JP2021117092A - Risk evaluation system, risk evaluation method and risk evaluation program for nuclear power plant - Google Patents

Abstract

To perform risk evaluation for a nuclear power plant with a light calculation load.SOLUTION: A risk evaluation system for a nuclear power plant includes a core damage prevention system for preventing a core from being damaged, and a containment vessel damage prevention system for preventing a containment vessel from being damaged. The risk evaluation system for the nuclear power plant comprises: a first analysis model that has a first event tree including events about the core damage prevention system and the containment vessel damage prevention system, and that outputs scenario information including success or failure of each event; and a second analysis model that has a second event tree including events about physicochemical phenomena related to loss of functions of the containment vessel, and that outputs the probability frequency at which functions of the containment vessel are lost. These models are used to calculate the probability frequency at which functions of the containment vessel are lost in each scenario.SELECTED DRAWING: Figure 2

Description

This disclosure relates to a nuclear plant risk assessment system, risk assessment methods, and risk assessment programs.

In nuclear plants, probabilistic risk assessment (PRA) is used to consider, for example, the uncertainty of results due to internal events inherent in plant operation management and external events such as earthquakes and tsunamis. After that, the effects of different safety measures are compared and the safety of the facility is comprehensively evaluated.

Patent Document 1 discloses one of the evaluation examples of PRA in a nuclear power plant. PRA in a nuclear plant is mainly divided into three levels, and level 1 PRA identifies accident scenarios leading to core damage and evaluates the frequency of occurrence of those scenarios. In Level 2 PRA, in addition to Level 1 PRA, accident scenarios in which a large amount of radioactive material diffuses from the reactor containment vessel are identified, and hydrothermal power and radioactive materials inside the reactor cooling system and in the containment vessel in the accident scenario are identified. The behavior of the radioactive material is predicted, and the type and amount of radioactive substances released into the environment and their frequency are evaluated. In Level 3 PRA, in addition to Level 2 PRA, the transfer of radioactive materials released from nuclear plants into the environment is predicted in consideration of weather conditions, etc., and the health effects of exposure on the general public are evaluated. In Patent Document 1, a method of systematically evaluating the risk at the time of evacuation response when an emergency event occurs in a nuclear plant is studied in consideration of the risk evaluation by Level 3 PRA.

JP-A-2018-60255

Atomic Energy Society of Japan, Atomic Energy Society of Japan Standard "Implementation Standards for Probabilistic Risk Assessment for Output Operating Conditions of Nuclear Power Plants (Level 1 PRA): 2013" Atomic Energy Society of Japan, Atomic Energy Society of Japan Standard "Implementation Standards for Probabilistic Risk Assessment for Output Operating Conditions of Nuclear Power Plants (Level 2 PRA): 2016" Atomic Energy Society of Japan, Atomic Energy Society of Japan Standard "Implementation Standards for Probabilistic Safety Evaluation of Nuclear Power Plants (Level 3 PSA Edition): 2008"

As mentioned above, the PRA of a nuclear power plant is divided into three levels for risk assessment. For example, in Level 2 PRA, in addition to Level 1 PRA, accident scenarios in which a large amount of radioactive material diffuses from the reactor containment vessel are specified. Therefore, each scenario specified in the event tree corresponding to Level 1 PRA is changed to Level 2 PRA. By inheriting to the corresponding event tree, the probability frequency for each scenario is calculated. The calculation of such a probability frequency has conventionally been performed based on an event tree integrally configured over Level 1 PRA and Level 2 PRA. However, in such a method, since the level 1 PRA and the level 2 PRA are evaluated all at once, the event tree becomes huge and the calculation load becomes huge. For example, in level 1 PRA, about 1000 patterns of scenarios are assumed depending on the combination of success or failure of each event, and if these are individually handed over to level 2 PRA, the amount of calculation becomes enormous. Since the calculation load is enormous, the time required for the calculation becomes enormous, and in some cases, the behavior of the computer device becomes unstable and stops.

At least one embodiment of the present disclosure has been made in view of the above circumstances, and provides a risk assessment system, a risk assessment method, and a risk assessment program for a nuclear plant capable of performing risk assessment in a nuclear plant with a small computational load. The purpose is to do.

The nuclear plant risk assessment system according to at least one embodiment of the present disclosure is designed to solve the above problems.
A nuclear plant risk assessment system that includes a core damage prevention system to prevent core damage and a containment damage prevention system to prevent containment damage.
It has a first event tree containing a plurality of events related to the core damage prevention system and the containment damage prevention system, and the success or failure of each event related to the containment damage prevention system leading to the damage of the containment vessel. The first analysis model that outputs scenario information including
A second event tree including a plurality of events related to a physicochemical phenomenon related to the loss of function of the containment vessel is provided, and the probability frequency of loss of function of the containment vessel is output for each scenario corresponding to the scenario information. Analytical model and
A success / failure probability setting unit that sets a success / failure probability of success or failure for each event included in the first event tree and the second event tree.
When the occurrence frequency of the causative event is input to the first analysis model, the scenario information output from the first analysis model is input to the second analysis model, so that each scenario based on the scenario information is input. The probability frequency calculation unit that calculates the probability frequency of loss of function of the storage container in
To be equipped.

