Szczegóły publikacji
Opis bibliograficzny
Neutronic analysis of mixed thorium-uranium fuel bundle for CANDU reactors / M. OETTINGEN, K. SKOLIK, M. MALICKI // IOP Conference Series: Earth and Environmental Science ; ISSN 1755-1307. — 2019 — vol. 214 no. 1 art. no. 012024, s. 1–8. — Bibliogr. s. 7–8, Abstr. — Publikacja dostępna online od: 2019-01-24. — 2nd International Conference on the Sustainable Energy and Environmental Development : 14–17 November 2017, Krakow, Poland
Autorzy (3)
Dane bibliometryczne
| ID BaDAP | 119853 |
|---|---|
| Data dodania do BaDAP | 2019-02-12 |
| Tekst źródłowy | URL |
| DOI | 10.1088/1755-1315/214/1/012024 |
| Rok publikacji | 2019 |
| Typ publikacji | referat w czasopiśmie |
| Otwarty dostęp |  |
| Creative Commons | |
| Czasopismo/seria | IOP Conference Series: Earth and Environmental Science |
Abstract
The paper shows neutronic analysis of the CANDU (CANada Deuterium Uranium) nuclear reactor fuel channel with mixed thorium-uranium fuel bundles. The numerical model of the fuel channel was designed using The Monte Carlo Continuous Energy Burn-up Code – MCB developed at the AGH University of Science and Technology, Faculty of Energy and Fuels, Department of Nuclear Energy. The super-computer Prometheus available in the frame of the Pl-Grid Infrastructure at the Academic Computer Centre Cyfronet AGH was used for multi-scale calculations. The fuel bundles are composed of two clusters of fuel rods. The neutronic analysis considers detailed numerical simulation of neutron transport in fully heterogeneous geometry of the fuel channel. Moreover, burnup simulations were performed using Transmutation Trajectory Analysis method implemented in the MCB code. In the analysis we mainly consider time evolutions of neutron multiplication factor, fissile 233U, 235U and 239Pu and fertile 238U and 232Th. The simulations were performed for eight scenarios with various fuel composition.
