JP6869747B2 - Reactor risk management equipment, reactor risk management methods, and reactor risk management programs - Google Patents

Description

The present invention provides a nuclear reactor risk management device, a nuclear reactor risk management method, which can promptly predict an accident situation in the event of an emergency such as a large-scale earthquake in a facility having a nuclear reactor such as a nuclear power plant. And the reactor risk management program.

In a nuclear power plant, radioactive materials are fissioned in a nuclear reactor, and the heat energy generated at that time is used to rotate a steam turbine to generate electricity. Since this fission reaction involves the generation of enormous heat energy, it is necessary to continuously cool the reactor by constantly supplying water to control the temperature and pressure of the reactor and containment vessel. However, if the heat removal function is lost due to a large-scale earthquake such as the Great East Japan Earthquake that occurred in recent years, it is expected that the pressure in the containment vessel will rise. Then, when the pressure inside the reactor containment vessel is likely to exceed the permissible value, the containment vessel vent (an emergency measure to reduce the pressure by discharging a part of the gas containing radioactive substances to the outside) is carried out to reduce the pressure. However, it is necessary to estimate the time of the containment vessel vent, predict how much radioactive material will be released at that time, and transmit information at an appropriate timing.

For example, in Patent Document 1, transmission data collected from each part of a nuclear facility, non-detection values not measured in a plant are estimated using the transmission data and a predetermined calculation model, and initial settings are made based on the transmission data and the non-detection values. A nuclear emergency calculation system and a program that perform simulations while tracking plant conditions in real time using the analysis code are disclosed.

JP-A-2007-10375

However, in the nuclear emergency calculation system and program of Patent Document 1, after updating the MAAP (Modular Accident Analysis Program) code initialization data based on the transmission data collected from the nuclear facility, precise calculation by the MAAP code is performed. I do. The MAAP code is a simultaneous and consistent analysis of the release and migration behavior of hydrothermal power / fission products (FP) in the reactor pressure vessel, reactor containment vessel, and reactor building during a severe accident in a light water reactor. It is an analysis program that can be used. Then, in order to obtain the reactor state prediction data using the MAAP code, it is necessary to set a large number of parameters. In addition, since calculation time is required to predict the progress of the situation, it is assumed that it will be difficult to predict the accident situation promptly in the event of an emergency.

An object of the present invention is to solve such a problem, and an object of the present invention is a nuclear reactor risk management device capable of promptly predicting an accident situation in the event of an emergency, and a nuclear reactor risk. To provide management methods and reactor risk management programs.

The present invention
Scrum occurrence time and
Accident scenario and
Reactor water injection stop time and
The water level when the reactor water injection is stopped and
Reactor pressure vessel rapid decompression time and
Reactor pressure vessel rapid pressure reducing valve number and
An input section for inputting input parameters including the presence or absence of pedestal water injection,
A storage unit that stores a plurality of state prediction source data calculated based on pre-calculation parameters that are larger than the input parameters.
Based on the input parameters, the state prediction source data matching the accident phase is selected from the plurality of state prediction source data based on the pre-calculation parameters, and the reactor water level, based on the selected state prediction source data, It is equipped with a control unit that predicts at least one of the timing of reactor pressure vessel damage, the reactor containment vessel pressure, the environmental release time of radioactive materials, and the amount of radioactive materials released to the environment .
The reactor pressure vessel rapid pressure reducing valve number is a reactor risk management device which is the number of pressure reducing valves opened according to the accident phase out of the total number of pressure reducing valves installed in the plant.

Further, it is preferable that the control unit corrects the prediction data based on the timing of the reactor pressure vessel breakage and at least one actually measured value of the reactor containment vessel pressure.

Further, it is preferable that the control unit predicts the time until the reactor containment vessel pressure reaches a predetermined value based on the selected state prediction source data.

Further, the input parameters include the presence / absence of reactor water injection after the reactor pressure vessel is damaged, the presence / absence of spray to the reactor containment vessel, the environmental release path of radioactive materials, the presence / absence of filter vent, and the presence / absence of pH control. seen at least 1 Tsuo含of out,
The pH control is preferably to supply an alkaline substance into the reactor containment vessel.

In addition, the control unit considers that after the injection of cooling water into the reactor is stopped, the water injection is continuously stopped, such as the reactor water level, the timing of reactor pressure vessel damage, and the reactor containment vessel pressure. It is preferable to predict the time of environmental release of radioactive material and at least one of the amount of radioactive material released into the environment.

Further, it is preferable that the control unit predicts the amount of radioactive material released to the environment in consideration of pedestal water injection, reactor water injection, and dry well spray after the injection of cooling water into the reactor is stopped.

In addition, the present invention
Depending on the input section
Scrum occurrence time and
Accident scenario and
Reactor water injection stop time and
The water level when the reactor water injection is stopped and
Reactor pressure vessel rapid decompression time and
Reactor pressure vessel rapid pressure reducing valve number and
Steps to enter input parameters, including with or without pedestal water injection,
A step of storing a plurality of state prediction source data calculated based on pre-calculation parameters that are larger than the input parameters by the storage unit, and
A step of selecting a state prediction source data matching the accident phase from a plurality of state prediction source data based on the pre-calculation parameter based on the input parameter by the control unit.
Of the reactor water level, reactor pressure vessel damage timing, reactor containment vessel pressure, radioactive material environmental release time, and radioactive material environmental release amount, based on the state prediction source data selected by the control unit. look including a step of predicting at least one of the,
The reactor pressure vessel rapid pressure reducing valve number is a reactor risk management method which is the number of pressure reducing valves opened according to the accident phase out of the total number of pressure reducing valves installed in the plant.

In addition, the present invention
On the computer
Scrum occurrence time and
Accident scenario and
Reactor water injection stop time and
The water level when the reactor water injection is stopped and
Reactor pressure vessel rapid decompression time and
Reactor pressure vessel rapid pressure reducing valve number and
Procedures for entering input parameters, including with or without pedestal water injection,
A procedure for storing a plurality of state prediction source data calculated based on pre-calculation parameters that are larger than the input parameters, and a procedure for storing the plurality of state prediction source data.
A procedure for selecting a state prediction source data that matches the accident phase from a plurality of state prediction source data based on the pre-calculation parameters based on the input parameters, and
Based on the selected state predictor data, at least one of the reactor water level, the timing of reactor pressure vessel breakage, the reactor containment pressure, the time of environmental release of radioactive material, and the amount of radioactive material released to the environment. to execute the steps of predicting,
The reactor pressure vessel rapid pressure reducing valve number is a reactor risk management program which is the number of pressure reducing valves opened according to the accident phase out of the total number of pressure reducing valves installed in the plant.

According to the present invention, it is possible to provide a nuclear reactor risk management device, a nuclear reactor risk management method, and a nuclear reactor risk management program capable of promptly predicting an accident situation in the event of an emergency.

It is a block diagram which shows an example (boiling water type light water furnace) of the structure of the reactor system which is the prediction target by the reactor risk management apparatus which concerns on 1st Embodiment of this invention. It is a block diagram which shows the structure of the reactor risk management apparatus which concerns on 1st Embodiment of this invention. It is a figure which shows an example of the input screen of an input parameter. (A) is an example of the state prediction source data (water level in the core) based on the pre-calculation parameter, and (b) is an example of the state prediction source data (hydrogen generation amount) based on the pre-calculation parameter. (A) to (d) are examples of state prediction source data (PCV pressure) based on pre-calculation parameters for each accident phase. It is a figure which shows the PCV pressure prediction data from the start of water injection to after RPV breakage, which was generated by selecting the state prediction source data which matched each accident phase from the state prediction source data based on a plurality of pre-calculation parameters. It is a figure which shows an example of the output screen (prediction result list) of the prediction result by the reactor risk management apparatus which concerns on 1st Embodiment of this invention. (A) to (c) are diagrams showing the comparison result between the prediction by the reactor risk management device according to the first embodiment of the present invention and the prediction using the MAAP code. (A) and (b) are diagrams showing an example of an output screen (by prediction result) of a prediction result by the reactor risk management device according to the first embodiment of the present invention. It is a flowchart which shows the operation flow of the reactor risk management apparatus which concerns on 1st Embodiment of this invention. It is a figure which shows the state which the state prediction source data is approximated by the number of data which is different for each accident phase. It is a block diagram which shows an example (pressurized water type light water reactor) of the structure of the nuclear reactor system which is the prediction target by the nuclear reactor risk management apparatus which concerns on 2nd Embodiment of this invention.

Hereinafter, the present invention will be described in more detail with reference to the drawings.

(First Embodiment)
FIG. 1 is a block diagram showing an example of the configuration of the reactor system 100, which is the prediction target by the reactor risk management device 200 according to the first embodiment of the present invention. In the present embodiment, the reactor system 100 will be described by taking a boiling water reactor (BWR) as an example, but an advanced boiling water reactor (ABWR) or a pressurized water reactor (ABWR) or a pressurized water reactor (ABWR) It may be another type of reactor such as PWR: Pressurized Water Reactor). A reactor pressure vessel 1 provided with fuel and a core 3 which are radioactive substances, a reactor containment vessel 20 which surrounds the reactor pressure vessel 1 and suppresses the release of radioactive substances to the outside when a coolant is lost, and an atom. It includes a reactor building 40 that surrounds the reactor containment vessel 20. Further, FIG. 1 shows a water supply pump 5 that supplies light water as a coolant into the reactor pressure vessel 1, and a filter vent portion for filtering and venting radioactive substances in the reactor containment vessel 20. 30 is also shown.

