JPH0480356B2 - - Google Patents

Description

DETAILED DESCRIPTION OF THE INVENTION [Objective of the Invention] (Industrial Field of Application) The present invention provides a means for reducing the power plant or This invention relates to a reactor power control device for a boiling water nuclear power plant that enables continuous operation of a nuclear reactor.

(Prior art) In general, in a boiling water nuclear power plant, in order to continue isolated operation within the plant in the event of an accident in the power system outside the plant, the reactor recirculation pump is tripped at the same time as the accident occurs, and the reactor recirculation pump is Emergency insertion of control rods reduces reactor power,
On the other hand, by rapidly opening the turbine bypass valve and allowing steam generated from the reactor to escape to the condenser, it is possible to continue operation without stopping the reactor.

FIG. 6 is a schematic system diagram of a reactor power control device for a conventional boiling water nuclear power plant.

Steam generated from the nuclear reactor 1 is fed through a main steam pipe 2 to a turbine 4 via a main steam control valve 3, and the turbine 4 is driven. turbine 4
The rotation is transmitted to the generator 5, where it is converted into electricity, and the electricity is transmitted to the grid 8 via the main transformer 6 and the main disconnector 7. On the other hand, the steam that has performed work in the turbine 4 is condensed in a condenser 9, and then returned to the reactor 1 by a reactor feed water pump 10. Further, between the main steam pipe 2 and the condenser 9, there is a bypass conduit 12 that bypasses the main steam control valve 3 and the turbine 4 and has a bypass valve 11.
The steam passing through the bypass conduit 12 is also condensed in the condenser 9.

In the boiling water nuclear power plant configured in this way, if any failure occurs in the system 8 or the generator 5, the main disconnector 7 and the main transformer 6 will open, and the open signal for the main transformer 6 will be activated. Load breakage detection circuit 1
Sent to 3. When the open signal of the main transformer 6 is detected by the load shedding detection circuit 13 in this way, the load shedding detection circuit 13 generates a load shedding signal and the main steam control valve 3 is activated. At the same time as the bypass valve 11 is suddenly closed, the bypass valve 11 is suddenly opened. However,
Steam sent from the nuclear reactor 1 to the turbine 4 is quickly cut off by the main steam control valve 3 and sent to the condenser 9 via the bypass valve 11.

On the other hand, the sudden closing of the main steam regulating valve 3 is detected by the main steam regulating valve sudden closing detection device 14, and the drive motor 16 of the reactor recirculation pump 15 is quickly stopped in response to the sudden closing of the main steam regulating valve 3. At the same time, the selective control rod insertion device 17a is activated, and some selected control rods 18 are urgently inserted into the reactor core 19.
The output of the nuclear reactor 1 is reduced. Further, the signal from the main steam control valve quick closing device 14 is sent to the feed water pump control device 20, and the signal from the main steam control valve quick closing device 14 is sent to the feed water pump control device
0, one or more reactor feed water pumps 10 are stopped, and the reactor is shifted to in-station independent operation. In order to ensure the safety of the reactor 1, for example, when the water level L in the reactor decreases and reaches a set value, the water level detector 21 is activated to issue a low water level signal, and the emergency shutdown circuit 22 is activated. All the control rods 23 are inserted into the reactor core 19, and the reactor is shut down.

In this way, in a typical boiling water reactor, when a failure occurs in the system 8 or the generator 5, the bypass valve 11 is suddenly opened, the reactor recirculation pump 15 is stopped, and the selective control rod insertion circuit is closed. 17a
is activated, the selected control rod 18 is urgently inserted into the reactor core, the output of the reactor is reduced, and the reactor water level rises due to the stoppage of the reactor recirculation pump 15 by stopping the reactor feed water pump 10. By suppressing this, the transition to isolated operation of the nuclear reactor is carried out stably.

(Problems to be Solved by the Invention) Figure 2 shows one quarter of the radial cross section of the core, and each square represents one control rod.
Also, the number inside this square indicates the group to which the control rod belongs. This group divides all the control rods in the core into multiple groups in advance,
The control rods are operated for each group. Further, the number columns on the lower and left side of FIG. 2 indicate I coordinates and J coordinates for indicating the radial position of the control rod in the reactor core using coordinates (I, J). These coordinate numbers are given to each fuel assembly and the position between the fuel assemblies, that is, the position where the control rods or nuclear instrumentation equipment are placed. The origin is at the bottom left when expanded to .

According to the grouping of control rods with such a configuration, the control rods that should be inserted urgently are determined by the operator who performs an artificial nuclear thermal analysis for each fuel combustion cycle. It was necessary to make a selection by operation, making driving operations complicated.

