JPS6150278B2 - - Google Patents

Description

[Detailed description of the invention]

The present invention relates to a boiling water reactor (BWR),
In particular, the present invention relates to an operation control method and apparatus for controlling the output distribution. In a BWR plant, steam generated within the reactor pressure vessel is supplied directly to a turbine, and a generator is connected to the turbine. Inside the reactor pressure vessel, there is a reactor core loaded with a large number of fuel assemblies. The cooling water is supplied into the reactor core from below the reactor core, and while rising within the reactor core, it is heated and turned into steam while cooling the fuel rods in the fuel assemblies. After passing through a steam separator and dryer above the core, the steam is sent from the reactor pressure vessel to the turbine. After the steam that gives rotational power to the turbine is discharged from the turbine, it is condensed in a condenser. This condensed water is supplied back into the reactor core through a jet pump in the reactor pressure vessel, which is heated by steam extracted from the turbine in a feedwater heater. Generally, nuclear reactors are designed so that the power distribution in the core is as flat as possible in order to maintain the integrity of the fuel rods. However, in a BWR plant, the state of extraction of steam from the turbine differs between rated output and partial output. Therefore,
During partial power such as startup (low flow rate and low power), sub-cooling at the BWR core inlet becomes larger than during rated power due to heat balance, resulting in a large power peak at the bottom of the core. . The threshold value at which the control rods can be pulled out of the reactor core and the output increased while preventing damage to the fuel rods and maintaining their integrity has been determined through experiments based on the linear thermal power density.
It is known that it is 8kw/ft. Power increases at linear thermal power densities of 8 kw/ft or higher are achieved by increasing the flow rate of cooling water flowing through the reactor core. At startup, the linear heat power density exceeds 8kw/ft due to the large power peaking in the lower part of the core. For this reason, if control rods are withdrawn from the reactor core at startup to increase output, control rod withdrawal must be stopped when the area of high output peaking at the bottom of the reactor core reaches 8kw/ft.
Thereafter, U.S. Patent Application No. 762248 (Japanese Unexamined Patent Publication No.
141990), the flow rate of cooling water flowing through the reactor core is increased to increase the linear thermal power density to an arbitrary value, and xenon, a fission product, is accumulated in the fuel rods and cooled. Reduce water flow rate to reduce output. In the presence of the xenon accumulated through this operation, the control rods are withdrawn from the reactor core, and the linear thermal power density is increased again to 8 kw/ft. That is, such operations are repeated. As mentioned above, when the power peaking in the lower part of the reactor core is large, the amount of control rods that can be pulled out at one time decreases, and the number of operations described above increases in order to increase the reactor output to the rated output. . The operation to increase the output when starting the BWR is extremely complicated, and it takes a long time for the output to reach the rated output. Therefore, the utilization rate of BWR decreases. The present invention aims to eliminate these drawbacks and improve the utilization rate of a BWR plant, and aims to improve the steam pressure in the reactor vessel, the flow rate of cooling water flowing through the reactor core in the reactor vessel, and the reactor. In a boiling water reactor operation control method in which heated feed water is supplied to the reactor vessel by measuring the flow rate of steam supplied from the vessel to the turbine, the temperature of the feed water is measured;
The first feature is that the enthalpy of the cooling water at the inlet of the reactor core is set in advance, a desirable feed water temperature is calculated from the measured and set value, and the feed water temperature is controlled to the desired value. , a first measuring means for measuring the steam pressure in the reactor vessel, a second measuring means for measuring the flow rate of cooling water flowing through the reactor core in the reactor vessel, and steam supplied from the reactor vessel to the turbine. a third measuring means for measuring the flow rate of the cooling water, a setting means for setting the enthalpy of the cooling water at the inlet of the reactor core, and a means for adjusting the enthalpy of the feed water supplied into the reactor vessel; , the pressure measured by second and third measuring means,
