JPS62272192A - Stability stabilizer in core of boiling water type reactor - Google Patents

Abstract

(57) [Summary] This bulletin contains application data before electronic filing, so abstract data is not recorded.

Description

Detailed Description of the Invention 3. Detailed Description of the Invention [Object of the Invention] (Industrial Application Field) The present invention relates to an in-core stability stable coloring δ of a boiling water reactor, and in particular, This invention relates to an in-core stability stabilization device using a feed water temperature signal.

(Prior Art) Generally, in a boiling water reactor, a reactor power control system provided for AI control of the reactor power has a configuration schematically shown in FIG.

This reactor power control system is composed of a reactivity control system 1 and a turbine control system 2, and the reactivity control system 1 is composed of control rods 3, a control rod drive system 4, and a recirculation flow rate control system 5. The control rod 3 and control rod drive system J 4t have output control and output distribution adjustment functions. These adjustments are made by remotely controlling the control rod drive system 4 by selecting the control 4413 to be operated by the control signal 6 manually controlled by the operator n via the f, II 110 rod drive system operation gA7. However, for output control, control rod 3
The power distribution is adjusted by changing the pattern of control rod positions.

The output tII1111 by adjusting the recirculation flow rate utilizes the characteristic that the output changes almost proportionally to the core flow rate, and the recirculation flow rate is normally adjusted by using the recirculation flow opening control system operation panel 8. The speed command signal of the recirculation pump 9 is changed by the control signal 6 manually controlled by the operator r1 through the recirculation flow rate control system 5, and the power frequency of the recirculation pump drive type e machine (not shown) is changed through the recirculation flow rate control system 5. This is done by varying the speed of the recirculation pump 9.

While the reactor output is being changed by the reactivity control system 1, the turbine steam control valve 10 and the turbine Bypass valve 11 is adjusted,
Since the amount of steam generated in the reactor 13 that is supplied to the turbine 12 is adjusted, the output of the turbine power generator 1114 changes by an amount corresponding to the change in the amount of reactor steam generation.

In addition, in a boiling water nuclear reactor plant (hereinafter referred to as a full capacity bypass plant) that has a turbine bypass valve 11 with a capacity equal to the flow rate of air generated from the reactor 13 at rated output, if the generator load is lost, the water supply Since the heater 15 is no longer used, low-temperature feed water flows into the reactor core 17 through the water supply pipe 16. In order to prevent a situation where the neutron flux in the core increases due to the applied reactivity resulting in a scram, a selective control rod insertion mechanism is used to insert the control rod 3 selected in advance when the generator is loaded. (hereinafter referred to as SRI) mechanism 18 is provided.

In addition, a plurality of neutron sensors 19 and 20 are arranged in the reactor core 17 to detect neutron flux at various locations in the reactor core 17, and each signal detected by each of these sensors 19 and 20 is a local neutron Bundle detection signal (hereinafter referred to as LPRM signal) 2
1 and taken out to the outside.

Further, the coolant recirculation system is provided with a flow meter (for example, an elbow meter) 22 for measuring the recirculation flow rate of the coolant, as shown in FIG. 3, and a signal converter 23 provided as necessary. The signal is converted into a distribution signal 24 via the .

The LPRM signal 21 and Takigawa signal 24 are guided to a monitoring device shown in FIG. That is, a plurality of neutron sensors 1
Each LPRM signal 21 detected by 9.20 (Y) is averaged by an averaging circuit 25 using an amplification amplifier as shown in FIG. The average neutron flux signal (hereinafter referred to as the APRM signal) 26 is output (this APRM signal 26 and the flow!n signal 24 are heat output as shown in FIG. 4) j monitor scrum block ( (hereinafter referred to as FTPM scram block) 27, APRM scram block 28, annunciator, and control rod river bed prevention block 29. Then, by the actuation signal 30 output from the annunciator, i, lIm rod withdrawal prevention block 29,
The annunciator 31 of the alarm device C and the υ1111 rod handling prevention device 32 are driven.

