PristupačnostNa FER-u postoji više zaposlenika s imenom
Analysis of the HI-TRAC VW Transfer Cask Dose Rates During Nuclear Power Plant Krsko Spent Fuel Dry Storage Campaign One
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
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
Evaluation of the NEK SFDS Cask Model Using Hybrid Shielding Methodology
Dose Calculation for Emergency Control Room HVAC Filter
Influence of Spacer Grids Homogenization on Core Reactivity and Axial Power Distribution
ENERGETSKA TRANZICIJA PREMA TRANSFORMACIJI ENERGETIKE
Comparison of Measured and Calculated Data for NPP Krško CILR Test
Reactor Vessel Modelling with the MELCOR Code
Contribution of FER to the WP4: summary report for the first year, TA2/SARNET ASCOM Technical report
Point Kernel Modification Including Support Vector Regression Neutron Buildup Factor Models
Calculation of Radioactive Inventory for Severe Accident and Consequence Codes
Dose Calculation for Emergency Control Room HVAC Filter
Influence of Spacer Grids Homogenization on Core Reactivity and Axial Power Distribution
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
Effectiveness of SFP Spray Cooling during Loss of Coolant Accidents
NPP Krsko IB Modelling with GOTHIC and MELCOR Codes
Comparison Between ORIGEN2.2 and ORIGEN-S Calculated Source Term
Comparison of Measured and Calculated Data for NPP Krško CILR Test
Point Kernel Modification Including Support Vector Regression Neutron Buildup Factor Models
Hybrid Shielding Methods Validation Using Graphite Shielding Measurements
Analysis of Spent Fuel Pool Loss of Coolant Inventory Accident Progression
Radioactive Inventory Data for Severe Accident and Consequence Calculation Codes
I2S-LWR Concept Update
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
Neutron Buildup Factors Calculation for Support Vector Regression Application in Shielding Analysis
Spent Fuel Pool Dose Rate Calculations Using Point Kernel and Hybrid Deterministic-Stochastic Shielding Methods
Nuclear and thermal hydraulic calculation of a representative I2S-LWR first core
Preparation of Radioactive Inventory Data for Severe Accident and Consequence Calculation Codes
Proceedings of the 11th International Conference of the Croatian Nuclear Society
NPP Krško Post-UFC Transient Response during MSLB
Spent Fuel Pool Dose Rate Calculations Using Point Kernel and Hybrid Deterministic-Stochastic Shielding Methods
Nuclear and thermal hydraulic calculation of a representative I2S-LWR first core
NPP Krško Station Blackout Analysis after Safety Upgrade Using MELCOR Code
Verification of GOTHIC Multivolume Containment Model during NPP Krško DBA LOCA
Severe accident gamma dose distribution through NPP Krško containment and auxiliary building calculated using SCALE6/MAVRIC sequence
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
Radiation Doses Estimation For Hypothetical NPP Krško Accidents Without And With PCFV Using RASCAL Software
Fuel Depletion Modeling of Reconstituted NEK Fuel Assembly Using Lattice Cell Programs
Hot Leg Streaming Calculation for Two-Loop PWR Plant
NPP Krško Containment Modelling with the ASTEC Code
Calculation of Hydrogen Concentration in Containment during LOCA Accident
Optimization of OPDT Protection for Overcooling Accidents
Decay Heat Calculation for Spent Fuel Pool Application
Spatial Distribution of Hydrogen in NEK Containment
Radiation heat transfer in a pressurized water reactor lower head filled with molten corium
Coupled code calculation of rod withdrawal at power accident
Severe Accident Analysis in the NPP Krško with the ASTEC Code
Severe Accident Analysis in a Two-Loop PWR Nuclear Power Plant with the ASTEC 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
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
Lattice Codes Pin Power Prediction Comparison
Xenon Correction in Homogenized Neutron Cross Sections
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