The risk assessment method for a nuclear plant according to at least one embodiment of the present disclosure is to solve the above problems.
A risk assessment method for a nuclear plant that includes a core damage prevention system to prevent core damage and a containment damage prevention system to prevent containment damage.
(I) A first event tree including a plurality of events related to the core damage prevention system and the containment damage prevention system, and each event related to the containment damage prevention system leading to damage to the containment vessel. It has a first analysis model that outputs scenario information including success or failure, and (ii) a second event tree that includes a plurality of events related to physicochemical phenomena related to the loss of function of the containment vessel, and corresponds to the scenario information. A process of preparing a second analysis model that outputs the probability frequency of the loss of function of the containment vessel for each scenario to be performed, and a process of preparing a second analysis model.
A step of setting a success or failure probability for each event included in the first event tree and the second event tree, and
A step of inputting the frequency of occurrence of the causative event into the first analysis model, and
By inputting the scenario information output from the first analysis model into the second analysis model, a step of calculating the probability frequency of loss of function of the storage container in each scenario based on the scenario information, and a step of calculating the probability frequency.
To be equipped.

The nuclear plant risk assessment program according to at least one embodiment of the present disclosure is designed to solve the above problems.
A nuclear plant risk assessment program that includes a core damage prevention system to prevent core damage and a containment damage prevention system to prevent containment damage.
With a computer device
(I) A first event tree including a plurality of events related to the core damage prevention system and the containment damage prevention system, and each event related to the containment damage prevention system leading to damage to the containment vessel. It has a first analysis model that outputs scenario information including success or failure, and (ii) a second event tree that includes a plurality of events related to physicochemical phenomena related to the loss of function of the containment vessel, and corresponds to the scenario information. A process of preparing a second analysis model that outputs the probability frequency of the loss of function of the containment vessel for each scenario to be performed, and a process of preparing a second analysis model.
A step of setting a success or failure probability for each event included in the first event tree and the second event tree, and
A step of inputting the frequency of occurrence of the causative event into the first analysis model, and
By inputting the scenario information output from the first analysis model into the second analysis model, a step of calculating the probability frequency of loss of function of the storage container in each scenario based on the scenario information, and a step of calculating the probability frequency.
Is feasible.

According to at least one embodiment of the present disclosure, it is possible to provide a risk assessment system, a risk assessment method, and a risk assessment program for a nuclear plant capable of performing risk assessment in a nuclear plant with a small computational load.

It is a schematic block diagram of a nuclear power plant. It is an overall block diagram of the risk evaluation system which concerns on at least one Embodiment of this disclosure. It is a figure which shows the 1st analysis model. It is a figure which shows the 2nd analysis model together with a part of the 1st analysis model. This is an example of setting scenario information. This is an example of the frequency of occurrence for each causative event. It is a flowchart which shows the risk evaluation method which concerns on at least one Embodiment of this disclosure for each process.

Hereinafter, some embodiments of the present invention will be described with reference to the accompanying drawings. However, the dimensions, materials, shapes, relative arrangements, etc. of the components described as embodiments or shown in the drawings are not intended to limit the scope of the present invention to this, and are merely explanatory examples. do not have.

The risk assessment system 100 according to at least one embodiment of the present disclosure targets the nuclear plant 1. FIG. 1 is a schematic configuration diagram of a nuclear power plant 1.

The nuclear plant 1 includes a reactor containment vessel 2 for generating steam by the thermal energy generated in the fission reaction. The reactor containment vessel 2 includes a reactor vessel 4 (pressure vessel) having a core containing nuclear fuel. A fuel rod (not shown) containing pelletized nuclear fuel (for example, uranium fuel, MOX fuel, etc.) is housed in the reactor vessel 4, and the thermal energy generated in the fission reaction of this fuel causes the reactor vessel 4 to contain the fuel rod (not shown). The primary cooling water inside is heated. The reactor vessel 4 is provided with control rods 8 for absorbing and adjusting the number of neutrons generated in the core including nuclear fuel in order to control the reactor output.

The primary cooling loop 10 through which the primary cooling water flows includes a pressurizer 12, a steam generator 14, and a primary coolant pump 16 in addition to the above-mentioned reactor vessel 4. The pressurizer 12 is configured to pressurize the primary cooling water in the primary cooling loop 10 so that the primary cooling water does not boil. In the steam generator 14, the secondary cooling water is heated to generate steam by heat exchange between the primary cooling water flowing through the primary cooling loop 10 and the secondary cooling water (secondary coolant) flowing through the secondary cooling loop 18. Let me. The primary coolant pump 16 is configured to circulate the primary cooling water in the primary cooling loop 10.
The pressurizer 12, the steam generator 14, and the primary coolant pump 16 are housed in the reactor containment vessel 2 together with the reactor vessel 4 described above.

The pipes 21 and 22 constituting the secondary cooling loop 18 are connected to a steam turbine (not shown). The steam turbine is driven by steam sent from the steam generator 14 via the pipe 21 to generate electricity. The steam that drives the steam turbine is cooled by a condenser (not shown) to be condensed, and is returned to the steam generator 14 through the pipe 22.

The reactor containment vessel 2 shown in FIG. 1 is a pressurized water reactor (PWR). In other embodiments, the reactor containment vessel 2 may be a boiling water reactor (BWR) or, unlike a light water reactor including a pressurized water reactor and a boiling water reactor, decelerates. It may be a type of nuclear reactor that uses a substance other than light water as a material or a cooling material.

