Fast reactor piping rupture detection device and piping rupture detection method

Abstract

To detect breakage of a pipeline in a reactor in a tank type reactor.SOLUTION: A pipeline breakage detection device 100 for a fast reactor comprises: a storage unit 120 which stores characteristic information indicating a relation between a voltage value of voltage supplied to a pump motor 7 of a primary system pump 5 when a core 3 is driven with each of a plurality of heat outputs and a trip set value being effective power when starting a protection operation of the fast reactor; a current measurement unit 134 which measures a current value of the current flowing into the pump motor 7; a voltage measurement unit 133 which measures a voltage value of the voltage supplied to the pump motor 7; a power calculation unit 135 which calculates the effective power consumed in the pump motor 7 on the basis of the measured current value and voltage value; and an output unit 137 which outputs a breakage detection signal indicating that a pipeline 6 in a reactor is broken when the calculated effective power becomes equal to or less than the trip set value associated with the measured voltage value.SELECTED DRAWING: Figure 3

Description

[Technical Field]

[0001] The present invention relates to a pipe rupture detection device and a pipe rupture detection method for detecting rupture of in-core piping of a fast reactor. [Background technology]

[0002] Tank-type nuclear reactors are known in which primary system equipment such as an intermediate heat exchanger, a pump, and primary coolant (liquid sodium) is housed in a main vessel (see, for example, Patent Document 1). In tank-type nuclear reactors, the primary system pump is driven to supply the primary coolant to the reactor core through in-reactor piping to raise its temperature, and the intermediate heat exchanger exchanges heat between the heated primary coolant and the secondary coolant, thereby transferring thermal energy to the secondary coolant. The heat transferred to the secondary coolant is converted into power in a steam generator. [Prior art documents] [Patent documents]

[0003] [Patent Document 1] Japanese Patent Application Publication No. 2018-194350 Summary of the Invention [Problem to be solved by the invention]

[0004] If the in-core piping is damaged and the leakage of primary coolant from the in-core piping increases, the amount of primary coolant supplied to the core decreases, causing the core temperature to rise and resulting in damage to the core. For this reason, there is a demand for a system that can detect damage to the in-core piping.

[0005] Conventional methods for detecting damage to in-core piping include a method for detecting damage to in-core piping based on the core temperature, a method for detecting damage to in-core piping based on the flow rate of the in-core piping, and a method for providing an outer enclosing pipe surrounding the in-core piping and detecting damage to the in-core piping based on the liquid level of the primary coolant in the outer enclosing pipe or the pressure in the outer enclosing pipe.

[0006] The method of detecting damage to in-core piping based on the core temperature has the problem of a long response time due to the heat conduction and heat transfer of the thermometer guide tube, the gas in the guide tube, and the thermometer body. Furthermore, if the gas in the guide tube has good thermal conductivity to shorten the response time, additional equipment for gas replacement and monitoring will be required. Furthermore, if the thermometer is pressed into the guide tube or the guide tube is opened to bring the thermometer into direct contact with the primary coolant, there is a problem of difficulty in replacing the thermometer.

[0007] The method of detecting a rupture of an in-core piping based on the flow rate of the in-core piping and the method of detecting a rupture of an in-core piping based on the liquid level of the primary coolant in the outer envelope pipe or the pressure in the outer envelope pipe have a problem of being difficult to implement because they require the installation of a flow meter in the primary coolant or the installation of an outer envelope pipe, and therefore there is a need for other methods to detect a rupture of an in-core piping.

