CN114239279B - Reactor thermal safety design cooperative device, method, terminal and storage medium - Google Patents
Reactor thermal safety design cooperative device, method, terminal and storage medium Download PDFInfo
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- 238000000034 method Methods 0.000 title claims abstract description 54
- 238000013461 design Methods 0.000 title claims abstract description 32
- 239000000446 fuel Substances 0.000 claims abstract description 121
- 238000004458 analytical method Methods 0.000 claims abstract description 62
- 230000001052 transient effect Effects 0.000 claims abstract description 36
- 238000005253 cladding Methods 0.000 claims abstract description 34
- 230000008859 change Effects 0.000 claims abstract description 10
- 238000009835 boiling Methods 0.000 claims description 40
- 239000008188 pellet Substances 0.000 claims description 35
- 239000002826 coolant Substances 0.000 claims description 31
- 230000006399 behavior Effects 0.000 claims description 17
- 238000004364 calculation method Methods 0.000 claims description 17
- 230000008569 process Effects 0.000 claims description 12
- 238000012546 transfer Methods 0.000 claims description 12
- 239000003381 stabilizer Substances 0.000 claims description 10
- 238000004590 computer program Methods 0.000 claims description 6
- 238000000605 extraction Methods 0.000 claims description 6
- 239000004973 liquid crystal related substance Substances 0.000 claims description 6
- 238000004134 energy conservation Methods 0.000 claims description 4
- 230000001105 regulatory effect Effects 0.000 claims description 3
- 230000005540 biological transmission Effects 0.000 abstract description 8
- 230000008054 signal transmission Effects 0.000 description 5
- 238000003672 processing method Methods 0.000 description 3
- 238000012938 design process Methods 0.000 description 2
- 238000005516 engineering process Methods 0.000 description 2
- 230000006870 function Effects 0.000 description 2
- 239000000463 material Substances 0.000 description 2
- 230000009286 beneficial effect Effects 0.000 description 1
- 238000002485 combustion reaction Methods 0.000 description 1
- 238000004891 communication Methods 0.000 description 1
- 238000012986 modification Methods 0.000 description 1
- 230000004048 modification Effects 0.000 description 1
- 230000003287 optical effect Effects 0.000 description 1
- 238000012545 processing Methods 0.000 description 1
- 239000007787 solid Substances 0.000 description 1
- 239000000126 substance Substances 0.000 description 1
Classifications
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- G—PHYSICS
- G06—COMPUTING; CALCULATING OR COUNTING
- G06F—ELECTRIC DIGITAL DATA PROCESSING
- G06F30/00—Computer-aided design [CAD]
- G06F30/20—Design optimisation, verification or simulation
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- G—PHYSICS
- G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
- G21C—NUCLEAR REACTORS
- G21C17/00—Monitoring; Testing ; Maintaining
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- G—PHYSICS
- G06—COMPUTING; CALCULATING OR COUNTING
- G06F—ELECTRIC DIGITAL DATA PROCESSING
- G06F2119/00—Details relating to the type or aim of the analysis or the optimisation
- G06F2119/08—Thermal analysis or thermal optimisation
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- Y—GENERAL TAGGING OF NEW TECHNOLOGICAL DEVELOPMENTS; GENERAL TAGGING OF CROSS-SECTIONAL TECHNOLOGIES SPANNING OVER SEVERAL SECTIONS OF THE IPC; TECHNICAL SUBJECTS COVERED BY FORMER USPC CROSS-REFERENCE ART COLLECTIONS [XRACs] AND DIGESTS
- Y02—TECHNOLOGIES OR APPLICATIONS FOR MITIGATION OR ADAPTATION AGAINST CLIMATE CHANGE
- Y02E—REDUCTION OF GREENHOUSE GAS [GHG] EMISSIONS, RELATED TO ENERGY GENERATION, TRANSMISSION OR DISTRIBUTION
- Y02E30/00—Energy generation of nuclear origin
- Y02E30/30—Nuclear fission reactors
Abstract
The application discloses a reactor thermal safety design cooperative device, a method, a terminal and a storage medium, which comprise a thermal hydraulic transient analysis module, a reactor core thermal hydraulic subchannel analysis module and a fuel element behavior analysis module, wherein the thermal hydraulic transient analysis module is used for acquiring system transient characteristics in an accident state and a change curve of system parameters with time, the reactor core thermal hydraulic subchannel analysis module is used for determining fuel burning share, and the fuel element behavior analysis module is used for acquiring a fuel cladding temperature peak value and a core block temperature peak value; according to the application, the work in the thermal safety design is decomposed into the thermal hydraulic transient analysis module, the reactor core thermal hydraulic subchannel analysis module and the fuel element behavior analysis module, and the information among the modules is transmitted through data transmission, so that the proper module work can be selected according to specific design requirements, a large amount of input data from different professions is prevented from being processed, and the work efficiency and quality are improved.