The nuclear power plant 1 having such a configuration includes a core damage prevention system for preventing damage to the core (reactor vessel 4), a containment damage prevention system for preventing damage to the reactor containment vessel 2, and the containment vessel damage prevention system. To be equipped. The core damage prevention system includes a plurality of devices MS1, MS2, ... That function to prevent damage to the core, and FIG. 1 shows a core injection system 15 as a partial example. The containment vessel damage prevention system includes a plurality of devices CS1, CS2, ... That function to prevent damage to the containment vessel 2, and FIG. 1 shows the containment vessel spray system 17 as a partial example. Has been done.
The devices MS1, MS2, ... Constituting the core damage prevention system and the devices CS1, CS2, ... Constituting the containment vessel damage prevention system may be different from each other or may overlap each other. good.

Next, a risk evaluation system 100 for evaluating the nuclear power plant 1 having the above configuration will be described. In the risk assessment system 100, when a core damage occurs in the nuclear power plant 1, the core damage progresses, the reactor containment vessel 2 is damaged, and the reactor containment vessel 2 loses its function. Quantitative risk assessment is performed by calculating the probability frequency of each scenario for each damage state of the reactor containment vessel 2. Such a risk assessment system 100 has, for example, a hardware configuration including an electronic arithmetic unit such as a computer, and by installing the risk assessment program according to at least one embodiment of the present disclosure, at least the present disclosure. The risk assessment method according to one embodiment is configured to be feasible.

The risk assessment program may be recorded in a predetermined storage medium, for example. In this case, the risk assessment system 100 is realized by installing the risk assessment program in the hardware configuration by reading the storage medium. A storage medium on which such a risk assessment program is recorded is also included in one embodiment of the present disclosure.

FIG. 2 is an overall configuration diagram of the risk assessment system 100 according to at least one embodiment of the present disclosure. The risk evaluation system 100 includes a causal event occurrence frequency acquisition unit 102, an analysis model storage unit 104, a success / failure probability setting unit 106, a probability frequency calculation unit 108, and an analysis unit 110.

Although the internal configuration of the risk evaluation system 100 is shown by functional blocks in FIG. 2, these functional blocks are merely an example of defining each step of the risk evaluation method described later, and some of them are used. The functional blocks may be integrated or further subdivided into a plurality of functional blocks.

The causal event occurrence frequency acquisition unit 102 is configured to acquire the occurrence frequency of the causal event that may cause damage to the core of the nuclear plant 1. The causative event is an event that can be assumed as a factor of damaging the core of the nuclear power plant 1, and includes, for example, a large break LOCA in an internal event and a reactor building damage in an external event. The frequency of occurrence of each causative event acquired by 102 copies of the frequency of occurrence of the causative event may be an actual value based on past data or a virtual value calculated by simulation or the like.

FIG. 6 is an example of the frequency of occurrence for each causative event. FIG. 6 shows a plurality of possible causal events A, B, ... Regarding core damage. Further, the occurrence frequencies NA, NB, ... Are specified in advance for each of the causative events A, B, .... Such a combination of the causative event A, B, ... And the occurrence frequency NA, NB, ... Is readable and stored in a storage device accessible in advance, and is read out by the causative event occurrence frequency acquisition unit 102. It may be acquired by the operator, or what is input via the input interface (not shown) by the operator's operation may be acquired by the cause event occurrence frequency acquisition unit 102.

The analysis model storage unit 104 is configured to store the analysis model 112 used for carrying out risk evaluation, and the hardware configuration is composed of a storage device such as a database, for example. The analysis model 112 has an event tree containing a plurality of events constituting each scenario from the core damage generated in the nuclear plant 1 to the damage of the reactor containment vessel 2. This event tree accepts the occurrence frequencies NA, NB, ... For each of the cause events A, B, ... Acquired by the cause event occurrence frequency acquisition unit 102 as input, and receives the occurrence frequency NA, NB, ... For each scenario of the reactor containment vessel 2. It is configured to be able to output the probability frequency of functional loss.

3 and 4 are diagrams conceptually showing the analysis model 112 stored in the analysis model storage unit 104. The analysis model 112 includes a first analysis model 112A and a second analysis model 112B, FIG. 3 shows the first analysis model 112A, and FIG. 4 shows the second analysis model 112B as the first analysis model 112A. Shown with some.

The first analysis model 112A has a first event tree ET1 that receives as input the occurrence frequencies NA, NB, ... For each of the cause events A, B, ... Acquired by the cause event occurrence frequency acquisition unit 102. , The output of the first event tree ET1 is configured to be input to the second analysis model 112B. The second analysis model 112B has a second event tree ET2 that receives the output from the first event tree ET1 of the first analysis model 112A as an input, and determines the probability frequency of loss of function of the reactor containment vessel 2 for each scenario. It is configured to output.

The first analysis model 112A has a first event tree ET1 including a plurality of events related to the core damage prevention system included in the nuclear plant 1 and the containment vessel damage prevention system. The first event tree includes a core damage prevention event tree ET1A including a plurality of events related to the core damage prevention system and a containment vessel damage prevention event tree ET1B including a plurality of events related to the containment vessel damage prevention system. The core damage prevention event tree ET1A is an event tree corresponding to level 1 PRA among the three levels of PRA in a nuclear plant, and the containment vessel damage prevention event tree ET1B is a part of the event tree corresponding to level 2 PRA related to the system. (That is, the first event tree ET1 has a configuration in which the system portion of the level 2 PRA is combined with the conventional level 1 PRA).

