JPH02170099A - Controller for nuclear reactor - Google Patents

Abstract

PURPOSE:To maintain the safety of a core throughout operation by varying the rate of control rod insertion by a control rod driving controller and adjusting the output of a nuclear reactor when a feed water heating loss and a feed water temperature drop are caused. CONSTITUTION:For example, the feed water loss is generated during operation at a point X1 in a nuclear reactor recirculation flow rate %-to-output % characteristic diagram. The temperature of feed water drops gradually. The width DELTATFW of this temperature drop is monitored by a feed water temperature detector 14 and inputted to the control rod driving device 19 as a signal 18. When a signal 20 indicating the feed water drop is inputted, on the other hand, the program of a recirculation flow rate control curve calculating device 17 is started and the recirculation flow rate control curve at the operation point X1 is calculated with signals from a neutron instrumentation system 10 and a reactor flow rate instrumentation system 3 and the signal of a control rod pattern from a control rod driving mechanism to vary the insertion rate of the control rod according to the degree of the current output of the segment at this time.

Description

[Detailed description of the invention] [Purpose of the invention] (Industrial application field) The present invention relates to a control device for a nuclear reactor.

(Prior Art) Generally, in a boiling water reactor, a reactor pressure vessel is housed in the reactor containment vessel, and fuel assemblies containing a large number of nuclear fuels are loaded into the reactor pressure vessel and the reactor core is The department is made up of: Normally, the output of a nuclear reactor is controlled by changing the flow rate of control rods, which are neutron absorbers, and cooling water, which is a moderator. Output control using control rods is
This is done based on the concept of minimum output ratio as shown below.

Generally, in a boiling water reactor, under normal operating conditions, boiling within the reactor core is in a state of nucleate boiling. When the boiling is nucleate boiling, the heat transfer coefficient is good, so the temperature difference between the cladding tube and the coolant is small, the cladding tube temperature is sufficiently suppressed, and the integrity of the fuel rod is maintained.

However, as the reactor power increases and the heat flux increases, the state shifts from nucleate boiling to transition boiling, and the cladding temperature begins to rise.As the heat flux increases further, the state shifts to film boiling, which can lead to cladding failure. There is a possibility that it will come. This state of transition from nucleate boiling to transition boiling is boiling transition.

The minimum critical power ratio (hereinafter referred to as MCPR) is used as an index representing the thermal margin of a nuclear reactor. This is the ratio of the expected fuel assembly power at which a boiling transition begins to occur and the actual power. That is, the minimum critical power ratio (MCPR).

Therefore, in order to operate the nuclear reactor safely, an operating limit value is set for MCPH, and the reactor is always operated with a certain margin. This MCPH operation limit value (hereinafter OLMCP
As shown in Figure 7, MCPH (hereinafter referred to as SLMCPR) is calculated by adding the maximum value of ΔMCPR (decreasing amount of MCPR) due to various transient changes expected during operation. I'm looking for it.

By the way, BWR plants currently include conventional scram and high-speed scram plants. In conventional scram plants, ΔMCPR tends to be larger at the end of the cycle than at the beginning. This is because at the end of the cycle, the control rods are withdrawn to increase the reactivity of the reactor core, so if a transient event occurs, there is a delay in inserting the control rods even if a scram signal is received. Therefore, in a conventional scram plant, the ΔMCPR is at its maximum when the bypass valve is inoperative during generator load shedding, where the reactivity is most rapidly applied among various transient changes. This event occurs due to a loss of generator load due to a power system accident, etc. To protect the turbine, the turbine steam control valve is abruptly closed and the reactor is scrammed.

Reactor pressure increases due to main steam shutoff, and neutron flux increases due to positive reactivity applied by collapsing voids, but the recirculation pump stops at the same time as the reactor scrams due to the signal of the turbine steam control valve rapid closure. Therefore, the core flow rate decreases rapidly and the number of voids increases rapidly. As a result, the transient increase in neutron flux is suppressed due to the negative reactivity applied by the scram and the negative reactivity of the void. (See Figure 8). On the other hand, in plants that employ high-speed scrams, the scram insertion time is early, so the 8MCPR does not change at the early or late stages of the cycle, and reaches its maximum when the feedwater heating is lost.

