CN116361972A - Filter screen failure PSA modeling method and system for passive nuclear power plant - Google Patents
Filter screen failure PSA modeling method and system for passive nuclear power plant Download PDFInfo
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- 239000007795 chemical reaction product Substances 0.000 claims abstract description 12
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- 239000002826 coolant Substances 0.000 claims description 11
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Abstract
The invention relates to a filter screen failure PSA modeling method and system for a passive nuclear power plant, comprising the following steps: determining the total amount x of potential fragments of the nuclear power plant according to the inherent fragment amount of the nuclear power plant and the fragment amount in chemical reaction products after an expected accident; analyzing migration paths of fragments according to the quantity and the types of fragments generated at different break positions in the nuclear power plant; obtaining probability distribution of pit blocking under a certain accident type by utilizing a nuclear power plant equipment reliability database, and obtaining a process of transferring and accumulating fragments to a filter screen according to the relation between the filter screen failure rate r and the potential fragment quantity x, the break size y and the break position z of the nuclear power plant; dividing working conditions according to the sizes of x, y and z, determining the value of the filter screen blockage failure rate under each working condition, and performing PSA modeling according to the value results under different working conditions.
Description
Technical Field
The invention relates to the technical field of nuclear power safety, in particular to a filter screen failure PSA modeling method and system for a passive nuclear power plant.
Background
The statements in this section merely provide background information related to the present disclosure and may not necessarily constitute prior art.
And (3) carrying out probability safety evaluation (PSA) on the nuclear power plant, and correlating the failure probability of the whole nuclear power system with the failure probability of subsystems, components, external conditions and the like of each level through logical reasoning of the structure, so as to find out the occurrence frequency of various accidents and carry out safety evaluation.
In the third-generation AP series passive nuclear power plant PSA, the blockage of the pit filter screen and the filter screen with the built-in material-changing water tank has higher importance on the damage frequency of the reactor core of the power plant, the quantity of fragments generated in the nuclear power plant is different under different accident working conditions, the data of the blockage of the filter screen are different, and the current direct adoption of the same blockage data under various accident working conditions brings greater uncertainty.
For example, in existing PSA models, pit screen plugging failure employs a recommended value for plugging failure rate during normal filter operation in an advanced light water reactor user requirements file; or the filter screens corresponding to the IRWST (built-in material-changing water tank) are provided with pipelines which are connected with each other, and the filter screens are blocked only when the filter screens are blocked at the same time, so that the blocking failure of the IRWST filter screen is modeled as a basic event and is connected into an IRWST subsystem fault tree, and a blocking failure probability value is obtained; therefore, under different accident conditions, the quantity of fragments generated in the nuclear power plant is different, the corresponding filter screen blockage data are different, and the current direct adoption of the same blockage data under various accident conditions brings greater uncertainty.
Disclosure of Invention
In order to solve at least one technical problem in the background art, the invention provides the filter screen failure PSA modeling method and system for the passive nuclear power plant, which can fully reflect the design characteristics, filter screen types and arrangement, the types, the number and the migration paths of fragments under different accident working conditions and the like of the third-generation AP series passive nuclear power plant, effectively reduce the uncertainty brought by the original filter screen modeling method and obtain the risk insight more conforming to the actual power plant.
In order to achieve the above purpose, the present invention adopts the following technical scheme:
a first aspect of the present invention provides a method for modeling filter failure PSA in a passive nuclear power plant, comprising the steps of:
determining the total amount x of potential fragments of the nuclear power plant according to the inherent fragment amount of the nuclear power plant and the fragment amount in chemical reaction products after an expected accident;
analyzing migration paths of fragments according to the quantity and the types of fragments generated at different break positions in the nuclear power plant;
obtaining probability distribution of pit blocking under a certain accident type by utilizing a nuclear power plant equipment reliability database, and obtaining a process of transferring and accumulating fragments to a filter screen according to the relation between the filter screen failure rate r and the potential fragment quantity x, the break size y and the break position z of the nuclear power plant;
dividing working conditions according to the sizes of x, y and z, determining the value of the filter screen blockage failure rate under each working condition, and performing PSA modeling according to the value results under different working conditions.
