Patent Number: 043371183
Section: description

DESCRIPTION OF THE PREFERRED EMBODIMENT In FIG. 1, there is shown a nuclear power plant including a boiling-water reactor (BWR) 2 provided with a core, not shown, in a pressure vessel 4. A plurality of control rods, not shown, are selectively moved into and out of the core by a rod drive system 6. Contained in the pressure vessel 4 is a coolant (light water) which is recirculated through the core by recirculation pumps 8 which receive part of the coolant and forces it to flow into jet pumps within the pressure vessel so that the coolant flows upward through fuel assemblies in the core. The heat produced by the fuel assemblies is transferred to the coolant and a head of steam is produced in the upper portion of the pressure vessel 4. The steam is supplied to a turbine 10 which drives an electrical generator 12. The turbine 10 exhausts to a condenser 14 and the resulting condensate is returned as feedwater to the pressure vessel 4 through conduit means, not shown. Located on the discharge side of each recirculation pump 8 is a control valve 16 having its opening varied by a flow control system 18, to adjust the recirculation flow rate of coolant and thus control the core coolant flow rate. Alternatively, control of the core coolant flow rate may be effected by controlling the number of revolutions of the recirculation pumps 8. The flow control system 18 will be described in detail. A reactor power change demand signal is applied to a main controller 22 either manually or as a load speed deflection signal from a turbine control mechanism 20. A neutron flux controller 26 produces a flow rate demand signal as a function of the difference between an output signal of the main controller 22 and a detected value signal of a neutron monitoring system 24. A flow rate controller 30 supplies a signal through a function generator 32 to hydraulic control means 34 for the control valves 16 so as to bring the difference between the output signal of the neutron flux controller 26 and the detected value signal of a recirculation flow rate measurement system 28 to nil. The openings of the control valves 16 are adjusted by this signal to thereby control the recirculation flow rate and thus the core flow rate to a demanded level. Control is effected by a similar system when control of the recirculation flow rate is effected by adjusting the number of revolutions of the recirculation pump 8. 36 is a turbine bypass valve. One example of the operation plan of a BWR which may be practiced by controlling the core coolant flow rate by this flow control system 18 will be described by referring to FIG. 2. In FIG. 2, the abscissa represents the core flow rate, and the ordinate indicates the reactor power level. As aforesaid, although the core flow does not show the same rate as the recirculation flow, there is a uniform relation between them. Thus, it will be noted that the core flow rate and the recirculation flow rate can be substituted for each other, and the core flow rate can be detected by the recirculation flow rate measurement system 28. At initial stages of operation of the reactor, the reactor is operated at point A on a flow rate control curve 38. With the lapse of time, a reduction is caused in reactivity owing to fuel consumption, resulting in a fall in reactor power. To compensate for this reduction in reactor power, the core flow rate is increased to maintain the reactor power at a high level by utilizing a change in the manner in which voids are formed in the reactor. By gradually increasing the core flow rate in this way, it is possibel to maintain the reactor power at a desired level for about one to two months. After the core flow rate has reached 100% or when operation of the reactor is performed at point B, no further increase in flow rate is permissible, so that the core flow rate is temporarily reduced to move the reactor operating point to C at which reactor power is reduced. By changing the control rod insertion ratio at point C, the reactor operating point moves from C to D, and then returns to point A following an increase in the core flow rate. Thus the reactor is operated in a cycle lasting one to two months. In order to prevent an excessive rise in reactor power which might result from deviation of the operation of the reactor from the aforesaid plan during operation of the plant, there is provided a core monitoring system including an average power range monitor (APRM) 40, a thermal power monitor (TPM) 42 and a rod block monitor (RBM) 44, which receive signals from the neutron monitoring system 24 and recirculation flow rate measurement system 28 and transmits a rod block signal or scram signal to the rod drive system 6. The core