This event occurs when an abnormality occurs in the feed water temperature control system and the feed water temperature decreases.The temperature of the coolant at the core inlet decreases and the volume ratio of bubbles in the core decreases, causing the coolant density to increase. The reactor power becomes abnormally high, eventually leading to a scram due to the high neutrons (see Figure 9). Therefore 0I
JCPR is different for conventional scram and high-speed scram; high-speed scram: 01, MCPR = SLMCPR+ (△MCPR at the time of loss of feed water heating), and operation is performed so as not to exceed OLMCPR during operation.

On the other hand, control using the core flow rate is performed by utilizing the characteristic that the core flow rate changes almost in proportion to the output (No. 10).
Flow control curve A) in the figure.

This flow rate control curve is set so that the core can be maintained safely during operation due to safety constraints and the minimum pump speed curve E, which is the lower limit when controlling output by flow rate.

By the way, the safety of boiling water reactors depends on local channel stability and overall core stability.

be. Channel stability refers to the thermal-hydraulic stability within the fuel channel box where oscillations in the channel flow rate flowing within the feed channel box impede heat transfer to the moderator and locally oscillate the reactor power. , the stability of the gas-liquid two-phase flow in the channel box determined by the channel inlet flow rate, the transport delay between the channel pressure drop and the feedback effect. On the other hand, core stability refers to the core average neutron flux stability, which is determined by the transport delay between the neutron flux of the entire core and the amount of voids in the core, and the reactivity feedback effect of the entire reactor. stability.

The first indicator of core stability and channel stability.
The stability reduction ratio x2/Xo shown in FIG. 1 is used. The stability reduction ratio indicates a response capacity relationship to a step-like disturbance, and is expressed as a ratio of overshoot amounts. If the stability attenuation ratio is greater than 1.0, the response becomes a divergent oscillation and becomes unstable. The minimum pump speed is the lower limit when output is controlled by flow rate, and stability is at its strictest at this point.

Therefore, this minimum pump speed is designed so that the stability has a certain margin Y at the maximum output H at the minimum pump speed, as shown in FIG.

However, attempts are currently being made to operate at high output and low flow rate (broken line D) as shown in FIG. 12 in order to expand the operating range. In this case, there is less margin at the minimum pump speed maximum output point H', and the natural circulation maximum output point (■') may enter the unstable region.

Therefore, when driving an alpine car at a low flow rate, Driving Guideline E is set to draw a contour line with a stability reduction ratio of 0.8 (with a margin of 0.2), and to avoid driving at a stability reduction ratio of 0.8 (with a margin of 0.2). I have to. In addition, even if a transient event such as tripping of two recirculation pumps occurs during operation and the state shifts to a natural circulation state, the 5RI (selective control rod) is used to prevent an unstable state.
The design is such that the output is lowered by inserting the

(Problem to be Solved by the Invention) It is assumed that during normal operation, particularly during high-output/low-flow operation, a loss of feed water heating occurs, for example, the feed water temperature drop is not so large (10 to 30° C.). In this case, the output increases gradually, but the change is very slow, and there is a possibility that the output does not increase and stabilizes until the neutron flux reaches a high scram. (Fig. 13 J-)J') At this time, multiple accidents occur, and although it seems to occur very rarely, if a recirculation pump trip were to occur,
According to the output-flow characteristics, point J' enters the unstable region such as H#, H#, etc. Also, if generator load shedding occurs at points J' and N, there is a possibility that OL MCP R will exceed SLMCPR. come out.

The present invention has been made in consideration of the above points, and the flow rate control curve (aWD + b) at the start of feedwater heating loss is observed from the operating point and control rod pattern, etc. The output level of the intercept and the drop in the feedwater temperature are determined. To provide a control device for a nuclear reactor that can keep the core stable at all times during operation and continue operation without exceeding SLMCPR by changing the insertion ratio of SRI depending on the width and adjusting the output increase. With the goal.

