Nuclear power plant and operational method of nuclear power plant

Abstract

To adjust a core flow rate higher than that in a stationary state when a power change operation of a natural circulation reactor is conducted by a control rod, thus improving a power change speed while maintaining a stable condition.SOLUTION: A nuclear power plant includes a large-diameter space chimney at an upper part of a reactor core within a shroud of a reactor pressure vessel. The nuclear power plant includes: a coolant flowmeter configured to measure a coolant flow flowing into the reactor core; a water level gauge configured to measure a water level of a coolant within the reactor pressure vessel; a water-supply flowmeter connected to the reactor pressure vessel and configured to measure a water-supply flow rate; a monitoring device configured to monitor the coolant flow, the water level and the water-supply flow rate; and a controlling device configured to control a control rod and the water-supply flow rate. The controlling device is configured to: calculate, in advance, a fluctuation margin of a core flow rate by a time lag of the water-supply flow rate; change a target value of the water level to be higher than a fluctuation margin of the water level corresponding to the fluctuation margin of the core flow rate so as to change the water-supply flow rate and then change the water level; and thereafter operate and control a position change of the control rod.SELECTED DRAWING: Figure 4

Description

[Technical Field]

[0001] The present invention relates to a nuclear power plant and a method for operating a nuclear power plant. [Background technology]

[0002] In a boiling water reactor, nuclear fuel rods are arranged in a square fuel assembly, where nuclear reactions boil the coolant (liquid) flowing between the fuel rods, and the resulting steam is used to generate electricity. The heat of the fuel rod reaction is transferred directly to the coolant, maintaining efficient steam generation and cooling of the fuel rods. At the bottom of the fuel assembly, the coolant passes through the lower tie plate and flows into the assembly, where it boils, forming a two-phase flow in the upper part of the assembly, where both the gas and liquid phases coexist.

[0003] The gas-liquid two-phase flow is separated into moisture through a separator and dryer installed in the upper part of the reactor, and the separated steam is sent to the turbine building through the main steam pipe as main steam. The hot water remains in the space outside the separator or outside the shroud (hereafter referred to as the downcomer) and flows downward through the downcomer at a specified flow rate by a recirculation pump, before flowing back into the reactor core.

[0004] Among boiling water reactors, those based on natural circulation (hereafter referred to as natural circulation reactors) eliminate the need for a cooling water recirculation pump by simplifying the reactor internal structures, and maintain the cooling of the reactor core by the natural circulation of the cooling water. Figure 13 shows a schematic diagram of a natural circulation reactor.

[0005] In Patent Document 1, a space called a chimney 4 is provided above the core in the shroud 2 to promote natural circulation. Steam 8 generated in the core 3 flows into the chimney 4. In the downcomer 7 through the cylindrical shroud 2, no steam exists and only hot water flows. As a result, the average density of the coolant (average density of steam and water) in the chimney 4 is lower than the density of single-phase water outside the shroud. A density difference occurs between the chimney 4 inside the shroud 2 and the downcomer 7 outside, which creates a hydrostatic head difference inside and outside the shroud, resulting in a pressure loss due to the natural circulation flow rate equivalent to the hydrostatic head difference.

[0006] From the above, in a natural circulation reactor, the recirculation flow rate of cooling water and the core flow rate are determined by the static pressure difference between the inside and outside of the shroud caused by the average volume fraction of steam in the shroud (hereinafter referred to as the void fraction), and the core flow rate is not directly controlled by a recirculation pump or the like.

[0007] During normal operation of a natural circulation reactor, the target value for the water level inside the reactor pressure vessel formed in the downcomer (hereafter referred to as downcomer water level 9) is set to a constant value, and the core flow rate appears as a response to the void fraction in the chimney according to the power. The downcomer water level 9 is an important monitoring parameter for detecting abnormalities in the plant's operating state, and determines the static head around the shroud. In a steady state, the downcomer water level 9 is maintained at a constant target value by using the feedwater flow rate, which is equivalent to the main steam flow rate.

