JPS607161B2 - Nuclear plant water supply methods and equipment - Google Patents

Description

DETAILED DESCRIPTION OF THE INVENTION The present invention relates to a method and apparatus for supplying feed water to a nuclear power plant having a once-through steam generator system.

For convenience of explanation, the prior art and its problems will be explained using a fast breeding power plant having a Y-coupled separated once-through steam generator system as an example. In such a fast breeder reactor power plant, the heat generated in the reactor is transferred to primary sodium (hereinafter referred to as primary Na) which is circulated through the primary cooling system that cools the reactor core by a primary sodium pump. This is then transferred to the secondary sodium (called secondary Na) which is also circulated through the secondary cooling system in the same manner as the primary cooling system via an intermediate heat exchanger installed outside the reactor. be done.

The secondary Na that has completed heat exchange in the intermediate heat exchanger is divided into a superheater and a reheater, where it is subjected to heat exchange with steam, and then combined again, and then led to an evaporator where it is converted into steam. After completing the heat exchange, the secondary Na pump returns to the intermediate heat exchanger, and this operation is repeated. The system consisting of the evaporator, superheater and reheater described above is
Although it is called a coupled-separated once-through steam generator system, the feed water supply to this system is generally carried out by a turbine-driven water pump system (hereinafter referred to as BFP), which is driven by steam diverted from the main turbine that drives the generator. -T system) and an electric motor-driven starting water supply pump system (hereinafter referred to as starting BFP-M system).

The superheated steam generated by the evaporator and superheater and the steam reheated by the reheater are each guided to the main turbine to perform work and generate electricity.

The steam that has completed its work in the main turbine and BFP-T is led to a condenser, cooled and condensed, and then used again as feed water for the steam generator system. In general, in nuclear power plants, in order to ensure that the reactor decay heat is removed after the reactor has scrammed, the plant's output is rapidly reduced at the same time as the reactor is scrammed, and the plant is temporarily operated independently at the plant's internal load. The so-called fast cut back (Fast Cut mother ck) operation method is not adopted.
In this case, the turbine generator is tripped and the supply of driving power for the cooling system power equipment is switched from receiving power from the external power line.
Alternatively, it is common to switch to an emergency power source (generally a diesel generator is used) and perform decay heat removal operation. On the other hand, a two-system power transmission system is normally used, but if a system load shedding accident occurs simultaneously in this two-system system, it will be an external power loss accident for the nuclear power plant. All power equipment trips and loses all functionality while the emergency power source is turned on. In a fast breeder reactor power plant, when an external power loss accident occurs in the same machine as mentioned above, the entire system of the plant trips, so the primary Na pump and secondary N
The a pump also trips and the flow rate coasts down.

However, as for the Na pump, the drive is immediately switched to a small electric motor driven by an uninterruptible power source (storage battery, etc.) after tripping to ensure a sufficient Na flow rate to remove decay heat. At the same time, the emergency power supply is turned on and the auxiliary core cooling system (hereinafter referred to as ACCS) is activated, putting it into a standby state for heat removal.
Since it takes a relatively long time of several minutes until the ACCS can perform the decay heat removal operation, the evaporator and superheater perform the decay heat removal operation during this time.

Note that the reheater does not contribute to the decay heat removal operation because it is isolated from the secondary cooling system by rapidly closing the Na flow control valve when the main turbine trips. On the other hand, in the event of an external power loss accident, the BFP-T also trips because the condenser loses its cooling capacity. As a result, the water supply to the evaporator and superheater is cut off within a few seconds after the accident occurs. On the other hand, as soon as an accident occurs, the power supply of the startup BFP-M is switched to an emergency power supply in order to quickly restore the water supply capacity, but it takes several tens of seconds before the BFP-M can supply water. That is, when an external power loss accident occurs, there is no water supply to the evaporator or superheater for several tens of seconds after the accident occurs.

As a result, since the evaporator and superheater are once-through type and have a small amount of internal water, they suddenly become dry-out (referred to as dry-out), and the steam pressure rises rapidly, and the temperature of the heat transfer tubes also rises to about the Na temperature. . However, after several tens of seconds, sufficient water is supplied by the starting BFP 1M to remove the decay heat, so the heat exchanger tube is completely turned from the dry-out state and is rapidly cooled down. For this reason, the heat exchanger tubes and their supporting structures are exposed to severe thermal and mechanical conditions, and the evaporators,
The service life of the superheater may be significantly reduced. '
2 - As a result of the main turbine tripping, heat exchange in the reheater is almost instantaneous, so until the Na flow control valve of the reheater is fully closed, high-temperature Na is removed from the reheater. flows out. Furthermore, as a result of the water supply to the steam generator being interrupted, heat exchange cannot be performed in the superheater during this time, so Na at approximately the same high temperature flows out from the superheater as in the case of the reheater. Furthermore, since these phenomena occur almost simultaneously, the Na temperature at the evaporator inlet may rise beyond the design temperature of the evaporator, which may have a significant adverse effect on the service life of the evaporator. [3' When supplying water by BFP-M to a place where the evaporator or superheater is in a dry-out state and the steam pressure is rising, the steam pressure may fluctuate, although it can be controlled to some extent by the main steam relief valve or safety valve. Because of the large amount of water, there is a risk that the water supply will not be stable. (4) Furthermore, as a result of rapidly supplying water to the evaporator in a dry-out state, a thermal-hydraulic flow instability phenomenon may occur, and the decay heat removal operation by the steam generator may become unstable. We have explained the problems in detail using a fast breeder reactor plant with a combined steam generator system as an example, but these problems are common to nuclear power plants with once-through steam generator systems, and It is not affected by the configuration of the evaporator system, such as integrated evaporator/superheater type, separated type, reheat type, or non-reheat type.