Based on the cooling water flow rate, the steam flow rate, and the enthalpy of the cooling water set by the setting means, H _fw = Hf (1-Wc/W _STM ) + H _IN (Wc/W _STM
) Here, H _fw : Enthalpy of feed water (cal/g) Hf : Enthalpy of saturated cooling water (cal/g) Wc : Flow rate of cooling water in the reactor core (t/h) W _STM : Flow rate of steam sent to the turbine ( t/
h) A second feature is that the feed water enthalpy adjustment means is controlled so as to satisfy the relationship: H _IN :Enthalpy at the entrance of the reactor core (cal/g); The system includes a water supply system and a turbine connected to the reactor vessel, the water supply system having a heater for feed water fed back to the reactor vessel, and a heater for measuring the dome pressure P _R of the reactor vessel. A boiling water reactor having a first measuring device, a second measuring device for measuring the cooling water flow rate Wc in the core, and a second measuring device for measuring the steam flow rate W _STM sent to the turbine. In the operation control device, the feed water temperature T at the inlet of the reactor core is
The steam pressure W _R , the cooling water flow rate Wc, the steam flow rate W _STM , and the enthalpy H _IN of the cooling water at the inlet of the reactor core are determined by these values. The measured temperature T and the desired temperature Ts are provided to a comparator that calculates the desired feed water temperature Ts in thermal equilibrium, and the measured temperature T and desired temperature Ts are provided to a comparator.
A third feature is that a regulator is provided to adjust the heater according to the output of the comparator. Here, we will explain the results of examining the characteristics of the core of a BWR operated based on the conventional operation control method. The changes in reactor output during BWR startup will be explained based on Fig. 1. The horizontal and vertical axes of this figure show the cooling water flow rate (%) in the reactor core and the reactor power (%), respectively. When the flow rate flowing through the reactor core is held constant at 20% and the control rods are withdrawn from the reactor core, the reactor power begins to rise from point _A1 .
When the reactor power reaches the _A2 point, the control rod withdrawal operation is stopped. Point _A is the reactor power corresponding to the linear thermal power density of 8kw/ft of the fuel rods loaded in the reactor core. If the reactor power is increased beyond the _A2 point by withdrawing the control rods, there is a greater risk that the fuel rods will be damaged. After control rod withdrawal is stopped, the flow rate of cooling water flowing through the reactor core is increased. The reactor power increases along the line A ₂ − _{A 3} . The increase in reactor power between A ₂ and _{A 3} is performed below the critical rate of increase that causes fuel rod failure, as described in US Pat. No. 4,057,466. When the reactor output reaches point _A3 , the cooling water flow rate is reduced. Reactor power drops to _A4 point. The reason why the reactor power decreases to point A ₄ is due to the production of xenon within the fuel rods, as described in US Patent Application No. 762,248. In the presence of this xenon, the control rods are withdrawn and the reactor power is increased from point _A4 to point _A2 . When the reactor power reaches point _A , control rod withdrawal is stopped. The reactor power is then
It gradually increases as xenon disappears. Increase the reactor output to _A5 point by increasing the cooling water flow rate. Thereafter, the operations of reducing the cooling water flow rate, withdrawing the control rod, stopping the withdrawal, and increasing the cooling water flow rate are repeated. As a result, the reactor power is A ₅ −A ₆ −A ₂
−A ₇ −A ₈ −A ₂ −A ₉ , and the reactor output can be increased to 100% of the rated output (A ₉ point). By the above operations, the reactor power is increased from 0% to 100%. In a BWR plant, the state of steam extraction from the turbine differs between the reactor's rated output (e.g. 100% reactor output) and partial reactor output (e.g. 67.4% reactor output). As shown, the flow rate of each part of the plant,
The states of enthalpy, pressure, etc. are different. Table 1 shows the values of each part of the plant at rated output and partial output for a BWR with a power density of 50kw/l.
In other words, during partial power such as startup (low flow rate and low power), due to heat balance, cooling water is sub-cooled (hereinafter referred to as
(Simply referred to as subcooling) becomes larger than at rated power, so the power peaking in the lower part of the core increases. The results are shown in Table 1.