[0008] When performing power control operations (hereafter referred to as load-following operations) that change the reactor's power output in response to startup operations or demand commands in a natural circulation reactor, the core flow rate cannot be directly adjusted, so the thermal power of the core is adjusted primarily by manipulating the control rods, and the core flow rate is determined as a response to changes in steam generation rate in response to changes in core power and changes in the average void fraction in the chimney. Figure 14 shows the relationship between core flow rate and core power in a natural circulation reactor. The horizontal axis is the core flow rate normalized by the core flow rate at 100% rated operation, and the vertical axis is the core flow rate normalized by the core power at rated operation. The dashed line is the steady-state line, which represents the steady state. Normally, the relationship between core flow rate and power output is one-to-one, and power is controlled within a planned range based on the steady-state line predicted in advance. [Prior art documents] [Patent documents]

[0009] [Patent Document 1] Japanese Patent Application Laid-Open No. 2007-225530 Summary of the Invention [Problem to be solved by the invention]

[0010] However, if the power change rate is high during load-following operation of a natural circulation reactor, the void fraction in the chimney responds to the time change in power, and not only the core flow rate but also the downcomer water level 9 may respond in an attempt to maintain the static head difference inside and outside the shroud. Furthermore, since the feedwater flow rate responds with a time delay to control the downcomer water level 9, the downcomer water level 9 may fluctuate. Since the power change range per unit time increases as the power change rate increases, the downcomer water level 9 changes by a time delay due to the feedwater flow rate, which may affect stable plant operation. In addition, if load-following operation is stopped due to control rod operation shutdown, the power change rate may decrease.

[0011] An object of the present invention is to provide a nuclear power plant and an operating method for a nuclear power plant that can adjust the core flow rate higher than that in a steady state when a power change operation of a natural circulation reactor is performed using control rods, thereby improving the power change rate while maintaining a stable state. [Means for solving the problem]

[0012] The nuclear power plant of the present invention is a nuclear power plant having a large-diameter spatial chimney above the core within the shroud of a reactor pressure vessel, and is equipped with a coolant flow meter that measures the coolant flow rate flowing into the reactor core, a water level gauge that measures the water level of the coolant in the reactor pressure vessel, a feedwater flow meter connected to the reactor pressure vessel and that measures the feedwater flow rate, a monitoring device that monitors the coolant flow rate, the water level, and the feedwater flow rate, and a control device that controls the control rod and the feedwater flow rate, wherein the control device calculates in advance the fluctuation range of the core flow rate due to a time delay of the feedwater flow rate, and changes the feedwater flow rate and the water level by changing the target value of the water level to be greater than or equal to the fluctuation range of the water level that corresponds to the fluctuation range of the core flow rate, and then controls by changing the position of the control rod.

[0013] Alternatively, the nuclear power plant of the present invention is a nuclear power plant having a large-diameter spatial chimney above the core within the shroud of a reactor pressure vessel, and comprises a coolant flow meter that measures the flow rate of coolant flowing into the reactor core, a water level gauge that measures the water level of the coolant in the reactor pressure vessel, a feedwater flow meter connected to the reactor pressure vessel and that measures the feedwater flow rate, a monitoring device that monitors the coolant flow rate, the water level, and the feedwater flow rate, and a control device that controls the control rods and the feedwater flow rate, wherein the control device calculates in advance a fluctuation range of the core flow rate due to a time delay of the feedwater flow rate and a fluctuation time of the core flow rate, and changes the target value of the reactor water level to at least the water level fluctuation range corresponding to the fluctuation range of the core flow rate from the control rod operation start time to the fluctuation time of the core flow rate, thereby changing the feedwater flow rate and changing the water level, and by the control rod operation end time, changes the target value of the water level to the original water level.