[0008] The present invention has been made in view of these points, and has as its object to detect breakage of in-core piping in a tank-type nuclear reactor. [Means for solving the problem]

[0009] A piping breakage detection device for a fast reactor according to a first aspect of the present invention is provided in a fast reactor that is a tank-type nuclear reactor, and detects breakage of in-core piping that supplies primary coolant circulating within the fast reactor to the reactor core from a primary system pump, and includes: a memory unit that stores characteristic information indicating the relationship between a voltage value of a voltage supplied to a pump motor provided in the primary system pump when the core is operated at each of a plurality of thermal outputs and a trip value that is an active power when a protective operation of the fast reactor is initiated; a current measurement unit that measures a current value of a current flowing into the pump motor; a voltage measurement unit that measures a voltage value of a voltage supplied to the pump motor; a power calculation unit that calculates the active power based on the current value measured by the current measurement unit and the voltage value measured by the voltage measurement unit; and an output unit that outputs a breakage detection signal that indicates that the in-core piping has been broken when the active power calculated by the power calculation unit becomes equal to or less than the trip value that is stored in the memory unit in association with the voltage value measured by the voltage measurement unit.

[0010] The fast reactor piping break detection device may further include a setting unit that sets the voltage corresponding to the thermal power of the core, and a motor control unit that controls the pump motor using a VVVF inverter so that the voltage set by the setting unit is applied to the pump motor at a frequency that keeps a constant ratio to the voltage.

[0011] A piping failure detection method for a fast reactor according to a second aspect of the present invention is executed by a computer that is installed in the fast reactor, which is a tank-type nuclear reactor, and that detects a failure of in-core piping that supplies primary coolant circulating within the fast reactor to the core from a primary system pump, the computer comprising the steps of: measuring a current value of a current flowing into a pump motor installed in the primary system pump when the core of the fast reactor is operated at each of a plurality of thermal outputs; measuring a voltage value of a voltage supplied to the pump motor; calculating active power consumed by the pump motor based on the measured current value and voltage value; and referring to characteristic information that indicates a relationship between the voltage value of the voltage supplied to the pump motor when the core is operated at each of a plurality of thermal outputs and a trip value that is active power at which a protective operation of the fast reactor is initiated, and outputting a failure detection signal that indicates that the in-core piping has been damaged when the calculated active power becomes equal to or less than the trip value associated with the measured voltage value. [Effects of the Invention]

[0012] According to the present invention, it is possible to detect damage to in-core piping in a tank-type nuclear reactor. [Brief explanation of the drawings]

[0013] [Figure 1] 1 is a diagram showing a schematic configuration of a tank-type reactor. [Figure 2] FIG. 10 is a diagram showing the relationship between an operation plan value, which is a power value of active power relative to a voltage supplied to a pump motor, and a trip setting value. [Figure 3] FIG. 1 is a diagram illustrating a configuration of a pipe breakage detection device. [Figure 4] FIG. 10 is a diagram showing an example of a simulation result when a damage detection signal is output. [Figure 5] 4 is a flowchart showing a process flow relating to detection of a pipe breakage of an in-furnace pipe in the pipe breakage detection device. DETAILED DESCRIPTION OF THE INVENTION

[0014] [Structure of Tank-Type Reactor 1] The fast reactor pipe rupture detection device 100 according to this embodiment is installed in a fast reactor, which is a tank-type reactor 1, and is a device for detecting rupture of in-core piping for supplying sodium, which serves as primary coolant circulating within the fast reactor, from a primary system pump to the reactor core. In explaining the fast reactor pipe rupture detection device 100, first, the structure of the tank-type reactor 1 will be explained with reference to the drawings. FIG. 1 is a diagram showing a schematic configuration of the tank-type reactor 1. In FIG. 1, the flow of sodium as coolant is indicated by arrows. In the following explanation, the fast reactor pipe rupture detection device 100 will be simply referred to as the pipe rupture detection device 100.

[0015] The tank-type reactor 1 is, for example, a fast reactor. As shown in Fig. 1, a reactor core 3, an intermediate heat exchanger 4, a primary system pump 5, and in-reactor piping 6 are provided inside a main vessel 2 of the tank-type reactor 1. Sodium, which is a liquid metal, is contained inside the main vessel 2 as a primary coolant. A pump motor 7 is connected to the primary system pump 5, and an inverter 8 is connected to the pump motor 7.

