Heat pipe micro-reactor probability safety target formulating method based on quantitative prediction model

Abstract

A heat pipe micro-stack probability safety target formulating method based on a quantitative prediction model comprises the following steps: determining an originating event list and an originating event frequency, establishing an event sequence model, establishing a fault tree of accident response measure failure corresponding to event tree branches, and calculating a failure probability, boolean operation is carried out on the event tree to obtain the final state frequency of the event sequence, totaling and random sampling are carried out on the event sequence with the released radionuclide activity exceeding the dose constraint limit value to obtain the frequency constraint limit value, and therefore the dose constraint limit value-frequency constraint limit value coupling safety target of the heat pipe micro-stack is obtained. According to the method, the probabilistic safety target is calculated from the structure of the heat pipe micro-stack, the technical practice of the heat pipe micro-stack is met, and effective guidance can be provided for design, manufacturing and structure optimization of the heat pipe micro-stack.

Description

Technical Field

[0001] The invention belongs to the field of nuclear power, and in particular relates to a method for formulating probabilistic safety targets for a heat pipe micro-reactor based on a quantitative prediction model. Background Art

[0002] At present, regulatory authorities generally set corresponding probabilistic safety target thresholds for existing nuclear power plant facilities. For example, the U.S. Nuclear Regulatory Commission (NRC) has given probabilistic safety targets from the perspectives of core damage frequency (CDF) and large early release frequency (LERF) of nuclear power plants: CDF<10 -4 / reactor-year, LERF<10 -5 / reactor year. The Safety Assessment and Verification of Nuclear Power Plants (HAD102 / 17) issued by the China National Nuclear Safety Administration in 2006 recommended a probabilistic safety assessment target for new nuclear power plants of core damage frequency <10 -5 / reactor year, frequency of large radioactive releases <10 -6 Reactor year. However, as a new type of micro mobile nuclear power source, the heat pipe micro reactor has better inherent safety than traditional nuclear power plants due to its low power, small nuclear fuel loading, and all-solid core. The above-mentioned probabilistic safety evaluation targets can only be given from a regulatory perspective. It is impossible to propose guiding probabilistic safety targets for heat pipe micro reactors from a technical perspective, and it is difficult to constrain and evaluate the safety level of heat pipe micro reactor technical solutions. Therefore, a reasonable and feasible method for formulating probabilistic safety targets for heat pipe micro reactors based on quantitative prediction models is provided to determine the probabilistic safety evaluation targets that reflect the design characteristics of heat pipe micro reactors, which is of great value to the design traction and safety assurance of heat pipe micro reactors. Summary of the invention

[0003] The purpose of the present invention is to provide a method for formulating a probabilistic safety target for a heat pipe micro-reactor based on a quantitative prediction model, and to determine the probabilistic safety target according to the structural characteristics of the heat pipe micro-reactor.

[0004] According to an embodiment of the present invention, a method for formulating probabilistic safety goals for a heat pipe micro-reactor based on a quantitative prediction model is provided, the method comprising the following steps: determining a list of initiating events and the corresponding frequencies of initiating events. According to the list of initiating events, an event sequence model is established, the event sequence model comprising an event tree caused by the initiating event and a judgment criterion for each level of the event tree branch; wherein the event tree comprises an envelope event of each initiating event in the heat pipe micro-reactor, the event tree is determined according to the system response logic of the heat pipe micro-reactor; each level of the event tree branch comprises an accident non-deterioration branch and an accident deterioration branch, the judgment criterion of the branch is the response effectiveness of the accident response measures of the heat pipe micro-reactor corresponding to the event branch. Establish a fault tree for the failure of the accident response measures corresponding to each level of the event tree branch, and calculate the failure probability of the accident response measures. Perform Boolean operations on the event tree to obtain the frequencies of different accident end states of the event sequence model. Calculate the activity of radionuclides released to the environment by the accident end state. Provide a dose constraint limit for the release of radionuclides into the environment, screen the accident end states that exceed the dose constraint limit and add them up to obtain the frequency of the over-limit event; perform random sampling analysis on the over-limit event frequency to obtain the probability distribution of the accident end state that exceeds the dose constraint limit, and obtain the frequency constraint limit. Obtain the probabilistic safety target of the heat pipe micro-reactor, which includes the coupling of the dose constraint limit and the frequency constraint limit.

[0005] Through the above method, starting from the device characteristics and design characteristics of the heat pipe micro-reactor, the probabilistic safety target of a given heat pipe micro-reactor can be predicted by quantitative calculation, thereby improving the technical rationality of setting the probabilistic safety target. Furthermore, this method combines the characteristics of the heat pipe micro-reactor's all-solid-state core, and uses the dose constraint limit and the frequency constraint limit as the indicator of the probabilistic safety target of the heat pipe micro-reactor, which is more in line with the actual technical characteristics of the heat pipe micro-reactor than the safety indicator of the core damage frequency of conventional nuclear power facilities.

[0006] Furthermore, in some embodiments, the method for determining the initiating event list includes FMEA analysis and comparison and decomposition of the deterministic accident list, and the process of determining the initiating event list also includes the step of establishing a main logic diagram for integrity verification.

[0007] Furthermore, in some embodiments, after determining the initiating event list, the step of classifying and merging the initiating event list into event groups is also included, and the event groups include: accidental shutdown, transient loss of thermoelectric conversion system, general transient, heat pipe rupture in the core area, heat pipe rupture in the heat exchanger area, core and reactor body failure, heat exchanger failure, and radioactive shielding and containment failure.

[0008] Further, in some embodiments, the frequency of the initiating event is determined by the following method: for transient events, the transient events include accidental shutdown, loss of thermal power conversion system transient or general transient, F T =F GT ×P T , where F T is the frequency of transient initiating events, F GT is the frequency of transient initiating events in the general database, P T is the percentage of the corresponding transient events in all transient events. For the core area heat pipe rupture and the heat exchanger area heat pipe rupture, F P =R j ×L×Y, where F P is the frequency of heat pipe rupture initiation events, R j is the failure rate of heat pipe rupture per unit length, L is the length of the heat pipe, and Y is the reactor-year time. For failures of the core and reactor body, and failures of the heat exchanger, the event frequency is calculated by establishing a fault tree. For failures of radioactive shielding and containment, F RC =F TK ×Y×f c , where F RC is the frequency of initiating events of radioactive shielding and containment failure, F TK is the external leakage data of the central pressure vessel in the general database, Y is a bunch of years, f c is a given correction factor.