[0014] Alternatively, the method for operating a nuclear power plant of the present invention is a method for operating a nuclear power plant having a large-diameter spatial chimney above the core in a shroud of a reactor pressure vessel, and includes a flow rate measuring step of measuring a coolant flow rate flowing into the reactor core, a water level measuring step of measuring a water level of the coolant in the reactor pressure vessel, a feedwater flow rate measuring step connected to the reactor pressure vessel and measuring a feedwater flow rate, a monitoring step of monitoring the coolant flow rate, the water level and the feedwater flow rate, and a control step of controlling the control rods and the feedwater flow rate, wherein the control step calculates in advance a fluctuation range of the core flow rate due to a time delay of the feedwater flow rate, and The control step is characterized in that the control step changes the target value of the water level by changing the feedwater flow rate to a value greater than the fluctuation range of the water level corresponding to the fluctuation range of the amount of feedwater, thereby changing the water level, and then controls by changing the position of the control rod, or the control step is characterized in that the control step calculates in advance the fluctuation range of the core flow rate due to the time delay of the feedwater flow rate and the time of the fluctuation of the core flow rate, changes the target value of the reactor water level by changing the feedwater flow rate to a value greater than the water level fluctuation range corresponding to the fluctuation range of the core flow rate from the control rod operation start time to the fluctuation time of the core flow rate, thereby changing the water level, and then changes the target value of the water level to the original water level by the control rod operation end time. [Effects of the Invention]

[0015] According to the present invention, when a power change operation of a natural circulation reactor is performed using a control rod, the core flow rate can be adjusted to be higher than that in a steady state, and it is possible to provide a nuclear power plant and an operating method for a nuclear power plant that are capable of improving the power change rate while maintaining a stable state. [Brief explanation of the drawings]

[0016] [Figure 1] 1 shows the configuration of a reactor pressure vessel in a first embodiment. [Figure 2] 1 shows the relationship between power and core flow rate in a comparative example. [Figure 3] 1 shows the relationship between power and core flow rate in the first embodiment. [Figure 4] 1 shows a flowchart in the first embodiment. [Figure 5] 1 shows the change over time when the output is increased in the comparative example and the first embodiment. [Figure 6] 10 shows the configuration of the fuel assembly inlet in the second embodiment. [Figure 7] 10 shows the relationship between power and core flow rate in the second embodiment. [Figure 8] 10 shows a flowchart in a second embodiment. [Figure 9] 1 shows a schematic diagram of a natural circulation furnace in Example 3. [Figure 10] 10 shows a flowchart in a third embodiment. [Figure 11] 10 shows a flowchart in a fourth embodiment. [Figure 12] 10 shows an image of the change over time when the output is increased in Example 4. [Figure 13] A schematic diagram of a natural circulation furnace is shown. [Figure 14] The relationship between core flow rate and core power in a natural circulation reactor is shown below. DETAILED DESCRIPTION OF THE INVENTION

[0017] Hereinafter, an embodiment of the present invention will be described with reference to the drawings. [Example]

[0018] Figure 1 shows the configuration of a reactor pressure vessel in embodiment 1. In this embodiment, the required feedwater flow rate into the reactor associated with power control operations is predicted based on a monitoring device 16 and a control system, and the target control value for the water level in the reactor is adjusted in advance, thereby aiming to keep the water level, which fluctuates during control rod operation, within an acceptable range. In a natural circulation reactor, a reactor pressure vessel 1 has a core 3 installed in the lower part of a shroud 2, and a chimney 4 installed above the core 3. The upper part of the shroud 2 is connected to a separator 6, and steam separated by the separator 6 flows to the top of the reactor pressure vessel 1 via a dryer 5 above it, and the steam is sent to the turbine building via a main steam system line 10.

[0019] In the feedwater system line 13, hot water is sent by the feedwater pump 11 to the reactor pressure vessel 1 via the turbine, condenser, etc. In such a reactor, a monitoring device 16 is provided to acquire and monitor the measurement data of a coolant flow meter 14 that measures the flow rate of the coolant flowing into the reactor core 3, a water level gauge 18 that measures the downcomer water level 9 in the pressure vessel, and a feedwater flow meter 12 that measures the feedwater flow rate. In addition, a control device 17 is provided to process the data from the monitoring device 16 and send instructions to the feedwater pump 11, feedwater flow control valve 19, and control rods 15 based on the calculation results.

