The main target of this research was to estimate the possibility for improving the performance of... more The main target of this research was to estimate the possibility for improving the performance of Boiling Water Reactor (BWR) cores by using uranium thorium dioxide fuels instead of the commonly used oxide fuel. MCNPX (Monte Carlo N-Particle eXtend) computer code based on Mont Carlo method, is used to design two models for a BWR fuel bundle in a typical operating temperature and pressure conditions, the first is the oxide fuel bundle model and the second is the blanket-seed (BS) bundle model. A test case was compared with a previously published data calculated by HELIOS code and good agreement was found. The two models are used to determine the multiplication factor, pin-by-pin power distribution and thermal neutron flux distribution for both UO 2 and (UZr/ThO 2 ) fuels. BS bundle model improves the performance of BWR in different ways: (a) by increasing the energy extracted per fuel assembly, (b) by reducing the uranium ore, (c) reducing the plutonium accumulated in BWR through burn-up and continues to grow throughout the world, and (d) by increasing fuel burn-up and reactor lifetime. The spent fuel in the reactor can be recycled, and the plutonium and its isotopes can be extracted. The extracted plutonium can be used as a Mixed Oxide (MOX) fuel.
Design Boiling Water Reactor Core Model Using MCNPX for Studying the Burnable Poisons and the Axial Enrichment Fuel Effect on the Neutronic Characteristics for BWR, Gd 2 O 3 –MCNPX
The accelerator-driven system (ADS) is an innovative reactor that is being considered as a dedica... more The accelerator-driven system (ADS) is an innovative reactor that is being considered as a dedicated high-level-waste burner in a double-strata fuel cycle. ("Double-strata fuel cycle" means a partitioning and transmutation system for long-lived radioactive nuclides.) The target is the physical and functional interface between the accelerator and the subcritical reactor in the ADS, so it is probably the most innovative component of the ADS. Key parameters of ADS are the number of neutrons emitted per incident proton, the neutron multiplicity (n/p), the mean energy deposited in the target for neutrons produced, the neutron energy spectrum, and the spallation product spatial distribution. This paper focuses on the production of neutrons in the spallation reactions. The neutrons produced in the spallation reactions can be characterized by their energy and spatial distributions and multiplicity. The present calculations have been performed using the Monte Carlo code MCNPX. The Monte Carlo simulations have been performed to investigate the neutron multiplicity as a function of incident proton beam energy, as well as a function of target material and target size. Neutron flux distributions at the target surface are calculated and compared with different target materials and proton energies. A comparison of MCNPX with experimental results is made.
Monte Carlo simulation of the ETRR-2 research reactor using the MCNP Code
Kerntechnik, May 1, 2004
The MCNP computer Code is used to model the ETRR-2 research reactor. A computer program was desig... more The MCNP computer Code is used to model the ETRR-2 research reactor. A computer program was designed to evaluate the axial burn-up of the fuel elements. The excess reactivity of the reactor core is calculated for different core configurations and compared with the existing measurements. The thermal flux is also calculated and compared with measurements. Several factors that affect the safety of the reactor such as power peak and the effect of control rod insertion on the reactor power and flux were studied and analysed. The agreement between the MCNP results and the experimentally determined values is good.
Study the Effect of Gd2O3 on the Neutronic Characteristics of Boiling Water Reactor using MCNPX Code
MCNPX code ( M onte C arlo N – P article code e x tended) is used to design Boiling Water Reactor... more MCNPX code ( M onte C arlo N – P article code e x tended) is used to design Boiling Water Reactor (BWR) bundle model. This model is designed to study the effect of Gd 2 O 3 on the reactivity effect of voiding, void coefficient, neutron flux and normalized power. The incorporation of Gd 2 O 3 directly into the UO 2 fuel is the most attractive process, because of its significant effect on the fuel life in the core of the reactor. The effect of axial fuel enrichment on the normalized power is analyzed. In this study, the influences of the burnable poisons on the main parameters of the reactor such as multiplication factor and burnup are investigated. Results are discussed to assess the effect of Gd 2 O 3 on fuel cycle characteristics. Keywords: Axial Fuel Enrichment, Boiling Water Reactor, Neutronic characteristics for BWR, Gd 2 O 3 , MCNPX, Reactivity effect of voiding Cite this Article Abdelghafar Galahom, Bashter II, Moustafa Aziz. Study the Effect of Gd 2 O 3 on the Neutronic Characteristics of Boiling Water Reactor using MCNPX Code. Journal of Nuclear Engineering and Technology . 2015; 5(3): 26-32p.
