Pristupačnost
A Comparison of the Radioactive Waste Produced for Different Nuclear Energy Development Scenarios
Exploring the Factors Influencing Expansion of Nuclear Energy in Croatia
Dynamics for Sustainable Nuclear Buildup Based on LWR and FBR Technologies and Its Impact on CO2 Emission Reduction
Radioactive waste management in Croatia - public opinion, legal framework, and policy
PCA Benchmark Analysis with ADVANTG3.0.1. and MCNP6.1.1b Codes
Student Polling on Nuclear Energy and Radioactive Waste Management
Student opinion survey on nuclear energy and radioactive waste
PARTISN5.97 Code Verification Using keff Analytical Benchmarks
Total Ambient Dose Equivalent Neutron Buildup Factor Calculation Using TART
PCA Benchmark Analysis with ADVANTG3.0.1. and MCNP6.1.1b Codes
Public opinion survey in Republic of Croatia on nuclear energy and radioactive waste
Carbon Emission Impact for Energy Strategy in Which All Non-CCS Coal Power Plants Are Replaced by NPPs
I2S-LWR Concept Update
Draft Version of the DOCPAGANSA GUI and Control Module
National Survey on Nuclear Energy and Radioactive Waste in Croatia
I2S-LWR Activation Analysis of Heat Exchangers Using Hybrid Shielding Methodology With SCALE6.1
Carbon Emission Impact for Energy Strategy in which All Non-CCS Coal Power Plants Are Replaced by Nuclear Power Plants
Long Term Fuel Sustainable Fission Energy Perspective Relevant for Combating Climate Change
Long Term Fuel Sustainable Fission Energy Perspective Relevant for Combating Climate Change
I2S-LWR Activation Analysis of Heat Exchangers Using Hybrid Shielding Methodology With SCALE6.1
Public Opinion Survey - Energy - The Present and the Future – 2015.
PWR Containment Shielding Calculations with SCALE6.1 using Hybrid Deterministic-Stochastic Methodology
I2S-LWR Pressure Vessel Fast Fluence Calculations
Survey and analysis of public opinion on energy sources
The potential of fission nuclear energy in resolving global climate change
Severe accident gamma dose distribution through NPP Krško containment and auxiliary building calculated using SCALE6/MAVRIC sequence
Dose rates modeling of pressurized water reactor primary loop components with SCALE6.0
SCALE6.1 HYBRID SHIELDING METHODOLOGY FOR THE SPENT FUEL DRY STORAGE
RESEARCH AND DEVELOPMENT OF POINT-KERNEL METHOD APPLICATION IN RADIATION SHIELDING
Boration modeling of the PWR biological shield using SCALE6.1 hybrid shielding methodology
Application of Support Vector Regression on Neutron Buildup Factors
Upgrade of the NEMIS portal 2014
Molten Salt Thorium Reactor - A Promising Nuclear Technology to Stop Global Warming
A View on the Future of Nuclear Fission Energy
Public Opinion Survey - Energy – The Present and the Future – 2012/2013
SCALE6 Hybrid Deterministic-Stochastic Shielding Methodology for PWR Containment Calculations
Modeling of PWR Biological Shield Boration Using SCALE6.1 Hybrid Shielding Methodology
Three Years’ Experience of the Web Site NEMIS
Modeling of the ORNL PCA benchmark using SCALE6.0 hybrid deterministic-stochastic methodology
Dose Rates Modeling of Pressurized Water Reactor Primary Loop Components with SCALE6.0
Continued Research and Analysis of Public Opinion on Nuclear Energy
Monte Carlo Codes for Neutron Buildup Factors
24-Month Operating Cycle Containing Gadolinium Integral Burnable Absorbers for NPP Krško
Full Core Criticality Modeling of Gas-Cooled Fast Reactor using the SCALE6.0 and MCNP5 Code Packages
Lattice Codes Pin Power Prediction Comparison
Xenon Correction in Homogenized Neutron Cross Sections
Long Term Sustainability of Nuclear Fuel Resources
Full Core Criticality Modeling of Gas-Cooled Fast Reactor Using the SCALE6.0 and MCNP5 Code Packages
Monte Carlo Codes for Neutron Buildup Factors
PWR Pressure Vessel and Biological Shield Dose Rates Modelling Using SCALE6/FW-CADIS Methodology
Sustainability of Nuclear Fuel Resources
Upgrade of NEMIS Portal
On the Perspective of Nuclear Energy Following the Fukushima Accident
Modeling of H.B.Robinson-2 Pressure Vessel Benchmark
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”
The potential of fission nuclear power in resolving global climate change under the constraints of nuclear fuel resources and once-through fuel cycles
Editorial Selected Papers from TopSafe 2008: Safety at Nuclear Installations
On the Potential of Nuclear Fission Energy for Effective Reduction of Carbon Emission Under the Constraint of Uranium Resources Use without Spent Fuel Reprocessing
Spectral Codes Pin Power Prediction Comparison
Prediction Capabilities of Spectral Codes DRAGON, FA2D, NEWT
Modeling of Pool Critical Assembly Pressure Vessel Facility Benchmark
On the potential of nuclear fission energy for effective reduction of carbon emission under the constraint of uranium resources use without spent fuel reprocessing
Learning Support Vector Regression Models for Fast Radiation Dose Rate
Upgrade of the FUMACS 2005 Code Package
Upgrade of FUMACS Code Package for Modeling of NGF and Gadolinium, Final Report
Upgrade of the web portal NEMIS - Nuclear energy mysticism and reality
Sensitivity Studies of Fuel Pin Temperature for PWR Fuel Assemblies Containing Burnable Absorbers
Zero electric energy island concept in Croatia - preliminary study for the island of Vis
Nuclear Energy Information Portal NEMIS - nuclear energy; mysticism and reality
Sufficiency of the Nuclear Fuel
Availability of nuclear fuel for long-term expansion of nuclear power
FA2D Prediction Capability for NPP Krsko Fuel Assembly Calculation
Verification of FA2D Prediction Capability Using Fuel Assembly Benchmark
