Nuclear power plant and method for operating nuclear power plant

Abstract

According to the present invention, when an operation for adjusting the output of a natural circulation reactor is performed using control rods, the core flow rate can be adjusted higher than the steady state, and the rate of output change can be improved while maintaining a stable state. Provided is a nuclear power plant having a large-diameter space chimney in the upper part of a core in a shroud of a reactor pressure vessel. The nuclear power plant comprises a coolant flow meter that measures the flow rate of a 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 which is connected to the reactor pressure vessel and which measures the flow rate of feedwater, a monitoring device that monitors the coolant flow rate, the water level, and the feedwater flow rate, and a control device that controls control rods and the feedwater flow rate. The control device calculates in advance a range of fluctuation in a core flow rate due to the time delay of the feedwater flow rate, and manipulates and controls a change in the positions of the control rods after changing the feedwater flow rate and changing the water level by changing a target value for the water level so as to be equal to or greater than the range of fluctuation in the water level, which corresponds to the range of fluctuation in the core flow rate.

Description

Nuclear power plant and method of operating a nuclear power plant

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

[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 (hereinafter 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. Therefore, 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 causes 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 and core flow rate of the cooling water 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 of the water level inside the reactor pressure vessel formed in the downcomer (hereinafter 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 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 (hereinafter referred to as load-following operations) in which the reactor power is changed in response to startup operations or demand commands in a natural circulation reactor, the core flow rate cannot be directly adjusted. Therefore, the thermal power of the core is adjusted mainly by operating the control rods, and the core flow rate is determined as a response to changes in the steam generation rate and the average void fraction in the chimney in response to changes in core power. Figure 14 shows the relationship between core flow rate and core power in a natural circulation reactor. The horizontal axis represents the core flow rate normalized by the core flow rate at 100% rated operation, and the vertical axis represents the core flow rate normalized by the core power at rated operation. The dashed line represents the steady-state line, which represents the steady state. Normally, the relationship between core flow rate and power is one-to-one, and power is controlled within a planned range based on the steady-state line predicted in advance.

[0009] Japanese Patent Application Laid-Open No. 2007-225530

[0010] However, when 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 change range of the downcomer water level 9 due to the time delay in the feedwater flow rate increases, which may affect stable operation of the plant. 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 the 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.

[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.

[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.

[0016] 1 shows the configuration of a reactor pressure vessel in Example 1. 2 shows the relationship between power and core flow rate in a comparative example. 3 shows the relationship between power and core flow rate in Example 1. 4 shows a flowchart in Example 1. 5 shows the time change when power is increased in a comparative example and Example 1. 6 shows the configuration of a fuel assembly inlet in Example 2. 7 shows the relationship between power and core flow rate in Example 2. 8 shows a flowchart in Example 2. 9 shows a schematic diagram of a natural circulation reactor in Example 3. 10 shows a flowchart in Example 3. 11 shows a flowchart in Example 4. 12 shows an image of the time change when power is increased in Example 4. 13 shows a schematic diagram of a natural circulation reactor. 14 shows a relationship between core flow rate and core power in a natural circulation reactor.

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

[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 allowable 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 through a main steam system line 10.

[0019] In the feedwater system line 13, hot water is sent by a feedwater pump 11 to the reactor pressure vessel 1 via a turbine, a condenser, etc. In such a reactor, a monitoring device 16 is provided to acquire and monitor measurement data from 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 and control the feedwater pump 11, feedwater flow rate 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 static head difference between the inside and outside of the shroud 2 balances mainly with pressure losses such as friction losses, and therefore the core flow rate also changes even if the downcomer water level 9, which determines the static head outside the shroud 2, changes. In general, 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 following equation relates the feedwater flow rate (mass flow rate) Wfw and the generated steam flow rate Wms: Wfw = Wms + Δh (1), where Δh is the difference [m] between the target and measured downcomer water level 9, and is close to zero in the steady state when the target value has been reached. The above equation means that when the downcomer water level 9 is constant, the coolant flow rate that boils and flows out of the reactor as main steam to the steam turbine is supplemented by feedwater. If the downcomer water level 9 changes, Wms and Δh in the above equation also change, so the feedwater flow rate is automatically adjusted in a normal control system to match the downcomer water level 9 to the target value.