Further, the present invention can be applied to a nuclear power plant whose main purpose is not to generate electricity, as long as the plant has a once-through steam generator system and a turbine generator. The purpose of the present invention is to prevent the evaporator and superheater from drying out in the event of an external power loss accident in a nuclear power plant having a once-through steam generator system and a turbine generator, and to operate the decay heat removal operation of the reactor. It is an object of the present invention to provide a method and a device for stably performing the heat exchanger and suppressing the thermal shock applied to the heat exchanger tubes and structures of the evaporator and superheater.

Embodiments of the present invention will be described in detail below with reference to the drawings.

FIG. 1 is a flow diagram of a fast breeder reactor power plant incorporating a device of the present invention having a Y-coupled once-through steam generator.

In the figure, 1 is a nuclear reactor, 2 is a primary Na pump, 3 is an intermediate heat exchanger, 4 is a secondary Na pump, 5 is a superheater, 6 is a reheater, 7 is an evaporator, 8 is a main turbine 9 1 is a generator, 1 is a condenser, 11 is a deaerator, 12 is a BFP-T, 13 is a B
It is FP-M. Since these devices have been explained previously, their explanation will be omitted here. In the present invention, in addition to the conventional BFP-T 12 and starting BFP-M 13, the water supply system is connected to a water supply pump 15 that is driven to the rotor main shaft of the main turbine 8 via a joint 14.

The type of coupling 14 may be a direct coupling, a gear drive, or a variable distance drive such as a fluid coupling. The water supply pump 15 is driven by the rotational power of the main turbine during normal operation, and supplies all or part of the water supply to the evaporator 7 . Its capacity shall be at a minimum sufficient to ensure a flow rate of water supply sufficient to remove reactor decay heat after an external power loss accident occurs. When an external power loss accident occurs, the water supply pumps other than the water supply pump 15 connected to the turbine generator rotor shaft lose their driving source, so they immediately trip and lose their water supply capacity.

However, since the water supply pump 15 is driven by the large rotational inertia of the turbine generator rotor, it does not stop rotating instantly, but performs decay heat removal movement for several tens of seconds until the emergency diesel generator is started. Sufficient feed water can be supplied to the steam generator. As a result, there is no risk of the evaporator or superheater drying out due to water supply cutoff, and heat exchange continues in the evaporator and superheater, so the Na temperature at the evaporator inlet will never exceed the design temperature of the evaporator. There isn't. In the present invention, if the steam generator is an integrated once-through type or a separated once-through type and ventilation is provided to the superheater 5, in order to keep the main steam pressure constant, the main steam pipe 16 leading to the turbine generator is A main steam pressure control device consisting of a pressure gauge 17 that measures the pressure of the main steam, a pressure controller 18 that receives and controls the pressure signal, and a pressure control relief valve 19 is provided in front of the main steam relief valve 20.

In addition, if the steam generator is a separate once-through type and steam that does not meet the ventilation conditions for the superheater 5 flows out from the evaporator 7, and the superheater 5 is isolated, the steam separator A pressure gauge 22 that measures the pressure inside 21, a pressure controller 23 that receives the signal and controls the pressure, and a steam dump valve 2.
4. Provide a steam separator pressure control device consisting of a drain valve 26.

When controlling the main steam pressure and the steam-water separator pressure at a constant level using the above pressure control method and control device, the water supply to the steam generator after an external power loss accident will drop almost linearly as time 0 elapses. I can do it. The reason for this will be explained below using a mathematical formula. Note that in the following description, the steam generator will be described as a separate once-through type for the sake of simplicity.
After an external power loss accident occurs, the relationship between the kinetic energy possessed by the turbine-generator rotary rut-connected water supply pump 15 and the water supply flow rate can be expressed by the following equation, ignoring mechanical loss.

-ZWdW=day.

W. dt (1'g 0 here J: moment of inertia kg force g: gravitational acceleration skin/sec2 w: water supply pump rotational angular velocity rad/secH: water supply pump discharge water head W: water supply flow rate kg/
sect: Time Sec Also, the relationship between the water supply pump discharge head, rotational angular velocity, and water supply flow rate can be expressed by the following equation.

H=A・V2−B・W2 ■0 where A, B; positive constant On the other hand, the relationship between the main steam pressure or the steam pressure at the steam/moisture broiler outlet, the water pump discharge head, and the pressure loss is approximately expressed by the following equation. Express.