monitoring system further includes an operating region monitor (ORM) 46 which is operative, when an excessive rise in reactor power is caused particularly by an increase in the core flow rate, to block the increase in the core flow rate or run-back the flow rate upon the power level reaching a predetermined threshold short of the scram threshold. More specifically, ORM 46 receives signals from the neutron monitoring system 24 and recirculation flow rate measurement system 28 and transmits a coolant block signal or run-back signal to the flow controller 32 of the flow control system 18 subsequently to be described. APRM 40 will now be described by referring to FIG. 3. APRM 40 includes an averaging circuit 48 for receiving signals from the neutron monitoring system 24 including a plurality of local power range monitors (LPRMs) and averaging these signals to produce the power level of the reactor. The signal from the averaging circuit 48 is transmitted to a comparator 50. Meanwhile a rod block threshold circuit 52 is set beforehand at a power level of rod block threshold as a function of the core coolant flow rate as shown at a line 54 in FIG. 2. The rod block threshold circuit 52 receives a signal from the recirculation flow rate measurement system 28 and transmits to the comparator 50 a threshold level signal corresponding to the prevailing flow rate. Upon receiving these signals from the two circuits 48 and 52, the comparator 50 compares them and transmits a comparison signal to a signal generator 55 which transmits, when the power level is higher than the threshold level, a rod block signal to the rod drive system 6. The signal from the averaging circuit 48 is also transmitted to another comparator 56 which also receives a signal from a scram threshold circuit 58. The scram threshold circuit 58 is set beforehand at a power level of scram threshold as shown at a line 60 in FIG. 2. The second comparator 56 compares the signals from the circuits 48 and 58 and transmits a comparison signal to the signal generator 55 which transmits, when the power level is higher than the threshold level, a scram signal to the rod drive system 6. Thus, APRM 40 monitors a rise in the power level of the reactor transmits a rod block signal to the drive system 6 when the power level has reached the rod block threshold line 54 shown in FIG. 2, to thereby block control rod withdrawing. For example, when the power level reaches about 106% of the rated power in a rated power operation, control rod withdrawing is blocked. Also, APRM 40 monitors the power level of the reactor which might be caused primarily by control rod withdrawing, an increase in flow rate and a rise in the pressure in the pressure vessel 4 caused by shutoff of the load or the like. When this power level reaches the scram threshold line 60 shown in FIG. 2, APRM 40 transmits a scram signal to the rod drive system 6 to scram the reactor. Scramming takes place when the power level reaches about 120% of the rated power, for example. TPM 42 will now be described by referring to FIG. 4. Like APRM 40, TPM 42 includes an averaging circuit 62 for receiving signals from LPRMs of the neutron monitoring system 24 and averaging local power levels to produce the power level of the reactor. The averaging circuit 62 supplies a signal to a time delay circuit 64 for conversion to a thermal power level. The delay circuit 64 transmits a signal to a comparator 66 to which a signal from a scram threshold circuit 68 is also supplied. The scram threshold circuit 68 is set beforehand at a power level of scram threshold as a function of the core coolant flow rate as indicated by a line 70 in FIG. 2, for example, and transmits to the comparator 66 a threshold level signal corresponding to the prevailing core coolant flow rate upon receipt of a signal from the recirculation flow rate measurement system 28. The comparator 66 compares these two signals from the circuits 64 and 68 and transmits a comparison signal to a signal generator 72 which transmits, when the thermal power level is higher than the threshold level, a scram signal to the rod drive system 6. Thus, TPM 42 monitors a rise in the thermal power level which might be cause primarily by control rod withdrawing and a rise in the flow rate, and supplies a scram signal to the rod drive system 6 when the thermal power level has reached the scram threshold line 70 shown in FIG. 2, thereby scramming the reactor. The reactor is scrammed in rated power operation when the thermal level reaches about 115% of the rated power, for example. RBM 44 will now be described by referring to FIG. 5. RBM 44 includes an LPRMs signal selecting circuit 74 for receiving signals from LPRMs of the neutron monitoring system 24 for selection