JP7237869B2 - Reactor water level measurement system and reactor water level measurement method - Google Patents

Description

TECHNICAL FIELD Embodiments of the present invention relate to a reactor water level measurement system and a reactor water level measurement method.

When a severe accident occurs in a nuclear reactor plant, it is necessary to grasp the situation inside the reactor in order to determine countermeasures. In particular, information on conditions inside the reactor, such as the water level inside the reactor pressure vessel (RPV), is extremely important for understanding the state of fuel cooling or damage. As a method for measuring the water level in the reactor, in a typical boiling water reactor (BWR), a condensation tank and a reference water column detection pipe connected to the vapor phase portion of the RPV and a fluctuating water column detection pipe connected to the liquid phase portion of the RPV are used. A differential pressure type water level gauge that detects the differential pressure is used. However, in the accident at the nuclear power plant caused by the Great East Japan Earthquake, water in the reference water column detection pipe and fluctuating water column detection pipe inside the containment vessel (PCV) became hot and evaporated, making it impossible to measure the reactor water level. rice field. After this accident, several water level gauges with different principles have been proposed to accurately measure the reactor water level even in severe accidents.

JP 2007-205912 A JP 2008-170223 A

While water level gauges have been proposed assuming severe accidents, the conventional differential pressure type is also effective for measuring the reactor water level during normal operation, and is considered highly reliable from many years of experience. . During a severe accident, it is extremely important to measure the reactor water level.

The embodiment of the present invention has been made in consideration of such circumstances, and aims to provide a reactor water level measurement technique capable of consistently measuring the reactor water level from normal operation to severe accident. aim.

A nuclear reactor water level measurement system according to an embodiment of the present invention includes an accident analysis unit that analyzes a severe accident of a nuclear reactor and analyzes the water level in the reactor corresponding to the progress of the severe accident; A water level measurement value analysis unit that analyzes the water level measurement values measured according to the progress by a plurality of types of water level gauges each having a different measurement principle, and the reactor water level analyzed by the accident analysis unit. a prediction model creation unit that analyzes the relationship between the water level measurement values analyzed by the water level measurement value analysis unit and creates a prediction model that is used to predict the in-core water level from a plurality of the water level measurement values; Created by a data accumulation unit that accumulates the water level measurement values actually measured in the reactor by the water level gauge of the type, and the water level measurement values accumulated in the data accumulation unit and the prediction model creation unit a reactor water level prediction unit that predicts the actual in-core water level of the nuclear reactor based on the prediction model; and an internal information output unit.

Embodiments of the present invention provide a reactor water level measurement technique capable of consistently measuring the reactor water level from normal operation to severe accident.

1 is a block diagram showing a reactor water level measurement system of this embodiment; FIG. Sectional drawing which shows a nuclear reactor. Explanatory drawing which shows the measurement principle of a thermocouple-type water level gauge. The graph which shows the example of the output signal of a thermocouple-type water level gauge. A graph showing the measurement principle of a radiological water level gauge. Explanatory drawing which shows the in-core situation data set. Explanatory drawing which shows a measured value data set. A graph showing the reactor water level at the time of a severe accident. A graph showing the reactor pressure at the time of a severe accident. A graph showing the temperature inside the reactor at the time of a severe accident. Graph showing measurements of multiple types of water level gauges. Graph showing the actual water level in the reactor and the predicted water level in the reactor. 4 is a flowchart showing a reactor water level measurement method;

Hereinafter, embodiments of a reactor water level measurement system and a reactor water level measurement method will be described in detail with reference to the drawings.

Reference numeral 1 in FIG. 1 denotes a reactor water level measurement system of this embodiment. This reactor water level measurement system 1 is used to consistently measure the reactor water level in a nuclear reactor plant from the time of normal operation to the time of a severe accident.

A reactor water level measurement system 1 includes a water level measurement unit 2 , an in-core measurement unit 3 and a water level prediction computer 4 .

The water level measurement unit 2 includes a plurality of types (three types) of water level gauges 5 to 7 each having a different measurement principle. The water level measurement unit 2 of the present embodiment includes a differential pressure type water level gauge 5 as a first class water level gauge, a thermocouple water level gauge 6 as a second type water level gauge, and a radiation type water level gauge as a third type water level gauge. 7.

At least two of these water level gauges 5 to 7 should be provided. By doing so, it is possible to provide a plurality of types of water gauges 5 to 7 having different durability against severe accidents. Preferably, at least one water gauge 7 operates without loss of function from normal operation to severe accidents. In addition to these water level gauges 5 to 7, water level gauges having other measurement principles may be provided.

The in-furnace measurement unit 3 includes an in-furnace thermometer 8 and an in-furnace pressure gauge 9 . Furthermore, the in-furnace measurement unit 3 may be equipped with other measuring devices for grasping the conditions inside the furnace. The furnace thermometer 8 measures the temperature of the atmosphere inside the pressure vessel 22 (FIG. 2) (gas phase temperature) and the temperature of the reactor water 24 (liquid phase temperature). The furnace pressure gauge 9 measures the air pressure inside the pressure vessel 22 .