The core damage prevention event tree ET1A is configured to include at least a plurality of events MS1, MS2, ... Corresponding to the devices MS1, MS2, ... It is set for each event A, B, .... More specifically, each core damage prevention event tree ET1A has an event corresponding to a device that functions when a related causal event occurs among the devices MS1, MS2, ... It is arranged in a sequence order and has a branching structure corresponding to success or failure at each event. That is, the events included in the core damage prevention event tree ET1A are configured by selecting each event corresponding to the constituent equipment of the core damage prevention system related to the causative event corresponding to the core damage prevention event tree ET1A ( It should be noted that at least a part of the events included in the core damage prevention event tree ET1A corresponding to the different causative events may be duplicated).

The core damage prevention event tree ET1A outputs a core damage state PDS for each scenario defined by the branched structure including each event MS1, MS2, .... The core damage state PDS output from the core damage prevention event tree ET1A is classified based on a predetermined rule. In the example of FIG. 3, the core damage state PDS is classified into a plurality of core damage states PDSa, PDSb, ..., According to the characteristics based on some viewpoints. In this way, by classifying the core damage state PDS output from the core damage prevention event tree ET1A according to a predetermined rule, the amount of data transferred to the subsequent containment damage prevention event tree ET1B can be suppressed, and the calculation load can be effectively reduced. It can be mitigated.

The containment vessel damage prevention event tree ET1B is configured to include, at least partially, a plurality of events CS1, CS2, ... Corresponding to the devices CS1, CS2, ... It is set for each core damage state PDS, which is the output of the prevention event tree ET1A. More specifically, each containment vessel damage prevention event tree ET1B corresponds to the equipment that functions when the related core damage state PDS occurs among the equipment CS1, CS2, ... The events to be performed are arranged in chronological order, and each event has a branching structure corresponding to success or failure. That is, as the events included in the containment vessel damage prevention event tree ET1B, each event corresponding to the component equipment of the containment vessel damage prevention system related to the core damage state PDS input to the containment vessel damage prevention event tree ET1B is selected. (Note that at least a part of the events included in the containment vessel damage prevention event tree ET1B corresponding to the different core damage states PDS may be duplicated).

As described above, in the first event tree ET1, even if the core damage state PDSs are output from different core damage prevention event trees ET1A, the core damage state PDSs classified into the common category are the common containment vessel damage prevention event tree. It is input to ET1B. Explaining with reference to FIG. 3, a scenario in which the core damage state PDSa is output from the core damage prevention event tree ET1A corresponding to the cause event A, and the core damage state from the core damage prevention event tree ET1A corresponding to the cause event B. The scenario in which PDSa is output is guided to the common storage container damage prevention event tree ET1B that inputs the core damage state PDSa. As a result, compared to the case where the containment vessel damage prevention event tree ET1B is individually prepared for each of the scenarios included in the core damage prevention event tree ET1A set for each causative event, the containment vessel damage prevention event tree ET1B The number of is small, and the calculation load can be effectively reduced.

The containment damage prevention event tree ET1B is an event related to the containment damage prevention system up to each damage state of the reactor containment vessel 2 for each scenario via a branched structure including each event CS1, CS2, .... Outputs scenario information SI including success or failure of CS1, CS2, .... Here, FIG. 5 is a setting example of the scenario information SI. In FIG. 5, in each event CS1, CS2, ... Included in the containment vessel damage prevention event tree ET1B, "0" is given if it succeeds, and "1" is given if it fails. The rules for defining the scenario information SI by the combination of these are shown in. For example, in a scenario in which each of the events CS1, CS2, ... Is successful, "PDSa-000" is output as the scenario information SI. Further, in the scenario in which the events CS1 and CS2 succeed and the event CS3 fails, "PDSa-001" is output as the scenario information SI. Further, in the scenario in which the events CS1 and CS3 succeed and the event CS2 fails, "PDSa-010" is output as the scenario information SI.

The containment vessel damage prevention event tree ET1B is configured to be able to output MCS information in addition to such scenario information SI. The MCS information is information on the equipment damage state that causes success or failure in each event included in at least one of the core damage prevention event tree ET1A or the containment vessel damage prevention event tree ET1B. Specifically, the MCS information includes the equipment damage state in at least a part of the equipment MS1, MS2, ..., Which constitutes the core damage prevention system corresponding to each event included in the core damage prevention event tree ET1A. , Equipment damage state in at least a part of the equipment CS1, CS2, ..., Which constitutes the storage container damage prevention system corresponding to each event included in the storage container damage prevention event tree ET1B. By sending these MCS information to the analysis unit 110, it can be used for the importance / uncertainty analysis described later.

The second analysis model 112B has a second event tree ET2 including a plurality of events PH1, PH2, ... Regarding physicochemical phenomena related to the loss of function of the reactor containment vessel 2. The plurality of events PH1, PH2, ... Included in the second event tree ET2 relate to physicochemical phenomena corresponding to the plurality of events CS1, CS2, ... In the second event tree ET2 of the present embodiment, in addition to a plurality of events PH1, PH2, ... Related to the physicochemical phenomenon related to the loss of function of the reactor containment vessel 2, the containment vessel damage prevention event tree ET1B By being configured to include a plurality of events CS1, CS2, ... Included in, the two are linked. Such a second event tree ET2 is an event tree corresponding to level 2 PRA among the three levels of PRA in the nuclear power plant.

