CN117076434A - EPR nuclear power unit accident initiation event list optimization method and system - Google Patents
EPR nuclear power unit accident initiation event list optimization method and system Download PDFInfo
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Abstract
The invention relates to a method and a system for optimizing an accident initiation event list of an EPR nuclear power unit, wherein the method comprises the following steps: collecting an initial event list of each nuclear power unit; performing comparison analysis on each initial event list, and determining an optimization object of an accident initial event list of the EPR nuclear power unit according to comparison results; performing frequency analysis on accident-originating events in the optimization object; reclassifying the accident by combining the EPR accident classification standard and the frequency analysis result; optimizing the initial event list according to the classification result; and generating an optimized originating event list. By implementing the invention, after the accident initiation event list is optimized for the EPR nuclear power unit, the EPR accident analysis and the technical specification of the operation of the EPR nuclear power unit are optimized, the unit is maintained on line, and the economic benefit of the nuclear power plant is further improved on the premise of ensuring the safety.
Description
Technical Field
The invention relates to the field of analysis of nuclear power unit originating events, in particular to an EPR nuclear power unit accident originating event list optimization method and system.
Background
The VVER nuclear power unit designed for four rows of safety systems can realize online maintenance of one row of safety systems and effectively reduce the workload in overhaul.
According to current programming principles for EPR operating specifications, a nuclear power plant Final Safety Analysis Report (FSAR) provides an overall input of safety requirements for the programming of operating specifications. Therefore, on the premise of ensuring safety, in order to improve the flexibility of unit operation and realize on-line maintenance of more system equipment so as to shorten the overhaul period, deep comparison, analysis and research are needed for the FSAR accident analysis of the EPR, the relation between the accident analysis and the EPR operation technical specification programming is further clarified, and the technical specification programming methodology is optimized to realize on-line maintenance so as to improve the operation performance level of the power station.
Disclosure of Invention
The technical problem to be solved by the present invention is to address at least one of the drawbacks of the related art mentioned in the background art mentioned above: and according to the optimized initial event list, performing optimal EPR accident analysis and EPR nuclear power unit operation technical specification, and providing an EPR nuclear power unit accident initial event list optimization method and system.
The technical scheme adopted for solving the technical problems is as follows: the invention provides an EPR nuclear power unit accident initiation event list optimization method, which comprises the following steps:
step S10: collecting an initial event list of each nuclear power unit;
step S20: performing comparison analysis on each initial event list, and determining an optimization object of the EPR nuclear power unit accident initial event list according to comparison results;
step S30: performing frequency analysis on the accident-originated event in the optimization object;
step S40: reclassifying the accident by combining the EPR accident classification standard and the frequency analysis result;
step S50: optimizing the originating event list according to the classification result;
step S60: and generating the optimized originating event list.
Preferably, the step S10 further includes:
acquiring the design safety requirement of the EPR nuclear power unit;
the step S20 further includes:
dividing the operation range of the EPR nuclear power unit into a plurality of states according to the design safety requirements, transversely comparing an initial event list of the EPR nuclear power unit with initial event lists of other nuclear power units, identifying difference items among the initial event lists, and selecting a corresponding initial event from the plurality of states as an optimization object of the accident initial event list according to an identification result.
Preferably, the incidents include non-isolatable and isolatable water loss incidents;
the step S30 includes:
and carrying out frequency analysis on the non-isolatable water loss accident and the isolatable water loss accident in the corresponding originating event to obtain the frequency analysis result of the corresponding originating event.
Preferably, the unable to isolate the loss of water accident comprises: the reactor coolant system pipeline breaks and the pressure stabilizer safety valve is blocked;
the isolatable water loss event comprises: and the pipeline of the safety shell inner waste heat discharging system is broken and the pipeline of the safety shell outer waste heat discharging system is broken.
Preferably, the step S20 further includes:
step S21: and analyzing the frequency source of the accident originating event in the optimization object, and executing the step S30 if the frequency source is judged to be not in accordance with the preset standard.
Preferably, the frequency source comprises: small break accidents can not be isolated in the containment and break openings which can be isolated by a waste heat discharging system inside and outside the containment.