The water level prediction computer 4 includes a data accumulation unit 10, an accident analysis unit 11, a water level measurement value analysis unit 12, a prediction model creation unit 13, an in-core water level prediction unit 14, an in-core information output unit 15, an information input unit 16, and a storage unit. 17 and a main control unit 18 .

The data accumulation unit 10 accumulates water level measurement values (indicated values) actually measured in the nuclear reactor by a plurality of types of water level gauges 5-7. Each of the water level gauges 5 to 7 independently measures the water level in the reactor and transmits the water level measurement value to the data accumulation unit 10 .

The data accumulation unit 10 of the present embodiment accumulates in-core temperature measurement values actually measured in the reactor by the in-core thermometer 8 . Further, the data accumulation unit 10 accumulates measured values of the in-core pressure actually measured in the reactor by the in-core pressure gauge 9 .

The data accumulation unit 10 accumulates the time when the water level measurement value was measured in each of the water level gauges 5 to 7 in association with each water level measurement value. In this way, the water level in the reactor can be predicted based on the water level measurement value corresponding to the situation of the severe accident progressing in chronological order. The accumulated time may be the elapsed time from the occurrence of the severe accident.

The accident analysis unit 11 analyzes the severe accident of the nuclear reactor in advance and analyzes the reactor water level (true value) corresponding to the progress of the severe accident.

The accident analysis unit 11 of the present embodiment simulates a severe accident, and changes over time corresponding to progress of each parameter indicating at least one of the reactor temperature, reactor pressure, molten state of fuel, and the reactor water level. is analyzed, and a reactor situation data set (Fig. 6) is created in which the elapsed time from the occurrence of the severe accident is associated with the parameter. In this way, the water level in the reactor can be obtained by grasping the reactor conditions corresponding to the elapsed time of the severe accident. The simulation performed by the accident analysis unit 11 includes not only the situation during a severe accident but also the situation inside the reactor during normal operation.

Based on the severe accident analysis using the accident analysis unit 11, the water level measurement value analysis unit 12 analyzes the water level measurement values measured by the respective water level gauges 5 to 7 according to the progress of the severe accident.

The water level measurement value analysis unit 12 of the present embodiment analyzes water level measurement values measured by multiple types of water level gauges 5 to 7 based on the in-core situation data set (FIG. 6) created by the accident analysis unit 11. Then, a measured value data set (FIG. 7) is created in which the elapsed time and the water level measured value are associated with each other. In this way, the water level in the reactor can be obtained based on the measured water level corresponding to the elapsed time of the severe accident.

The prediction model creation unit 13 analyzes the relationship between the reactor water level analyzed by the accident analysis unit 11 and the water level measurement value analyzed by the water level measurement value analysis unit 12 . In addition, a prediction model is created that is used to predict the reactor water level from multiple water level measurements.

The prediction model is used to predict reactor water level from multiple water level measurements by performing multiple regression analysis. In this way, the in-core water level can be predicted from a plurality of water level measurement values by arithmetic processing based on the prediction model. The prediction model is a function for predicting the reactor water level, that is, a function for obtaining the reactor water level (true value) from the water level measurement values (indicated values) measured by multiple types of water level gauges 5 to 7. Including configuration.

The prediction model creation unit 13 of the present embodiment uses the reactor water level as the objective variable and the water level measurement value as the explanatory variable based on the reactor situation data set (FIG. 6) and the measured value data set (FIG. 7). A prediction model is developed that is used to predict reactor water level from multiple water level measurements by performing regression analysis. In this way, a prediction model corresponding to elapsed time can be created.

Further, the predictive model creation unit 13 may create a predictive model that is used to predict the reactor water level by adding values other than the water level measurement value as explanatory variables. The values other than the water level measurement value include at least one of the furnace pressure value and the furnace temperature value. In this way, a value other than the water level measurement value is used to predict the in-core water level, so prediction accuracy can be improved.

Also, the predictive model creation unit 13 analyzes the relationship between the water level measurement value measured by the water level gauge after being damaged in the severe accident and the reactor water level, and creates a predictive model. In this way, the reactor water level can be predicted based on information including the fact that the water level gauges 5 and 6 have been damaged.

The in-core water level prediction unit 14 predicts the actual in-reactor water level based on the water level measurement values accumulated in the data accumulation unit 10 and the prediction model created by the prediction model creation unit 13 .

The in-core information output unit 15 outputs in-core information about the in-core water level predicted by the in-core water level prediction unit 14 .

The water level prediction computer 4 of this embodiment includes a display device such as a display for outputting analysis results. That is, the in-furnace information output unit 15 controls the image displayed on the display. The display may be separate from the computer main body, or may be integrated with the computer main body. Further, the in-core information output unit 15 may control the image displayed on the display of another computer connected via the network.

In addition, although a display is exemplified as a display device in the present embodiment, other modes may be used. For example, information may be displayed using a projector. Furthermore, a printer that prints information on a paper medium may be used as an alternative to the display. In other words, the target controlled by the in-furnace information output unit 15 may include a projector or a printer.

Predetermined information is input to the information input unit 16 according to the operation of the user who uses the water level prediction computer 4 . The information input unit 16 includes an input device such as a mouse or keyboard. That is, predetermined information is input to the information input unit 16 according to the operation of these input devices.