Calculation of control plates life time and worth at the ETRR-2 research reactor
Kerntechnik, Aug 1, 2004
Abstract A detailed three-dimensional MCNP model of the ETRR-2 research reactor has been develope... more Abstract A detailed three-dimensional MCNP model of the ETRR-2 research reactor has been developed for the analysis of the neutronic parameters of the reactor. The model was used to determine the partial control plates reactivity worth and the behavior of the worth with time. The results of the present model were compared with the experimental measurements for partial control plates reactivity worth and with the design calculations for control plates behavior with time. The comparisons indicate satisfactory agreement.
Neutronic and burn-up calculations of heterogeneous Thorium/Uranium fuel in pressurized water reactors
Kerntechnik, Sep 1, 2010
Abstract The purpose of this work is to study the feasibility of the Thorium/Uranium fuel cycle i... more Abstract The purpose of this work is to study the feasibility of the Thorium/Uranium fuel cycle in heterogeneous Pressurized Water Reactors (PWR) core design. This paper focuses on the neutronic and burn-up analysis of the Thorium/Uranium fuel using the computer codes MCNPX and WIMS. The design is based on the Whole Assembly Seed and Blanket (WASB) concept, in which the individual seed (Uranium) and blanket (Thorium-Uranium) units occupy one full-size PWR assembly in a checkerboard core configuration. The results of the present models were compared with the solution of benchmark problems and satisfactory agreement was found.
Calculation of the pin power distribution for a thorium reactor assembly and benchmarking
Kerntechnik, Mar 1, 2007
A computer model was developed to perform neutronic and burn-up analysis for an assembly of a tho... more A computer model was developed to perform neutronic and burn-up analysis for an assembly of a thorium reactor. The MCNP computer code was used to model the geometry of the assembly and to determine both, the power and flux distribution. A system of ordinary differential equations which represents all fuel isotopes was solved numerically to evaluate the time behavior of fuel composition and burn-up. The results of the present model were compared with the solutions of benchmark problems and satisfactory agreement was found.
Neutronic calculations for the new fuel configuration of the ETRR-1 research reactor
Kerntechnik, Nov 1, 2005
Abstract Neutronic calculations were performed for the new loading configuration of the ETRR-1 re... more Abstract Neutronic calculations were performed for the new loading configuration of the ETRR-1 research reactor. The MCNP three dimensions Monte Carlo code and the two dimensions CITATION code are used to model the reactor. The power and thermal flux distributions in the reactor core are calculated. The power peak factor and the effect of control rod insertion on both flux and power profiles in the reactor core are determined and analyzed. The partial and total control rods worth are calculated. It was found that the difference between MCNP and CITATION in power distributions is 4 to 8 % and for thermal flux ranges between 3 to 14 %.
Comparative analysis between homogeneous and heterogeneous models of gas cooled fast reactor core (GFR-2400)
Kerntechnik
The main objective of the presented work is to investigate the effectiveness of homogenization of... more The main objective of the presented work is to investigate the effectiveness of homogenization of the GFR-2400 core in simulating its neutronic characteristics. To explore and evaluate the neutronic behavior of the large scale Gas cooled Fast Reactor GFR-2400 core, two computer models (homogeneous and heterogeneous) were designed using the MCNPX code. The designed heterogeneous model has been validated by comparing its results with a previously published paper. The results of both models were compared with each other to study the effect of fuel homogeneity on the radial flux and power distribution through the core. As part of a safety analysis of the reactor core for both designs, the reactivity worth of control rods, Effective delayed neutron fraction (ß eff), Doppler constant and Depressurization effect have been analyzed and compared. The variations of the effective multiplication factor (k eff), the minor actinides concentration and the most important fission products as a funct...
Comparative neutronic study for heterogeneous and homogeneous fuel assembly in a lead-cooled fast reactor
IOP Conference Series: Materials Science and Engineering, 2021
ALFRED (Advanced Lead-cooled Fast Reactor European Demonstrator) is a (300 MW th ) pool-type reac... more ALFRED (Advanced Lead-cooled Fast Reactor European Demonstrator) is a (300 MW th ) pool-type reactor with closed hexagonal Fuel Assemblies (FAs) which are divided into two radial zones (inner and outer core zones). In this work two three dimensional models (3D-heterogenous and 3D-homogeneous) for inner assembly (IA) and outer assembly (OA) have been designed using MCNPX transport code to simulate the neutronic behavior and to study the fuel performance and the effect of homogenization inside the fuel assemblies. The results of the present work for power distribution and fuel burn-up show good agreement between 3D-homogeneous and 3D-heterogeneous models.
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Papers by Moustafa Aziz