Pristupačnost
Verification of the IRIS Numerical Model for the Severe Accident Calculation
Application of the ASYST Code in Estimating Uncertainty of the QUENCH-03 Experiment Calculation
Tridimensional core depletion calculation for NPP Krsko cycle 34
Small break LOCA studies for different layouts of passive safety systems in the IRIS reactor
Quantification of liquid radioactive effluents environmental impact
TRACE/RELAP5 Calculation of NPP Krško SGTR Accident Under Realistic and SRP Conditions
NPP Krško Steam Generator Tube Rupture (SGTR) Accident Analysis using MELCOR Code
Development of the Numerical Model of the IRIS Reactor for Severe Accident Analysis
Calculation and Uncertainty Quantification of the QUENCH-03 Experiment Simulation Using the ASYST Code
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
Radiation dose rate analysis of conceptual solution for the Croatian low- and intermediate-level radioactive waste
Natural Heat Transfer Performance of Helium Filled Spent Fuel Multi Purpose Container Calculated for Different Initial Pressures
Benchmark Calculation of FHR Fuel Assembly Phase I-C Depletion Exercises
Gamma Dose Rate Calculations for RCC4 and RCC8 Containers using Monte Carlo Codes
Calculation of Unmitigated Small Break Loss of Coolant Accident for the IRIS Reactor
Assessment of potential impact of combustible gases from reactor core damage on risk of outside containment spent fuel pool damage
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
Usage of Monte Carlo Code Serpent2 for Calculation of FHR Fuel Assembly
Uncertainty Study of the In-Vessel Phase of a Severe Accident in a Pressurized Water Reactor
Estimation of Dose Rates Around Dry Storage Building During Campaign One Loading in Nuclear Power Plant Krsko
Radiation shielding analysis of the HI-STORM FW storage cask
Assessment of extending operation of Nuclear Power Plant Krško from 2023 to 2043 – techno economic, ecological and power flow and system dynamics influence
Calculation of the QUENCH-02 Experiment with the ASYST Code Including the Uncertainty Evaluation
Dose Rate Assessment Around The PCFV Release Line During Severe Accident Conditions in Nuclear Power Plant Krsko
Influence of reactivity feedback modelling in RELAP5 NEK LONF ATWS calculation
Thermal Response of External Duct Piping in Case PCFV Actuation During Prolonged NEK SBO
NEK 3 inch Cold Leg Break LOCA Calculation using TRACE 5.0p5 and RELAP5/MOD 3.3 Codes
Thermal Model of HI STORM FW Cask for COBRA-SFS Code
Methodology to calculate radiological impact for NPP Krško life time extension environmental impact assessment
Analysis of the HI-TRAC VW Transfer Cask Dose Rates During Nuclear Power Plant Krsko Spent Fuel Dry Storage Campaign One
NPP Krško Large Break Loss of Coolant Accident using MELCOR Code
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
Analysis of the HI-TRAC VW Transfer Cask Dose Rates During Nuclear Power Plant Krsko Spent Fuel Dry Storage Campaign One
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
ENERGY TRANSITION TOWARDS ENERGY TRANSFORMATION
Evaluation of the NEK SFDS Cask Model Using Hybrid Shielding Methodology
Contribution of FER to the WP4: summary report for the first year, TA2/SARNET ASCOM Technical report
Influence of Spacer Grids Homogenization on Core Reactivity and Axial Power Distribution
Reactor Vessel Modelling with the MELCOR Code
Comparison of Measured and Calculated Data for NPP Krško CILR Test
Dose Calculation for Emergency Control Room HVAC Filter
Hazard Assessment of NPP Krško for Republic of Croatia
Point Kernel Modification Including Support Vector Regression Neutron Buildup Factor Models
Application of Fractional Scaling Analysis for development and design of Integral Effects Test facility
A new heuristics for the event ordering in binary decision diagram applied in fault tree analysis