[0020] The core flow rate of a natural circulation reactor is determined so that the difference in static head between the inside and outside of the shroud 2 balances mainly with pressure losses such as friction losses, so the core flow rate also changes even if the downcomer water level 9, which determines the static head outside the shroud 2, changes. Generally, the downcomer water level 9 is operated to be kept constant by adjusting the feedwater flow rate, so the water level is not used to control the core flow rate.

[0021] During steady-state operation, the downcomer water level 9 in the reactor is maintained at a constant target value. The relationship between the feedwater flow rate (mass flow rate) Wfw and the generated steam flow rate Wms is given by the following equation: Wfw=Wms+Δh (1), where Δh is the difference [m] between the target and measured value of downcomer water level 9, and is close to zero in steady state when the target value has been reached. The above equation means that when downcomer water level 9 is constant, the flow rate of coolant that boils and flows out of the reactor to the steam turbine as main steam is supplemented by feedwater. If the downcomer water level 9 changes, Wms and Δh in the above equation will also change, so in a normal control system the feedwater flow rate will be automatically adjusted to match the downcomer water level 9 to the target value.

[0022] When the output changes from P1 to P2 due to output control, the change in the feedwater flow rate ΔWfw is expressed by the following equation: ΔWfw = ΔWms ≒ (P2 - P1) / Hsat (2) Here, Hsat is the latent heat of saturation [J / kg]. The above equation is positive when the output is increased and negative when the output is decreased. In reality, the feedwater flow rate responds with a time delay. The feedwater flow rate delay ΔWdelay is expressed by the following equation. ΔWdelay=(ΔWfw / Δt12)T···(3) Here, W1 and W2 are the steady-state core flow rates corresponding to power outputs P1 and P2, T is the feedwater time delay [s], and Δt12 is the target value [s] for the power adjustment time. T is the time difference between issuing a command to change the feedwater flow rate and the actual change in the feedwater flow rate into the reactor, and is an inherent value that arises from the design specifications of the feedwater pump, feedwater piping length, and feedwater control system characteristics. Furthermore, the above equation means that the feedwater flow rate required to maintain a constant downcomer water level 9 is delayed by ΔWdelay, making it impossible to maintain the downcomer water level 9 constant. It is known that the following relationship generally holds between the reactor water level change width Δh and the core flow rate ΔWcore. ΔWcore∝Δh (4) In addition, when the core flow rate is constant, the variation range ΔWfw of the feedwater flow rate and the variation range of the downcomer water level 9 have the following relationship: Δh=ΔWfw / (Ad·ρl)···(5) where Ad is the cross-sectional area of ​​the flow passage outside the shroud [m 2 ], ρl is the liquid density [kg / m 3 If ΔWfw = ΔWdelay is substituted into the above equation, the water level change width caused by the time delay in the water supply flow rate becomes Δhdelay.

[0023] From the above equation, the change in core flow rate ΔWc,d during power control, taking into account the time delay in the feedwater flow rate, can be predicted using the following equation: ΔWc,d=W2−W1+f(ΔWdelay)···(6) Here, f is a function consisting of equations (4) and (5) that expresses the relationship between the range of change in core flow rate and the range of change in feedwater flow rate. In the above equation, (W2-W1) indicates the range of change in core flow rate due to power change, and based on the power-core flow rate relationship shown in Figure 10, it can be ignored if the change in core flow rate relative to the power change is small.

[0024] By calculating the fluctuation range of the core flow rate based on the information on the change range of the core flow rate due to the output change and the change range of the feed water flow rate as in equation (6), it is possible to grasp the change range of the core flow rate during output control considering the time delay of the feed water flow rate.