Each block 27, 28, 29 operates according to the settings shown in FIG. That is, TPM Scrum Block 27
The region above the straight line 33 including the break point is the trip region, and the A PRM scram block 28 has a certain constant value 34 that exceeds the rated value of the △PRM output, which is not dependent on the flow rate.
The area exceeding this is the trip area. Upon entering these trip regions, each block 27, 28 issues a trip signal to the reactor protection system, not shown, causing the reactor 13 to scram.

The annunciator and control rod withdrawal prevention block 29 is the fifth
In the figure, the area above the straight line 35 is the operating area, and when it enters this operating area, the work is done! The lJJ signal 30 is output. This activation signal 30 activates an annunciator 31, which is an alarm device, and also activates a control rod river bed blocking device 32.

In a boiling water reactor, the inherent stability of the reactor is compromised under any expected operating conditions, such as natural circulation conditions where there is loss of on-site ffi and the reactor recirculation pump loses me. The post is set up so that there is no such thing.

Reactor-specific stability includes hydraulic stability in the channels where oscillations in coolant flow prevent heat transfer to the moderator and thereby increase reactor power. and the core stability due to the reactivity feedback effect of the entire reactor.

Normally, reduction ratios are used to evaluate the stability of boiling water reactors. FIG. 6 shows the definition of the reduction ratio. That is, the amplitudes X2 and X of the response song FA36 of the system to the disturbance.

The reduction ratio is defined as the ratio X2/Xo, and x2/X0<i
, ot- the system is stable and X 2 / X o > 1
, OT: The system oscillates and is in an unstable state. Also, X
, 2 /Xo=1 indicates the stability limit.

These stability deteriorates as the coolant flow decreases and as the reactor output increases, the reduction ratio increases.

Also, when considering channel stability, the lower the coolant flow and the higher the reactor power, the larger the void volume ratio becomes, and the ratio of two-phase pressure drop to single-phase pressure drop becomes larger, worsening channel stability. .

Moreover, when the proportion of void volume is large, the void reaction coefficient becomes negative 1 and becomes large, which deteriorates the core stability.

In actual operation, it is desirable that the width reduction ratio be approximately 0.8 or less, so a contour line 37 with a width reduction ratio of 0.8 is provided as shown in Figure 5, and the reactor recirculation pump loses its function. In such cases, care is taken to prevent unstable operation on the high output, low flow rate side from the contour line 37.

(Problem to be Solved by the Invention) By the way, if the generator load is lost in a full-capacity bypass plant, if the reactor recirculation pump is tripped or blown, the oil from the turbine to the feedwater heater will be After SRI, which aims to suppress the increase in neutron flux in the reactor core due to the lack of steam and the drop in feed water temperature, the reactor shifts to operation with only the station load (hereinafter referred to as in-station isolated operation).

After isolated operation within the plant, we start up the recirculation pump in order to connect the generator as quickly as possible from the standpoint of power generation efficiency, but if for some reason the start-up of the recirculation pump is delayed. Since the neutron flux in the core continues to increase due to the decrease in the feed water temperature, in other words, the output continues to increase, it is necessary to prevent operation in an unstable region from a stability standpoint, and to improve the TPM.
To prevent scrams from occurring, operators constantly monitor neutrons, etc., and limit operation based on continuous rotation conditions.

Therefore, there are problems in that not only is there a heavy burden on the operators, but the range in which the reactor can be operated is narrow.

Also, in normal boiling water reactor power plants,
If the reactor recirculation pump trips, the core flow rate decreases and may enter an unstable region, resulting in the same problem as above.

The present invention has been developed with these points in mind, and aims to improve reactor operating performance by reducing the burden on operators and easing operational constraints, as well as to prevent unstable conditions from occurring with higher reliability. The purpose of the present invention is to provide a device for stabilizing stability within a nuclear reactor core, which is capable of preventing such problems and improving safety performance.