The invention also constructs an EPR nuclear power unit accident initiation event list optimizing system, which comprises:
the acquisition module is used for acquiring an initial event list of each nuclear power unit;
the comparison module is used for carrying out comparison analysis on each initial event list and determining an optimization object of the EPR nuclear power unit accident initial event list according to comparison results;
the frequency analysis module is used for carrying out frequency analysis on the accident-originating event in the optimization object;
the classification module is used for reclassifying the accidents by combining the EPR accident classification standard and the frequency analysis result;
the optimizing module is used for optimizing the initial event list according to the classification result;
and the result generation module is used for generating the optimized initial event list.
Preferably, the acquisition module is further configured to:
acquiring the design safety requirement of the EPR nuclear power unit;
the contrast module is further to:
dividing the operation range of the EPR nuclear power unit into a plurality of states according to the design safety requirements, transversely comparing an initial event list of the EPR nuclear power unit with initial event lists of other nuclear power units, identifying difference items among the initial event lists, and selecting a corresponding initial event from the plurality of states as an optimization object of the accident initial event list according to an identification result.
The present invention also constructs an electronic apparatus, characterized by comprising:
one or more processors;
storage means for storing one or more programs which, when executed by the one or more processors, cause the one or more processors to implement the EPR nuclear power unit incident initiation event manifest optimization method as defined in any one of the preceding claims.
The present invention also constructs a storage medium having a computer program stored thereon, characterized in that the computer program, when executed by a processor, implements the EPR nuclear power unit incident initiation event list optimization method as described in any of the above.
By implementing the invention, the following beneficial effects are achieved:
the invention discloses an EPR nuclear power unit accident initiation event list optimization method and system. After the initial event list of the EPR nuclear power unit is optimized by the method, the EPR accident analysis and the EPR nuclear power unit operation technical specification are optimized, the unit is maintained on line, and the economic benefit of the nuclear power plant is further improved on the premise of ensuring the safety.
Drawings
The invention will be further described with reference to the accompanying drawings and examples, in which:
FIG. 1 is a flow chart of the EPR nuclear power unit accident onset event list optimization method of the present invention;
FIG. 2 is a schematic diagram of a flow chart of calculation of the stuck open frequency of a safety valve of the voltage regulator of the present invention;
FIG. 3 is a block diagram of a module of the EPR nuclear power unit incident initiation event list optimization system of the present invention.
Detailed Description
For a clearer understanding of technical features, objects and effects of the present invention, a detailed description of embodiments of the present invention will be made with reference to the accompanying drawings.
It should be noted that the flow diagrams depicted in the figures are merely exemplary and do not necessarily include all of the elements and operations/steps, nor are they necessarily performed in the order described. For example, some operations/steps may be decomposed, and some operations/steps may be combined or partially combined, so that the order of actual execution may be changed according to actual situations.
The block diagrams depicted in the figures are merely functional entities and do not necessarily correspond to physically separate entities. That is, the functional entities may be implemented in software, or in one or more hardware modules or integrated circuits, or in different networks and/or processor devices and/or microcontroller devices.
Wherein abbreviations and key terms of the present invention are defined as:
DBC: designing a reference working condition; DEC: designing an expansion working condition; EPR: advanced pressurized water reactor in Europe; FSAR: final security analysis report; PWR: pressurized water reactor; LOCA: a loss of water accident; RCP: a reactor coolant system; RCV: a reaction and volume control system; RHR: a waste heat removal system; RIS: a safe injection system.
In this embodiment, as shown in fig. 1, the invention provides a method for optimizing an accident initiation event list of an EPR nuclear power unit, which comprises the following steps:
step S10: collecting an initial event list of each nuclear power unit;
step S20: performing comparison analysis on each initial event list, and determining an optimization object of the EPR nuclear power unit accident initial event list according to comparison results;
step S30: performing frequency analysis on accident-originating events in the optimization object;
step S40: reclassifying the accident by combining the EPR accident classification standard and the frequency analysis result;
step S50: optimizing the initial event list according to the classification result;
step S60: and generating an optimized originating event list.
Specifically:
in this embodiment, step S10 further includes:
acquiring the design safety requirement of the EPR nuclear power unit;
step S20 further includes:
dividing the operation range of the EPR nuclear power unit into a plurality of states according to design safety requirements, transversely comparing an initial event list of the EPR nuclear power unit with initial event lists of other nuclear power units, identifying difference items among the initial event lists, and selecting corresponding initial events from the plurality of states as optimization objects of the accident initial event list according to identification results.
