A5061000700- Nuclear reactor physics and engineering
- Graphite, nuclear technology, radiation studies
- Nuclear Materials and Properties
- Nuclear Physics and Applications
- Field-Flow Fractionation Techniques
- Heat transfer and supercritical fluids
- Nuclear and radioactivity studies
- Advanced Multi-Objective Optimization Algorithms
Saigon University
2023-2024
Văn Hiến University
2023
Ho Chi Minh City University of Technology and Education
2011-2021
Institute for Nuclear Science and Technology
2006
This article presents results obtained from a research into an application of simulated annealing method to the in-core fuel reloading pattern optimization for reactor. The decision variable problem is next cycle after present finishes. objective function maximizes effective multiplication factor keff at beginning while it established include important safety paramater – power peaking factor, in search process. A procedure searching optimal solutions was formed and computer code developed...
Abstract This paper presents a detailed description of new variant differential evolution for nuclear reactor refueling optimization problem. combines the elitism strategy with discrete evolution. The allows non-dominated solutions found during search and stored in archive to participate operation. population size is same as size, number participating at particular generation controlled by specific probability. proposed method successfully applied research its first time optimal loading...
This paper investigates the performance of genetic algorithm (GA) with improved selection techniques, i.e. Tournament and Roulette Wheel, applied to in-core fuel management Dalat nuclear research reactor (DNRR). Numerical calculations have been performed based on DNRR core 100 HEU bundles. The optimal fitness function was chosen maximize keff minimize power peaking factor. statistical analysis using Mann-Whitney test shows that GA is advantageous over Wheel in ICFM problem DNRR....
This paper presents results of the evaluated group constants for fuel and other important materials Miniature Neutron Source Reactor (MNSR) moderator temperature coefficient reactivity through global reactor calculation. In this study, were calculated with WIMSD code calculation is accomplished by CITATION code. work also a method evaluation at different temperatures it’s average value in range directly values MNSRs. provides simple analytical representation convenient kinetics safety assessment.
This paper presents results of the evaluated group constants for fuel and other important materials Miniature Neutron Source Reactor (MNSR) moderator temperature coefficient reactivity through global reactor calculation. In this study, were calculated with WIMSD code calculation is accomplished by CITATION code. work also a method evaluation temperatures directly values MNSRs. provides simple analytical representation convenient kinetics safety assessment.
Abstract A discrete differential evolution (DE) method has been applied to the problem of fuel loading pattern optimization Dalat Nuclear Research Reactor (DNRR). classic strategy DE/rand/l/bin was chosen for mutation DE method. Numerical calculations have performed based on core configuration 100 highly enriched uranium (HEU) bundles with various burnup levels. Comparison performance between and a genetic algorithm (GA) also carried out. The optimal LPs obtained from two methods are...