Determining the total amount x of potential fragments of the nuclear power plant according to the inherent fragment amount of the nuclear power plant and the fragment amount in the chemical reaction product after the expected accident, wherein the total amount x is specifically as follows:
determining characteristics of potential fragments according to fragment samples in the nuclear power plant;
determining the fragment load rate of the surfaces of the structures in the running state of the pressurized water reactor according to the characteristics of potential fragments, the surface characteristics of the structures where the fragments in the containment are possibly deposited and the fragment samples, and obtaining the inherent fragment quantity of the nuclear power plant;
and determining the composition and the quantity of expected chemical precipitates of the nuclear power plant according to the chemical reactants after the accident to obtain the quantity of fragments in the chemical reaction products after the expected accident, and adding the quantity of fragments with the inherent quantity of fragments of the nuclear power plant to obtain the total quantity x of potential fragments of the nuclear power plant.
The method comprises the steps of analyzing a fragment migration path according to the quantity and the type of fragments generated at different break positions in a nuclear power plant, wherein the fragment migration path comprises the following specific steps:
the pressure vessel direct injection line breaks, producing a maximum amount of debris that migrates to the core;
the automatic pressure relief system pipeline at the top of the pressure stabilizer breaks, and the maximum amount of fragments migrating to the filter screen of the built-in refueling water tank is generated;
the core makeup tank inlet line breaks, producing a maximum amount of debris that migrates to the pit screen;
the filter failure rate r is related to the potential amount of fragments x, the break size y and the break position z of the power plant, i.e. r=f (x, y, z).
Dividing working conditions according to the sizes of x, y and z, and determining the value of the filter screen blocking failure rate under each working condition, wherein the value is specifically as follows:
when y is more than or equal to 0 and less than y1, y1 is a set first threshold value, and the working condition is a first type of working condition; the first working condition is a transient state with the least amount of generated fragments and a small crack loss type accident working condition of the coolant;
when y1 is less than or equal to y2, y2 is a set second threshold value, and the working condition is a second-class working condition; the second working condition is a medium break coolant loss accident and an automatic pressure relief system misoperation event working condition which generate medium fragment quantity;
when y2 is less than or equal to y3, y3 is a set third threshold value, and the working condition is a third type of working condition; the third type of conditions are large break coolant loss incidents that produce the greatest amount of debris, core makeup tank pipe break and safety injection line break conditions.
And determining the failure rate of the pit filter screen under three working conditions based on pit blockage failure data in the nuclear power plant operation database, and obtaining the multiplication factors of the failure rate of the filter screen under the three working conditions.
A second aspect of the invention provides a screen failure PSA modeling system for a passive nuclear power plant, comprising:
a potential debris source analysis unit configured to: determining the total amount x of potential fragments of the nuclear power plant according to the inherent fragment amount of the nuclear power plant and the fragment amount in chemical reaction products after an expected accident;
the crack position and fragment migration path analysis unit is configured to: analyzing migration paths of fragments according to the quantity and the types of fragments generated at different break positions in the nuclear power plant;
a screen and plug pattern analysis unit configured to: obtaining probability distribution of pit blocking under a certain accident type by utilizing a nuclear power plant equipment reliability database, and obtaining a process of transferring and accumulating fragments to a filter screen according to the relation between the filter screen failure rate r and the potential fragment quantity x, the break size y and the break position z of the nuclear power plant;
the working condition dividing and risk evaluating unit is configured to: dividing working conditions according to the sizes of x, y and z, determining the value of the filter screen blockage failure rate under each working condition, and performing PSA modeling according to the value results under different working conditions;
a PSA modeling unit configured to: and building house-shaped events to respectively represent three accident working conditions, and selecting corresponding filter screen blockage failure data by calling corresponding boundary conditions in each event tree to realize PSA modeling of filter screen blockage failure.
Compared with the prior art, the above technical scheme has the following beneficial effects:
1. the method combines the types, the quantity and the migration paths of fragments in the nuclear power plant under different accident working conditions, analyzes the application condition of a general database, and obtains a scheme for modeling the filter screen blocking failure probability under various accident working conditions, so that the contribution share of various accident working conditions to CDF is more reasonable, the influence of filter screen blocking failure on CDF is more consistent with the actual condition of the nuclear power plant, and the PSA analysis is facilitated to obtain more reasonable and reliable risk insight.