Description
Technical Field
The application relates to the field of thermal safety of reactors, in particular to a device, a method, a terminal and a storage medium for collaborative design of thermal safety of a reactor.
Background
In the thermal safety design process of the nuclear reactor, a large amount of input data from different professions needs to be processed, a plurality of professional computing programs are adopted for computing and analyzing, the data amount generated in the computing and analyzing process is large, a data processing method and a transmission process are complex, and how to efficiently manage the data and a scheduling flow is one of the keys for restricting the efficiency and the quality of the thermal safety design of the reactor.
Most of thermal safety design cooperative devices in the current stage input a plurality of data at one time, carry out overall safety judgment, have no splitting function, and cannot select proper cooperative flow according to specific conditions.
Disclosure of Invention
The application aims to provide a reactor thermal safety design collaboration device, a method, a terminal and a storage medium, and solve the problem of digital management of the reactor thermal safety analysis design, wherein the technical problems are that the data volume generated in the calculation and analysis process in the thermal safety design process is large, the data processing method and the transmission process are complex, and the device and the method have no universality.
The application is realized by the following technical scheme:
a reactor thermal safety design collaboration apparatus, comprising:
the thermal hydraulic transient analysis module is used for acquiring system transient characteristics in an accident state and acquiring a change curve of system parameters along with time;
a core thermodynamic hydraulic subchannel analysis module for determining fuel burn-out fractions;
a fuel element behavior analysis module for obtaining a fuel cladding temperature peak and a pellet temperature peak;
the signal output end of the thermal hydraulic transient analysis module is connected with the signal input end of the reactor core thermal hydraulic subchannel analysis module and the signal input end of the fuel element behavior analysis module.
Specifically, the core thermodynamic and hydraulic subchannel analysis module comprises:
a first calculation module for obtaining a deviated nucleate boiling ratio corresponding to each point in time;
the first judging module is used for judging whether the minimum value of the deviation nucleate boiling ratio is smaller than a preset limit value;
the adjusting module is used for adjusting the enthalpy rise factor when the minimum value of the deviation nucleate boiling ratio is smaller than a preset limit value, and inputting the adjusted enthalpy rise factor into the first calculating module;
the enthalpy rise factor determining module is used for obtaining the enthalpy rise factor meeting the criterion of deviating from the limit value of the nucleate boiling ratio when the minimum value of the deviating from the nucleate boiling ratio is larger than the preset limit value;
the fuel statistical curve extraction module is used for obtaining a fuel statistical curve and obtaining an enthalpy rise factor-fuel burnout share curve;
a fuel burn-out portion determination module for determining a fuel burn-out portion in a corresponding fuel statistical curve based on the obtained enthalpy rise factor.
Specifically, the fuel element behavior analysis module includes:
the input module is used for obtaining the fuel cladding temperature and the pellet temperature corresponding to each time point in the accident process;
a second calculation module for obtaining a fuel clad temperature peak and a pellet temperature peak.
A reactor thermal safety design collaboration method, comprising:
acquiring system transient characteristics under an accident state, acquiring a change curve of system parameters along with time, and synchronously inputting the system transient characteristics to a reactor core thermal hydraulic sub-channel analysis module and a fuel element behavior analysis module;
determining a fuel burn-out fraction;
obtaining a fuel cladding temperature peak value and a pellet temperature peak value;
and judging whether the safety limit criterion is met or not according to the fuel burning share and the fuel cladding temperature peak value and the pellet temperature peak value.
Specifically, the system parameters include thermal power, core inlet temperature, stabilizer pressure, coolant volumetric flow fuel burn-out fraction, nuclear power, coolant temperature, and coolant density.
The calculating method of the transient characteristic of the system comprises the following steps:
establishing a system model;
dividing a system model into a plurality of sections;
solving corresponding mass, energy and momentum conservation equations and a system transient characteristic calculation model;
and combining the initial condition and the boundary condition of the accident to obtain the transient characteristic of the system in the accident state.
