Fluoride-Salt-Cooled High-Temperature Test Reactor Thermal-Hydraulic Licensing and Uncertainty Propagation Analysis

Tritium Production Analysis and Management Strategies for a Fluoride - salt - cooled High - temperature Test Reactor ( FHTR )

The Fluoride-salt-cooled High-temperature Test Reactor (FHTR) is a test reactor concept that aims to demonstrate the neutronics, thermal-hydraulics, materials, tritium management, and to address other reactor operational and maintenance issues before a commercial Fluoride-salt-cooled High-temperature Reactor (FHR) can be deployed. The MIT Nuclear Systems Design class proposed a design for a 100...

Variable Electricity with Base-load Reactor Operations Fluoride-salt-cooled High-temperature Reactor (FHR) with Nuclear Air-Brayton Combined Cycle (NACC) and Firebrick Resistance-Heated Energy Storage

In this century man will transition to a low-carbon energy future—either in first half of the century because of concerns about global climate and ocean pH (acidity) changes or in the second half of the century because of depletion of fossil resources in a world of 10 billion people. Since the caveman discovered fire, our energy policy has been to have a storable supply of a carbon fuel (wood, ...

The Advanced High-Temperature Reactor: High-Temperature Fuel, Molten Salt Coolant, and Liquid-Metal-Reactor Plant

The Advanced High-Temperature Reactor is a new reactor concept that combines four existing technologies in a new way: (1) coated-particle graphite-matrix nuclear fuels (traditionally used for helium-cooled reactors), (2) Brayton power cycles, (3) passive safety systems and plant designs from liquid-metal-cooled fast reactors, and (4) low-pressure molten-salt coolants with boiling points far abo...

High Temperature Gas Cooled Reactors

Thermal Hydraulic Limits Analysis for the MIT Research Reactor Low Enrichment Uranium Core Conversion Using Statistical Propagation of Parametric Uncertainties

The MIT Research Reactor (MITR) is evaluating the conversion from highly enriched uranium (HEU) to low enrichment uranium (LEU) fuel. In addition to the fuel element redesign from 15 to 18 plates per element, a reactor power upgraded from 6 MW to 7 MW is proposed in order to maintain the same reactor performance of the HEU core. Previous approaches in analyzing the impact of engineering uncerta...