In the safety analysis of the probability of a reactor, an initial event is defined as a disturbance which occurs in the reactor and can potentially cause the damage of the reactor core, the current three-generation pressurized water reactor nuclear power units at home and abroad mainly comprise AP1000, EPR, hualong No. one and the like, the initial event list of the three nuclear power units has differences, and the differences of the EPR nuclear power unit and the initial event list of other three-generation nuclear power units and the pressurized water reactor nuclear power plant are determined through data analysis.
The main data to be collected include: EPR design safety requirements, an AP1000 initial event list, a Hua-long first nuclear power unit initial event list, an initial event frequency international general data source, pressurized water reactor nuclear power plant shutdown condition safety analysis and commercial nuclear power plant pipeline failure frequency.
In this embodiment, the EPR unit operating range is divided into states A, B, C, D, E and F according to EPR design safety requirements. The specific partitioning of the 6 states is shown in table 1.
Table 1: EPR nuclear power unit operating range
The design requirement of a newly added originating event (related accident of a pool containing spent fuel) is determined by analyzing an upstream design source (French regulation or supervision requirement) of the originating event of the EPR unit, and the newly added originating event is transversely compared with the originating event of the nuclear power unit such as AP1000, hualong No. one and the like, the difference item of an originating event list is identified, and an optimized originating event list is analyzed. By adopting an engineering judgment method, referring to the initial event list of three generations of pressurized water reactors at home and abroad, such as the initial event list of EPR power plants (US EPR and Finland OL3 power plants), three generations of other power plants (Fuqing Hualong, AP 1000) and the like, the optimized initial event list is analyzed. The EPR nuclear power unit initiation event list and other types of PWR nuclear power comparison analysis results are shown in Table 2.
Table 2: EPR nuclear power unit and other unit originating event list comparison
As can be seen by comparison, the FSAR originated event list of the EPR unit is more detailed than that of the Hualong unit and the AP1000 unit. Compared with a typical three-generation pressurized water reactor nuclear power unit, the EPR nuclear power unit is added with small break LOCA (less than or equal to DN 50) in a state C, D, E, including the break of an injection pipeline of an emergency boride system; under state C, D, E, the internal (external) waste heat removal system pipeline of the containment is broken (isolatable) (. Ltoreq.DN 250); in terms of spent fuel pool drainage in state E, either a small break (< 50 mm) that is not isolatable or a RIS break (< 250 mm) that is isolatable in RHR mode.
The accident of the spent fuel pool is simpler than the accident mechanism of the reactor core, and the accident process is relatively slow. Under the conditions of non-break and break accidents, when the cooling of the spent fuel pool is completely lost, the spent fuel pool is boiled and exposed for a long time under different working conditions. The identified spent fuel pool is therefore not suitable for the analysis presented in this patent. In contrast to other third generation nuclear reactor types, such as AP1000, middle-core version No. huaron, no break-type initiation event is considered during shutdown. The low-mode primary loop and the small break accident of the pipeline connected with the low-mode primary loop of the EPR nuclear power unit are respectively listed as design reference working conditions, and compared with other three-generation type nuclear power units, the working conditions are strictly divided, so that the operation and maintenance management of the EPR nuclear power unit after operation by a manufacturer is not facilitated. Therefore, LOCA accidents in the state C, D, E are regarded as optimization targets.
As can be seen by comparing with the initial event list of a typical pressurized water reactor nuclear power unit, the LOCA accident is increased in the state C, D, E of the EPR nuclear power unit and other third-generation nuclear power units, and the LOCA accident comprises a small break accident which cannot be isolated in a containment and a break which can be isolated by a residual heat removal system inside and outside the containment.
Further, the incidents include non-isolatable and isolatable water loss incidents; for the small break accident which cannot be isolated in the containment of the EPR nuclear power unit under the state C, D, E and the break which can be isolated by the internal and external waste heat discharging system of the containment, the originating event frequency can be determined according to the analysis methods of the LOCA which cannot be isolated and the originating event of the LOCA accident which can be isolated.
