In this thesis, a code that can be used to perform 3 dimensional neutronics analysis of nuclear reactor cores is developed. The code discretizes two group neutron diffusion equations using finite volume method for the geometry and diffusion parameters defined in the input file. The large sparse linear system obtained after discretization is solved effectively by using Krylov Subspace Methods. In order to verify the code developed, BEAVRS benchmark problem was modelled in the code. In order to model the problem, two group neutron diffusion parameters were generated for 165 different regions. The reaction rates, flux and current distributions obtained from the Serpent code, a genetic algorithm and the code developed were used to generate the ...
Criticality eigenvalue and power distributions of a medium-sized sodium-cooled fast reactor core wer...
This paper presents an assessment of three deterministic core simulators with the focus on the neutr...
Calculation of the neutron flux in a nuclear reactor core is ideally performed by solving the neutro...
Tez (Yüksek Lisans) -- İstanbul Teknik Üniversitesi, Enerji Enstitüsü, 2001Thesis (M.Sc.) -- İstanbu...
summary:We present a method for solving the equations of neutron transport with discretized energeti...
Diffusion approximation is an important approximation used to model a nuclear reactor core with a qu...
This research utilizes the MATLAB PDE Toolbox for modeling 3D static reactors with hexagonal fuel la...
Neutron transport calculations by Monte Carlo methods are finding increased application in nuclear r...
Nodal diffusion codes have been successfully used for decades as a primary tool of commercial power ...
The objective of this thesis is to introduce a new methodology for two{dimensional multi{ group neut...
Tez (Yüksek Lisans) -- İstanbul Teknik Üniversitesi, Bilişim Ensititüsü, 2003Thesis (M.Sc.) -- İstan...
In this work it is presented a solution of the stationary multi-layer multi-group neutron diffusion...
Tez (Doktora) -- İstanbul Teknik Üniversitesi, Enerji Enstitüsü, 2003Thesis (Ph.D.) -- İstanbul Tech...
[EN] The spatial distribution of the neutron flux within the core of nuclear reactors is a key facto...
International audienceThe Laboratory for Reactor Physics and Systems Behaviour at the PSI and the EP...
Criticality eigenvalue and power distributions of a medium-sized sodium-cooled fast reactor core wer...
This paper presents an assessment of three deterministic core simulators with the focus on the neutr...
Calculation of the neutron flux in a nuclear reactor core is ideally performed by solving the neutro...
Tez (Yüksek Lisans) -- İstanbul Teknik Üniversitesi, Enerji Enstitüsü, 2001Thesis (M.Sc.) -- İstanbu...
summary:We present a method for solving the equations of neutron transport with discretized energeti...
Diffusion approximation is an important approximation used to model a nuclear reactor core with a qu...
This research utilizes the MATLAB PDE Toolbox for modeling 3D static reactors with hexagonal fuel la...
Neutron transport calculations by Monte Carlo methods are finding increased application in nuclear r...
Nodal diffusion codes have been successfully used for decades as a primary tool of commercial power ...
The objective of this thesis is to introduce a new methodology for two{dimensional multi{ group neut...
Tez (Yüksek Lisans) -- İstanbul Teknik Üniversitesi, Bilişim Ensititüsü, 2003Thesis (M.Sc.) -- İstan...
In this work it is presented a solution of the stationary multi-layer multi-group neutron diffusion...
Tez (Doktora) -- İstanbul Teknik Üniversitesi, Enerji Enstitüsü, 2003Thesis (Ph.D.) -- İstanbul Tech...
[EN] The spatial distribution of the neutron flux within the core of nuclear reactors is a key facto...
International audienceThe Laboratory for Reactor Physics and Systems Behaviour at the PSI and the EP...
Criticality eigenvalue and power distributions of a medium-sized sodium-cooled fast reactor core wer...
This paper presents an assessment of three deterministic core simulators with the focus on the neutr...
Calculation of the neutron flux in a nuclear reactor core is ideally performed by solving the neutro...