Patent Number: 048760579
Section: summary

DESCRIPTION The present invention relates to a control process for a nuclear reactor comprising controlling the neutron flux and the power distribution in the reactor core. This control comprises a stage of determining the neutron flux and the power distribution and a regulation stage through the displacement of control rods in the reactor core and/or modification of the boron concentration in the reactor primary circuit, as a function of the determined neutron flux and power distribution. The control can in particular be carried out as a function of the result of the comparison between the neutron flux and the power distribution as determined and desired values for said neutron flux and said power distribution, said desired values representing limit values or the safety criteria of the reactor. The process according to the invention makes it possible to very rapidly obtain neutron flux and power distribution values. This permits on-line applications, such as: detection of abnormalities in connection with the position of the control bars and the boron concentration, simulation of the behaviour of the core in the transient mode and investigating an optimum core control strategy, assisting the monitoring of the physical state of the core and interpretation of measurements on the core, e.g. for reactivity measurements. This property can also be used in out-of-line applications, such as: core design calculations, analysis of normal or abnormal transients, detailed calculations of the evolution of isotopic compositions of the fuel as a function of the irradiation level, calculations of the recharging of fuel assemblies into the cores of operating reactors, new core control studies, studies of new core control strategies. In general terms, the nuclear reactor core comprises assemblies constituted by fissile materials, with which the neutrons react in accordance with the following main processes: impact on nuclei of fissile materials with modification of the direction and velocity of the neutrons undergoing the shock, trapping by a nucleus, without emission of new neutrons, absorption by a nucleus, with emission of new neutrons. Fission is the preponderant phenomenon in nuclear reactors and is associated with the giving off of a large amount of energy. In the time interval separating two interaction events, the displacement of the neutrons is rectilinear and uniform. One state of the core can be characterized by the properties of the fissile materials contained in it, the distribution of the neutrons in the core and their interaction with the nuclei of fissile materials. Thus, one state of the core is defined by a number of parameters and more specifically by the value of each of these parameters at each point of the core. In practice, the state of the core is defined by the value of each parameter in a plurality of zones, meshes or networks, all of which represent the core volume. The meshes are preferably identical and can e.g. consist of cubic volumes. In this case, for a 900 MWe pressurized water nuclear reactor, the core e.g. corresponds to a total of approximately 22,000 cubic meshes, each having a side length of approximately 10 cm. Throughout the remainder of the text the total number of meshes is designated J and the mesh index j (1.ltoreq.j.ltoreq.J). For the practical determination of the power and flux distribution in the nuclear reactor core, the range of neutron velocities is conventionally broken down into finite intervals, designated by the index g, in which g varies from 1 to G, G being the total number of velocity groups. The magnitudes describing the distribution of the neutrons in the core are then: the neutron density n(j,g), which represents the number of neutrons per volume unit which, in the mesh j, are in the velocity group g. the neutron flux .phi.(j,g) in the mesh j which is, by definition, the product of the neutron density n(j,g) by the mean velocity of the neutrons of the velocity group g. The magnitudes describing the core characteristics with respect to the interaction with the neutrons (hereinafter called interaction probabilities) are: the effective differential macroscopic diffusion section .SIGMA.s(j,g.fwdarw.g') equal to Rs(j,g.fwdarw.g')/.phi.(j,g), in which Rs(j,g.fwdarw.g') is the number of neutrons of the velocity group g undergoing, per time and volume unit, a shock in the mesh j and being diffused, after the shock, with a new velocity in the velocity group g', the total effective macroscopic diffusion section .SIGMA.s(j,g) defined by ##EQU1## and equal to Rs(j,g)/.phi.(j,g), in which Rs(j,g) is the number of neutrons of the velocity group g undergoing a shock per volume and time unit, the effective macroscopic trapping section .SIGMA.s(j,g) equal to Rc(j,g)/.phi.(j,g), in which Rc(j,g) represents the number of neutrons of the velocity group g absorbed per time and volume unit without reemission of new neutrons, the effective macroscopic fission section .SIGMA.s(j,g) equal to Rf(j,g)/.phi.(j,g), in which Rf(j,g) represents the number of fissions, per time and volume unit, induced by the neutrons of the velocity group: during a fission, there is an emission of on average new neutrons and a giving off of approximately 200 MeV of energy by fission: subsequently it will be considered for simplification purposes that the ratio K/.nu. is constant, the effective macroscopic absorption section .SIGMA.a(j,g) in group g equal to .SIGMA.c(j,g)+.SIGMA.