Step S30 includes:
and carrying out frequency analysis on the non-isolatable water loss accident and the isolatable water loss accident in the corresponding originating event to obtain a frequency analysis result of the corresponding originating event.
Wherein, unable isolation water loss accident includes: the reactor coolant system pipeline breaks and the pressure stabilizer safety valve is blocked;
the isolatable water loss event comprises: and the pipeline of the safety shell inner waste heat discharging system is broken and the pipeline of the safety shell outer waste heat discharging system is broken.
The obtained safety analysis data of the shutdown condition of the pressurized water reactor nuclear power plant is used for the analysis, and LOCA accident analysis of the EPR nuclear power unit under the state C, D, E is as follows:
(1) Unable to isolate LOCA
In the state C, D, E of the EPR nuclear power unit, under the shutdown condition that an injection and residual heat removal system (RIS) is connected in a Residual Heat Removal (RHR) mode, a small crack loss accident that a crack is positioned in a reactor coolant system (RCP) or an emergency boride system causes the loss of coolant and the pressure drop of the reactor coolant system (RCP). The accident analyzed was caused by an inexpensible break comprising a break of the injection line of the emergency boride system (RBS), with an equivalent diameter less than or equal to DN50 (equivalent area less than or equal to 20cm 2).
The non-isolatable LOCA includes: the RCP pipeline breaks (to a second valve of a related system) and the safety valve of the voltage stabilizer is blocked to form a small break which can not return to the seat.
Wherein RCP pipe rupture is expressed as: according to the international general data source, the typical frequency value of the PWR small break accident under the power working condition is 5E-04 per year. During shutdown conditions, the frequency of rupture of the RCP pipe is lower than under power conditions because of the lower stress. In mode Ca, the LOCA accident frequency of the RCP pipeline according to the study is 1/28 of the power condition. In states Cb, D, E, where the primary circuit is already at or near atmospheric pressure, the RCP system pipe is less likely to have LOCA accidents and therefore the frequency of LOCA accidents in the primary circuit pipe is no longer calculated.
In the state Ca, the frequency of LOCA occurrence in the one-circuit pipeline can be calculated by the following equation 1:
wherein F is A For the next loop pipeline of state AThe frequency of the raw LOCA; tc (tc) a The running time of the EPR unit under the state Ca; t is t A The running time of the EPR unit in the A state is shown.
The frequency of small cracks of RCP under the state Ca is calculated to be 1.24E-07/year of pile according to the operation time of the EPR nuclear power unit under each state in the table 1 and the formula 1. Whereas for states Cb, D and E, stress analysis considers that RCP is already at or near atmospheric pressure in this state, the RCP system piping is less likely to experience LOCA accidents.
The safety valve of the voltage stabilizer is clamped and opened and expressed as: the EPR nuclear power unit is provided with three pressure relief pipelines at the top of the voltage stabilizer, and each pipeline is provided with a safety valve. When the RHR is connected, the pressure regulator relief valve provides a circuit cold overpressure protection to prevent the pressure vessel cylinder from failing due to brittle fracture at low temperatures before reaching the design pressure of the coolant system. When the loop pressure exceeds the regulator relief valve setting, the regulator relief valve will open. If the regulator relief valve is stuck in an open position, a loop LOCA event will be generated through the relief valve. A loop coolant will enter the regulator pressure relief tank through the regulator pressure relief tube. The temperature, pressure and level indications in the pressure regulator relief tank will alert the operator to open the valve catch and if the operator closes the shut-off valve, the LOCA will terminate.
The LOCA frequency caused by the stuck-open regulator relief valve can be quantified by the event tree in fig. 2. In state C, D, E, after the RHR system is put into operation, the RCP system pressure protection is completed by the RHR safety valve. Before RHR is disconnected, the RCP system pressure protection is completed with a regulator relief valve. Therefore, only LOCA caused by the opening of the surge relief valve in the state Ca1 needs to be considered.
Under the state Ca1, the frequency of accidental opening of the safety valve of the voltage stabilizer is 1.79E-06/year, the probability of incapability of closing the safety valve of the voltage stabilizer after opening is 3.5E-05, the probability of not closing the stop valve by an operator is 1E-02, and the probability of failure of closing the stop valve is 1.51E-05. Thus, in state Ca1, the regulator relief valve snaps open resulting in a LOCA frequency of 1.89E-12/year of pile.