【table】

[Table] In a BWR plant, the subcooling at the core inlet is 11.4 cal/g at rated power, and at partial power when the reactor power is 67.4% and the core flow rate is 40% (point B in Figure 1). ) subcooling at the core inlet is 23.4 cal/g. Figure 2 shows changes in subcooling at the core inlet in relation to the flow rate of cooling water in the core. The horizontal and vertical axes of the figure are the flow rate of cooling water in the core (%) and the subcooling of cooling water at the inlet of the core (cal/g), respectively.
The section between CD and D is a section where the reactor power increases due to control rod withdrawal, and the section between DE and E is a section where the reactor output increases due to an increase in the flow rate of cooling water in the reactor core. As the cooling water flow rate increases and the reactor power increases, the subcooling at the core inlet decreases. 3 and 4 respectively show the power distribution and void ratio at rated output and partial output when the control rod patterns are the same. The horizontal axes in FIGS. 3 and 4 represent the axial position of the core (0 and 24 correspond to the lower and upper ends of the core, respectively). The vertical axis indicates the relative output in FIG. 3 and the void ratio in FIG. 4. The characteristics and in Figure 3 indicate the relative power at the rated power and partial power of the reactor, respectively, and the characteristics and in Figure 4 indicate the void fraction at the rated power and partial power of the reactor, respectively. It shows. From Figure 3, the output peaking at rated output is 1.63, while the output peaking at partial output is 1.95. Therefore, the output peaking at partial output is approximately 20% larger than that at rated output. When the reactor is at partial power, the subcooling at the core inlet is large as described above, so as shown in FIG. 4, the void fraction in the upper part of the core increases compared to when the reactor is at rated power. However, the void fraction in the lower part of the core near the boiling start point hardly changes during partial power and rated power of the reactor. In a BWR plant, an increase in void fraction has the property of causing a decrease in reactor output, which increases the power peaking in the upper part of the reactor core. As mentioned above, the threshold at which power can be increased by withdrawing the control rods from the reactor core while maintaining the integrity of the fuel rods is 8 kW/ft in terms of linear thermal power density. In a BWR plant with a power density of 50kw/l, the linear thermal power density corresponding to a power peak of 1.95 at the bottom of the core at partial power (reactor power 67.4%, core flow rate: 40%) is 8.2kw/ft. The threshold for control rod withdrawal is exceeded. The operation to increase the reactor output to the rated output becomes complicated. For this reason, it takes a long time to increase the output of the nuclear reactor, which is one of the reasons why the operation rate of BWR plants decreases. Next, in order to reduce the power peaking in the lower part of the core at partial load during reactor startup,
Various studies were conducted. As a result, by controlling the subcooling of the core inlet, that is, the enthalpy (hereinafter simply referred to as enthalpy) of the cooling water at the core inlet, a flat axial power distribution can be achieved even during partial power of the reactor. I understand. The results of the study are described below. Figure 5 shows power peaking and changes in reactor power due to changes in enthalpy at the core inlet during partial power (core flow rate is 40%) in a BWR plant with a power density of 50kw/l. The horizontal axis shows the enthalpy at the core inlet (cal/g), and the vertical axis shows power peaking and reactor power (%). Characteristics and indicate power peaking and reactor power, respectively. The enthalpy at the core inlet of a conventional BWR plant at partial power is 276 cal/g.
It is. When the enthalpy at the core inlet increases,
Since the subcooling at the core inlet is reduced, the boiling start point moves to the lower end of the core, and the void ratio in the lower core increases significantly. However, the change in void fraction in the upper part of the core is small. Therefore, the power ratio in the lower part of the reactor core decreases, and the power peaking at partial power of the reactor becomes smaller. Also, when the enthalpy at the core inlet increases,
As the average void fraction in the reactor core increases, the reactor power decreases. If the core inlet enthalpy at partial power is 293 cal/g, the same as at rated power, the power peaking will be 1.62 and the reactor power will be 49.1%.
becomes. Compared to a conventional BWR plant, the power peaking at partial power is reduced to 83% of the conventional example, the reactor output is reduced to 73% of the conventional example, and the power peaking is about the same as at rated output. The linear heat power density in this case is 5.8kw/ft, which is about the same as that of a conventional BWR plant.
