Burnup calculations using serpent code in accelerator driven thorium reactors
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M. E. Korkmaz
, M. Yiğit and O. Ağar
Abstract
In this study, burnup calculations have been performed for a sodium cooled Accelerator Driven Thorium Reactor (ADTR) using the Serpent 1.1.16 Monte Carlo code. The ADTR has been designed for burning minor actinides, mixed 232Th and mixed 233U fuels. A solid Pb-Bi spallation target in the center of the core is used and sodium as coolant. The system is designed for a heating power of 2000 MW and for an operation time of 600 days. For burnup calculations the Advanced Matrix Exponential Method CRAM (Chebyshev Rational Approximation Method) and different nuclear data libraries (ENDF7, JEF2.2, JEFF3.1.1) were used. The effective multiplication factor change from 0.93 to 0.97 for different nuclear data libraries during the reactor operation period.
Kurzfassung
In dieser Arbeit wurden Abbrandrechnungen für einen Natrium-gekühlten beschleunigergetriebenen Thorium-Reaktor (ADTR) mit Hilfe des Monte-Carlo Codes Serpent 1.1.16 durchgeführt. Der ADTR wurde ausgelegt für minore Aktinide, gemischte 232Th und 233U Brennstoffe. Im Zentrum des Kerns wurde ein Pb-Bi Target und als Kühlmittel Natrium verwendet. Das System wurde ausgelegt für eine Heizleistung von 2000 MW und eine Betriebszeit von 600 Tagen. Für die Abbrandberechnungen wurden die Advanced Matrix Exponential Method CRAM (Chebyshev Rational Approximation Method) und verschiedene Kerndatenbibliotheken (ENDF7, JEF2.2, JEFF3.1.1) verwendet. Der effektive Multiplikationsfaktor kann sich während der Betriebszeit von 0.93 auf 0.97 erhöhen je nach Kerndatenbibliothek.
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- Technical Notes/Technische Mitteilungen
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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