[Structure of the invention]

(Means for Solving the Problems) The present invention provides a function setting means for setting a function for adjusting the neutron flux in the reactor core from the reactor feed water temperature, and a function from the function setting means and a reactor. A calculation means for calculating the neutron flux by inputting the feed water heating signal; a detection means for detecting the neutron flux in the reactor core; Comparison means for comparing the neutron flux and outputting a difference signal when the detected neutron flux exceeds the detected neutron flux, and inserting a specific control body into the reactor core by inputting the difference signal from the comparison means. and control rod insertion signal generating means for outputting an information signal.

(Function) The in-core stability stabilization device for a boiling water reactor according to the present invention uses a feed water temperature signal to monitor the sub-reactor using a separate configuration and equipment automation technology, and prevents instability due to instability. When the reactor approaches an unstable region, it automatically prevents the reactor from entering an unstable region, and if it unavoidably enters an unstable region, it automatically adjusts the power distribution for stabilization. It will be planned. Therefore, it is possible to improve the reactor operating performance by reducing the burden on the operators, relaxing the operation 1-1 regulations, and improving the stability performance.

(Embodiment) An embodiment of the present invention will be described below with reference to FIGS. 1 and 2.

FIG. 1 shows an example of the in-core stability stabilization device for a boiling water reactor according to the present invention, and the reference numeral 21 in the figure indicates the third
Multiple L from each LPRM sensor 19°20 shown in the figure
This LPRM signal 21, which is a PPM signal, is input to an averaging circuit 40, subjected to an average calculation, and amplified by an amplifier 41, and is input as an APRM signal 42 to a comparator circuit 43.

Further, in FIG. 1, reference numeral 44 is a function setting circuit that sets a function for calculating the peak of the neutron flux in the reactor core from the reactor inlet feed water temperature. It is input to the calculation circuit 46 together with the feed water temperature signal 45 from the water supply temperature sensor (not shown) installed in the reactor 13 population section of the water supply pipe 16 shown in FIG. It is now possible to do so. This neutron flux P is then input to the comparison circuit 43 shown in FIG. 1 and compared with the PRM signal @42, and if the APRM signal @42 is greater than the A difference signal 47 is output and input to a control rod insertion signal generation circuit 48.

As shown in FIG. 1, a control rod designation signal 50 from a control rod designation circuit 49 is input to the control rod insertion signal generation circuit 48.
An information signal 51 regarding one control rod insertion outputted from 8 is input to a control rod insertion signal generating device 52 and an alarm device 53 that issues an alarm for automatic control rod insertion.

Next, the operation of this embodiment will be explained.

Each LPRM signal 21 is manually inputted to an averaging circuit 40 to be averaged, and is amplified by an amplifier 41, and the APR
It is input to the comparison circuit 43 as the M signal 42.

On the other hand, the function set by the function setting circuit 44 is manually inputted to the calculation circuit 46 for the water supply inlet temperature 45 and 6, and the neutron flux P is calculated here. This neutron flux P is then manually input to the comparison circuit 43.

The comparison circuit 43 compares the APRM signal @42 and the neutron flux P, and if the APRM signal 42 is larger than the neutron flux 1, the difference signal 47 is outputted to the control rod insertion signal generation circuit 48. is input. A control rod designation signal 50 from a control rod designation circuit 49 is input to the control rod insertion signal generation circuit 48 .

By the way, when a generator load occurs in a full-capacity bypass plant, the neutron flux P, which is a function of the reactor inlet feed water temperature set by the function setting circuit 44, is shown as a set value characteristic line 54 in FIG. For example, it can be determined by the following equation.

When T > To −-200℃, P=Po=30 (%)・・・・・・・・・・・・(1
) When TNT1=1oO℃, P=P1 =50 (%)・・・・・・・・・・・・(
2) When T ≧T≧To, P=-0, 2T+70 (3)
That is, when T > To and T <'1-1, regardless of the IH degree T of the reactor inlet water supply, the threshold value 1] and P for the neutron flux P are the limits, respectively, and T1≧T
≧To, the limit of the neutron flux P is a value expressed by a monotonically decreasing function with the reactor inlet feed water temperature d as an independent variable.

