[en] During a criticality accident, significant exposure is generated by 4 radiation origins: radiation directly induced by fissions (prompt neutrons and gamma), gamma radiations induced by (n, γ) reactions in crossed materials (capture gamma) and gamma radiations emitted by fission products. Due to boiling of the solution, a fraction of fissions products is airborne and is deposited in the ventilation shafts. 5.10¹⁸ fissions are considered in a dissolution tank containing uranyl nitrate by using the deterministic ATTILA radiation transport code. Instantaneous radiations rates are evaluated as a function of the distance and compared with data available in the literature. Dose rates induced behind various shielding materials such as concrete, steel or glass are assessed. In all cases, relative contributions of prompt or capture radiations is detailed. (author)

[en] In order to avoid criticality risks, a large number of facilities using spent fuels have been designed considering the fuel as fresh. This choice has obviously led to considerable safety margins. In the early 80's, a method was accepted by the French Safety Authorities allowing to consider the changes in the fuel composition during the depletion with some very pessimistic hypothesis: only actinides were considered and the amount of burnup used in the studies was equal to the mean burnup in the 50-least-irradiated centimeters. As many facilities still want to optimize their processes (e.g. transportation, storage, fuel reprocessing), the main companies involved in the French nuclear industry, researchers and IRSN set up a Working Group in order to study the way burnup could be taken into account in the criticality calculations, considering some fission products and a more realistic axial profile of burnup. The first of this article introduces the current French method used to take burnup into account in the criticality studies. The second part is devoted to the studies achieved by the Working Group to improve this method, especially concerning the consideration of the neutron absorption of some fission products and of an axial profile of burnup: for that purpose, some results are presented related to the steps of the process like the depletion calculations, the definition of an axial profile and the criticality calculation. In the third part, some results (keff) obtained with fission products and an axial profile are compared to those obtained with the current one. The conclusions presented are related to the present state of knowledge and may differ from the final conclusions of the Working Group. (author)