The reactor pressure vessel 1 (RPV: Reactor Pressure Vessel) includes enriched uranium as a fuel and a core 3 inside. The reactor pressure vessel 1 is a steel structure and has a thickness so that radiation from a radioactive substance does not leak to the outside. Further, as shown in FIG. 1, the reactor pressure vessel 1 is connected to a water supply passage 5a and a steam supply passage 7a in order to circulate light water, which is a coolant supplied from the water supply pump 5, into the reactor pressure vessel 1. .. The cooling material supplied via the water supply channel 5a receives the heat of the core 3 and becomes high-temperature steam, and then is sent to a steam turbine (not shown) via the steam supply path 7a and the pressure regulating valve 7 to steam. Power is generated by rotating a turbine.

The reactor containment vessel 20 (PCV: Primary Containment Vessel) has a role as a barrier so that the radioactive material leaked from the reactor pressure vessel 1 does not leak to the outside when the coolant is lost or the like. As shown in FIG. 1, the reactor containment vessel 20 includes the above-mentioned reactor pressure vessel 1 and suppression pools 22 arranged at one location on each side of the bottom of the reactor containment vessel 20. Further, a relief flow path 9a that branches from the above-mentioned steam supply path 7a and guides the coolant to the suppression pool 22 via the main steam relief valve 9 is provided.

The suppression pool 22 has water inside, and has a function of suppressing the pressure rise by condensing the steam released from the reactor pressure vessel 1 via the flow path 9a with water. Further, the suppression pool 22 suppresses the pressure rise in the reactor containment vessel 20 by similarly condensing steam containing radioactive substances leaking from the reactor pressure vessel 1 and spreading in the reactor containment vessel 20 with water. doing. The suppression pool 22 not only suppresses the pressure of the introduced steam, but also plays a role of removing radioactive substances contained in the steam. That is, in the present embodiment, the radioactive substances contained in the steam can be removed by passing the steam in the reactor containment vessel 20 through the water in the suppression pool 22.

The steam whose pressure has been reduced by passing it through water in the suppression pool 22 is introduced into the filter vent portion 30 via the discharge path 22a, as shown in FIG. The steam passed through the water in the filter vent portion 30 is discharged to the outside from the open air passage 30a after the radioactive substances are further removed. It can also be discharged directly to the outside from the open air passage 22b branching from the discharge passage 22a.

Next, the reactor risk management device 200 will be described. FIG. 2 is a block diagram showing the configuration of the reactor risk management device 200 according to the first embodiment of the present invention. The reactor risk management device 200 stores the input unit 210 for inputting input parameters by the user and the state prediction source data calculated by using the MAAP (Modular Accident Analysis Program) code based on the pre-calculation parameters described later. Storage unit 250, and control unit 230 that predicts the reactor storage vessel pressure, etc. by selecting the state prediction source data that matches the accident phase from the state prediction source data in the storage unit 250 based on the input parameters and the like. It is provided with an output unit 290 that outputs a prediction result.

The operation by the control unit 230 can be configured by, for example, executing a program by a CPU (Central Processing Unit) mounted on a microcomputer or a PC (Personal Computer). Further, the operation by the control unit 230 may be realized by hardware.

The storage unit 250 can be realized, for example, by controlling an IC (Integrated Circuit) memory or the like provided in the microcomputer constituting the control unit 230 by a CPU (Central Processing Unit).

Further, the storage unit 250 may be composed of, for example, a recording / reproducing device connected to a PC and a storage medium capable of recording / reproducing the recording / reproducing device. Examples of the storage medium include IC memory, HDD (registered trademark) (Hard Disc Drive), CD (Compact Disc), DVD (registered trademark) (Digital Versatile Disc), BD (Blu-ray (registered trademark) Disc), and the like. Can be used. In this case, the storage unit 250 can be configured to be controlled by a CPU or the like mounted on the PC.

The user inputs input parameters from the input unit 210 in order to predict the state inside the reactor by the reactor risk management device 200 of the present embodiment. The input of this input parameter can be input by, for example, the user from the input GUI (see FIG. 3) displayed on the display by a keyboard or the like. Further, the control unit 230 may be configured to read the input parameter from the storage unit 250. Further, the control unit 230 of the reactor risk management device 200 is configured to communicate with the reactor control unit (not shown) of the reactor system 100 via the communication unit 270 and acquire input parameters from the reactor system 100. You may. The reactor risk management device 200 may be incorporated in the reactor system 100.

In this embodiment, 12 input parameters are used as shown in Table 1. The outline of each input parameter is as shown in Table 1.

Hereinafter, each input parameter will be described.

The scrum time is the time when the fission reaction is in an emergency stop state by inserting a scrum, that is, an absorption rod that absorbs neutrons into the gap between the fuel rods. From this point in time, in the core 3, the decay heat generated by the decay of fission products and the chemical reaction heat may be considered as heat sources, and this scrum time is an input parameter for starting the decay heat evaluation start.

An accident scenario is an input parameter that selects whether or not a Loss of Coolant Accident (LOCA) has occurred. The loss-of-coolant accident is an accident in which the piping of the reactor is damaged and the coolant flows out. Since the pressure rise behavior of the reactor containment vessel differs greatly depending on the presence or absence of a coolant loss accident, it is used as an input parameter.

The reactor water injection stop time and water source are the time when the supply of the cooling material to the reactor pressure vessel 1 is stopped and the water source of the cooling material. Due to the suspension of coolant supply, the water level of the coolant in the reactor pressure vessel drops, which causes the core to be exposed. Therefore, by using the time when the supply of the coolant is stopped as an input parameter, it is possible to predict the decrease in the water level of the coolant in the reactor pressure vessel. The water source for water injection into the reactor is assumed to be the suppression pool 22 inside the reactor containment vessel 20 and the external water source outside the reactor containment vessel 20, but the PCV pressure depends on which water source is used for the following reasons. It is used as an input parameter because it affects the factors such as. That is, when the water source in the reactor containment vessel 20 is used, the rate of temperature rise of the suppression pool 22 becomes faster than that in the case of the external water source due to the decay heat. When the temperature of the suppression pool 22 rises, the vapor pressure rises and the PCV pressure also rises. In addition, the temperature of the water injected into the reactor pressure vessel 1 also increases. On the other hand, when an external water source is used, the water injection temperature does not rise with the passage of time, and the temperature rise rate of the suppression pool 22 slows down, while the suppression pool water level rises. The water level when the reactor water injection is stopped is used as an input parameter because it affects the time margin from the reactor water injection stop time to the core exposure.

The RPV rapid decompression time is the time when the steam in the reactor pressure vessel 1 is released into the reactor storage vessel 20 when the supply of the cooling material to the reactor is stopped. Since the emergency depressurization of the reactor pressure vessel 1 greatly affects the evaluation of the water level drop in the reactor pressure vessel 1, it is used as an input parameter. The rapid depressurization of RPV means that the steam in the reactor pressure vessel 1 is released into the suppression pool 22 via the steam supply path 7a, the main steam relief valve 9, and the relief flow path 9a in FIG.

The number of RPV rapid pressure reducing valves is the number of valves (not shown in FIG. 1) that are opened when the steam in the reactor pressure vessel 1 is released into the reactor containment vessel 20. Since the rate of decrease in the pressure inside the reactor pressure vessel 1 changes depending on the number of valves opened during this decompression, it is used as an input parameter. The input options: 1 to 8 valves shown in Table 1 are examples in the case where the reactor system 100 is an advanced boiling water reactor (ABWR).

The pedestal water injection is to supply an additional coolant in the pedestal to cool the core 3 that has fallen to the lower part of the reactor containment vessel 20. The pedestal refers to the space below the reactor pressure vessel in the reactor containment vessel. The pedestal is the region directly below the reactor pressure vessel 1 in FIG. 1, and the presence or absence of this pedestal water injection has a great influence on the pressure rise in the reactor containment vessel 20 when the reactor pressure vessel 1 is damaged. It is said.

Reactor water injection after RPV damage is the presence or absence of coolant supply into the reactor pressure vessel 1 after the reactor pressure vessel 1 is damaged. After the reactor pressure vessel 1 is damaged, both the fuel and the coolant are leaking into the reactor containment vessel 20, but whether or not to supply the additional coolant into the reactor pressure vessel 1 Since it has a large effect on the source term (type, amount, physical / chemical form, etc. of radioactive material that may be released to the outside of the reactor system 100), it is used as an input parameter.

The spray to the reactor containment vessel 20 reduces the radioactive material released into the reactor containment vessel 20 in the event of an accident, and cools the reactor containment vessel 20 in the D / W in order to suppress the temperature and pressure rise. It is carried out by spraying the material in a mist form. This spray also has a large effect on the source term, so it is used as an input parameter.

