Radiological consequence analysis method, apparatus, device, medium, and program product

Abstract

This application discloses a method, apparatus, equipment, medium, and program product for radioactive consequences analysis, relating to the field of nuclear power accident analysis technology. The radioactive consequences analysis method includes: obtaining the core stockpile; based on a standard source term framework, determining the time-varying release fraction of source terms from the core to the containment based on the core stockpile, to obtain the release rate; employing a simplified natural removal model, calculating the time-varying environmental release source terms based on the release rate and containment leakage rate; wherein the simplified natural removal model includes a predefined set of natural removal coefficients; inputting the environmental release source terms into a radioactivity dose assessment model to calculate the radioactive consequences. This application solves the problem of high computational resource consumption in current methods for analyzing the radioactive consequences of passive nuclear power plant accidents by breaking down the complex overall simulation into a modular process and using simplified relationships from the standard framework and the simplified natural removal model to replace complex mechanistic models.

Description

Technical Field

[0001] This application relates to the field of nuclear power accident analysis technology, and in particular to methods, apparatus, equipment, media and procedures for analyzing the consequences of radioactivity. Background Technology

[0002] The analytical techniques for the radiological consequences of severe accidents at passive nuclear power plants have evolved from early deterministic safety analysis to probabilistic safety assessment (PSA), and have gradually introduced integrated system programs to simulate accident processes. With the deepening of the passive safety concept, related analyses are developing towards more efficient and refined dynamic simulations of the entire plant.

[0003] Currently, in China, the main approach to analyzing the radiological consequences of severe accidents at passive nuclear power plants is to use integrated severe accident analysis programs, such as MELCOR and MAAP, to simulate accident sequences and assess radioactive source terms. These programs, by coupling modules on thermal-hydraulic processes, core meltdown, and fission product behavior, enable the reproduction of severe accident conditions at passive nuclear power plants.

[0004] However, the current integrated program model is complex and has the problem of huge computational resource consumption: the integrated program model is complex, and it takes several hours to several days to simulate an accident sequence for the whole plant, which is inefficient and makes it difficult to perform large-scale parameter sensitivity analysis and uncertainty analysis.

[0005] Therefore, there is an urgent need for a new method for analyzing the radioactive consequences of passive nuclear power plant accidents, which suffers from high computational resource consumption.

[0006] The above content is only used to help understand the technical solution of this application and does not represent an admission that the above content is prior art. Summary of the Invention

[0007] The main objective of this application is to provide a method, apparatus, equipment, medium, and program product for analyzing the consequences of radioactive events, aiming to solve the technical problem of high computational resource consumption in current methods for analyzing the consequences of radioactive events in passive nuclear power plant accidents.

[0008] To achieve the above objectives, this application proposes a method for analyzing radioactive consequences, the method comprising: Obtain the core accumulation; Based on the standard source term framework, and according to the core accumulation, the time-varying source term release fraction from the core into the containment is determined to obtain the release rate. Using a simplified natural removal model, the environmental release source term varying over time is calculated based on the release rate and containment leakage rate; wherein, the simplified natural removal model includes a predefined set of natural removal coefficients; The environmental release source term is input into the radioactivity dose assessment model to calculate the radioactive consequences.

[0009] In one embodiment, the step of obtaining the core accumulation includes: Based on the state of the reactor operating at full power until the end of its equilibrium cycle life, a rough estimate of the accumulation of core fission products is calculated. The core stockpile is obtained by multiplying the rough estimate of the stockpile by the power uncertainty coefficient.

[0010] In one embodiment, the step of calculating the environmental release source term over time using a natural removal simplified model, based on the release rate and the containment leakage rate, includes: The environmental release source term is obtained by solving a set of differential equations on the change of radioactivity within the containment over time, based on the release rate, the radioactive decay constant of the nuclide, the natural removal coefficient set, and the containment leakage rate.

[0011] In one embodiment, the differential equations relating the radioactivity within the containment to time are as follows: dA1 / dt = R - (λT + λS + L) * A1 dA2 / dt = L * A1 In the formula, A1 represents the radioactivity of the radionuclide within the containment, A2 represents the radioactivity of the radionuclide in the environment, R is the release rate, λT is the radioactive decay constant of the radionuclide, λS is the natural removal coefficient group, and L is the containment leakage rate.

[0012] In one embodiment, the natural removal coefficient group includes the elemental iodine deposition removal coefficient and the radioactive aerosol natural removal coefficient.

[0013] In one embodiment, the step of inputting the environmental release source term into a radioactivity dose assessment model and calculating the radioactive consequences includes: The environmental release source term is input into the radioactivity dose assessment model to calculate the radiation dose that the public may receive at different time periods and distances; wherein, the distance is the distance between the public and the containment vessel.