[0022] When power output changes from P1 to P2 due to power control, the change in 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 during power increase operation and negative during power decrease operation. 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 core flow rates in the steady state corresponding to power outputs P1 and P2, T is the feedwater time delay [s], and Δt12 is the target value for the power adjustment time [s]. T is the time lag between issuing a command to change the feedwater flow rate and the actual change in the feedwater flow rate into the reactor. It is an inherent value that arises from the design specifications of the feedwater pump, feedwater piping length, and feedwater control system. Furthermore, the above equation means that the feedwater flow rate required to maintain the downcomer water level 9 constant is delayed by ΔWdelay, which means that the downcomer water level 9 cannot be maintained constant. It is known that the relationship between the reactor water level variation Δh and the core flow rate ΔWcore is roughly expressed by the following equation: ΔWcore ∝ Δh (4) Furthermore, when the core flow rate is constant, the relationship between the feedwater flow rate variation ΔWfw and the downcomer water level 9 variation is expressed by the following equation: Δh = ΔWfw / (Ad ρl) (5) Here, Ad is the flow path cross-sectional area [m 2 ], ρl is the liquid density [kg / m 3 If ΔWfw = ΔWdelay is substituted into the above equation, the water level change width Δhdelay caused by the time delay in the water supply flow rate is obtained.

[0023] From the above relational expressions, the variation width ΔWc,d of the core flow rate during power control, taking into account the time delay of the feedwater flow rate, can be predicted by the following equation: ΔWc,d = W2 - W1 + f (ΔWdelay) ... (6) Here, f is a function composed of equations (4) and (5) and expresses the relationship between the variation width of the core flow rate and the variation width of the feedwater flow rate. In the above equation, (W2 - W1) indicates the variation width of the core flow rate due to a change in power, and from 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 information on the fluctuation range of the core flow rate due to power changes and the fluctuation range of the feedwater flow rate as in equation (6), it is possible to grasp the fluctuation range of the core flow rate during power control taking into account the time delay of the feedwater flow rate.

[0025] Figure 2 shows the relationship between power and core flow rate in a comparative example. The dotted lines represent the steady state at water level a and water level b (a < b), respectively. When increasing power by withdrawing control rods from State 1 at water level b to State 2 for power increase operation, if the power change rate is slow, the power increases along the dotted line. On the other hand, if the power change rate is fast and Δt12 cannot be ignored relative to T in equation (3), there is a possibility that the core flow rate will decrease, equivalent to the feedwater flow delay ΔWdelay, as shown by the solid line. Furthermore, at this time, there is a possibility that the core flow rate will fall below the steady state at water level a.

[0026] In boiling water reactors, a water level limit is usually set to detect an emergency shutdown or a tendency toward an emergency shutdown. If water level a is the limit for detecting an abnormality, control rod operation must be stopped when the water level reaches the dotted line of water level a, which may result in a shutdown of the power increase operation. In the case of power decrease operation, the above-mentioned magnitude relationship is reversed, and the core flow rate may increase by ΔWdelay.

[0027] FIG. 3 shows the power-core flow rate relationship in Example 1. A steady state of water level c (a < b < c) is added to FIG. 2 , and on the dotted line representing the steady state of water level c, states corresponding to the same power as State 1 and State 2 are designated State 1' and State 2', respectively. In this example, before performing control rod manipulation (withdrawal in the case of power increase operation), the target value of the downcomer water level 9 is increased from water level b to water level c to change from State 1 to State 1'. This increases the feedwater flow rate, and according to the relationship in Equation (5), the downcomer water level 9 is increased from water level b to water level c. Subsequently, by performing control rod manipulation, even if the core flow rate decreases, the water level reaches State 2' without falling below water level a, thereby completing power increase operation. In the case of power decrease operation, by lowering the water level from c to b in advance from State 2' to State 2, it is possible to operate the reactor while allowing for the increase in core flow rate that occurs when control rods are inserted.