The environmental release route selects whether the radioactive material is released to the outside from the dry well (D / W) or the wet well (W / W) of the reactor containment vessel 20. Here, the dry well (D / W) refers to the upper portion where steam is dominant in the reactor containment vessel 20, and the wet well (W / W) is where water is dominant in the reactor containment vessel 20. Refers to the lower part that becomes. The W / W vent line shown in Table 1 is a vent route via the wet well, and in the present embodiment, radioactive substances are discharged from the discharge path 22a via the suppression pool 22 as shown in FIG. Refers to the path of release. On the other hand, the D / W vent line shown in Table 1 is a vent route (not shown) via a dry well, and is used as a vent route when, for example, some trouble occurs in the W / W vent line. Will be done. Since the release via the wet well (W / W) has a greater effect on the source term than the release via the dry well (D / W) because the radioactive substances are removed by water and the amount of release is small. It is an input parameter.

Regarding the availability of the filter vent, as shown in FIG. 1, the radioactive material in the steam discharged from the wet well (W / W) of the reactor containment vessel 20 via the discharge path 22a is the filter vent portion. Since it is further removed in No. 30 and the amount released is small, it has a great influence on the source term, so it is used as an input parameter.

The pH control is used as an input parameter because the volatility of iodine is suppressed by supplying an alkaline substance into the reactor containment vessel 20 and the source term is greatly affected.

Since the above 12 input parameters have a great influence on the main prediction data such as the water level in the RPV, the amount of hydrogen generated, the PCV pressure, the PCV gas phase temperature, and the source term evaluation, in the present embodiment, when an emergency occurs. Prediction data is obtained by inputting only these input parameters. An example of the input screen of these 12 input parameters is shown in FIG.

In the example of FIG. 3, individual input parameters can be set for each of the plants from Unit 1 to Unit 7. The "S / P water source" in the input parameter "reactor water injection stop time" refers to a form in which water from the suppression pool 22 (Suppression Pool) is supplied into the reactor pressure vessel 1 as a coolant, and is an "external water source". Refers to a form in which the coolant is supplied into the reactor pressure vessel 1 from the outside.

Further, in FIG. 3, the user inputs the water level when the reactor water injection is stopped.

Further, in FIG. 3, the "RPV rapid water reactor number" is a maximum of 8 in the case of ABWR as an example, and a number from 1 to 8 is input according to the number of valves.

The user does not necessarily have to input all the above 12 input parameters. For example, when the user wants to predict the time until the pressure in the reactor containment vessel 20 reaches the limit pressure, among the input parameters, "scram time", "accident scenario", "reactor water injection stop time and water source" , "Water level when reactor water injection is stopped", "RPV rapid decompression time", "Reactor pressure vessel rapid decompression valve number", and "Presence or absence of pedestal water injection" have a great influence on the predicted data. On the other hand, although the other input parameters affect the amount of radioactive material released, they have little effect on the predicted pressure data in the reactor containment vessel 20. Therefore, the user may input only the above seven input parameters. In that case, predetermined values are used for the remaining five input parameters.

In addition, it is conceivable that the water injection will be restored after the reactor water injection is stopped. However, since the purpose of this embodiment is to promptly transmit evacuation information, the water injection is continuously stopped without considering such restoration. Make a prediction as if you were doing it. As a result, the calculation model is simplified, and the prediction result can be obtained quickly. However, in the evaluation of the amount of radioactive material released to the environment at the same time as venting, the presence or absence of pedestal water injection, reactor water injection, and D / W spray after the reactor pressure vessel is damaged has an effect of orders of magnitude. It may be configured to take into account the parameters of.

When the user completes the input of the input parameter, the control unit 230 reads out the state prediction source data stored in the storage unit 250 based on the pre-calculation parameter. The pre-calculation parameters are parameters required for MAAP code calculation, which are larger in number than the above-mentioned input parameters. Although not selected as input parameters, the parameters required for MAAP code calculation include, for example, reactor water injection water level, reactor water injection start time, water injection pump characteristics, reactor water injection destination, coolant water source temperature, etc. There are initial gas phase components and temperature in the PCV, D / W spray flow rate, and the like. In this embodiment, the total number of pre-calculation parameters required to be input by the user in order to obtain the state prediction source data using the MAAP code is about 300, while the number of input parameters is the pre-calculation parameters. It is sufficiently small to be less than 1/15 of the number. For example, under the assumption that the reactor is submerged during the duration of water injection, the water level at the time of water injection in the reactor has a sufficiently small effect on the event progress, and the effect on the forecast data as a representative value is also small. .. Further, regarding the reactor water injection start time and the water injection pump characteristics, it is not necessary to input the pump characteristics because the core damage does not occur during the water injection period under the above assumption. However, it should be noted that in reality, depending on the water injection situation, the core may be exposed. Furthermore, the water injection destination inside the reactor (inside the shroud / outside the shroud, etc.) is not specified because it does not significantly affect the possibility of core flooding except for LOCA under the above assumption. It should be noted that the influence of the water injection point cannot be ignored in the gap between whether or not the core damage can be prevented. In addition, the coolant water source temperature has a small effect on the behavior of the reactor water level after water injection is stopped, and the effect on the PCV pressure is not large. Therefore, by using a slightly strict representative value, the water source in the individual evaluation It is possible to omit the temperature input. Further, the initial gas phase component and temperature in the PCV have little influence on the reactor water level behavior and the PCV pressure behavior as typical values. Furthermore, regarding the D / W spray flow rate, under the assumption that "a flow rate that fully exerts the source term reduction effect can be expected", even if the value described in the accident response procedure manual etc. is used as a representative value, it can be used as the prediction data. The impact is small. For this reason, in the example of this embodiment, the above parameters are not used as input parameters but as parameters for pre-calculation.

FIGS. 4A and 4B show specific examples of the state prediction source data calculated by using the MAAP code based on the parameters for pre-calculation. Although the state prediction source data calculated by using the MAAP code is shown as a graph in FIGS. 4A and 4B, it is stored as numerical data in the storage unit 250. In FIG. 4A, the horizontal axis is the time [h] and the vertical axis is the water level [m] in the core (RPV). It is a plot. Although the behavior of the water level decrease in the RPV changes depending on the water injection period, the tendency of the change is explained by the decrease behavior of the decay heat and shows a continuous change tendency. Therefore, it is possible to estimate the water level in the core by interpolating the data of the water injection period of 12 hours, the water injection period of 24 hours, and the water injection period of 36 hours. In this embodiment, the reactor water injection start time is not used as the input parameter as described above, but instead, if the time from the scram generation time to the reactor water injection stop time, which is the input parameter, is set as the water injection period, it collapses. It is possible to calculate the water level drop behavior after the reactor water injection is stopped due to heat. The TAF (Top of Active Fuel) in FIG. 4A represents the top of the effective fuel, and indicates that the water level in the core is at the upper end position of the fuel loaded on the fuel rods in the reactor. Further, BAF (Bottom of Active Fuel) represents the bottom of the effective fuel, and indicates that the water level in the core is at the lower end position of the fuel loaded on the fuel rods in the reactor.

In FIG. 4B, the horizontal axis is time [h] and the vertical axis is hydrogen generation amount [kg], the RPV pressure and its history, the number of RPV rapid pressure reducing valves, and the hydrogen generation amount for each water level when the reactor water injection is stopped. It is a plot of the state prediction source data of. According to FIG. 4B, it can be seen that the time at which the amount of hydrogen generated rapidly increases varies greatly depending on the RPV pressure and its history, the number of RPV rapid pressure reducing valves, and the water level when the reactor water injection is stopped. Therefore, in the present embodiment, the number of RPV rapid pressure reducing valves and the water level when the reactor water injection is stopped are acquired as input parameters, and based on the input number of RPV rapid pressure reducing valves and the water level when the reactor water injection is stopped, the MAAP code is used in advance. It is configured to select the state prediction source data that matches the accident phase from the plurality of calculated state prediction source data. As a result, instead of newly performing the calculation by the MAAP code, the state prediction source data matching the accident phase is selected from the state prediction source data based on the pre-calculation parameters stored in the storage unit 250 based on the input parameters. You only have to do this, so you can quickly make the necessary predictions. In this embodiment, in addition to the input parameters as described above, prediction data such as the reactor water level, hydrogen generation amount, relocation timing, and reactor pressure vessel damage timing predicted by the input parameters are also taken into consideration. The optimum state prediction source data is selected. The relocation refers to a phenomenon in which the core is melted by heat and falls to the bottom of the reactor pressure vessel 1.

In addition to the data shown in FIGS. 4 (a) and 4 (b), the state prediction source data calculated using the MAAP code based on the pre-calculation parameters includes a large number of data such as source term, RPV pressure, and PCV pressure. Is being prepared.