In addition, P and PO2 in the above equations (1) and (2) are the applied reactivity due to SRI at the time of generator electric load cut-off in J'3 for the entire 1-bypass blunt, and the change in output distribution during isolated operation in the station. It is determined taking into account the following.

Moreover, the formula (3) is the general formula p=aT+b (a.

b is a constant), but it is not necessarily limited to this, and any other function may be used as long as it is a monotonically decreasing function.

Furthermore, as the feed water temperature T, the reactor 13 shown in FIG.
Although it is common to use the reactor inlet feed water temperature of the nearby feed water pipe 16, it is sufficient that it is related to the feed water temperature. The feed water temperature measured by the method may also be used.

Also, constant a=-0, 2°b=70 in equation (3).
<1) P in the formula. -30, P 1 in equation (2)
= 50 is an example, and it is possible to change it to a more appropriate value specific to the plant.

In this way, the combination of the υItil rods that allows the output to be lower than the limit value of the neutron flux determined by the set value characteristic line 54 according to the a, II till collar insertion low insertion signal generation circuit 48 and the difference signal 47. , the position is selected from the control rod designation circuit 49, and an information signal 51 regarding the control rod is provided to a control rod insertion signal generator 52 and an alarm device 53 that issues an alarm for automatic control rod insertion.

(Effects of the Invention) As described above, the present invention provides a function setting means for setting a function for calculating the neutron flux in the reactor core from the reactor feed water temperature, and a function from the function setting means and the reactor feed water temperature. a calculation means for calculating the neutron flux based on the input of the neutron flux, a detection means for detecting the neutron flux in the reactor core, and a comparison between the neutron flux detected by this detection means and the neutron flux calculated by the calculation means, A comparison means that outputs a difference signal when the neutron flux exceeds the calculated neutron flux, and a control rod insertion signal that outputs an information signal for inserting a specific control rod into the reactor core based on the input of the difference signal from the comparison means. Since the reactor is equipped with a generating means, it is possible to improve the operating performance of the nuclear reactor, such as reducing negative lII on the operating day and easing operational constraints, and it is also possible to improve safety.

[Brief explanation of the drawing]

Fig. 1 is a block diagram of an in-core stability stabilization device for a boiling water reactor showing an embodiment of the present invention, and Fig. 2 is a setting characteristic diagram showing the boundary of the unstable region in the device of Fig. 1. , 3rd
Figure 4 is a schematic diagram of the reactor power control system in a typical boiling water nuclear coupler, Figure 4 is a block diagram showing the conventional processing system for reactor neutron flux detection signals, and Figure 5 is the conventional monitoring system Figure 6 is an explanatory diagram showing the definition of the reduction ratio. 42... APRM signal, 43... comparison circuit, 44.
... Function setting circuit, 45 ... Water supply inlet temperature signal, 46
...Arithmetic circuit, 47...Difference No. 13, 48...Control rod insertion signal, 49...Control rod designation circuit, 50...
Control rod designation signal, 51... Information signal, 52... Control rod insertion signal generator, 53... Warning device, P... Calculated neutron flux. Applicant's representative ili Fuji - male 4.8-7K
, 4 degrees (0C) 2nd

Claims (1)

[Claims] 1. Function setting means for setting a function for calculating the neutron flux in the reactor core from the reactor feed water temperature; A calculation means for calculating the neutron flux, a detection means for detecting the neutron flux in the core, and a comparison between the neutron flux detected by the detection means and the neutron flux calculated by the calculation means, and the detected neutron flux is calculated. and a control rod insertion signal generating means that outputs an information signal for inserting a specific control rod into the reactor core based on the input of the difference signal from the comparison means. An in-core stability stabilizing device for a boiling water reactor, characterized by comprising: 2. In-core stability of a boiling water reactor as set forth in claim 1, wherein the control rod insertion signal generation circuit has control rod selection means for selecting a specific control rod. Stabilizer.