2. The method can fully reflect the design characteristics, the filter screen types and arrangement, the types, the number and the migration paths of fragments under different accident working conditions and the like of the third-generation AP series passive nuclear power plant, effectively reduce the uncertainty brought by the original filter screen modeling method, and obtain the risk insight which is more in line with the reality of the nuclear power plant.
Drawings
The accompanying drawings, which are included to provide a further understanding of the invention and are incorporated in and constitute a part of this specification, illustrate embodiments of the invention and together with the description serve to explain the invention.
FIG. 1 is a schematic illustration of a PSA modeling analysis of a nuclear power plant filter failure in accordance with one or more embodiments of the present invention;
FIG. 2 is a schematic diagram of a PS A fault tree model for pit screen plugging failure provided by one or more embodiments of the present invention;
fig. 3 is a graph of a fold amount x of debris and a filter failure rate r fitted according to one or more embodiments of the present invention.
Detailed Description
The invention will be further described with reference to the drawings and examples.
It should be noted that the following detailed description is exemplary and is intended to provide further explanation of the invention. Unless defined otherwise, all technical and scientific terms used herein have the same meaning as commonly understood by one of ordinary skill in the art to which this invention belongs.
It is noted that the terminology used herein is for the purpose of describing particular embodiments only and is not intended to be limiting of exemplary embodiments according to the present invention. As used herein, the singular is also intended to include the plural unless the context clearly indicates otherwise, and furthermore, it is to be understood that the terms "comprises" and/or "comprising" when used in this specification are taken to specify the presence of stated features, steps, operations, devices, components, and/or combinations thereof.
In the third generation AP series passive core plant design, in order to achieve adequate core cooling (including short-term water make-up and long-term cooling) in the event of an accident condition, the passive core cooling system (PXS) is required to be normally put into operation and to operate effectively in a long-term recirculation mode. In the process, various fragments can migrate to a pit filter screen or a built-In Refueling Water Storage Tank (IRWST) filter screen and gradually accumulate to form a fragment bed, so that the containment recirculation long-term cooling flow and the IRWST injection flow are reduced, and the core can lose sufficient cooling, thereby threatening the safety of the nuclear power plant.
For example, in a PSA model of the Western-style house AP1000, the blockage failure of a pit filter screen adopts a recommended value of 1.0E-05/h of blockage failure rate during the operation of a common filter in an advanced light water reactor user requirement file, while in a PSA model of a construction design stage of a domestic AP1000 power plant, because 3 filter screens corresponding to IRWST are provided with pipelines which are connected with each other, only 3 filter screens are blocked at the same time to cause filter screen blockage, so that the blockage failure rate of the IRWST filter screen is modeled as a basic event, and the basic event is connected into an IRWST subsystem fault tree, wherein the blockage failure probability value is 1.0E-05/d; similarly, since the pit is provided with a pipe interconnection for 2 screens, the pit recirculation screen blockage failure is modeled as a fundamental event, connected into the recirculation subsystem failure tree, and the failure probability value is taken to be 3.0E-05/d, considering that the amount of debris of the pit screen is much greater than the IRWST screen.
Thus, PSA results indicate that: the modeling method of filter screen blockage and the failure rate value thereof have great influence on the Core Damage Frequency (CDF) of the nuclear power plant.
Therefore, the following embodiment provides a filter screen failure PSA modeling method and system for a passive nuclear power plant, which can fully reflect design characteristics, filter screen types and arrangement, types, quantity, migration paths and the like of fragments under different accident working conditions of a third-generation AP series passive nuclear power plant, effectively reduce uncertainty brought by the original filter screen modeling method, and obtain risk insight which is more in line with the reality of the power plant.