Influence of NPP Krško Core Nuclear Characteristics on RCCA Ejection Accident
On Input Vector Representation for the SVR Model of Reactor Core Loading Pattern Critical Parameters
Public Opinion Survey - Energy - The Present and the Future
Public Opinion Survey : "Energy – The Present and the Future"
Machine learning of the reactor core loading pattern critical parameters
Public opinion survey on nuclear energy
Solved Problems in Mechanics and Heat
Nuclear energy and environment
Support vector regression model for the estimation of γ -ray buildup factors for multi-layer shields
Neutronic modelling of nuclear fuel containing integral burnable absorbers
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
New Features Implemented in FUMACS 2005 Code Package and Future Perspectives in Development of the New Versions
Analysis of nuclear fuel procurement for NPP Krško
Gastric pentadecapeptide BPC 157 - effective therapy of muscle crush injury in rat, given intraperitoneally or applied locally as a cream
BPC 157 accelerates healing of transected rat quadriceps muscle
IRIS - Advanced Integral Nuclear Reactor with Modular Construction
Application of Support Vector Regression in Estimation of Buildup Factors for Double-Layered Shields
Analysis of long-term storage of low and intermediate level waste in NPP Krško
Modelling of Chashma NPP Reactor Core Using FUMACS Computer Code Package
Depletion Modeling of Integral Burnable Absorbers Containing Enriched Boron
Modeling of CHASHMA NPP Reactor Core Using FUMACS Computer Code Package
Three-Batch Reloading Scheme for IRIS Reactor Extended Cycles
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
Techno-Economical Aspects of Ultra Long Working Cycles
IRIS Core Criticality Calculations
Techno-economical Aspects of Ultra-long Working Cycles
IRIS Core Neutronics Modeling
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
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
Radiation Dose Evaluation for Hypothetical Accident with Transport Package Containing Iridium-192 Source
The impact of technical and financial constraints of high radioactive waste disposal on the development of nuclear energy
Upgrade of the FUMACS Code Package
Model of a Dry Storage Facility for a Medium Nuclear Power Plant
Modernization of the FUMACS code package
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
Investigating a Possibility to Extend an Operating Cycle Length of NPP Krško
Radiological Implications of Spent Fuel Pool Capacity Increase by Spent Fuel Consolidation
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
DEVELOPMENT OF POWER REACTOR SAFETY ANALYSES COUPLED SYSTEM FOR PERSONAL COMPUTER ENVIRONMENT
Radiation Dose Rates in the Vicinity of the NPP Krško Spent Fuel Pool
In-core fuel management in NE Krško
The analysis of core moderator temperature coefficient in NE Krško
Investigating a possibility to implement 24-month cycle in NPP Krško
Equilibrium cycles of NPP Krško after power uprate
Recent improvements in the design and manufacture of LWR fuel
Investigating a possibility to implement 24-month cycle in NPP Krško
Equilibrium Cycles of NPP Krško after Power Uprate
RELAP5/mod3.2 - QUABOX/CUBBOX-HYCA Coupling
Reracking Possibilities of the NPP Krško Spent Fuel Pool
Reracking Possibilities of the NPP Krško Spent Fuel Pool
Development of methodes and programs for nuclear radiation shields computation
Consolidation of spent fuel rods as an option to increase the capacity of spent fuel pool
Analysis of the 24-month operating cycle of the NPP Krško under uprated power conditions
Calculation of thermohydraulic parameters of the spent fuel pool of the NPP Krško with the assumption of denser fuel storage and spent fuel burnup
Generation of cross sections for normal operation and safety analysis of the Krško Nuclear Power Plant
RELAP5/Mod3.2 - Quabox/Cubbox-Hyca Coupled Code
Analysis of techical and energy system elements of importance for decision on NPP Krško status.
Impact of burnup credit on the NPP Krško spent fuel pool criticality
Impact of Burnup on the NPP Krško Spent Fuel Pool Criticality
Analysis of storage of the spent fuel accumulated during the entire working life of NPP Krško in the existing spent fuel pool
Capability of the QUABOX/CUBBOX-ATHLET Coupled Code System
Evaluation of Individual Risk from Normal Operation of Nuclear Power Plant Krško
Risk assessment of energy and other complex industrial systems in the City of Zagreb (1)
Risk assessment of energy and other complex industrial systems in the City of Zagreb (2)
Conceptual study of the spent fuel intermediate storage on the NPP site
Case Study Zagreb Project - Main Experiences and Results
Risk Assessment of the Krško NPP Normal Operation on the Public in the Zagreb Area
Evaluation of Ecological Results and Risk from Early Shut-down of Krško Nuclear Power Plant
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
Reactor Safety Research Applied to NPP Krško in 1988/1989
Nuclear fuel cycles
Techno-economic considerations for a small or medium nuclear system built within a limited time interval
Analysis, calculation and testing of shielding for radioactive waste deposition
Modelling transient from uncontrolled reduction of boric acid concentration for NPS Krško
Some possibilities of fuel cost reductions for light water reactors (PWR) and heavy water reactors
Calculation of gamma shielding of medium active nuclear power station waste
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
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 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