The second event tree ET2 is provided for each core damage state PDS, and a scenario having a common core damage state PDS is configured to be guided to a common second event tree ET2. Specifically, the scenarios in which "PDSa-000", "PDSa-001", "PDSa-010", ... Are output as scenario information SI from the containment vessel damage prevention event tree ET1B are common to these. It is input to the second event tree ET2 corresponding to the core damage state PDSa. In the second event tree ET2, each scenario is specified by specifying the branch destination of the branch structure based on the scenario information SI "PDSa-000", "PDSa-001", "PDSa-010", ... .. For example, when the scenario information SI is "PDSa-000", a successful scenario is specified for each of the events CS1, CS2, ... Included in the second event tree ET2. When the scenario information SI is "PDSa-001", a scenario that succeeds in events CS1 and CS2 and fails in event CS3 is specified. When the scenario information SI is "PDSa-010", a scenario that succeeds in events CS1 and CS3 and fails in event CS2 is specified.

In this way, in the second event tree ET2, the scenario is specified based on the scenario information SI output from the storage container damage prevention event tree ET1B of the first event tree ET1, so that the tree structure can be simplified. The calculation load can be effectively reduced.

Returning to FIG. 2, the success / failure probability setting unit 106 sets the success / failure probability of success or failure for each event of the event tree included in the analysis model stored in the analysis model storage unit. Specifically, as described above, the analysis model storage unit 104 stores the first analysis model 112A and the second analysis model 112B, and has the first event tree ET1 and the second event tree ET2, respectively. The success / failure probability setting unit 106 sets the success / failure probability of success or failure for each event included in the first event tree ET1 and the second event tree ET2.

The success / failure probability setting in the success / failure probability setting unit 106 may be readable and stored in an accessible storage device in advance, and may be set by being read out by the success / failure probability setting unit 106, or may be set by the operator. It may be set by inputting a success / failure probability via a predetermined interface (not shown) by operation. Further, the success / failure probability in each event set by the success / failure probability setting unit 106 may be an actual value based on past data or a virtual value calculated by simulation or the like.

The probability frequency calculation unit 108 uses the analysis model stored in the analysis model storage unit 104 to calculate the probability frequency CF in which each scenario leading to the loss of function of the containment vessel occurs. Specifically, the probability frequency calculation unit 108 inputs the occurrence frequency for each causative event to the analysis model (see FIGS. 3 and 4), and then, for each scenario defined in the event tree of the analysis model. Calculate the probability frequency CF. By calculating the probability frequency CF in this way, it is possible to quantitatively evaluate the risk of how often each scenario may occur.
The detailed calculation method of the probability frequency by the probability frequency calculation unit 108 will be described later.

The analysis unit 110 has a function of performing importance / uncertainty analysis based on MCS information obtained in the process of calculating the probability frequency using the analysis model. The importance / uncertainty analysis is obtained by identifying the factors that control the probability frequency using MCS information and evaluating the uncertainty range. This makes it possible to carry out the importance / uncertainty analysis required by the PRA Society standard.

Subsequently, a risk evaluation method that can be implemented using the risk evaluation system 100 having the above configuration will be described. FIG. 7 is a flowchart showing the risk evaluation method according to at least one embodiment of the present disclosure for each process.

First, the probability frequency calculation unit 108 acquires an analysis model from the analysis model storage unit 104 (step S1). The analysis model described above is stored in advance in the analysis model storage unit 104 with reference to FIGS. 3 and 4, and the probability frequency calculation unit 108 acquires the analysis model by accessing the analysis model storage unit 104.

The analysis model stored in the analysis model storage unit 104 is a causal event assumed in the nuclear plant 1, and the constituent devices MS1 and MS2 of the core damage prevention system functioning to prevent the core damage caused by each causal event. Based on the containment vessel damage prevention system components CS1, CS2, ..., And the physicochemical phenomena PH1, PH2, ..., Which function to prevent damage to the containment vessel due to core damage. Will be built. The construction of the analysis model may be performed automatically using an electronic arithmetic unit such as a computer, or may be performed manually by the work of the operator.

Subsequently, the success / failure probability setting unit 106 sets the success / failure probability for each event included in the event tree included in the analysis model 112 acquired in step S1 (step S2). The success / failure probability setting unit 106 includes each event included in the first event tree ET1 included in the first analysis model 112A included in the analysis model and each event included in the second event tree ET2 included in the second analysis model 112B. On the other hand, the success / failure probability is set. As a result, the probability of success or failure at the branch point of each event tree included in the analysis model 112 is set.

