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Preliminary design and assessment of a heat pipe residual heat removal system for the reactor driven subcritical facility

Nuclear Engineering and Technology, 2021, 53(12), , pp.3879-3891

2021

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The effects of neutron and ionizing irradiation on the aqueous corrosion of SiC

Journal of nuclear materials, 2020-08, Vol.536 (C), p.152190, Article 152190

2020

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Experimental demonstration of a data-driven control system for subcritical nuclear facility

Progress in nuclear energy (New series), 2024-03, Vol.168 (C), p.104977, Article 104977

2024

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Experimental validation of a high fidelity Monte Carlo neutron transport model of the MIT graphite exponential pile

Progress in nuclear energy (New series), 2022-10, Vol.152 (C), p.104368, Article 104368

2022

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Tritium generation, release, and retention from in-core fluoride salt irradiations

Progress in nuclear energy (New series), 2021-01, Vol.131 (C), p.103576, Article 103576

2021

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Machine learning based system performance prediction model for reactor control

Annals of nuclear energy, 2018-03, Vol.113 (C), p.270-278

2018

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The effects of neutron and ionizing irradiation on the aqueous corrosion of SiC

Journal of nuclear materials, 2020-08, Vol.536 (C)

2020

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Experimental validation of a high fidelity Monte Carlo neutron transport model of the MIT graphite exponential pile

Progress in nuclear energy (New series), 2022-08, Vol.152

2022

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The effects of neutron and ionizing irradiation on the aqueous corrosion of SiC

Journal of nuclear materials, 2020-05, Vol.536 (1)

2020

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Tritium generation, release, and retention from in-core fluoride salt irradiations

Progress in nuclear energy (New series), 2020-11, Vol.131 (C)

2020

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Neutronics feasibility of an MIT Reactor–driven subcritical facility for the Fluoride‐salt–cooled High‐temperature Reactor

International journal of energy research, 2017-11, Vol.41 (14), p.2248-2257

2017

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Thermal‐hydraulic design and analysis of a small modular molten salt reactor (MSR) with solid fuel

International journal of energy research, 2018-03, Vol.42 (3), p.1098-1114

2018

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Neutronic Design Features of a Transportable Fluoride-Salt-Cooled High-Temperature Reactor

Journal of nuclear engineering and radiation science, 2016-07, Vol.2 (3)

2016

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Miniature Sensor Irradiation Tests Under Steady-State and Transient Conditions at MIT Research Reactor (MITR) and Transient Reactor Test Facility (TREAT)

Nuclear technology, 2019-12, Vol.206 (6)

2019

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Thermal-hydraulic analyses of MIT reactor LEU transition cycles

Progress in nuclear energy (New series), 2020-01, Vol.118, p.103117, Article 103117

2020

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Preliminary design and assessment of a heat pipe residual heat removal system for the reactor driven subcritical facility

Nuclear engineering and technology, 2021, Vol.53 (12), p.3879-3891

2021

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Thermal-hydraulic analyses of MIT reactor LEU transition cycles

Progress in nuclear energy (New series), 2019-08, Vol.118 (C)

2019

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Neutronics modeling and analysis of the TMSR-SF1 fuel lattice and full core with explicit fuel particle distribution and random pebble loadings

Progress in nuclear energy (New series), 2018-11, Vol.109, p.171-179

2018

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Miniature Sensor Irradiation Tests Under Steady-State and Transient Conditions at MIT Research Reactor (MITR) and Transient Reactor Test Facility (TREAT)

Nuclear technology, 2020-06, Vol.206 (6), p.924-937

2020

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Thermal-Hydraulic Analyses of Transportable Fluoride Salt-Cooled High-Temperature Reactor with CFD Modeling

Nuclear technology, 2016-10, Vol.196 (1), p.34-52

2016

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Transitional cores and fuel cycle analyses in support of MIT reactor low enriched uranium fuel conversion

Progress in nuclear energy (New series), 2019-10, Vol.119 (C)

2019

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Uncertainty analysis of Transportable Fluoride-salt-cooled High-temperature Reactor (TFHR) using coupled DAKOTA with RELAP-3D method

Nuclear engineering and design, 2017-12, Vol.324 (C), p.269-279

2017

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Transitional cores and fuel cycle analyses in support of MIT reactor low enriched uranium fuel conversion

Progress in nuclear energy (New series), 2020-01, Vol.119 (C), p.103171, Article 103171

2020

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Estimate of radiation release from MIT reactor with un-finned LEU core during Maximum Hypothetical Accident

Progress in nuclear energy (New series), 2017-05, Vol.97 (C), p.231-244

2017

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Estimate of radiation release from MIT reactor with un-finned LEU core during Maximum Hypothetical Accident

Progress in nuclear energy (New series), 2017-05, Vol.97 (C)

2017

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Preliminary design and assessment of a heat pipe residual heat removal system for the reactor driven subcritical facility

The effects of neutron and ionizing irradiation on the aqueous corrosion of SiC

Experimental demonstration of a data-driven control system for subcritical nuclear facility

Experimental validation of a high fidelity Monte Carlo neutron transport model of the MIT graphite exponential pile

Tritium generation, release, and retention from in-core fluoride salt irradiations

Machine learning based system performance prediction model for reactor control

The effects of neutron and ionizing irradiation on the aqueous corrosion of SiC

Experimental validation of a high fidelity Monte Carlo neutron transport model of the MIT graphite exponential pile

The effects of neutron and ionizing irradiation on the aqueous corrosion of SiC

Tritium generation, release, and retention from in-core fluoride salt irradiations

Neutronics feasibility of an MIT Reactor–driven subcritical facility for the Fluoride‐salt–cooled High‐temperature Reactor

Thermal‐hydraulic design and analysis of a small modular molten salt reactor (MSR) with solid fuel

Neutronic Design Features of a Transportable Fluoride-Salt-Cooled High-Temperature Reactor

Miniature Sensor Irradiation Tests Under Steady-State and Transient Conditions at MIT Research Reactor (MITR) and Transient Reactor Test Facility (TREAT)

Thermal-hydraulic analyses of MIT reactor LEU transition cycles

Preliminary design and assessment of a heat pipe residual heat removal system for the reactor driven subcritical facility

Thermal-hydraulic analyses of MIT reactor LEU transition cycles

Neutronics modeling and analysis of the TMSR-SF1 fuel lattice and full core with explicit fuel particle distribution and random pebble loadings

Miniature Sensor Irradiation Tests Under Steady-State and Transient Conditions at MIT Research Reactor (MITR) and Transient Reactor Test Facility (TREAT)

Thermal-Hydraulic Analyses of Transportable Fluoride Salt-Cooled High-Temperature Reactor with CFD Modeling

Transitional cores and fuel cycle analyses in support of MIT reactor low enriched uranium fuel conversion

Uncertainty analysis of Transportable Fluoride-salt-cooled High-temperature Reactor (TFHR) using coupled DAKOTA with RELAP-3D method

Transitional cores and fuel cycle analyses in support of MIT reactor low enriched uranium fuel conversion

Estimate of radiation release from MIT reactor with un-finned LEU core during Maximum Hypothetical Accident

Estimate of radiation release from MIT reactor with un-finned LEU core during Maximum Hypothetical Accident

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