Computational modeling of monoenergetic neutral particle inverse transport problems in slab geometry
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R. R. Gomes
Abstract
An analytical numerical method applied to three different types of monoenergetic neutral particle inverse transport problems in the discrete ordinates (SN) formulation is presented: (a) boundary condition estimation; (b) interior source estimation; and (c) effective slab length estimation. These three types of inverse problems governed by the linear integrodifferential transport equation in SN formulation are related respectively to medical physics; nuclear waste storage; and non-destructive testing in industry. Numerical results and a brief discussion are given to conclude the paper.
Kurzfassung
Eine analytische numerische Methode zur Lösung inverser Transportprobleme für drei verschiedene Arten neutraler monoenergetischer Teilchen wird in Form diskreter Ordinaten (SN) vorgestellt: (a) Näherungswerte mit Randbedingungen; (b) Näherungswerte innenliegender Quellen; und (c) Näherungswerte für effektive Stablängen. Diese drei Arten inverser Probleme, die durch die lineare integrodifferentiale Transportgleichung in SN Form bestimmt werden, gilt es zu lösen in den Bereichen Medizinphysik (a); Lagerung radioaktiver Abfälle (b); und zerstörungsfreie Materialprüfung in der Industrie (c). Numerische Ergebnisse und eine kurze Diskussion beschliessen den Beitrag.
References
1 Lewis, E. E.; Miller, W. F.: Computational Methods of Neutron Transport. American Nuclear Society, Illinois, 1993Search in Google Scholar
2 Barros, R. C.; Larsen, E. W.: A Numerical Method for One-Group Slab-Geometry Discrete Ordinates Problems with No Spatial Truncation Error. Nuclear Science and Engineering104 (1990) 199–208Search in Google Scholar
3 Burden, R. L.; Faires, J. D.: Numerical Analysis. 8. ed, New York, Dover, 1993Search in Google Scholar
4 Duderstadt, J. J.; HamiltonL.J.: Nuclear Reactor Analysis. New York: John Wiley & Son., 1976Search in Google Scholar
5 McCormick, N. J.: Inverse radiative transfer problems: a review. Nuclear Science and Engineering112 (1992) 185Search in Google Scholar
© 2013, Carl Hanser Verlag, München
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- Computational modeling of monoenergetic neutral particle inverse transport problems in slab geometry
- Effects on criticality of selected scattering phase functions in neutron transport equation using the Chebyshev approximation
- U1 and P1 approximations to neutron transport equation for diffusion length calculation
- Technical Notes/Technische Mitteilungen
- TN approximation for the critical size of one-speed neutrons in a slab with anisotropic scattering
- Albedo and constant source problems for extremely anisotropic scattering
Articles in the same Issue
- Contents/Inhalt
- Contents
- Summaries/Kurzfassungen
- Summaries
- Technical Contributions/Fachbeiträge
- Chemistry aspects of the source term formation for a severe accident in a CANDU type reactor
- Analytical study on degraded core quenching
- Experimental investigations on control of flow instability in single-phase natural circulation loop
- Burnup calculations using serpent code in accelerator driven thorium reactors
- Investigation of neutronic effects in structural material of a hybrid reactor by using the MCNPX Monte Carlo transport code
- Nuclear aspects and cyclotron production of the positron emitter 55Co
- Calculation of age-dependent effective doses for external exposure using the MCNP code
- Effect of Cu2+/Al3+ mole ratio on structure of Cu – Al bimetallic nanoparticles prepared by radiation induced method
- A numerical method for resonance integral calculations
- Computational modeling of monoenergetic neutral particle inverse transport problems in slab geometry
- Effects on criticality of selected scattering phase functions in neutron transport equation using the Chebyshev approximation
- U1 and P1 approximations to neutron transport equation for diffusion length calculation
- Technical Notes/Technische Mitteilungen
- TN approximation for the critical size of one-speed neutrons in a slab with anisotropic scattering
- Albedo and constant source problems for extremely anisotropic scattering
