Benchmark Calculation of FHR Fuel Assembly Phase I-C Depletion Exercises

Radiation dose rate analysis of conceptual solution for the Croatian low- and intermediate-level radioactive waste

Gamma Dose Rate Calculations for RCC4 and RCC8 Containers using Monte Carlo Codes

Usage of Monte Carlo Code Serpent2 for Calculation of FHR Fuel Assembly

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

Analysis of the HI-TRAC VW Transfer Cask Dose Rates During Nuclear Power Plant Krsko Spent Fuel Dry Storage Campaign One

Exploring the Factors Influencing Expansion of Nuclear Energy in Croatia

Analysis of the HI-TRAC VW Transfer Cask Dose Rates During Nuclear Power Plant Krsko Spent Fuel Dry Storage Campaign One

Influence of Spacer Grids Homogenization on Core Reactivity and Axial Power Distribution

Evaluation of the NEK SFDS Cask Model Using Hybrid Shielding Methodology

Comparison Between ORIGEN2.2 and ORIGEN-S Calculated Source Term

Student Polling on Nuclear Energy and Radioactive Waste Management

Ispitivanje studentskog mnijenja o nuklearnoj energiji i radioaktivnom otpadu

Influence of Spacer Grids Homogenization on Core Reactivity and Axial Power Distribution

Calculation of Radioactive Inventory for Severe Accident and Consequence Codes

National Survey on Nuclear Energy and Radioactive Waste in Croatia

Radioactive Inventory Data for Severe Accident and Consequence Calculation Codes

Ispitivanje i analiza javnog mnijenja u Republici Hrvatskoj o nuklearnoj energiji i radioaktivnom otpadu

The free fall experiment

Nuclear and thermal hydraulic calculation of a representative I2S-LWR first core

Public Opinion Survey - Energy - The Present and the Future – 2015.

Spent Fuel Pool Dose Rate Calculations Using Point Kernel and Hybrid Deterministic-Stochastic Shielding Methods

I2S-LWR Activation Analysis of Heat Exchangers Using Hybrid Shielding Methodology With SCALE6.1

I2S-LWR Activation Analysis of Heat Exchangers Using Hybrid Shielding Methodology With SCALE6.1

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

Ispitivanje i analiza javnog mnijenja o izvorima energije

Proton beam irradiation effects on PET and PET nanocomposites with carbon based nanoparticles

Fuel Depletion Modeling of Reconstituted NEK Fuel Assembly Using Lattice Cell Programs

Public Opinion Survey - Energy – The Present and the Future – 2012/2013

Nadogradnja NEMIS portala 2014

Full Core Criticality Modeling of Gas-Cooled Fast Reactor using the SCALE6.0 and MCNP5 Code Packages

Three Years’ Experience of the Web Site NEMIS

Nastavak ispitivanja i analize javnog mnijenja o nuklearnoj energiji

Xenon Correction in Homogenized Neutron Cross Sections

Nadogradnja portala NEMIS

Full Core Criticality Modeling of Gas-Cooled Fast Reactor Using the SCALE6.0 and MCNP5 Code Packages

Lattice Codes Pin Power Prediction Comparison

Fuel Depletion Modeling of a Gas-cooled Fast Reactor Using the SCALE6.0 Code Package

Two Years Experience of the Web Site NEMIS - “Nuclear Energy – Mysticism and Reality”

Upgrade of the FUMACS 2005 Code Package

Spectral Codes Pin Power Prediction Comparison

Prediction Capabilities of Spectral Codes DRAGON, FA2D, NEWT

Upgrade of FUMACS Code Package for Modeling of NGF and Gadolinium, Final Report

Nadogradnja informatičkog portala NEMIS - Nuklearna energija mistika i stvarnost

Sensitivity Studies of Fuel Pin Temperature for PWR Fuel Assemblies Containing Burnable Absorbers

10 godina rada hrvatskih znanstvenika na problemima razminiranja

Informacijski portal o nuklearnoj energiji NEMIS - nuklearna energija; mistika i stvarnost

Public Opinion Survey : "Energy – The Present and the Future"

Public Opinion Survey - Energy - The Present and the Future

Verification of FA2D Prediction Capability Using Fuel Assembly Benchmark

FA2D Prediction Capability for NPP Krsko Fuel Assembly Calculation

Rad na razvoju sustava za otkrivanje mina brzim i sporim neutronima

Ispitivanje i analiza javnog mnijenja o nuklearnoj energiji

Neutroničko modeliranje nuklearnog goriva koje sadrži integralne sagorive apsorbere

Pregled statusa nuklearne energetike u EU i SAD, te prikladnost pojedinih tehnologija za RH

Analiza nabave nuklearnog goriva za NE Krško

Three-Batch Reloading Scheme for IRIS Reactor Extended Cycles

Nuclear Method for Humanitarian Demining

Analiza dugoročnog smještaja nisko i srednje aktivnog otpada u NE Krško

Radiation Dose rates in the Vicinity of the NPP Krško Reracked Spent Fuel Pool in Real Case Scenarios

Tehničke komponente uređaja PELAN

IRIS Core Criticality Calculations

Tehno-ekonomska optimizacija odlaganja ozračenog goriva u NE Krško u redovnom i produljenom životnom vijeku

Ekonomsko-energetska opravdanost i tehničke mogućnosti uvođenja 18-mjesečnog radnog ciklusa za NE Krško

Neutroničko modeliranje jezgre IRIS reaktora

Upgrade Of The Fumacs Code Package

Core Design Calculations of IRIS Reactor Using Modified CORD-2 Code Package

Defects in polycristalline silicon studied by IBICC

Calculation of Core Design Benchmark 44 for IRIS Reactor Using Modified CORD-2 Code Package

Analiza prijelaznih radnih ciklusa NE Krško u uvjetima povećane snage

Utjecaj tehničkih i financijskih ograničenja odlaganja visoko radioaktivnog otpada na razvoj nuklearne energetike

Upgrade of the FUMACS Code Package

Thermal-Hydraulic Modeling of Reracked Spent Fuel Pool

Public opinion survey "Nuclear energy - the present and the future"

Radiological shielding analysis of special building for the old NPP Krško steam generators

THERMAL-HYDRAULIC MODELING OF RERACKED SPENT FUEL POOL

Modernizacija paketa računalskih programa FUMACS

DEVELOPMENT OF POWER REACTOR SAFETY ANALYSES COUPLED SYSTEM FOR PERSONAL COMPUTER ENVIRONMENT

Proračun štita spremišta za privremenu pohranu dotrajalih parogeneratora NE Krško

Validation of SCALE Code Package on High Performance Neutron Shields

Utjecaj mogućnosti povećanja kapaciteta bazena za istrošeno gorivo NE Krško uz pretpostavku gušćeg smještaja goriva i odgora istrošenog goriva na brzine doza

Validation of QAD-CGGP for Shielding Analyses of Intermediate and Low Level Waste Drums

Concrete spent fuel storage casks dose rates

Analiza 24 mjesečnog radnog ciklusa NE Krško u uvjetima povećanja snage

Razvoj metoda i programa za proračun štitova od nuklearnog zračenja

Proračun termohidrauličkih parametara bazena za istrošeno gorivo NE Krško uz pretpostavku gušćeg smještaja goriva i odgora istrošenog goriva

Extended Burnup Impact on the TN24 Spent Fuel Storage Cask Main Parameters

Generiranje udarnih presjeka za redoviti pogon i sigurnosne analize Nuklearne elektrane Krško

Analiza pohrane istrošenog goriva akumuliranog tijekom čitavog radnog vijeka NE Krško u postojećem bazenu za istrošeno gorivo