[0025] Fig. 2 shows the relationship between the output and the core flow rate in the comparative example. The dotted lines represent the steady states at water levels a and b (a < b), respectively. When increasing the output by the control rod withdrawal operation from state 1 to state 2 at water level b for the output increase operation, if the output change rate is small, the output increases along the dotted line. On the other hand, when the output change rate is large and Δt12 cannot be ignored with respect to T in equation (3), the core flow rate corresponding to the delay ΔWdelay of the feed water flow rate may decrease as shown by the solid line. Furthermore, at this time, the core flow rate may fall below the steady state at water level a.

[0026] In a boiling water reactor, usually, water level limit values are provided to detect abnormal plant shutdown or its tendency. Assuming that water level a is the limit value for abnormal detection, the control rod operation must stop when the dotted line at water level a is reached, and the output increase operation may stop. In the case of the output decrease operation, the above-described magnitude relationship is reversed, and the core flow rate may increase by ΔWdelay.

[0027] Fig. 3 shows the relationship between the output and the core flow rate in Example 1. The steady state at water level c (a < b < c) is added to Fig. 2, and on the dotted line indicating the steady state at water level c, the states corresponding to the same output as states 1 and 2 are defined as states 1' and 2', respectively. In this example, before performing the control rod operation (pulling out operation in the case of the output increase operation), the target value of the downcomer water level 9 is increased from water level b to water level c so as to change from state 1 to state 1'. This increases the feed water flow rate and increases the downcomer water level 9 from water level b to water level c according to the relationship in equation (5). Then, by performing the control rod operation, it is possible to reach state 2' without falling below water level a even if a decrease in the core flow rate occurs, and complete the output increase operation. In the case of the output decrease operation, by previously decreasing the water level from c to b from state 2' to state 2, it is possible to allow the increase in the core flow rate that occurs when the control rod is inserted and operate.

[0028] A flowchart of this embodiment is shown in Figure 4. From the monitored data of power, core flow rate, and downcomer water level 9, the current power P1, core flow rate W1, and downcomer water level h1, which is the current water level, are obtained. In addition, the target power P2 and target power change time Δt12 are obtained, and the core flow rate W2 at the target core power is obtained from the relationship between power and core flow rate in the steady state. Next, the water level change width Δhdelay due to the delay in the feedwater flow rate is calculated.

[0029] It is determined whether the water level fluctuation range is within the allowable range Δhallow, and if it is within the allowable range, control rod operation is carried out as usual. If it is outside the allowable range, the target water level is set to be increased to or above h1 + (Δhdelay - Δhallow) and maintained there. The downcomer water level 9 and core flow rate then increase. After that, power increase operation is carried out by control rod operation, and when the target power P2 is reached, control rod operation is stopped. Δhallow is, for example, the difference between the low water level warning water level and the normal water level. Note that the same operation is possible in the case of reduced power operation, where Δhallow is, for example, the difference between the high water level warning water level and the normal water level. After control rod operation is stopped, the target water level is returned to h1 and maintained there, and the water level and core flow rate are reduced.

[0030] In addition, in Figure 4, when the water level has both a lower limit hlow and an upper limit hhigh, the water level target value htarget is set within the range of the following equation. The output increase operation is expressed by the following equation. h1+Δhdelay―|h1―hlow| <hterget<hhigh···(7) The reduced power operation is expressed by the following equation: hlow <hterget ​FIG. 5 shows the time changes during power increase in the comparative example and Example 1. (a) is the comparative example, and (b) is the present example. In the comparative example, after control rod operation begins, the core and downcomer water levels 9 drop at a certain time and fall below water level a. If water level a is at the alarm water level, control rod operation may be stopped. In this example, before control rod operation begins, the target water level is increased from water level b to water level c, thereby increasing the downcomer water level 9 from water level b to water level c (State 1 → State 1'). After that, although the core flow rate decreases due to a delay in the feedwater flow rate, it is prevented from reaching water level a. After the target power is reached (State 1' → State 2') and control rod operation is terminated, the target value of the downcomer water level 9 is returned to the original water level b, thereby restoring the downcomer water level 9 and core flow rate (State 2' → State 2). [Example]

[0031] 6 is a schematic diagram of the lower inlet portion of the fuel assembly used in this embodiment. In the first embodiment, the core flow rate is limited by the allowable range of the downcomer water level 9, as in the relationship between the downcomer water level 9 and the core flow rate expressed by equation (4). In contrast, the purpose of this embodiment is to selectively expand the allowable range of the core flow rate even within the same allowable range of the downcomer water level 9 during power increase operation.