[0009] Furthermore, in some embodiments, the fault tree is established by the following method: determining the target system that fails corresponding to the failure of the accident response measure, taking the failure of the target system as the top event, determining the sufficient cause event and the necessary cause event of the top event as the secondary event, further determining the sufficient cause event and the necessary cause event that lead to the occurrence of the secondary event as the tertiary event, repeating this process until the bottom event of all cause events is determined, and connecting all cause events through logic gates.

[0010] Further, in some embodiments, the method for calculating the failure probability of the accident response measures includes the steps of determining the component failure probability and determining the common cause event probability. The component failure includes demand failure and operation failure, the demand failure includes the demand failure of the periodic test type component, and the demand failure event probability is Q mean =1-(1-e -rTI ) / rTI, where r is the failure rate and TI is the test time interval; the probability of an operational failure event is Q mean =1-e -rTM, where TM is the mission time. The probability of a common cause event is determined by the following method: determine a common cause component group, where the components in the common cause component group meet the following conditions: have the same initial conditions and operating characteristics, have the same role or function, have the same failure mode, have the same component operating environment or test maintenance related error factors; the probability of a common cause event is Where Q T is the probability of the common cause of the common cause event, m is the number of components in the common cause component group, α i (m) is the alpha factor for common cause failure of i components, and k is the minimum number of common factors.

[0011] Furthermore, in some embodiments, the accident response measures include reactivity control, decay heat removal and radioactive containment; the fault tree for failure of reactivity control is established based on the shutdown system, the fault tree for failure of decay heat removal is established based on the residual heat removal system, and the fault tree for failure of radioactive containment is established based on the stack container system.

[0012] Furthermore, in some embodiments, before performing Boolean operations on the event tree, a step of merging the accident final states is also included, and the merged accident final states include: N0: the cladding, core matrix and stack container of the heat pipe micro-reactor are all failed, and the cladding temperature does not exceed the allowable temperature; N1: the cladding and core matrix of the heat pipe micro-reactor are failed, the stack container is valid, and the cladding temperature does not exceed the allowable temperature; N3: the cladding, core matrix and stack container of the heat pipe micro-reactor are all valid, and the cladding temperature does not exceed the allowable temperature; S0: the cladding, core matrix and stack container of the heat pipe micro-reactor are all failed, and the maximum temperature of the cladding exceeds the allowable temperature but the duration is shorter than the service limit of the cladding at the allowable temperature; S 1: : The cladding and core matrix of the heat pipe micro-reactor fail, the stack container is valid, the maximum temperature of the cladding exceeds the allowable temperature but the duration is shorter than the service limit of the cladding at the allowable temperature; S3: The cladding, core matrix and stack container of the heat pipe micro-reactor are all valid, the maximum temperature of the cladding exceeds the allowable temperature but the duration is shorter than the service limit of the cladding at the allowable temperature; L0: The cladding, core matrix and stack container of the heat pipe micro-reactor fail, the maximum temperature of the cladding exceeds the allowable temperature and the duration reaches or exceeds the service limit of the cladding at the allowable temperature; L1: The cladding and core matrix of the heat pipe micro-reactor fail, the stack container is valid, the maximum temperature of the cladding exceeds the allowable temperature and the duration reaches or exceeds the service limit of the cladding at the allowable temperature.

[0013] Furthermore, in some embodiments, the activity of radionuclides released into the environment in the final state of the accident is calculated by the following method:

[0014]

[0015] Among them, R i is the activity of radionuclide i released into the environment after the accident, T is the time after the accident, q i is the activity of radionuclide i released from the fuel into the stack per unit time, λ di is the seashore constant of radionuclide i, and L is the stack leakage rate.

[0016] Further, in some embodiments, the dose constraint limit is calculated by the following method:

[0017]

[0018] Where D is the personal effective dose after the accident, DCF in,i is the conversion factor of the effective dose of internal radiation of nuclide i, DCF T,I is the thyroid dose conversion factor for radionuclide i, R i,j is the activity of nuclide i released into the environment in time period j, Br j is the breathing rate of the personnel in time period j; the D value is corrected by using the personnel radiation biological effect influence parameter to obtain the dose constraint limit. BRIEF DESCRIPTION OF THE DRAWINGS

[0019] Figure 1 A flow chart of a method for formulating a probabilistic safety target for a heat pipe micro-repile based on a quantitative prediction model in one embodiment;

[0020] Figure 2 A main logic diagram of an unexpected radioactive release in an embodiment;

[0021] Figure 3 A logic diagram of radioactive release of a reactor module in one embodiment;

[0022] Figure 4 A fault tree of a heat exchanger failure initiation event in an embodiment;

[0023] Figure 5 This is a schematic diagram of an event tree structure in an embodiment;

[0024] Figure 6 A schematic diagram of a transient event tree of a loss of a thermoelectric conversion system in one embodiment;

[0025] Figure 7 A schematic diagram of a shutdown failure fault tree in an embodiment;

[0026] Figure 8 Schematic diagram of the frequency distribution of events exceeding the dose constraint limit in one embodiment.

[0027] The purpose of the above drawings is to explain the present invention in detail so that those skilled in the art can understand the technical concept of the present invention, but it is not intended to limit the present invention. DETAILED DESCRIPTION

[0028] The present invention is further described in detail below through specific embodiments in conjunction with the accompanying drawings.

[0029] The heat pipe micro-reactor is a new type of nuclear reactor. Unlike the traditional pressurized water reactor, the core of the heat pipe micro-reactor is a fully solid-state structure. The fuel rods enclosed in the reactor compartment are inserted into the solid-state core, and heat is exported to the heat exchanger through heat pipes to generate electricity. It is light in weight and occupies a small area, and can be flexibly deployed as a micro mobile nuclear power source. Due to the structural differences between the heat pipe micro-reactor and the traditional reactor, as well as the characteristics of the heat pipe micro-reactor with low power, small fuel loading, and low system complexity, the heat pipe micro-reactor has higher inherent safety. Therefore, the various safety targets formulated based on traditional nuclear reactors cannot fully reflect the technical characteristics of the heat pipe micro-reactor, and it is difficult to guide the reliability design and safety level of the heat pipe micro-reactor.