(2) Isolatable breach
In the normal operating states C and D, the power plant is cooled by a safety injection and residual heat removal system (RIS) in Residual Heat Removal (RHR) mode. A circuit coolant is drawn from a reactor cooling system (RCP) hot leg by a low pressure ampere-injection (LHSI) pump, cooled, and finally injected into the RCP cold leg. The cause of the accident is due to the occurrence of an isolatable break in one RIS line during RHR mode operation, the break being of a size less than or equal to DN250. The isolatable breach inside the containment may be located:
downstream of the RIS isolation valve second closest to the RCP, on the RIS drain line connected to the RCP hot leg, or upstream of the check valve second closest to the RCP, on the RIS fill line connected to the RCP cold leg.
The isolated break outside the containment refers to an accident that the system is connected with the RCP system to cause loss of reactor coolant, and is a type of accident that can cause overpressure and rupture of the system connected with the reactor coolant outside the containment. Loss of reactor coolant accidents due to interface systems have been a concern in terms of public health risks, as radioactive fission products may bypass containment structures and be released directly into the environment.
The isolatable breach considered in the accident analysis at EPR nuclear power unit condition C, D, E includes an inner containment RHR system pipeline breach and an outer containment RHR system pipeline breach. The frequency of the initiation event for the two types of isolatable breach can be calculated from the frequency of pipeline rupture of the RHR system.
The obtained commercial nuclear power plant pipeline failure frequency data is used for the use according to international data sources, the pipeline rupture frequency of the safety related system is 1.92E-10/section hour, and according to statistics, the RHR system has 42 sections of pipelines in the containment and 47 sections of pipelines outside the containment. The RHR system breaks within the containment at a frequency of 8.06E-09/hour and breaks outside the containment at a frequency of 9.02E-09/hour.
According to table 1, the frequency of cracking the inner and outer pipelines of the containment vessel of the RHR system can be calculated, and then the frequency of cracking the inner pipeline of the containment vessel of the RHR system in state C, D, E is respectively as follows: 8.95E-07/year of heap, 3.55E-07/year of heap, 9.75E-07/year of heap. The frequency of cracking the outer tube of the RHR system containment at state C, D, E is: 1.00E-06/year of the pile, 3.95E-07/year of the pile, 1.09E-06/year of the pile.
The system connected to the RCP system results in that LOCA cannot estimate the frequency of occurrence as a single originating event, but rather the evaluation is carried out in a series of contexts. These events take into account factors such as the likelihood of high system pressures, the likelihood of rupture, and operator recovery. It is expected that the isolatable LOCA frequency of an EPR nuclear power plant unit in state C, D, E will be less than 1E-06 per year of the stack.
In the present embodiment, step S20 further includes:
step S21: and (3) analyzing the frequency source of the accident-originated event in the optimization object, and if the frequency source is judged to be not in accordance with the preset standard, executing step S30.
In addition, in the present embodiment, the frequency sources include: a small break accident can not be isolated in the containment and a break which can be isolated by a waste heat discharging system inside and outside the containment;
small break accidents cannot be isolated in the containment:
the occurrence frequency of the small break accident of the EPR nuclear power unit in the state C, D, E is calculated to give LOCA frequency values in different states based on international general data. The LOCA event initiation event frequency value is mainly given in combination with the pipe size and the run time under shutdown conditions. The temperature and pressure conditions are used as key causes of pipeline breakage, and are not considered as variables affecting the accident occurrence frequency. The conservation degree of calculation of the shutdown breach frequency according to time halving based on the NUREG-5750 database is higher.
The obtained source of the global universal data of the originating event frequency is used for this, the median value of the originating event frequency value of small LOCA is 5E-04/year in NUREG-5750 under the power condition, and the estimated value of the originating event frequency of small LOCA given in the reference power plant is less than 5E-04/year.
For small LOCA incidents including boron implanted tubing at states C, D and E, the frequency of the occurrence is primarily given in terms of LOCA frequency values given in NUREG-5750 in combination with tubing size and run time. The frequency of small LOCA incidents in the full power state is 5E-04 per year, these pipeline operation events account for about 1% per year, so the small LOCA frequency in EPR nuclear power unit states C, D and E is 5E-06 per year or less. In practice, however, the small LOCA frequency will be significantly reduced at states C, D and E because the one-circuit pressure boundary pressure is much reduced compared to the full power state.
