Dr. Sören Kliem

Reactor Safety
Phone: +49 351 260 2318

Department of Reactor Safety


Neutron physics and reactor dynamics

  • Validation and application to light water reactors and innovative reactor concepts of the Monte Carlo code SERPENT2
  • Application of the deterministic lattice code HELIOS-2
  • Development, verification and application of the in-house reactor dynamics code DYN3D
  • Extension of the DYN3D code to innovative reactor concepts
  • Coupling of DYN3D to the system code ATHLET
  • Coupling of DYN3D to the Computational Fluid Dynamics codes ANSYS CFX and TRIO_U

Plant dynamics and severe accident analysis

  • Accident analysis and analysis of plant dynamics using the ATHLET system code
  • Assessment of the impact of severe accident management measures on the progression of severe accidents in PWRs and VVERs

Monte-Carlo simulations (n- / γ-field calculations)

  • Development of the in house Monte Carlo code TRAMO
  • Fluence calculations of the of the reactor pressure vessel and internals using MCNP and TRAMO

Latest Publication

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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FWOR Jörg Konheiser Evgeny Nikitin André Gommlich Dr. Alexander Grahn Reuven Rachamin Dr. Silvio Baier Kerstin Kurde Dr. Emil Fridman Dr. Frank Schäfer Dr. Sören Kliem


NameBld./Office+49 351 260Email
Dr. Sören Kliem250/1082318


NameBld./Office+49 351 260Email
Dr. Silvio Baier250/1113034
Dr. Yurii Bilodid250/2092020
Dr. Emil Fridman250/2092167
Matthias Jobst250/1143572
Jörg Konheiser250/1092416
Kerstin Kurde250/1073025
Dr. Evgeny Nikitin250/2072906
Dr. Erik Pönitz250/1152067
Dr. Reuven Rachamin250/1103291
Dr. Frank Schäfer250/1172069


Dr. Sören Kliem

Reactor Safety
Phone: +49 351 260 2318