[0030] In order to solve the above problems, an embodiment of the present invention provides a method for formulating a probabilistic safety target of a heat pipe micro-reactor based on a quantitative prediction model. The calculation process is as follows: Figure 1 shown.

[0031] The method specifically comprises the following steps:

[0032] Step 1: Identify the list of initiating events.

[0033] For heat pipe micro reactors, the list of initiating events can be determined in the following ways:

[0034] 1) Conduct system engineering assessment to identify potential initiating events. Use the failure mode and effects analysis (FMEA) method to conduct initiating events that may be caused by a certain system failure in the reactor, and find the failure mode that will directly or in combination with other failures lead to radioactive release, including but not limited to operational failure, malfunction, fracture, leakage, etc. In one embodiment, the results of the system-level FMEA analysis for a certain model of heat pipe micro-reactor are shown in Table 1.

[0035] 2) The accident list of heat pipe micro-reactors given by deterministic analysis is used as a comparison, and the accident list of heat pipe micro-reactors under deterministic analysis is analyzed, evaluated and further refined from the perspective of probability theory, as shown in Table 2. The accident list given by deterministic analysis divides heat pipe micro-reactor accidents into five categories: increased heat output of the heat pipe reactor system, reduced heat removal capacity of the thermoelectric conversion system, abnormal reactivity and power distribution, decreased heat exchange capacity of the matrix / heat pipe, and expected transients that failed to shut down the reactor in an emergency.

[0036]

[0037] Table 1 FEMA evaluation table for heat pipe micro-reactor system

[0038]

[0039] Table 2 Comparison of the list of deterministic accidents and probabilistic initiating events

[0040] By comparing Table 2, it can be seen that the probabilistic initiating event list is more detailed than the deterministic accident list. For example, according to the different reactor responses and radioactive consequences, the deterministic heat pipe failure accident is subdivided into heat pipe rupture in the core area and heat pipe rupture in the heat exchanger area. In addition, the probabilistic initiating event also takes into account the failure of the core matrix, damage to the stack container, failure of the electrical system, etc., and the analysis is more complete.

[0041] 3) Use the main logic diagram to verify the integrity of the finalized initiating event list. To further confirm the integrity of the initiating event list, based on the similarity of the accident process or consequences, establish the following Figure 2 The main logic diagram of unexpected radioactive release shown is used to classify all internal initiating events. The structure of the diagram is similar to the logical structure of the fault tree analysis. The top event represents the unexpected radioactive release (UR) caused by all initiating events. The shapes under each bottom event represent different event types: diamonds represent non-developing events, such as unexpected UR (UR-EXE) caused by external events, which need to be combined with specific application scenarios and will not be further analyzed in the main logic diagram, while low power and shutdown events (IE-LPSD) need to be combined with specific information such as the reactor refueling cycle and will not be further analyzed in the main logic diagram; unexpected radioactive release caused by internal initiating events (UR-INE) may be caused by insufficient core cooling or radioactive release from reactor modules; circles represent the bottom-level events of the analysis, such as the expected transient state (ATWS) that failed to shut down the reactor in an emergency; and triangles represent that the event can be further disassembled into a new main logic diagram, such as the radioactive release (R-RR) of the reactor module, which can be further established as Figure 3 The main logic diagram is shown in Figure 1. A complete analysis is performed using the main logic diagram to trace each event branch back to the non-development event or the bottom event, thus completing the integrity verification of the initiating event list. Figure 3Taking the R-RR shown in the figure as an example, according to the system structure and working logic analysis of the heat pipe micro-reactor, the possible causes of R-RR include radioactive release of the core fuel system (COR-UR), radioactive release of the heat transfer system (PHX-UR), radioactive release of the radiation shielding system (RRS-UR), and radioactive release of the stack system (RCC-UR). Any one of these events will cause R-RR to occur, so they are connected through an OR gate. COR-UR has three further possible causes: core matrix failure, core body support failure, and core area heat pipe rupture; PHX-UR has two further possible causes: heat pipe rupture in the heat exchanger area and heat exchanger failure; RRS-UR has a single cause of radiation shielding failure; RCC-UR has a single cause of stack container damage.

[0042] 4) Group and merge the initiating events to obtain the initiating event group. The merging of initiating events is carried out according to the following principles: a) Initiating events with the same response system and success criteria (including the system's automatic response and operator action) to avoid unexpected radioactive releases can be grouped together; b) In some cases, the relative importance is determined according to the frequency of the initiating events, and the non-dominant initiating events are merged into the dominant initiating events, and there is no need to build an event tree separately; c) During the grouping process, it is found that some categories of initiating events need to be further subdivided because the reactor response is significantly different or there is a possibility of more serious radioactive release; d) For initiating events that may directly lead to unexpected radioactive releases, since their frequency of occurrence is very low and there are no mitigation measures, it is not necessary to specially build an event tree for analysis, but they must be included in the frequency of unexpected radioactive releases. According to the above principles, in some embodiments, the list of initiating events is divided into 8 categories: SSR, transient loss of thermal power conversion system (TRAN-PCS), general transient

[0043] (GTRAN), heat pipe rupture in the core area (HP-COR), heat pipe rupture in the heat exchanger area (HP-PHX), core and reactor body failure (PHX), and radioactive shielding and containment failure (RRC).

[0044] Step 2, determine the frequency of occurrence of each initiating event of the heat pipe micro-reactor.

[0045] For the above 8 types of initiating event groups, the following method is used in the preferred embodiment to quantify the initiating event frequency:

[0046] For transient initiating events similar to those in conventional nuclear power plants, such as accidental shutdown, transient event of loss of thermal power conversion system, general transient event, etc., the calculation formula is: F T =F GT ×P T Among them, F Tis the frequency of transient initiating events (unit: / reactor-year), F GT is the frequency of common transient initiating events in the general database (unit: / reactor-year), P T It is the percentage of transient events such as accidental shutdown, turbine shutdown, and other general transient events (excluding the contribution of events such as reactor coolant loss that are not applicable to micro-mobile nuclear power sources) in the general transient events. This proportion can be queried in the general database (such as NUREG / CR-5750).