Next, the method of generating the prediction data from the state prediction source data will be described more specifically by taking the PCV pressure as an example. Tables 2 to 5 show the database configuration of the state prediction source data related to the PCV pressure in each accident phase of "water injection period", "after water injection stop", "after relocation", and "after RPV damage", respectively. .. For example, the database in the "water injection period" shown in Table 2 has seven databases (DB) from "HPOS" to "LOSF" for each RPV pressure and water injection source. Here, the RPV pressure: high pressure refers to a high pressure state in which the RPV rapid depressurization is not performed, and the RPV pressure: low pressure refers to a low pressure state after the RPV rapid decompression is performed. The RPV pressure can be acquired, for example, by the control unit 230 acquiring the measured value of the RPV pressure from the reactor system 100 via the communication unit 270. In addition, the RPV pressure state is acquired from the RPV pressure state prediction source data calculated using the MAAP code based on the pre-calculation parameters, and the RPV pressure that best matches each accident phase based on the input parameters and the like. This may be done by selecting the state prediction source data. The three databases in the lower part of Table 2, LISX, LOSX, and LOSF, are all data in LOCA, and three cases are assumed depending on the water source and the presence or absence of liquid phase outflow. In the "water injection period" shown in Table 2, the state prediction source data matching the accident phase can be selected only from the RPV pressure and the water injection source for the following reasons.

The radioactive material in the reactor pressure vessel 1 emits radiation and decays, and the energy released at that time is changed to heat, which is called decay heat. The decay heat Q is the amount of heat released during the time when the reactor is in the shutdown state, and therefore depends on the water injection shutdown time. Assuming a model in which light water in the reactor pressure vessel 1 is changed to steam with a steam generation amount W by the decay heat Q, the steam generation amount W tends to be proportional to the decay heat Q and inversely proportional to the latent heat of vaporization hf of water. There is.

Here, the latent heat of vaporization hf is the amount of heat absorbed when light water per unit mass, which is a coolant, changes from a liquid to a vapor. The latent heat of vaporization hf of light water depends on the internal pressure in the reactor pressure vessel 1. That is, when the pressure in the reactor pressure vessel 1 is high, the latent heat of vaporization hf becomes small, and even if the same decay heat Q is absorbed, the amount of steam generated W increases, while when the pressure in the reactor pressure vessel 1 is low, it evaporates. The latent heat hf becomes large, and even if the same decay heat Q is absorbed, the amount of steam generated W decreases. Therefore, since the amount of steam generated W differs depending on the level of the RPV pressure, the PCV pressure is affected. The control unit 230 determines whether the RPV pressure is high pressure or low pressure from information such as the reactor rapid decompression time and the number of reactor rapid decompression valves, which are input parameters.

Further, the control unit 230 recognizes whether the water injection water source is inside or outside from the input parameter. When the water injection water source is outside, the amount of water in the PCV increases, while the water temperature decreases, which affects the amount of steam generated. The control unit 230 selects the state prediction source data matching the accident phase from the state prediction source data related to the PCV pressure calculated based on the pre-calculation parameters based on the RPV pressure determination result and the water injection water source recognition result. To do. More specifically, a database matching the RPV pressure and the water injection source is selected from the databases HPOS to LPIS in Table 2.

In FIG. 5 (a), the horizontal axis is time [h], the vertical axis is D / W pressure (PCV pressure) [Pa (a)] ((a) means absolute pressure), the RPV pressure, and the reactor. It is a plot of the state prediction source data of the D / W pressure for each water source of water injection. According to FIG. 5A, it can be seen that the rate of increase in the D / W pressure varies greatly depending on the RPV pressure and the water source of the reactor water injection. Therefore, in the present embodiment, the water source of the reactor water injection is acquired as an input parameter, and a plurality of state prediction source data calculated in advance by the MAAP code based on the input water source of the reactor water injection and the determination result of the RPV pressure. Therefore, it is configured to select the state prediction source data that matches the accident phase. This makes it possible to make the necessary predictions quickly.

In the example of KK6 shown in FIG. 3, the scrum time was 13:00 on May 1, 2016, and the reactor water injection stop time was 20:00 on May 1, 2016. Therefore, the "water injection period" from the scrum time, that is, the time when the fission reaction stops and only the decay heat starts to be generated, to the reactor water injection stop time is 7 hours. Further, in KK6, since the rapid depressurization is performed at the same time as the scrum time, the RPV pressure is "low pressure" from the beginning. Also, from the input parameters, the water injection source is "inside", that is, the suppression pool 22.

From the graphs shown in FIG. 5A, the control unit 230 selects a graph corresponding to the example of KK6 in FIG. 3, RPV pressure: low pressure, and water injection water source: inside. Further, as described above, since the water injection period is 7 hours, the portion 0 to 7 [h] on the horizontal axis in FIG. 5A is used as the “state prediction source data matching the accident phase” in the “water injection period”. select. The selected state prediction source data is stored in the storage unit 250 and used for PCV pressure prediction data generation, which will be described later.

Next, the database "after water injection is stopped" will be described. In the database "after water injection stop" shown in Table 3 below, 11 databases from "L2HH012" to "LLOCA00" for each reactor water injection stop water level, RPV pressure transition when water level drops, number of pressure reducing valves, and water injection period. Has (DB). In the example of this embodiment, the water level when the reactor water injection is stopped is represented by seven stages of L1, L1.5, L2, L3, L4, NML, and L8, which means that L8 has the highest water level. .. In addition, NWL represents the normal water level.

The RPV pressure transition when the water level drops indicates how the RPV pressure changed during the period “after the water injection was stopped”. For example, in the first half of the period "after the water injection is stopped", the rapid depressurization is not performed yet and the RPV pressure is kept in a high pressure state, while the rapid decompression is performed during the period "after the water injection is stopped". When the RPV pressure becomes a low pressure state in the latter half of the period "after the water injection is stopped", the RPV pressure transition when the water level drops becomes "high pressure → low pressure". Further, if rapid depressurization has not yet been performed even at the end of the "after water injection stop" period, the RPV pressure "after water injection stop" remains high. Therefore, the RPV pressure transition when the water level drops in that case is “high pressure → high pressure”. Further, when the rapid depressurization is performed at the beginning of the "water injection period" or the "after the water injection stop" period, the RPV pressure remains low for almost the entire period "after the water injection stop". Therefore, the RPV pressure transition when the water level drops in that case is “low pressure → low pressure”.

In this "after water injection stop" accident phase, the database for the water level when the reactor water injection is stopped, the RPV pressure transition when the water level drops, the number of pressure reducing valves, and the water injection period is provided for the following reasons. That is, if the water level is high when the reactor water injection is stopped, it takes time to expose the fuel, so that the timing of the PCV pressure rise due to hydrogen generation after the fuel is exposed is delayed. In addition, the amount of steam transferred to the PCV changes depending on the level of the RPV pressure transition, which affects the PCV pressure. If the number of pressure reducing valves changes, the speed of change of the RPV pressure changes, which affects the PCV pressure. Since the water injection period affects the decay heat, it affects how the water level decreases and the amount of steam generated. Therefore, it also affects the PCV pressure.

Since the databases: LLOCA00, TQUAX000, and TQUV000 assume that the accident will progress immediately after normal operation without the water injection period, the water level when water injection is stopped is the same NWL as during normal operation, and the water injection period. Is 0 hours.

In the example of KK6 in FIG. 3, the water level when water injection is stopped is NWL, and the water injection period is 7 hours as described above. Further, as described above, the rapid depressurization has already been carried out during the water injection period, and the RPV pressure has become low. Therefore, since the RPV pressure remains low even after this "after the water injection is stopped", the RPV pressure transition when the water level drops is "low pressure → low pressure".

However, as shown in Table 3 and FIG. 5 (b), a plurality of state prediction source data (PCV pressures) calculated based on the pre-calculation parameters “after water injection stop” calculated in advance using the MAAP code. ) Is calculated only for L2 and L8 as the water level when water injection is stopped (excluding LLOCA00, TQUAX000, and TQUV000, which assume that the accident will progress immediately after the operation without the water injection period). Also, the water injection period is calculated only for 12 hours, 24 hours or 36 hours (excluding LLOCA00, TQUAX000, and TQUV000). Therefore, it is necessary to calculate the data of the water level at the time of stopping the water injection: NWL and the water injection period: 7 hours by interpolation (extrapolation) from these state prediction source data.

First, the state prediction source data of RPV pressure (low pressure → low pressure) _ water level (L2) _ water injection (12 hours) shown in FIG. 5 (b) and RPV pressure (low pressure → low pressure) _ water level (L8) _ water injection (12 hours). Using the state prediction source data of time), the difference data of the state prediction source data at the water level NWL with respect to the water level L2 is calculated by interpolation. More specifically, the control unit 230 calculates the difference data of the state prediction source data in the water level NWL with respect to the water level L2 by the following equation (1).

Difference data of state prediction source data in water level NWL with respect to water level L2 = {(RPV pressure (low pressure → low pressure) _ water level (L8) _ state prediction source data of water injection (12 hours))-(RPV pressure (low pressure → low pressure) _ Water level (L2) _ water injection (12 hours) state prediction source data)} / (water level L8-water level L2) x (water level NWL-water level L2) (1)

Next, the control unit 230 uses the RPV pressure (high pressure → high pressure) _ water level (L2) _ water injection (12 hours) state prediction source data shown in FIG. 5 (b) and the RPV pressure (high pressure → high pressure) _ water level ( L2) _ Using the state prediction source data of water injection (24 hours), the difference data of the state prediction source data in the water injection period of 7 hours with respect to the water injection period of 12 hours is calculated by extrapolation. More specifically, the control unit 230 calculates the difference data of the state prediction source data in the water injection period of 7 hours with respect to the water injection period of 12 hours by the following equation (2).