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
Influence of Detailed PCFV Model on Containment Behaviour During SBO
Influence of Spacer Grids Homogenization on Core Reactivity and Axial Power Distribution
Hybrid Shielding Methods Validation Using Graphite Shielding Measurements
Point Kernel Modification Including Support Vector Regression Neutron Buildup Factor Models
Comparison of Measured and Calculated Data for NPP Krško CILR Test
Dose Calculation for Emergency Control Room HVAC Filter
Effectiveness of SFP Spray Cooling during Loss of Coolant Accidents
Comparison Between ORIGEN2.2 and ORIGEN-S Calculated Source Term
Calculation of Radioactive Inventory for Severe Accident and Consequence Codes
NPP Krsko IB Modelling with GOTHIC and MELCOR Codes
Radioactive Inventory Data for Severe Accident and Consequence Calculation Codes
Application of ASTEC, MELCOR, and MAAP Computer Codes for Thermal Hydraulic Analysis of a PWR Containment Equipped with the PCFV and PAR Systems
I2S-LWR Concept Update
Calculation of NEK CILR Test Using GOTHIC Code
Analysis of Spent Fuel Pool Loss of Coolant Inventory Accident Progression
Operation and Performance Analysis of Steam Generators in Nuclear Power Plants
Nuclear and thermal hydraulic calculation of a representative I2S-LWR first core
Nuclear and thermal hydraulic calculation of a representative I2S-LWR first core
Neutron Buildup Factors Calculation for Support Vector Regression Application in Shielding Analysis
Proceedings of the 11th International Conference of the Croatian Nuclear Society
Spent Fuel Pool Dose Rate Calculations Using Point Kernel and Hybrid Deterministic-Stochastic Shielding Methods
Spent Fuel Pool Dose Rate Calculations Using Point Kernel and Hybrid Deterministic-Stochastic Shielding Methods
Preparation of Radioactive Inventory Data for Severe Accident and Consequence Calculation Codes
NPP Krško Post-UFC Transient Response during MSLB
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
Severe accident gamma dose distribution through NPP Krško containment and auxiliary building calculated using SCALE6/MAVRIC sequence
Environmental Consequences of Radioactive Gas Effluent Release from Nuclear Power Plants
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
Fuel Depletion Modeling of Reconstituted NEK Fuel Assembly Using Lattice Cell Programs
Radiation Doses Estimation For Hypothetical NPP Krško Accidents Without And With PCFV Using RASCAL Software
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
Spatial Distribution of Hydrogen in NEK Containment
Development of NPP Krško Primary Loop Model for TRACE code
Optimization of OPDT Protection for Overcooling Accidents
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
Severe Accident Analysis in a Two-Loop PWR Nuclear Power Plant with the ASTEC Code
Uncertainty Evaluation of the Rod Withdrawal at Power Accident Analysis including 3D Neutron Kinetics
Physical conditioning programme of the top junior tennis player during preparation period
Lattice Codes Pin Power Prediction Comparison
Xenon Correction in Homogenized Neutron Cross Sections
Analysis of Steam Generator Tube Rupture (SGTR) Accident for NPP Krško
Prediction of Local Hydrogen Concentrations in PWR Containment Using GOTHIC Code
SPES3 Facility RELAP5 Sensitivity Analyses on the Containment System for Design Review
Analysis of Rod Withdrawal at Power (RWAP) Accident using ATHLET Mod 2.2 Cycle A and RELAP5/mod 3.3 Codes
Liquid-salt-cooled Reactor Start-up with Natural Circulation
Coupled Code Calculation of Rod Withdrawal at Power Accident
Testing of an Integral Layout SMR on the SPES3 Facility: BDBE Simulation and Influence of the PCC Actuation Delay on the Accident Recovery
Analysis of Different Containment Models for IRIS Small Break LOCA using GOTHIC and RELAP5 codes
Precursor Based PTS Screening Methodology of the EOP Operator Actions for PWR Plant