Break that can keep apart of the inside and outside waste heat discharge system of containment:
the break event which can be isolated by the internal and external waste heat discharge system of the containment of the EPR nuclear power unit under the state C, D, E is determined to be a DBC-4 event according to qualitative judgment, and the occurrence frequency of the break event is not determined.
The accident condition refers to a condition which deviates from normal operation and is lower in frequency but more serious than the expected operation event. Accident conditions include design base accident and (DBC) Design Extension Conditions (DEC). Design basis accidents refer to hypothetical accidents that lead to accident conditions in a nuclear power plant, where the radioactive emissions of these accidents are within acceptable limits. The design expansion working condition refers to an accident working condition which is not in the design reference accident consideration range, the accident working condition is considered according to the optimal estimation method in the design process, and the radioactive substance release of the accident working condition is within the acceptable limit value. Design extension conditions include conditions that do not cause significant core damage and core fusion (severe accident) conditions.
The EPR nuclear power unit selects the events according to the potential risks caused by the accidents on the following main safety functions, and selects the corresponding accidents as the object of accident analysis: reactivity and power control, heat removal from the fuel element, and radioactivity containment.
The design reference accidents are divided into four types of design reference working conditions according to the annual occurrence frequency of the accidents, and four types of DBC accident division criteria are as follows:
DBC-1: normal operation related transients; DBC-2: expected operational event (10-2/year of heap < f); DBC-3: rare accidents (10-4/year of the stack < f < 10-2/year of the stack); DBC-4: extreme incidents (10-6/year < f < 10-4/year). Note f-generation event/accident frequency.
The DEC accident is divided into a working condition causing obvious damage to the reactor core and a core melting (serious accident) working condition according to the state caused by the accident to the reactor core:
DEC-se:Sub>A: design spread conditions/over design benchmark incidents with unmelted core and DEC-B: in particular, severe Accidents (SA) lead to melting of the core.
The initial event which is lower in frequency than the design reference accident occurs, and is not used as the design reference accident, and can be used as DEC-A (without the design expansion working condition which causes obvious damage to the reactor core). Typically events occur less frequently than 10-6 per year of the heap, the event may be divided into DEC events.
In this embodiment, in combination with the analysis results of the frequencies of the non-isolatable LOCA and isolatable LOCA originating events in state C, D, E, the non-isolatable small break accident in the containment of the EPR nuclear power unit in state C, D, E and the isolatable break of the residual heat removal system inside and outside the containment can be categorized as DEC type events.
In this embodiment, according to the analysis result of the LOCA accident originating event frequency, the LOCA originating event list is optimized again according to the EPR nuclear power unit accident classification standard, and the LOCA accident originating event list after optimization is shown in table 3.
Table 3: LOCA accident initiation event list after EPR nuclear power unit optimization
In this embodiment, as shown in fig. 3, the present invention further constructs an EPR nuclear power unit accident initiation event list optimization system, including:
the acquisition module is used for acquiring an initial event list of each nuclear power unit;
the comparison module is used for carrying out comparison analysis on each initial event list and determining an optimization object of the EPR nuclear power unit accident initial event list according to comparison results;
the frequency analysis module is used for carrying out frequency analysis on accident-originating events in the optimization object;
the classification module is used for reclassifying the accidents by combining the EPR accident classification standard and the frequency analysis result;
the optimizing module is used for optimizing the initial event list according to the classification result;
and the result generation module is used for generating an optimized initial event list.
In this embodiment, the acquisition module is further configured to:
acquiring the design safety requirement of the EPR nuclear power unit;
the comparison module is further for:
dividing the operation range of the EPR nuclear power unit into a plurality of states according to design safety requirements, transversely comparing an initial event list of the EPR nuclear power unit with initial event lists of other nuclear power units, identifying difference items among the initial event lists, and selecting corresponding initial events from the plurality of states as optimization objects of the accident initial event list according to identification results.