[0032] This embodiment is mainly composed of a fuel assembly 40, a lower tie plate 43, a fuel support bracket 44, and a fuel support plate 45. Another feature is the provision of a cylindrical lattice-shaped orifice (hereinafter referred to as a variable resistance orifice 46). In a fuel assembly equipped with a variable resistance orifice 46, as shown in the figure, the flow resistance decreases as the flow transitions from laminar to turbulent at a certain cooling water inlet flow velocity Vth. This has the effect of increasing the core flow rate even at the same downcomer water level 9. In other words, by providing a fuel support bracket that supports the fuel assembly and the lower tie plate and providing an orifice made of a cylindrical member at the inlet of the fuel support bracket, the core flow rate can be increased.

[0033] In equation (4) shown in the first embodiment, the proportionality coefficient between the range of change in the downcomer water level 9 and the range of change in the core flow rate, which are in a proportional relationship, can be used with any core flow rate as the boundary. However, while the effect of increasing the core flow rate is obtained, there is a possibility that channel stability will decrease due to a decrease in the fluid resistance at the core inlet. Therefore, in this embodiment, by determining in advance the core flow rate Wth (hereinafter referred to as the core flow rate threshold) corresponding to the inlet flow velocity Vth, the effect of increasing the core flow rate can be selectively obtained by the variable resistance orifice 46 as needed, and the influence of a decrease in the core flow rate due to a delay in the feedwater flow rate can be avoided.

[0034] 7 shows the relationship between power and core flow rate, demonstrating the effect of this embodiment. During power ramp-up operation, which increases power from P1 to target power P2, the target value of the downcomer water level 9 is increased from water level b to water level c before the control rod withdrawal operation and maintained there, thereby increasing the downcomer water level 9. If the core flow rate exceeds the core flow rate threshold Wth in the range of power P1 to P2 in the steady state at water level c without the variable resistance orifice 46, then with the variable resistance orifice 46, the steady state at water level c is expanded by the amount of the decrease in fluid resistance compared to when the variable resistance orifice 46 is not provided. This allows the allowable range of the core flow rate to be expanded relative to the range of core flow rate reduction caused by a delay in the feedwater flow rate, within the same allowable range of the downcomer water level 9.

[0035] 8 shows a flowchart of this embodiment. It is generally the same as the first embodiment. When the water level fluctuation range is outside the allowable range, it is determined whether it is necessary to increase the core flow rate above the core flow rate threshold value Wth of the variable resistance orifice 46. hth is a value obtained by converting Wth into the downcomer water level 9. In this way, the resistance reduction effect of the variable resistance orifice 46 is utilized only when the range of decrease in the core flow rate exceeds the allowable range of the normal downcomer water level 9.

[0036] As described above, when the fluctuation range of the predicted core flow rate exceeds the core flow rate threshold at which the fluid resistance of the orifice decreases, the target water level is changed to a water level equivalent to the core flow rate threshold at which the fluid resistance of the orifice decreases, thereby expanding the allowable range of the core flow rate. [Example]

[0037] Figure 9 shows a schematic diagram of a natural circulation reactor embodying this embodiment. In this embodiment, a decrease in the core flow rate due to a shortage of feedwater flow is avoided by ensuring the amount of water in the feedwater line in advance during increased power operation. In the primary cooling system of a boiling water reactor under normal operating conditions, the coolant flow rate is maintained at a constant level, and the main steam generated in the reactor is used to generate electricity in the steam turbine, and then flows back into the reactor mainly via the condenser and feedwater heater.

