A burnup credit approach for irradiated fast-neutron reactor fuels nuclides of interest and fuel storage application - Archive ouverte HAL Access content directly
Conference Papers Year : 2015

A burnup credit approach for irradiated fast-neutron reactor fuels nuclides of interest and fuel storage application

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Abstract

The concept of taking into account the reduction of the reactivity of nuclear used fuel due to their burnup is referred to as Burnup Credit (BUC). Economic incentives offered by the implementation of a BUC methodology in criticality analyses of Light Water Reactor (LWR) used fuel facilities are nowadays well-demonstrated.The current prospect of reprocessing the fissile and fertile subassemblies of the PHENIX reactor in France, and the development of the 4th-Generation Fast-neutron Reactors (FR) enable research projects in relevant fuel cycle operations. Under these circumstances, the CEA and AREVA-NC have decided to study a burnup credit approach for irradiated fuels of the PHENIX reactor.This paper focuses on a preliminary investigation on the use of burnup credit for used FR fuel operations. The first step consists in analyzing and selecting the nuclides of interest for burnup calculation, on the basis of the 27 BUC nuclides (12 actinides and 15 fissions products) chosen for LWR-MOx fuels. These nuclides are then involved in a criticality calculation (a used fuel pool) to highlight the interest of a BUC approach. The opportunity to take account of only one fission product is highly promising and will considerably simplify the way to determine a penalizing fuel inventory. Moreover, it is to notice that results for different configurations may differ from those obtained for a fuel storage application. Thus, specific analyses have to be carried out in order to study the applicability of the conclusions of this paper.
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Dates and versions

cea-02492566 , version 1 (27-02-2020)

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  • HAL Id : cea-02492566 , version 1

Cite

C. Carmouze, C. Riffard, G. Grassi. A burnup credit approach for irradiated fast-neutron reactor fuels nuclides of interest and fuel storage application. ICNC 2015 - 9th International Conference on Nuclear Criticality Safety, Sep 2015, Charlotte, United States. ⟨cea-02492566⟩

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