TY - GEN
T1 - Fuel Flow Modeling under Thermal Fluidics Considerations within Operational Domains of Molten Salt Reactors
AU - Abuqudaira, Thabit M.
AU - Tsvetkov, Pavel V.
AU - Sabharwall, Piyush
N1 - Publisher Copyright:
© 2023 Proceedings of the 20th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2023. All rights reserved.
PY - 2023
Y1 - 2023
N2 - Molten Salt Reactors (MSRs) offer significant versatility supported by their design flexibility and safety advantages due to high temperatures, design customization capabilities, strong negative reactivity feedback effects, and salt chemistry characteristics.Different control schedules and different mass flow rates in the MSR primary systems impact neutronics and thermal hydraulics coupling.The result is twofold.First, it leads to adaptability in various applications when the customization of a MSR unit is desired.Second, it calls for the quantification of dynamics and safety characteristics to assess the operational domain and make sure reactor stability and inherent safety characteristics are maintained.This paper provides an in-depth transient analysis of the operational implications of MSRs having heat-generating fuel and heat-transporting salt mixed forming liquid fuel salts contained within primary systems.A coupling model was developed to simulate fuel flow in molten salt reactors.A zero-dimensional reactor kinetics model was used with a one-dimensional heat transfer model in the reactor core.The temperature reactivity feedbacks of the fuel salt resulting from the thermal-hydraulics model were used in the reactor kinetics model to complete the coupled code.The model was first validated with the experimental results from the Molten Salt Reactor Experiment (MSRE).Then, it was applied to simulate the Molten Salt Breeder Reactor (MSBR) response at steady state and transients.The developed model was shown to be a suitable tool for the dynamic analysis of molten salt reactors.
AB - Molten Salt Reactors (MSRs) offer significant versatility supported by their design flexibility and safety advantages due to high temperatures, design customization capabilities, strong negative reactivity feedback effects, and salt chemistry characteristics.Different control schedules and different mass flow rates in the MSR primary systems impact neutronics and thermal hydraulics coupling.The result is twofold.First, it leads to adaptability in various applications when the customization of a MSR unit is desired.Second, it calls for the quantification of dynamics and safety characteristics to assess the operational domain and make sure reactor stability and inherent safety characteristics are maintained.This paper provides an in-depth transient analysis of the operational implications of MSRs having heat-generating fuel and heat-transporting salt mixed forming liquid fuel salts contained within primary systems.A coupling model was developed to simulate fuel flow in molten salt reactors.A zero-dimensional reactor kinetics model was used with a one-dimensional heat transfer model in the reactor core.The temperature reactivity feedbacks of the fuel salt resulting from the thermal-hydraulics model were used in the reactor kinetics model to complete the coupled code.The model was first validated with the experimental results from the Molten Salt Reactor Experiment (MSRE).Then, it was applied to simulate the Molten Salt Breeder Reactor (MSBR) response at steady state and transients.The developed model was shown to be a suitable tool for the dynamic analysis of molten salt reactors.
KW - Fuel Flow
KW - Fueled Salt
KW - Molten Salt Reactors
KW - Reactor Operation
KW - Transient Behavior
UR - https://www.scopus.com/pages/publications/85202872808
U2 - 10.13182/NURETH20-40919
DO - 10.13182/NURETH20-40919
M3 - Conference contribution
AN - SCOPUS:85202872808
T3 - Proceedings of the 20th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2023
SP - 3208
EP - 3221
BT - Proceedings of the 20th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2023
PB - American Nuclear Society
T2 - 20th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2023
Y2 - 20 August 2023 through 25 August 2023
ER -