JPS60225092A - Emergency core cooling device - 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 [Field of Application of the Invention] The present invention relates to a boiling water nuclear reactor, and relates to an emergency core cooling system suitable for operating in the event of a loss of coolant accident and efficiently cooling the core.

[Background of the invention]

The emergency core cooling system (ECC8) of a conventional boiler-powered nuclear reactor is composed of a high-pressure spray system, a low-pressure spray system, a low-pressure water injection system, and the like. Figure 1 shows the state of core cooling by the low-pressure water injection system in the event of a loss of coolant accident due to a pipe rupture.

The low-pressure water injection system consists of a pipe 51 and a pump 52, and in the event of a loss of coolant accident, unsaturated cooling water is injected from the pipe 51 into the bypass region 4 between the shroud wall 11 and the fuel assembly 30 in the pressure vessel 1. The area is flooded, and the inside of the lower plenum 2 and fuel assembly 3 is also flooded.

The flow path through which the cooling water injected into the bypass region 4 by the ECC 8 flows into the fuel assembly 3 and the lower blennium 2 will be explained with reference to FIG. 2. The lower tie plate 31 at the lower end of the fuel assembly 3 has a leak hole 32, and the cooling water in the core bypass region 4 flows through the leak hole 32 to the fuel assembly 3.
flow inside.

On the other hand, a fuel support 33 that supports the fuel assembly 3 is placed on a core support plate 41, and has an inlet orifice 2 at its lower end.
There is 1. The cooling water that has flowed into the integrated body 3 flows into the lower plenum 2 through this orifice 21 and falls.

As shown in FIG. 1, the cooling water 71 flowing into and falling into the lower plenum 2 through the inlet orifice 21 is suppressed from falling by the steam 72 blown up from the lower plenum 2 by boiling under reduced pressure (this phenomenon is called the CCFL phenomenon). A part of the kite assembly 3 is accumulated inside the kite assembly 3. Therefore, the interior of the assembly 3 is flooded before the lower plenum 2 is flooded, making it possible to achieve early cooling.

Such an effect is particularly noticeable when saturated water flows in from the leak hole 32. However, when sub-cooled water flows in from the leak hole 32, steam condensation occurs, so even with the same upward steam flow rate, the falling water flow rate increases, a so-called CCFL break, which causes the cooling water in the assembly 3 to flow into the lower plenum. 2, resulting in a delay in re-flooding of the reactor core. Koushi 90 C1” About L% property 3rd
As shown in the figure. The horizontal axis is the blow-up steam flow rate, and the vertical axis is the falling water flow rate, with the solid line representing saturated water and the broken line representing unsaturated water. As is clear from the figure, for a constant blow-up steam flow rate, the falling water flow rate is larger in the case of subcooled water than in the case of saturated water. This causes the cooling water in the assembly 3 to fall into the lower plenum 2, delaying the re-flooding of the core.

[Purpose of the invention]

The purpose of the present invention is to maintain the temperature of the cooling water flowing from the bypass through the leak hole to the lower part of the assembly at a saturation temperature corresponding to the pressure of the system during operation of the low-pressure core water injection system, thereby reducing the amount of cooling water accumulated in the assembly. An object of the present invention is to provide an emergency core cooling system that prevents water from falling into a lower plenum by blowing up steam and achieves early re-flooding of a reactor core.

[Summary of the invention]

In the present invention, the temperature of the cooling water in the bypass area is measured, and the saturation temperature corresponding to the pressure in the pressure vessel is compared, and the low-pressure core water injection system is adjusted so that the temperature in the lower part of the bypass does not become unsaturated. The flow rate is controlled to maintain the temperature of the cooling water flowing from the bypass through the leak hole to the lower part of the assembly at the saturation temperature. In order to do so, the CCFL is maintained at the lower orifice with the lower assembly, and cooling water within the assembly is prevented from falling into the lower blennium.

[Embodiments of the invention]

Hereinafter, the present invention will be explained in more detail with reference to Examples. FIG. 4 is a schematic diagram showing an embodiment of the present invention. Since the parts given the same numbers as in FIG. 1 have the same functions, their explanations will be omitted.

