicon carbide fuel cladding designs tested under simulated PWR conditions by Stempien, John D. (John Dennis)

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Aspects of a high intensity neutron source by Chapman, Peter H. (Peter Henry)

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