Embodiment one:
as shown in fig. 1, the method for modeling the filter failure PSA of the passive nuclear power plant comprises the following steps:
determining the total amount x of potential fragments of the nuclear power plant according to the inherent fragment amount of the nuclear power plant and the fragment amount in chemical reaction products after an expected accident;
analyzing migration paths of fragments according to the quantity and the types of fragments generated at different break positions in the nuclear power plant;
obtaining probability distribution of pit blocking under a certain accident type by utilizing a nuclear power plant equipment reliability database;
analyzing the migration and accumulation process of fragments onto the filter screen according to the relation between the filter screen failure rate r and the potential fragment quantity x, the break size y and the break position z of the nuclear power plant and the probability distribution of pit blockage under a certain accident type; dividing working conditions according to the sizes of x, y and z, and determining the value of the filter screen blocking failure rate under each working condition;
and performing PSA modeling according to the value results under different working conditions.
Specific:
the following scheme is adopted in the embodiment to determine the filter screen modeling method:
in a first step, the sources of debris that may be generated by the power plant are evaluated.
The amount of potential debris of the power plant is set as variable x. During construction, testing and operation of a nuclear power plant, a portion of the fiber and particulate debris accumulates on the surface of the structure, and these "potential debris" characteristics can be determined from a sample test of debris in the operating power plant, and various surface debris loading rates can be determined from the surface characteristics of the structure (horizontal surfaces, vertical wall surfaces, equipment and piping, etc.) where debris may be deposited within the containment vessel, actual debris samples taken from similar surfaces of an existing operating nuclear power plant, and existing pressurized water reactor operating power plant interview observations.
In addition, it is also necessary to evaluate the fragments generated by the post-accident chemical reaction to obtain the composition and quantity of the expected chemical precipitations of the power plant. The amount of inherent fragments of the power plant is added to the amount of chemical sediment after the expected accident to obtain the total amount of potential fragments of the power plant.
And secondly, evaluating the quantity, the type and the migration path of fragments generated at different break positions in the power plant.
The size of a break of the power plant is set as a variable y, and the position of the break is set as a variable z. Due to the unique design of third generation AP series nuclear power plants, the number and variety of fragments generated at different break locations of a loss of coolant accident (LOCA) are substantially the same. Reasons include:
a) The AP1000 does not use conventional insulation (such as fiberglass) that is easily damaged by LOCA spray;
b) At the location where LOCA spray may occur, the AP1000 uses a metal reflective insulation (RMI);
c) The density of the RMI material ensures that any debris from its damage will settle to the bottom of the pit and not migrate to the screen. However, at different break locations, the amount of debris reaching the pit screen or the built-In Refueling Water Storage Tank (IRWST) screen varies due to the different path of migration of the debris and the different characteristics of the different debris as the recirculating water flows.
Through investigation and analysis, the maximum amount of fragments migrating to the pit screen can be generated by the broken inlet pipeline of the core makeup tank, because water can flow out of the core makeup tank and downwards into a PXS (passive core cooling) system compartment where the pit screen is located; the fracturing of the regulator top 1, 2, 3 stage automatic pressure relief system lines will produce the maximum amount of debris migrating to the IRWST screen, and the loop compartment breach (reactor pressure vessel direct injection line breach, etc.) will produce the maximum amount of debris directed to the core, as the breach may be submerged causing some debris to bypass the recirculation screen directly into the core.
From the above analysis, it can be deduced that the filter failure rate r is related to the potential amount of fragments x, the break size y and the break position z of the power plant, i.e. r=f (x, y, z). The evaluation lays a foundation for working condition division of filter screen blockage modeling.
And thirdly, determining the applicability of the data according to a probability safety evaluation (PSA) related general database and the data adopted by domestic and foreign operating power plants.
By researching the modeling conditions of the pit filter screens in a large number of power plant PSA models, the fact that the PSA of most power plants does not select different filter screen blocking failure parameters according to different accident conditions at present, but adopts the same pit filter screen blocking failure parameters under various accident conditions is known.
Sources include: data in French EDF handbook, NUREG/CR-4550, NUREG/CR-6928 (2015 edition), etc. The analysis calculation example given by the document NUREG/CR-6771 divides accident conditions into 4 types of conditions, namely LLOCA, MLOCA & SLOCA_3, SLOCA_2 and transient, analyzes accident process parameters of 69 nuclear power plants, and arranges the accident types of LLOCA, MLOCA & SLOCA_3, SLOCA_2 and transient according to the occurrence probability of pit blockage from large to small.