Specifically, the specific coordination operation process between each module in the EPR nuclear power unit accident initiation event list optimization system may refer to the above-mentioned EPR nuclear power unit accident initiation event list optimization method, and will not be described herein.
In addition, the electronic equipment comprises a memory and a processor; the memory is used for storing a computer program; the processor is used for executing a computer program to realize the EPR nuclear power unit accident onset event list optimization method according to any one of the above. In particular, according to embodiments of the present invention, the processes described above with reference to flowcharts may be implemented as computer software programs. For example, embodiments of the present invention include a computer program product comprising a computer program embodied on a computer readable medium, the computer program comprising program code for performing the method shown in the flowcharts. In such an embodiment, the computer program may perform the above-described functions defined in the methods of embodiments of the present invention when downloaded and installed and executed by an electronic device. The electronic equipment in the invention can be a terminal such as a notebook, a desktop, a tablet computer, a smart phone and the like, and also can be a server.
In addition, a storage medium of the present invention has a computer program stored thereon, which when executed by a processor, implements the EPR nuclear power unit incident initiation event list optimization method of any one of the above. In particular, it should be noted that the storage medium of the present invention may be a computer readable signal medium or a computer readable storage medium, or any combination of the two. The computer readable storage medium can be, for example but not limited to, an electronic, magnetic, optical, electromagnetic, infrared, or semiconductor system, apparatus, or device, or a combination of any of the foregoing. More specific examples of the computer-readable storage medium may include, but are not limited to: an electrical connection having one or more wires, a portable computer diskette, a hard disk, a Random Access Memory (RAM), a read-only memory (ROM), an erasable programmable read-only memory (EPROM or flash memory), an optical fiber, a portable compact disc read-only memory (CD-ROM), an optical storage device, a magnetic storage device, or any suitable combination of the foregoing. In the context of this document, a computer readable storage medium may be any tangible medium that can contain, or store a program for use by or in connection with an instruction execution system, apparatus, or device. In the present invention, however, the computer-readable signal medium may include a data signal propagated in baseband or as part of a carrier wave, with the computer-readable program code embodied therein. Such a propagated data signal may take any of a variety of forms, including, but not limited to, electro-magnetic, optical, or any suitable combination of the foregoing. A computer readable signal medium may also be any computer readable medium that is not a computer readable storage medium and that can communicate, propagate, or transport a program for use by or in connection with an instruction execution system, apparatus, or device. Program code embodied on a computer readable medium may be transmitted using any appropriate medium, including but not limited to: electrical wires, fiber optic cables, RF (radio frequency), and the like, or any suitable combination of the foregoing.
The computer readable medium may be contained in the electronic device; or may exist alone without being incorporated into the electronic device.
In the present specification, each embodiment is described in a progressive manner, and each embodiment is mainly described in a different point from other embodiments, and identical and similar parts between the embodiments are all enough to refer to each other. For the device disclosed in the embodiment, since it corresponds to the method disclosed in the embodiment, the description is relatively simple, and the relevant points refer to the description of the method section.
Those of skill would further appreciate that the various illustrative elements and algorithm steps described in connection with the embodiments disclosed herein may be implemented as electronic hardware, computer software, or combinations of both, and that the various illustrative elements and steps are described above generally in terms of functionality in order to clearly illustrate the interchangeability of hardware and software. Whether such functionality is implemented as hardware or software depends upon the particular application and design constraints imposed on the solution. Skilled artisans may implement the described functionality in varying ways for each particular application, but such implementation decisions should not be interpreted as causing a departure from the scope of the present invention.
The steps of a method or algorithm described in connection with the embodiments disclosed herein may be embodied directly in hardware, in a software module executed by a processor, or in a combination of the two. The software modules may be disposed in Random Access Memory (RAM), memory, read Only Memory (ROM), electrically programmable ROM, electrically erasable programmable ROM, registers, hard disk, a removable disk, a CD-ROM, or any other form of storage medium known in the art.
By implementing the invention, the following beneficial effects are achieved:
the invention discloses an EPR nuclear power unit accident initiation event list optimization method and system. After the initial event list of the EPR nuclear power unit is optimized by the method, the EPR accident analysis and the EPR nuclear power unit operation technical specification are optimized, the unit is maintained on line, and the economic benefit of the nuclear power plant is further improved on the premise of ensuring the safety.