The emergency core cooling system according to the present invention includes a pipe 51 that supplies cooling water to the core bypass region, a pump 52, a drive motor 54, and a flow rate control device 60. The flow rate control device 60 includes a temperature measuring device 61 .
A pressure measuring device 62 in the pressure vessel is used as a measuring device, and a comparator 63 compares the coolant temperature in the bypass region with a coolant saturation temperature corresponding to the pressure. It is a feedback control system that controls the supply flow rate of cooling water by changing the rotation speed. In an emergency, unsaturated water is supplied through the entire pipe 51 through an opening 53 in the peripheral wall above the bypass area.

As this cooling water falls through the bypass area and flows toward the center, it is given heat by heat transfer from the structural materials and fuel assembly, and its temperature increases. In the case of a BWR5 plant, the temperature increases by about 10 C during the movement of cooling water from the mid-height position of the bypass center area and the lower position of the bypass peripheral area to the lower position of the bypass center area. Therefore, by controlling the flow rate so as to keep the temperature at these locations approximately 10 C lower than the saturation temperature corresponding to the pressure of the pressure vessel, the coolant temperature at the lower portion of the bypass central region can be maintained at the saturation temperature. As a result, unsaturated water flows in from the leak hole 32 and CC occurs at the aggregate inlet orifice 21.
FL break can be prevented.

On the other hand, pressure measurement can be performed at any arbitrary position because the pressure inside the pressure vessel is approximately uniform.

In addition, the saturation temperature for that pressure can be calculated using a simple table or mathematical formula.

The temperature measuring device in the bypass area consists of a thermocouple.
Instrumentation piping 70 penetrating from the lower end of the pressure vessel to the bypass area
Insert it inside and measure the temperature. It is desirable to periodically take out and inspect this thermocouple to maintain high reliability.

The relationship between the input signal and the output signal in the comparator 63 is shown in FIG. The input signal is a temperature difference, and the saturation temperature T m
a Represents the difference between S and bypass temperature T.

The output signal is the amount of change ΔN in the rotational speed of the flow pump drive motor 54, and is expressed as the difference between the rotational speed at a new point in time after control and the current rotational speed.

7N=9 New point - Current point The control method is as shown in the figure, as ∆T increases, ∆N
Use methods to reduce That is, when the bypass temperature decreases and ΔT increases, ΔN becomes a negative value, and the injection flow rate decreases while the motor rotational speed is lower than the current value.

As a result, the temperature increases due to heat exchange in the bypass area,
Control can be performed so that the increase in ΔT is suppressed and maintained at a constant value (approximately 10c).

FIG. 6 shows another embodiment of the invention. In this embodiment, the temperature of the cooling water in the lower brenum is used as the saturation temperature corresponding to the pressure inside the pressure vessel. In the lower plenum, boiling under reduced pressure continues as the system pressure decreases, and the temperature of the cooling water is maintained at approximately the saturation temperature. Therefore, the temperature is directly measured and used for flow control. The other configurations are the same as in the previous embodiment.

FIG. 7 shows another embodiment of the invention. In this embodiment, a flow rate adjustment valve 55 is used as a method of changing the flow rate of the emergency core cooling system. In this embodiment, the control signal from the control system is the amount of change ΔA in the valve opening A.
, and there is a relationship of 0 = "new time point -" current time.

The relationship between ΔA and temperature difference ΔT is shown in FIG. A method is used in which ΔA decreases as ΔT increases. That is,
When the bypass temperature decreases and ΔT increases, ΔT becomes a negative value, the opening degree of the valve becomes smaller than at present, and the injection flow rate decreases. As a result, the temperature increases due to heat exchange in the bypass area, suppressing the increase in ΔT, and keeping it at a constant value (approximately 10C).
can be controlled so that it is maintained at

FIG. 9 shows an example of the effects resulting from the implementation of the present invention. At the time of a loss of coolant accident due to a guillotine rupture in a recirculation system piping, the surface temperature of the cladding changes as shown in Figure 9. In the example of the temperature change in the conventional technology shown by the broken line, the CCF after about 120 seconds
The temperature continued to rise due to the L break, and after the lower plenum was flooded, the assembly was submerged again and the temperature decreased. On the other hand, in the case of the present invention shown by the solid line, the cooling water flowing from the bypass region through the leak hole into the aggregate is saturated water, so the CCFL
No break occurred, and the aggregate re-flooded before the lower Blenheim re-flooded, reducing the maximum cladding temperature by approximately 2
It is possible to suppress the temperature to a low 00C.

