PristupačnostNa FER-u postoji više zaposlenika s imenom
Dose Rate Assessment Around the PCFV Release Line During Severe Accident Conditions in Nuclear Power Plant Krsko
Uncertainty and Sensitivity Evaluation of the QUENCH-02 Experiment Simulation Using the ASYST Code
Analysis of Steam Generator Tube Rupture (SGTR) Accident using RELAP5/MOD 3.3 and TRACE 5.0p5 Codes
The Assessment of Dose Rates during MPC Loading and Drying in frame of the Nuclear Power Plant Krško First SFDS Loading Campaign
NEK 3 inch Cold Leg Break LOCA Calculation using TRACE 5.0p5 and RELAP5/MOD 3.3 Codes
Influence of reactivity feedback modelling in RELAP5 NEK LONF ATWS calculation
Uncertainty Study of the In-Vessel Phase of a Severe Accident in a Pressurized Water Reactor
Dose Rate Assessment Around The PCFV Release Line During Severe Accident Conditions in Nuclear Power Plant Krsko
Calculation of the QUENCH-02 Experiment with the ASYST Code Including the Uncertainty Evaluation
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
Graduate Geothermal training in the European Economic Area
Preliminary Uncertainty Assessment of a Severe Accident Scenario Using the ASYST Code
Spent Fuel Dry Storage Loading Plan Based on Uniform Decay Heat Distribution among Casks
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
Toplinski izračun donjeg plenuma reaktorske posude
Izračun termohidrauličkih uvjeta u reaktorskoj zgradi nuklearne elektrane
Room Classification Based on EMC Conditions in Nuclear Power Plants
Analysis of the upflow conversion modification and influence on selected LOCA accidents in a PWR plant
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
Modeliranje goriva nuklearnog reaktora u programu ANSYS
NPP Krško 3 inch Cold Leg Break LOCA Calculation using RELAP5/MOD 3.3 and MELCOR 1.8.6 Codes
Overview of techno-economic issues of enhanced geothermal systems implementation and integration
Influence of Detailed PCFV Model on Containment Behaviour During SBO
Comparison of Measured and Calculated Data for NPP Krško CILR Test
Fuel Handling Building Response to SFP Loss of Cooling
NPP Krsko IB Modelling with GOTHIC and MELCOR Codes
Economic and environmental assessment for enhanced geothermal systems integration into energy systems Decision-making support tool for optimal usage of geothermal energy
Comparison Between ORIGEN2.2 and ORIGEN-S Calculated Source Term
Proceedings of the 12th International Conference of the Croatian Nuclear Society
Effectiveness of SFP Spray Cooling during Loss of Coolant Accidents
Calculation of NEK CILR Test Using GOTHIC Code
Operation and Performance Analysis of Steam Generators in Nuclear Power Plants
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
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
Decay Heat Calculation for Spent Fuel Pool Application
NPP Krško Containment Modelling with the ASTEC Code
Optimization of OPDT Protection for Overcooling Accidents
Hot Leg Streaming Calculation for Two-Loop PWR Plant
Spatial Distribution of Hydrogen in NEK Containment
Calculation of Hydrogen Concentration in Containment during LOCA Accident
NPP Krško Containment Modelling and Calculation with the ASTEC Code
Severe Accident Analysis in the NPP Krško with the ASTEC Code
Coupled code calculation of rod withdrawal at power accident
Radiation heat transfer in a pressurized water reactor lower head filled with molten corium
ASTEC Computer Code Application to NPP Krško
Severe Accident Analysis in a Two-Loop PWR Nuclear Power Plant with the ASTEC Code
Coupled Code Calculation of Rod Withdrawal at Power Accident
Corium Behaviour and the Lower Head Thermal Response after a Core Meltdown
Influence of Modelling Options in RELAP5/SCDAPSIM and MAAP4 Computer Codes on Core Melt Progression and Reactor Pressure Vessel Integrity
Comparison of R5G Coupled Code and Classical “Two-steps” Containment Calculation
Model izračuna reaktorske faze teških nesreća u nuklearnoj elektrani
RELAP5 Modeling of PWR Reactor RTD Bypass
QUENCH-11 Experiment Analysis with RELAP5/SCDAPSIM Code
Analysis of In-Vessel Severe Accident Phenomena in NPP Krško
Influence of Modelling Options in RELAP5/SCDAPSIM and MAAP4 Computer Codes on Core Melt Progression and Reactor Pressure Vessel Integrity
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 SBO Accident in NPP Krško using RELAP5/SCDAPSIM/MOD3.2
Evaluation of the Safety Margins during Shutdown for NPP Krško
Analysis of Phebus FPT1 Experiment with RELAP5/SCDAPSIM Computer Code
ISP-46 Analysis with RELAP/SCDAPSIM Computer Code
QUENCH-06 RELAP5/SCDAPSIM Nodalization Notebook
RELAP5/SCDAPSIM Analysis of the QUENCH-06 Experiment
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
Nastava
Sveučilišni preddiplomski
- Elektroenergetika (Nositelj)
- Elektroenergetika (Nositelj)
- Okoliš, održivi razvoj i ublažavanje klimatskih promjena (Nositelj, Nositelj)
- Okoliš, održivi razvoj i ublažavanje klimatskih promjena (Nositelj, Nositelj)
- Projekt (Predavanja)
- Projekt E (Predavanja)
- Završni rad (Predavanja)
- Završni rad (Predavanja)
Sveučilišni diplomski
- Dinamika fluida i prijenos topline (Nositelj)
- Nuklearna sigurnost (Nositelj)
- Nuklearno inženjerstvo (Nositelj)
- Numerički postupci u dinamici fluida (Nositelj)
- Numerički postupci u prijenosu topline (Nositelj)
- Obnovljivi izvori i pohrana energije (Nositelj)
- Obnovljivi izvori i pohrana energije (Nositelj)
- Procjena rizika (Nositelj)
- Procjena rizika (Nositelj)
- Diplomski projekt (Predavanja)
- Diplomski rad (Predavanja)
- Diplomski rad (Predavanja)
- Projekt (Predavanja)
- Seminar 1 (Predavanja)
- Seminar 2 (Predavanja)
- Laboratorij elektroenergetike 2 (Laboratorijske vježbe)
Poslijediplomski doktorski
- Raspoloživost elektroenergetskih podsustava (Nositelj)
- Vjerojatnosna procjena tehnološkog rizika (Nositelj)
Kompetencije
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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