[0047] For the initiating events caused by heat pipe rupture, which is unique to heat pipe micro-reactors, including heat pipe rupture in the core area and heat pipe rupture in the heat exchanger area, the pipeline / valve analysis method is used to determine the initiating event frequency by multiplying the heat pipe rupture failure rate by the corresponding heat pipe length. The calculation formula for the heat pipe rupture initiating event frequency Fp (unit: / reactor year) is F p =R i ×L×Y. Among them, R i is the rupture failure rate of each heat pipe section (unit: / hour), L is the length of the heat pipe (unit: section), and Y is a pile-year time of 8760 hours.

[0048] For the initiating events caused by the failure of the core and the reactor body, as well as the failure of the heat exchanger, which are unique to the heat pipe micro-reactor, the initiating event fault tree is established to calculate the frequency by analyzing the failure logic based on the failure mode and reliability data of the corresponding system / equipment. Figure 4 As shown, in one embodiment, a fault tree analysis is performed for the initiating event of heat exchanger failure (IE-RCS-HX), and the main factors of its failure include failure of heat exchanger related components (@IE-RCS-HX-1) such as failure of inlet and outlet pipes, and failure of the heat exchanger body (@IE-RCS-HX-2). Among them, the failure factors of heat exchanger inlet pipe failure (@IE-RCS-HX-3) and heat exchanger outlet pipe failure (@IE-RCS-HX-4) further include fixed pipe blockage (RCS-IN / OUT-SP-GPY), connector structure failure (RCS-IN / OUT-JO-Y) and metal hose thermal failure (@IE-RCS-HX-5 / 6). Among them, the fixed hose blockage and the connector structure failure are inseparable bottom events, and the metal hose thermal failure can be further caused by the bottom events of metal hose blockage (RCS-IN / OUT-MI-GPY), metal hose external leakage (small) (RCS-IN / OUT-MI-YSY) and metal hose external leakage (large) (RCS-IN / OUT-MI-YLY). The cause of the heat exchanger body failure (@IE-RCS-HX-2) is the heat exchanger structure failure (RCS-HX-Y). According to the occurrence frequency of all the above bottom events, the failure frequency of the top event can be calculated according to the initiating event fault tree.

[0049] For the initiating events of radioactive shielding and containment failure, combined with the similarity of equipment categories and the mechanism of the initiating events, the large external leakage data of unpressurized containers in the NUREG / CR-6928 general database is used as a reference, and after certain corrections, it is used as the frequency of the initiating event. The calculation formula is: F RC =F TK ×Y×f c Among them, F RC is the frequency of initiating events of radioactive shielding and containment failure (unit: / reactor-year), F TK is the external leakage data of non-pressure vessels in the general database (unit: / h), Y is a pile of years 8760 hours, f c is a given correction factor.

[0050] Step 3: Conduct event sequence modeling to determine the potential event sequence corresponding to each initiating event.

[0051] After the initiating event occurs, corresponding mitigation methods need to be adopted to avoid the event from deteriorating and ensure that the heat pipe micro-reactor is in a safe and stable state. Using the event tree tool, the event sequence model of the micro mobile nuclear power source is developed through the small event tree-large fault tree model development method.

[0052] First, for each initiating event group of the heat pipe micro-reactor, determine the safety functions that need to be performed to prevent radioactive release. In some embodiments, the three safety functions that need to be performed normally to maintain a safe and stable state of the heat pipe micro-reactor are: reactivity control, decay heat removal, and radioactive containment.

[0053] Next, the systems that perform the above safety functions are identified. In some embodiments, the systems that perform reactivity control are two independent shutdown systems, namely the drum and the control rods; the system that performs decay heat removal is the residual heat removal system; and the system that performs radioactive containment is the stack container.

[0054] The next step is to determine the final state of the event sequence. Partial or complete failure of the above safety functions will lead to different consequences. The various consequences can be classified and merged from the two dimensions of core damage degree and radioactive material release, which can be divided into 8 types of accident final states:

[0055] N0: The cladding, core matrix and stack container of the heat pipe micro-reactor are all failed, and the cladding temperature does not exceed the allowable temperature;

[0056] N1: The cladding and core matrix of the heat pipe micro-reactor fail, the stack container is valid, and the cladding temperature does not exceed the allowable temperature;

[0057] N3: The cladding, core matrix and stack container of the heat pipe micro-reactor are all effective, and the cladding temperature does not exceed the allowable temperature;

[0058] S0: The cladding, core matrix and stack container of the heat pipe micro-reactor are all failed, and the maximum temperature of the cladding exceeds the allowable temperature but the duration is shorter than the service limit of the cladding at the allowable temperature;

[0059] S1: The cladding and core matrix of the heat pipe micro-reactor fail, the stack container is valid, the maximum temperature of the cladding exceeds the allowable temperature but the duration is shorter than the service limit of the cladding at the allowable temperature;

[0060] S3: The cladding, core matrix and stack container of the heat pipe micro-reactor are all effective, and the maximum temperature of the cladding exceeds the allowable temperature but the duration is shorter than the service limit of the cladding at the allowable temperature;

[0061] L0: The cladding, core matrix and stack container of the heat pipe micro-reactor all fail, the maximum temperature of the cladding exceeds the allowable temperature and the duration reaches or exceeds the service limit of the cladding at the allowable temperature;

[0062] L1: The cladding and core matrix of the heat pipe micro-reactor fail, the stack container is valid, the maximum temperature of the cladding exceeds the allowable temperature and the duration reaches or exceeds the service limit of the cladding at the allowable temperature.

[0063] In some embodiments, the allowable temperature of the cladding material is 800° C., and the service limit time of the cladding material at 800° C. can be determined through experiments.