[0016] The main vessel 2 is a vessel with a diameter of approximately 15 to 20 meters. The reactor core 3 is supported horizontally inside the main vessel 2. The reactor core 3 is provided with core fuel containing fissionable material and control rods for controlling the core reactivity. The control rods are driven by a control rod drive mechanism. The control rod drive mechanism controls the amount of insertion of the control rods into the core fuel. This controls the nuclear fission of the core fuel and the thermal output in the reactor core 3. The reactor core 3 heats up sodium, which serves as the primary coolant. In the following explanation, sodium before it is heated is referred to as low-temperature sodium, and sodium that has been heated to a high temperature is referred to as high-temperature sodium.

[0017] The intermediate heat exchanger 4 has an inlet window through which high-temperature sodium flows in and an outlet window through which low-temperature sodium flows out after heat exchange. The intermediate heat exchanger 4 exchanges heat between the high-temperature sodium that flows in through the inlet window and the sodium that functions as the secondary coolant. Specifically, the high-temperature sodium, whose temperature has risen to approximately 550°C in the reactor core 3, flows into the inlet of the intermediate heat exchanger 4 by the action of the primary system pump 5. The high-temperature sodium that flows in exchanges heat with the secondary system sodium that functions as the secondary coolant, thereby becoming low-temperature sodium whose temperature has dropped to approximately 400°C. The low-temperature sodium flows out through the outlet window to the bottom of the main vessel 2. The secondary system sodium flows into the steam generator 9, where it heats water and generates steam to drive a turbine.

[0018] There are multiple primary pumps 5, and although not shown, they are provided on the circumference where multiple intermediate heat exchangers 4 are installed. The primary pumps 5 are driven by pump motors 7 that are variable speed controlled by an inverter 8, i.e., VVVF (Variable Voltage Variable Frequency) inverter controlled. The inverter 8 controls the pump motors 7 by constant V / F control that changes the output voltage in proportion to the frequency.

[0019] The primary system pumps 5 pump the low-temperature sodium that has flowed out of the intermediate heat exchanger 4 into the in-core piping 6, thereby supplying the low-temperature sodium to the reactor core 3. The in-core piping 6 guides the low-temperature sodium pumped by the primary system pumps 5 to the reactor core 3.

[0020] In a fast reactor, the flow rate of sodium as primary coolant is set to be proportional to the reactor thermal power output from partial load to rated power operation. For this reason, the pump motor 7 of the primary system pump 5 is VVVF inverter controlled so that the flow rate of low-temperature sodium increases in proportion to the reactor thermal power output from partial load to rated power operation.

[0021] The load torque applied to the pump motor 7 due to pressure loss when sodium passes through the in-core piping 6, the reactor core 3, the intermediate heat exchanger 4, etc. is proportional to the square of the rotation speed of the pump motor 7. The rotation speed of the pump motor 7 is roughly proportional to the output frequency of the inverter 8. Furthermore, because the control method for the pump motor 7 is constant V / F control, the output frequency of the inverter 8 is proportional to the output voltage. Therefore, the load torque applied to the pump motor 7 due to the pressure loss of sodium is proportional to the square of the output voltage. Furthermore, the torque of the pump motor 7, which balances the load torque, is proportional to the current (torque current) flowing into the pump motor 7. Therefore, the current flowing into the pump motor 7 increases roughly as a square characteristic of the voltage, and the active power of the pump motor 7 increases roughly as a cube characteristic of the voltage.

[0022] If an abnormality such as breakage occurs in the in-core piping 6, sodium will bypass the reactor core 3 and intermediate heat exchanger 4, which have high resistance, and therefore its pressure loss will decrease, and the load torque generated by the pressure loss of sodium will also decrease. As a result, the torque of the pump motor 7, which balances the load torque, will also decrease. In contrast, a voltage is applied to the pump motor 7 so that low-temperature sodium is supplied to the reactor core 3 in accordance with the reactor thermal power, and the rotation speed of the pump motor 7 does not change much, so the current flowing into the pump motor 7 decreases. As a result, if an abnormality such as breakage occurs in the in-core piping 6, the active power at the same voltage value will decrease compared to when no abnormality such as breakage occurs in the in-core piping 6. In the following explanation, breakage in the in-core piping 6 is, for example, a guillotine rupture, in which the in-core piping 6 is instantly cut in half.