Difference data of the state prediction source data in the water injection period of 7 hours with respect to the water injection period of 12 hours = {(RPV pressure ((high pressure → high pressure) _ water level (L2) _ state prediction source data of water injection (24 hours))-(RPV pressure (RPV pressure ( High pressure → high pressure) _ water level (L2) _ water injection (12 hours) state prediction source data)} / (water injection period 24 hours-water injection period 12 hours) x (water injection period 7 hours-water injection period 12 hours) (2)

The control unit 230 uses the difference between the equations (1) and (2) to obtain the prediction source data for the water level L2, the RPV pressure (low pressure → low pressure), and the water injection period of 12 hours, which are the conditions that serve as the basis for the above two interpolation calculations. By adjusting using the data, the prediction source data for the water level when the reactor water injection is stopped: NWL, the RPV pressure transition (low pressure → low pressure) when the water level drops, and the water injection period: 7 hours are calculated. "Adjust using the difference data of equations (1) and (2)" means, for example, the equation (1) in the state prediction source data of RPV pressure (low pressure → low pressure) _ water level (L2) _ water injection (12 hours). ) And (2) can be performed by simply adding the difference data. Further, when adding the difference data of the equation (2), adjustment may be made by multiplying by a coefficient corresponding to the difference between the RPV pressure (low pressure → low pressure) and the RPV pressure (high pressure → high pressure) and then adding. Good.

When the water injection period is extrapolated and calculated by the formula (2), the state in the RPV pressure transition (high pressure → high pressure) different from the RPV pressure transition (low pressure → low pressure) which is the state of the reactor "after the water injection is stopped". Prediction source data is used. This is because the difference in PCV pressure due to the water injection period does not differ significantly between the RPV pressure transition (low pressure → low pressure) and the RPV pressure transition (high pressure → high pressure), so the water injection period is as shown in Table 3 and FIG. 5 (b). : As the state prediction source data for 24 hours and 36 hours, only the data in the RPV pressure transition (high pressure → high pressure) is prepared so that the number of state prediction source data to be stored is reduced as much as possible.

In addition to the data on the time change of the D / W pressure (PCV pressure) shown in FIG. 5B, the plurality of databases shown in Table 3 show the time required from the stop of water injection to the relocation calculated by the MAAP code. I have the data. The control unit 230 calculates the time from the water injection stop to the relocation according to the current state of the reactor by performing interpolation calculation and the like for the data of the time required from the water injection stop to the relocation as well as the PCV data. To do. In the present embodiment, since the time from the water injection stop to the relocation is 8.3 hours, the control unit 230 inputs the PCV pressure state prediction source data adjusted by the above-mentioned interpolation and extrapolation data from the water injection stop. Only 8.3 hours are selected and stored in the storage unit 250. This state prediction source data is used for generating PCV pressure prediction data, which will be described later, as "state prediction source data matching the accident phase" after "after water injection is stopped".

In the above calculation example (example using the input parameter of KK6 in FIG. 3), only the state prediction source data of RPV pressure transition (low pressure → low pressure) and RPV pressure transition (high pressure → high pressure) is dealt with, so that it is rapid. The number of valves at the time of decompression does not matter (in Table 3, the database of RPV pressure transition (low pressure → low pressure) and RPV pressure transition (high pressure → high pressure) are both displayed as the number of pressure reducing valves: 0 valves). However, when rapid depressurization is performed during this "after water injection stop" period, the RPV pressure changes (high pressure → low pressure), and the depressurization speed changes depending on the number of pressure reducing valves. The control unit 230 selects a database that matches the number of pressure reducing valves in Table 3 based on the number of RPV rapid pressure reducing valves input as an input parameter. If there is no pressure reducing valve number that matches the number of RPV rapid pressure reducing valves entered in Table 3, the database of one pressure reducing valve number and the number of pressure reducing valves are the same as in the above water level and water injection period when water injection is stopped. Interpolate the 8-valve database. Then, it is used for generating PCV pressure prediction data, which will be described later, as "state prediction source data matching the accident phase" after "after water injection is stopped".

Next, the database "after relocation" will be described. The database "after relocation" shown in Table 4 below has seven databases (DB) from "L2H012" to "TQUV00" for each reactor water injection shutdown water level, RPV pressure, and water injection period.

In this "post-relocation" accident phase, a database is provided for each water level when the reactor water injection is stopped because the water level when the reactor water injection is stopped affects the water level at the bottom of the pressure vessel when the relocation occurs. This is because it affects the PCV pressure behavior through the amount of steam transferred to the PCV later.

The databases: LLOCA0, TQUAX00, and TQUV00 assume that the accident will progress immediately after normal operation without the water injection period, so the water level when water injection is stopped is the same NWL as during normal operation, and the water injection period. Is 0 hours.

In the example of KK6 in FIG. 3, the water level when water injection is stopped is NWL, and the water injection period is 7 hours as described above. Further, as described above, the rapid depressurization has already been carried out during the water injection period, and the RPV pressure has become low. Therefore, the RPV pressure is low even after this “after relocation”.

However, as shown in Table 4 and FIG. 5 (c), a plurality of state prediction source data (PCV pressures) calculated based on the pre-calculation parameters “after relocation” calculated in advance using the MAAP code. Is calculated only for L2 and L8 as the water level when water injection is stopped (excluding LLOCA0, TQUAX00, and TQUV00, which assume that the accident will progress immediately after the operation without the water injection period). The water injection period is calculated only for 12 hours, 24 hours or 36 hours (excluding LLOCA0, TQUAX00, and TQUV00). Therefore, in FIG. 5C, the water level L2 when water injection is stopped, RPV pressure: low pressure, and the state prediction source data (RPV pressure (low pressure) _ water level (L2) _ water injection (12 hours)) in which the water injection period is 12 hours are used as a reference. , The difference data with respect to it is acquired by interpolation or extrapolation, and the data of water level when water injection is stopped: NWL and water injection period: 7 hours is calculated.

First, the state prediction source data of RPV pressure (high pressure) _ water level (L2) _ water injection (12 hours) and RPV pressure (high pressure) _ water level (L8) _ water injection (24 hours) shown in FIG. 5 (c) are used. Then, the difference data of the state prediction source data in the water injection period of 7 hours with respect to the water injection period of 12 hours is calculated by extrapolation. More specifically, the control unit 230 calculates the difference data of the state prediction source data in the water injection period of 7 hours with respect to the water injection period of 12 hours by the following equation (3).

Difference data of the state prediction source data in the water injection period of 7 hours with respect to the water injection period of 12 hours = {(RPV pressure (high pressure) _ water level (L8) _ state prediction source data of water injection (24 hours))-(RPV pressure (high pressure) _ Water level (L2) _ State prediction source data of water injection (12 hours)} / (Water injection period 24 hours-Water injection period 12 hours) x (Water injection period 7 hours-Water injection period 12 hours) (3)

Here, the two state prediction source data for generating the difference data by the equation (3) are the state prediction source data in which the water levels (L2 and L8) at the time of stopping the water injection are different in addition to the water injection period (12 hours and 24 hours). Is. The RPV pressure is also the state prediction source data under high pressure conditions. However, in this accident phase "after relocation", if the water level is lower than the fuel, it does not change in the sense that the fuel temperature rises abnormally, and the difference in PCV pressure due to the water injection period changes greatly depending on the water level when water injection is stopped. There is no such thing. Further, the difference in PCV pressure due to the water injection period is not significantly affected by the RPV pressure. Therefore, it can be considered that the difference data of the above two state prediction source data substantially calculates the difference in the water injection period. In this way, even in the accident phase "after relocation", as shown in Table 4 and FIG. 5 (c), only the data at the RPV pressure (high pressure) is prepared as the state prediction source data for the water injection period: 24 hours and 36 hours. I try to reduce the number of state prediction source data to be stored as much as possible.

The control unit 230 uses the difference data obtained as a result of the above interpolation calculation for the RPV pressure (low pressure) _ water level (L2) _ water injection (12 hours), which is the reference state prediction source data. By adjusting, the prediction source data for the water injection period: 7 hours is calculated. "Adjustment using the difference data obtained as a result of the interpolation calculation" can be performed, for example, by simply adding the difference data of the equation (3). Further, when adding the difference data of the equation (3), after multiplying the difference between the RPV pressure (low pressure) and the RPV pressure (high pressure) and the coefficient corresponding to the difference in the water level when the water injection is stopped (NWL and L2). Adjustment may be made by adding or the like.

In addition to the data on the time change of the D / W pressure (PCV pressure) shown in FIG. 5 (c), the plurality of databases shown in Table 4 show the time required from the relocation calculated by the MAAP code to the RPV damage. I have the data. The control unit 230 calculates the time from relocation to RPV damage according to the current state of the reactor by performing interpolation calculation and the like for the data of the time from this relocation to RPV damage in the same manner as the PCV data. To do. In the present embodiment, since the time from this relocation to the RPV damage is 7.1 hours, the control unit 230 relocates the PCV pressure state prediction source data adjusted by the above-mentioned interpolation (extrapolation) data. 7.1 Only one hour is selected and stored in the storage unit 250. This state prediction source data is used for PCV pressure prediction data generation, which will be described later, as "state prediction source data matching the accident phase" in "after relocation".