[0014] Furthermore, to achieve the above objectives, this application also proposes a radioactive consequences analysis device, which includes: The acquisition module is used to obtain the core accumulation amount; The determination module is used to determine, based on the standard source term framework, the release rate from the core to the containment that varies over time corresponding to the core accumulation. The calculation module is used to calculate the environmental release source term as a function of time, based on the release rate and the containment leakage rate, using a natural removal simplified model; wherein the natural removal simplified model includes a predefined set of natural removal coefficients. The consequences assessment module is used to input the environmental release source terms into the radioactivity dose assessment model and calculate the radioactive consequences.

[0015] In addition, to achieve the above objectives, this application also proposes a radioactive consequences analysis device, the device comprising: a memory, a processor, and a computer program stored in the memory and executable on the processor, the computer program being configured to implement the steps of the radioactive consequences analysis method as described above.

[0016] In addition, to achieve the above objectives, this application also proposes a storage medium, which is a computer-readable storage medium, on which a computer program is stored, which, when executed by a processor, implements the steps of the radioactive consequences analysis method described above.

[0017] In addition, to achieve the above objectives, this application also provides a computer program product, which includes a computer program that, when executed by a processor, implements the steps of the radioactive consequences analysis method described above.

[0018] One or more technical solutions proposed in this application have at least the following technical effects: The technical solution of this application first obtains the core accumulation, providing basic source term data for all subsequent calculations; then, based on the standard source term framework, it quickly determines the release rate over time, avoiding complex thermal-hydraulic calculations and realizing modular estimation of the source term release process; next, it uses a simplified model containing a predefined set of natural removal coefficients to calculate the environmental release source term. Compared with the integrated program, this simplified natural removal model can significantly simplify the simulation of complex physical processes within the containment, significantly reducing computational resource consumption and time costs; finally, it inputs the environmental release source term into a general dose model to calculate the radioactive consequences, achieving the goal of rapidly assessing radioactive risks.

[0019] Overall, this application successfully solves the technical problems of high computational resource consumption and low efficiency in current methods for analyzing the radioactive consequences of passive nuclear power plant accidents by breaking down the complex overall simulation into modular processes and replacing complex mechanistic models with simplified relationships from standard frameworks and natural removal simplification models. It provides an efficient and practical alternative for the rapid analysis of the radioactive consequences of severe passive nuclear power plant accidents. Attached Figure Description

[0020] The accompanying drawings, which are incorporated in and form part of this specification, illustrate embodiments consistent with this application and, together with the description, serve to explain the principles of this application.

[0021] To more clearly illustrate the technical solutions in the embodiments of this application or the prior art, the drawings used in the description of the embodiments or the prior art will be briefly introduced below. Obviously, for those skilled in the art, other drawings can be obtained based on these drawings without creative effort.

[0022] Figure 1 This is a flowchart illustrating an embodiment of the radioactive consequences analysis method of this application. Figure 2 This is a flowchart illustrating Embodiment 2 of the radioactive consequences analysis method of this application; Figure 3 This is a flowchart illustrating Embodiment 3 of the radioactive consequences analysis method of this application; Figure 4 This is a flowchart illustrating Example 5 of the radioactive consequences analysis method of this application; Figure 5 This is a flowchart illustrating Example 6 of the radioactive consequences analysis method of this application; Figure 6 This is a flowchart illustrating Example 7 of the radioactive consequences analysis method of this application. Figure 7 This is a schematic diagram of the module structure of the radioactive consequences analysis device according to an embodiment of this application; Figure 8 This is a schematic diagram of the equipment structure of the hardware operating environment involved in the radioactive consequences analysis method in the embodiments of this application.

[0023] The purpose, features, and advantages of this application will be further explained in conjunction with the embodiments and with reference to the accompanying drawings. Detailed Implementation

[0024] It should be understood that the specific embodiments described herein are merely illustrative of the technical solutions of this application and are not intended to limit this application.

[0025] To better understand the technical solution of this application, a detailed description will be provided below in conjunction with the accompanying drawings and specific implementation methods.

[0026] The main solution of this application embodiment is as follows: obtaining the core accumulation amount; based on the standard source term framework, determining the time-varying source term release share from the core to the containment according to the core accumulation amount to obtain the release rate; using a natural removal simplified model, calculating the time-varying environmental release source term according to the release rate and containment leakage rate; wherein, the natural removal simplified model includes a predefined set of natural removal coefficients; inputting the environmental release source term into the radioactivity dose evaluation model to calculate the radioactive consequences.

[0027] In this embodiment, for ease of description, the following description uses a computing terminal as the execution subject.