Specifically, the method of determining the fuel burn-up share includes:
obtaining a deviation nucleate boiling ratio corresponding to each time point;
judging whether the minimum value of the deviated nucleate boiling ratio is smaller than a preset limit value, and if so, obtaining an enthalpy rise factor meeting the criterion of the deviated nucleate boiling ratio;
if the minimum value of the deviation nucleate boiling ratio is smaller than the preset limit value, the deviation nucleate boiling ratio corresponding to each time point is obtained again after the enthalpy rise factor is regulated, and whether the minimum value of the deviation nucleate boiling ratio is smaller than the preset limit value is judged;
obtaining a fuel statistical curve and an enthalpy rise factor-fuel burnout share curve;
and determining the fuel burning share in the corresponding fuel statistical curve according to the obtained enthalpy rise factor.
Specifically, the method for obtaining the deviated nucleate boiling ratio corresponding to each time point comprises the following steps:
establishing a three-dimensional model of the reactor core, the three-dimensional model taking as inputs reactor thermal power, core inlet temperature, stabilizer pressure, and coolant volumetric flow fuel burnout fraction;
obtaining heat transfer parameters of a reactor core and a coolant, wherein the heat transfer parameters are obtained by solving a mass and energy conservation equation and a heat transfer relation;
a deviation from nucleate boiling ratio was obtained.
Specifically, the method for obtaining the fuel cladding temperature and the pellet temperature comprises the following steps:
obtaining the temperature of the fuel cladding and the temperature of the pellets corresponding to each time point in the accident process;
the fuel cladding temperature peak value and the pellet temperature peak value are obtained by the following steps:
taking reactor nuclear power, pressurizer pressure, coolant temperature, coolant density and coolant flow as inputs, solving a heat transfer equation for the fuel;
obtaining a temperature profile of the fuel pellets and the cladding;
the fuel cladding temperature peak and pellet temperature peak are obtained.
A computer readable storage medium storing a computer program, wherein the computer program when executed by a processor implements the steps of a reactor thermal safety design collaboration method described above.
An electronic device, comprising: at least one processor; the method comprises the steps of,
a memory communicatively coupled to the at least one processor; wherein, the liquid crystal display device comprises a liquid crystal display device,
the memory stores instructions executable by the at least one processor to cause the at least one processor to process the steps of a reactor thermal safety design coordination method described above.
Compared with the prior art, the application has the following advantages and beneficial effects:
according to the application, the work in the thermal safety design is decomposed into the thermal hydraulic transient analysis module, the reactor core thermal hydraulic subchannel analysis module and the fuel element behavior analysis module, and the information among the modules is transmitted through data transmission, so that the proper module work can be selected according to specific design requirements, a large amount of input data from different professions is prevented from being processed, and the work efficiency and quality are improved.
Drawings
The accompanying drawings, which are included to provide a further understanding of the application and are incorporated in and constitute a part of this specification, illustrate exemplary embodiments of the application and together with the description serve to explain the principles of the application.
Fig. 1 is a flow chart of a reactor thermal safety design collaboration method in accordance with the present application.
Fig. 2 is a flow chart of a method for determining a fuel burn-up share according to the present application.
Fig. 3 is a flow chart of a method of obtaining fuel cladding temperature peaks and pellet temperature peaks according to the present application.
Detailed Description
The present application will be described in further detail with reference to the drawings and embodiments, for the purpose of making the objects, technical solutions and advantages of the present application more apparent. It is to be understood that the specific embodiments described herein are merely illustrative of the substances, and not restrictive of the application.
It should be further noted that, for convenience of description, only the portions related to the present application are shown in the drawings.
Embodiments of the present application and features of the embodiments may be combined with each other without conflict. The present application will be described in detail below with reference to the accompanying drawings in conjunction with embodiments.
The stuck shaft accident generally causes the fuel rod to deviate from nucleate boiling (DNB), and when the thermal safety analysis of the reactor is carried out, the fuel burning share of the fuel rod with DNB and the highest temperature of the fuel cladding and the pellets in the transient process need to be calculated for judging whether the corresponding limit criterion is met.
For design analysis scenes involving a plurality of professional computing programs, large data volume, a data processing method and complex transmission process, the design analysis scenes are necessary to be decomposed into task flows based on standardized flows, so that a designer can orderly plan the design flows and manage data, and the thermal engineering safety design work of the reactor is completed under the driving of the task flows. Therefore, a collaborative design method and an implementation process for the thermal safety design of the reactor based on task flow control are provided.
