Annals of Nuclear Energy
ISSN: 0306-4549
Datasets associated with articles published in Annals of Nuclear Energy
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140 results
- Data for: A Study on Steam Cycle Optimization for Integrating Energy Storage System to Nuclear Power PlantRaw result data in forms of MATLAB files
- Data for: On the Nuclear Safety improvement by post-inerting Small Modular Reactor with stainless steel claddingParabolic rate constant for the reactions studied. Components volume fraction in containment. Shapiro diagram for the studied cases.
- Data for: Verification of On-the-fly Homogenization Method Based on the Analysis of HTTRraw date used by figures
- Data for: Numerical study on the uncertainty reduction for thermal power estimation of an open pool research reactor with natural convection core coolingIn this study, the calorimetric method is used to study the thermal power estimation in an open pool research reactor during power operation under natural convection core cooling. The pool temperature measuring points are used as a key factor to minimize the uncertainty of the thermal power of the research reactor. The Computational Fluid Dynamics (CFD) technique is used to analyze the natural convection flow pattern in the reactor pool. The results show how the natural convection flow pattern influences the uncertainty of the thermal power estimation in terms of the flow stabilization and how the optimal temperature measuring point can be chosen. This paper shows that an appropriate thermal power estimation with a low uncertainty can be obtained based on the temperature measurement in the lower section of the reactor pool with the proper time criteria in view of the temperature propagation.
- Data for: Numerical study on the uncertainty reduction for thermal power estimation of an open pool research reactor with natural convection core coolingIn this study, the calorimetric method is used to study the thermal power estimation in an open pool research reactor during power operation under natural convection core cooling. The pool temperature measuring points are used as a key factor to minimize the uncertainty of the thermal power of the research reactor. The Computational Fluid Dynamics (CFD) technique is used to analyze the natural convection flow pattern in the reactor pool. The results show how the natural convection flow pattern influences the uncertainty of the thermal power estimation in terms of the flow stabilization and how the optimal temperature measuring point can be chosen. This paper shows that an appropriate thermal power estimation with a low uncertainty can be obtained based on the temperature measurement in the lower section of the reactor pool with the proper time criteria in view of the temperature propagation.
- Data for: Numerical study on the uncertainty reduction for thermal power estimation of an open pool research reactor with natural convection core coolingIn this study, the calorimetric method is used to study the thermal power estimation in an open pool research reactor during power operation under natural convection core cooling. The pool temperature measuring points are used as a key factor to minimize the uncertainty of the thermal power of the research reactor. The Computational Fluid Dynamics (CFD) technique is used to analyze the natural convection flow pattern in the reactor pool. The results show how the natural convection flow pattern influences the uncertainty of the thermal power estimation in terms of the flow stabilization and how the optimal temperature measuring point can be chosen. This paper shows that an appropriate thermal power estimation with a low uncertainty can be obtained based on the temperature measurement in the lower section of the reactor pool with the proper time criteria in view of the temperature propagation.
- Data for: Two-phase modelling of liquid-gas flow of dispersed and segregated flowsThe following data were attached: -Processed data -Table and graphics -Experimental and numerical data -Cases configured to simulate each of the experiments in OpenFOAM -The modified twoPhaseEulerFoam code. One of the modifications made is the addition of the Grace Dredge model
- Data for: A Probabilistic Model-Based Diagnostic Framework for Nuclear Engineering SystemsSimulation data for a single-phase counterflow heat exchanger
- Data for: Loose parts detection method combining blind deconvolution with support vector machineThis dataset includes the computer codes developed in Matlab and raw data of all figures for the manuscript.
- Data for: Fault identification and diagnosis based on KPCA and similarity clustering for nuclear power plantsThe dataset is Qinshan PWR data under a cold leg rupture with tiny degree. Also, we provide our simulation results.
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