This decreases to 71%. Figure 6 shows a comparison of the power distribution at rated power and partial power (core cooling water flow rate 40%) when the control rod pattern and core inlet enthalpy are the same (293 cal/g). There is. The horizontal axis of the figure represents the axial position from the lower end of the core (0 and 24 correspond to the lower and upper ends of the core, respectively), and the vertical axis represents the relative power. Characteristics and indicate relative outputs at rated output and partial output, respectively. This figure shows that the two output distributions match well, as described above. FIG. 7 also shows a comparison of void distributions. The horizontal axis is the same as in FIG. 6, and the vertical axis is the void ratio. "Characteristics" and "Characteristics" respectively indicate void rates at rated output and partial output under the same conditions as in FIG. In other words, in the lower part of the reactor core near the boiling start point when the reactor is at rated power, the void fraction increases by about 15% when the reactor is at partial power. On the other hand, in the upper part of the core, the void ratio at partial power is about 5
%To increase. During partial output, such void distribution flattens the output distribution. Therefore, by controlling the enthalpy at the core inlet, it is possible to flatten the axial power distribution in the core even during partial power such as during startup. In addition, since the reactor power is also reduced, the linear thermal power density is significantly reduced, and the margin for the control rod withdrawal threshold is increased. Therefore, the procedure for starting up the nuclear reactor becomes simple, and the time required to achieve the rated output can be shortened. The enthalpy at the core inlet can be calculated from the heat balance of the BWR plant using the following formula. H _IN = Hf (1-W _STM /Wc) + H _fw (W _STM /Wc
)...(1) Here, H _IN : Enthalpy at the core inlet (cal/g) Hf: Enthalpy of saturated cooling water (cal/g) W _STM : Steam flow rate sent to the turbine (t/g)
h) Wc: Cooling water flow rate in the reactor core (t/h) H _fw : Enthalpy of feed water (cooling water supplied to the reactor pressure vessel) (cal/g) Considering the heat balance of the BWR plant in detail, The steam carrier ratio, the amount of heat added by the control rod drive, the amount of heat lost in the purification system, the amount of heat added by the recirculation pump, and the amount of heat escaping from the reactor pressure vessel wall will be included in equation (1). However, these values are small and can be ignored. The enthalpy of saturated cooling water (hereinafter simply referred to as saturated water) is a value determined by the dome pressure of the reactor pressure vessel, and the steam flow rate is determined by the reactor output. Additionally, the flow rate of cooling water in the reactor core is manually changed to vary the reactor output. Therefore, in order to control the enthalpy at the core inlet, it is sufficient to control the enthalpy of the feed water. The enthalpy of water supply is given by the following equation by converting equation (1). H _fw = Hf (1-Wc/W _STM ) + H _IN (Wc/W _STM
)…(2) The measured values of the BWR plant can be used for the cooling water flow rate and steam flow rate in the core. The enthalpy of saturated water can be determined using measurements of the reactor pressure vessel dome pressure and steam tables. Therefore, by setting the enthalpy at the core inlet using equation (2), it is possible to determine how much the enthalpy of the feed water should be controlled in a certain reactor operating state. Cooling water is supplied into the reactor pressure vessel as feed water in an amount equivalent to the amount of water that leaves the reactor pressure vessel in the form of steam. A condenser is used to condense steam used in a turbine to reduce turbine exhaust pressure and increase the thermal efficiency of the turbine. The temperature of the condensed water at the outlet of the condenser is close to the temperature of the seawater cooling the steam. Therefore, it is not possible to control the enthalpy of feed water (condensed water) supplied from the condenser to the reactor pressure vessel by controlling the flow rate on the feed water side of the feed water heater. The enthalpy of feedwater can be easily changed by controlling the flow