Example 1
The embodiment provides a reactor thermal safety design cooperative device, which comprises a thermal hydraulic transient analysis module, a reactor core thermal hydraulic subchannel analysis module and a fuel element behavior analysis module.
In the analysis work of the stuck shaft accident, the whole analysis task is decomposed into three subtasks: (1) calculating system transient characteristics under an accident; (2) determining the fuel burn-up fraction at which DNB occurs; (3) The fuel cladding temperature peak and pellet temperature peak are calculated.
In this embodiment, the three subtasks are carried by three modules, and corresponding parameters are obtained in the three modules.
The thermal hydraulic transient analysis module is used for acquiring the transient characteristics of the system in an accident state and acquiring a change curve of system parameters along with time;
the system parameters include thermal power, core inlet temperature, stabilizer pressure, coolant volumetric flow fuel burn-out fraction, nuclear power, coolant temperature, and coolant density, among others.
And after the obtained system parameters are matched with time, generating a change curve.
Inputting a change curve corresponding to system parameters such as thermal power, core inlet temperature, voltage stabilizer pressure, coolant flow fuel burnout share and the like into a core thermal hydraulic sub-channel analysis module to determine the fuel burnout share;
inputting a change curve corresponding to system parameters such as nuclear power, voltage stabilizer pressure, coolant temperature, coolant density, coolant flow fuel burnout share and the like into a fuel element behavior analysis module to obtain a fuel cladding temperature peak value and a pellet temperature peak value;
the signal output end of the thermal hydraulic transient analysis module is connected with the signal input end of the reactor core thermal hydraulic subchannel analysis module and the signal input end of the fuel element behavior analysis module, and data transmission is carried out among the three modules through a signal transmission device.
The three modules can be arranged in one host for data transmission through cables, and also can be arranged at different positions for data transmission through wireless devices.
Example two
The present embodiment provides a reactor thermal safety design cooperation method based on the first embodiment, including:
the thermal hydraulic transient analysis module acquires the transient characteristics of the system in an accident state, acquires a change curve of system parameters along with time, and synchronously inputs the transient characteristics of the system to the reactor core thermal hydraulic sub-channel analysis module and the fuel element behavior analysis module;
the calculating method of the transient characteristic of the system comprises the following steps: in the thermal hydraulic transient analysis module, a model is built for the system and divided into a plurality of sections, and the transient characteristics of the system in the accident state are obtained by solving corresponding mass, energy and momentum conservation equations and related calculation models and combining initial conditions and boundary conditions about the accident.
In this embodiment, solving the corresponding mass, energy conservation equation and related calculation model according to the system parameters is a common calculation method for those skilled in the art, and will not be described again.
The reactor core thermal hydraulic sub-channel analysis module determines the fuel burning share according to the input related data;
the fuel element behavior analysis module obtains a fuel cladding temperature peak value and a pellet temperature peak value according to the input related data;
and judging whether the safety limit criterion is met or not according to the fuel burning share and the fuel cladding temperature peak value and the pellet temperature peak value.
The safety limit rule is obtained according to the regulation, and if the safety limit rule is not met, the reactor thermal safety design is proved to be unqualified.
Example III
The specific structure of the reactor core thermal hydraulic sub-channel analysis module in the first embodiment comprises a first calculation module, a first judgment module, an adjustment module, an enthalpy rise factor determination module, a fuel statistical curve extraction module and a fuel burning share determination module.
The first calculation module is used for obtaining a deviation nucleate boiling ratio corresponding to each time point;
the first judging module is used for judging whether the minimum value of the deviation nucleate boiling ratio is smaller than a preset limit value;
the adjusting module is used for adjusting the enthalpy rise factor when the minimum value of the deviated nucleate boiling ratio is smaller than a preset limit value, and inputting the adjusted enthalpy rise factor into the first calculating module;
the enthalpy rise factor determining module is used for obtaining the enthalpy rise factor meeting the criterion of deviating from the limit value of the nucleate boiling ratio when the minimum value of the deviating from the nucleate boiling ratio is larger than the preset limit value;
the fuel statistical curve extraction module is used for obtaining a fuel statistical curve and an enthalpy rise factor-fuel burnout share curve, wherein the combustion statistical curve is an existing parameter, and the existing parameter is input into the fuel statistical curve library in advance and extracted through the fuel statistical curve extraction module.
