Areva Np Gmbh

Framatome Anp Gmbh

Hubert Salaün

Jérémy Galpin

Elodie Mery De Montigny

Nicolas Goreaud

Benjamin Chazot

Michel Bonnamour

Jacques Gauthier

Bruno Bonnamour

A method for simulating the fluid flow in a vessel of a nuclear reactor is provided. The reactor includes a core inside the vessel, the core including a lower plate, an upper plate and fuel assemblies extending between the plates, and having a volume axially delimited by first and second interfaces corresponding to the plates. The method includes computing, for the core volume, the fluid pressure and speed, from an initial value of the speed or pressure in the first interface and respectively in the second interface, and using the fluid mass, movement quantity and energy balance equations. The method includes determining an additional volume inside the vessel, the additional volume being outside the core volume, axially at one end thereof and axially delimited by two interfaces, one of which is the first or second interface; and the computing, for the additional volume and using the equations, the pressure and speed, from an initial value of the speed or pressure in each of the interfaces of the additional volume.

Methods for simulating the flow of a fluid in a vessel of a nuclear reactor and for calculating the mechanical deformation of assemblies of a nuclear reactor core, and associated computer program products

A method for simulating the flow of a fluid in a vessel of a nuclear reactor is provided. The nuclear reactor includes the vessel and a core inside the vessel, the core including nuclear fuel assemblies, each one extending in an axial direction, including nuclear fuel rods and a grid for maintaining the rods, and being spaced apart from another by a clearance between the grids in a transverse direction. This method for simulating a fluid flow in the vessel of a nuclear reactor includes determining of head loss coefficients in the core, and computing the fluid pressure and speed component(s) in the core using the equation: &#x2207;P=&#x2212;K&#xd7;V where P is the component of the fluid pressure, K is a matrix including the determined head loss coefficients, and V is a vector including the fluid speed component(s). A transverse head loss coefficient in the assemblies is determined from a transverse Reynolds number, and an axial head loss coefficient in the clearance is from the dimension of the clearance in the transverse direction.

A spacer grid for a fuel assembly in a light-water-cooled nuclear reactor, for providing the transverse retention of a bundle of fuel rods in mutually parallel arrangements, having an array of cells juxtaposed and placed in a regular lattice, each bounded and separated from neighboring cells by at least one peripheral wall which is open at two opposed ends, along the direction of an axis of the cell, so as to receive a fuel rod of cylindrical general shape passing along the cell along its axis parallel to the peripheral wall.

Spacer grid for a fuel unit in a nuclear reactor cooled by light water

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Lyons, France

Hubert Salaün