Corium Behaviour and the Lower Head Thermal Response after a Core Meltdown
Spectral Codes Pin Power Prediction Comparison
Condensation within Small Compartments during Design Basis Accidents
SPES3 Facility and IRIS Reactor Numerical Simulations For The SPES3 Final Design
Comparison of R5G Coupled Code and Classical “Two-steps” Containment Calculation
RELAP5 Modeling of PWR Reactor RTD Bypass
ORIGEN2.1 Cycle Specific Calculation of Krško Nuclear Power Plant Decay Heat and Core Inventory
Upgrade of the FUMACS 2005 Code Package
Prediction Capabilities of Spectral Codes DRAGON, FA2D, NEWT
The SPES3 Experimental Facility Design for the IRIS Reactor Simulation
Upgrade of FUMACS Code Package for Modeling of NGF and Gadolinium, Final Report
Analysis of Different Containment Models for IRIS Small Break LOCA, using GOTHIC and RELAP5 Codes
Analysis of OECD/CSNI ISP-42 Phase A PANDA Experiment Using RELAP5/mod3.3 and GOTHIC 7.2a Codes
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Sensitivity Studies of Fuel Pin Temperature for PWR Fuel Assemblies Containing Burnable Absorbers
Influence of Thermal-Hydraulic Model to Fuel Management Core Calculation
Modelling of Condensation in Vertical Tubes for Passive Safety Systems
Verification of FA2D Prediction Capability Using Fuel Assembly Benchmark
Analysis of In-Vessel Severe Accident Phenomena in NPP Krško
Influence of NPP Krško Core Nuclear Characteristics on RCCA Ejection Accident
Potential advantages and disadvantages of sequential building small nuclear units instead of a large nuclear power plant
NPP Krško Core Calculations to Improve Operational Safety
High voltage overhead lines and application of the ordinance for protection from electromagnetic fields
RAZVOJ MODELA ZA PRORAČUN HOMOGENIZIRANIH NEUTRONSKIH KONSTANTI REFLEKTORA NUKLEARNE ELEKTRANE PWR TIPA
Sensitivity analysis of control assembly ejection accident on NPP Krško core nuclear characteristics
Comparison of economy and emissions of nuclear power plants and combined gas and wind electricty generators
Evaluation of Accident Sequences in IRIS Relevant for Revising the Need for Relocation and Evacuation Measures Unique to NPPs for Innovative SMRs
IRIS – Progress in Licensing and Toward Deployment
Hydrogen Behaviour in NEK Containment Calculated Using GOTHIC Lumped and Distributed Volume Options
3D Simulation of a High Pressure Water Injection Into a Circular Cold Leg Pipe Flow
Main Steam Line Break – Hot Full Power Analysis for NPP Krško Using RELAP5/PARCS Coupled Code
Possible Role of Advanced Nuclear Reactor IRIS in Croatian Power System
Preparation of NPP Krško input data for program TRACE
Anticipated Transient Without Scram Analysis for NPP Krško using Point Kinetics Option and Coupled Code
IRIS - Advanced Integral Nuclear Reactor with Modular Construction
IRIS - Advanced Medium Power Nuclear Reactor with Improved Safety and Economy
IRIS - advanced medium power nuclear reactor with advanced security and econimics
IRIS (International Reactor Innovative and Secure) - Design Overview and Deployment Prospects
3D Simulation of a High Pressure Water Injection Into a Circular Cold Leg Pipe Flow
IRIS – ADVANCED MEDIUM POWER NUCLEAR REACTOR WITH IMPROVED SAFETY AND ECONOMY
Analysis of future nuclear power plants competitive investment costs with stochastic methods
Applicability of Coupled Code RELAP5/GOTHIC to NPP Krško MSLB Calculation
IRIS Design Overview and Status Update
Modelling of thermal energy losses in industrial facilities with controlled environmental temperature
IRIS Small Break LOCA Phenomena Identification and Ranking Table (PIRT) Addendum 1 - IRIS Small Break LOCA Sensitivity Report for PIRT Development
The design and safety features of the IRIS reactor
Radiation Dose Specification for Equipment Qulaification
Modeling of CHASHMA NPP Reactor Core Using FUMACS Computer Code Package