Next, the database "after RPV damage" will be described. The database "after RPV damage" shown in Table 5 below has five databases (DB) from "PXH" to "P00" for each of the presence or absence of pedestal water injection and the RPV pressure before RPV damage.

In this "after RPV damage" accident phase, a database is provided for each presence or absence of pedestal water injection and RPV pressure before RPV damage for the following reasons. That is, while there is an effect of lowering the temperature inside the PCV depending on the presence or absence of pedestal water injection, it can be a factor of increasing the amount of steam generated, which affects the PCV pressure. Further, when the RPV pressure before the RPV breakage is in a high pressure transition (high pressure for a long time from a relatively early stage), steam tends to flow into the PCV and the PCV pressure tends to rise. On the other hand, when the RPV pressure changes to a low voltage (it is lowered from an early stage to a low pressure), heat flows into the PCV side from an early stage, so that the heat is easily absorbed by the wall of the reactor containment vessel 20, so that the PCV Pressure rise is unlikely to occur.

In the example of KK6 in FIG. 3, rapid depressurization has already been performed during the water injection period as described above, and the RPV pressure is low. Further, as is clear from the description in the column of KK6 in FIG. 3, pedestal water injection is not performed. Therefore, the control unit 230 selects “RPV pressure transition (long-term low pressure) _no pedestal water injection” from the plurality of prediction source data shown in FIG. 5 (d). The selected state prediction source data is stored in the storage unit 250. Then, it is used for generating PCV pressure prediction data, which will be described later, as "state prediction source data matching the accident phase" in "after RPV damage".

FIG. 6 shows that the prediction data is generated from the state prediction source data by taking the PCV pressure as an example. That is, from the PCV pressure state prediction source data calculated using the MAAP code based on the pre-calculation parameters illustrated in FIGS. 5A to 5D, the input parameters and other data for each accident phase are obtained. It is the PCV pressure prediction data obtained by selecting the most matching PCV pressure state prediction source data based on the prediction data and accumulating the selected state prediction source data for each accident phase. More specifically, the prediction data of FIG. 6 is generated by each of the following steps.

First, the control unit 230 is stored in the storage unit 250 for each accident phase of "water injection period", "after water injection stop", "after relocation", and "after RPV damage", and is in a state corresponding to the "accident phase". Read "Prediction source data". In the example of KK6 in FIG. 3, the water injection period is 7 hours, so the control unit 230 first reads out the state prediction source data for 7 hours in the "water injection period" stored in the storage unit 250. The horizontal axis 0 to 7 hours in FIG. 6 is a plot of the selected state prediction source data in this “water injection period”. In FIG. 6, the PCV pressure on the vertical axis is converted from the absolute pressure in FIG. 5 to the gauge pressure.

Subsequently, the control unit 230 reads out the "state prediction source data matching the accident phase" stored in the storage unit 250 "after the water injection is stopped". The state prediction source data in this "after water injection stop" is the equation (1) and (1) with respect to the state prediction source data of RPV pressure (low pressure → low pressure) _ water level (L2) _ water injection (12 hours) as described above. It is the state prediction source data adjusted using 2). Then, when the control unit 230 reads out the state prediction source data for the 8.3 hours, the state of "after water injection is stopped" after the state prediction source data in the "water injection period" which is the "accident phase" before that. Accumulate forecast source data. That is, as shown in FIG. 6, when the time 0 [h] is set at the start of water injection, the PCV pressure value at the time 7 [h] at the end of water injection is the initial value of the PCV pressure "after the water injection is stopped". Therefore, the increase or decrease of the PCV pressure in the accident phase "after the water injection is stopped" is accumulated on the PCV pressure value in the "water injection period".

The same processing is performed for the subsequent accident phases "after relocation" and "after RPV damage". That is, by accumulating the increase or decrease of the PCV pressure value in the current accident phase with respect to the PCV pressure value in the previous accident phase, from the start of water injection to after the RPV damage as shown in FIG. It is possible to generate data showing the relationship between each accident phase (time) of the above and the D / W pressure (PCV pressure) prediction data.

The state prediction source data stored in the storage unit 250 and calculated using the MAAP code based on the pre-calculation parameters is approximated by the number of data according to the complexity of the change with respect to the time axis for each accident phase. It is stored. Therefore, when generating the prediction data as shown in FIG. 6, it is used to generate the prediction data after appropriately interpolating the state prediction source data stored in the storage unit 250.

Further, in FIG. 6, 2Pd is the limit pressure in the reactor containment vessel 20. Further, 1Pd is a PCV design pressure smaller than the limit pressure for warning the occurrence of relocation in the reactor. In this embodiment, it is assumed that venting is performed before the PCV pressure reaches 2Pd.

FIG. 7 is a diagram showing an example of an output screen of the prediction result generated by the control unit 230. As shown in FIG. 7, the predicted results are the time when the top of active fuel (TAF) of the reactor water level was reached, the time when the bottom of active fuel (BAF) was reached at the reactor water level, and the lower part of the RPV was damaged. The time, the PCV pressure 1Pd arrival time, the PCV pressure 2Pd arrival time, the environmental release time, and the environmental release amount. This output screen is output when the control unit 230 displays and prints the prediction data on the output unit 290. In FIG. 7, DIANA (Dose Information Analysis for Nuclear Accident) is three-dimensional around the nuclear power plant every 10 minutes, assuming that radioactive rare gas / iodine and particulate matter are released into the atmosphere. It is a system that can simulate the translocation and diffusion phenomenon and evaluate the air dose rate at any point.

8 (a) to 8 (c) show the comparison result between the prediction data by the reactor risk management device 200 according to the present embodiment and the prediction data using only the MAAP code. 8 (a) to 8 (c) show the time [time] on the horizontal axis, and the vertical axis shows (a) RPV pressure, (b) PCV pressure, and (c) RPV water level, respectively. ing. In any of the prediction results, the prediction result by the reactor risk management device 200 of the present embodiment (shown by the solid line in FIGS. 8A to 8C) is the prediction result using the MAAP code (FIG. 8A). The tendency of time change is similar to (shown by a broken line in (c)), and the MAAP calculation result is almost reproduced for the absolute value.

9 (a) and 9 (b) are diagrams showing an example of an output screen of the prediction result by the reactor risk management device 200 according to the first embodiment of the present invention for each prediction result. 9 (a) and 9 (b) both show the time on the horizontal axis, and the vertical axis shows (a) PCV pressure and (b) environmental release amount, respectively. In addition, each horizontal axis indicates the date and time when the three accidents of water injection stop, relocation, and RPV damage occurred. Also in the example of FIG. 9A, as described with reference to FIG. 6, the state prediction source data (state prediction source data based on the pre-calculation parameters) for each accident phase during the water injection period, after the water injection is stopped, after the relocation, and after the RPV is damaged. PCV pressure prediction data is generated by selecting state prediction source data that matches each accident phase from (PCV pressure) and accumulating increments of PCV pressure in each accident phase. According to the prediction result of FIG. 9A, although there is no significant change in the transition of the PCV pressure when the water injection is stopped, the PCV pressure rises significantly instantly at the time of relocation, and the PCV pressure rises faster after the RPV is damaged. Is increasing monotonically. Then, at the time when the PCV pressure reaches the limit pressure (2Pd), as shown in FIG. 9B, various radioactive substances are predicted to be released to the environment.

In the examples of FIGS. 9A and 9B, the prediction result is obtained using only the state prediction source data based on the parameters for pre-calculation using the MAAP code, but the present invention is not limited to this mode. For example, the control unit 230 may be configured to acquire actual measurement data such as RPV damage time and PCV pressure from the reactor system 100 via the communication unit 270, and appropriately correct the prediction result based on the acquired data. ..

FIG. 10 is a flowchart showing an operation flow of the reactor risk management device 200 according to the present embodiment. First, the control unit 230 stores the state prediction source data calculated based on the pre-calculation parameters in the storage unit 250 (step S101). The state prediction source data calculated based on this pre-calculation parameter is, for example, the state prediction of each RPV pressure and D / W pressure (PCV pressure) for each water source of reactor water injection as shown in FIG. 5 (a). The original data. Next, the input unit 210 receives the input parameter input by the user using the keyboard or the like from the input screen shown in FIG. 3, for example (step S102), and stores it in the storage unit 250. The control unit 230 reads the input parameters stored in the storage unit 250 and the state prediction source data based on the pre-calculation parameters, and based on the input parameters and the like, the state prediction based on the plurality of read pre-calculation parameters. From the original data, the state prediction source data that matches the accident phase is selected (step S103). The selection of the state prediction source data that matches this accident phase is, for example, from the state prediction source data of each RPV pressure and the D / W pressure (PCV pressure) for each water source of the reactor water injection as shown in FIG. 5 (a). , The RPV pressure most suitable for each accident phase and the state prediction source data of the D / W pressure (PCV pressure) corresponding to the water source of the reactor water injection are selected and adjusted as necessary. Then, the control unit 230 determines the reactor water level, the timing of the reactor pressure vessel breakage, the reactor containment vessel pressure, the environmental release time of the radioactive material, and the environmental release amount of the radioactive material based on the selected state prediction source data. At least one of them is predicted (step S104). This prediction is shown in FIG. 6, for example, by accumulating the state prediction source data of the D / W pressure (PCV pressure) most suitable for each accident phase selected from FIGS. 5 (a) to 5 (d) for each accident phase. It is performed by generating predicted data of PCV pressure such as. By executing step S104, the operation flow of the reactor risk management device 200 ends.