[0064] Finally, graphical event tree logic is used. An event tree is established for each initiating practice group. The starting point (top event) of the event tree is the initiating event, and the end point is a certain accident final state. Each event tree describes the response of the micro-mobile nuclear power source to the most extreme event (i.e., the envelope event) in the event group. When determining the response of the micro-mobile nuclear power source, all systems are considered as long as they are actually expected to respond to the event. In one embodiment, the event tree structure is as follows: Figure 5 As shown, for the initiating event ZZZ, the top event is represented by the header IEV-ZZZ, and the top event is divided into multiple levels of branch events. Each level of branch events corresponds to two different event trends. Header 1 represents the second-level event XXX, whose failure mark is XXX1. The upward branch of XXX1 represents that XXX1 did not occur (success), while the downward branch represents that XXX1 occurred and the accident worsened. Similarly, header 2 represents the third-level event YYY. YYY has two failure conditions YYY1 and YYY2, and finally forms 4 bottom events and their corresponding final states I-N3, II-S3, III-N3, and IV-S0. In the above event tree, when IEV-XXX occurs, XXX1 does not occur (success), and YYY1 occurs, resulting in the accident final state S3; when IEV-XXX occurs, XXX1 occurs, and YYY2 occurs, resulting in the accident final state S0.

[0065] Based on the characteristics of the initiating event, or the state of the previous system, the success criteria of a header in one branch may be different from the success criteria of the same header in other branches, or even different in the same event tree. In this case, the same header may have different input identifiers on different branches of the same event tree. These identifiers are also called "conditions". The most common example of conditions is radioactive containment, whose successful configuration (such as the number of valves and pipelines required to be closed) and startup method (such as automatic and / or manual) largely depend on the progress of previous events. For heat pipe micro-reactors, when the residual heat removal is effective, the substrate and heat pipes are considered as barriers, depending on their own material properties; in the case of residual heat removal failure, consider closing the relevant valves to ensure the integrity of the chamber.

[0066] Step 4: Determine the success criteria for each event sequence header and obtain quantitative input.

[0067] The success criteria are the minimum level required for the system. The number of columns or devices that need to be operated is determined by system reliability, diversity and redundancy.

[0068] In one embodiment, the process of modeling the event sequence for the transient (TRAN-PCS) initiating event of the heat pipe micro-pile loss of heat and electricity conversion system and obtaining the quantitative input of the header success criteria is as follows: Establish an event tree model such as Figure 6As shown in FIG. 1 , the event tree includes the following headings: the initiating event TRAN-PCS; the secondary event RTRIP (reactor shutdown), which indicates that the reactor is successfully shut down by the drum or the safety rod. According to the system design, the failure of one drum is allowed and the shutdown can also be achieved by the safety rod. Therefore, the success of this heading requires that 11 of the 12 drums are successfully actuated or the safety rod is inserted to a sufficient shutdown depth; the tertiary event JNA (residual heat removal), which indicates that the reactor temperature is controllable by removing the residual heat of the reactor through air convection. According to the system design, the residual heat removal system has three valves, and the design capacity of each valve is 50 %, so the success of this title requires that two of the three columns of the residual heat removal system are successfully put into operation; the fourth level event BF (radioactivity containment barrier intact), during the event, high temperature and high pressure may occur in the stack container. If the heat pipe loses its integrity, it will lead to the loss of a barrier to contain radioactivity. In the case of failure of the residual heat removal system, it is necessary to consider isolating the residual heat removal and closing the relevant penetrations on the stack container to achieve radioactivity containment. The success of this title requires closing the penetrations on the stack container in the case of residual heat removal failure. In the case of successful operation of the residual heat removal, the pressure in the heat pipe is lower than its pressure limit and the temperature in the heat pipe is lower than its temperature limit. For TRAN-PCS, RTRIP, JNA, and BF are all successful, and the final state is N3; only JNA fails, and the final state is L1; only JNA and BF fail, and the final state is L0; only RTRIP fails, and the final state is S1; only RTRIP and BF fail, and the final state is S0; and if both RTRIP and JNA fail, BF will inevitably fail, resulting in the final state L0.

[0069] In order to obtain the cumulative frequency of each accident sequence, in addition to determining the frequency of the initiating event, it is also necessary to determine the conditional probability of the sequence (the probability of the sequence occurring under the condition of the initiating event). The conditional probability of the sequence needs to be multiplied by the branch probability of each header on the event sequence. For the event tree header RTRIP (reactor shutdown), its success criterion is that under accident conditions, the reactor shutdown system successfully triggers the shutdown and reaches the required shutdown depth (the drum is successfully reset to the required number or the safety rod is successfully inserted). The failure criterion of the residual discharge system is determined based on the design characteristics of the system and the requirements of the success criterion, that is, the number of successful resets of the drum is less than the required number and the safety rod insertion fails. The fault tree logic structure is constructed according to this criterion.

[0070] Establish a fault tree for RTRIP failure, such as Figure 7As shown in Figure 7, RTRIP failure is taken as the top event, and the judgment criteria for its failure are determined, that is, the number of successful drum resets is less than the required number and the safety rod insertion fails. Find all necessary and sufficient direct cause events that cause the top event, place them in the corresponding event symbols, draw them in the second row of the fault tree, and then connect them with appropriate logic gates according to the logical relationship between these direct cause events and the top event in the actual system. For the top event of reactor shutdown failure, the top event occurs when the drum shutdown failure (failure of 2 or more drum actions, @RTRIP-2) and the shutdown rod shutdown failure (failure of safety rod insertion, @RTRIP-3) occur at the same time. Therefore, the second row of the fault tree is shown in Figure 7, and the two failure events are connected by an "AND" gate. Next, find the necessary and sufficient direct cause events of each failure event in the second row, similar to the third row of the fault tree described in step 2, and then use appropriate logic gates to connect these direct cause events (third row) with the failure event (second row). In this way, follow the tree building rules and the actual situation of the system to expand downward layer by layer until all cause events are bottom events. Figure 7 In the third row of the fault tree shown in the figure, the causes of the jamming of two or more drums and the failure of the insertion of the safety rod are found respectively. For the jamming of the drum, it may be caused by the failure of the drive of the drum drive mechanism (@RTRIP-5) or the failure of the trigger of the instrumentation and control system (PMS) (RTPMS), so the third row is connected by the "OR" gate of the above two failure events. In this way, it is expanded downward layer by layer until all the cause events are bottom events. Finally, the occurrence of the top event is transformed into the Boolean algebra between all the bottom events: RTRIP = (@RTRIP-5∪ZGQD-CCF2)∪RTPMS∩(KZBQD-CM∪RTPMS). Among them, KZBQD-CM is the event control rod drive mechanism cannot operate, ZGQD-CCF2 is the common cause failure of the operation of two or more drums, and ZGQD-DM-x is the operation failure of the xth drum drive mechanism.

