TY - JOUR
T1 - Neutronic Evaluation and Optimization of the Centrifugal Nuclear Thermal Rocket Concept
AU - Walters, William J.
N1 - Publisher Copyright:
© 2023 American Nuclear Society.
PY - 2023
Y1 - 2023
N2 - The centrifugal nuclear thermal rocket is a concept for a liquid-fueled nuclear system that would allow for a much higher specific impulse than the more traditional solid-fueled nuclear thermal propulsion designs. Although some preliminary neutronics analyses have been done on conceptual designs, this work seeks to perform a more systematic analysis and optimization of design parameters and to investigate additional neutronics properties such as power distributions and reactivity coefficients. This work used OpenMC for neutronics analysis and Dakota for the parametric study and optimization. Inter- and intra-fuel element power distributions were calculated, and a strong radial dependence was noted within fuel elements that may pose a challenge to thermal constraints. A positive moderator temperature coefficient of 3.78 (Formula presented.) 0.16 pcm/K was calculated for the reference model, which may pose a challenge for system design and control. The optimization study of reflector size, fuel spacing, fuel mass, and fuel element radius indicated many trade-offs in the design considerations, and that the baseline model can be significantly improved in all respects. Positive reactivity feedback can be minimized by reducing moderation, and peaking factors can be reduced by limiting the amount of fuel per fuel element, which also minimizes the system mass.
AB - The centrifugal nuclear thermal rocket is a concept for a liquid-fueled nuclear system that would allow for a much higher specific impulse than the more traditional solid-fueled nuclear thermal propulsion designs. Although some preliminary neutronics analyses have been done on conceptual designs, this work seeks to perform a more systematic analysis and optimization of design parameters and to investigate additional neutronics properties such as power distributions and reactivity coefficients. This work used OpenMC for neutronics analysis and Dakota for the parametric study and optimization. Inter- and intra-fuel element power distributions were calculated, and a strong radial dependence was noted within fuel elements that may pose a challenge to thermal constraints. A positive moderator temperature coefficient of 3.78 (Formula presented.) 0.16 pcm/K was calculated for the reference model, which may pose a challenge for system design and control. The optimization study of reflector size, fuel spacing, fuel mass, and fuel element radius indicated many trade-offs in the design considerations, and that the baseline model can be significantly improved in all respects. Positive reactivity feedback can be minimized by reducing moderation, and peaking factors can be reduced by limiting the amount of fuel per fuel element, which also minimizes the system mass.
UR - http://www.scopus.com/inward/record.url?scp=85148215156&partnerID=8YFLogxK
UR - http://www.scopus.com/inward/citedby.url?scp=85148215156&partnerID=8YFLogxK
U2 - 10.1080/00295639.2022.2161805
DO - 10.1080/00295639.2022.2161805
M3 - Article
AN - SCOPUS:85148215156
SN - 0029-5639
VL - 197
SP - 2150
EP - 2160
JO - Nuclear Science and Engineering
JF - Nuclear Science and Engineering
IS - 8
ER -