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

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

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

Radiation shielding analysis of the HI-STORM FW storage cask

Estimation of Dose Rates Around Dry Storage Building During Campaign One Loading in Nuclear Power Plant Krsko

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

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

Calculation of Radioactive Inventory for Severe Accident and Consequence Codes

Student Polling on Nuclear Energy and Radioactive Waste Management

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

Student opinion survey on nuclear energy and radioactive waste

Radioactive Inventory Data for Severe Accident and Consequence Calculation Codes

Public opinion survey in Republic of Croatia on nuclear energy and radioactive waste

National Survey on Nuclear Energy and Radioactive Waste in Croatia

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

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

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.

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

Survey and analysis of public opinion on energy sources

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

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

Upgrade of the NEMIS portal 2014

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

Continued Research and Analysis of Public Opinion on Nuclear Energy

Three Years’ Experience of the Web Site NEMIS

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

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

Xenon Correction in Homogenized Neutron Cross Sections

Lattice Codes Pin Power Prediction Comparison

Upgrade of NEMIS Portal

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”

Spectral Codes Pin Power Prediction Comparison

Prediction Capabilities of Spectral Codes DRAGON, FA2D, NEWT

Upgrade of the FUMACS 2005 Code Package

Upgrade of the web portal NEMIS - Nuclear energy mysticism and reality

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

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

Public Opinion Survey - Energy - The Present and the Future

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

FA2D Prediction Capability for NPP Krsko Fuel Assembly Calculation

10 years of work of Croatian scientists on the problems of demining

On the development of the device for mine detection with fast and slow neutrons

Nuclear Energy Information Portal NEMIS - nuclear energy; mysticism and reality

Verification of FA2D Prediction Capability Using Fuel Assembly Benchmark

Neutronic modelling of nuclear fuel containing integral burnable absorbers

Public opinion survey on nuclear energy

An overview of the status of nuclear energy in the EU and the USA, and the suitability of certain technologies for the Republic of Croatia

Analysis of nuclear fuel procurement for NPP Krško

Analysis of long-term storage of low and intermediate level waste in NPP Krško

Three-Batch Reloading Scheme for IRIS Reactor Extended Cycles

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

Nuclear Method for Humanitarian Demining

IRIS Core Criticality Calculations

IRIS Core Neutronics Modeling

Techno-economic optimization of spent fuel storage in NPP Krško during regular and extended life time

Economic and energy justification and technical possibilities of introducing an 18-month operating cycle for NPP Krško

PELAN Technical Components

Analysis of transit operating cycles of Krško NPP in conditions of power uprate

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

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

Defects in polycristalline silicon studied by IBICC

Upgrade Of The Fumacs Code Package

The impact of technical and financial constraints of high radioactive waste disposal on the development of nuclear energy

Upgrade of the FUMACS Code Package

THERMAL-HYDRAULIC MODELING OF RERACKED SPENT FUEL POOL

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

Modernization of the FUMACS code package

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

Thermal-Hydraulic Modeling of Reracked Spent Fuel Pool

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

Validation of SCALE Code Package on High Performance Neutron Shields

Impact of the possibility of increasing the capacity of the spent fuel pool of the NPP Krško with the assumption of denser fuel storage and burnup of spent fuel at dose rates

Calculation of the storage shield for temporary storage of replaced steam generators of NPP Krško

Analysis of the 24-month operating cycle of the NPP Krško under uprated power conditions

Generation of cross sections for normal operation and safety analysis of the Krško Nuclear Power Plant

Development of methodes and programs for nuclear radiation shields computation

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

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

Calculation of thermohydraulic parameters of the spent fuel pool of the NPP Krško with the assumption of denser fuel storage and spent fuel burnup

Concrete spent fuel storage casks dose rates

Analysis of storage of the spent fuel accumulated during the entire working life of NPP Krško in the existing spent fuel pool