rate of steam extracted from the turbine and supplied to the feedwater heater, and controlling and adjusting the amount of heat exchange between the steam side and the feedwater side in the feedwater heater. By the way, the enthalpy of water supply cannot be directly measured. However, since the feed water is high-pressure subcooled water and has a specific heat value close to 1.0, it is sufficient to control the feed water temperature, which can be directly measured, instead of the enthalpy of the feed water. The present invention has been made based on the above study results. A specific embodiment of the present invention will be described based on FIG. 8. Figure 8 shows a schematic diagram of the BWR plant. A reactor core 1 exists within a reactor pressure vessel 2 . A plurality of jet pumps 3 are arranged around the reactor core 1 within the reactor pressure vessel 2 . One end of a recirculation pipe 4, which has one end connected to the reactor pressure vessel 2, passes through the reactor pressure vessel 1 and faces an inlet of the jet pump 3. A recirculation pump 5 is attached to the recirculation piping 4. Main steam pipe 6 connects reactor pressure vessel 2 and turbine 7 . Condenser 8 is connected to the steam exhaust of turbine 7 . A feed water pipe 12 connected to the condenser 8 is connected to a feed water spargeer 13 disposed within the reactor pressure vessel 2 via a demineralizer 9, a feed water heater 10, and a feed water pump 11. The extraction pipe 14 connected to the turbine 7 has a flow rate control valve 15.
It is connected to the feed water heater 10 via. A drain pipe 16 connects the feed water heater 10 and the condenser 8. 17 is a feed water temperature control device, the structure of which is shown in detail in FIG. Feed water temperature control device 1
7 is a function generator 18, a calculator 19, an adder 20
and a function generator 21. When the recirculation pump 5 is driven, the cooling water in the reactor pressure vessel 2 flows through the recirculation pipe 4 and is ejected into the inlet of the jet pump 3. The cooling water passes through the jet pump 3, reaches the lower part of the reactor core 1, and flows into the reactor core 1. While rising within the core 1, the cooling water turns into steam by cooling the fuel rods in the fuel assemblies arranged in the core 1. The steam passes through a steam separator and a dryer (not shown), and is led to a turbine 7 via a main steam pipe 6. Steam exhausted from the turbine 7 is condensed in a condenser 8. The condensed liquid cooling water is sequentially guided to the demineralizer 9, the feedwater heater 10, and the feedwater pump 11 through the water supply pipe 12, and is supplied from the feedwater spargeer 13 into the reactor pressure vessel 2. Steam extracted from the turbine 7 is guided into the feedwater heater 10 by an extraction pipe 14. Cooling water (feed water) flowing through the water supply piping 12 in the feed water heater 10 is heated by the extracted steam. After heating the feed water, the steam becomes drain water and is supplied to the condenser 8 through the drain pipe 16. The operation control of the BWR in this embodiment using the feed water temperature control device 17 will be explained below. When the nuclear reactor starts operating, the recirculation pump 5 is driven and cooling water is supplied to the reactor core 1 as described above. The flow rate of cooling water flowing through the reactor core 1 is adjusted to 20%. Control rod 24 inserted into reactor core 1
is withdrawn and the reactor's output is increased. Let's take as an example the situation when the reactor output is 49.1% and the cooling water flow rate in the core is 40%. The dome pressure P _R within the reactor pressure vessel 2 is measured by a pressure gauge. The measured dome pressure P _R is determined by the function generator 1
8, where it is converted to the enthalpy of saturated water, Hf. A steam table may be used to convert the enthalpy of saturated water Hf from the dome pressure _PR . However, if the dome pressure P _R is in the range of 60 to 75 Kg/cm ² , the dome pressure P _R and the enthalpy Hf of saturated water have the following relationship. Hf=294.8+1.244(P _R −65) …(3) In this example, the dome pressure P _R is calculated based on equation (3).
is converted into the enthalpy of saturated water Hf by a function generator 18. The enthalpy Hf of saturated water is calculated using calculator 1.