Furthermore, when restoring the power plant's output after an accident, the reactor power must be lowered further and the emergency inserted control rods must be pulled out in order to follow the operating procedures for fuel adjustment, resulting in a change in the reactor power distribution. as well as distorting the fuel and having a detrimental effect on the fuel.
There were problems such as making operation complicated and reducing the operating rate of the power plant.

The purpose of the present invention is to quickly shift to isolated operation within the power plant when a failure occurs in the power system outside the power plant.
Boiling water can reduce the burden on operators when performing emergency insertion operations of selected control rods, and can smoothly restore reactor power from the above-mentioned isolated operation to recovery from the above-mentioned accident without adversely affecting the fuel. The objective is to obtain a reactor output control device for a nuclear power plant.

[Structure of the invention]

(Means for Solving the Problems) In order to achieve the above object, in the present invention, when a failure occurs in the power system outside a nuclear power plant and the system is rapidly transitioned to standalone operation, a The present invention provides a reactor power control device for a boiling water nuclear power plant, characterized by having a selective control rod insertion circuit for urgently inserting selective control rods made up of some or all of the control rods that have been removed into the reactor core.

(Function) In a device configured as described above, the selective control rods that should be inserted into the reactor core in an emergency manner are placed in the outermost part of the reactor core in order to quickly shift the reactor to isolated operation in the event of a power system accident. Since the configuration is made up of part or all of the installed control rods, it is possible to always obtain a nearly constant control rod value, and as the fuel combustion cycle progresses, the selection of the selected control rod to be urgently inserted is possible. No changes need to be made.

Furthermore, it is possible to reduce the distortion of the core power distribution when withdrawing the selected control rod during power recovery after failure recovery, and it is possible to secure thermal margin for the fuel.

(Example) Hereinafter, an example of the present invention will be described with reference to the drawings.

FIG. 1 is a schematic system diagram of one embodiment. still,
Here, the same parts and locations as in FIG. 6 are denoted by the same reference numerals, and explanations of their configurations will be omitted. In this embodiment, a water pump abnormality detection circuit 25 is connected to the water pump control device 20.
If any one of the reactors fails and a water supply volume commensurate with the output of the reactor cannot be secured, this abnormality is detected. When this abnormality occurs, the selective control rod insertion circuit 17 is activated via the feedwater pump abnormality detection circuit 25, and a part of the control rods is inserted into the reactor core 19 in an emergency manner. When the emergency shutdown detection circuit 26 is activated, the emergency shutdown circuit 22 is activated and all control rods are inserted into the reactor in an emergency manner.

Further, FIG. 3 shows grouping of control rods in an embodiment of the present invention. In the figure, the same parts as in FIG. 2 are denoted by the same reference numerals, and explanations thereof will be omitted.
In Figure 3, 5 groups and 6 groups at the outermost periphery of the core
The control rods 30 represented by groups are grouped independently from groups other than the outermost periphery. In this embodiment, the emergency insertion groups of control rods are divided into groups as shown in FIG. 3, and the outermost control rods 30 of the core are divided into one or more, A part or all of the control rods are selective control rods, and are configured to be operated by a selective control rod insertion circuit 17.

However, if the system 8 or the generator 5 etc. breaks down and the main transformer 6 etc. is opened, the main steam control valve 3 is suddenly closed via the load failure detection circuit 13 and the bypass valve 11 is closed. It is suddenly opened and the steam from the reactor 1 is released to the condenser 9. Meanwhile, the reactor recirculation pump 15 stops in response to the sudden closing of the main steam control valve 3, and approximately 30 seconds later, the output of the reactor 1 decreases to 50-60%.
The output drops to the rated output. In addition, in response to the sudden closing of the main steam control valve 3, the selective control rod insertion circuit 17 is activated, and in a few seconds, part or all of the outermost control rods 30 are urgently inserted, and the atomic force is increased by approximately 30% of the rated output. The power of the furnace 1 is reduced. In this way, by stopping the reactor recirculation pump 15 and urgently inserting the outermost control rod 30, the output of the reactor 1 is reduced to 20 to 30% of the rated output, and the steam generated in the reactor 1 is forced to pass through the bypass valve 11. is discharged to the condenser 9 through the reactor 1.
The system is separated from the system 8 and becomes an isolated operating state within the station. At this time, by stopping the operation of the predetermined water supply pump 10, the rise in the water level L due to the stopping of the reactor recirculation pump 15 is suppressed, and the emergency stop circuit 22 is not activated.

