TAILIEUCHUNG - Comparison of SERPENT and SCALE methodology for LWRs transport calculations and additionally uncertainty analysis for cross-section perturbation with SAMPLER module

In nuclear safety research, the quality of the results of simulation codes is widely determined by the reactor design and safe operation, and the description of neutron transport in the reactor core is a feature of particular importance. | Comparison of SERPENT and SCALE methodology for LWRs transport calculations and additionally uncertainty analysis for cross-section perturbation with SAMPLER module EPJ Nuclear Sci. Technol. 2 11 2016 Nuclear Sciences A. Labarile et al. published by EDP Sciences 2016 amp Technologies DOI epjn e2016-50002-7 Available online at http REGULAR ARTICLE Comparison of SERPENT and SCALE methodology for LWRs transport calculations and additionally uncertainty analysis for cross-section perturbation with SAMPLER module Antonella Labarile Nicolas Olmo Rafael Miró Teresa Barrachina and Gumersindo Verd u Institute for Industrial Radiophysical and Environmental Safety ISIRYM Universitat Politècnica de València Camí de Vera s n 46022 Valencia Spain Received 28 April 2015 Received in final form 30 October 2015 Accepted 12 January 2016 Published online 18 March 2016 Abstract. In nuclear safety research the quality of the results of simulation codes is widely determined by the reactor design and safe operation and the description of neutron transport in the reactor core is a feature of particular importance. Moreover for the long effort that is made there remain uncertainties in simulation results due to the neutronic data and input specification that need a huge effort to be eliminated. A realistic estimation of these uncertainties is required for finding out the reliability of the results. This explains the increasing demand in recent years for calculations in the nuclear fields with best-estimate codes that proved confidence bounds of simulation results. All this has lead to the Benchmark for Uncertainty Analysis in Modelling UAM for Design Operation and Safety Analysis of LWRs of the NEA. The UAM-Benchmark coupling multi-physics and multi- scale analysis using as a basis complete sets of input specifications of boiling water reactors BWR and pressurized water reactors PWR . In this study the results of the transport calculations carried out using the .

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• Nuclear safety research
• Uncertainty Analysis in Modelling
• UAM Benchmark
• Coupling multi physics
• Boiling water reactors
• Nuclear engineering and technology
• Nordic research and development cooperation
• Strengthen nuclear reactor safety after the Fukushima accident
• The Fukushima accident
• Nordic Nuclear Safety Research
• Neutronic analysis of control rod effect on safety parameters
• Tehran Research Reactor
• The control shim safety rods
• Power peaking factor
• Moderator temperature coefficient
• Nuclear engineering and design
• Brayton cycle
• Supercritical carbon dioxide
• Nuclear engineering
• Safety analysis
• Control strategy
• Safety classification of systems structures
• Components for pool type research reactors
• Pool type research reactors
• Safety importance and significance
• Thermal hydraulics
• Safety analyses
• Nuclear research reactor
• Design analyses for full core conversion
• Cladding temperature
• Real time interactive transient simulator
• Dalat nuclear research reactor
• Human machine interface
• The safety analysis report
• Best estimate thermal hydraulic
• Thermal hydraulic analysis of core flow bypass
• Typical research reactor
• Core flow bypass
• Safety margins during normal operation
• Nuclear Medical Imaging
• Imaging Techniques
• Recent Developments
• Magnetic Resonance Imaging
• Safety Standards
• Integrated Level 1eLevel 2
• Probabilistic risk assessment for boiling water reactors
• Large Early Release Frequency
• Level 1 Probabilistic Risk Assessment
• Level 2 Probabilistic Risk Assessment
• Nuclear energy
• Boiling water reactor
• Loading pattern optimisation
• Simulated annealing
• Nuclear reactors
• Fuel performance
• Fuel Assemblies
• Reactivity Control
• Reactor Pressure Vessel
• BWR Generations
• Control rods