Science and Technology of Nuclear Installations

Comparative Analysis of the Dalat Nuclear Research Reactor with HEU Fuel Using SRAC and MCNP5

Numerical Study on Hydrogen Flow Behavior in Two Compartments with Different Connecting Pipes

Postprocessing of Accidental Scenarios by Semi-Supervised Self-Organizing Maps

Prediction of Flow and Temperature Distributions in a High Flux Research Reactor Using the Porous Media Approach

Stand-Alone Containment Analysis of the Phébus FPT Tests with the ASTEC and the MELCOR Codes: The FPT-0 Test

Dispersed Two-Phase Flow Modelling for Nuclear Safety in the NEPTUNE_CFD Code

Engineering Design of a Voloxidizer with a Double Reactor for the Hull Separation of Spent Nuclear Fuel Rods

Implementation and Comparison of High-Resolution Spatial Discretization Schemes for Solving Two-Fluid Seven-Equation Two-Pressure Model

Calculation of Lead-Iron Double-Layer Thickness for Gamma-Ray Shielding by MATLAB Program

Analysis of NEA-NSC PWR Uncontrolled Control Rod Withdrawal at Zero Power Benchmark Cases with NODAL3 Code

Development and Application of the Power Plant Real-Time Temperature and Stress Monitoring System

Seismic Vulnerability Assessment of Site-Vicinity Infrastructure for Supporting the Accident Management of a Nuclear Power Plant

Safety Features of High Temperature Gas Cooled Reactor

Modeling Loss-of-Flow Accidents and Their Impact on Radiation Heat Transfer

Feasibility Studies on Pyro-SFR Closed Fuel Cycle and Direct Disposal of Spent Nuclear Fuel in Line with the Latest National Policy and Strategy of Korea

Science and Technology of Nuclear Installations

Table of Contents

Table of Contents: 2017

Comparative Analysis of the Dalat Nuclear Research Reactor with HEU Fuel Using SRAC and MCNP5

Numerical Study on Hydrogen Flow Behavior in Two Compartments with Different Connecting Pipes

Postprocessing of Accidental Scenarios by Semi-Supervised Self-Organizing Maps

Prediction of Flow and Temperature Distributions in a High Flux Research Reactor Using the Porous Media Approach

Stand-Alone Containment Analysis of the Phébus FPT Tests with the ASTEC and the MELCOR Codes: The FPT-0 Test

Dispersed Two-Phase Flow Modelling for Nuclear Safety in the NEPTUNE_CFD Code

Engineering Design of a Voloxidizer with a Double Reactor for the Hull Separation of Spent Nuclear Fuel Rods

Implementation and Comparison of High-Resolution Spatial Discretization Schemes for Solving Two-Fluid Seven-Equation Two-Pressure Model

Calculation of Lead-Iron Double-Layer Thickness for Gamma-Ray Shielding by MATLAB Program

Analysis of NEA-NSC PWR Uncontrolled Control Rod Withdrawal at Zero Power Benchmark Cases with NODAL3 Code

Development and Application of the Power Plant Real-Time Temperature and Stress Monitoring System

Seismic Vulnerability Assessment of Site-Vicinity Infrastructure for Supporting the Accident Management of a Nuclear Power Plant

Safety Features of High Temperature Gas Cooled Reactor

Modeling Loss-of-Flow Accidents and Their Impact on Radiation Heat Transfer

Feasibility Studies on Pyro-SFR Closed Fuel Cycle and Direct Disposal of Spent Nuclear Fuel in Line with the Latest National Policy and Strategy of Korea