The reactor risk management device 200 can be configured to approximate the PCV pressure state prediction source data based on the same pre-calculation parameters with data having a different number of points for each accident phase. For example, as shown in FIG. 11, when the PCV pressure increases monotonically in the accident phase 1 while the PCV pressure increases in a cubic curve in the accident phase 2, the accident phase 1 is approximated by two-point data. However, the accident phase 2 may be configured to approximate with 4-point data. As a result, the amount of data to be stored in the storage unit 250 can be reduced, and highly accurate prediction can be performed with a small amount of calculation.

In this embodiment, an example is described in which state prediction source data matching the accident phase is selected from the state prediction source data, and PCV pressure prediction data is generated based on the selected state prediction source data. However, it is not limited to this example. Prediction data can be generated by the same procedure for the reactor water level, the timing of reactor pressure vessel damage, the environmental release time of radioactive materials, and the environmental release amount of radioactive materials.

As described above, in the present embodiment, at least seven input parameters are input, and the accident phase is matched from the state prediction source data calculated based on the pre-calculation parameters that are larger than the input parameters stored in the storage unit 250. Select the state prediction source data to be performed, and based on the selected state prediction source data, the reactor water level, the timing of reactor pressure vessel damage, the reactor containment vessel pressure, the environmental release time of radioactive materials, and the environment of radioactive materials. It was configured to predict at least one of the emissions. As a result, it is not necessary to input many parameters and execute the calculation by the MAAP code when an emergency occurs, and the state that matches the accident phase based on a small number of input parameters from the already calculated state prediction source data. The desired prediction data can be obtained by selecting the prediction source data. Therefore, when an emergency occurs, it is possible to quickly obtain necessary forecast data and transmit information.

Further, in the present embodiment, since the number of input parameters is less than 1/15 of the number of pre-calculation parameters that need to be input, it is possible to more quickly acquire the prediction data required in the event of an emergency. .. In particular, in one example of the present embodiment, after the cooling water injection is stopped, the prediction data can be obtained instantly after the input parameters are input, so that the information can be transmitted with a margin for the vent timing.

In the present embodiment, the input parameters are parameters other than the seven essential input parameters among the parameters for pre-calculation, that is, the scram occurrence time, the accident scenario, the reactor water injection stop time, the reactor water injection stop time water level, and the atom. It may be configured not to include at least one parameter other than the reactor pressure vessel rapid decompression time, the number of reactor pressure vessel rapid pressure reducing valves, and the presence or absence of pedestal water injection. Examples of parameters not included in the input parameters include, for example, the presence or absence of reactor water injection after the reactor pressure vessel is damaged, the presence or absence of spray into the reactor containment vessel, the environmental release path of radioactive substances, and the presence or absence of filter vents. , With or without pH control, water level at reactor water injection, reactor water injection start time, water injection pump characteristics, reactor water injection destination, coolant water source temperature, initial gas phase component and temperature in PCV, dry well spray flow rate, etc. .. By not including at least one of these parameters in the input parameter, it is possible to omit the input of the parameter having a small influence on the event progress and acquire the necessary prediction data more quickly.

Further, in the present embodiment, the prediction data is corrected based on the timing of the damage to the reactor pressure vessel and the measured value of at least one of the reactor containment vessel pressures. Thereby, the accuracy of the prediction data can be further improved.

Further, in the present embodiment, it is configured to predict the time until the reactor containment vessel pressure reaches a predetermined value based on the selected state prediction source data. As a result, the vent timing can be predicted quickly.

Further, in the present embodiment, the input parameters include the presence / absence of reactor water injection after the reactor pressure vessel is damaged, the presence / absence of spray in the reactor containment vessel, the environmental release path of radioactive materials, the presence / absence of filter vent, and pH control. It was configured to include at least one of the presence or absence. This makes it possible to more accurately predict the amount of radioactive substances released into the environment.

Further, in the present embodiment, the state prediction source data is configured to be approximated by the number of data according to the accident phase. As a result, the state prediction source data can be configured with a small amount of data, so that the storage data in the storage unit 250 can be reduced and the calculation amount required for prediction can be reduced to quickly obtain the prediction data.

Further, in the present embodiment, after the injection of the cooling water into the reactor is stopped, the reactor water level, the timing of the reactor pressure vessel breakage, the reactor containment vessel pressure, and the like, assuming that the water injection is continuously stopped. It was configured to predict the time of environmental release of radioactive material and at least one of the amount of radioactive material released into the environment. This eliminates the need to deal with complicated events that recover after water injection is stopped, so necessary forecast data can be obtained quickly.

Further, in the present embodiment, the amount of radioactive material released to the environment is predicted in consideration of the pedestal water injection, the reactor water injection, and the dry well spray after the injection of the cooling water into the reactor is stopped. This makes it possible to accurately predict the amount of radioactive substances released into the environment.

Next, the reactor system 300 according to the second embodiment of the present invention will be specifically described with reference to the drawings.

(Second Embodiment)
FIG. 12 is a block diagram showing an example of the configuration of the reactor system 300, which is the prediction target by the reactor risk management device 400 according to the second embodiment of the present invention. In this embodiment, the reactor system 300 will be described by taking a pressurized water reactor (PWR) as an example, and the differences from the boiling water reactor of the first embodiment will be mainly described.

The nuclear reactor system 300 surrounds the reactor pressure vessel 301 provided with the fuel and the core 303, which are radioactive substances, and the reactor pressure vessel 301, and suppresses the release of the radioactive substance to the outside when the coolant is lost or the like. The containment vessel 320, the pressurizer 310 that pressurizes the pressurized water that is the primary cooling material heated in the reactor pressure vessel 301, and the heated and pressurized pressurized water are introduced into the heat exchanger 341 to be used as the secondary cooling material. It includes a steam generator 340 that heats a certain light water to generate steam, and a reactor building (not shown) that surrounds the reactor containment vessel 320. Further, in FIG. 12, a steam generator uses a primary coolant pump 305 that supplies pressurized water as a primary coolant into the reactor pressure vessel 301, and light water that is a secondary coolant cooled via a turbine or the like. The secondary coolant pump 345 supplied to the 340 is also shown.

As shown in FIG. 12, in the reactor pressure vessel 301, pressurized water, which is the primary coolant cooled by the heat exchanger 341 in the steam generator 340, is supplied to the inside by the primary coolant pump 305 via the primary water supply channel 305a. ing. Further, the pressurized water heated in the reactor pressure vessel 301 is supplied to the pressurizer 310 via the pressurized water channel 307a to be pressurized, and then supplied to the steam generator 340. The heated and pressurized pressurized water heats and vaporizes the light water, which is the secondary coolant, by radiating heat in the heat exchanger 341 in the steam generator 340, and at the same time, it cools itself and returns to the reactor pressure vessel 301. Return.

The light water heated by the pressurized water in the heat exchanger 341 becomes high-temperature steam, and then is sent to the steam turbine via the steam supply path 347a and a pressure regulating valve (not shown) to generate electricity by rotating the steam turbine. Do.

The reactor containment vessel 320 has a role as a barrier so that the radioactive material leaked from the reactor pressure vessel 301 does not leak to the outside when the coolant is lost or the like. A pressurizer relief tank (not shown) is provided in the reactor containment vessel 320, and when the pressure in the pressurizer 310 exceeds a predetermined threshold value, the rapid pressure reducing valve of the pressurizer 310 is opened to apply pressure. The pressurized water in the pressure device 310 is discharged into the pressure device release tank. When the pressure in the steam generator 340 exceeds a predetermined threshold value, the rapid pressure reducing valve of the steam generator 340 is opened, and steam is discharged into the environment from, for example, the steam supply path 347a.

Here, the relationship between the reactor system 300 according to the second embodiment and the reactor system 100 according to the first embodiment is summarized in Table 6.

In a pressurized water reactor, the space directly below the reactor pressure vessel 301 is called a reactor cavity instead of the pedestal in the boiling water reactor. Reactor water injection water source in a pressurized water reactor: “Inside” means to use the water accumulated in the reactor cavity as a water source for reactor water injection after an accident occurs.

In the pressurized water reactor, the water injection stop time and the water level at the time of water injection stop are considered separately for the primary system and the secondary system. That is, the water injection stop time of the primary system is the time when the water injection into the pressurizer 310 of the pressurized water which is the primary coolant is stopped, and the water injection stop time of the secondary system is the steam of the light water which is the secondary coolant. It is the time when the water injection into the generator 340 is stopped. Similarly, the water level when the water injection of the primary system is stopped is the water level in the pressurizer 310 when the water injection into the pressurizer 310 of the pressurized water which is the primary coolant is stopped, and the water level when the water injection of the secondary system is stopped is , The water level in the steam generator 340 when the injection of light water, which is a secondary coolant, into the steam generator 340 is stopped.

