Measurement of the Efficiency of a 6LiD Thermal-to-14 MeV Neutron Converter in the Experimental Channel of an IVV-2M Reactor

The results of measurements of the efficiency of a ⁶LiD thermal-to-fast neutron converter for neutrons from DT and ⁶LiT fusion reactions with energy 14 MeV in the experimental channel of an IVV-2M reactor are presented. The first experimental estimates of the conversion coefficients for the corresponding fusion reactions are obtained: K ^D 2.11·10^–4 and K ^Li 1.36·10^–4. The value found in this work for the total conversion coefficient 3.47·10^–4 is approximately 1.7 times greater than the previously measured value and about 20% greater than the maximum computed estimate for a ⁶LiD converter.An experimental apparatus with a ⁶LiD converter in an IVV-2M channel is an accessible, comparatively inexpensive, and unique (for now) source of 14 MeV neutrons that can provide continuous and approximately uniform irradiation of Ø4 × 20 cm samples by neutrons from DT and ⁶LiT fusion 2.7·10¹⁰ sec^–1·cm^–2 for a comparatively long time (500 h).     

Effects of various spacer grid modeling on the neutronic parameters of the VVER-1000 reactor

The main objective of this paper is to study the effects of various spacer grid models on the neutronic parameters of a VVER-1000 reactor. Specifically, the data of the nuclear power plant at the Bushehr site, which is of a VVER-1000 type, will be studied. Three models, representing the spacer grids along the fuel assemblies are presented. These three models are the homogeneous and the heterogeneous local spacer grid models and the shroud spacer grid model. In the homogeneous and the heterogeneous models, the spacer grids are considered at their actual locations in the axial direction. The only difference between the two models is that in the homogeneous model, the spacer grids are homogenized with the coolant while in the heterogeneous model, the spacer grids are modeled around the fuel cells at their exact axial positions. In the shroud model, the spacer grids are modeled in the shroud region containing the coolant and are not necessarily placed at their appropriate axial positions.     

The determination of reactor kinetic parameters in a two-core reactor

The kinetic parameters, α the coupling coefficient and gt the mean neutron transit time have been determined using a reactor oscillator on the coupled-core of the Queen Mary College research reactor. By using correlation techniques it has proved possible to use detectors small enough to be inserted in the fuel tanks. It is shown that the simplified Baldwin model with one-group diffusion theory is inadequate to describe the kinetic behaviour and the experimentally-determined parameters are dependent upon the positioning of the detectors.     

Kinetic parameters of a material test research reactor fueled with various low enriched uranium dispersion fuels

The effects of using different low enriched uranium fuels, having same uranium density, on the kinetic parameters of a material test research reactor were studied. For this purpose, the original aluminide fuel (UAl_x-Al) containing 4.40 gU/cm³ of an MTR was replaced with silicide (U₃Si-Al and U₃Si₂-Al) and oxide (U₃O₈-Al) dispersion fuels having the same uranium density as of the original fuel. Simulations were carried out to calculate prompt neutron generation time, effective delayed-neutron fraction, core excess reactivity and neutron flux spectrum. Nuclear reactor analysis codes including WIMS-D4 and CITATION were used to carry out these calculations. It was observed that both the silicide fuels had the same prompt neutron generation time 0.02% more than that of the original aluminide fuel, while the oxide fuel had a prompt neutron generation time 0.05% less than that of the original aluminide fuel. The effective delayed-neutron fraction decreased for all the fuels; the decrease was maximum at 0.06% for U₃Si₂-Al followed by 0.03% for U₃Si-Al, and 0.01% for U₃O₈-Al fuel. The U₃O₈-Al fueled reactor gave the maximum ρ_excess at BOL which was 21.67% more than the original fuel followed by U₃Si-Al which was 2.55% more, while that of U₃Si₂-Al was 2.50% more than the original UAl_x-Al fuel. The neutron flux of all the fuels was more thermalized, than in the original fuel, in the active fuel region of the core. The thermalization was maximum for U₃O₈-Al followed by U₃Si-Al and then U₃Si₂-Al fuel.     

世界核电主要堆型技术沿革

介绍了世界核电主要反应堆堆型的工作原理、研发历史、现状及发展趋势,重点放在我国和世界核电的主力堆型——压水堆上。还介绍了正在研发的第四代核能系统。     

世界核电主要堆型技术沿革

介绍了世界核电主要反应堆堆型的工作原理、研发历史、现状及发展趋势,重点放在我国和世界核电的主力堆型——压水堆上。还介绍了正在研发的第四代核能系统。     

世界核电主要堆型技术沿革

介绍了世界核电主要反应堆堆型的工作原理、研发历史、现状及发展趋势,重点放在我国和世界核电的主力堆型——压水堆上。还介绍了正在研发的第四代核能系统。     

世界核电主要堆型技术沿革

介绍了世界核电主要反应堆堆型的工作原理、研发历史、现状及发展趋势,重点放在我国和世界核电的主力堆型——压水堆上。还介绍了正在研发的第四代核能系统。