Publications de Konstantin Mikityuk

Simulation of flow and heat transfer in a differentially heated cubical cavity using coarse Large Eddy Simulation

Two widely used sub-grid scale models namely: the standard and the dynamic Smagorinsky models were tested in a simulation of the flow in a thermally driven 3D cavity at turbulent Rayleigh number Ra = 10(9). The cubical cavity of 0.7 m side-length is set to ...

2023

Interface Tracking Investigation of the Sliding Bubbles Effects on Heat Transfer in the Laminar Regime

Konstantin Mikityuk, Lubomír Bures

Helium gases are utilized to remove fission products from the molten salt fast reactor (MSFR) core during operation. Helium gases and other volatile fission products may be introduced into the intermediate heat exchanger channels. The effect of these gases ...

2022

Optimization of multi-group energy structures for diffusion analyses of sodium-cooled fast reactors assisted by simulated annealing-Part II: Methodology application

Konstantin Mikityuk, Vincenzo Anthony Di Nora

Part I of this study introduced a novel methodology for the optimization of energy group structures to be used in diffusion calculations of Sodium cooled Fast Reactors. The goal of the optimization is to speed up calculations, particularly in transient ana ...

2021

Optimization of multi-group energy structures for diffusion analyses of sodium-cooled fast reactors assisted by simulated annealing - Part I: Methodology demonstration

Konstantin Mikityuk, Vincenzo Anthony Di Nora

This study presents an approach to the selection of optimal energy group structures for multi-group nodal diffusion analyses of Sodium-cooled Fast Reactor cores. The goal is to speed up calculations, particularly in transient calculations, while maintainin ...

2021

A coarse-mesh methodology for modelling of single-phase thermal-hydraulics of ESFR innovative assembly design

Andreas Pautz, Konstantin Mikityuk, Carlo Fiorina, Stefan Radman

The GeN-Foam multi-physics code enables coupled thermal-hydraulic, neutronic, and thermal-mechanical analysis of fast reactor cores with 3-D unstructured deformable meshes. Advances within the Horizon2020 ESFR-SMART project, as well as trends towards highe ...

ELSEVIER SCIENCE SA2019

Analytical solutions to a coupled fluid dynamics and neutron transport problem with application to GeN-Foam verification

Konstantin Mikityuk

Code verification is an important part of software development process that asserts code correctness. In this work, analytical solutions are used in order to develop a simple multiphysics problem involving coupled fluid dynamics and neutron transport to be ...

2018

A Simplified Numerical Benchmark For Pool-Type Sodium Fast Reactors

Andreas Pautz, Konstantin Mikityuk, Carlo Fiorina, Stefan Radman

The current work presents a simplified benchmark for a pool-type Sodium Fast Reactor based on a 2D (r-z) geometry for testing tightly-coupled spatial neutron transport, thermal-hydraulics and thermal-mechanics modeling. The new benchmark is motivated by de ...

AMER SOC MECHANICAL ENGINEERS2018

Creation Of An Openfoam Fuel Performance Class Based On Fred And Integration Into The Gen-Foam Multi-Physics Code

Andreas Pautz, Konstantin Mikityuk, Carlo Fiorina

The FRED code is an in-house tool developed at the Paul Scherrer Institut for the so-called 1.5-D nuclear fuel performance analysis. In order to extend its field of application, this code has been re-implemented as a class of the OpenFOAM numerical library ...

AMER SOC MECHANICAL ENGINEERS2018

Coupled 3-D Neutronics/Thermal-Hydraulics Optimization Study For Improving The Response Of A 3600 Mw(Thermal) Sfr Core To An Unprotected Loss-Of-Flow Accident

Rakesh Chawla, Aurélia Chenu, Konstantin Mikityuk, Jiri Krepel, Kaichao Sun

The sodium-cooled fast reactor (SFR), as a fast neutron spectrum system, is characterized by several performance advantages. In particular, the long-term operation of an SFR core in a closed fuel cycle will lead to an equilibrium state, where both reactivi ...

Amer Nuclear Soc2013

Analysis of selected Phenix EOL tests with the FAST code system - Part II: Unprotected phase of the Natural Convection Test

Rakesh Chawla, Aurélia Chenu, Konstantin Mikityuk

Experimental results from the unprotected phase of the Natural Convection (NC) Test performed in the Phenix reactor in the frame of end-of-life (EOL) experiments have been used to qualify the FAST code system for Sodium-cooled Fast Reactor (SFR) transient ...

Elsevier2012