[0038] However, if the power increases in a short period of time due to power control, a temporary feedwater flow rate greater than normal may be required, which may result in a shortage. If the feedwater flow rate is insufficient, it becomes difficult to increase the downcomer water level 9 as shown in Examples 1 and 2, which may result in a further decrease in the core flow rate. In this example, a feedwater tank 30 is provided on the feedwater line, and the water volume is monitored by a water level gauge 31. In addition, the feedwater tank valve is opened and closed according to a command signal from the control device.

[0039] Figure 10 shows a flowchart of this embodiment. It is generally the same as embodiment 1. During power increase operation, after control rod manipulation begins, a shortage of feedwater flow rate can be avoided by opening the feedwater tank valve and increasing the amount of water in the feedwater line. That is, a feedwater tank connected to a feedwater pipe connected to a feedwater pump and a water level gauge for measuring the amount of cooling water in the feedwater tank are provided, and the cooling water in the feedwater tank is released into the feedwater pipe between the start and end of control rod manipulation, thereby avoiding a shortage of feedwater flow rate. [Example]

[0040] In this embodiment, the purpose is to shorten the interval between power increase or decrease operation and to perform continuous power changes by simultaneously adjusting the downcomer water level 9 and operating the control rods while avoiding a decrease in the core flow rate due to a delay in the feedwater flow rate. In Examples 1 to 3, the downcomer water level 9 needs to be operated before and after the control rod operation, so intervals are required before the start and after the end of power control. In order to respond immediately when a power control command is received immediately after the end of a power change, it is necessary to shorten the interval.

[0041] FIG. 11 shows a flowchart of this embodiment. FIG. 12 shows the effect of this embodiment. In this embodiment, the timing at which a decrease in core flow rate occurs (core flow rate decrease time tth) is predicted. Next, the target value of the downcomer water level 9 is increased from the control rod operation start time t1 to time tth, thereby mitigating the decrease in core flow rate. Furthermore, the target water level value is adjusted so that the downcomer water level 9 returns to its original level at the same time as the control rod operation end time t2. The core flow rate decrease time tth is the timing at which the increase in downcomer water level 9 due to the increase in the void fraction in the chimney caused by the power increase balances with the decrease in downcomer water level 9 due to the main steam flow rate, and is expressed by the following relationship. (P2-P1)Hc / Pz=(W2-W1)(tth-t1) / Ad·ρl··(9) tth=t1+(P2−P1)Ad·ρl·Hc / (Pz(W2−W1))··(10) Here, Pz is the output at rated operation [W], and Hc is the chimney height [m].

[0042] The time of change in core flow rate can be determined by calculating the time when the increase in water level due to the increase in void fraction in the chimney caused by an increase in power balances with the decrease in water level due to the main steam flow rate. [Explanation of symbols]

[0043] 1... reactor pressure vessel, 2... shroud, 3... reactor core, 4... chimney, 5... dryer, 6...Separator, 7...Downcomer, 8...Steam, 9...Downcomer water level, 10...Main steam system line, 11...Feedwater pump, 12...Feedwater flow meter, 13...Feedwater system line, 14...coolant flow meter, 15...control rod, 16...monitoring device, 17...control device, 18...water level gauge, 19...water supply flow control valve, 30...water supply tank, 32...water supply tank valve, 40... fuel assembly, 41... channel box, 42... fuel rod, 43...lower tie plate, 44...fuel support bracket, 45...fuel support plate, 46...Variable resistance orifice

Claims

1. In a nuclear power plant having a large-diameter space chimney above the reactor core in the shroud of the reactor pressure vessel, a coolant flow meter that measures the flow rate of coolant flowing into the reactor core; a water level meter for measuring the water level of the coolant in the reactor pressure vessel; a feedwater flow meter connected to the reactor pressure vessel and configured to measure a feedwater flow rate; a monitoring device for monitoring the coolant flow rate, the water level, and the feedwater flow rate; a control rod and a control device for controlling the feedwater flow rate; The control device calculates in advance the fluctuation range of the core flow rate due to a time delay in the feedwater flow rate, and changes the target value of the water level to a value equal to or greater than the fluctuation range of the water level corresponding to the fluctuation range of the core flow rate, thereby changing the feedwater flow rate and changing the water level, and then controls the control rod by operating to change the position of the control rod.