[0023] When the in-core piping 6 is damaged, the current flowing into the pump motor 7 decreases, and so it is conceivable to detect damage to the in-core piping 6 by monitoring the current. However, the current flowing into the pump motor 7 fluctuates greatly, which can lead to a problem of falsely detecting damage to the in-core piping 6. If a large margin is secured for the current value that serves as the reference for starting protective operations of the fast reactor in order to prevent this problem, a problem arises in that it takes a long time from when the in-core piping 6 is damaged until protective operations of the fast reactor are started.

[0024] In contrast, the active power calculated by multiplying the voltage and current offsets the fluctuations of the voltage and current, so the margin of the reference value for starting protective action for the fast reactor can be made smaller. As a result, when detecting damage to the in-reactor piping 6 by monitoring the active power supplied to the pump motor 7, protective action for the fast reactor can be started in a shorter time than when detecting damage to the in-reactor piping 6 by monitoring the current flowing into the pump motor 7.

[0025] To determine the trip setting, which is a reference value for detecting a break in the in-core piping 6 due to a drop in active power that requires the initiation of protective operation of the fast reactor, for example, the primary pump 5 is operated before reactor power operation, and the voltage values ​​supplied to the pump motors 7 corresponding to each of the multiple thermal outputs of the reactor core 3 and the current values ​​flowing into the pump motors 7 are measured. The trip setting is determined by identifying a relationship between the active power corresponding to the voltage value supplied to the pump motors 7 when no break has occurred in the in-core piping 6, based on the measured voltage and current values ​​and differences in coolant density due to differences in coolant temperature, and adding a margin to the identified result. The trip setting is set taking into account errors in the current and voltage measurements and fluctuations in the voltage of the pump motors 7 that occur when no break has occurred in the in-core piping 6.

[0026] Fig. 2 is a diagram showing the relationship between the operation plan value, which is the value of the active power for the voltage supplied to the pump motor 7, and the trip setting value. The operation plan value is the value of the active power when the reactor is operated in a normal state with no damage to the in-reactor piping 6. As shown in Fig. 2, it can be seen that the operation plan value increases with an increase in voltage, and that the trip setting value is set to a value lower than the operation plan value for each of the multiple voltages.

[0027] The piping breakage detection device 100 measures the voltage value of the pump motor 7 and the active power during operation of the fast reactor, and detects breakage in the in-reactor piping 6 by determining whether the measured active power is equal to or less than a trip setting value corresponding to the measured voltage value. When the measured active power falls below the trip setting value, the piping breakage detection device 100 generates a breakage detection signal as a trip signal for initiating protective operations for the fast reactor, and starts protective operations for the fast reactor by transmitting the breakage detection signal to a trip control device (not shown) that causes the reactor shutdown system to perform an emergency shutdown of the fast reactor. As described above, by using the active power, the piping breakage detection device 100 generates a breakage detection signal in a short time after a break occurs in the in-reactor piping 6, and therefore protective operations for the fast reactor are started promptly.

[0028] [Configuration of pipe breakage detection device 100] Next, the configuration of the pipe breakage detection device 100 will be described. Fig. 3 is a diagram showing the configuration of the pipe breakage detection device 100. The pipe breakage detection device 100 is, for example, a computer, and has a communication unit 110, a storage unit 120, and a control unit 130. The control unit 130 has a setting unit 131, an electric motor control unit 132, a voltage measurement unit 133, a current measurement unit 134, a power calculation unit 135, a determination unit 136, and an output unit 137.

[0029] The communication unit 110 is, for example, a communication interface that enables the pipe break detection device 100 to communicate with other devices such as a voltmeter 10 and an ammeter 11. The voltmeter 10 includes, for example, an instrument transformer. The voltmeter 10 is provided between the inverter 8 and the pump motor 7 and measures the primary voltage of the pump motor 7. The ammeter 11 includes, for example, a current transformer. The ammeter 11 is provided between the inverter 8 and the pump motor 7 and measures the primary current of the pump motor 7.

