Pristupačnost
Small break LOCA studies for different layouts of passive safety systems in the IRIS reactor
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
Radiation dose rate analysis of conceptual solution for the Croatian low- and intermediate-level radioactive waste
Dose Rate Assessment Around the PCFV Release Line During Severe Accident Conditions in Nuclear Power Plant Krsko
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
Uncertainty and Sensitivity Evaluation of the QUENCH-02 Experiment Simulation Using the ASYST Code
Development of the Numerical Model of the IRIS Reactor for Severe Accident Analysis
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
Analysis of Steam Generator Tube Rupture (SGTR) Accident using RELAP5/MOD 3.3 and TRACE 5.0p5 Codes
Assessment of potential impact of combustible gases from reactor core damage on risk of outside containment spent fuel pool damage
Usage of Monte Carlo Code Serpent2 for Calculation of FHR Fuel Assembly
The Assessment of Dose Rates during MPC Loading and Drying in frame of the Nuclear Power Plant Krško First SFDS Loading Campaign
Thermal Response of External Duct Piping in Case PCFV Actuation During Prolonged NEK SBO
Calculation of the QUENCH-02 Experiment with the ASYST Code Including the Uncertainty Evaluation
Thermal Model of HI STORM FW Cask for COBRA-SFS Code
Estimation of Dose Rates Around Dry Storage Building During Campaign One Loading in Nuclear Power Plant Krsko
Dose Rate Assessment Around The PCFV Release Line During Severe Accident Conditions in Nuclear Power Plant Krsko
Uncertainty Study of the In-Vessel Phase of a Severe Accident in a Pressurized Water Reactor
Influence of reactivity feedback modelling in RELAP5 NEK LONF ATWS calculation
Assessment of extending operation of Nuclear Power Plant Krško from 2023 to 2043 – techno economic, ecological and power flow and system dynamics influence
Radiation shielding analysis of the HI-STORM FW storage cask
NEK 3 inch Cold Leg Break LOCA Calculation using TRACE 5.0p5 and RELAP5/MOD 3.3 Codes
Spent Fuel Dry Storage Loading Plan Based on Uniform Decay Heat Distribution among Casks
NPP Krško Large Break Loss of Coolant Accident using MELCOR Code
Preliminary Uncertainty Assessment of a Severe Accident Scenario Using the ASYST Code
Analysis of the HI-TRAC VW Transfer Cask Dose Rates During Nuclear Power Plant Krsko Spent Fuel Dry Storage Campaign One
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
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
Hazard Assessment of NPP Krško for Republic of Croatia
Dose Calculation for Emergency Control Room HVAC Filter
Application of Fractional Scaling Analysis for development and design of Integral Effects Test facility
Influence of Spacer Grids Homogenization on Core Reactivity and Axial Power Distribution
Point Kernel Modification Including Support Vector Regression Neutron Buildup Factor Models
A new heuristics for the event ordering in binary decision diagram applied in fault tree analysis
ENERGY TRANSITION TOWARDS ENERGY TRANSFORMATION
Evaluation of the NEK SFDS Cask Model Using Hybrid Shielding Methodology
Reactor Vessel Modelling with the MELCOR Code
Comparison of Measured and Calculated Data for NPP Krško CILR Test
NPP Krško 3 inch Cold Leg Break LOCA Calculation using RELAP5/MOD 3.3 and MELCOR 1.8.6 Codes
Hybrid Shielding Methods Validation Using Graphite Shielding Measurements
Calculation of Radioactive Inventory for Severe Accident and Consequence Codes
Fuel Handling Building Response to SFP Loss of Cooling
Point Kernel Modification Including Support Vector Regression Neutron Buildup Factor Models
Comparison Between ORIGEN2.2 and ORIGEN-S Calculated Source Term
Influence of Spacer Grids Homogenization on Core Reactivity and Axial Power Distribution