And fourthly, comprehensively considering the potential debris amount x, the break size y, the break position z and other factors of the power plant, and determining the PSA (pressure swing adsorption) partial condition modeling method for the filter screen blockage failure.
Analyzing the process of how various fragments migrate and accumulate on a filter screen through the types of fragments, the amounts of fragments, the migration conditions of fragments and the like possibly generated by the power plant under different accident conditions, and classifying the initial events into three categories according to the amount of the fragments accumulated on the filter screen after each initial event possibly generated by the power plant occurs:
when y is small or 0, the x-quantity is considered to contain only the inherent power plant debris quantity, the amount of debris generated by chemical precipitation is negligible, and the influence of the variable z is also negligible. The working conditions are recorded as a first type of working conditions, and correspond to the working conditions such as transient state with minimum debris quantity, small crack loss of coolant accident (SLOCA) and the like;
when y is increased to a certain degree, x is also increased rapidly due to the increase of the generation amount of the sediment in the chemical reaction; the influence of the variable z increases synchronously, but in PSA analysis the influence of the break position z is estimated as the most conservative case. The working conditions are recorded as second working conditions, and correspond to the working conditions of medium-fragment-quantity medium-break coolant loss accidents (MLOCAs), automatic pressure relief system malfunction events (SPADS) and the like;
in addition, the y value corresponding to the working condition with the largest z contribution may be in a middle range, but is also conservatively classified into a third class of working conditions, and the working conditions correspond to the working conditions such as large-break coolant loss accident (LLOCA), core makeup tank pipeline rupture (CMTLB), injection pipeline rupture (SI-LB) and the like which can generate the largest fragment amount.
Fifthly, determining the failure rate of filter screen blockage under each working condition.
Because rare accidents such as LLOCA and MLOCA do not occur in the operation history of the nuclear power plant, the pit blockage failure data counted in the NUREG/CR-6928 general database is considered to be only applicable to working condition 1, namely SLOCA and transient events; in addition, as the 2 pit filter screens of the AP series passive nuclear power plant are provided with the pipeline interconnection, the pit blockage can be caused only if 2 filter screens are blocked simultaneously, so that the filter screen blockage failure is modeled as a basic event. The value of this basic event (Q 1 ) Based on pit blocking failure data 3.42E-07/h counted in 2015 NUREG/CR-6928 general database, and considering 24 hours task time, the failure rate of the pit filter screen under the first working condition is calculated as follows:
Q 1 ＝γ sump ×24h＝3.42E-07/h×24h＝8.2E-06/d。
under the accident condition that the coolant is lost due to the pipeline break, high-energy injection, the damp and hot environment in the containment, chemical action and the like, the quantity of fragments possibly generated in the containment and migrated to the filter screen is far higher than that under the normal operation condition, and the conservatively estimated maximum fragment quantity under the large break can reach about 3 times of that under the normal operation condition.
The failure rate of the filter screen is related to x, y and z, the influence of y and z is comprehensively considered under the working condition determined in the step 4, the amount of fragments in the first working condition is set to be 1 time, the amount of fragments under other working conditions is set to be x, the failure rate of the filter screen is r, and the relation between r and x is obtained through fitting as shown in figure 3.
According to investigation statistics, the amount of fragments under the second type of working conditions is about 2 times of that under the first type of working conditions, the amount of fragments under the third type of working conditions is about 3 times of that under the first type of working conditions, and according to the fitting formula, the recommendation of the multiplication factor of the failure rate of the filter screen under the following 3 types of working conditions is shown in a table 1.
Table 1: filter screen failure rate multiplication factor recommendation under class 3 working conditions
| Working conditions of | Multiplication factor of failure rate of filter screen |
| First category: SLOCA and |
1 |
| The second category: accident conditions such as MLOCA and SPADS | 5 |
| Third category: accident conditions of LLOCA, CMTLB and SILB | 10 |
For a containment in-tank refueling water tank (IRWST) filter, since in the third generation AP-series passive nuclear power plant, there is one separate filter placed within the IRWST on each of the two IRWST injection lines. The other filter screen is connected with the two filter screens through the pipeline, so that even if one filter screen is blocked, the other two filter screens can be used and connected with the two IRWST injection pipelines through the pipeline, and the possibility of failure caused by blockage of the filter screens is greatly reduced. For the failure probability of the IRWST filter screen, according to analysis, the total amount of fragments reaching the IRWST filter screen is about one fourth of that of the pit filter screen, and certain uncertainty exists in the data; in addition, in the AP series passive nuclear power plant, three filter screens are arranged in the IRWST, and only 3 filter screens are blocked at the same time to cause filter screen blocking, so that after comprehensive consideration and conservation allowance are taken, the blocking failure of all the three IRWST filter screens is also modeled as a basic event, and the probability of the blocking failure is taken as one third of the failure probability of the pit filter screen.