[0028] 4 shows a flowchart of this embodiment. Based on 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. Furthermore, 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, and 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] 4, when the water level has both a lower limit hlow and an upper limit hhigh, the target water level htarget is set within the range of the following equation. Power increase operation is expressed by the following equation: h1 + Δhdelay - |h1 - hlow| < htarget < hhigh ... (7) Power decrease operation is expressed by the following equation: hlow < htarget < h1 - Δhdelay - |hhigh - h1| ... (8) FIG. 5 shows the time changes during power increase in the comparative example and in Example 1. (a) is the comparative example, and (b) is this example. In the comparative example, after control rod operation begins, the core and downcomer water level 9 drops at a certain time and falls below water level a. If water level a is at the warning water level, there is a possibility that control rod operation will be stopped. In this embodiment, 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'), and although the core flow rate subsequently 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 completed, the target value of the downcomer water level 9 is returned to the original water level b, thereby returning the downcomer water level 9 and the core flow rate to their original state (state 2' → state 2).

[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 primarily 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 composed 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, reducing the fluid resistance at the core inlet may result in a decrease in channel stability. 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 using 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] FIG. 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 the downcomer water level 9 is increased by maintaining this level. If the variable resistance orifice 46 is not provided, and 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, the variable resistance orifice 46 provides a larger steady state at water level c than the case without the variable resistance orifice 46 due to the reduced fluid resistance. 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, which is generally the same as that of 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.

[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, then flows back into the reactor mainly via the condenser and feedwater heater.

[0038] However, if the power is increased in a short period of time due to power control, a feedwater flow rate greater than normal may be temporarily 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 in response to an instruction signal from the control device.

[0039] A flowchart of this embodiment is shown in Figure 10. It is generally the same as that of the first embodiment. During power increase operation, after control rod manipulation is initiated, a shortage of feedwater flow rate can be avoided by opening the feedwater tank valve to increase 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.

[0040] In this embodiment, the purpose is to shorten the interval between power increase or decrease operations and to implement 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 embodiments 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 the 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 the 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.

[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... feedwater flow control valve, 30... feedwater tank, 32... feedwater 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. A nuclear power plant having a large-diameter spatial chimney above the core within the shroud of a reactor pressure vessel, comprising: 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 control rods and the feedwater flow rate, wherein 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 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 range of fluctuation in the core flow rate, and then controls by changing the position of the control rod.

2. A nuclear power plant having a large-diameter spatial chimney above the core within the shroud of a reactor pressure vessel, comprising: a coolant flow meter that measures the flow rate of coolant flowing into the reactor core; a water level meter 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 control rods and the feedwater flow rate, wherein 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 the fluctuation in the core flow rate, and changes the target value of the reactor water level to more than the water level fluctuation range equivalent to the range of fluctuation in the core flow rate from the start time of control rod operation to the time of the fluctuation in the core flow rate, thereby changing the feedwater flow rate and changing the water level, and by the end time of control rod operation, changes the target value of the water level to the original water level.

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

4. A nuclear power plant according to claim 2, wherein the time of change in the core flow rate is calculated as the timing at which the increase in the water level due to an increase in the void fraction in the chimney caused by an increase in power output and the decrease in the water level due to the main steam flow rate are balanced.

5. A nuclear power plant according to claim 1 or 2, 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 power changes and the fluctuation range of the feedwater flow rate.

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

7. A nuclear power plant as claimed in claim 6, characterized in that when the predicted fluctuation range of the core flow rate exceeds the core flow rate threshold 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 at which the fluid resistance of the orifice decreases.

8. A nuclear power plant according to claim 1 or 2, comprising a water supply tank connected to a water supply pipe connected to a water supply pump, and a water level meter for measuring the amount of cooling water in the water supply tank, wherein the cooling water in the water supply tank is released into the water supply pipe between the start and end of control rod operation.

9. A method of operating a nuclear power plant having a large-diameter spatial chimney above the core within the shroud of a reactor pressure vessel, comprising: a flow rate measuring step of measuring the flow rate of coolant flowing into the reactor core; a water level measuring step of measuring the water level of the coolant in the reactor pressure vessel; a feedwater flow rate measuring step connected to the reactor pressure vessel and measuring the 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 a control rod 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, changes the feedwater flow rate by changing the target value of the water level to a fluctuation range of the water level equivalent to or greater than the fluctuation range of the core flow rate, and then controls by operating to change 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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