2. In a nuclear power plant having a large-diameter space chimney above the reactor core in the shroud of the reactor pressure vessel, a coolant flow meter that measures the flow rate of coolant flowing into the reactor core; a water level meter for measuring the water level of the coolant in the reactor pressure vessel; a feedwater flow meter connected to the reactor pressure vessel and configured to measure a feedwater flow rate; a monitoring device for monitoring the coolant flow rate, the water level, and the feedwater flow rate; a control rod and a control device for controlling the feedwater flow rate; The control device calculates in advance the range of fluctuation in the core flow rate due to a time delay in the feedwater flow rate and the time of fluctuation of the core flow rate, changes the target value of the reactor water level by changing the feedwater flow rate to a water level fluctuation range equivalent to the range of fluctuation in the core flow rate or more from the control rod operation start time to the fluctuation time of the core flow rate, and changes the target value of the water level to the original water level by the control rod operation end time.

3. 2. The nuclear power plant according to claim 1, The nuclear power plant is characterized in that the control device controls the water level to be changed to the water level before the control rod operation after operating the control rod position change.

4. 3. The nuclear power plant according to claim 2, A nuclear power plant characterized in that the time of change in the core flow rate is calculated as the timing when the increase in the water level due to an increase in the void fraction in the chimney caused by an increase in power balances with the decrease in the water level due to the main steam flow rate.

5. The nuclear power plant according to claim 1 or 2, A nuclear power plant, characterized in that the fluctuation range of the core flow rate is calculated based on information on the fluctuation range of the core flow rate due to a change in power and the fluctuation range of the feedwater flow rate.

6. The nuclear power plant according to claim 1 or 2, 1. A nuclear power plant comprising: a fuel support fitting for supporting a fuel assembly and a lower tie plate; and an orifice formed of a cylindrical member at an inlet of said fuel support fitting.

7. 7. The nuclear power plant according to claim 6, A nuclear power plant characterized in that, when the predicted fluctuation range of the core flow rate exceeds the core flow rate threshold value at which the fluid resistance of the orifice decreases, the target value of the water level is changed to the water level corresponding to the core flow rate threshold value at which the fluid resistance of the orifice decreases.

8. The nuclear power plant according to claim 1 or 2, a water supply tank connected to a water supply pipe connected to the water supply pump; a water level meter for measuring the amount of cooling water in the water supply tank; 1. A nuclear power plant comprising: a reactor for discharging cooling water from said water supply tank into said water supply pipe during a period from the start of control rod operation to the end of control rod operation.

9. A method for operating a nuclear power plant having a large-diameter spatial chimney above a reactor core in a shroud of a reactor pressure vessel, comprising: a flow rate measuring step of measuring a flow rate of coolant flowing into the reactor core; a water level measuring step of measuring a water level of the coolant in the reactor pressure vessel; a feedwater flow rate measuring step connected to the reactor pressure vessel and measuring a feedwater flow rate; a monitoring step of monitoring the coolant flow rate, the water level, and the feedwater flow rate; a control step of controlling a control rod and the feedwater flow rate; The control step calculates in advance a fluctuation range of the core flow rate due to a time delay of the feedwater flow rate, and changes the target value of the water level to a fluctuation range of the water level equivalent to the fluctuation range of the core flow rate or more, thereby changing the feedwater flow rate and changing the water level, and then controls by operating a change in the position of the control rod, or The control step includes calculating in advance the fluctuation range of the core flow rate due to a time delay in the feedwater flow rate and the time of the fluctuation of the core flow rate, changing the target value of the reactor water level by changing the feedwater flow rate to a water level fluctuation range equivalent to the fluctuation range of the core flow rate or more from the control rod operation start time to the fluctuation time of the core flow rate, thereby changing the feedwater flow rate and changing the water level, and changing the target value of the water level to the original water level by the control rod operation end time.

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