Testing of an Integral Layout SMR on the SPES3 Facility: BDBE Simulation and Influence of the PCC Actuation Delay on the Accident Recovery
Coupled Code Calculation of Rod Withdrawal at Power Accident
ORIGEN2.1 Cycle Specific Calculation of Krško Nuclear Power Plant Decay Heat and Core Inventory
Condensation within Small Compartments during Design Basis Accidents
SPES3 Facility and IRIS Reactor Numerical Simulations For The SPES3 Final Design
Spectral Codes Pin Power Prediction Comparison
Prediction Capabilities of Spectral Codes DRAGON, FA2D, NEWT
Comparison of R5G Coupled Code and Classical “Two-steps” Containment Calculation
Corium Behaviour and the Lower Head Thermal Response after a Core Meltdown
RELAP5 Modeling of PWR Reactor RTD Bypass
Upgrade of the FUMACS 2005 Code Package
Analysis of Different Containment Models for IRIS Small Break LOCA, using GOTHIC and RELAP5 Codes
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
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
Potential advantages and disadvantages of sequential building small nuclear units instead of a large nuclear power plant
Verification of FA2D Prediction Capability Using Fuel Assembly Benchmark
Influence of NPP Krško Core Nuclear Characteristics on RCCA Ejection Accident
Analysis of In-Vessel Severe Accident Phenomena in NPP Krško
Visokonaponski nadzemni elektroenergetski vodovi i primjena Pravilnika o zaštiti od elektromagnetskih polja
Comparison of economy and emissions of nuclear power plants and combined gas and wind electricity generators
NPP Krško Core Calculations to Improve Operational Safety
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
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
IRIS – Progress in Licensing and Toward Deployment
Moguća uloga naprednog nuklearnog reaktora IRIS u hrvatskom elektroenergetskom sustavu
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
IRIS (International Reactor Innovative and Secure) - Design Overview and Deployment Prospects
Analysis of future nuclear power plants competitive investment costs with stochastic methods
IRIS - napredni nuklearni reaktor srednje snage s poboljšanom sigurnošću i ekonomičnošću
IRIS - napredni nuklearni reaktor srednje snage s poboljšanom sigurnošću i ekonomičnošću
IRIS Design Overview and Status Update
IRIS - NAPREDNI NUKLEARNI REAKTOR SREDNJE SNAGE S POBOLJŠANOM SIGURNOŠĆU I EKONOMIČNOŠĆU
Applicability of Coupled Code RELAP5/GOTHIC to NPP Krško MSLB Calculation
IRIS - Napredni integralni nuklearni reaktor modularne izvedbe
Radiation Dose Specification for Equipment Qulaification
Small Break Loss of Cooloant Accident Analysis for the International Reactor Innovative and Secure (IRIS)
Coupled RELAP5/GOTHIC Model for IRIS SBLOCA Analysis
The Design and Safety Features of the IRIS Reactor
IRIS Small Break LOCA Phenomena Identification and Ranking Table (PIRT) Addendum 1 - IRIS Small Break LOCA Sensitivity Report for PIRT Development
Use of the Deterministic Safety Analyses in Support to the NPP Krško Modifications
Modelling of Chashma NPP Reactor Core Using FUMACS Computer Code Package
Modeling of CHASHMA NPP Reactor Core Using FUMACS Computer Code Package
Three-Batch Reloading Scheme for IRIS Reactor Extended Cycles
Development of RELAP5 Nodalization for IRIS Non-LOCA Transient Analyses
Overview of Computational Challenges in the Development of Evaluation Models for the IRIS Reactor
The Design and Safety Features of the IRIS Reactor
Coupled RELAP5/GOTHIC Model for IRIS Reactor SBLOCA Analysis
Three Mile Island Unit 1 Main Steam Line Break Three-Dimensional Neutronics-Thermal-Hydraulics Analysis
IRIS RELAP 5 mod 3.3 Nodalization and Steady State Qualification
LOCA Analysis of the IRIS Reactor
IRIS Core Criticality Calculations
Neutroničko modeliranje jezgre IRIS reaktora
Analysis of Environmental Conditions for NPP Krško DC Battery and Battery Charger Rooms Following SG Blowdown Processing System Line Break