[Structure of the invention]

(Means for Solving the Problems) The control rod drive control device according to the present invention uses a feed water temperature detector that detects the range of decrease in feed water temperature, and when a loss of feed water heating occurs, a signal transmitted from the feed water temperature detector is used to regenerate the current time. The recirculation flow control curve is calculated using a recirculation flow control curve calculation program that calculates the operating point and control rod pattern. The present invention is characterized by being equipped with a device that adjusts the output increase by changing the SRI insertion rate based on the relationship between these two points, such as the fact that the larger the temperature decrease, the higher the output increase rate.

(Function) As shown above, the nuclear reactor control device according to the present invention detects the feed water temperature while constantly observing the operating state, so no matter at which operating point loss of feed water heating occurs, it is not necessary to waste water. Always maintain core stability and SLMCPR without suppressing output.
It becomes possible to drive without getting up.

(Example) Regarding the example of the control rod drive control device in the present invention,
This will be explained using FIGS. 1 to 6.

As shown in FIG. 1, a reactor core 2 loaded with a large number of fuel assemblies is formed inside a reactor pressure vessel of a boiling water reactor. The reactor core 2 is immersed in a coolant (moderator W), and a plurality of in-core neutron detectors 11 are disposed in the reactor core 2, which constitute a neutron instrumentation system 10 that measures the reactor power. Ru. The reactor pressure vessel 1 is also connected to a turbine 5 via a main steam pipe 4, and the steam generated by boiling reactor water in the reactor core 2 in the reactor pressure vessel 1 is transferred to the turbine 5 via the main steam pipe 4. The steam that has worked here is condensed into condensate in the turbine condenser 6, and this condensate is heated in the feed water heater 8, and the pressure is raised by the feed water pump 7 as feed water, and the steam is sent through the water feed pipe 9. Circulate into the reactor pressure vessel. On the other hand, the output of the reactor is controlled by a control rod drive mechanism 12 and a recirculation pump 13 that changes the flow rate of cooling water. The control rod drive control device 19 according to the present invention includes a feed water temperature detector 14 that measures the feed water temperature, a signal 10 from the neutron instrumentation system, a signal 15 from the core flow rate instrumentation system 3, and a control rod pattern signal. It has a recirculation flow rate control curve calculation device 17 which calculates a recirculation flow rate control curve at an operating point from 16, and changes the insertion ratio of control rods according to the logic shown in FIG. For example, as shown in Figure 3,
, X2 exemplify the above process. First, it is assumed that a loss of heating of the feed water occurs during operation at point X□ as shown in FIG. Along with this, the water supply temperature gradually decreases. This temperature decrease range ΔTFw is monitored by the feed water temperature detector 14 shown in FIG. 1 and inputted as a signal 18 to the control rod drive controller 19.

On the other hand, when the feed water temperature drop signal 20 is input, the program of the recirculation flow control curve calculation device 17 is started, and the signal from the neutron instrumentation system 10 and the core flow instrumentation system 3 and the control rod drive mechanism are input. The operating point X is determined by the control rod pattern signal.
The recirculation flow rate control curve (alID+b,) at is calculated, and the output at the cutting edge at this time is input to the control rod drive controller 19. The control rod drive control device 19 in the present invention uses these two signals, feed water temperature decrease width ΔTF
The insertion ratio of the control rod is changed depending on the magnitude of W and the output b at the cutting edge, and when the output b at the cutting edge is high as at point X, even if the feed water temperature decrease width is small. Even if multiple failures occur and the occurrence frequency is considered to be extremely low, if a recirculation pump trip occurs, there is a risk that the system will enter an unstable region according to the broken line in FIG. 3. Therefore, it becomes necessary to insert SRI. In other words, according to the logic in Figure 9, since the cutting edge b1 is large, △T
Even if the FW is not very large, SRI requires moderate insertion. On the other hand, even if feed water heating loss occurs during operation at X2 in Figure 3, the recirculation flow rate control curve a Wd +
Since b2 has sufficient margin compared to the recirculation flow rate side mouth groove curve A of the 100% control rod pattern F in normal operation, there is a possibility that it will fall into the unstable region even if the feed water temperature drop width △TFυ2 is somewhat large. . Therefore, it is only necessary to change the insertion ratio according to the stage of this temperature reduction width ΔTFW2, and since the larger the reduction width becomes, the more severe the SR process insertion ratio is. The above concept is briefly shown in FIGS. 4, 5, and 6. In the figure, the vertical axis (left side) is the rate of reactivity of the control rod, the vertical axis (right side) is the magnitude of the output of the cutting edge at each operating point, and the horizontal axis is the feed water temperature decrease width △TF. shows. For example, when the operating range is expanded, during operation on the flow control curve D (see Figure 3) on the 100% control rod pattern, the ratio shown in the shaded area in Figure 4 (when the cutting edge is at its maximum) is The reactivity of the control rod is determined according to the range of decrease in the feed water temperature.