It is to be understood that the above examples only represent preferred embodiments of the present invention, which are described in more detail and are not to be construed as limiting the scope of the invention; it should be noted that, for a person skilled in the art, the above technical features can be freely combined, and several variations and modifications can be made without departing from the scope of the invention; therefore, all changes and modifications that come within the meaning and range of equivalency of the claims are to be embraced within their scope.
Claims (10)
1. The EPR nuclear power unit accident initiation event list optimization method is characterized by comprising the following steps of:
step S10: collecting an initial event list of each nuclear power unit;
step S20: performing comparison analysis on each initial event list, and determining an optimization object of the EPR nuclear power unit accident initial event list according to comparison results;
step S30: performing frequency analysis on the accident-originated event in the optimization object;
step S40: reclassifying the accident by combining the EPR accident classification standard and the frequency analysis result;
step S50: optimizing the originating event list according to the classification result;
step S60: and generating the optimized originating event list.
2. The EPR nuclear power unit accident initiation event list optimization method according to claim 1, wherein the step S10 further comprises:
acquiring the design safety requirement of the EPR nuclear power unit;
the step S20 further includes:
dividing the operation range of the EPR nuclear power unit into a plurality of states according to the design safety requirements, transversely comparing an initial event list of the EPR nuclear power unit with initial event lists of other nuclear power units, identifying difference items among the initial event lists, and selecting a corresponding initial event from the plurality of states as an optimization object of the accident initial event list according to an identification result.
3. The method of optimizing an incident initiation event list for an EPR nuclear power unit according to claim 2, wherein the incidents include a non-isolatable and isolatable water loss incident;
the step S30 includes:
and carrying out frequency analysis on the non-isolatable water loss accident and the isolatable water loss accident in the corresponding originating event to obtain the frequency analysis result of the corresponding originating event.
4. The EPR nuclear power unit incident initiation event list optimization method of claim 3, wherein the inability to isolate a loss of water incident comprises: the reactor coolant system pipeline breaks and the pressure stabilizer safety valve is blocked;
the isolatable water loss event comprises: and the pipeline of the safety shell inner waste heat discharging system is broken and the pipeline of the safety shell outer waste heat discharging system is broken.
5. The EPR nuclear power unit accident initiation event list optimization method according to claim 1, wherein the step S20 further comprises:
step S21: and analyzing the frequency source of the accident originating event in the optimization object, and executing the step S30 if the frequency source is judged to be not in accordance with the preset standard.
6. The EPR nuclear power unit incident initiation event list optimization method of claim 5, wherein the frequency source comprises: small break accidents can not be isolated in the containment and break openings which can be isolated by a waste heat discharging system inside and outside the containment.
7. An EPR nuclear power unit accident initiation event inventory optimization system, comprising:
the acquisition module is used for acquiring an initial event list of each nuclear power unit;
the comparison module is used for carrying out comparison analysis on each initial event list and determining an optimization object of the EPR nuclear power unit accident initial event list according to comparison results;
the frequency analysis module is used for carrying out frequency analysis on the accident-originating event in the optimization object;
the classification module is used for reclassifying the accidents by combining the EPR accident classification standard and the frequency analysis result;
the optimizing module is used for optimizing the initial event list according to the classification result;
and the result generation module is used for generating the optimized initial event list.
8. The EPR nuclear power unit incident initiation event inventory optimization system of claim 7, wherein the acquisition module is further configured to:
acquiring the design safety requirement of the EPR nuclear power unit;
the contrast module is further to:
dividing the operation range of the EPR nuclear power unit into a plurality of states according to the design safety requirements, transversely comparing an initial event list of the EPR nuclear power unit with initial event lists of other nuclear power units, identifying difference items among the initial event lists, and selecting a corresponding initial event from the plurality of states as an optimization object of the accident initial event list according to an identification result.
9. An electronic device, comprising:
one or more processors;
storage means for storing one or more programs which, when executed by the one or more processors, cause the one or more processors to implement the EPR nuclear power unit incident initiation event list optimization method of any one of claims 1 to 6.
10. A storage medium having stored thereon a computer program which when executed by a processor implements the EPR nuclear power unit incident initiation event list optimization method of any one of claims 1 to 6.
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