Pristupačnost
Analysis of the upflow conversion modification and influence on selected LOCA accidents in a PWR plant
Room Classification Based on EMC Conditions in Nuclear Power Plants
Contribution of FER to the WP4: summary report for the first year, TA2/SARNET ASCOM Technical report
Reactor Vessel Modelling with the MELCOR Code
Comparison of Measured and Calculated Data for NPP Krško CILR Test
Overview of techno-economic issues of enhanced geothermal systems implementation and integration
Economic and environmental assessment for enhanced geothermal systems integration into energy systems Decision-making support tool for optimal usage of geothermal energy
Comparison of Measured and Calculated Data for NPP Krško CILR Test
Comparison Between ORIGEN2.2 and ORIGEN-S Calculated Source Term
NPP Krsko IB Modelling with GOTHIC and MELCOR Codes
Proceedings of the 12th International Conference of the Croatian Nuclear Society
Effectiveness of SFP Spray Cooling during Loss of Coolant Accidents
Influence of Detailed PCFV Model on Containment Behaviour During SBO
Fuel Handling Building Response to SFP Loss of Cooling
NPP Krško 3 inch Cold Leg Break LOCA Calculation using RELAP5/MOD 3.3 and MELCOR 1.8.6 Codes
Calculation of NEK CILR Test Using GOTHIC Code
Mathematical model of the NPP Krško PCFV system for the RELAP5 computer code
Application of ASTEC, MELCOR, and MAAP Computer Codes for Thermal Hydraulic Analysis of a PWR Containment Equipped with the PCFV and PAR Systems
Operation and Performance Analysis of Steam Generators in Nuclear Power Plants
Verification of GOTHIC Multivolume Containment Model during NPP Krško DBA LOCA
NPP Krško Station Blackout Analysis after Safety Upgrade Using MELCOR Code
NPP Krško DVI LOCA Calculation using RELAP5/mod 3.3 and FRAPTRAN to Assess UFC Modification Influence
Independent Review of NPP Modifications and Safety Upgrades
Spatial Distribution of Hydrogen in NEK Containment
NPP Krško Containment Modelling and Calculation with the ASTEC Code
Decay Heat Calculation for Spent Fuel Pool Application
Optimization of OPDT Protection for Overcooling Accidents
Calculation of Hydrogen Concentration in Containment during LOCA Accident
NPP Krško Containment Modelling with the ASTEC Code
Hot Leg Streaming Calculation for Two-Loop PWR Plant
Severe Accident Analysis in a Two-Loop PWR Nuclear Power Plant with the ASTEC Code
Severe Accident Analysis in the NPP Krško with the ASTEC Code
ASTEC Computer Code Application to NPP Krško
Coupled code calculation of rod withdrawal at power accident
Radiation heat transfer in a pressurized water reactor lower head filled with molten corium
Coupled Code Calculation of Rod Withdrawal at Power Accident
RELAP5 Modeling of PWR Reactor RTD Bypass
Corium Behaviour and the Lower Head Thermal Response after a Core Meltdown
Comparison of R5G Coupled Code and Classical “Two-steps” Containment Calculation
Calculational Model for In-Vessel Phase of Severe Accidents in Nuclear Power Plant
Influence of Modelling Options in RELAP5/SCDAPSIM and MAAP4 Computer Codes on Core Melt Progression and Reactor Pressure Vessel Integrity
Influence of Modelling Options in RELAP5/SCDAPSIM and MAAP4 Computer Codes on Core Melt Progression and Reactor Pressure Vessel Integrity
QUENCH-11 Experiment Analysis with RELAP5/SCDAPSIM Code
Analysis of In-Vessel Severe Accident Phenomena in NPP Krško
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
ISP-46 Analysis with RELAP/SCDAPSIM Computer Code
QUENCH-06 RELAP5/SCDAPSIM Nodalization Notebook
RELAP5/SCDAPSIM Analysis of the QUENCH-06 Experiment
Applicability Of Relap5/Scdapsim Code For Thermal-Hydraulic Analysis of Pakistan Research Reactor 1
Analysis of International Standard Problem ISP-45 on QUENCH Facility Using RELAP5/SCDAPSIM/MOD3.2 Computer Code
Teaching duties
University undergraduate
- Electric Power Engineering (Lecturer in charge)
- Enviromental Sustainability and Climate Change Mitigation (Lecturer in charge)
- BSc Thesis (Lectures)
- BSc Thesis (Lectures)
- Project (Lectures)
University graduate
- Energy Conversion (Lecturer in charge)
- Enviromental Sustainability and Climate Change Mitigation (Lecturer in charge)
- Nuclear Engineering (Lecturer in charge)
- Nuclear Safety (Lecturer in charge)
- Reliability and Availability Assessment Methods (Lecturer in charge)
- Risk Assessment (Lecturer in charge)
- Graduation Thesis (Lectures)
- Laboratory of Electrical Power Engineering 2 (Lecturers)
- Project (Lectures)
- Mass and Heat Transfer (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