Influence of Detailed PCFV Model on Containment Behaviour During SBO
Effectiveness of SFP Spray Cooling during Loss of Coolant Accidents
Dose Calculation for Emergency Control Room HVAC Filter
NPP Krsko IB Modelling with GOTHIC and MELCOR Codes
Comparison of Measured and Calculated Data for NPP Krško CILR Test
Operation and Performance Analysis of Steam Generators in Nuclear Power Plants
I2S-LWR Concept Update
Analysis of Spent Fuel Pool Loss of Coolant Inventory Accident Progression
Calculation of NEK CILR Test Using GOTHIC Code
Application of ASTEC, MELCOR, and MAAP Computer Codes for Thermal Hydraulic Analysis of a PWR Containment Equipped with the PCFV and PAR Systems
Radioactive Inventory Data for Severe Accident and Consequence Calculation Codes
Spent Fuel Pool Dose Rate Calculations Using Point Kernel and Hybrid Deterministic-Stochastic Shielding Methods
Verification of GOTHIC Multivolume Containment Model during NPP Krško DBA LOCA
NPP Krško Station Blackout Analysis after Safety Upgrade Using MELCOR Code
Preparation of Radioactive Inventory Data for Severe Accident and Consequence Calculation Codes
Neutron Buildup Factors Calculation for Support Vector Regression Application in Shielding Analysis
NPP Krško Post-UFC Transient Response during MSLB
Nuclear and thermal hydraulic calculation of a representative I2S-LWR first core
Nuclear and thermal hydraulic calculation of a representative I2S-LWR first core
Spent Fuel Pool Dose Rate Calculations Using Point Kernel and Hybrid Deterministic-Stochastic Shielding Methods
Proceedings of the 11th International Conference of the Croatian Nuclear Society
Environmental Consequences of Radioactive Gas Effluent Release from Nuclear Power Plants
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
NPP Krško DVI LOCA Calculation using RELAP5/mod 3.3 and FRAPTRAN to Assess UFC Modification Influence
Radiation Doses Estimation For Hypothetical NPP Krško Accidents Without And With PCFV Using RASCAL Software
Hot Leg Streaming Calculation for Two-Loop PWR Plant
Fuel Depletion Modeling of Reconstituted NEK Fuel Assembly Using Lattice Cell Programs
Spatial Distribution of Hydrogen in NEK Containment
Decay Heat Calculation for Spent Fuel Pool Application
Optimization of OPDT Protection for Overcooling Accidents
Development of NPP Krško Primary Loop Model for TRACE code
Calculation of Hydrogen Concentration in Containment during LOCA Accident
NPP Krško Containment Modelling 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
Severe Accident Analysis in the NPP Krško with the ASTEC Code
SPES3 Facility RELAP5 Sensitivity Analyses on the Containment System for Design Review
Xenon Correction in Homogenized Neutron Cross Sections
Physical conditioning programme of the top junior tennis player during preparation period
Lattice Codes Pin Power Prediction Comparison
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
Analysis of Steam Generator Tube Rupture (SGTR) Accident for NPP Krško
Uncertainty Evaluation of the Rod Withdrawal at Power Accident Analysis including 3D Neutron Kinetics
Prediction of Local Hydrogen Concentrations in PWR Containment Using GOTHIC Code
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
ORIGEN2.1 Cycle Specific Calculation of Krško Nuclear Power Plant Decay Heat and Core Inventory
Corium Behaviour and the Lower Head Thermal Response after a Core Meltdown
Prediction Capabilities of Spectral Codes DRAGON, FA2D, NEWT
Spectral Codes Pin Power Prediction Comparison
Condensation within Small Compartments during Design Basis Accidents
RELAP5 Modeling of PWR Reactor RTD Bypass
SPES3 Facility and IRIS Reactor Numerical Simulations For The SPES3 Final Design
Upgrade of the FUMACS 2005 Code Package
Comparison of R5G Coupled Code and Classical “Two-steps” Containment Calculation
The SPES3 Experimental Facility Design for the IRIS Reactor Simulation