[0030] The storage unit 120 is, for example, a read-only memory (ROM) and a random access memory (RAM). The storage unit 120 stores various programs for causing the pipe breakage detection device 100 to function. For example, the storage unit 120 stores programs for causing the control unit 130 of the pipe breakage detection device 100 to function as a setting unit 131, a motor control unit 132, a voltage measurement unit 133, a current measurement unit 134, a power calculation unit 135, a determination unit 136, and an output unit 137.

[0031] The storage unit 120 also stores characteristic information indicating the relationship between the voltage value of the voltage supplied to the pump motor 7 provided in the primary pump 5 when the core 3 is operated at each of a plurality of thermal outputs and the trip setting value, which is the active power when starting the protective operation of the fast reactor. The characteristic information is information indicating the relationship between the voltage values ​​of the plurality of voltages supplied to the pump motor 7 and the trip setting value, as shown by the dashed lines in Fig. 2, and is, for example, a table associating the trip setting value with each of a plurality of voltage values.

[0032] The control unit 130 is, for example, a CPU (Central Processing Unit). The control unit 130 executes an output control program stored in the storage unit 120, thereby functioning as a setting unit 131, a motor control unit 132, a voltage measurement unit 133, a current measurement unit 134, a power calculation unit 135, a determination unit 136, and an output unit 137.

[0033] The setting unit 131 sets a voltage to be supplied to the pump motor 7 corresponding to the thermal output of the reactor core 3. For example, the memory unit 120 stores voltage setting information indicating voltages corresponding to a plurality of thermal outputs. The setting unit 131 receives an instruction value for the thermal output of the reactor core 3 from a control device (not shown) that controls the output of the reactor via the communication unit 110. The setting unit 131 refers to the voltage setting information and sets the voltage associated with the received instruction value as the voltage to be supplied to the pump motor 7.

[0034] The motor control unit 132 performs VVVF inverter control on the pump motor 7 so that the voltage set by the setting unit 131 is applied to the pump motor 7 at a frequency that keeps a constant ratio to the voltage. The motor control unit 132 uses the inverter 8 connected to the pump motor 7 to control the pump motor 7 by V / F constant control that changes the output voltage in proportion to the frequency.

[0035] In the present embodiment, the pipe breakage detection device 100 sets the voltage to be supplied to the pump motor 7 and controls the pump motor 7 so that the voltage is applied to the pump motor 7, but this is not limiting. For example, a device other than the pipe breakage detection device 100 may set the voltage to be supplied to the pump motor 7 and control the pump motor 7 so that the voltage is applied to the pump motor 7.

[0036] The voltage measurement unit 133 measures the voltage value of the voltage supplied to the pump motor 7. For example, the voltage measurement unit 133 measures the voltage value of the voltage supplied to the pump motor 7 by acquiring voltage value information indicating the primary voltage of the pump motor 7 measured by a voltmeter 10 provided between the pump motor 7 and the inverter 8.

[0037] The current measurement unit 134 measures the current value of the current flowing into the pump motor 7. For example, the current measurement unit 134 measures the current value of the primary current flowing into the pump motor 7 by acquiring current value information that indicates the current value of the current flowing into the pump motor 7, measured by an ammeter 11 provided between the pump motor 7 and the inverter 8.

[0038] The power calculation unit 135 calculates the active power consumed by the pump motor 7 based on the voltage value of the voltage supplied to the pump motor 7 measured by the voltage measurement unit 133 and the current value of the current flowing into the pump motor 7 measured by the current measurement unit 134. The power calculation unit 135 calculates the active power consumed by the pump motor 7 by multiplying the voltage value of the voltage measured by the voltage measurement unit 133 by the current value of the current measured by the current measurement unit 134, taking into account the phase difference between the voltage and the current.