[0028] Currently, in China, the main approach to analyzing the radiological consequences of severe accidents at passive nuclear power plants is to use integrated severe accident analysis programs, such as MELCOR and MAAP, to simulate accident sequences and assess radioactive source terms. These programs, by coupling modules on thermal-hydraulic processes, core meltdown, and fission product behavior, enable the reproduction of severe accident conditions at passive nuclear power plants.

[0029] However, the current integrated program model is complex and has the problem of huge computational resource consumption: the integrated program model is complex, and it takes several hours to several days to simulate an accident sequence for the whole plant, which is inefficient and makes it difficult to perform large-scale parameter sensitivity analysis and uncertainty analysis.

[0030] Therefore, there is an urgent need for a new method for analyzing the radioactive consequences of passive nuclear power plant accidents, which suffers from high computational resource consumption.

[0031] Based on this, this application provides a solution that involves obtaining the core stockpile; determining the time-varying source term release share from the core to the containment based on the core stockpile according to the standard source term framework, to obtain the release rate; employing a natural removal simplified model, calculating the time-varying environmental release source term based on the release rate and containment leakage rate; wherein the natural removal simplified model includes a predefined set of natural removal coefficients; and inputting the environmental release source term into a radioactivity dose assessment model to calculate the radioactive consequences.

[0032] The technical solution of this application first obtains the core accumulation, providing basic source term data for all subsequent calculations; then, based on the standard source term framework, it quickly determines the release rate over time, avoiding complex thermal-hydraulic calculations and realizing modular estimation of the source term release process; next, it uses a simplified model containing a predefined set of natural removal coefficients to calculate the environmental release source term. Compared with the integrated program, this simplified natural removal model can significantly simplify the simulation of complex physical processes within the containment, significantly reducing computational resource consumption and time costs; finally, it inputs the environmental release source term into a general dose model to calculate the radioactive consequences, achieving the goal of rapidly assessing radioactive risks.

[0033] Overall, this application successfully solves the technical problems of high computational resource consumption and low efficiency in current methods for analyzing the radioactive consequences of passive nuclear power plant accidents by breaking down the complex overall simulation into modular processes and replacing complex mechanistic models with simplified relationships from standard frameworks and natural removal simplification models. It provides an efficient and practical alternative for the rapid analysis of the radioactive consequences of severe passive nuclear power plant accidents.

[0034] It should be noted that the executing entity in this embodiment can be a computing service device with data processing, network communication, and program execution functions, such as a tablet computer, personal computer, or mobile phone, or an electronic device or computing terminal capable of performing the above functions. The following description uses a computing terminal as an example to illustrate this embodiment and the subsequent embodiments.

[0035] Based on this, embodiments of this application provide a method for analyzing radioactive consequences, referring to... Figure 1 , Figure 1 This is a flowchart illustrating the first embodiment of the radioactive consequences analysis method of this application.

[0036] In this embodiment, the radioactive consequences analysis method includes steps S10 to S50: Step S10: Obtain the core accumulation amount; It should be noted that core stock refers to the total amount of radioactive fission products generated and accumulated in the reactor core of a nuclear power plant due to the fission reaction of nuclear fuel. This total amount is usually measured in becquerels (Bq). Core stock is the basic input data for all subsequent calculations of radioactive source terms and consequences.

[0037] In step S10, the computing terminal first needs to determine a baseline core stockpile. This value is typically calculated based on the reactor's design parameters and operating conditions. Specifically, it can be based on the fission product stockpile at maximum full-power operation, considering the fuel cycle reaching equilibrium (i.e., the end of the equilibrium cycle life). To ensure the conservatism of the analysis results, a power uncertainty coefficient can be introduced into this calculation result to obtain a core stockpile for subsequent analysis.

[0038] In one example, for a typical 1000 MW pressurized water reactor nuclear power plant, the total activity of various key fission products, such as cesium-137 and iodine-131, in the core at the end of its equilibrium cycle life is first calculated based on parameters such as thermal power, fuel enrichment, and burnup depth. Then, this total activity is multiplied by a power uncertainty factor, typically 1.02, to obtain the core stockpile for severe accident analysis of the nuclear power plant.

[0039] Step S30: Based on the standard source term framework, determine the time-varying source term release share from the core to the containment according to the core accumulation amount, so as to obtain the release rate. It should be noted that the standard source term framework refers to a predefined, standardized time-series model of the release of radionuclides from the reactor core to the containment. This model divides the severe accident process into different stages and assigns a release fraction for each radionuclide group at each stage; this fraction is a proportionality coefficient between 0 and 1. The release rate refers to the radioactivity released from the reactor core into the containment per unit time, measured in becquerels per hour (Bq / h), and is calculated as the product of the core stock and the derivative of the source term release fraction with respect to time.