In addition, the time of rapid decompression and the number of rapid decompression valves are also considered separately for the primary system and the secondary system in the pressurized water reactor. That is, the rapid decompression time of the primary system is the time when the pressure in the pressurizer 310 exceeds a predetermined threshold value, so that the rapid decompression of the pressurizer 310 is performed, and the rapid decompression time of the secondary system is the steam generator 340. It is the time when the steam generator 340 was rapidly depressurized because the pressure inside exceeded a predetermined threshold value. Similarly, the number of rapid decompression valves of the primary system is the number of valves opened when the pressure in the pressurizer 310 exceeds a predetermined threshold value and the pressurizer 310 is rapidly depressurized, and the number of rapid decompression valves of the secondary system The number is the number of valves opened when the pressure in the steam generator 340 exceeds a predetermined threshold value and the steam generator 340 is rapidly depressurized.

In the rapid depressurization of the pressurized water reactor, the rapid depressurization of the primary system corresponds to the rapid depressurization of the RPV in the BWR. The secondary system does not exist in the BWR, but if this secondary coolant is lost, the heat removing function of the primary system is lost, and as a result, the water level of the primary system drops. Therefore, it is necessary to consider the secondary system as well.

In a pressurized water reactor, there is no facility to inject water directly into the reactor cavity. However, when blowout water or PCV spray water by LOCA is accumulated in the reactor cavity, the conditions are similar to those of pedestal water injection in a boiling water reactor. Therefore, in the pressurized water reactor, "presence or absence of water supply to the reactor cavity" is read as "presence or absence of pedestal water injection" and used as an input parameter.

In the pressurized water reactor, there is no input parameter corresponding to this because no countermeasure is taken by injecting water into the reactor after the RPV is damaged.

Further, in the pressurized water reactor, there is no wet well, and the release route of the radioactive substance is limited to the dry well, so no input parameter corresponding to this is provided.

Further, pH control is performed in both a boiling water reactor and a pressurized water reactor. In the boiling water reactor, an alkaline substance is charged into the suppression pool 22, whereas in the pressurized water reactor, the PCV spray water source is alkaline. The substance is being charged.

Since the reactor risk management device 400 according to the second embodiment has the configuration shown in FIG. 2 as in the first embodiment, detailed description here will be omitted. The user inputs input parameters from the input unit 210 when predicting the state inside the reactor by the reactor risk management device 400 of the present embodiment. The input of this input parameter is different from the input parameter of the secondary system in addition to the primary system for the water injection stop time, the water level at the time of water injection stop, the rapid depressurization time, and the number of rapid decompression valves as compared with the first embodiment. You may be able to enter in. In addition, it is not necessary to enter the presence or absence of reactor water injection after RPV damage and the release route of radioactive materials. Further, as described above, the input parameter may be input by replacing "presence or absence of water supply to the reactor cavity" with "presence or absence of pedestal water injection".

The transition conditions between the "water injection period" and "after the water injection is stopped" shown in the first embodiment can be arbitrarily determined. For example, the accident phase may be changed from the "water injection period" to "after the water injection is stopped" when the water injection of the primary system is stopped, or the water injection of the primary system and the water injection of the secondary system may be stopped. The accident phase may be changed from the "water injection period" to "after the water injection is stopped" when both of the above are stopped. Further, the accident phase may be shifted when either the primary system water injection or the secondary system water injection is stopped, and further, the accident phase may be changed at the timing when the secondary system water injection is stopped. Good.

In addition, the database for each accident phase is configured to handle values of the secondary system as well as the primary system for the water injection stop time (water injection period), water level at the time of water injection stop, rapid decompression time, number of rapid pressure reducing valves, etc. You may. For example, the database of the accident phase “after water injection stop” shown in Table 3 in the first embodiment may be configured to have a database for each water level at the time of water injection stop of the primary system and the secondary system. Further, it may be configured to have a database for each number of rapid pressure reducing valves of the primary system and the secondary system, or for each water injection period.

Although the present invention has been described with reference to the drawings and examples, it should be noted that those skilled in the art can easily make various modifications or modifications based on the present disclosure. Therefore, it should be noted that these modifications or modifications are within the scope of the present invention. For example, the functions included in each component can be rearranged so as not to be logically inconsistent, and a plurality of components can be combined or divided into one.

1 Reactor pressure vessel 3 Core 5 Water supply pump 5a Water supply channel 7 Pressure regulating valve 7a Steam supply channel 9 Main steam relief valve 9a Relief flow path 20 Reactor containment vessel 22 Suppression pool 22a Discharge channel 22b Atmosphere open path 30 Filler vent section 30a Open path 40 Reactor building 100 Reactor system 200 Reactor risk management device 210 Input unit 230 Control unit 250 Storage unit 270 Communication unit 290 Output unit 300 Reactor system 301 Reactor pressure vessel 303 Core 305 Primary coolant pump 305a Primary water supply Road 307a Pressurized water channel 310 Pressurizer 320 Reactor containment vessel 340 Steam generator 341 Heat exchanger 345 Secondary coolant pump 345a Secondary water supply channel 347a Steam supply channel

Claims (8)

Scrum occurrence time and
Accident scenario and
Reactor water injection stop time and
The water level when the reactor water injection is stopped and
Reactor pressure vessel rapid decompression time and
Reactor pressure vessel rapid pressure reducing valve number and
An input section for inputting input parameters including the presence or absence of pedestal water injection,
A storage unit that stores a plurality of state prediction source data calculated based on pre-calculation parameters that are larger than the input parameters.
Based on the input parameters, the state prediction source data matching the accident phase is selected from the plurality of state prediction source data based on the pre-calculation parameters, and the reactor water level, based on the selected state prediction source data, It is equipped with a control unit that predicts at least one of the timing of reactor pressure vessel damage, the reactor containment vessel pressure, the environmental release time of radioactive materials, and the amount of radioactive materials released to the environment .
The reactor pressure vessel rapid pressure reducing valve number is the number of pressure reducing valves opened according to the accident phase out of the total number of pressure reducing valves installed in the plant . The reactor risk management device according to claim 1, wherein the control unit corrects the predicted data based on the timing of the reactor pressure vessel breakage and at least one measured value of the reactor containment vessel pressure. The reactor risk management device according to claim 1 or 2, wherein the control unit predicts the time until the reactor containment vessel pressure reaches a predetermined value based on the selected state prediction source data. The input parameters further include the presence or absence of reactor water injection after the reactor pressure vessel is damaged, the presence or absence of spray to the reactor containment vessel, the environmental release path of radioactive materials, the presence or absence of filter vent, and the presence or absence of pH control. seen at least 1 Tsuo含,
The reactor risk management apparatus according to any one of claims 1 to 3, wherein the pH control is to supply an alkaline substance into the reactor containment vessel. After the injection of cooling water to the reactor is stopped, the control unit assumes that the injection of water is continuously stopped, such as the reactor water level, the timing of reactor pressure vessel damage, the reactor containment vessel pressure, and radioactive materials. The reactor risk management apparatus according to any one of claims 1 to 4, which predicts at least one of the environmental release time and the amount of radioactive material released into the environment. Any of claims 1 to 5, wherein the control unit predicts the amount of radioactive material released to the environment in consideration of pedestal water injection, reactor water injection, and dry well spray after the injection of cooling water into the reactor is stopped. The reactor risk management device described in item 1. Depending on the input section
Scrum occurrence time and
Accident scenario and
Reactor water injection stop time and
The water level when the reactor water injection is stopped and
Reactor pressure vessel rapid decompression time and
Reactor pressure vessel rapid pressure reducing valve number and
Steps to enter input parameters, including with or without pedestal water injection,
A step of storing a plurality of state prediction source data calculated based on pre-calculation parameters that are larger than the input parameters by the storage unit, and
A step of selecting a state prediction source data matching the accident phase from a plurality of state prediction source data based on the pre-calculation parameter based on the input parameter by the control unit.
Of the reactor water level, reactor pressure vessel damage timing, reactor containment vessel pressure, radioactive material environmental release time, and radioactive material environmental release amount, based on the state prediction source data selected by the control unit. look including a step of predicting at least one of the,
The reactor pressure vessel rapid pressure reducing valve number is the number of pressure reducing valves opened according to the accident phase out of the total number of pressure reducing valves installed in the plant . Reactor risk management method. On the computer
Scrum occurrence time and
Accident scenario and
Reactor water injection stop time and
The water level when the reactor water injection is stopped and
Reactor pressure vessel rapid decompression time and
Reactor pressure vessel rapid pressure reducing valve number and
Procedures for entering input parameters, including with or without pedestal water injection,
A procedure for storing a plurality of state prediction source data calculated based on pre-calculation parameters that are larger than the input parameters, and a procedure for storing the plurality of state prediction source data.
A procedure for selecting a state prediction source data that matches the accident phase from a plurality of state prediction source data based on the pre-calculation parameters based on the input parameters, and
Based on the selected state predictor data, at least one of the reactor water level, the timing of reactor pressure vessel breakage, the reactor containment pressure, the time of environmental release of radioactive material, and the amount of radioactive material released to the environment. to execute the steps of predicting,
The reactor pressure vessel rapid pressure reducing valve number is the number of pressure reducing valves opened according to the accident phase out of the total number of pressure reducing valves installed in the plant, which is a reactor risk management program.

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