Patent Number: 046541861
Section: summary

FIELD OF THE INVENTION The present invention relates to the determination of the power of a pressurized water nuclear reactor which comprises at least one cooling loop and preferably three or four loops. It is important to know the power of a nuclear reactor at any time, particularly for safety reasons. PRIOR ART The present practice is to employ neutron detectors installed outside the reactor vessel for measuring the power. Such detectors give signals with a fast response but have the disadvantage of giving signals of low precision, particularly during periods of transient operation. To obtain more precise results, the reactor power can be computed from a heat balance calculated from measurements of the temperature of the primary fluid in the cold branch and in the hot branch of the primary loops. However, while the results obtained in this way are precise, they are nevertheless slow because of the high time constant of the temperature variations. Devices which permit the power of a nuclear reactor to be correctly determined in respect of both precision and speed are known, e.g. from, French Pat. No. 2,373,057, which describes an apparatus permitting the reactor power to be determined from the increase in core enthalpy, this enthalpy increase being computed by making use of the velocity of sound in the fluid in the hot and cold branches of the primary circuit. However, this apparatus involves installation of sensors on the primary circuit pipework, the major disadvantage being the increase in the number of connections to the said pipework and the resulting increase in the difficulty of lagging it. U.S. Pat. No. 3,752,735 describes a device which makes it possible to produce a signal representing the thermal power of the core by using the measurement of hot branch temperature and cold branch temperature and compensating these temperature measurements dynamically, following a formula which involves the time derivative of the difference between the hot branch temperature and the cold branch temperature. Such a device permits a fast response signal to be obtained, but the formula which is employed to obtain this signal is only approximate and renders the signal somewhat inaccurate. French Patent Application No. 2,416,531 describes a process for determining the power of a nuclear reactor in which a thermal power signal which has a high time constant but is relatively precise is combined with a neutron power signal which has a fast response but is less precise. Nevertheless this process does not produce very precise results in a transient regime, because it merely adjusts the thermal power signal by means of a neutron power signal which is delayed by a gain control unit 6 and an integrating unit 7 (FIG. 1). It does not take into account numerous data which can vary in a transient regime. Furthermore, to date there has been no known simple device enabling both the primary power and the secondary power of a nuclear reactor to be obtained rapidly and precisely both in steady state operation and in transient operation. SUMMARY OF THE INVENTION The object of the present invention is to remedy the disadvantages of the aforesaid processes and devices. It relates to a device for fast and precise determination of the power of a pressurized water nuclear reactor having one or more cooling loops, in steady state operation and during periods of transient operation. Moreover, in a preferred embodiment, it also makes it possible to obtain, in an equally simple, fast and precise manner, the secondary power, i.e., the power of the steam generators in each cooling loop of the reactor. According to the invention the device comprises, for each of its cooling loops: means for measuring, on the one hand, the neutron flux and, on the other hand, the temperature of the primary fluid at a point in the cold branch and at a point in the hot branch, means for computing the enthalpy of the primary fluid in the cold branch and in the hot branch from the said temperature measurements, a register which computes the enthalpy increase of the primary fluid as it crosses the core, from the difference between the enthalpy in the hot branch and the enthalpy in the cold branch delayed by a time shift operator expressing the average time of transit of a molecule of fluid between the temperature measurements points in the cold branch and in the hot branch, a multiplier of the increase in the enthalpy by the flow rate of the primary fluid, a comparator of the thermal power signal which is obtained and of the neutron power signal which is measured and made dynamically equivalent to the thermal power signal, and a corrector of the neutron power measurement signal depending on the signal produced by the said comparator. In a preferred embodiment of the invention, the neutron power measurement signal is made dynamically equivalent to the thermal power signal by means of a point model of heat transfer between the nuclear flux and the thermal flux to the primary fluid, of a point model of heat transfer of the fluid in the core corresponding to the time of transit of a molecule of primary fluid from the center of the core, to the outlet of the core and a time shift operator expressing the time of transit of a molecule of primary fluid from the core outlet to the point of measurement of the hot branch temperature. The corrector of the neutron power measurement signal is preferably an integrator. Moreover, it is preferable to correct the neutron power measurement signal for variations of the temperature measured in the cold branch before this signal is compared to the thermal power signal. It is advantageous to supplement the device according to the invention with a filter of the primary flow rate signal which is situated upstream of the said multiplier in order to allow for variation of the average time of transit of a molecule of primary fluid from the cold branch to the hot branch depending on the primary flow rate. It is further preferred to supplement the device for fast and precise determination of the primary power of a nuclear reactor as described above, by a device for rapid and precise determination of the secondary power of the reactor, comprising, for each loop, in addition to the aforesaid means for measuring temperature and for computing the enthalpy in the cold branch and in the hot branch, imprecise but fast means for determining the thermal power produced by the steam generator associated with the reactor from measurements of the temperature and the flow rate of the feed water and of the pressure and the flow rate of the steam from the steam generator, a register computing the reduction in the enthalpy of the primary fluid as it crosses the steam generator from the difference between the enthalpy in the hot branch and in the cold branch, the enthalpy in the hot branch being delayed by a time shift operator expressing the time of transit of a molecule of fluid between the two temperature measurement points, a multiplier of the decrease of enthalpy by the flow rate of the primary fluid, a comparator of the signal, which is obtained at the multiplier output, of the thermal power absorbed by the steam generator to the said signal, as determined above, of the thermal power produced by the steam generator, this signal being made dynamically equivalent to the signal of thermal power absorbed by the steam generator, and a corrector of the signal of power produced by the steam generator depending on the signal produced by the said comparator. Preferably, the measurement signal of the thermal power produced by the steam generator is made dynamically equivalent to the signal of the thermal power absorbed by the steam generator by means of a point model of heat transfer between the secondary fluid and the primary fluid, a point model of heat transfer of the primary fluid in the steam generator corresponding to the time of transit of a molecule of primary fluid from the center of the steam generator to the outlet of the steam generator, and a time shift operator expressing the time of transit of a molecule of primary fluid from the outlet of the steam generator to the point of measurement of the cold branch temperature. It is preferred to use an integrator to correct the measurement signal of the thermal power which is absorbed by the steam generator. The dynamics of the measurements of the primary fluid temperature in the cold branch and the hot branch are furthermore compensated by a phase lead corrector which is placed at the output of the multiplier of the variation of the enthalpy of the primary fluid by the flow rate of the primary fluid .