Analysis of OECD/CSNI ISP-42 Phase A PANDA Experiment Using RELAP5/mod3.3 and GOTHIC 7.2a Codes
Analysis of Different Containment Models for IRIS Small Break LOCA, using GOTHIC and RELAP5 Codes
Upgrade of FUMACS Code Package for Modeling of NGF and Gadolinium, Final Report
Sensitivity Studies of Fuel Pin Temperature for PWR Fuel Assemblies Containing Burnable Absorbers
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Influence of NPP Krško Core Nuclear Characteristics on RCCA Ejection Accident
Verification of FA2D Prediction Capability Using Fuel Assembly Benchmark
Analysis of In-Vessel Severe Accident Phenomena in NPP Krško
Modelling of Condensation in Vertical Tubes for Passive Safety Systems
Influence of Thermal-Hydraulic Model to Fuel Management Core Calculation
Potential advantages and disadvantages of sequential building small nuclear units instead of a large nuclear power plant
High voltage overhead lines and application of the ordinance for protection from electromagnetic fields
NPP Krško Core Calculations to Improve Operational Safety
Neutronic model development for calculus of a PWR reflector
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
Anticipated Transient Without Scram Analysis for NPP Krško using Point Kinetics Option and Coupled Code
Hydrogen Behaviour in NEK Containment Calculated Using GOTHIC Lumped and Distributed Volume Options
Preparation of NPP Krško input data for program TRACE
IRIS – Progress in Licensing and Toward Deployment
Evaluation of Accident Sequences in IRIS Relevant for Revising the Need for Relocation and Evacuation Measures Unique to NPPs for Innovative SMRs
3D Simulation of a High Pressure Water Injection Into a Circular Cold Leg Pipe Flow
Possible Role of Advanced Nuclear Reactor IRIS in Croatian Power System
Main Steam Line Break – Hot Full Power Analysis for NPP Krško Using RELAP5/PARCS Coupled Code
IRIS – ADVANCED MEDIUM POWER NUCLEAR REACTOR WITH IMPROVED SAFETY AND ECONOMY
IRIS - Advanced Integral Nuclear Reactor with Modular Construction
Analysis of future nuclear power plants competitive investment costs with stochastic methods
3D Simulation of a High Pressure Water Injection Into a Circular Cold Leg Pipe Flow
IRIS (International Reactor Innovative and Secure) - Design Overview and Deployment Prospects
IRIS - advanced medium power nuclear reactor with advanced security and econimics
IRIS - Advanced Medium Power Nuclear Reactor with Improved Safety and Economy
IRIS Design Overview and Status Update
Applicability of Coupled Code RELAP5/GOTHIC to NPP Krško MSLB Calculation
Coupled RELAP5/GOTHIC Model for IRIS SBLOCA Analysis
Use of the Deterministic Safety Analyses in Support to the NPP Krško Modifications
Small Break Loss of Cooloant Accident Analysis for the International Reactor Innovative and Secure (IRIS)
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
Modeling of CHASHMA NPP Reactor Core Using FUMACS Computer Code Package
Three-Batch Reloading Scheme for IRIS Reactor Extended Cycles
Calculation of local thermal-hydraulic conditions for qualification of safety class electrical equipment
Radiation Dose Specification for Equipment Qulaification
Modelling of thermal energy losses in industrial facilities with controlled environmental temperature
Modelling of Chashma NPP Reactor Core Using FUMACS Computer Code Package
IRIS Core Neutronics Modeling
Three Mile Island Unit 1 Main Steam Line Break Three-Dimensional Neutronics-Thermal-Hydraulics Analysis
IRIS Core Criticality Calculations
Coupled RELAP5/GOTHIC Model for IRIS Reactor SBLOCA Analysis
The Design and Safety Features of the IRIS Reactor
Overview of Computational Challenges in the Development of Evaluation Models for the IRIS Reactor
IRIS RELAP 5 mod 3.3 Nodalization and Steady State Qualification
LOCA Analysis of the IRIS Reactor
Development of RELAP5 Nodalization for IRIS Non-LOCA Transient Analyses
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
Design Of A Four-Year Straight-Burn Core For The Generation-IV Iris Reactor
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
Analysis of Environmental Conditions for NPP Krško DC Battery and Battery Charger Rooms Following SG Blowdown Processing System Line Break
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
Method for Analysis of Asymmetric Accidents in Nuclear Power Plants
Benchmark Performed by Different Coupled 3-D Neutronics Thermal-Hydraulic Codes
VERIFICATIONS OF ELECTRICAL AND MECHANICAL PROPERTIES FOR COMPACT TRANSMISSION LINE TOWER HEAD DESIGN
Evaluation of a Multisource Option Introduced into QAD-CGGP Code
Status of CAMP Activities in Croatia
Upgrade of the FUMACS Code Package
Application of the coupled code RELAP5- QUABOX/CUBBOX in Best-Estimate Analyses of Nuclear Power Plants
TMI-1 MSLB COUPLED 3-D NEUTRONICS/THERMALHYDRAULICS ANALYSIS: APPLICATION OF RELAP5-3D AND COMPARISON WITH DIFFERENT CODES
THERMAL-HYDRAULIC MODELING OF RERACKED SPENT FUEL POOL
Thermal-Hydraulic Modeling of Reracked Spent Fuel Pool
Modernization of the FUMACS code package
NPP KRSKO STEAM LINE BREAK ANALYSIS BY MEANS OF COUPLED CODE RELAP5-QUABOX/CUBBOX
DEVELOPMENT OF POWER REACTOR SAFETY ANALYSES COUPLED SYSTEM FOR PERSONAL COMPUTER ENVIRONMENT
VERIFICATION OF THE COUPLED CODE RELAP5-QUABOX/CUBBOX BY LARGE LOAD REJECTION ANALYSIS FOR NPP KRŠKO
ANALYSIS OF INADVERTENT CONTAINMENT SPRAY ACTUATION FOR NPP KRŠKO
NPP KRŠKO CONTAINMENT RESPONSE FOLLOWING MAIN STEAM LINE BREAK
ASSESSMENT OF THE MSLB ACCIDENT SAFETY MARGINS FOR NPP KRŠKO
APPLICATION OF THE COUPLED CODE RELAP5-QUABOX/CUBBOX IN BEST-ESTIMATE ANALYSES OF NUCLEAR POWER PLANTS
CORE-WIDE DNBR CALCULATION FOR NPP KRŠKO MSLB
SLB Analysis of NPP Krško Using Coupled RELAP5/MOD3.2-QUABOX/CUBBOX CODE
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
Quality of Education and Education of Quality in Croatia
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
RELAP5/mod2 Code for Win95/NT Environment with F90 Capabilities
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RELAP5/MOD2 Code for Win95/NT Environment with F90 Capabilities
RELAP5/Mod3.2 - Quabox/Cubbox-Hyca Coupled Code
Use of GOTHIC Code for Assessment of Equipment Environmental Qualification
NISA FEM Support in Seismic Qualification of Small Class 1E Electric Motors
Capability of the QUABOX/CUBBOX-ATHLET Coupled Code System
The Development of the Safety Related Valve Class 1E Electric Motor, the Target and the Results
RELAP5/mod2 Analysis of Inadvertent Closing of the Main Steam Isolation Valve in NPP Krško
Concrete Spent Fuel Cask Criticality Calculation
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
Investigation of intermediate conversion pressurized water reactor for small or medioum nuclear systems
Evaluation of ecological consequences of closing NP Krško
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
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
Analysis of transients caused by fast reactivity increase
IAEA Coordinated research programme on requirements for application of advanced reactors, Second annual progress report
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)
- Power Plants (Lecturer in charge)
- BSc Thesis (Lecturers)
- Project (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)
- Laboratory of Electrical Power Engineering 1 (Lecturer in charge)
- Nuclear Engineering (Lecturer in charge)
- Nuclear Safety (Lecturer in charge)
- Power Generation (Lecturer in charge)
- Power Generation (Lecturer in charge)
- Graduation Thesis (Lecturers)
- Project (Lecturers)
- Project (Lecturers)
- Project (Lecturers)
- Seminar 1 (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