The storage unit 17 stores predetermined information. This storage unit 17 is composed of a memory or an HDD. The storage unit 17 may store a database, which is a collection of information organized so that it can be searched or stored. For example, the storage unit 17 may store the prediction model created by the prediction model creation unit 13 .

The main control unit 18 controls the water level prediction computer 4 in an integrated manner. Note that the main control section 18 may control the water level measurement unit 2 .

The water level prediction computer 4 of this embodiment has hardware resources such as a CPU, ROM, RAM, and HDD, and the CPU executes various programs to implement information processing by software using the hardware resources. . Furthermore, the reactor water level measurement method of this embodiment is implemented by causing a computer to execute various programs.

At least the accident analysis unit 11, the water level measurement value analysis unit 12, the prediction model generation unit 13, and the in-core water level prediction unit 14 are realized by executing programs stored in the memory or HDD by the CPU.

As shown in FIG. 2, a nuclear reactor 19 provided in a nuclear power plant contains a core 21 composed of a plurality of fuel assemblies 20, a pressure vessel 22 that stores the core 21, and the pressure vessel 22. A reactor containment vessel 23 is provided.

The pressure vessel 22 is a container for the reactor core 21, and is a structure made of stainless steel that allows water as a coolant to flow between it and the outside while withstanding the high temperature and high pressure inside. Furthermore, the reactor containment vessel 23 that houses the pressure vessel 22 also has a function of blocking the outside so that the radioactive materials and radiation generated in the reactor core 21 do not leak. The water level in the reactor indicates the water level of the reactor water 24 inside the pressure vessel 22 . That is, it indicates the value of the height from the bottom of the pressure vessel 22 to the water surface.

Inside the pressure vessel 22, a steam separator 25 for separating steam from the reactor water 24 heated in the reactor core 21 and a steam dryer 26 for drying the separated steam are provided. In addition, many other core internals are provided. Further, a recirculation pump 27 for circulating the reactor water 24 inside the pressure vessel 22 is provided at the bottom of the pressure vessel 22 .

The nuclear reactor 19 of the present embodiment is a boiling water reactor ABWR (advanced boiling water reactor) in which the containment vessel 23 and the suppression chamber are integrally provided. It can be. For example, the present embodiment may be applied to a boiling water reactor (BWR) in which the containment vessel 23 and the suppression chamber (torus chamber) are separately provided, or to a pressurized water reactor (PWR). This embodiment may be applied.

The nuclear reactor 19 of the present embodiment includes a differential pressure type water level gauge 5 as a first class water level gauge, a thermocouple type water level gauge 6 as a second type water level gauge, and a radiation type water level gauge as a third class water level gauge. 7 are provided.

The differential pressure type water level gauge 5 measures the reactor water level inside the pressure vessel 22 of the reactor 19 based on the pressure difference between the upper portion and the lower portion of the pressure vessel 22 .

The differential pressure type water gauge 5 includes a condensation tank 28, a reference water column detection pipe 29, a first fluctuating water column detection pipe 30, a second fluctuating water column detection pipe 31, a broadband differential pressure detector 32, and a fuel zone differential pressure detector 33. .

The condensation tank 28 is provided inside the containment vessel 23 and outside the pressure vessel 22 . This condensation tank 28 is connected to the upper portion of the pressure vessel 22 . Water is accumulated in the condensation tank 28 .

The reference water column detection pipe 29 is provided outside the pressure vessel 22 . This reference water column detection pipe 29 extends obliquely downward from the condensation tank 28 and extends to the outside of the reactor containment vessel 23 . The reference water column detection pipe 29 branches in two directions. Each tip of the reference water column detection pipe 29 is connected to a broadband differential pressure detector 32 and a fuel range differential pressure detector 33 .

The first fluctuating water column detection pipe 30 is provided outside the pressure vessel 22 . The first fluctuating water column detection pipe 30 is connected to the vertical central portion of the pressure vessel 22 , extends obliquely downward, and further extends to the outside of the reactor containment vessel 23 . The tip of the first fluctuating water column detection pipe 30 is connected to a broadband differential pressure detector 32 .

The second fluctuating water column detection pipe 31 is provided outside the pressure vessel 22 . This second fluctuating water column detection pipe 31 is connected to the bottom of the pressure vessel 22 , extends obliquely downward, and further extends to the outside of the containment vessel 23 . The tip of the second fluctuating water column detection pipe 31 is connected to the fuel zone differential pressure detection section 33 .

Water accumulates inside the reference water column detection pipe 29 , the first fluctuating water column detection pipe 30 , and the second fluctuating water column detection pipe 31 . By detecting the pressure of the water inside as a differential pressure, the water level in the reactor can be measured. For example, when the reactor water level drops, the differential pressure between the reference water column detection pipe 29 and the fluctuating water column detection pipes 30 and 31 increases. , 31 is reduced.

The broadband differential pressure detector 32 detects changes in the internal pressure of the reference water column detection pipe 29 and the internal pressure of the first fluctuating water column detection pipe 30 as differential pressure. This wideband differential pressure detector 32 can be used to monitor changes in the water level during normal operation and during transient changes.