9 is input. Cooling water flow rate flowing through reactor core 1
Wc is determined from the cooling water flow rate _W1 measured by the flow meter 26 attached to the jet pump 3. If n jet pumps 3 are arranged, the cooling water flow rate Wc is determined by _nW1 . The cooling water flow rate W ₁ is input to the multiplier 27, and is converted into the cooling water flow rate Wc (=nW ₁ ) by the multiplier 27. Cooling water flow rate
Wc is input to the arithmetic unit 19. The steam flow rate W _STM sent to the turbine 7 measured by the flow meter 28 attached to the main steam pipe 6 is determined by the computing unit 19
is input. In addition, the computing unit 19 has the core inlet enthalpy H _IN predetermined by the setting unit 22.
is input. The predetermined value of core inlet enthalpy H _IN is constant. Arithmetic unit 19
calculates the enthalpy of feed water H _fw based on equation (2) using the input values of saturated water enthalpy Hf, cooling water flow rate Wc, steam flow rate W _STM and core inlet enthalpy H _IN . The details of this calculation will be explained with reference to FIG. Wc and W _STM are input to the divider 29, and (Wc/W
_STM ) is output. (Wc/W _STM ) is the multiplier 30
is input. The multiplier 30 is inputted by Hf.
Outputs Hf (Wc/W _STM ). Hf and Hf (Wc/W
_STM ) is input to the adder 31. The adder 31 is
Outputs Hf (1-Wc/W _STM ). Adder 31
The output of is input to the adder 32. Multiplier 33
inputs H _IN and the output (Wc/W _STM ) of the divider 29, and outputs H _IN (Wc/W _STM ) as an output. This output H _IN (Wc/ _WSTM ) is input to the adder 32. As the output of the adder 32, H _fw [=Hf(1-
Wc/W _STM )+H _IN (Wc/W _STM )] is obtained.
In this example, the enthalpy H _IN at the core inlet set by the setting device 22 is 293 cal/g. Therefore, the output _Hfw of the adder 32 is [Hf(1-
Wc/W _STM )+293(Wc/W _STM )]. H _fw
is input to an adder 20 shown in FIG. The adder 20 also receives the feed water temperature Tfw measured by the thermometer 34 . As mentioned above, the feed water is high-pressure subcooled water and the proportionality value is close to 1.0, so there is a slight difference from the actual enthalpy, but it can be considered that the enthalpy value of the feed water is the same as the value of the feed water temperature Tfw. good. That is, if the feed water temperature is 200°C, the enthalpy of the feed water is approximately 200 cal/g. An accurate value for the feed water enthalpy can be obtained by determining the feed water enthalpy corresponding to the feed water temperature using a steam table. However, the thermometer 34 can be said to be a type of enthalpy detection means for the supplied water. Adder 2
ΔHfw (=Hfw−Tfw) is output from 0. Δ
Hfw is input to the function generator 21. The function generator 21 outputs an output V corresponding to ΔHfw. That is, the function generator 21 outputs a signal to open the flow control valve 15 if ΔHfw is a positive value, and a signal to close the flow control valve 15 if ΔHfw is a negative value. The output V of the function generator 21 is transmitted to the amplifier 23
The voltage is amplified and raised to the voltage required to start the motor 35 that opens and closes the flow control valve 15. The motor 35 is driven by the output of the amplifier 23, and the flow control valve 15 is opened and closed. For example, Δ
When Hfw is a positive value, the motor 35
The opening degree of the flow control valve 15 increases. The amount of steam supplied to the feedwater heater 10 through the bleed pipe 14 increases. The temperature of the feed water heated by the feed water heater 10 rises, and the enthalpy of the feed water increases. If ΔHfw is a negative value,
Conversely, the opening degree of the flow control valve 15 becomes smaller, and the amount of steam supplied to the feed water heater 10 decreases. The temperature of the feed water decreases, and the enthalpy of the feed water also decreases.
