A study on the criticality of modified neutron transport equation by using alternative scattering phase functions
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A. Kara
Abstract
PN approximation is known as a proper method to solve neutron transport equation when literature is taken into consideration. Generally, conventional scattering function is used to solve criticality and diffusion problems in Legendre polynomial approximation. In this study, instead of conventional scattering function, Henyey-Greenstein (HG) and Anlı-Gungor phase functions (AG) are used in slab geometry transport equation and some critical thicknesses of the slab are calculated as an application with Legendre polynomial (PN) approximation and Marshak boundary condition. Results obtained from HG and AG scattering functions are compared and the correlations and discrepancies between the two functions are presented in the tables.
Kurzfassung
Die PN Approximation ist aus der Literatur als geeignete Methode zur Lösung der Neutronentransportgleichung bekannt. Im Allgemeinen werden konventionelle Streufunktionen zur Lösung von Kritikalitäts- und Diffusionsproblemen bei der Approximation mit Legendre Polynomen verwendet. In dieser Arbeit werden statt konventioneller Streufunktionen, Henyey-Greenstein (HG) und Anlı-Güngör Phasenfunktionen (AG) bei der Lösung der Transportgleichung für Stabgeometrie verwendet. Einige kritische Halbwertsdicken werden berechnet als Anwendung der Approximation mit Legendre Polynomen (PN) unter Marshak Randbedingungen. Die mit den HG und AG Streufunktionen erhaltenen Ergebnisse werden verglichen und Korrelationen und Abweichungen zwischen den beiden Fuintkionen in tabellarischer Form dargestellt.
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Artikel in diesem Heft
- Contents/Inhalt
- Contents
- Summaries/Kurzfassungen
- Summaries
- Technical Contributions/Fachbeiträge
- Evaluation of advanced displacement cross-sections for the major EUROFER constituents based on an atomistic modelling approach
- Tarapur atomic power station: analysis of station blackout scenario
- Estimating steady state and transient characteristics of molten salt natural circulation loop using CFD
- Thermal hydraulic analysis of ETRR-2 using RELAP5 code
- Transient performance during stopping the research reactor primary loop pump
- Neutronic performance calculations with alternative fluids in a hybrid reactor by using the Monte Carlo method
- A study on the criticality of modified neutron transport equation by using alternative scattering phase functions
- Production cross–section calculations of medical 32P, 117Sn, 153Sm and 186,188Re radionuclides used in bone pain palliation treatment
- Study on carbonated hydroxyapatite as a thermoluminescence dosimeter
- Social impact theory based modeling for security analysis in the nuclear fuel cycle
- Analysis of a homogenous and heterogeneous stylized half core of a CANDU reactor
