Patent Number: 043199593
Section: description

DESCRIPTION OF THE PREFERRED EMBODIMENT As diagrammatically shown in FIG. 1, in a boiling water type nuclear reactor, a plurality of fuel rod 2 (for brevity only one of them is shown) are regularly arranged in the form of a grid in a fuel assembly 1 of the reactor core i.e., a fuel bundle. Each fuel rod contains fissionable substance and it is now supposed that it generates a heat quantity of q Kcal/sec. per unit length. This heat quantity is transmitted to the coolant, in this case water, which flows through a fuel channel 4 via a sheath 3 whereby a two-phase flow consisting of water and steam is generated in the fuel channel. The coolant, or water flows into the fuel assembly 1 from the inlet plenum chamber, not shown, through the inlet opening 5 at the lower end of the fuel assembly. Then, the water flows into the fuel channel 4 through an inlet orifice 6 and forms the two-phase flow of water and steam, which is discharged into the outlet plenum chamber, not shown, through an outer orifice 7. As a plurality of fuel channels 4 are provided for the reactor core, the pressure differential .DELTA.p between the inlet and outlet plenum chambers is maintained at a definite value for all channels irrespective of some channel flow divergence in a specific fuel channel. From the view point of nuclear characteristics, if an oscillation occurs in a specific channel independently of the other channels, its effect on the stability of the surrounding channels due to nuclear feedback is negligibly small. The thermo-hydrodynamic oscillation of the two-phase flow in the fuel channels 4 in fuel assemblies 1 under these conditions is just equal to the stability of a heated two-phase flow loop having the pressure differential .DELTA.p between the inlet and outlet plenum chambers, and there are numerous experimental data regarding thereto. The limit of the channel stability is determined by such parameters as (1) the thermal power of each fuel channel and its distribution, (2) inlet subcooling, (3) the pressure in the fuel channel, (4) the hydrodynamic characteristic of the fuel channel including the inlet and outlet orifices, and (5) the inlet flow quantity. Accordingly, if it were possible to measure all of the five parameters described above regarding all fuel assemblies in a nuclear reactor under operation, it would be possible to determine the present stability margin in view of the stability limit described above. Considering the flow quantity, a flow quantity W* that gives the lower stability limit is determined by using all measured parameters other than the flow quantity. When the measured flow quantity is expressed by W, a ratio W/W* gives an evaluation for the stability margin. According to the prior art method, the power distribution and the flow quantity distribution of each fuel assembly of a boiling water type nuclear reactor have been determined by repeatedly calculating signals produced by neutron flux detectors disposed in the reactor core by taking into consideration the feedback effect between the thermal power, the flow quantity and the void effect. The following reference describes the detail of the prior art method. J. F. Grew "Process Computer Performance Evaluation Accuracy" NEDO-20340, June 1974, General Electric Co. Monta, et al have developed a method of directly determining the flow quantity in the fuel assembly by determining the propagation time of a disturbance along the flow by utilizing the mutual correlation function of the signals produced by a plurality of neutron flux detectors installed in the reactor as disclosed in Japanese Laid Open Patent Specification No. 139897 of 1977. FIG. 2 shows one embodiment of this invention based on a prior art method of calculating the flow quantity of the coolant. A computer 10 is connected to calculate the power distribution q.sub.i and the flow quantity distribution W.sub.i of the core by repeating a calculation regarding the relationship among the power distribution in the core, the void distribution, and the flow quantity distribution by using signals 11 produced by all neutron flux detectors in the core, and signals 12, the process variables indicating the inserted positions of control rods, the flow quantity of recirculation water and the thermal power of the core determined by the heat balance of the power plant. A channel stability supervisory device 13 is connected to receive a signal 14 representing the mean pressure of the channel and a signal 15 representing the subcooling of the recirculation water in addition to the signals representing the power distribution q.sub.i and the flow quantity distribution W.sub.i and functions to determine the channel stability in accordance with an equation to be described latter based on the degree of openings of the inlet and outlet orifices and other predetermined hydrodynamic data, for example two-phase flow friction factor of the fuel assemblies, or to determine the stability margin by comparing a calculated value of the stability limit based on certain designated parameters with the present value thereof. FIG. 3 shows one embodiment of this invention utilizing the above described method developed by the inventors. In this case, the mutual correlation function of the outputs of two neutron flux detectors 16 and 17 provided along the flow passage of the coolant for the purpose of measuring the flow quantity thereof is determined by a correlation meter 18 so as to obtain the propagation time .tau. of the disturbance between the two neutron flux detectors, thereby determining the flow quantity of the coolant in the fuel channels of the fuel assembly adjacent the neutron detectors. The output .tau. of the correlation meter 18, the outputs 11 of the neutron flux detectors and the process variables 12 are applied to computer 10. The connection and operation of the channel stability supervisory device 13 is the same as those of the supervisory device shown in FIG. 