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
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
PROVJERE OSNOVNIH ELEKTRIČKIH I MEHANIČKIH SVOJSTAVA KOMPAKTIRANIH GLAVA DALEKOVODNIH STUPOVA
Influence of Reactor Vessel Nodalizations in the Coupled Code Analysis of Asymmetric Main Feedwater Isolation
Benchmark Performed by Different Coupled 3-D Neutronics Thermal-Hydraulic Codes
Metoda za analizu nesimetričnih kvarova u nuklearnim elektranama
Status of CAMP Activities in Croatia
Upgrade of the FUMACS Code Package
Evaluation of a Multisource Option Introduced into QAD-CGGP Code
Evaluation Of A Multisource Option Introduced Into Qad-Cggp Code
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
TMI-1 MSLB COUPLED 3-D NEUTRONICS/THERMALHYDRAULICS ANALYSIS: APPLICATION OF RELAP5-3D AND COMPARISON WITH DIFFERENT CODES
APPLICATION OF THE COUPLED CODE RELAP5-QUABOX/CUBBOX IN BEST-ESTIMATE ANALYSES OF NUCLEAR POWER PLANTS
NPP KRSKO STEAM LINE BREAK ANALYSIS BY MEANS OF COUPLED CODE RELAP5-QUABOX/CUBBOX
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
CORE-WIDE DNBR CALCULATION FOR NPP KRŠKO MSLB
DEVELOPMENT OF POWER REACTOR SAFETY ANALYSES COUPLED SYSTEM FOR PERSONAL COMPUTER ENVIRONMENT
THERMAL-HYDRAULIC MODELING OF RERACKED SPENT FUEL POOL
Thermal-Hydraulic Modeling of Reracked Spent Fuel Pool
Croatian Experience in the Use of Coupled 3D Code With RELAP5
Role of the Exercises in the Training of the Technical Support Center Members in Croatia
DEVELOPMENT OF POWER REACTOR SAFETY ANALYSES COUPLED SYSTEM FOR PERSONAL COMPUTER ENVIRONMENT
Kakvoća u izobrazbi i izobrazba o kakvoći u Hrvatskoj
RELAP5/mod3.2 - QUABOX/CUBBOX-HYCA Coupling
Proračun termohidrauličkih parametara bazena za istrošeno gorivo NE Krško uz pretpostavku gušćeg smještaja goriva i odgora istrošenog goriva
RELAP5/MOD2 Code for Win95/NT Environment with F90 Capabilities
RELAP5/mod2 Code for Win95/NT Environment with F90 Capabilities
Plan i program postupaka za slučaj nesreće u NE Krško i NE Paks
RELAP5/Mod3.2 - Quabox/Cubbox-Hyca Coupled Code
Use of GOTHIC Code for Assessment of Equipment Environmental Qualification
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
RELAP5/mod2 Analysis of Inadvertent Closing of the Main Steam Isolation Valve in NPP Krško
Capability of the QUABOX/CUBBOX-ATHLET Coupled Code System
Concrete Spent Fuel Cask Criticality Calculation
Transient Temperature Behaviour Of The MO Suger Arrester Aged By Repeated Application Of Current Impulses And Analyzed By Means Of Computer Aided Thermal Modelling
Konceptualna studija privremenog odlagališta istrošenog goriva na lokaciji nuklearne elektrane
Modeliranje prijelaznih temperaturnih polja u strukturama pomoću prijenosnih funkcija
Procjena ekoloških posljedica zatvaranja NE Krško
Investigation of intermediate conversion pressurized water reactor for small or medioum nuclear systems
Razmatranje mogućnosti suhog skladištenja istrošenog goriva na lokaciji NE
Analiza, provjera i proračun štitova za odlaganje radioaktivnog otpada
Modeliranje prelazne pojave nekontroliranog smanjenja koncentracije borne kiseline za NE Krško
Analiza prijelazne pojave nekontroliranog smanjenja koncentracije borne kiseline
IAEA Coordinated Research Programme on Requirements for Future Application of Advanced Reactors, Second Annual Progress Report
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
Analiza prijelaznih pojava brzog povećanja reaktivnosti
IAEA Coordinated Research Programme on Requirements for Future Application of Advanced Reactors, Progress Report, September 1985
Analiza prijelazne pojave nekontroliranog izvlačenja banke kontrolnih štapova PWR reaktora pri punoj snazi
Analiza prijelaznih pojava nekontroliranog izvlačenja banke kontrolnih štapova na punoj snazi i izbacivanja snopa kontrolnih štapova
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Nuclear and plasma sciences
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