And sixthly, performing PSA modeling, calculating the influence of the PSA modeling on the power plant risk, and proving the rationality of the modeling mode.
The fault tree model is shown in fig. 2, taking a pit filter screen as an example, building house-shaped events HS-MLOCA and HS-LLOCA to represent second-class (MLOCA, SPADS, etc.) and third-class (LLOCA, CMTLB, SILB, etc.) accident conditions respectively, and the two house-shaped events are both inverted to represent other accident conditions such as SLOCA, transient, etc. And in each event tree, the corresponding filter screen blockage failure data is selected by calling the corresponding boundary conditions.
The embodiment comprehensively considers the difference between the third-generation AP series passive nuclear power plant and the second-generation pressurized water reactor, combines the types, the number and the migration paths of fragments in the power plant under different accident conditions, analyzes the application condition of a general database, and finally provides a scheme for modeling the filter screen blockage failure probability under various accident conditions. The contribution share of various accident conditions to CDF is more reasonable, the influence of filter screen blockage failure on CDF is more in line with the actual power plant, and the PSA analysis is facilitated to obtain more reasonable and reliable risk insight.
Embodiment two:
a screen failure PSA modeling system for a passive nuclear power plant, comprising:
a potential debris source analysis unit configured to: determining the total amount x of potential fragments of the nuclear power plant according to the inherent fragment amount of the nuclear power plant and the fragment amount in chemical reaction products after an expected accident;
the crack position and fragment migration path analysis unit is configured to: analyzing migration paths of fragments according to the quantity and the types of fragments generated at different break positions in the nuclear power plant;
a screen and plug pattern analysis unit configured to: obtaining probability distribution of pit blocking under a certain accident type by utilizing a nuclear power plant equipment reliability database, and obtaining a process of transferring and accumulating fragments to a filter screen according to the relation between the filter screen failure rate r and the potential fragment quantity x, the break size y and the break position z of the nuclear power plant;
the working condition dividing and risk evaluating unit is configured to: dividing working conditions according to the sizes of x, y and z, determining the value of the filter screen blockage failure rate under each working condition, and performing PSA modeling according to the value results under different working conditions;
a PSA modeling unit configured to: and building house-shaped events to respectively represent three accident working conditions, and selecting corresponding filter screen blockage failure data by calling corresponding boundary conditions in each event tree to realize PSA modeling of filter screen blockage failure.
Various modifications and variations of the present invention will be apparent to those skilled in the art. Any modification, equivalent replacement, improvement, etc. made within the spirit and principle of the present invention should be included in the protection scope of the present invention.
Claims (10)
1. The filter screen failure PSA modeling method of the passive nuclear power plant is characterized by comprising the following steps of:
determining the total amount x of potential fragments of the nuclear power plant according to the inherent fragment amount of the nuclear power plant and the fragment amount in chemical reaction products after an expected accident;
analyzing migration paths of fragments according to the quantity and the types of fragments generated at different break positions in the nuclear power plant;
obtaining probability distribution of pit blocking under a certain accident type by utilizing a nuclear power plant equipment reliability database, and obtaining a process of transferring and accumulating fragments to a filter screen according to the relation between the filter screen failure rate r and the potential fragment quantity x, the break size y and the break position z of the nuclear power plant;
dividing working conditions according to the sizes of x, y and z, determining the value of the filter screen blockage failure rate under each working condition, and performing PSA modeling according to the value results under different working conditions.