The fuel burn-out portion determination module is used for determining the fuel burn-out portion in the corresponding fuel statistical curve according to the obtained enthalpy rise factor.
The signal transmission method between the modules can be determined to be executed in one overall module or executed in a plurality of different sub-modules according to specific situations, and the signal transmission method between the modules comprises the following steps:
the output end of the first calculation module is connected with the input end of the first judgment module, the output end of the first judgment module is respectively connected with the input ends of the adjustment module and the enthalpy rise factor determination module, the output end of the adjustment module is connected with the input end of the first calculation module, and the output end of the enthalpy rise factor determination module and the output end of the fuel statistics curve extraction module are both connected with the input end of the fuel burnout share determination module.
The specific working method comprises the following steps:
obtaining a deviation nucleate boiling ratio corresponding to each time point; the method for obtaining the deviation nucleate boiling ratio corresponding to each time point comprises the following steps:
the reactor core is built into a three-dimensional model by taking the reactor thermal power, the reactor core inlet temperature, the pressure of a pressure stabilizer and the fuel burnout share of the volume flow of the coolant, which are calculated by the thermal transient analysis module, as input, and the heat transfer condition between the reactor core and the coolant is obtained by solving a mass and energy conservation equation and a heat transfer relation, so that the deviated nucleate boiling ratio is obtained.
Judging whether the minimum value of the deviated nucleate boiling ratio is smaller than a preset limit value, and if so, obtaining an enthalpy rise factor meeting the criterion of the deviated nucleate boiling ratio;
if the minimum value of the deviation nucleate boiling ratio is smaller than the preset limit value, the deviation nucleate boiling ratio corresponding to each time point is obtained again after the enthalpy rise factor is regulated, and whether the minimum value of the deviation nucleate boiling ratio is smaller than the preset limit value is judged;
obtaining a fuel statistical curve and an enthalpy rise factor-fuel burnout share curve;
and determining the fuel burning share in the corresponding fuel statistical curve according to the obtained enthalpy rise factor.
Example III
The specific structure of the fuel element behavior analysis module in the first embodiment is described in this embodiment, and the fuel element behavior analysis module includes an input module and a second calculation module.
The input module is used for obtaining the fuel cladding temperature and the pellet temperature corresponding to each time point in the accident process;
a second calculation module for obtaining a fuel cladding temperature peak and a pellet temperature peak;
the signal transmission method between the modules can be determined to be executed in one overall module or executed in a plurality of different sub-modules according to specific situations, and the signal transmission method between the modules comprises the following steps:
the signal output end of the output module is connected with the signal input end of the second calculation module.
The specific working method comprises the following steps:
the method for obtaining the corresponding fuel cladding temperature and pellet temperature at each time point in the accident process comprises the following steps:
and taking the reactor nuclear power, the pressure of the voltage stabilizer, the temperature of the coolant, the density of the coolant and the flow of the coolant which are calculated by the thermal transient analysis module as inputs, and obtaining the temperature distribution of the fuel pellets and the cladding by solving a heat transfer equation about the fuel, thereby obtaining a fuel cladding temperature peak value and a pellet temperature peak value.
Obtaining a fuel cladding temperature peak value and a pellet temperature peak value;
after obtaining the fuel burn-out share in the second embodiment and the cladding temperature peak and pellet temperature peak in the third embodiment, the core thermodynamic and hydraulic subchannel analysis module and the fuel element behavior analysis module output them to a host server or the like and compare them with safety limit criteria.
Example IV
A computer readable storage medium storing a computer program, wherein the computer program when executed by a processor implements the steps of a reactor thermal safety design collaboration method described above.
The memory may be used to store software programs and modules, and the processor executes various functional applications of the terminal and data processing by running the software programs and modules stored in the memory. The memory may mainly include a storage program area and a storage data area, wherein the storage program area may store an operating system, an execution program required for at least one function, and the like.
The storage data area may store data created according to the use of the terminal, etc. In addition, the memory may include high-speed random access memory, and may also include non-volatile memory, such as at least one magnetic disk storage device, flash memory device, or other volatile solid-state storage device.
An electronic device, comprising: at least one processor; the method comprises the steps of,
a memory communicatively coupled to the at least one processor; wherein, the liquid crystal display device comprises a liquid crystal display device,
the memory stores instructions executable by the at least one processor to cause the at least one processor to process the steps of a reactor thermal safety design coordination method as described above.