(j,g), the total effective macroscopic section .SIGMA.t(j,g) in group g equal to .SIGMA.a(j,g)+.SIGMA.s(j,g). The total number S(j) of neutrons emitted by the mesh j, also called neutron source, per time and volume unit is defined by: ##EQU2## In the same way, the power P(j) emitted by the mesh j is defined by P(j)=(K/.nu.).S(j) (if it is considered that the ratio K/.nu. is constant). It is known that the knowledge of the parameters .SIGMA.s(j,g.fwdarw.g'), .SIGMA.a(j,g), .SIGMA.t(j,g) and .nu.. .SIGMA.f(j,g) in each mesh j, 1.ltoreq.j.ltoreq.J makes it possible to determine the value of the quantity .phi.(j,g), S(j) and P(j). A first known process consists of calculating the neutron flux .phi.(j,g) in each mesh j by a method with finite differences. This process is described in the works "Nuclear Reactor Theory" by George Bell and Samuel Glasstone, Van Nostrand Reinhold Company and "Introduction a l'analyse numerique matricielle et a l'optimisation" by P. G. Ciarlet, published by Masson. A second known process consists of calculating the neutron flux .phi.(j,g) in each mesh j by a nodal method. These two processes are based on iterative methods, whereof the main calculation stage consists of reversing a matrix of size J.times.J, in which J represents the number of meshes and is equal to a few tens of thousands. It is clear that this calculation requires powerful processing means making it costly. In order to reduce the costs, it is possible to use less powerful processing means, but it is then necessary to accept a longer calculation time for determining the neutron flux and power in the reactor core, which is not satisfactory in the case of on-line monitoring of said core. The object of the invention is to permit a real time control of the reactor core. This is achieved by a rapid determination method for the neutron flux and the power without any need for reversing a large size matrix. In general terms, the invention consists of determining the neutron flux in the form of a sum between a first neutron flux component, associated with a predetermined core state, and a second neutron flux component, associated with the real state of the core. More specifically, the invention relates to a process for the control of a nuclear reactor having a core containing fissile material assemblies, the fission of the nuclei of said materials being brought about by interactions with neutrons and producing in turn neutrons, said nuclear reactor also having a means for regulating the neutron flux in the reactor core and the power given off by said reactor core, said regulating means comprising at least one assembly of control rods displaceable in said reactor core, said process being characterized in that: (A) iteratively there is a determination of the neutron fluxes .phi.(j,g) for each zone or mesh j 1.ltoreq.j.ltoreq.J of a group of meshes corresponding to the volume of the reactor core and for each velocity group g 1.ltoreq.g.ltoreq.G of a plurality of velocity groups for the neutrons; and the numbers of neutrons or sources S(j) emitted by each mesh j, 1.ltoreq.j.ltoreq.J, per volume and time unit, the determination of said values consisting of repeating the following sequence of operations until said values converge: PA0 (B) the regulating means is controlled as a function of the neutron fluxes .phi.(j,g) and powers P(j). (a) calculation of the first components of the neutron flux .phi..sup.0 (j,g), 1.ltoreq.j.ltoreq.J as a function of the predetermined coupling matrixes [.psi.g] 1.ltoreq.g.ltoreq.G and sources S(j), each element .psi.g(j,k), 1.ltoreq.k.ltoreq.K and 1.ltoreq.j.ltoreq.J expressing for neutrons of the velocity group g 1.ltoreq.g.ltoreq.G, the coupling between the mesh j and the adjacent meshes k corresponding to predetermined interaction probabilities of the neutrons with the fissile materials of the core for each mesh j, 1.ltoreq.j.ltoreq.J, PA1 (b) calculation of the real interaction probabilities of the neutrons in the core as a function of physical parameters describing the state of the core, PA1 (c) calculation of second neutron flux components .phi..sup.1 (j,g), 1.ltoreq.j.ltoreq.J as a function of the neutron fluxes .phi..sup.0 (j,g) of the predetermined interaction probabilities of the neutrons in the core and the real interaction probabilities of the neutrons in the core, PA1 (d) calculation of the neutron fluxes .phi.(j,g) in each mesh j, as the sum of the first and second neutron flux components, PA1 (e) calculation of the new sources NS(j), 1.ltoreq.j.ltoreq.J associated with each neutron flux .phi.(j,g) and new values of sources S(j) by the relation: ##EQU3## (f) calculation of the powers P(j) emitted in each mesh j as a function of the total power P emitted and the sources S(j), In preferred manner, the interaction probability of the neutrons in the core is defined by all the following parameters for each mesh j, 1.ltoreq.j.ltoreq.J: effective macroscopic diffusion sections .SIGMA.s(j,g.fwdarw.g') of the neutrons of the velocity group g, 1.ltoreq.g.ltoreq.G in the velocity group g', g'.noteq.g, effective macroscopic absorption section .SIGMA.s(j,g) of the neutrons of velocity g, 1.ltoreq.g.ltoreq.G, effective macroscopic fission section .SIGMA.f(j,g) of the neutrons of velocity g, 1.ltoreq.g.ltoreq.G. In the same way, the predetermined interaction probabilities of the neutrons in the core can be defined for each mesh j, 1.ltoreq.j.ltoreq.J and each velocity group, g, 1.ltoreq.g.ltoreq.G by: effective macroscopic diffusion sections .SIGMA.s.sup.0 (j,g g'), whose values are the possible values for said core, an effective macroscopic absorption section .SIGMA.a.sup.0 (j,g), whose value is a possible value for said core, an effective macroscopic fission section .SIGMA.f.sup.0 (j,g) of value equal to zero. In preferred manner, for each mesh j, 1.ltoreq.j.ltoreq.J, all the second neutron flux components, .phi..sup.1 (j,g), 1.ltoreq.g.ltoreq.G is determined as the solution of the system with G linear equations: