Idioma:

Spent fuel characterization analysis using various nuclear data libraries

Čalič, Dušan ; Kromar, Marjan

Nuclear Engineering and Technology, 2022, 54(9), , pp.3260-3271 [Periódico revisado por pares]

Texto completo disponível

Citações Citado por

Título:
Spent fuel characterization analysis using various nuclear data libraries
Autor: Čalič, Dušan ; Kromar, Marjan
Assuntos: Decay heat ; Monte Carlo ; Nuclide uncertainties ; Serpent ; Spent nuclear fuel ; 원자력공학
É parte de: Nuclear Engineering and Technology, 2022, 54(9), , pp.3260-3271
Notas: KISTI1.1003/JNL.JAKO202229960617019
Descrição: Experience shows that the solution to waste management in any national programme is lengthy and burdened with uncertainties. There are several uncertainties that contribute to the costs associated with spent fuel management. In this work, we have analysed the impact of the current nuclear data on the isotopic composition of the spent fuel and consequently their influence on the main spent fuel observables such as decay heat, activity, neutron multiplication factor, and neutron and photon source terms. Nuclear libraries based on the most general nuclear data ENDF/B-VII.0, ENDF/B-VII.1, ENDF/B-VIII.0 and JEFF-3.3 are considered. A typical NPP Krško fuel assembly is analysed using the Monte Carlo code Serpent 2. The analysis considers burnup of up to 60 GWd/tU and cooling times of up to 100 years. The comparison of results showed significant differences, which should be taken into account when selecting the library and evaluating the uncertainty in determining the characteristics of the spent fuel.
Editor: Elsevier B.V
Idioma: Inglês;Coreano

Links

Voltar para lista de resultados