Computer readable media may include computer storage media and communication media without loss of generality. Computer storage media includes volatile and nonvolatile, removable and non-removable media implemented in any method or technology for storage of information such as computer readable instruction data structures, program modules or other data. Computer storage media includes RAM, ROM, EPROM, EEPROM, flash memory or other solid state memory technology, CD-ROM, DVD or other optical storage, magnetic cassettes, magnetic tape, magnetic disk storage or other magnetic storage devices. Of course, those skilled in the art will recognize that computer storage media are not limited to the ones described above. The above-described system memory and mass storage devices may be collectively referred to as memory.
In the description of the present specification, reference to the terms "one embodiment/manner," "some embodiments/manner," "example," "a particular example," "some examples," etc., means that a particular feature, structure, material, or characteristic described in connection with the embodiment/manner or example is included in at least one embodiment/manner or example of the application. In this specification, the schematic representations of the above terms are not necessarily for the same embodiment/manner or example. Furthermore, the particular features, structures, materials, or characteristics described may be combined in any suitable manner in any one or more embodiments/modes or examples. Furthermore, the various embodiments/modes or examples described in this specification and the features of the various embodiments/modes or examples can be combined and combined by persons skilled in the art without contradiction.
Furthermore, the terms "first," "second," and the like, are used for descriptive purposes only and are not to be construed as indicating or implying a relative importance or implicitly indicating the number of technical features indicated. Thus, a feature defining "a first" or "a second" may explicitly or implicitly include at least one such feature. In the description of the present application, the meaning of "plurality" means at least two, for example, two, three, etc., unless specifically defined otherwise.
It will be appreciated by persons skilled in the art that the above embodiments are provided for clarity of illustration only and are not intended to limit the scope of the application. Other variations or modifications of the above-described application will be apparent to those of skill in the art, and are still within the scope of the application.
Claims (6)
1. A reactor thermal safety design coordination device, comprising:
the thermal hydraulic transient analysis module is used for acquiring system transient characteristics in an accident state and acquiring a change curve of system parameters along with time;
a core thermodynamic hydraulic subchannel analysis module for determining fuel burn-out fractions;
a fuel element behavior analysis module for obtaining a fuel cladding temperature peak and a pellet temperature peak;
the signal output end of the thermal hydraulic transient analysis module is connected with the signal input end of the reactor core thermal hydraulic subchannel analysis module and the signal input end of the fuel element behavior analysis module;
wherein, the reactor core thermal hydraulic sub-channel analysis module comprises:
a first calculation module for obtaining a deviated nucleate boiling ratio corresponding to each point in time;
the first judging module is used for judging whether the minimum value of the deviation nucleate boiling ratio is smaller than a preset limit value;
the adjusting module is used for adjusting the enthalpy rise factor when the minimum value of the deviation nucleate boiling ratio is smaller than a preset limit value, and inputting the adjusted enthalpy rise factor into the first calculating module;
the enthalpy rise factor determining module is used for obtaining the enthalpy rise factor meeting the criterion of deviating from the limit value of the nucleate boiling ratio when the minimum value of the deviating from the nucleate boiling ratio is larger than the preset limit value;
the fuel statistical curve extraction module is used for obtaining a fuel statistical curve and obtaining an enthalpy rise factor-fuel burnout share curve;
a fuel burn-out portion determination module for determining a fuel burn-out portion in a corresponding fuel statistical curve based on the obtained enthalpy rise factor;
wherein the fuel element behavior analysis module comprises:
the input module is used for obtaining the fuel cladding temperature and the pellet temperature corresponding to each time point in the accident process;
a second calculation module for obtaining a fuel clad temperature peak and a pellet temperature peak;
the reactor nuclear power, the pressure of the pressure stabilizer, the temperature of the coolant, the density of the coolant and the flow of the coolant which are calculated by the thermal hydraulic transient analysis module are taken as inputs, and the temperature distribution of the fuel pellets and the cladding is obtained by solving a heat transfer equation about the fuel, so that the fuel cladding temperature peak value and the pellet temperature peak value are obtained.