[0071] After the fault tree is established, in order to perform quantitative analysis on the fault tree, the probability of occurrence of each basic event needs to be given. Specifically, the failure probability of the component and the probability of the common cause event need to be determined.

[0072] Determine the probability of failure of a component.

[0073] For the bottom events of component failure in the top event fault tree, they can be mainly divided into two categories: demand failure type and operation failure type. The corresponding calculation method of the basic event failure probability is divided into five major types. They are continuously monitored repairable components, components for periodic testing, components with a fixed unavailability, components with a fixed mission time, and non-repairable components.

[0074] The failure probability of a component is determined by using different reliability models and reliability parameters according to different situations.

[0075] The most common demand failure events in the fault tree are determined directly by the "component type with a constant unavailability" and the parameter q (failure probability), or by the "component type for periodic testing" and r (failure rate), TI (test time interval). The probability of a demand failure event is Q mean =1-(1-e -rTI ) / rTI, where r is the failure rate and TI is the test time interval; the operational failure event is determined by the fourth type of model and parameters r (failure rate) and TM (mission time), and the probability of the operational failure event is Q mean =1-e -rTM , where TM is the task time.

[0076] Since the structure of heat pipe micro reactor is different from that of traditional pressurized water reactor, the selection of reliability data needs to be carried out for the equipment type in this reactor type. Reliability tests and analysis should be carried out on the special equipment of heat pipe micro reactor to calculate the equipment reliability parameters; for general equipment, general data of similar equipment of similar size and similar working environment can be used. General data is mainly taken from NUREG / CR-6928.

[0077] When failure data cannot be obtained from the above data sources, or it is considered that the failure data is not applicable to this design, refer to the following data sources in order:

[0078] 1) URD

[0079] 2) NUREG / CR-2728

[0080] 3) Other data sources

[0081] Other data sources include US sodium-cooled reactor data, etc. If the failure data is not taken from NUREG / CR-6928, a note should be provided.

[0082] For the data taken from NUREG / CR-6928, it is assumed that the demand failure probability obeys the beta distribution and the failure rate obeys the gamma distribution, and the recommended α and β values ​​are used as the parameters of the two distributions; for the data taken from other data sources, it is assumed that they all obey the lognormal distribution, and the error factor value is selected in the following principles: for the parameters that are directly obtained by statistical analysis but the analysis process and specific distribution are not given in the relevant literature, EF is taken as 3; for the data obtained by simple calculation from multiple different data sources, or the data calculated by the same method as the general data, the EF value can be taken as 10 or an approximate value of EF in the general data; for the data without any basis and only judged by experts, if the failure probability judged to be small, EF is taken as 30.

[0083] For the reactor shutdown problem, the failure rate of the drum drive mechanism jamming is 4.23×10 -7 / h, obeys log-normal distribution, EF is 10, and the failure probability is calculated using the periodic test type, with a periodic test cycle of one month. mean =1-(1-e -rTI ) / rTI, the failure probability of the drum drive mechanism jamming is 1.54×10 -4 .

[0084] Determine the probability of common cause events.

[0085] Inter-component correlation, also known as common cause failure, is a root cause event that directly causes multiple component failures due to a common cause. Common cause failures require special attention because they can cause redundant components to fail simultaneously (or nearly simultaneously, i.e., within the mission time), thereby reducing the theoretical maximum reliability increase obtained by adding redundant components / columns. The process of common cause failure analysis is as follows:

[0086] First, identify the common cause component group. A typical component group includes similar components that may cause a system redundant loop failure with the same failure mode. If necessary, first determine the assumptions of the common cause failure analysis, and then qualitatively screen the components of the same type (such as pumps, valves, etc.) according to the following conditions: have the same initial conditions (i.e., normally open, normally closed, charged, power off, etc.) and operating characteristics (normal operation, standby); have the same role or function, such as system isolation, flow regulation, parameter sensing, prime mover, etc.; have the same failure mode (failure to open when required, failure to start when required, etc.); have the same component operating environment, test maintenance related errors and other factors. For the drum drive mechanism jam in the reactor shutdown header, it has the same initial conditions and operating characteristics, the same function and the same failure mode and operating environment, etc., and the common causes between component failures need to be considered. The common cause failure analysis of the process system in the heat pipe micro-reactor is carried out using the alpha factor method.

[0087] Next, evaluate the common cause failure parameters. Select the relevant common cause failure analysis method and appropriate data source, and process the general data with the power plant specific data to obtain the appropriate common cause failure parameters. In the analysis of the heat pipe micro-reactor system, refer to the general parameter library NUREG / CR-5497 and NUREG / CR-5500. The alpha factor method is used to calculate the probability of common cause failure, and the formula is as follows: Where Q T is the probability of the common cause of the common cause event, m is the number of components in the common cause component group, α i (m)is the alpha factor of common cause failure of i components, and k is the minimum number of common causes. For the ZGQD-CCF2 (common cause failure of 2 or more drums), the failure criterion is that any 2 or more of the 12 drums are stuck, and the alpha factor method formula is used: Where Q T The probability of the drum drive getting stuck.

[0088] Step 5: quantify the event sequence and obtain the cumulative frequency of each event sequence.

[0089] Combined with the micro-mobile nuclear power source initiating event list, event sequence model and event sequence header input determined in the above steps, the frequency of occurrence of severe accident sequences of micro-mobile nuclear power sources is quantitatively calculated through a quantitative calculation model.

[0090] For the event sequences with final states of N1, S1, L1, N0, S0, and LO, the frequency of each final state is calculated by considering the frequency of the initiating events of the micro-mobile nuclear power source corresponding to the accident sequence, the header inputs of the two independent shutdown systems of the control rod and the drum, the residual heat removal system, the cabin and other mitigation systems.