2. The method for modeling filter screen failure PSA of a passive nuclear power plant as claimed in claim 1, wherein the total amount x of potential fragments of the nuclear power plant is determined based on the inherent fragment amount of the nuclear power plant and the amount of fragments in the chemical reaction product after the expected accident, specifically:
determining characteristics of potential fragments according to fragment samples in the nuclear power plant;
and determining the fragment load rate of the surfaces of the structures in the running state of the pressurized water reactor according to the characteristics of the potential fragments, the surface characteristics of the structures where the fragments are possibly deposited in the containment vessel and the fragment samples, and obtaining the inherent fragment quantity of the nuclear power plant.
3. The method for modeling filter screen failure PSA of a passive nuclear power plant of claim 2, wherein determining a total amount x of potential fragments of the nuclear power plant based on the amount of inherent fragments of the nuclear power plant and the amount of fragments in the chemical reaction product after the expected accident, further comprises:
and determining the composition and the quantity of expected chemical precipitates of the nuclear power plant according to the chemical reactants after the accident to obtain the quantity of fragments in the chemical reaction products after the expected accident, and adding the quantity of fragments with the inherent quantity of fragments of the nuclear power plant to obtain the total quantity x of potential fragments of the nuclear power plant.
4. The method for modeling filter failure PSA of a passive nuclear power plant according to claim 1, wherein the migration path of the generated fragments according to the number and types of fragments at different break positions in the nuclear power plant is specifically:
the pressure vessel direct injection line breaks, producing a maximum amount of debris that migrates to the core;
the automatic pressure relief system pipeline at the top of the pressure stabilizer breaks, and the maximum amount of fragments migrating to the filter screen of the built-in refueling water tank is generated;
the core makeup tank inlet line breaks, producing a maximum amount of debris that migrates to the pit screen;
the filter failure rate r is related to the potential amount of fragments x, the break size y and the break position z of the power plant, i.e. r=f (x, y, z).
5. The method for modeling filter screen failure PSA of a passive nuclear power plant according to claim 1, wherein the values of filter screen blockage failure rates under the working conditions are determined by dividing the working conditions according to the sizes of x, y and z, specifically:
when y is more than or equal to 0 and less than y1, y1 is a set first threshold value, and the working condition is a first type of working condition;
when y1 is less than or equal to y2, y2 is a set second threshold value, and the working condition is a second-class working condition;
when y2 is less than or equal to y3, y3 is a set third threshold, and the working condition is a third type of working condition.
6. The method for modeling filter screen failure PSA of a passive nuclear power plant of claim 5, wherein the first type of condition is a transient and small break coolant loss type of accident condition that produces minimal amounts of debris.
7. The method for modeling filter screen failure PSA in a passive nuclear power plant of claim 5, wherein said second type of condition is a medium break coolant loss event and automatic pressure relief system malfunction event condition that generates a medium amount of debris.
8. The method for modeling filter screen failure PSA of a passive nuclear power plant of claim 5, wherein the third type of condition is a large breach coolant loss event, core makeup tank pipe breach, and safety injection line breach condition that produces a maximum amount of debris.
9. A screen failure PSA modeling system for a passive nuclear power plant, comprising:
a potential debris source analysis unit configured to: determining the total amount x of potential fragments of the nuclear power plant according to the inherent fragment amount of the nuclear power plant and the fragment amount in chemical reaction products after an expected accident;
the crack position and fragment migration path analysis unit is configured to: analyzing migration paths of fragments according to the quantity and the types of fragments generated at different break positions in the nuclear power plant;
a screen and plug pattern analysis unit configured to: obtaining probability distribution of pit blocking under a certain accident type by utilizing a nuclear power plant equipment reliability database, and obtaining a process of transferring and accumulating fragments to a filter screen according to the relation between the filter screen failure rate r and the potential fragment quantity x, the break size y and the break position z of the nuclear power plant;
the working condition dividing and risk evaluating unit is configured to: dividing working conditions according to the sizes of x, y and z, determining the value of the filter screen blockage failure rate under each working condition, and performing PSA modeling according to the value results under different working conditions.
10. The passive nuclear power plant filter screen failure PSA modeling system of claim 9, further comprising a PSA modeling unit configured to: and building house-shaped events to respectively represent three accident working conditions, and selecting corresponding filter screen blockage failure data by calling corresponding boundary conditions in each event tree to realize PSA modeling of filter screen blockage failure.
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