FIG. 4 shows the control pattern during rated output operation. Similar to FIGS. 2 and 3, FIG. 4 shows 1/4 of the radial cross section of the core, and each square represents one control rod. The number inside the square indicates the control rod's insertion ratio in the axial direction, and the height in the axial direction (fuel effective part).
It is divided into 48 parts, with 0 representing total insertion and 48 representing total withdrawal. For example, in Fig. 4, the control rod at the position I=30, J=31 is 4, which is 4 from the bottom in the axial direction.
It shows that it has been inserted up to 4/48. Note that the control rod for full withdrawal should originally be indicated as 48, but this is omitted in Fig. 4 and is left blank.

Incidentally, when the reactor is operating at the rated output, the control rod pattern is such that the outermost control rod 30 is normally always at the fully withdrawn position, as shown in FIG. Therefore, the neutron absorption capacity of the control rods hardly decreases depending on the operating period of the nuclear reactor, and the selective control rod insertion circuit 17 allows the control rods to be inserted urgently according to the fuel combustion cycle, as in the past. A predetermined negative reactivity can always be applied to the reactor 1 by inserting the outermost control rod 30 without changing the selection of control rods.

Furthermore, the power distribution of the fuel channel around the outermost control rod 30 becomes a low value due to neutron leakage outside the core, resulting in a state as shown by the solid line a in Figure 5, and that value is changed due to fuel adjustment. The threshold value (dotted line b in Figure 5) in the operating procedure is not exceeded. Therefore, the failure of system 8 etc. is restored,
When transitioning from in-house isolated operation to output operation, the emergency inserted outermost control rod can be withdrawn with a margin within the thermal limit value of the fuel and the above threshold, and the output increases to the initial output. can do.

By the way, the conventional selection control insertion device operates only in the event of an accident outside the plant;
For example, when a reactor feedwater pump malfunctions and stops, there are inconveniences such as the inability to urgently reduce the reactor output to continue operating the turbine generator and the like. In this embodiment, when a part of the feed water pump fails during power operation of the reactor 1, the abnormality is detected by the abnormality detection circuit 25, the selective control rod insertion circuit 17 is activated, and the Part or all of the control rods 30 are urgently inserted into the reactor core 19, the output of the reactor 1 is reduced by about 30% of the rated output, and output operation is continued with the water supply capacity excluding the failed water supply pump. Therefore, after the above-mentioned water pump failure is restored, the output is restored without being affected by the thermal limit value of the fuel, etc. by withdrawing the outermost control rod, just as when restoring from isolated operation within the station. can be set.

〔Effect of the invention〕

In the present invention, in order to rapidly shift the reactor to isolated operation in the event of a power system accident, the selective control rods that should be inserted into the reactor core in an emergency manner are selected from among the control rods located at the outermost part of the reactor core. Since the control rod value is always almost constant, there is no need to change the selection of the selected control rod to be inserted urgently as the fuel combustion cycle progresses. Emergency control rod insertion operations can be performed easily.

Furthermore, it is possible to reduce the distortion of the core power distribution when withdrawing the selected control rod during power recovery after failure recovery, and it is possible to secure thermal margin for the fuel.

[Brief explanation of drawings]

Fig. 1 is a schematic system diagram of an embodiment of the reactor power control device for a boiling water nuclear power plant according to the present invention, and Fig. 2 shows an example of conventional control rod emergency insertion grouping for a 1/4 core. Explanatory drawings, FIG. 3 is a diagram showing an example of grouping of selected control rods according to an embodiment of the present invention, FIG. 4 is a diagram showing an example of a control rod pattern during reactor power operation, and FIG. FIG. 6, which is a diagram showing the power distribution of the fuel channel around the control rods, is a schematic system diagram of a reactor power control device for a conventional boiling water nuclear power plant. 1... Nuclear reactor, 3... Main steam control valve, 4... Turbine, 8... System, 9... Condenser, 10... Reactor feed water pump, 11... Bypass valve, 13...
Load breakage detection circuit, 14...Main steam control valve sudden closing detection device, 15...Reactor recirculation pump, 17...
...Selective control rod insertion device, 18...Control rod, 19...
... Core, 20 ... Water pump control device, 30 ...
Outermost control rod.

Claims (1)

[Claims] 1. In the event of a failure in the power system outside a nuclear power plant, in order to rapidly transition to isolated operation, selective control rods consisting of part or all of the control rods installed at the outermost part of the reactor core may be urgently inserted into the reactor core. 1. A reactor power control device for a boiling water nuclear power plant, characterized by having a selective control rod insertion circuit.