2. The detail of the operation of the channel stability supervisory device 13 will now be described. This device provides an information regarding the stability limit of the thermo-hydrodynamic oscillation of the two-phase flow of water and steam in the fuel channels of a boiling water type nuclear reactor and determines the channel stability periodically, or when requested, based on the present state. The information regarding the stability limit can be given by calculating the stability limit for the power distribution, flow quantity, inlet subcooling and pressure of a fuel channel by using stability analysis codes regarding a boiling water type nuclear reactor which correspond to STABLE, FABLE CODES developed by A. B. Jones, and by giving the calculated value in the form of an approximate function or a table, or can be given by an experimental equation. For example, according to Ishii, the equilibrium phase change number N.sub.pck.eg and subcool number N.sub.sub are defined respectively by the following equations. ##EQU1## where A.sub.c : sectional area of the channel flow per fuel rod, l: length of the heated channel portion, gw: heat flow flux of the wall of the heat conduction tube, V.sub.fi : flow velocity at the channel inlet, .rho..sub.g : steam density, .rho..sub.f : density of boiling water, .DELTA..sub.92 : .DELTA..sub.f -.rho..sub.g, .DELTA..sub.ifg : latent heat of evaporation, .DELTA..sub.isub : inlet subcooling, .xi..sub.h : peripheral length of the heated portion. The approximate stability limit line for a large subcooling is given by the following equation. ##EQU2## where K.sub.i : inlet orifice coefficient, K.sub.e : outlet orifice coefficient, f.sub.m : friction coefficient of two-phase flow, D.sub.h *: dimensionless hydraulic diameter=Dh/l (Dh: hydraulic diameter). In equation (3), the region in which the lefthand term is larger than the right-hand term represents an unstable region. FIG. 4 is a graph showing the stability in a phase plane in which the ordinate represents the subcool number N.sub.sub and the abscissa the equilibrium phase change number N.sub.pch.eq Region A to the left of the stability limit border line 20 expressed by equation 3 shows stable state whereas region B on the opposite side unstable state. Regarding the detail of the graph reference is made to M. Ishii's paper "Thermally Induced Instabilities in Two-phase Mixtures in Thermal Equilibrium", PhD Thesis, School of Mechanical Engineering, Georgia Institute of Technology, Atlanta, Georgia, June, 1971. Saha has proposed an approximate stability limit border line which holds even for smaller subcool numbers by expanding Ishii's equation for judging the stability as shown by the following paper. P. Saha, M. Ishii and N. Zuber, "An Experimental Investigation of the Thermally Induced Flow Oscillation in Two-phase Systems" ASME Winter Annual Meeting, 1975. Since the right-hand side of equation (3) is expressed in terms of the data of the hydrodynamic characteristic of each fuel channel, the stability can be judged by using measured values of the inlet flow velocity V.sub.fi, inlet subcooling .DELTA..sub.isub, heat flux g.sub.w and the mean pressure which are the parameters for determining N.sub.pck.eq and N.sub.sub on the left-hand side. Furthermore, when a predetermined quantity is evaluated to hold equation 3, for example V.sub.fi * (that is at the stability limit border line) it is possible to qualitatively evaluate the stability by comparing the limit value V.sub.fi * and the present value V.sub.fi of the inlet flow velocity. As shown in FIG. 5 a channel stability protective device 22 is added to receive stability margin S produced by the channel stability supervisory device 13 for comparing the stability margin S with a predetermined reference value S*. When the stability margin is smaller than the reference value, an instruction for inserting control rods near the fuel channel is produced by the protective device 22. Thus, as the thermal power generated by a given channel decreases, the channel stability margin is recovered as can be noted from equation 3 thereby stabilizing the operation of the nuclear reactor. It is well known that the computer 10 for calculating the core power distribution and the flow quantity distribution, and the channel stability supervisory device can be realized by programming in process computers. The judgment of the stability can be made by any other method other than those described hereinabove without departing from the scope of this invention. For example, an decay ratio may be used as the stability allowance. Since the actual operating condition of a nuclear reactor varies in a complicated manner, as the degree of the burning of the fuel, the insertion pattern of control rods, the reactor power and other parameters vary, when one tries to ensure a desired stability for all cases, the operating range would be limited. In contrast, according to this invention, since the stability of respective fuel channels are supervised in accordance with operating condition of the core it is possible to avoid excessively large stability margin and hence to widen the operating range, thereby enabling more flexible operation than is possible by the prior art method. In addition, according to this invention, as it is possible to detect critical core regions having a small channel stability margin it is possible to increase the stability margin by changing the control rods pattern in such and nearby regions without decreasing the overall power of the reactor, this also contributing to the widening of the stable operating range. Furthermore, according to this invention, when the stability allowance of any fuel channel decreases below a predetermined value it is possible to automatically recover the stability allowance by inserting rapidly or slowly a limited number of control rods near said fuel channel thereby preventing generation of channel instability which otherwise would result in scram of the reactor.