Coupled RELAP5/GOTHIC Model for IRIS SBLOCA Analysis
Small Break Loss of Cooloant Accident Analysis for the International Reactor Innovative and Secure (IRIS)
Modelling of Chashma NPP Reactor Core Using FUMACS Computer Code Package
Use of the Deterministic Safety Analyses in Support to the NPP Krško Modifications
Three-Batch Reloading Scheme for IRIS Reactor Extended Cycles
Proračun lokalnih termohidrauličkih uvjeta za potrebe kvalificiranja električne opreme sigurnosne klase
Coupled RELAP5/GOTHIC Model for IRIS Reactor SBLOCA Analysis
Overview of Computational Challenges in the Development of Evaluation Models for the IRIS Reactor
Development of RELAP5 Nodalization for IRIS Non-LOCA Transient Analyses
The Design and Safety Features of the IRIS Reactor
Three Mile Island Unit 1 Main Steam Line Break Three-Dimensional Neutronics-Thermal-Hydraulics Analysis
IRIS Core Neutronics Modeling
IRIS RELAP 5 mod 3.3 Nodalization and Steady State Qualification
IRIS Core Criticality Calculations
LOCA Analysis of the IRIS Reactor
Analysis of Environmental Conditions for NPP Krško DC Battery and Battery Charger Rooms Following SG Blowdown Processing System Line Break
Design Of A Four-Year Straight-Burn Core For The Generation-IV Iris Reactor
Design Of A Four-Year Straight-Burn Core For The Generation-IV Iris Reactor
Core Design Calculations of IRIS Reactor Using Modified CORD-2 Code Package
Upgrade Of The Fumacs Code Package
Calculation of Core Design Benchmark 44 for IRIS Reactor Using Modified CORD-2 Code Package
Design of a Four-Year Straight-Burn Core for the Generation IV IRIS Reactor
Evaluation Of A Multisource Option Introduced Into Qad-Cggp Code
Upgrade of the FUMACS Code Package
Verifications of electrical and mechanical properties for compact transmission line tower head design
Status of CAMP Activities in Croatia
Method for Analysis of Asymmetric Accidents in Nuclear Power Plants
Benchmark Performed by Different Coupled 3-D Neutronics Thermal-Hydraulic Codes
Influence of Reactor Vessel Nodalizations in the Coupled Code Analysis of Asymmetric Main Feedwater Isolation
Evaluation of a Multisource Option Introduced into QAD-CGGP Code
APPLICATION OF THE COUPLED CODE RELAP5-QUABOX/CUBBOX IN BEST-ESTIMATE ANALYSES OF NUCLEAR POWER PLANTS
SLB Analysis of NPP Krško Using Coupled RELAP5/MOD3.2-QUABOX/CUBBOX CODE
TMI-1 MSLB COUPLED 3-D NEUTRONICS/THERMALHYDRAULICS ANALYSIS: APPLICATION OF RELAP5-3D AND COMPARISON WITH DIFFERENT CODES
CORE-WIDE DNBR CALCULATION FOR NPP KRŠKO MSLB
ASSESSMENT OF THE MSLB ACCIDENT SAFETY MARGINS FOR NPP KRŠKO
NPP KRŠKO CONTAINMENT RESPONSE FOLLOWING MAIN STEAM LINE BREAK
ANALYSIS OF INADVERTENT CONTAINMENT SPRAY ACTUATION FOR NPP KRŠKO
VERIFICATION OF THE COUPLED CODE RELAP5-QUABOX/CUBBOX BY LARGE LOAD REJECTION ANALYSIS FOR NPP KRŠKO
Application of the coupled code RELAP5- QUABOX/CUBBOX in Best-Estimate Analyses of Nuclear Power Plants
Thermal-Hydraulic Modeling of Reracked Spent Fuel Pool
DEVELOPMENT OF POWER REACTOR SAFETY ANALYSES COUPLED SYSTEM FOR PERSONAL COMPUTER ENVIRONMENT
NPP KRSKO STEAM LINE BREAK ANALYSIS BY MEANS OF COUPLED CODE RELAP5-QUABOX/CUBBOX
Modernization of the FUMACS code package
THERMAL-HYDRAULIC MODELING OF RERACKED SPENT FUEL POOL
Role of the Exercises in the Training of the Technical Support Center Members in Croatia
Croatian Experience in the Use of Coupled 3D Code With RELAP5
DEVELOPMENT OF POWER REACTOR SAFETY ANALYSES COUPLED SYSTEM FOR PERSONAL COMPUTER ENVIRONMENT
RELAP5/mod3.2 - QUABOX/CUBBOX-HYCA Coupling
Calculation of thermohydraulic parameters of the spent fuel pool of the NPP Krško with the assumption of denser fuel storage and spent fuel burnup
Quality of Education and Education of Quality in Croatia
RELAP5/MOD2 Code for Win95/NT Environment with F90 Capabilities
Nije dostupan
RELAP5/Mod3.2 - Quabox/Cubbox-Hyca Coupled Code