H=Top (p+△P)'3' y △P=C・W2 '4) Here P: Main steam pressure or steam water and 9/〆Steam pressure at separator outlet △P: Pressure loss k9/mey : Specific weight of water supply k9/elbow C: Positive constant expression (Substituting 3' and {4} into the expression [2 and rearranging it,
We get the following equation.

P=H': Aw2-BW2 '2}' y where H: Water head corresponding to pressure P B: Positive constant On the other hand, by substituting formula {3) into the right side of formula {1), formula t2)'
When rearranged using -萱WdW=(H+Total P)●W-d
t.

However, during normal operation, m' can be approximated by the following equation, since the rod is H'.

−ZWdW=日・. W. dt or g Here, since H' is a substantially constant value because constant pressure control is performed, the following equation is obtained by differentiating both sides of equation {2}' with respect to time.

AW・group−B′w group=o′2) Therefore, by eliminating w from equations m'' and '2}'', the following equation is obtained.

dW dt=-D・H' (5} However, D is a positive constant and can be expressed by the following equation.

By integrating the D〒band '6' equation {51 with respect to time, the following equation is finally obtained.

W=-D・H′・t+W.

'7) Here, Wo is an integral constant, and is the water supply flow rate supplied by the turbine generator rotor shaft connected water supply pump at the moment when the external power loss accident occurs. In other words, the water supply flow rate after the external power loss accident (Formula '7)
) will fall almost linearly over time as shown in Figure 2, so water can be stably supplied and the decay heat removal operation by the evaporator and superheater can be performed smoothly. As shown in Figure 2, even if an external power loss accident occurs, we explained that a stable water supply flow rate can be ensured. As a result of the feed water flow being large compared to the secondary Na flow rate, there is a possibility that supercooling may occur in the evaporator.

In order to further prevent this, the present invention includes a thermometer 28 at the steam outlet of the evaporator 7 and a temperature controller 29 that receives the signal.
A temperature control device consisting of the following is provided, and the water supply control valve 30 is opened and closed according to the output of the controller 29 to adjust the flow rate of the water supply. In the above explanation, the steam temperature at the evaporator outlet was taken as an example, but in addition, a temperature signal obtained by any one of the evaporator outlet evaporator tube plate temperature, the evaporator inlet Na temperature, or an appropriate combination of the three can be used in the temperature controller. 29 may feed back to the water supply control valve 30 or the variable joint 14 to control the water supply flow rate. FIG. 3 shows the temporal change in the thus obtained feed water flow rate, and it can be seen that the feed water flow rate is close to the secondary Na flow rate, and that the evaporation temperature at the evaporator outlet can be controlled to be almost constant.

This not only prevents overcooling of the evaporator 7 but also suppresses thermal shock applied to the heat exchanger tubes and structures of the evaporator and superheater. In the above description, the type of turbine generator in a fast breeder reactor power plant is shown as a tandem compound, but a cross compound may also be used, in which case two systems of the water supply system of the present invention may be provided.

[Brief explanation of the drawing]

Fig. 1 is a flow diagram of a conventional fast breeder reactor power plant system incorporating the arrangement of the device of the present invention, and Figs. 2 and 3 show a case where the evaporator outlet steam temperature is not controlled and a case where the steam temperature at the outlet of the evaporator is controlled in the device of the present invention. It is a graph which shows the temporal change of the water supply flow rate and Na flow rate in the case. 14...Fitting, 15...Water pump, 16...
...0 Main steam pipe, 17 ... Pressure gauge, 18.
...Pressure controller, 19...Pressure control relief valve, 20...Main steam relief valve, 21...
・Steam water separator, 22... Pressure gauge, 23...
Pressure controller, 24... Steam/moisture broiler steam dump valve, 25... Flash tank, 26... Steam/water separator steam drain valve, 27... Bypass relief valve, 28.
...Thermometer, 29 ...Temperature controller, 30
...Water supply control valve. 2 figures of spear, 3 figures of power, 1 figure

Claims (1)

[Claims] 1. In a nuclear power plant having a once-through steam generator system, all or part of the amount of water to be supplied is borne by a water supply pump that is connected to and driven by the rotor shaft of a turbine generator and has a large rotational inertia. , When an external power loss accident occurs, the main steam pressure is controlled at a constant level while the superheater is vented, and after the superheater is isolated, the water and steam pressure in the steam separator continues to be controlled at a constant level. A method for supplying water to a nuclear power plant, further comprising controlling the steam temperature at the outlet of the evaporator to stabilize the flow rate of the water and prevent overcooling of the steam generator. 2. In a nuclear power plant with a once-through steam generator system, the water supply pump device is connected to and driven by the rotor shaft of the turbine generator and bears all or part of the water supply flow rate, and the superheater is vented in the event of an external power loss accident. a main steam pressure control device that controls the main steam pressure to a constant level while the superheater is being isolated; a steam separator pressure control device that controls the steam pressure in the steam separator after the superheater is isolated; A water supply system for a nuclear power plant, characterized in that a steam temperature control device is provided at the evaporator outlet.