of these signals. The circuit 74 supplies a signal to a comparator 76 to which a signal from a rod block threshold circuit 78 is also supplied. The rod block threshold circuit 78 is set at a power level of rod block threshold beforehand as a function of the core coolant flow rate and transmits to the comparator 76 a threshold level signal corresponding to the prevailing core coolant flow rate upon receipt of a signal from the recirculation flow rate measurement system 28. The power level of rod block threshold at which the circuit 78 is set is not shown in FIG. 2. However, the power level is generally below the line 54 by about 1-3%. The comparator 76 compares the signals from the two circuits 74 and 78 and transmits a comparison signal to a signal generator 80 which transmits, when the selected local power level is higher than the threshold level, a rod block signal to the rod drive system 6. Thus, RBM 44 monitors a rise in the local power level which might be caused by control rod withdrawing and transmits, when the local power level reaches the rod block threshold set beforehand, a rod block signal to the rod drive system 6 to block control rod withdrawing. The nuclear reactor continues its operation even if the control rod withdrawing is blocked by APRM 40 or RBM 44. It is possible to operate again the blocked control rods if other control rods are inserted or the core coolant flow rate is reduced to thereby reduce the power level. Last but not the least important is an operating region monitor (ORM) 46 which constitutes the characterizing feature of the present invention. Referring to FIG. 6, ORM 46 includes an averaging circuit 82 for receiving signals from LPRMs of the neutron monitoring system 24 and averaging the local power levels to produce the power level of the reactor. The averaging circuit 82 transmits a signal to a comparator 84. ORM 46 also comprises a coolant block threshold circuit 86 for receiving a signal from the recirculation flow rate measurement system 28. The circuit 86 is set at a power level of coolant block threshold determined as a function of the core coolant flow rate as indicated by a line 88 in FIG. 2, and transmits to the comparator 84 a threshold level signal corresponding to the prevailing core coolant flow rate upon receipt of a signal from the system 28. The comparator 84 compares the two signals from the circuits 82 and 86 supplies a comparison signal to a signal generator 90 which transmits, when the power level is higher than the threshold level, a coolant block signal to the flow rate controller 30 of the flow control system 18. The flow rate controller 30 adjusts the openings of the control valves 16 through the function generator 32 and hydraulic control means 34 so as to block the increase in the recirculation flow rate and thus the increase in the core coolant flow rate, thereby maintaining the core flow rate at the blocked level. Thereafter, the core flow rate is manually returned to a normal operating region 92 as shown in FIG. 2. Alternatively, the signal generator 90 may be modified to generate a coolant run-back signal. In this case, the flow controller 30 which receives the coolant run-back signal adjusts the openings of the control valves 16 through the function generator 32 and hydraulic control means so as to automatically run-back or reduce the recirculation flow rate and thus the core coolant flow rate to a minimum rate. The signal generator 90 may be further modified to selectively produce a coolant block signal or a coolant run-back signal. Thus, ORM 46 monitors the reactor power level which might be caused by an increase in the core coolant flow rate. When the reactor power level reaches a coolant block threshold line 88 shown in FIG. 2, ORM 46 transmits a coolant block signal or a coolant runback signal to the flow control system 18, to thereby block the increase in the core coolant flow rate or to thereby run-back the core coolant flow rate. Thus an excessive increase in reactor power which might otherwise be caused by an increase in the core coolant flow rate can be suppressed before the need to scram the reactor arises, and thus operation of the reactor can be continued. The circuit 86 is set beforehand at a power level of coolant block threshold by analysis in such a manner that when the power level is caused to rise by an increase in flow rate, the blocking or running-back can be effected to keep the core characteristics parameters such as the maximum linear heat generating rate and minimum critical power ratio from reaching their critical levels that may cause the breakdown of the fuel cladding. In the embodiment shown in FIG. 2, the coolant block threshold line 88 has the same starting point D as the flow rate