Varying the flow rate of steam extracted is to adjust the amount of heat added to the feed water by the steam. When the amount of steam extracted decreases, the amount of heat added to the feedwater by the steam supplied to the feedwater heater 10 decreases. By adjusting the enthalpy of the feed water in this way, the enthalpy H _IN at the inlet of the reactor core is approximately
It becomes 293cal/g. The void distribution in the reactor core 1 at a reactor power of 49.1% has the characteristics shown in FIG. 7, and the power distribution in the reactor core 1 has the characteristics shown in FIG. According to this embodiment, even during partial loads such as startup, the output distribution can be flattened, and the linear heat output density can be significantly reduced to 5.8 kw/ft. Therefore, the rate of increase in output due to control rod withdrawal can be increased, so in Fig. 1, output increases due to increase in cooling water flow rate, output decreases due to decrease in cooling water flow rate, and output increase up to point _A due to control rod withdrawal. The number of process repetitions can be reduced. Therefore, the time required to start up the nuclear reactor can be shortened.
The time required to start up the reactor can be shortened while maintaining the integrity of the fuel rods, and the procedure for withdrawing the control rods can be simplified. FIG. 11 shows the temporal changes in the reactor output at startup and the flow rate of cooling water in the core according to this embodiment (enthalpy at the core inlet: 293 cal/g) in comparison with the conventional method. The horizontal axis shows the elapsed time (days) after startup, and the vertical axis shows the reactor output (%) and the cooling water flow rate (%) in the reactor core. Characteristics K and L
1 and 2 respectively indicate the reactor output and the cooling water flow rate of the reactor core in this example. The characteristics M and N are
The figures show the reactor output and the cooling water flow rate in the core of the conventional method, respectively. In this embodiment, since the linear heat power density is reduced during partial power as described above, the time required to withdraw the control rod is negligibly small compared to the conventional method. In addition, in this example, since the power distribution is constant in each operating state, it is possible to increase the slope of the power increase due to the increase in the flow rate of cooling water in the core, and the time required to achieve the rated power is 2.25 days. It is. On the other hand, with the conventional method, it takes 1.25 days to increase the output by withdrawing the control rods, and 3.92 days to increase the output by increasing the flow rate of cooling water in the core, resulting in a total of 5.17 days required to achieve the rated output. This example allows the reactor power to be increased to the rated power in half the time of the conventional method.
It can greatly contribute to improving the utilization rate of BWR plants. Moreover, if this operation control method is used, the output distribution can be flattened under various operating conditions, so that the axial distribution of xenon concentration can also be flattened. Therefore,
Even in the case of output fluctuations such as during load-following control operation, the integrity of the fuel rods can be maintained, contributing to safe operation. Furthermore, when the flow rate of cooling water is low, the average void fraction in the reactor core increases and the reactor power decreases, so it is possible to have a wider range of power fluctuations than in the conventional method, which provides flexibility for power fluctuation operation. has the advantage of increasing sex. The function of the feed water temperature control device 17 can also be performed by a computer. Hfw and Tfw in adder 20
The opening degree of the flow control valve 15 may be adjusted according to the value of Hfw without determining the deviation of Hfw. As described above, the boiling water reactor operation control method and device of the present invention can reduce the time required to increase the reactor output to a predetermined maximum output in a BWR, and increase the utilization rate of the BWR plant. This has great industrial effects.

[Brief explanation of the drawing]

Figure 1 is a characteristic diagram showing the process of increase in reactor power during BWR startup in terms of the relationship between the cooling water flow rate in the core and the reactor output. A characteristic diagram showing the relationship between cooling water and subcooling. Figure 3 is a characteristic diagram showing the relative power distribution in the axial direction of the core in the conventional BWR operation control method. Figure 4 is a characteristic diagram showing the relative power distribution in the axial direction of the core in the conventional BWR operation control method. Figure 5 is a characteristic diagram showing the distribution of void fraction in the axial direction of the BWR, and Figure 6 is a characteristic diagram showing the relationship between enthalpy of cooling water at the core inlet, power peaking, and reactor output during BWR partial power. 7 is a characteristic diagram showing the relative power distribution in the axial direction of the core in the BWR operation control method of the present invention, FIG. 7 is a characteristic diagram showing the void fraction distribution in the axial direction of the core in the BWR operation control method of the present invention, Figure 8 is a system diagram of a BWR plant to which an embodiment of the present invention is applied;
Fig. 9 is a detailed system diagram of the feed water temperature control device shown in Fig. 8, Fig. 10 is a hardware configuration diagram of the computing unit shown in Fig. 9,
FIG. 11 is a characteristic diagram showing the relationship between the elapsed time after reactor startup, the flow rate of cooling water in the reactor core, and the reactor output. 1... Reactor core, 2... Reactor pressure vessel, 3...