Subsequently, the probability frequency calculation unit 108 inputs the occurrence frequency of each causal event acquired by the causal event occurrence frequency acquisition unit 102 into the analysis model 112 in which the success / failure probability is set in step S2 (step S3). The probability frequency CF of each scenario included in the event tree included in the analysis model 112 is calculated (step S4). Specifically, the occurrence frequency for each cause event acquired by the cause event occurrence frequency acquisition unit 102 is first input to the first event tree ET1 of the first analysis model 112A of the analysis model 112. The occurrence frequencies of the causative events A, B, ... Are input to the core damage prevention event tree ET1A corresponding to each causal event in the first event tree ET1. The core damage prevention event tree ET1A outputs a core damage state PDS to be handed over to the containment vessel damage prevention event tree ET1B for each scenario defined by the branch crack structure including each event MS1, MS2, ....

The containment damage prevention event tree ET1B is provided for each core damage state PDS output from the core damage prevention event tree ET1A, and the common core damage state PDS leads to the corresponding 1 containment damage prevention event tree ET1B. To be taken. As a result, compared to the case where the containment vessel damage prevention event tree ET1B is individually prepared for each of the scenarios included in the core damage prevention event tree ET1A set for each causative event, the containment vessel damage prevention event tree ET1B The number of is small, and the calculation load can be effectively reduced.

The containment vessel damage prevention event tree ET1B outputs the scenario information SI corresponding to each scenario, so that each scenario of the first event tree ET1 is taken over by the second event tree ET2. The second event tree ET2 has the physicochemical phenomena PH1, PH2, ... Corresponding to each event included in the storage container damage prevention event tree ET1B of the first event tree ET1. The scenario can be handed over between the second event tree ET2 with a small amount of information.

The probability frequency calculation unit 108 calculates the probability frequency CF using the analysis model 112 in which the amount of information inherited between the event trees is reduced in this way. As a result, the calculation load can be significantly reduced as compared with the case of using an analysis model in which a series of flows from the core damage to the damage of the reactor containment vessel 2 is composed of one event tree. As a result, the processing time for calculating the probability frequency CF of each scenario in the risk evaluation system 100 can be effectively shortened.

Further, the analysis unit 110 acquires the MCS information output in the calculation process of the analysis model 112 as described above (step S5), and performs the importance / uncertainty analysis using the MCS information (step S6). This makes it possible to carry out the importance / uncertainty analysis required by the PRA Society standard.

As described above, according to the above embodiment, an analysis model for calculating the probability frequency CF for each scenario from the core damage to the loss of function of the reactor containment vessel 2 with respect to the occurrence frequency of the event causing the core damage. In 112, the scenario is inherited by the scenario information SI between the first event tree ET1 of the first analysis model 112A and the second event tree ET2 of the second analysis model 112B. As a result, the amount of information inherited between the first event tree ET1 and the second event tree ET2 is significantly reduced, and the calculation load for calculating the probability frequency CF can be effectively reduced.

In addition, it is possible to replace the components in the above-described embodiments with well-known components as appropriate without departing from the spirit of the present disclosure, and the above-described embodiments may be combined as appropriate.

The contents described in each of the above embodiments are grasped as follows, for example.

(1) The risk assessment system for a nuclear plant according to at least one embodiment of the present disclosure is
A risk assessment system (for example, the above) of a nuclear plant (for example, the nuclear plant 1 of the above embodiment) including a core damage prevention system for preventing damage to the core and a containment damage prevention system for preventing damage to the containment vessel. The risk assessment system 100) of the embodiment.
A first event tree (eg, events MS1, MS2, ..., CS1, CS2, ...) Containing a plurality of events related to the core damage prevention system and the containment vessel damage prevention system (for example, the events MS1, MS2, ..., CS1, CS2, ... Of the above embodiment). Scenario information (for example, scenario information SI of the above embodiment) having the first event tree ET1) of the above embodiment and including the success or failure of each event related to the containment vessel damage prevention system leading to the damage of the containment vessel. (For example, the first analysis model 112A of the above embodiment) and
A second event tree (for example, the second event tree ET2 of the above embodiment) including a plurality of events (for example, events PH1, PH2, ... Of the above embodiment) related to a physicochemical phenomenon related to the loss of function of the storage container. A second analysis model (for example, the second analysis model 112B of the above embodiment) that has and outputs the probability frequency that the function loss of the storage container occurs for each scenario corresponding to the scenario information.
A success / failure probability setting unit (for example, the success / failure probability setting unit 106 of the above embodiment) that sets the success / failure probability of success or failure for each event included in the first event tree and the second event tree.
When the occurrence frequency of the causative event is input to the first analysis model, the scenario information output from the first analysis model is input to the second analysis model, so that each scenario based on the scenario information is input. The probability frequency calculation unit (for example, the probability frequency calculation unit 108 of the above embodiment) for calculating the probability frequency (for example, the probability frequency CF of the above embodiment) for the loss of function of the storage container in
To be equipped.

According to the aspect (1) above, in the analysis model for calculating the probability frequency for each scenario from the core damage to the loss of the function of the storage container with respect to the occurrence frequency of the event causing the core damage, the first analysis model. The scenario is inherited by the scenario information between the first event tree possessed by the user and the second event tree possessed by the second analysis model. As a result, the amount of information inherited between the first event tree and the second event tree is significantly reduced, and the calculation load for calculating the probability frequency can be effectively reduced.

(2) In one aspect, in the above aspect (1),
The first analysis model has the first event tree set for each of the causative events.
The second analysis model has the second event tree set for each damage state of the containment vessel output from the first event tree.

