Pristupačnost
The accident of reduced heat removal from the core of a nuclear reactor
Design and regulation of wind turbines
TRACE/RELAP5 Calculation of NPP Krško SGTR Accident Under Realistic and SRP Conditions
Small break LOCA studies for different layouts of passive safety systems in the IRIS reactor
Point-Kernel Code Development for Gamma-Ray Shielding Applications
The accident of increased heat removal from the core of a nuclear reactor
Development of the Numerical Model of the IRIS Reactor for Severe Accident Analysis
Application of the finite element method in the thermal calculation of the photovoltaic panel
Dose Rate Assessment Around the PCFV Release Line During Severe Accident Conditions in Nuclear Power Plant Krsko
Gamma Dose Rate Calculations for RCC4 and RCC8 Containers using Monte Carlo Codes
Uncertainty and Sensitivity Evaluation of the QUENCH-02 Experiment Simulation Using the ASYST Code
Calculation of Unmitigated Small Break Loss of Coolant Accident for the IRIS Reactor
NPP Krško Steam Generator Tube Rupture (SGTR) Accident Analysis using MELCOR Code
Calculation and Uncertainty Quantification of the QUENCH-03 Experiment Simulation Using the ASYST Code
Radiation dose rate analysis of conceptual solution for the Croatian low- and intermediate-level radioactive waste
The Assessment of Dose Rates during MPC Loading and Drying in frame of the Nuclear Power Plant Krško First SFDS Loading Campaign
Analysis of Steam Generator Tube Rupture (SGTR) Accident using RELAP5/MOD 3.3 and TRACE 5.0p5 Codes
Dose Rate Assessment Around The PCFV Release Line During Severe Accident Conditions in Nuclear Power Plant Krsko
Uncertainty assessment of a severe accident in a light-water reactor nuclear power plant
Uncertainty quantification of the QUENCH-02 experiment simulation results with the ASYST code
Graduate Geothermal training in the European Economic Area
Calculation of the QUENCH-02 Experiment with the ASYST Code Including the Uncertainty Evaluation
NEK 3 inch Cold Leg Break LOCA Calculation using TRACE 5.0p5 and RELAP5/MOD 3.3 Codes
Uncertainty Study of the In-Vessel Phase of a Severe Accident in a Pressurized Water Reactor
Proceedings of the 13th International Conference of the Croatian Nuclear Society
Thermal Response of External Duct Piping in Case PCFV Actuation During Prolonged NEK SBO
Influence of reactivity feedback modelling in RELAP5 NEK LONF ATWS calculation
Spent Fuel Dry Storage Loading Plan Based on Uniform Decay Heat Distribution among Casks
Preliminary Uncertainty Assessment of a Severe Accident Scenario Using the ASYST Code
Application of the RELAP5 code for analysis of the consequences of failures in a nuclear power plant
Methodology to calculate radiological impact for NPP Krško life time extension environmental impact assessment
NPP Krško Large Break Loss of Coolant Accident using MELCOR Code
Thermal calculation of the lower plenum of a reactor vessel
Room Classification Based on EMC Conditions in Nuclear Power Plants
Calculation of thermo-hydraulic conditions inside containment of a nuclear power plant
Analysis of the upflow conversion modification and influence on selected LOCA accidents in a PWR plant
Modelling of fuel in the nuclear reactor using the ANSYS code
Contribution of FER to the WP4: summary report for the first year, TA2/SARNET ASCOM Technical report
Comparison of Measured and Calculated Data for NPP Krško CILR Test
Reactor Vessel Modelling with the MELCOR Code
Economic and environmental assessment for enhanced geothermal systems integration into energy systems Decision-making support tool for optimal usage of geothermal energy
NPP Krško 3 inch Cold Leg Break LOCA Calculation using RELAP5/MOD 3.3 and MELCOR 1.8.6 Codes
Fuel Handling Building Response to SFP Loss of Cooling
Overview of techno-economic issues of enhanced geothermal systems implementation and integration
Proceedings of the 12th International Conference of the Croatian Nuclear Society
Effectiveness of SFP Spray Cooling during Loss of Coolant Accidents
Comparison of Measured and Calculated Data for NPP Krško CILR Test
Comparison Between ORIGEN2.2 and ORIGEN-S Calculated Source Term
Influence of Detailed PCFV Model on Containment Behaviour During SBO
NPP Krsko IB Modelling with GOTHIC and MELCOR Codes
Operation and Performance Analysis of Steam Generators in Nuclear Power Plants
Application of ASTEC, MELCOR, and MAAP Computer Codes for Thermal Hydraulic Analysis of a PWR Containment Equipped with the PCFV and PAR Systems
Calculation of NEK CILR Test Using GOTHIC Code
Mathematical model of the NPP Krško PCFV system for the RELAP5 computer code
NPP Krško Station Blackout Analysis after Safety Upgrade Using MELCOR Code