RELAP5/mod2 Code for Win95/NT Environment with F90 Capabilities
Capability of the QUABOX/CUBBOX-ATHLET Coupled Code System
RELAP5/mod2 Analysis of Inadvertent Closing of the Main Steam Isolation Valve in NPP Krško
Use of GOTHIC Code for Assessment of Equipment Environmental Qualification
Concrete Spent Fuel Cask Criticality Calculation
The Development of the Safety Related Valve Class 1E Electric Motor, the Target and the Results
NISA FEM Support in Seismic Qualification of Small Class 1E Electric Motors
Conceptual study of the spent fuel intermediate storage on the NPP site
Transient Temperature Behaviour Of The MO Suger Arrester Aged By Repeated Application Of Current Impulses And Analyzed By Means Of Computer Aided Thermal Modelling
Modelling of Transient Temperature Fields in Structures with System Transfer Function
Evaluation of ecological consequences of closing NP Krško
Investigation of intermediate conversion pressurized water reactor for small or medioum nuclear systems
A study of spent fuel storage at the nuclear power plant site
Analysis, calculation and testing of shielding for radioactive waste deposition
Modelling transient from uncontrolled reduction of boric acid concentration for NPS Krško
Analysis of the transient caused by uncontrolled reduction of boric acid concentration
IAEA Coordinated research programme on requirements for application of advanced reactors, Second annual progress report
Analysis of transients caused by fast reactivity increase
A case study of the effect of advanced fuel cycles and improvements in PWR and PHWR reactors on the resource requirements within small or medium nuclear power programme
IAEA Coordinated research programme on requirements for future application of advanced reactors, Progress Report, September 1985
Analysis of a transient resulting from uncontrolled withdrawal of a group of control rods of PWR reactor at full power
Analysis of transients following uncontrolled retraction of the bank of control rods and ejection of a group of control rods
Teaching
University undergraduate
- Electric Power Engineering (Lecturer in charge)
- Electric Power Engineering (Lecturer in charge)
- 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)
- Laboratory of Electrical Power Engineering 1 (Lecturer in charge)
- Nuclear Engineering (Lecturer in charge)
- Nuclear Safety (Lecturer in charge)
- Nuclear Safety (Lecturer in charge)
- Power Generation (Lecturer in charge)
- Power Generation (Lecturer in charge)
- Graduation Thesis (Lecturers)
- Mentorship Seminar (Lecturers)
- Project (Lecturers)
- Project (Lecturers)
Postgraduate doctoral study programme
- Innovative Nuclear Systems for Sustainable Development (Lecturer in charge)
- Nuclear power plant safety analyses (Lecturer in charge)
Competences
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Nuclear and plasma sciences
Accelerator magnets Radiation effects Biological effects of radiation Gamma-ray effects Ion radiation effects Neutron radiation effects Total ionizing dose Radiation monitoring Radiation dosage Radiation safety Radiation protection Reactor instrumentation -
Industry applications
Electrical accidents Industrial accidents Fuel cells Heat engines Pollution control Renewable energy sources Nuclear facility regulation Conductors Fires Flammability Hazardous materials Health and safety Radiation protection Radiation safety -
Instrumentation and measurement
Electromagnetic modeling Nuclear measurements Radiation monitoring -
Materials, elements, and compounds
Radioactive materials Nuclear fuels Radioactive decay Radioactive waste -
Power engineering and energy
Fuel cells Energy efficiency Geothermal energy Nuclear fuels Solar energy Wind energy Energy storage Geothermal power generation Hydroelectric power generation Nuclear power generation Fission reactors Wind power generation Power system reliability