According to the aspect (2) above, when the second event tree is provided for each scenario of the first event tree by providing the second event tree for each damage state of the containment vessel output from the first event tree. Compared with this, the number of the second event tree can be reduced. As a result, the amount of information inherited between the first event tree and the second event tree can be more effectively reduced, and the calculation load for calculating the probability frequency can be effectively reduced.

(3) In one aspect, in the above aspect (1) or (2),
The first analysis model outputs MCS information regarding a device damage state that causes success or failure in at least a part of the plurality of events included in the first event tree to the second analysis model.

According to the aspect (3) above, when calculating the probability frequency for each scenario, MCS information regarding the equipment damage state that causes the success or failure of each event is output. This makes it possible to carry out importance / uncertainty analysis using MCS information, and can appropriately respond to the requirements of the PRA Society standard.

(4) In one aspect, in any one of the above (1) to (3),
The second event tree includes events related to the storage container damage prevention system (for example, events CS1, CS2, ... Of the above-described embodiment) and events related to the physicochemical phenomenon (for example, events PH1, PH2, ... Of the above-described embodiment).・ ・) Including.

According to the aspect (4) above, the second event tree including the event related to the physicochemical phenomenon further includes the event related to the storage container damage prevention system included in the first event tree. As a result, the first event tree and the second event tree commonly include the events related to the storage container damage prevention system included in the first event tree, and each scenario in the first event tree is described by the scenario information. 2 It can be efficiently handed over to the event tree.

(5) In one aspect, in any one of the above (1) to (4),
The first event tree is
A core damage prevention event tree that includes a plurality of events related to the core damage prevention system and outputs the damage state of the core corresponding to each scenario, and a core damage prevention event tree.
A containment damage prevention event tree that includes a plurality of events related to the containment damage prevention system for each core damage state and outputs the damage state of the containment vessel corresponding to each scenario.
including.

According to the aspect (5) above, by constructing the first event tree including the core damage prevention event tree and the containment vessel damage prevention event tree, the calculation load for calculating the probability frequency is effectively reduced. can do.

(6) The risk assessment method for a nuclear plant according to at least one embodiment of the present disclosure is
It is a risk evaluation method of a nuclear power plant (for example, the nuclear power plant 1 of the above embodiment) including a core damage prevention system for preventing damage to the core and a containment damage prevention system for preventing damage to the containment vessel. ,
(I) A first event including a plurality of events related to the core damage prevention system and the containment vessel damage prevention system (for example, events MS1, MS2, ..., CS1, CS2, ... Of the above-described embodiment). Scenario information (eg, the scenario of the above embodiment) having a tree (for example, the first event tree ET1 of the above embodiment) and including the success or failure of each event related to the containment vessel damage prevention system leading to the damage of the containment vessel. A first analysis model that outputs information SI) (for example, the first analysis model 112A of the above embodiment) and (ii) a plurality of events related to physicochemical phenomena related to the loss of function of the containment vessel (for example, events of the above embodiment). It has a second event tree (for example, the second event tree ET2 of the above embodiment) including PH1, PH2, ...), And the probability frequency that the function loss of the containment vessel occurs for each scenario corresponding to the scenario information is determined. A step of preparing a second analysis model to be output (for example, the second analysis model 112B of the above embodiment) and
A step of setting a success or failure probability for each event included in the first event tree and the second event tree, and
The step of inputting the occurrence frequency of the causative event into the first analysis model and
By inputting the scenario information output from the first analysis model into the second analysis model, the probability frequency of loss of function of the storage container in each scenario based on the scenario information (for example, the probability frequency of the embodiment). The process of calculating CF) and
To be equipped.

According to the aspect (6) above, in the analysis model for calculating the probability frequency for each scenario from the core damage to the loss of the function of the storage container with respect to the occurrence frequency of the event causing the core damage, the first analysis model. The scenario is inherited by the scenario information between the first event tree possessed by the user and the second event tree possessed by the second analysis model. As a result, the amount of information inherited between the first event tree and the second event tree is significantly reduced, and the calculation load for calculating the probability frequency can be effectively reduced.

(7) The nuclear plant risk assessment program according to at least one embodiment of the present disclosure
A risk assessment program for a nuclear plant (eg, the nuclear plant 1 of the above embodiment) that includes a core damage prevention system to prevent core damage and a containment damage prevention system to prevent containment damage. ,
With a computer device
(I) A first event including a plurality of events related to the core damage prevention system and the containment vessel damage prevention system (for example, events MS1, MS2, ..., CS1, CS2, ... Of the above-described embodiment). Scenario information (eg, the scenario of the above embodiment) having a tree (for example, the first event tree ET1 of the above embodiment) and including the success or failure of each event related to the containment vessel damage prevention system leading to the damage of the containment vessel. A first analysis model that outputs information SI) (for example, the first analysis model 112A of the above embodiment) and (ii) a plurality of events related to physicochemical phenomena related to the loss of function of the containment vessel (for example, events of the above embodiment). It has a second event tree (for example, the second event tree ET2 of the above embodiment) including PH1, PH2, ...), And the probability frequency that the function loss of the containment vessel occurs for each scenario corresponding to the scenario information is determined. A step of preparing a second analysis model to be output (for example, the second analysis model 112B of the above embodiment) and
A step of setting a success or failure probability for each event included in the first event tree and the second event tree, and
A step of inputting the frequency of occurrence of the causative event into the first analysis model, and
By inputting the scenario information output from the first analysis model into the second analysis model, the probability frequency of loss of function of the storage container in each scenario based on the scenario information (for example, the probability frequency of the embodiment). The process of calculating CF) and
Is feasible.