Verification of GOTHIC Multivolume Containment Model during NPP Krško DBA LOCA
Independent Review of NPP Modifications and Safety Upgrades
NPP Krško DVI LOCA Calculation using RELAP5/mod 3.3 and FRAPTRAN to Assess UFC Modification Influence
NPP Krško Containment Modelling and Calculation with the ASTEC Code
Hot Leg Streaming Calculation for Two-Loop PWR Plant
Optimization of OPDT Protection for Overcooling Accidents
Spatial Distribution of Hydrogen in NEK Containment
Decay Heat Calculation for Spent Fuel Pool Application
Calculation of Hydrogen Concentration in Containment during LOCA Accident
NPP Krško Containment Modelling with the ASTEC Code
ASTEC Computer Code Application to NPP Krško
Severe Accident Analysis in the NPP Krško with the ASTEC Code
Radiation heat transfer in a pressurized water reactor lower head filled with molten corium
Severe Accident Analysis in a Two-Loop PWR Nuclear Power Plant with the ASTEC Code
Coupled code calculation of rod withdrawal at power accident
Coupled Code Calculation of Rod Withdrawal at Power Accident
RELAP5 Modeling of PWR Reactor RTD Bypass
Model izračuna reaktorske faze teških nesreća u nuklearnoj elektrani
Influence of Modelling Options in RELAP5/SCDAPSIM and MAAP4 Computer Codes on Core Melt Progression and Reactor Pressure Vessel Integrity
Corium Behaviour and the Lower Head Thermal Response after a Core Meltdown
Comparison of R5G Coupled Code and Classical “Two-steps” Containment Calculation
Influence of Modelling Options in RELAP5/SCDAPSIM and MAAP4 Computer Codes on Core Melt Progression and Reactor Pressure Vessel Integrity
Analysis of In-Vessel Severe Accident Phenomena in NPP Krško
QUENCH-11 Experiment Analysis with RELAP5/SCDAPSIM Code
Benchmark Exercise on QUENCH-11 Experiment
Analysis of Small Break LOCA During Mode 3 and Mode 4 Operation for NPP Krško
Evaluation of Accident Sequences in IRIS Relevant for Revising the Need for Relocation and Evacuation Measures Unique to NPPs for Innovative SMRs
Core Degradation and Hydrogen Production During a Severe Accident
RCP Seal Leakage Analysis During Station Blackout for NPP Krško
Analysis of Phebus FPT1 Experiment with RELAP5/SCDAPSIM Computer Code
Evaluation of the Safety Margins during Shutdown for NPP Krško
Analysis of SBO Accident in NPP Krško using RELAP5/SCDAPSIM/MOD3.2
QUENCH-06 RELAP5/SCDAPSIM Nodalization Notebook
RELAP5/SCDAPSIM Analysis of the QUENCH-06 Experiment
ISP-46 Analysis with RELAP/SCDAPSIM Computer Code
Analysis of International Standard Problem ISP-45 on QUENCH Facility Using RELAP5/SCDAPSIM/MOD3.2 Computer Code
Applicability Of Relap5/Scdapsim Code For Thermal-Hydraulic Analysis of Pakistan Research Reactor 1
Teaching
University undergraduate
- Electric Power Engineering (Lecturer in charge)
- Electric Power Engineering (Lecturer in charge)
- Environmental Sustainability and Climate Change Mitigation (Lecturer in charge, Lecturer in charge)
- Environmental Sustainability and Climate Change Mitigation (Lecturer in charge, Lecturer in charge)
- Project E (Lecturers)
- Project E (Lecturers)
University graduate
- Computational Fluid Dynamics (Lecturer in charge)
- Computational Heat Transfer (Lecturer in charge)
- Fluid Dynamics and Heat Transfer (Lecturer in charge)
- Fluid Dynamics and Heat Transfer (Lecturer in charge)
- Nuclear Engineering (Lecturer in charge)
- Nuclear Safety (Lecturer in charge)
- Nuclear Safety (Lecturer in charge)
- Renewable Energy and Energy Storage (Lecturer in charge)
- Renewable Energy and Energy Storage (Lecturer in charge)
- Risk Assessment (Lecturer in charge)
- Risk Assessment (Lecturer in charge)
- Graduation Thesis (Lecturers)
- Master Project (Lecturers)
- Mentorship Seminar (Lecturers)
- Project (Lecturers)
- Laboratory of Electrical Power Engineering 2 (Laboratory exercises)
Postgraduate doctoral study programme
- Availability Evaluation of Electric Power System Subsystems (Lecturer in charge)
- Probabilistic Assessment of Technological Risk (Lecturer in charge)
Competences
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Engineering – general
Thermal engineering -
Reliability
Availability Reliability theory -
Science – general
Thermodynamics -
Control systems
Heating systems Thermal analysis -
Industry applications
Industrial accidents Heat engines Steam engines Renewable energy sources Sustainable development Radiation safety -
Materials, elements, and compounds
Fluid dynamics Radioactive materials Nuclear fuels Radioactive decay -
Power engineering and energy
Energy consumption Energy conversion Photovoltaic cells Energy resources Nuclear fuels Solar energy Geothermal power generation Nuclear power generation Solar power generation Power system reliability