In addition, the rated flow value of the low-pressure water injection system is 1700 t/hr/unit in the case of a BWR5 plant, and the rotation speed is 3500 F-, but the temperature in the lower part of the bypass area is maintained at the saturation temperature until it is submerged again after reaching the rated flow rate. The flow rate of the reservoir is approximately 6% of the rated flow rate.
0% means dust, which is within the controllable range of the style control system.

〔Effect of the invention〕

According to the present invention, in the event of a loss of coolant accident in a boiling water reactor,
By keeping the temperature of the cooling water flowing into the fuel assembly from the core bypass area through the leak hole at the bottom of the fuel assembly at the saturation temperature, CCFL breakage at the assembly inlet orifice is prevented and a large amount of cooling water is accumulated within the assembly. This allows for early re-flooding. As a result, the maximum temperature of the cladding tube in the event of a loss of coolant accident can be kept low, and the safety of the reactor can be improved.

In addition, although the current BWR plant does not measure the temperature in the butterfly bypass area, it has the secondary effect of being able to measure the temperature during normal operation and use it as a monitor for plant abnormalities such as flow instability.

Furthermore, even in the event of a reactor accident, it is possible to obtain essential data such as temperature changes in the reactor core, which has a large effect on its use.

[Brief explanation of drawings]

Fig. 1 is a schematic diagram illustrating a conventional emergency core cooling system, Fig. 2 is a partial sectional view showing the flow path in the lower part of the core, Fig. 3 is a CCFL characteristic diagram at the inlet orifice, and Fig. 4 is an implementation of the present invention. A schematic diagram showing an example, FIG. 5 is a diagram showing control characteristics of the embodiment in FIG. 4, and FIG. 6 is a schematic diagram showing another embodiment of the present invention.
FIG. 7 is a schematic diagram showing yet another embodiment of the present invention;
FIG. 7 is a diagram showing the control characteristics of the embodiment, and FIG. 9 is a diagram showing the time change of the cladding surface temperature showing the effects of the present invention. DESCRIPTION OF SYMBOLS 1... Pressure vessel, 2... Lower plenum, 3... Fuel assembly, 4... Bypass area, 11... Shroud wall, 21... Core inlet orifice, 31... Lower type Rate, 32... Leak hole, 33... Fuel support, 41... Core support plate, 51... Low pressure water injection system piping, 52... Pump, 54... Drive motor, 6
0...Flow rate control device, 55...Flow rate adjustment valve, 61.
...Temperature measuring device, 62... Pressure measuring device, 63... Temperature comparator, 70... Piping for temperature instrumentation. Agent Patent Attorney Tatsuyuki Unuma 1121 γλ Figure 13I21 Fukiage Su゛Shoki ζTanban (tc3/, f) 14 Mouth Figure 5 Moxibustion of the number of revolutions ΔN (-Nli 1 Ginmoto-N1 Rimo-mura) No. 2 Figure A 7 Figure χ δ Mugii Otodo of mouth valve opening

Claims (1)

[Claims] 1. A thermometer that measures the temperature of the coolant in the core bypass region in an emergency core cooling system that injects coolant into the core bypass region of a nuclear reactor and drops it into the lower plenum through a leak hole in the event of an accident. and a measurement system for measuring a saturation temperature corresponding to the pressure value in the pressure vessel, and a circuit for comparing the saturation temperature corresponding to this pressure value with the coolant temperature in the bypass area, and a circuit that compares the saturation temperature corresponding to the pressure value with the coolant temperature in the bypass area, and measures the temperature difference between them. An emergency core cooling system characterized in that the flow rate of the coolant injected into the reactor is controlled to maintain the temperature of the coolant flowing from the leak hole at the saturation temperature. 2. In claim 1, the measurement system for measuring the saturation temperature corresponding to the pressure value in the pressure vessel includes a pressure gauge for measuring the pressure in the pressure vessel and a conversion of the saturation temperature for the pressure measurement value. An emergency core cooling system characterized by comprising a circuit that can be recharged. 3. In claim 1, the measurement system for measuring the saturation temperature corresponding to the pressure value in the pressure vessel includes a thermometer for measuring the temperature of the cooling water in the lower plenum, and determining the saturation temperature from the temperature value. An emergency core cooling system characterized by an empty circuit.