Department of Reactor Safety
Neutron physics and reactor dynamics
Plant dynamics and severe accident analysis
Monte-Carlo simulations (n- / γ-field calculations)
Impact of Thermal-Hydraulic Feedback and Differential Thermal Expansion on European SFR Core Power Distribution
Lindley, B.; Álvarez Velarde, F.; Baker, U.; Bodi, J.; Cosgrove, P.; Charles, A.; Fiorina, C.; Fridman, E.; Krepel, J.; Lavarenne, J.; Mikityuk, K.; Nikitin, E.; Ponomarev, A.; Radman, S.; Shwageraus, E.; Tollit, B.
The objective of this paper is to quantify the coupling effect on the power distribution of sodium-cooled
fast reactors (SFRs), specifically the European SFR. Calculations are performed with several state-of-the-art
reactor physics and Multiphysics codes (TRACE/PARCS, DYN3D, WIMS, COUNTHER and GeN-Foam) to build
confidence in the methodologies and validity of results. Standalone neutronics calculations were generally
in excellent agreement with a reference Monte Carlo-calculated power distribution (from Serpent). Next,
the impact of coolant density and fuel temperature Doppler feedback was calculated. Reactivity
coefficients for perturbations in the inlet temperature, coolant heat up and core power were shown to be
negative with values of around -0.5 pcm/°C, -0.3 pcm/°C and -3.5 pcm/% respectively. Fuel temperature
and coolant density feedback was found to introduce a roughly -1%/+1% in/out power tilt across the core.
Calculations were then extended to axial expansion for cases where fuel is linked and unlinked to the clad.
Core calculations are in good agreement with each other. The impact of differential fuel expansion is found
to be larger for fuel both linked and unlinked to the clad, with the in/out power tilt increasing to around -
4%/+2%. Thus, while broadly confirming the known result that standalone physics calculations give good
results, the expansion coupling effect is perhaps more than anticipated a priori. These results provide a
useful benchmark for the further development of Multiphysics codes and methodologies in support of
advanced reactor calculations.
Journal of Nuclear Engineering and Radiation Science (2023)
Online First (2023) DOI: 10.1115/1.4056930
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