[0091] For a sequence of events of a certain final state, the following quantitative calculation formula is used to calculate the sequence occurrence frequency of the final state:

[0092]

[0093] Among them, F CONi is the sum of the occurrence frequencies of each event sequence with the final state CONi, IE i The originating event type IE i The annual frequency of occurrence, Q FUN1i For IE i The probability value of the shutdown head (shutdown failure) under the initiating event sequence; Q FUN2i For IE i The probability value of the residual heat removal system failure (residual heat removal failure) under the initiating event sequence, Q FUN3i For IE i The probability value of the stack container header (stack containment failure) under the initiating event sequence. Among them, IE i The value of comes from step 2, Q FUN1i , Q FUN2i , Q FUN3i The value of comes from step 4.

[0094] Step 6, determine the activity of radionuclides released into the environment after the accident. The activity of radionuclides released into the environment after the heat pipe microreactor accident can be calculated by the following formula:

[0095]

[0096] Among them, R i is the activity of radionuclide i released into the environment after the accident, in Bq, T is the time after the accident, in h, q i is the activity of radionuclide i released from the fuel to the stack per unit time, in Bq / h, λ di is the decay constant of radionuclide i, in h -1 , L is the stack leakage rate, unit: h -1 .

[0097] The above formula can be used to calculate the activity of radionuclides released into the environment under different core damage states.

[0098] Step 7: Determine the dose constraint limit.

[0099] The radioactivity released into the environment after the accident will migrate and diffuse in the ambient atmosphere. Based on step 6, the degree of atmospheric contamination in the vicinity of the device after the accident is determined in combination with the meteorological conditions of the plant site, and the effective dose of the accident is determined by comprehensively considering the internal exposure caused by inhalation of airborne radioactive substances and the external exposure caused by immersion in the radioactive plume. The calculation model for the personal effective dose in the vicinity of the device after the accident is as follows:

[0100]

[0101] Where D is the personal effective dose after the accident, in Sv; DCF in,i is the conversion factor of the effective dose of internal radiation of nuclide i, in Sv / Bq; DCF T,I is the thyroid dose conversion factor of radionuclide i, in Sv / Bq; R i,j is the activity of nuclide i released into the environment in time period j, in Bq; Br j is the breathing rate of the person in time period j, in m 3 / s. Based on the array of personal effective doses in the vicinity of the device obtained by modeling and analysis, the D value is corrected using the parameters affecting the biological effects of radiation on personnel to obtain the effective dose constraint limit A.

[0102] Step 8: Determine the frequency constraint limit.

[0103] Through step 6, the radionuclide release dose corresponding to the final state caused by each initiating event sequence can be calculated, and the result can be compared with the effective dose constraint limit obtained in step 7. The final state frequencies exceeding the dose constraint limit are summed up to obtain the cumulative frequency F of the over-limit event sequence. c . After calculating F c After the point value, cUncertainty analysis is performed, that is, random sampling analysis is performed on the probability of each event in the over-limit event sequence. In some embodiments, Monte Carlo simulation can be performed more than 10,000 times to obtain F c The frequency distribution results are as follows Figure 8 As shown, the 95% score point is selected as the frequency constraint limit B.

[0104] Step 9, obtain the probabilistic safety target of the heat pipe micro-reactor based on the coupling of dose constraint limit and frequency constraint limit: the cumulative frequency of all accident sequences in which radioactive materials released into the environment during the accident cause the effective dose of individuals in the adjacent area to exceed A (mSv) should be less than B (reactor / year).

[0105] The method for formulating the probabilistic safety target of a heat pipe micro-reactor based on a quantitative prediction model provided in an embodiment of the present invention starts from the structural design of the heat pipe micro-reactor and proposes a method for determining the probabilistic safety target of the heat pipe micro-reactor through technical analysis. The safety target calculated by this method conforms to the technical characteristics of the heat pipe micro-reactor and has targeted guiding significance for the design, construction and structural optimization of the heat pipe micro-reactor.

[0106] The purpose of the above embodiments is to further explain the present invention in detail in conjunction with the accompanying drawings so that those skilled in the art can understand the technical concept of the present invention. Within the scope of the present invention, the optimization or equivalent replacement of the part structures and method steps involved, as well as the combination of implementation methods in different embodiments without conflicts in structure and principle, all fall within the protection scope of the present invention.

Claims

1. A method for formulating probabilistic safety targets for heat pipe micro-reactors based on a quantitative prediction model, characterized in that: The following steps are involved: Determine the list of initiating events and the corresponding initiating event frequencies; According to the initiating event list, an event sequence model is established, wherein the event sequence model includes an event tree caused by the initiating event and a determination criterion for each level of the event tree branch; wherein the event tree includes an enveloping event of each initiating event in the heat pipe micro-pile, and the event tree is determined according to the system response logic of the heat pipe micro-pile; each level of the event tree branch includes an accident non-deterioration branch and an accident deterioration branch, and the determination criterion of the branch is the response effectiveness of the accident response measures of the heat pipe micro-pile corresponding to the event branch; Establishing a fault tree of failure of the accident response measures corresponding to each level of the event tree branch, and calculating the failure probability of the accident response measures; Performing Boolean operations on the event tree to obtain frequencies of different accident final states of the event sequence model; Calculate the activity of radionuclides released into the environment in the final state of the accident; Provide dose constraint limits for the release of radionuclides into the environment, screen out the accident end states that exceed the dose constraint limits and sum them up to obtain the frequency of excess events; Performing random sampling analysis on the frequency of the over-limit events to obtain the probability distribution of the accident final state exceeding the dose constraint limit and obtain the frequency constraint limit; The probabilistic safety target of the heat pipe micro-pile is obtained, and the probabilistic safety target of the heat pipe micro-pile includes the coupling of the dose constraint limit value and the frequency constraint limit value.

2. The method for formulating probabilistic safety targets for heat pipe micro-reactors based on a quantitative prediction model according to claim 1, characterized in that: The method for determining the initiating event list includes FMEA analysis and comparison and decomposition of the deterministic accident list. The process of determining the initiating event list also includes the step of establishing a main logic diagram for integrity verification.