[0039] The determination unit 136 determines whether the active power calculated by the power calculation unit 135 is equal to or less than the trip set value that is associated with the voltage value measured by the voltage measurement unit 133 and stored in the storage unit 120. Specifically, the determination unit 136 first refers to the table stored in the storage unit 120 and identifies the trip set value that is associated with the voltage value measured by the voltage measurement unit 133. Next, the determination unit 136 determines whether the active power calculated by the power calculation unit 135 is equal to or less than the identified trip set value.

[0040] When the active power calculated by the power calculation unit 135 becomes equal to or less than the trip setting value stored in the memory unit 120 in association with the voltage value measured by the voltage measurement unit 133, the output unit 137 outputs a damage detection signal indicating that the in-reactor piping 6 has been damaged. For example, when the determination unit 136 determines that the active power calculated by the power calculation unit 135 is equal to or less than the trip setting value, the output unit 137 outputs a damage detection signal via the communication unit 110 to a trip control device that causes the reactor shutdown system to perform an emergency shutdown of the reactor, in order to start a protective operation for the reactor.

[0041] Fig. 4 is a diagram showing an example of a simulation result when a damage detection signal is output. In the example shown in Fig. 4, the horizontal axis represents the time elapsed since the in-core piping 6 was damaged, and the vertical axis represents the active power supplied to the pump motor 7. Fig. 4 shows a simulation result when a damage detection signal is output when the reactor is operated at rated output and the voltage supplied to the pump motor 7 is 6600 V. When the voltage supplied to the pump motor 7 is 6600 V, the active power when no damage has occurred in the in-core piping 6 is approximately 2150 kW, and the trip setting value is approximately 1750 kW as shown in Fig. 2.

[0042] 4, the solid line indicates the active power, the dashed line indicates the trip setting value, and the dotted line indicates the load torque applied to the pump motor 7. As shown in FIG. 4, it can be seen that the load torque applied to the pump motor 7 drops sharply in response to the occurrence of breakage in the internal piping 6. Although there is a slight delay with respect to the drop in the load torque applied to the pump motor 7, the active power supplied to the pump motor 7 drops sharply, and the active power becomes equal to or less than the trip setting value in about 0.15 seconds. As a result, the output unit 137 can output a breakage detection signal in about 0.15 seconds after the internal piping 6 is broken.

[0043] [Processing flow in the pipe breakage detection device 100] Next, a description will be given of the flow of processing related to detection of a pipe breakage of the in-furnace piping 6 in the pipe breakage detection device 100. Fig. 5 is a flowchart showing the flow of processing related to detection of a pipe breakage of the in-furnace piping 6 in the pipe breakage detection device 100. The processing shown in Fig. 5 is repeatedly executed during operation of the pipe breakage detection device 100.

[0044] First, the voltage measurement unit 133 measures the voltage value of the voltage supplied to the pump motor 7 (S1). Next, the current measurement unit 134 measures the current value of the primary current flowing through the pump motor 7 (S2). In this flowchart, the process related to S2 is executed after the process related to S1 is executed, but this is not limited to this. The process related to S1 may be executed after the process related to S2 is executed, or the process related to S1 and the process related to S2 may be executed in parallel.

[0045] Next, the power calculation unit 135 calculates the active power consumed by the pump motor 7 based on the voltage value of the voltage supplied to the pump motor 7 measured in S1 and the current value of the current flowing into the pump motor 7 measured in S2 (S3). Next, the determination unit 136 refers to the table stored in the storage unit 120 and identifies the trip setting value associated with the voltage value of the voltage supplied to the pump motor 7 measured in S1 (S4).

[0046] Next, the determination unit 136 determines whether the active power calculated in S3 is equal to or less than the trip set value specified in S4 (S5). If the determination unit 136 determines that the active power is equal to or less than the trip set value, it proceeds to S6, and if it determines that the active power is greater than the trip set value, it terminates the processing according to this flowchart. In S6, the output unit 137 outputs a damage detection signal indicating that the in-furnace piping 6 has been damaged.