control line 38 and is generally situated slightly above line 38. More specifically, the threshold level is about 105% of the rated power level at the rated flow rate and at flow rates adjacent to the rated flow rate and is about 103% of the power level on the flow rate control line 38 in a substantial range of flow rates below these flow rates. Generally, the threshold level at the rated flow rate and flow rates adjacent to the rated flow rate can be set at a value in the range between 103 and 108% of the rated power level, and the threshold level in a substantial range of flow rates below these flow rates can be set at a value in the range between 102 and 107% of the power level for the flow control line 38. As apparent from the foregoing, according to the invention, there is provided, in addition to the APRM 40, TPM 42 and RBM 44 of the conventional core monitoring system, the ORM 46 operative to prevent an excessive rise in the reactor power level caused by an increase in the core coolant flow rate, before the reactor is scrammed. As a result, various advantages are offered in operating a nuclear reactor by the present invention. Firstly, when an operator turns the wrong valves, or some equipment misoperates, for example, the core coolant flow rate may abnormally rise and the power may rapidly rise. When this phenomenon occurs, it is possible to inhibit an abnormal transient change in core characteristics (minimum critical power ratio, maximum linear heat generating rate, rated power, flow vibration characteristics, etc.) by blocking or running back an increase in the core coolant flow rate by controlling the recirculation pumps. Secondly, when TPM 42 and APRM 40 are the only monitoring devices used, the reactor is scrammed when the threshold power level is exceeded as a result of a rise in power caused by an increase in the core coolant flow rate. This makes it inevitable to interrupt the operation of the reactor. However, according to the invention, when the threshold power level (about 105% of rated power) of ORM 46 is exceeded, the increase in the core coolant flow rate is blocked or the flow rate is run-back, so that an excessive rise in power due to an increase in flow rate can be inhibited. After the inhibiting action is performed, the core coolant flow rate can easily be returned to a normal flow rate control condition. Thus the invention minimizes the number of times the reactor is scrammed and enables the reactor to be substantially continuously operated with minimum interruption. Another important advantage offered by the invention is that because of the provision of ORM in addition to TPM and APRM as a system for monitoring the power level caused by a rise of the core coolant flow rate, improvements are provided to the minimum critical power ratio which is the monitor index for preventing the thermal breakdown of the fuel cladding owing to the fact that the scram threshold of TPM and APRM is about 115-120% of the rated power at or near the rated flow rate but ORM has a coolant block and run-back threshold which is about 105% of the rated power and thus the range of variations in minimum critical power ratio before the threshold power level is reached can be reduced to 1/3-1/4 by taking as a reference the range of changes occurring until about 115-120% of the rated power is attained. The same goes for the maximum linear heat generating rate which is the monitor index for preventing the mechanical breakdown of the fuel cladding. Thus as compared with the nuclear reactor having no ORM as disclosed in the aforesaid U.S. Pat. No. 3,565,760, for example, the reactor provided with ORM according to the invention shows no increase in the core characteristics parameters such as minimum critical power ratio and maximum linear heat generating rate above their critical levels which might brought about the breakdown of the fuel cladding, even if the power level is raised in rated operation. Thus a nuclear reactor with ORM could develop higher power than a nuclear reactor of the same design having no ORM. This feature of the invention will be described in detail by referring to the drawings. Generally, in designing a nuclear reactor, the critical lever Lu of any one of core characteristics parameters that may brought about breakdown of the fuel cladding shown in FIG. 7 is first obtained. Then, the operation critical level Lo of the core characteristics parameter for normal operation is set such that critical lever Lu can be maintained even if an excessive rise in power is caused by the carelessness of an operator or misoperation of some equipment. More specifically, the operation critical level Lo is set in such a manner that, assumming that the core characteristics parameter X vary