Further, the fuel zone differential pressure detection unit 33 detects the change in the internal pressure of the reference water column detection pipe 29 and the internal pressure of the second fluctuating water column detection pipe 31 as a differential pressure. This fuel zone differential pressure detection unit 33 can be used for confirmation of core submergence at the time of a severe accident.

The differential pressure type water level gauge 5 converts the differential pressure detected by the broadband differential pressure detector 32 and the fuel zone differential pressure detector 33 into the reactor water level. A water level measurement value indicating the water level in the reactor is accumulated in the data accumulation unit 10 . It should be noted that a value indicating the differential pressure before being converted to the in-furnace water level may be accumulated in the data accumulation unit 10 as the water level measurement value.

As shown in FIG. 2, the thermocouple water level gauge 6 (differential thermocouple water level gauge) measures the temperature of the furnace inside the pressure vessel 22 by means of thermocouple elements 36 and 37 (FIG. 3) provided inside the pressure vessel 22. Measure the internal water level (air-liquid boundary).

The thermocouple water level gauge 6 has a plurality of sensor rods 34 provided between the fuel assemblies 20 in the core 21 . These sensor rods 34 are tubular components extending from the inside to the outside of the pressure vessel 22 . Since the sensor rod 34 is provided inside the pressure vessel 22, it is in contact with the reactor water 24 and can directly measure the water level in the reactor.

As shown in FIG. 3, one sensor rod 34 is internally provided with one heater 35 extending in its longitudinal direction. Further, the hot junction thermocouple element 36 and the cold junction thermocouple element 37 are provided in close proximity to each other.

In addition, a heat insulating portion 38 is provided to surround the thermocouple element 36 for the hot junction together with the heater 35 . A heat insulating portion 38 is not provided around the thermocouple element 37 for cold junction. A set of the sensor section 39 is composed of the heat insulating section 38, the thermocouple element 36 for the hot junction, and the thermocouple element 37 for the cold junction.

A specific measurement principle will be described. The heater 35 heats both the hot junction thermocouple element 36 and the cold junction thermocouple element 37 . Since the thermal conductivity is different between air and water, the manner in which the heat 40 of the heater 35 is released to the outside differs between when the sensor unit 39 is in the air and when it is in water.

For example, as shown in FIG. 3A, when the sensor unit 39 is in water, the amount of heat dissipation from the thermocouple element 36 for the hot junction and the amount of heat dissipation from the thermocouple element 37 for the cold junction The heat difference increases. As shown in FIG. 3B, when the sensor unit 39 is in the air, the amount of heat dissipation from the thermocouple element 36 for the hot junction and the amount of heat dissipation from the thermocouple element 37 for the cold junction The difference in heat dissipation becomes smaller. In this manner, the difference between the amount of heat released by the thermocouple element 36 for the hot junction and the amount of heat released by the thermocouple element 37 for the cold junction determines whether the sensor unit 39 is in the air or in the water. can be detected.

FIG. 4 shows a graph Q1 when the sensor section 39 is in water and a graph Q2 when the sensor section 39 is in the air. Thus, the output voltage of the output signal of the thermocouple water level gauge 6 differs depending on whether the sensor section 39 is in the air or in the water. For example, by setting a threshold value H in advance and determining whether or not the output voltage exceeds the threshold value H, it is possible to detect whether the sensor unit 39 is in the air or in the water.

The water level in the furnace can be measured by detecting whether each sensor unit 39 is in air or in water. A water level measurement value indicating the water level in the reactor is accumulated in the data accumulation unit 10 . It should be noted that the value indicating the output voltage of the output signal of each sensor section 39 may be accumulated in the data accumulation section 10 as the water level measurement value.

A plurality of sensor portions 39 are arranged along the longitudinal direction of the sensor rod 34 . For example, one sensor rod 34 is provided with nine sensor units 39 arranged vertically. These sensor units 39 also make it possible to measure at which height position between the core 21 and the bottom of the pressure vessel 22 the water surface of the reactor water 24 exists. It should be noted that at least it is sufficient to be able to measure at which height position from the effective fuel top (TAF) to the effective fuel bottom (BAF) in the core 21 the water surface of the reactor water 24 exists.

As shown in FIG. 2, the radiation type water level gauge 7 measures the reactor water level inside the pressure vessel 22 using gamma rays 41 (radiation rays) emitted from the pressure vessel 22 to the outside. The radiation type water level gauge 7 is provided inside the reactor containment vessel 23 (biological shielding section) and outside the pressure vessel 22 .

The radiation type water gauge 7 includes a plurality of dose detectors 42 arranged vertically along the pressure vessel 22 . These dose detectors 42 are arranged from the center to the bottom of the pressure vessel 22 . Each dose detection part 42 is provided, for example, every 1 m in the height direction. Each dose detector 42 detects, for example, gamma rays 41 emitted from the core 21 (fuel). The radioactive water level gauge 7 can measure the water level in the reactor by detecting the attenuation of the gamma rays 41 due to the shielding effect of water.

The intensity of the gamma rays 41 is measured by each dose detector 42, and the height at which the difference (Δγ) between the gamma ray intensities of the dose detectors 42 is maximum is determined as the water level in the reactor.