3. The method for formulating probabilistic safety targets for heat pipe micro-reactors based on a quantitative prediction model according to claim 2, characterized in that: After determining the initiating event list, the method also includes classifying and merging the initiating event list into event groups, wherein the event groups include: accidental shutdown, transient loss of thermoelectric conversion system, general transient, heat pipe rupture in the core area, heat pipe rupture in the heat exchanger area, failure of the core and the reactor body, failure of the heat exchanger, and radioactive shielding and containment failure.

4. The method for formulating probabilistic safety targets for heat pipe micro-reactors based on a quantitative prediction model according to claim 3 is characterized in that: The frequency of the initiating event is determined by: For transient events, the transient events include accidental shutdown, loss of thermal power conversion system transient or general transient, F T =F GT ×P T , where F T is the frequency of transient initiating events, F GT is the frequency of transient initiating events in the general database, P T is the percentage of the corresponding transient events in all transient events; For heat pipe rupture in the core area and heat pipe rupture in the heat exchanger area, F P =R j ×L×Y, where F P is the frequency of heat pipe rupture initiation events, R j is the failure rate of heat pipe rupture per unit length, L is the length of the heat pipe, and Y is a pile-year period; For failures of the core and reactor body, and heat exchanger, the event frequency is calculated by establishing a fault tree; For radioactive shielding and containment failures, F RC =F TK ×Y×f c , where F RC is the frequency of initiating events of radioactive shielding and containment failure, F TK is the external leakage data of the central pressure vessel in the general database, Y is a bunch of years, f c is a given correction factor.

5. The method for formulating probabilistic safety targets for heat pipe micro-reactors based on a quantitative prediction model according to any one of claims 1 to 4, characterized in that: The fault tree is established by the following method: determining the target system that fails corresponding to the failure of the accident response measure, taking the failure of the target system as the top event, determining the sufficient cause event and the necessary cause event of the top event as the secondary event, further determining the sufficient cause event and the necessary cause event that lead to the occurrence of the secondary event as the tertiary event, repeating this process until the bottom event of all cause events is determined, and connecting all cause events through logic gates.

6. The method for formulating probabilistic safety targets for heat pipe micro-reactors based on a quantitative prediction model according to claim 5, characterized in that: A method for calculating the failure probability of accident response measures, including the steps of determining the probability of component failure and determining the probability of common cause events; The component failure includes demand failure and operation failure, wherein the demand failure includes demand failure of periodic test type components, and the probability of the demand failure event is Q mean =1-(1-e -rTI ) / rTI, where r is the failure rate and TI is the test time interval; the probability of an operational failure event is Q mean =1-e -rTM , where TM is the task time; The probability of a common cause event is determined by the following method: determining a common cause component group, wherein the components in the common cause component group meet the following conditions: have the same initial conditions and operating characteristics, have the same role or function, have the same failure mode, have the same component operating environment or test maintenance related error factors; The probability of the common cause event is Where Q T is the probability of the common cause of the common cause event, m is the number of components in the common cause component group, α i (m) is the alpha factor for common cause failure of i components, and k is the minimum number of common factors.

7. The method for formulating probabilistic safety targets for heat pipe micro-reactors based on a quantitative prediction model according to any one of claims 1 to 4, characterized in that: The accident response measures include reactivity control, decay heat removal and radioactive containment; the fault tree for reactivity control failure is established based on the shutdown system, the fault tree for decay heat removal failure is established based on the residual heat removal system, and the fault tree for radioactive containment failure is established based on the stack container system.

8. The method for formulating probabilistic safety targets for heat pipe micro-reactors based on a quantitative prediction model according to any one of claims 1 to 4, characterized in that: Before performing Boolean operation on the event tree, the method further includes merging the accident final states, wherein the merged accident final states include: N0: The cladding, core matrix and chamber container of the heat pipe micro-reactor are all failed, and the cladding temperature does not exceed the allowable temperature; N1: The cladding and core matrix of the heat pipe micro-reactor fail, the stack container is valid, and the cladding temperature does not exceed the allowable temperature; N3: The cladding, core matrix and stack container of the heat pipe micro-reactor are all effective, and the cladding temperature does not exceed the allowable temperature; S0: The cladding, core matrix and stack container of the heat pipe micro-reactor are all failed, and the maximum temperature of the cladding exceeds the allowable temperature but the duration is shorter than the service limit of the cladding at the allowable temperature; S1: The cladding and core matrix of the heat pipe micro-reactor fail, the stack container is valid, the maximum temperature of the cladding exceeds the allowable temperature but the duration is shorter than the service limit of the cladding at the allowable temperature; S3: The cladding, core matrix and stack container of the heat pipe micro-reactor are all effective, and the maximum temperature of the cladding exceeds the allowable temperature but the duration is shorter than the service limit of the cladding at the allowable temperature; L0: The cladding, core matrix and stack container of the heat pipe micro-reactor all fail, the maximum temperature of the cladding exceeds the allowable temperature and the duration reaches or exceeds the service limit of the cladding at the allowable temperature; L1: The cladding and core matrix of the heat pipe micro-reactor fail, the stack container is valid, the maximum temperature of the cladding exceeds the allowable temperature and the duration reaches or exceeds the service limit of the cladding at the allowable temperature.

9. The method for formulating probabilistic safety targets for heat pipe micro-reactors based on a quantitative prediction model according to any one of claims 1 to 4, characterized in that: The activity of radionuclides released into the environment in the final state of the accident is calculated in the following way: Among them, R i is the activity of radionuclide i released into the environment after the accident, T is the time after the accident, q i is the activity of radionuclide i released from the fuel into the stack per unit time, λ di is the decay constant of radionuclide i, and L is the stack leakage rate.

10. The method for formulating probabilistic safety targets for heat pipe micro-reactors based on a quantitative prediction model according to any one of claims 1 to 4, characterized in that: The dose constraint limit is calculated by the following method: Where D is the personal effective dose after the accident, DCF in,i is the conversion factor of the effective dose of internal radiation of nuclide i, DCF T,I is the thyroid dose conversion factor for radionuclide i, R i,j is the activity of nuclide i released into the environment in time period j, Br j is the breathing rate of the person in time period j; The D value is corrected using the personnel radiation biological effect influence parameter to obtain the dose constraint limit.

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