as indicated by a line (a) in FIG. 7 and the range of variations of the core characteristics parameter are denoted by .DELTA.X, then Lo.ltoreq.Lu-.DELTA.X. In FIG. 7, a line (b) represents an unallowable operation condition, and a line (c) is an allowable operation condition in which operation efficiency is lower than in the operation condition represented by line (a). It is essential that in setting the operation critical level Lo, all the factors concerned in a rise in power and all the core characteristics parameters that constitute indices of breakdown of the fuel cladding should be taken into consideration. The principal factors concerned in a rise in power include withdrawing of control rods, an increase in the core coolant flow rate and a rise in the pressure in the core due to shutoff of the load. The indices of breakdown of the fuel cladding include the maximum linear heat generating rate and minimum critical power ratio. The latter can be expressed in terms of the fuel assembly power. FIG. 8 show variations .DELTA.P.sub.L of the maximum linear heat generating rate R.sub.L and variatins .DELTA.P.sub.B of the fuel assembly power P.sub.B occurring in a nuclear reactor provided with APRM and PBM when a rise in power is caused by the three factors referred to hereinabove. FIG. 9 is a view similar to FIG. 8 but showing the values obtained with a nuclear reactor provided with ORM according to the invention in addition to APRM and RBM. As can be clearly seen in FIG. 8, the provision of APRM and RBM enables .DELTA.P.sub.L and .DELTA.P.sub.B to be reduced as indicated by hatching when an excessive power rise is caused by control rod withdrawing and pressure rise, but .DELTA.P.sub.L and .DELTA.P.sub.B show no reduction when an excessive power rise is caused by an increase in the core coolant flow rate. This makes it inevitable to set the operation critical level L.sub.o for normal operation of the reactor by taking into consideration such relatively large values of .DELTA.P.sub.L and .DELTA.P.sub.B. Thus L.sub.o is limited to a low level after all. On the other hand, if ORM is additionally provided .DELTA.P.sub.L and .DELTA.P.sub.B can be reduced in all aspects and thus the operation power level of the reactor can be set at a high level. A further important advantage of the invention is that since the threshold power level at which ORM is set is determined as a function of the core coolant flow rate, a rise in power can be prevented by all means when the power level reaches the threshold power level corresponding to the prevalling flow rate regardless of the situation in which the power is increased by a rise in the core coolant flow rate. This feature of the invention will be described in detail by referring to FIGS. 10 and 11 and by comparing the power monitoring system according to the invention with the control system disclosed in Japanese Patent Publication No. 21518/79 referred to hereinabove in the background of the invention. The control system of the prior art is provided with means for resetting, in a normal operation mode, the recirculation coolant flow rate threshold M and core coolant flow rate threshold C only when the power density calculated at certain time intervals is higher in level than the value obtained by the preceding calculation, to thereby avoid an increase in flow rate above the threshold levels. In this control system, when the power level is reduced by reducing the core coolant flow rate after the threshold levels M and C are set at a high power level P.sub.H shown in FIG. 10 following a slow and gradual rise in power, the threshold level M would be kept at the high level. If, for example, the flow rate rises due to the failure of the flow control system after the period of a low power P.sub.L has lasted for some time, the flow rate would continue to rise until the level M or C is reached. A power level P.sub.H * attained at this time would be higher than the aforesaid high power level P.sub.H by an amount corresponding to a reduction in the amount of Xenon (neutron absorber) in the core occurring during the time the reactor is operated at the low power level P.sub.L. In the case of a reactor provided with ORM according to the invention, when the flow rate begins to rise from the low power P.sub.L under similar circumstances, the power level does not rise above the power level of coolant block threshold as shown in FIG. 11 and the rise in power is blocked at a threshold level P.sub.T corresponding to the prevailing flow rate. That is, according to the invention, even if the power level drops or the amount of Xenon shows a variation prior to the rise in power, it is possible to effectively suppress an excessive rise in power due to a rise in flow rate.