Calculation of the pin power distribution for a thorium reactor assembly and benchmarking
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M. Aziz
Abstract
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.
Kurzfassung
Ein Computermodell wurde entwickelt um Neutronenberechnungen und eine Analyse des Abbrands für die Brennelementanordnung eines Thoriumreaktors durchzuführen. Der Monte Carlo Code MCNP wurde verwendet zur Modellierung der Geometrie der Anordnung und zur Bestimmung der Leistungs- und Flussverteilung. Ein System gewöhnlicher Differentialgleichungen, das alle Brennelementisotope repräsentiert, wurde numerisch gelöst, um das Zeitverhalten der Brennelementzusammensetzung und den Abband zu bestimmen. Die Ergebnisse dieses Modells wurden verglichen mit den Ergebnissen von Benchmarkproblemen. Dabei wurde eine befriedigende Übereinstimmung gefunden.
References
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© 2007, Carl Hanser Verlag, München
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Articles in the same Issue
- Contents/Inhalt
- Contents
- Summaries/Kurzfassungen
- Summaries
- Technical Contributions/Fachbeiträge
- Wavelet techniques for the determination of the decay ratio in boiling water reactors
- Analytical and experimental investigations of shear stress in rod bundles with irregular cells
- Incineration of weapon grade plutoniumin a (DT) fusion driven hybrid reactor using various coolants
- Calculation of the pin power distribution for a thorium reactor assembly and benchmarking
- Comparative assessment of methods for the reactivity measurement in subcritical systems by pulsed experiments
- Thermal-hydraulic modeling of reactivity accidents in MTR reactors
- Prediction of the onset of flow instability in the ETRR-2 research reactor under loss of flow accident
- Time-dependent albedo problem for quadratic anisotropic scattering
- HN solutions of the time dependent linear neutron transport equation for a slab and a sphere
- Application of the UN method to the reflected critical slab problem for one-speed neutrons with forward and backward scattering
- The effects of different expansions of the exit distribution on the extrapolation length for linearly anisotropic scattering
- Technical Note
- Shadowing the earth from Lagrange Point L1
