Science and Technology of Nuclear Installations: Vol 2008, No 1

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Science and Technology of Nuclear Installations

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Selected Papers from the International Conference on Nuclear Energy for New Europe 2007

Modelling of the Production of Source Neutrons from Low-Voltage Accelerated Deuterons on Titanium-Tritium Targets

CFD Code Validation against Stratified Air-Water Flow Experimental Data

Use of Nuclear Data Sensitivity and Uncertainty Analysis for the Design Preparation of the HCLL Breeder Blanket Mockup Experiment for ITER

Simulation of MASPn Experiments in MISTRA Test Facility with COCOSYS Code

Simulation of Boiling Flow Experiments Close to CHF with the Neptune_CFD Code

Power Distribution and Possible Influence on Fuel Failure in WWER-1000

Decay Heat Removal and Transient Analysis in Accidental Conditions in the EFIT Reactor

Theoretical and Numerical Study of Heat Transfer Deterioration in High Performance Light Water Reactor

Machine Learning of the Reactor Core Loading Pattern Critical Parameters

Analysis of TROI-13 Steam Explosion Experiment

Fission Product Transport and Source Terms in HTRs: Experience from AVR Pebble Bed Reactor

Gamma Decay Heat Distribution in Core: A Known Issue Revisited

Using Safety Margins for a German Seismic PRA

RELAP5/MOD3.3 Code Validation with Plant Abnormal Event

Comparison of Methods for Dependency Determination between Human Failure Events within Human Reliability Analysis

Deterministic Safety Technology for RBMK Reactors

Coupling of Modular High-Temperature Gas-Cooled Reactor with Supercritical Rankine Cycle

Flow Instabilities in Boiling Two-Phase Natural Circulation Systems: A Review

Scaling, Uncertainty, and 3D Coupled Code Calculations in Nuclear Technology

Natural Circulation in Nuclear Reactor Systems

Wastes Management Through Transmutation in an ADS Reactor

Accident Management in VVER-1000

Effect of Loop Diameter on the Steady State and Stability Behaviour of Single-Phase and Two-Phase Natural Circulation Loops

Evaluation of Heat Removal from RBMK-1500 Core Using Control Rods Cooling Circuit

A Critical Review of the Recent Improvements in Minimizing Nuclear Waste by Innovative Gas-Cooled Reactors

Numerical Simulations and Design Optimization of the PHT Loop of Natural Circulation BWR

SBWR Model for Steady-State and Transient Analysis

Effect of Coolant Inventories and Parallel Loop Interconnections on the Natural Circulation in Various Heat Transport Systems of a Nuclear Power Plant during Station Blackout

Flow Stagnation under Single and Two-Phase Natural Circulation Conditions in the APEX-CE Test Facility

Natural Circulation Characteristics at Low-Pressure Conditions through PANDA Experiments and ATHLET Simulations

Numerical Analysis of General Trends in Single-Phase Natural Circulation in a 2D-Annular Loop

International Standard Problems and Small Break Loss-of-Coolant Accident (SBLOCA)

Use of the Natural Circulation Flow Map for Natural Circulation Systems Evaluation

International Course to Support Nuclear Licensing by User Training in the Areas of Scaling, Uncertainty, and 3D Thermal-Hydraulics/Neutron-Kinetics Coupled Codes: 3D S.UN.COP Seminars

Buckling of Imperfect Thin Cylindrical Shell under Lateral Pressure

Dynamic Analysis of Coolant Channel and Its Internals of Indian 540 MWe PHWR Reactor

GRS Method for Uncertainty and Sensitivity Evaluation of Code Results and Applications

A Statistical Methodology for Determination of Safety Systems Actuation Setpoints Based on Extreme Value Statistics

Reliability Assessment of 2400 MWth Gas-Cooled Fast Reactor Natural Circulation Decay Heat Removal in Pressurized Situations

Validation of Infinite Impulse Response Multilayer Perceptron for Modelling Nuclear Dynamics

Development Considerations of AREVA NP Inc.′s Realistic LBLOCA Analysis Methodology

Dynamics of Fluid Fuel Reactors in the Presence of Periodic Perturbations

Deterministic Analysis of Natural Circulation Events at the Ignalina NPP

Bootstrap and Order Statistics for Quantifying Thermal-Hydraulic Code Uncertainties in the Estimation of Safety Margins

Thermal-Hydraulic System Codes in Nulcear Reactor Safety and Qualification Procedures

An Overview of Westinghouse Realistic Large Break LOCA Evaluation Model

Analyses of Instability Events in the Peach Bottom-2 BWR Using Thermal-Hydraulic and 3D Neutron Kinetic Coupled Codes Technique

Thermal-Hydraulic Analysis Tasks for ANAV NPPs in Support of Plant Operation and Control

BWR Stability Issues in Japan

Approaches, Relevant Topics, and Internal Method for Uncertainty Evaluation in Predictions of Thermal-Hydraulic System Codes

Regulatory Scenario for the Acceptance of Uncertainty Analysis Methodologies for the LB-LOCA and the Brazilian Approach

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