In other words, during operation on the flow rate control curve, even if the feed water temperature does not drop below IO'lq, □ is put in the reactivity Δ, and then 20
8 at ℃, 2, 3 at 30℃, and a 50℃ drop leads to a scrum. If driving at point X1, b.

is rather large, so as shown in Figure 5, when the temperature is lowered by 10°C, the a control rod is inserted into Δ, and the reactivity is then added as shown in perspective in the figure. On the other hand, at point X2, as shown earlier, b2 is not so large, so normal operation is performed until the temperature drops by 30°C as shown in Figure 6. When the temperature drops by 30°C, SRI is started to be inserted, and the reactivity of 3b is shown in Δ. Put in. By moving parallel to the size of the cut edge in this way, it is possible to change the temperature and reactivity ratio at the start of SRI insertion, and maintain the safety of the reactor during operation without unnecessary output control. Become.

〔Effect of the invention〕

As described above, the reactor control system of the present invention is capable of operating when a loss of feedwater heating occurs such that the feedwater temperature drop is not so large, the output leading to scram does not increase, and the output stabilizes at a constant level. By adjusting the power increase by changing the SRI insertion rate based on the relationship between the flow rate control curve at the point and the width of the feed water temperature decrease, it is possible to maintain core stability during operation and continue operation without exceeding the SLMCPR. becomes possible.

[Brief explanation of the drawing]

FIG. 1 is a schematic diagram of a nuclear reactor control device according to the present invention, FIG. 2 is a flowchart according to the present invention, FIG. 3 is an output-core flow rate characteristic diagram according to the present invention, and FIG. 4 is a water supply according to the present invention. Figure 5 is a diagram showing the temperature decrease range vs. control rod reactivity ratio characteristic. Figure 5 is a control rod reactivity ratio characteristic diagram at the X operation point according to the present invention. Figure 6 is a control rod reactivity ratio characteristic diagram at the X2 operation point according to the present invention. Fig. 7 is a conventional MCPR characteristic diagram, Fig. 8 is a conventional output change characteristic diagram, Fig. 9 is a conventional scram characteristic diagram,
Figure 0 is a conventional power-core flow rate characteristic diagram, Figure 11 is a definition diagram of the width reduction ratio, Figure 12 is a conventional power-core flow rate characteristic diagram, and Figure 13 is a conventional power-core flow rate characteristic diagram. . 1 - Reactor pressure vessel 3 Core flow bar 1 system 5 Turbine 7... Water supply pump 9 Water supply piping 11 - In-reactor neutron detector 2... Reactor core 4 - Main steam pipe 6 - Condenser 8 Feed water heating Equipment 10...Neutron instrumentation system 12...Control rod drive mechanism 13...Recirculation pump 14...Feed water temperature detector 1
7... Recirculation flow control curve calculation device 19... Control rod drive control device

Claims (1)

[Claims] A feedwater temperature detector monitors the reactor feedwater temperature and detects the range of decrease from the temperature in normal operation, and a recirculation flow control curve is calculated from the operating point and control rod pattern at that time based on the signal from the feedwater temperature detector. By using a recirculation flow rate control curve calculation device and a signal from this recirculation flow rate control curve calculation device, the insertion ratio of selected control rods is changed based on the relationship between the flow rate control curve and the width of the feed water temperature decrease when feedwater heating loss occurs. A nuclear reactor control device comprising a control rod drive control device that prevents a rise in output.