FIG. 5A is a graph showing the relationship between the radioactive water level gauge 7, the reactor core 21 and the pressure vessel 22. FIG. FIG. 5B is a graph showing the gamma ray intensity at each height measured by the radiation water level gauge 7. FIG. FIG. 5C is a graph showing the difference in gamma ray intensity (Δγ).

For example, it is assumed that the reactor water level (WL) exists between the effective fuel top (TAF) and the effective fuel bottom (BAF) in the core 21 (FIG. 5A). Here, the gamma ray intensity is attenuated in the portion where the reactor water 24 exists. On the other hand, the gamma ray intensity in the portion where the fuel is exposed is not attenuated (Fig. 5(B)).

In each dose detection unit 42, the portion where the difference (Δγ) from the adjacent dose detection unit 42 above or below is maximized is determined. This difference (Δγ) becomes maximum when comparing the dose detection unit 42 immersed in the reactor water 24 and the dose detection unit 42 not immersed in the reactor water 24 (FIG. 5(C)).

The portion where this difference (Δγ) is maximum can be measured as the reactor water level (WL). A water level measurement value indicating the water level in the reactor is accumulated in the data accumulation unit 10 . Note that the gamma ray intensity values detected by the dose detectors 42 or the gamma ray intensity difference (Δγ) values may be accumulated in the data accumulation unit 10 as water level measurement values.

In the radiation type water gauge 7 of this embodiment, the water surface (gas-liquid boundary) of the reactor water 24 exists at any height position between the core 21 and the bottom of the pressure vessel 22 based on the intensity of the gamma rays 41. can be measured. Moreover, the radiation type water level gauge 7 can measure the water level in the reactor without contact from the outside of the pressure vessel 22 . Therefore, even if the fuel is exposed or melted in the event of a severe accident, the reactor water level (gas-liquid boundary) can be continuously measured.

In addition, when the fuel melts and falls to the bottom of the pressure vessel 22, the radiation pattern of gamma rays also changes. For example, when the reactor water level drops and the core 21 is exposed from the reactor water 24, the intensity of the gamma rays 41 increases. Furthermore, when the fuel is melted and melts down into the reactor water 24 at the bottom of the pressure vessel 22, the fuel is immersed in the reactor water 24 again, so the gamma rays 41 are attenuated. Then the intensity will decrease. That is, the radiation type water level gauge 7 can grasp the cooling state and melting state of the fuel at the time of the severe accident. It is also possible to grasp the distribution of fuel at the time of a severe accident.

As shown in FIG. 2, the differential pressure type water level gauge 5 loses its measuring function because the water inside the reference water column detection pipe 29 evaporates during a severe accident. In a severe accident, the thermocouple water level gauge 6 can maintain the measurement of the reactor water level before the fuel is melted. will be lost. On the other hand, the radiation type water level gauge 7 is provided outside the pressure vessel 22 and can measure the water level in the reactor without contacting the pressure vessel 22 and the core 21. is never lost.

Analysis results obtained by simulating a severe accident of the nuclear reactor 19 by the accident analysis unit 11 will be described with reference to the graphs of FIGS. 8 to 12. FIG.

The accident analysis unit 11 performs a simulation when a severe accident such as a fuel melting accident occurs in a reactor type (for example, a boiling water reactor) in which water level gauges 5 to 7 are installed, and analyzes the water level in the reactor at the time of the severe accident, Analyze changes in furnace temperature, furnace pressure, etc.

FIG. 8 is graphs G1 and G2 showing reactor water levels during a severe accident evaluated by the accident analysis unit 11 . The horizontal axis indicates the elapsed time (in seconds), and the vertical axis indicates the reactor water level (in meters). A collapse water level and a two-layer water level are used as values of the in-core water level in this embodiment. Graph G1 indicates the collapse water level, and graph G2 indicates the double layer water level. Note that the collapse water level and the two-layer water level have different definitions, so the two graphs G1 and G2 do not necessarily match.

When a severe accident occurs, cooling of the fuel is delayed, so the water level in the reactor begins to drop rapidly from the start of the severe accident (origin). Fuel is exposed from the reactor water 24 due to the decrease in the reactor water level. At time F when the fuel melts and falls into the reactor water 24, the two-layer water level graph G2 changes greatly.

FIG. 9 is a graph G3 showing the reactor internal pressure at the time of the severe accident evaluated by the accident analysis unit 11. As shown in FIG. The horizontal axis indicates elapsed time (unit: seconds), and the vertical axis indicates furnace pressure (unit: megapascal).

The reactor pressure rises at the beginning of the severe accident (origin) and decreases as the accident progresses. Then, at time F when the fuel melts and drops into the reactor water 24, the graph G3 of the reactor internal pressure rises sharply.

FIG. 10 is graphs G4 and G5 showing the in-core temperature at the time of the severe accident. The horizontal axis indicates elapsed time (unit: seconds), and the vertical axis indicates furnace temperature (unit: °C). Graph G4 indicates the gas phase temperature, and graph G5 indicates the liquidus temperature.

