Junior Research Group Reactor Physics
- Emil Fridman
- Yurii Bilodid
- Daniela Baldova
- Reuven Rachamin
We contribute to the development of a code system to be used for the simulation of steady-state and transient behavior of innovative Generation-IV (GEN-IV) nuclear reactors including High Temperature Gas cooled Reactors (HTGRs), Sodium cooled Fast Reactors (SFRs), Lead cooled Fast Reactors (LFR) and others. The code system is based on the reactor dynamics code DYN3D developed at the Department of Reactor Safety for the modeling of operating Light Water Reactors (LWRs).
GEN-IV reactors differ considerably from the conventional LWRs with respect to neutronic, thermal-hydraulic, and thermal-mechanical properties. Consequently, the LWR oriented DYN3D code should undergo significant methodological modifications before it will be applicable for the analysis of GEN-IV systems. In addition, the procedure for the generation of nuclear data (or so-called cross sections) to be employed by DYN3D has to be established.
- R. Rachamin, C. Wemple, E. Fridman, “Neutronic analysis of SFR core with HELIOS-2, Serpent, and DYN3D codes,” Annals of Nuclear Energy, Vol. 55, Pages 194-204, 2013.
- E. Fridman, E. Shwageraus, “Modeling of SFR cores with Serpent–DYN3D codes sequence,” Annals of Nuclear Energy, Vol. 53, Pages 354-363, 2013.
- U. Rohde, S. Baier, S. Duerigen, E. Fridman, S. Kliem, B. Merk, “Development and verification of the coupled 3D neutron kinetics/thermal-hydraulics code DYN3D-HTR for the simulation of transients in block-type HTGR,” Nuclear Engineering and Design, Vol. 251, Pages 412-422, 2012.
- E. Fridman, J. Leppänen, “On the use of the Serpent Monte Carlo code for few-group cross section generation,” Annals of Nuclear Energy, Vol. 38, , Pages 1399-1405, 2011.
Our group investigates the feasibility of using thorium based nuclear fuels in existing Pressurized Water Reactors (PWRs) while focusing on the two fuel cycle options. The first is the recycling of excess plutonium using Th-Pu mixed oxide fuel which is considered as an attractive alternative to the currently used U-Pu mixed oxide fuel. The second is the improving utilization of natural resources and the reduction in natural uranium demand through the use of the high conversion Th-U233 mixed oxide fuel.
- D. Baldova, E. Fridman, “High Conversion Th-U233 fuel assembly for current generation of PWRs,” Proc. PHYSOR 2012 - Advances in Reactor Physics, Knoxville, Tennessee, USA, 15.-20.04.2012.
- E. Fridman, S. Kliem, “Pu recycling in a full Th-MOX PWR core. Part I: Steady state analysis,” Nuclear Engineering and Design, Vol. 241, Pages 193-202, 2011.
- D. Volaski, E. Fridman, E. Shwageraus, “Thermal design feasibility of Th-233U PWR breeder,” Proc. GLOBAL 2009, Paris, France, 06.-11.09.2009.
- D. Volaski, E. Shwageraus, E. Fridman, “Investigation of fuel assembly design options for high conversion thorium fuel cycle in PWRs,” Proc. Advances in Nuclear Fuel Management IV, Hilton Head Island, South Carolina, USA, 12.-15.04.2009.