2. A method of reactor thermal safety design collaboration, comprising:
acquiring system transient characteristics under an accident state, acquiring a change curve of system parameters along with time, and synchronously inputting the system transient characteristics to a reactor core thermal hydraulic sub-channel analysis module and a fuel element behavior analysis module;
determining a fuel burn-out fraction;
obtaining a fuel cladding temperature peak value and a pellet temperature peak value;
judging whether a safety limit criterion is met or not according to the fuel burning share, the fuel cladding temperature peak value and the pellet temperature peak value;
wherein the method of determining the fuel burn-out fraction comprises:
obtaining a deviation nucleate boiling ratio corresponding to each time point;
judging whether the minimum value of the deviated nucleate boiling ratio is smaller than a preset limit value, and if so, obtaining an enthalpy rise factor meeting the criterion of the deviated nucleate boiling ratio;
if the minimum value of the deviation nucleate boiling ratio is smaller than the preset limit value, the deviation nucleate boiling ratio corresponding to each time point is obtained again after the enthalpy rise factor is regulated, and whether the minimum value of the deviation nucleate boiling ratio is smaller than the preset limit value is judged;
obtaining a fuel statistical curve and an enthalpy rise factor-fuel burnout share curve;
determining fuel burn-out portions in the corresponding fuel statistical curves according to the obtained enthalpy rise factors;
wherein the method of obtaining a fuel cladding temperature peak and a pellet temperature peak comprises:
obtaining the temperature of the fuel cladding and the temperature of the pellets corresponding to each time point in the accident process;
the fuel cladding temperature peak value and the pellet temperature peak value are obtained by the following steps: taking reactor nuclear power, pressurizer pressure, coolant temperature, coolant density and coolant flow as inputs, solving a heat transfer equation for the fuel; obtaining a temperature profile of the fuel pellets and the cladding; the fuel cladding temperature peak and pellet temperature peak are obtained.
3. The reactor thermal safety design coordination method of claim 2, wherein the system parameters include thermal power, core inlet temperature, stabilizer pressure, coolant volumetric flow fuel burnout fraction, nuclear power, coolant temperature, and coolant density;
the calculating method of the transient characteristic of the system comprises the following steps:
establishing a system model;
dividing a system model into a plurality of sections;
solving corresponding mass, energy and momentum conservation equations and a system transient characteristic calculation model;
and combining the initial condition and the boundary condition of the accident to obtain the transient characteristic of the system in the accident state.
4. A reactor thermal safety design synergy method according to claim 2, wherein said method for obtaining a deviated nucleate boiling rate corresponding to each time point comprises:
establishing a three-dimensional model of the reactor core, the three-dimensional model taking as inputs reactor thermal power, core inlet temperature, stabilizer pressure, and coolant volumetric flow fuel burnout fraction;
obtaining heat transfer parameters of a reactor core and a coolant, wherein the heat transfer parameters are obtained by solving a mass and energy conservation equation and a heat transfer relation;
a deviation from nucleate boiling ratio was obtained.
5. A computer readable storage medium storing a computer program, characterized in that the computer program when executed by a processor implements the steps of the method according to any of claims 2-4.
6. An electronic device, comprising: at least one processor; the method comprises the steps of,
a memory communicatively coupled to the at least one processor; wherein, the liquid crystal display device comprises a liquid crystal display device,
the memory stores instructions executable by the at least one processor to enable the at least one processor to implement the steps of the method of any one of claims 2-4.
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| Publication number | Priority date | Publication date | Assignee | Title |
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| KR20110045660A (en) * | 2009-10-27 | 2011-05-04 | 한국전력공사 | Nuclear reactor core assessment method using thermal hydraulic safety analysis code |
| CN105653869A (en) * | 2016-01-05 | 2016-06-08 | 中国核动力研究设计院 | Three-dimensional transient performance analysis method for supercritical water reactor core |
| CN108875213A (en) * | 2018-06-19 | 2018-11-23 | 哈尔滨工程大学 | A kind of reactor core thermal-hydraulic multiscale analysis method |
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| KR20110045660A (en) * | 2009-10-27 | 2011-05-04 | 한국전력공사 | Nuclear reactor core assessment method using thermal hydraulic safety analysis code |
| CN105653869A (en) * | 2016-01-05 | 2016-06-08 | 中国核动力研究设计院 | Three-dimensional transient performance analysis method for supercritical water reactor core |
| CN108875213A (en) * | 2018-06-19 | 2018-11-23 | 哈尔滨工程大学 | A kind of reactor core thermal-hydraulic multiscale analysis method |
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