The gas phase temperature graph G4 gradually rises from the start of the severe accident (origin) as the accident progresses. The liquidus temperature graph G5 gradually decreases as the reactor water 24 evaporates from the severe accident start point (origin). Then, at time F when the fuel melts and drops into the reactor water 24, the gas phase temperature graph G4 temporarily decreases, and the liquid phase temperature graph G5 temporarily increases.

As shown in FIG. 1, in the accident analysis unit 11, in advance, the reactor water level, reactor temperature, reactor pressure, and fuel content are determined in response to various accident scenarios assumed for the target nuclear reactor 19. Analyze the conditions inside the furnace, such as the melting state, and save the results. Furthermore, these in-core information are transmitted to the water level measurement value analysis unit 12 . Here, the accident analysis unit 11 creates an in-core situation data set and saves the in-core situation data set.

As shown in Fig. 6, the in-core situation data set contains information indicating the in-core water level, in-core temperature, in-core pressure, and fuel melting state in association with the elapsed time (progress) of the severe accident. be done. Note that other in-furnace information may be registered.

As shown in FIG. 1, the water level measurement value analysis unit 12 analyzes all types of water level gauges 5 to 7 installed in the target nuclear reactor 19 based on the in-core information analyzed by the accident analysis unit 11. Analyze the measured water level (indicated value).

11 are graphs G6 to G8 showing the water level measurement values of the water level gauges 5 to 7 analyzed by the water level measurement value analyzing unit 12. FIG. The horizontal axis indicates the elapsed time (in seconds), and the vertical axis indicates the measured water level (in meters). Graph G6 shows the water level measurement values of the differential pressure type water level gauge 5, graph G7 shows the water level measurement values of the thermocouple type water level gauge 6, and graph G8 shows the water level measurement values of the radiation type water level gauge 7.

Since the differential pressure type water level gauge 5 and the thermocouple type water level gauge 6 are damaged (failure) at time B, the respective water level measurement values indicate zero. In this embodiment, information indicating that the differential pressure type water level gauge 5 and the thermocouple type water level gauge 6 are damaged is also used to predict the reactor water level. In other words, even if the measured water level indicates zero, that value is used for multiple regression analysis.

The water level readings of each type of water gauge 5-7 have their own characteristics. For example, the water level measurement value of the differential pressure type water level gauge 5 indicates a value close to the collapsed water level (FIG. 8), and the thermocouple type water level gauge 6 indicates a value close to the two-phase water level (FIG. 8).

Further, since the differential pressure type water level gauge 5 and the thermocouple type water level gauge 6 are installed inside the pressure vessel 22, their operation is stopped when the fuel melts. On the other hand, the radiation type water level gauge 7 does not directly measure the reactor water level. The radiation water level gauge 7 estimates the water level in the reactor from the distribution of the radiation emitted from the fuel, which is the modulation of the radiation intensity caused by the difference in the water density in the height direction of the pressure vessel 22 .

In the case of the radiation type water level gauge 7, the water level can be measured with a certain degree of accuracy in the vicinity of the core 21. less accurate. However, since the radiation type water level gauge 7 is installed outside the pressure vessel 22, it operates without losing its function even when the fuel melts. The graph G8 of the radioactive water level gauge 7 also changes at the time F when the fuel melts and falls into the reactor water 24, and it can be seen that the water level measurement value can be obtained according to the conditions inside the reactor.

The water level measurement value analysis unit 12 creates a measurement value data set in which the reactor water level output from the accident analysis unit 11 and the water level measurement value analyzed by the water level measurement value analysis unit 12 are associated, and the prediction model creation unit 13 Send to

As shown in FIG. 7, the measured value data set contains the water level measured value of the differential pressure type water level gauge 5 (first class water level gauge) and the thermocouple type water level in association with the elapsed time (progress) of the severe accident. Information indicating the water level measurement value of the total 6 (second-class water level gauge) and the water level measurement value of the radiation type water level gauge 7 (third-class water level gauge) is registered.

As shown in FIG. 1, the prediction model creation unit 13 performs multiple regression analysis using the measured value data set created by the measured water level analysis unit 12 as the target variable for the reactor water level and the measured water level value as the explanatory variable. By doing so, a regression formula for predicting the reactor water level from the measured water level is created. A water level prediction model using this regression equation is transmitted to the in-core water level prediction unit 14 .

The in-core water level prediction unit 14 receives the water level measurement values accumulated in the data accumulation unit 10 and the prediction model based on the multiple regression analysis created by the prediction model creation unit 13 . By applying this prediction model to the measured water level, the predicted value of the water level in the reactor 19 is derived.

FIG. 12 is a graph P showing predicted values of the in-core water level obtained by the in-core water level prediction unit 14, and a graph G2 showing the true value of the in-core water level. In addition, the true value exemplifies the double layer water level.

Graph P of the predicted value of the reactor water level includes the effects of the differential pressure type water level gauge 5 and the thermocouple type water level gauge 6, which are effective in normal times, and the radiation type water level gauge 7, which is effective in the event of a severe accident. Therefore, it is possible to measure the reactor water level without loss of function from normal times to severe accidents.