Jet pump, 4... Recirculation piping, 5... Recirculation pump, 6... Main steam pipe, 7... Turbine, 8
... Condenser, 9 ... Desalination device, 10 ... Feed water heater, 11 ... Water supply pump, 12 ... Water supply pipe, 13
...Water supply spargeer, 14...Bleed air piping, 15...
...Flow rate control valve, 16...Drain piping, 17...Feed water temperature control device, 18...Function generator, 19...
Arithmetic unit, 20...adder, 21...function generator,
22...Setter, 23...Amplifier, 24...Control rod, 25...Pressure gauge, 26...Flowmeter, 27...
Multiplier, 28... Container total, 29... Divider, 30
... Multiplier, 31 ... Adder, 32 ... Adder,
33... Multiplier, 34... Thermometer, 35... Motor.

Claims (1)

[Scope of Claims] 1. The steam pressure in the reactor vessel, the flow rate of cooling water flowing through the reactor core in the reactor vessel, and the flow rate of steam supplied from the reactor vessel to the turbine are measured, and the heated feed water is In a method for controlling the operation of a boiling water reactor supplied to a reactor vessel, the temperature of the feed water is measured, the enthalpy of the cooling water at the inlet of the reactor core is set in advance, and the measured and set value is set. A method for controlling operation of a boiling water reactor, characterized in that a desirable feed water temperature is calculated from the above, and the feed water temperature is controlled to the desired value. 2. A first measuring means for measuring the steam pressure in the reactor vessel, a second measuring means for measuring the flow rate of cooling water flowing through the reactor core in the reactor vessel, and a second measuring means for measuring the flow rate of the cooling water flowing through the reactor core in the reactor vessel; a third measuring means for measuring the flow rate, a setting means for setting the enthalpy of the cooling water at the inlet of the reactor core, and a means for adjusting the enthalpy of the feed water supplied into the reactor vessel; 2nd and 3rd
Based on the pressure measured by the measuring means, the cooling water flow rate and the steam flow rate, and the enthalpy of the cooling water set by the setting means, H _fw = Hf (1-Wc/W _STM ) + H _IN (Wc/W _STM
) Here, H _fw : Enthalpy of feed water (cal/g) Hf : Enthalpy of saturated cooling water (cal/g) Wc : Flow rate of cooling water in the reactor core (t/h) W _STM : Flow rate of steam sent to the turbine ( t/
h) A method for controlling the operation of a boiling water reactor, characterized in that the feed water enthalpy adjusting means is controlled so as to satisfy the relationship: H _IN :enthalpy at the core inlet (cal/g). 3. A nuclear reactor core arranged in a reactor vessel, a water supply system and a turbine connected to the reactor vessel, the water supply system having a heater for feed water fed back to the reactor vessel, a first measuring device for measuring the dome pressure _PR of the reactor vessel;
An operation control device for a boiling water reactor, comprising a second measuring device for measuring a cooling water flow rate Wc in the reactor core and a third measuring device measuring a steam flow rate W _STM sent to the turbine, A fourth measuring device for measuring the feed water temperature T at the inlet of the reactor core, steam pressure W _R , cooling water flow rate Wc, steam flow rate W _STM , and similarly the enthalpy H _IN of the cooling water at the inlet of the reactor core,
These values are given to a computing unit that calculates the desired feed water temperature Ts in thermal equilibrium of the reactor vessel, the measured temperature T and the desired temperature Ts are given to a comparator, and the output of the comparator is used to calculate the 1. An operation control device for a boiling water reactor, characterized in that it is provided with a controller for adjusting a heater.