Obtaining the neutronic and thermal hydraulic parameters of the VVER-1000 Bushehr nuclear reactor core by coupling nuclear codes
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Mohsen Rafiei Karahroudi
Abstract
In this research, the simulation of one-sixth of the VVER-1000 reactor core is carried out by the WIMS-D4 nuclear code, based on the symmetry of the core and information obtained from the Final Safety Analysis Report. The atomic densities of some important nuclear materials and fission poisons are calculated by the WIMS-D4 code at the end of the first fuel cycle. In addition, the cross sections of some nuclides are obtained by WIMS-D4, and they are transferred into the CITATION code as inputs. In the next stage, neutron flux and reactor power are calculated by the CITATION code in Cold Zero Power and Hot Full Power status, and subsequently the heat flux of the core is obtained. Then the results are returned again into the extended program cycle. Finally, the heat flux of the core is inputted into the COBRA code, and the temperatures of fuel, clad and coolant are calculated along the various distances applying the COBRA thermal hydraulic code, through the results of the CITATION code and also initial data as default created from the Final Safety Analysis Report. In conclusion, there are some interesting outcomes resulting form the obtained results.
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© 2014, Carl Hanser Verlag, München
Artikel in diesem Heft
- Contents/Inhalt
- Contents
- Summaries/Kurzfassungen
- Summaries
- Technical Contributions/Fachbeiträge
- Evaluated gas production cross-section data for natural titanium irradiated with protons at energies up to 3 GeV
- Investigation of (n,γ) reaction in hybrid reactor zones
- Burnup analysis of the VVER-1000 reactor using thorium-based fuel
- Equivalent thermal conductivity of the storage basket with spent nuclear fuel of VVER-1000 reactors
- A study on empirical systematic for the (d,n) reaction cross sections at 8.6 MeV
- Recursive solutions for multi-group neutron kinetics diffusion equations in homogeneous three-dimensional rectangular domains with time dependent perturbations
- Thermoluminescent characteristics of nano-structure hydroxyapatite:Dy
- Performance evaluation of anion exchange resins Purolite NRW-5050 and Duolite A-611 by application of radioisotopic techniques
- A binary mixed integer coded genetic algorithm for multi-objective optimization of nuclear research reactor fuel reloading
- Development of an educational nuclear research reactor simulator
- Technical Notes/Technische Mitteilungen
- Obtaining the neutronic and thermal hydraulic parameters of the VVER-1000 Bushehr nuclear reactor core by coupling nuclear codes
Artikel in diesem Heft
- Contents/Inhalt
- Contents
- Summaries/Kurzfassungen
- Summaries
- Technical Contributions/Fachbeiträge
- Evaluated gas production cross-section data for natural titanium irradiated with protons at energies up to 3 GeV
- Investigation of (n,γ) reaction in hybrid reactor zones
- Burnup analysis of the VVER-1000 reactor using thorium-based fuel
- Equivalent thermal conductivity of the storage basket with spent nuclear fuel of VVER-1000 reactors
- A study on empirical systematic for the (d,n) reaction cross sections at 8.6 MeV
- Recursive solutions for multi-group neutron kinetics diffusion equations in homogeneous three-dimensional rectangular domains with time dependent perturbations
- Thermoluminescent characteristics of nano-structure hydroxyapatite:Dy
- Performance evaluation of anion exchange resins Purolite NRW-5050 and Duolite A-611 by application of radioisotopic techniques
- A binary mixed integer coded genetic algorithm for multi-objective optimization of nuclear research reactor fuel reloading
- Development of an educational nuclear research reactor simulator
- Technical Notes/Technische Mitteilungen
- Obtaining the neutronic and thermal hydraulic parameters of the VVER-1000 Bushehr nuclear reactor core by coupling nuclear codes