As shown in FIG. 1, the in-core information output unit 15 outputs in-core information including the predicted value of the in-core water level evaluated by the in-core water level prediction unit 14 and the water level measurement value measured by the water level measurement unit 2. Controls display on the display, etc. This display is installed in the central control room of a nuclear power plant, etc., and is consistently monitored by operators from normal operation to severe accidents.

Next, the reactor water level measuring method will be described with reference to the flow chart of FIG. Note that the block diagram shown in FIG. 1 will be referred to as appropriate.

As shown in FIG. 13, first, the user of the water level prediction computer 4 performs preliminary processing. This pre-processing creates a predictive model. In step S<b>11 , various data regarding the nuclear reactor 19 are input to the accident analysis unit 11 .

In the next step S<b>12 , the accident analysis unit 11 simulates a severe accident of the nuclear reactor 19 .

In the next step S13, the accident analysis unit 11 analyzes the severe accident of the nuclear reactor 19 and analyzes the water level in the reactor corresponding to the progress of the severe accident.

In the next step S14, based on the analysis of the severe accident performed by the accident analysis unit 11, the water level measurement value analysis unit 12 analyzes the progress of the severe accident using a plurality of types of water level gauges 5 to 7 each having a different measurement principle. Analyze the water level measurements taken accordingly.

In the next step S15, the prediction model creation unit 13 analyzes the relationship between the reactor water level analyzed by the accident analysis unit and the water level measurement value analyzed by the water level measurement value analysis unit 12, and determines the reactor water level from the water level measurement value. Create a prediction model that will be used to predict the

Here, the preliminary processing using the water level prediction computer 4 ends. Next, the water level prediction computer 4 is used to start actual in-reactor water level prediction processing of the reactor 19 .

In the next step S16, the data accumulation unit 10 accumulates water level measurement values actually measured in the nuclear reactor 19 by a plurality of types of water level gauges 5-7. Note that the data accumulation unit 10 may accumulate the in-core temperature measurement values actually measured in the reactor by the in-core thermometer 8 .

In the next step S17, the in-core water level prediction unit 14 calculates the actual in-core water level of the reactor 19 based on each water level measurement value accumulated in the data accumulation unit and the prediction model created by the prediction model creation unit. to predict.

In the next step S<b>18 , the in-core information output unit 15 outputs in-core information regarding the in-core water level predicted by the in-core water level prediction unit 14 . Then, the in-core water level prediction processing ends.

By repeating the in-core water level prediction processing from step S16 to step S18, the in-core water level can be consistently measured from the time of normal operation to the time of a severe accident, and the operator can grasp the in-core water level. can.

In addition, in the flowchart of the present embodiment, each step is exemplified in a form in which each step is executed in series. good. Also, some steps may be executed in parallel with other steps.

The water level prediction computer 4 of the present embodiment includes a dedicated chip, FPGA (Field Programmable Gate Array), GPU (Graphics Processing Unit), or CPU (Central Processing Unit). Storage devices such as (Read Only Memory) or RAM (Random Access Memory), external storage devices such as HDD (Hard Disk Drive) or SSD (Solid State Drive), display devices such as displays, and mouse or keyboard An input device and a communication interface are provided. This water level prediction computer 4 can be realized with a hardware configuration using a normal computer.

In addition, the program executed by the water level prediction computer 4 of this embodiment is pre-installed in a ROM or the like and provided. Alternatively, this program can be stored as an installable or executable file on a non-transitory computer-readable storage medium such as CD-ROM, CD-R, memory card, DVD, flexible disk (FD), etc. may be stored and provided.

Also, the program executed by the water level prediction computer 4 may be stored on a computer connected to a network such as the Internet, downloaded via the network and provided. The water level prediction computer 4 can also be configured by connecting and combining separate modules that independently perform the respective functions of the constituent elements by connecting them via a network or a dedicated line.

In this embodiment, each of the water level gauges 5 to 7 having different measurement principles is provided, but other modes are also possible. For example, two or more water gauges having the same type of measurement principle may be provided. For example, two or more radiation type water gauges 7 may be provided.

In this embodiment, the in-core situation data set (FIG. 6) and the measured value data set (FIG. 7) are created as separate data sets, but other modes may be used. For example, the in-core situation data set and the measured value data set may be created as one data set.

According to the embodiment described above, the relationship between the reactor water level analyzed by the accident analysis unit and the water level measurement value analyzed by the water level measurement value analysis unit is analyzed, and the reactor water level is predicted from a plurality of water level measurement values. By providing a predictive model preparation unit that prepares a predictive model used for this purpose, it is possible to consistently measure the reactor water level from the time of normal operation to the time of a severe accident.

While several embodiments of the invention have been described, these embodiments have been presented by way of example and are not intended to limit the scope of the invention. These embodiments can be implemented in various other forms, and various omissions, replacements, changes, and combinations can be made without departing from the scope of the invention. These embodiments and their modifications are included in the scope and spirit of the invention, as well as the scope of the invention described in the claims and equivalents thereof.