According to the aspect (7) above, in the analysis model for calculating the probability frequency for each scenario from the core damage to the loss of the function of the storage container with respect to the occurrence frequency of the event causing the core damage, the first analysis model. The scenario is inherited by the scenario information between the first event tree possessed by the user and the second event tree possessed by the second analysis model. As a result, the amount of information inherited between the first event tree and the second event tree is significantly reduced, and the calculation load for calculating the probability frequency can be effectively reduced.

1 Nuclear Plant 2 Reactor Containment Vessel 4 Reactor Vessel 8 Control Rod 10 Primary Cooling Loop 12 Pressurizer 14 Steam Generator 15 Core Injection System 16 Primary Coolant Pump 17 Containment Vessel Spray System 18 Secondary Cooling Loops 21 and 22 Piping 100 Risk evaluation system 102 Cause event occurrence frequency acquisition unit 104 Analysis model storage unit 106 Success / failure probability setting unit 108 Probability frequency calculation unit 110 Analysis unit 112 Analysis model 112A First analysis model 112B Second analysis model

Claims (7)

A nuclear plant risk assessment system that includes a core damage prevention system to prevent core damage and a containment damage prevention system to prevent containment damage.
It has a first event tree containing a plurality of events related to the core damage prevention system and the containment damage prevention system, and the success or failure of each event related to the containment damage prevention system leading to the damage of the containment vessel. The first analysis model that outputs scenario information including
A second event tree including a plurality of events related to a physicochemical phenomenon related to the loss of function of the containment vessel is provided, and the probability frequency of loss of function of the containment vessel is output for each scenario corresponding to the scenario information. Analytical model and
A success / failure probability setting unit that sets a success / failure probability of success or failure for each event included in the first event tree and the second event tree.
When the occurrence frequency of the causative event is input to the first analysis model, the scenario information output from the first analysis model is input to the second analysis model, so that each scenario based on the scenario information is input. The probability frequency calculation unit that calculates the probability frequency of loss of function of the storage container in
A nuclear plant risk assessment system. The first analysis model has the first event tree set for each of the causative events.
The risk assessment system for a nuclear power plant according to claim 1, wherein the second analysis model has the second event tree set for each damage state of the containment vessel output from the first event tree. The first analysis model outputs MCS information regarding a device damage state that causes success or failure in at least a part of the plurality of events included in the first event tree to the second analysis model. The risk assessment system for a nuclear plant according to item 1 or 2. The risk assessment system for a nuclear plant according to any one of claims 1 to 3, wherein the second event tree includes an event related to the containment vessel damage prevention system and an event related to the physicochemical phenomenon. The first event tree is
A core damage prevention event tree that includes a plurality of events related to the core damage prevention system and outputs the damage state of the core corresponding to each scenario, and a core damage prevention event tree.
A containment damage prevention event tree that includes a plurality of events related to the containment damage prevention system for each core damage state and outputs the damage state of the containment vessel corresponding to each scenario.
The risk assessment system for a nuclear plant according to any one of claims 1 to 4, including the above. A risk assessment method for a nuclear plant that includes a core damage prevention system to prevent core damage and a containment damage prevention system to prevent containment damage.
(I) A first event tree including a plurality of events related to the core damage prevention system and the containment damage prevention system, and each event related to the containment damage prevention system leading to damage to the containment vessel. It has a first analysis model that outputs scenario information including success or failure, and (ii) a second event tree that includes a plurality of events related to physicochemical phenomena related to the loss of function of the containment vessel, and corresponds to the scenario information. A process of preparing a second analysis model that outputs the probability frequency of the loss of function of the containment vessel for each scenario to be performed, and a process of preparing a second analysis model.
A step of setting a success or failure probability for each event included in the first event tree and the second event tree, and
A step of inputting the frequency of occurrence of the causative event into the first analysis model, and
By inputting the scenario information output from the first analysis model into the second analysis model, a step of calculating the probability frequency of loss of function of the storage container in each scenario based on the scenario information, and a step of calculating the probability frequency.
A nuclear plant risk assessment method. A nuclear plant risk assessment program that includes a core damage prevention system to prevent core damage and a containment damage prevention system to prevent containment damage.
With a computer device
(I) A first event tree including a plurality of events related to the core damage prevention system and the containment damage prevention system, and each event related to the containment damage prevention system leading to damage to the containment vessel. It has a first analysis model that outputs scenario information including success or failure, and (ii) a second event tree that includes a plurality of events related to physicochemical phenomena related to the loss of function of the containment vessel, and corresponds to the scenario information. A process of preparing a second analysis model that outputs the probability frequency of the loss of function of the containment vessel for each scenario to be performed, and a process of preparing a second analysis model.
A step of setting a success or failure probability for each event included in the first event tree and the second event tree, and
A step of inputting the frequency of occurrence of the causative event into the first analysis model, and
By inputting the scenario information output from the first analysis model into the second analysis model, a step of calculating the probability frequency of loss of function of the storage container in each scenario based on the scenario information, and a step of calculating the probability frequency.
A viable nuclear plant risk assessment program.

Images