[0047] [Effects of this embodiment] As described above, the piping breakage detection device 100 according to this embodiment stores in the memory unit 120 characteristic information indicating the relationship between the voltage value of the voltage supplied to the pump motor 7 provided in the primary system pump 5 when the reactor core 3 is operated at each of a plurality of thermal outputs and the trip setting value, which is the active power when protective operation of the fast reactor is initiated, and outputs a breakage detection signal indicating that the in-core piping 6 has been broken when the active power consumed in the pump motor 7 falls below the trip setting value associated with the voltage value of the voltage supplied to the pump motor 7.

[0048] In this way, damage to the in-core piping 6 in a tank-type nuclear reactor can be detected in a short time. In addition, the voltage transformer and current transformer that make up the voltmeter 10 and ammeter 11 can be installed in an electrical panel room with good environmental conditions, and replacement is easy. Therefore, the equipment for detecting damage to the in-core piping 6 can be simplified.

[0049] The present invention has been described above using embodiments, but the technical scope of the present invention is not limited to the scope described in the above embodiments, and various modifications and changes are possible within the scope of the gist of the present invention. All or part of the device can be configured by functionally or physically distributing or integrating in any unit. Furthermore, new embodiments resulting from any combination of multiple embodiments are also included in the embodiments of the present invention. The effects of new embodiments resulting from the combination combine the effects of the original embodiments. [Explanation of symbols]

[0050] 1. Tank-type reactor 2 Main vessel 3. Reactor core 4 Intermediate heat exchanger 5 Primary pump 6 Furnace piping 7. Pump motor 8 inverters 9. Steam Generator 10. Voltmeter 11 Ammeter 100 Pipe breakage detection device 110 Communications Department 120 Storage section 130 control section 131 Setting section 132 Motor control unit 133 Voltage measurement section 134 Current measurement section 135 Power calculation unit 136 Judgment section 137 Output section

Claims

1. A piping rupture detection device for a fast reactor, which is a tank-type nuclear reactor, is provided to detect rupture of an in-core piping for supplying a primary coolant circulating in the reactor of the fast reactor from a primary system pump to a reactor core, a storage unit that stores characteristic information indicating the relationship between a voltage value of a voltage supplied to a pump motor provided in the primary system pump when the reactor core is operated at each of a plurality of thermal outputs and a trip setting value that is an active power when a protective operation of the fast reactor is started; a current measuring unit that measures a current value of a current flowing into the pump motor; a voltage measuring unit that measures a voltage value of a voltage supplied to the pump motor; a power calculation unit that calculates the active power based on the current value measured by the current measurement unit and the voltage value measured by the voltage measurement unit; an output unit that outputs a damage detection signal indicating that the in-furnace piping has been damaged when the effective power calculated by the power calculation unit becomes equal to or less than the trip setting value that is associated with the voltage value measured by the voltage measurement unit and stored in the storage unit; A pipe break detection device for a fast reactor.

2. a setting unit that sets the voltage corresponding to the thermal power of the core; a motor control unit that controls the pump motor by a VVVF inverter so that the voltage set by the setting unit is applied to the pump motor at a frequency that keeps a constant ratio to the voltage; further comprising The pipe breakage detection device for a fast reactor according to claim 1.

3. A computer that is provided in a fast reactor that is a tank-type nuclear reactor and detects a break in an in-core piping that supplies primary coolant circulating in the fast reactor from a primary system pump to a reactor core, measuring a current value of a current flowing into a pump motor provided in the primary system pump when the core of the fast reactor is operated at each of a plurality of thermal outputs; measuring a voltage value of a voltage supplied to the pump motor; calculating an active power consumed by the pump motor based on the measured current value and voltage value; a step of referring to characteristic information indicating the relationship between a voltage value of a voltage supplied to the pump motor when the reactor core is operated at each of a plurality of thermal outputs and a trip setting value which is an active power when a protective operation of the fast reactor is initiated, and outputting a damage detection signal indicating that the in-core piping has been damaged when the calculated active power becomes equal to or less than the trip setting value associated with the measured voltage value; Method for detecting piping damage in a fast reactor.

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