DESCRIPTION OF SYMBOLS 1... Reactor water level measurement system, 2... Water level measurement unit, 3... In-reactor measurement unit, 4... Water level prediction computer, 5... Differential pressure type water level gauge, 6... Thermocouple type water level gauge, 7... Radiation type water level gauge, 8 Furnace thermometer 9 Furnace pressure gauge 10 Data accumulation unit 11 Accident analysis unit 12 Water level measurement value analysis unit 13 Prediction model creation unit 14 Furnace water level prediction unit 15 In-core information output unit 16 Information input unit 17 Storage unit 18 Main control unit 19 Reactor 20 Fuel assembly 21 Core 22 Pressure vessel 23 Reactor containment vessel 24... Reactor water, 25... Steam separator, 26... Steam dryer, 27... Recirculation pump, 28... Condensing tank, 29... Reference water column detection pipe, 30... First fluctuation water column detection pipe, 31... Second fluctuation water column detection pipe, 32 broadband differential pressure detector, 33 fuel zone differential pressure detector, 34 sensor rod, 35 heater, 36 thermocouple element for hot junction, 37 thermocouple element for cold junction, 38... Thermal insulation part, 39... Sensor part, 40... Heat, 41... Gamma ray, 42... Dose detection part.

Claims (10)

an accident analysis unit that analyzes a severe accident of a nuclear reactor and analyzes the water level in the reactor corresponding to the progress of the severe accident;
a water level measurement value analysis unit that analyzes the water level measurement values measured according to the development state by a plurality of types of water level gauges having different measurement principles, based on the analysis of the severe accident;
The relationship between the in-core water level analyzed by the accident analysis unit and the water level measurement value analyzed by the water level measurement value analysis unit is analyzed, and the in-core water level is predicted from a plurality of water level measurement values. a prediction model creation unit that creates a prediction model that is
a data accumulation unit for accumulating the water level measurement values actually measured in the nuclear reactor by a plurality of types of the water level gauges;
an in-reactor water level prediction unit that predicts the actual in-reactor water level of the nuclear reactor based on the respective water level measurement values accumulated in the data accumulation unit and the prediction model created in the prediction model creation unit; ,
an in-core information output unit that outputs in-core information about the in-core water level predicted by the in-core water level prediction unit;
comprising a
Reactor water level measurement system. The prediction model is used to predict the reactor water level from a plurality of the water level measurements by performing multiple regression analysis.
The reactor water level measurement system according to claim 1. The plurality of types of water level gauges are
a differential pressure type water level gauge for measuring the water level inside the pressure vessel from the pressure difference between the top and bottom of the pressure vessel of the nuclear reactor;
a thermocouple water level gauge for measuring the water level inside the pressure vessel with a thermocouple element provided in the pressure vessel;
a radiological water level gauge that measures the water level inside the pressure vessel by means of radiation emitted from the pressure vessel to the outside;
including at least two of
The reactor water level measurement system according to claim 1 or 2. The data accumulation unit accumulates the time when the water level measurement value is measured in association with each water level measurement value,
The reactor water level measurement system according to any one of claims 1 to 3. The accident analysis unit
Simulating the severe accident,
Analyzing changes over time corresponding to the progress of each parameter indicating at least one of the reactor temperature, the reactor pressure, the molten state of the fuel, and the reactor water level,
Creating a data set of conditions inside the reactor in which the elapsed time from the occurrence of the severe accident is associated with the parameter;
The reactor water level measurement system according to any one of claims 1 to 4. The water level measurement value analysis unit
analyzing the water level measurement values measured by a plurality of types of water level gauges based on the in-core situation data set created by the accident analysis unit;
creating a measurement data set that associates the elapsed time with the water level measurement;
The reactor water level measurement system according to claim 5. The prediction model creation unit
Based on the in-core situation data set and the measured value data set,
Creating the prediction model to be used for predicting the in-core water level from a plurality of the water level measured values by performing multiple regression analysis with the in-core water level as the objective variable and the water level measured value as the explanatory variable;
The reactor water level measurement system according to claim 6. The prediction model creation unit creates the prediction model used for predicting the reactor water level by adding values other than the water level measurement value as the explanatory variables.
The reactor water level measurement system according to claim 7. The prediction model creation unit analyzes the relationship between the water level measurement value measured by the water level gauge after damage due to the severe accident and the reactor water level, and creates the prediction model.
The reactor water level measurement system according to any one of claims 1 to 8. a step in which the accident analysis unit analyzes the severe accident of the nuclear reactor and analyzes the water level in the reactor corresponding to the progress of the severe accident;
a step in which the water level measurement value analysis unit analyzes the water level measurement values measured according to the progress by a plurality of types of water level gauges each having a different measurement principle, based on the analysis of the severe accident;
A prediction model creation unit analyzes the relationship between the reactor water level analyzed by the accident analysis unit and the water level measurement value analyzed by the water level measurement value analysis unit, and predicts the reactor water level from the water level measurement value. creating a predictive model to be used in
a step in which the data accumulation unit accumulates the water level measurement values actually measured in the nuclear reactor by a plurality of types of the water level gauges;
An in-core water level prediction unit predicts the actual in-reactor water level of the reactor based on the water level measurement values accumulated in the data accumulation unit and the prediction model created by the prediction model creation unit. and
an in-core information output unit outputting in-core information about the in-core water level predicted by the in-core water level prediction unit;
including,
Reactor water level measurement method.

Images