[0040] In step S30, the computing terminal selects or constructs a suitable standard source term framework based on the specific design characteristics of the nuclear power plant being evaluated. This framework clarifies how the proportion of different types of radionuclides (such as inert gases, volatile element iodine, semi-volatile metal cesium, and non-volatile ruthenium) released from the reactor core changes over time at different stages of an accident, such as gap release, early pressure vessel failure, and interaction between molten material and concrete. Then, combined with the core accumulation obtained in step S10, the source term release share over time is calculated, thereby obtaining the release rate over time.

[0041] Step S40: Using a simplified natural removal model, the environmental release source term changing over time is calculated based on the release rate and the containment leakage rate; wherein, the simplified natural removal model includes a predefined set of natural removal coefficients; It should be noted that the simplified natural removal model refers to a set of mathematical relationships used to simulate the removal of radioactive materials within a containment facility through passive means (i.e., natural physical and chemical processes without human intervention). The natural removal coefficient set is the core of this model, containing first-order removal coefficients describing the efficiency of different removal mechanisms (such as gravity settling, wall deposition, etc.), expressed in hours per hour (h). -1 The containment leakage rate refers to the rate at which internal gases leak from the containment to the external environment through non-rupture paths such as penetrations and welds under accident conditions. It is also expressed as a first-order removal factor, with units of hourly (h⁻¹). The environmental release source term refers to the activity of radioactive materials that ultimately leak from the containment into the external environment, and it also varies over time.

[0042] In step S40, the release rate obtained in step S30 is used as input, and a predefined set of natural removal coefficients and the containment leakage rate of the nuclear power plant are also obtained. These coefficients together constitute a differential equation describing the change in the total amount of radioactive material within the containment. Solving this differential equation allows for the calculation of the radioactivity remaining within the containment at each moment, as well as the cumulative radioactivity released into the environment through the leakage path, i.e., the environmental release source term. The core idea of ​​this step is that radioactive material within the containment not only decays but is also removed by natural processes, while a portion leaks out.

[0043] Step S50: Input the environmental release source term into the radioactivity dose assessment model and calculate the radioactive consequences.

[0044] It should be noted that the radioactive dose assessment model is a general calculation framework used to calculate the radiation dose received by an individual or group of people after exposure to radioactive materials. This model typically requires input parameters such as the radioactive source term released into the environment, the atmospheric dispersion factor, the exposure pathways (e.g., cloud infiltration, ground deposition, inhalation of internal radiation), and the dose conversion factor.

[0045] In step S50, the environmental release source term calculated in step S40 is used as input, combined with the meteorological conditions of the specific plant site and the atmospheric dispersion factor calculated by the atmospheric dispersion model, and then input into the radiation dose assessment model. This model calculates the corresponding radiation dose, such as the effective dose or thyroid equivalent dose, based on preset public locations (at different distances in downwind direction) and exposure time. This calculation result is used to assess the radiological consequences of the accident on public health.

[0046] In one example, the environmental release source term A2(t) is input into the dose assessment model recommended by the International Atomic Energy Agency. It is assumed that the atmospheric dispersion factor at 500 meters downwind after the accident is 1^10. -4The model calculates the effective dose from inhaling radioactive aerosols if the public remains at that location for 4 hours. The calculation involves integrating A2(t) over time, multiplying it by the atmospheric diffusion factor and respiration rate, and then multiplying it by the appropriate dose conversion factor to obtain the total effective dose value, for example, 50 millisieverts.

[0047] This embodiment provides a method for analyzing radioactive consequences. First, by obtaining the core accumulation, it provides basic source term data for all subsequent calculations. Then, based on the standard source term framework, it quickly determines the release rate over time, avoiding complex thermal-hydraulic calculations and realizing modular estimation of the source term release process. Next, it uses a simplified model containing a predefined set of natural removal coefficients to calculate the environmental release source term. Compared to the integrated program, this simplified natural removal model can significantly simplify the simulation of complex physical processes within the containment, significantly reducing computational resource consumption and time costs. Finally, it inputs the environmental release source term into a general dose model to calculate the radioactive consequences, achieving the goal of rapidly assessing radioactive risk.

[0048] Overall, this application successfully solves the technical problems of high computational resource consumption and low efficiency in current methods for analyzing the radioactive consequences of passive nuclear power plant accidents by breaking down the complex overall simulation into modular processes and replacing complex mechanistic models with simplified relationships from standard frameworks and natural removal simplification models. It provides an efficient and practical alternative for the rapid analysis of the radioactive consequences of severe passive nuclear power plant accidents.

[0049] It should be noted that the applicable objects of this application's radioactive consequence analysis method include, but are not limited to, the following: 1) Nuclear safety regulatory agencies: This radioactive consequence analysis method can be used as an independent verification calculation and validation tool to quickly assess the conservatism and rationality of severe accident analysis reports submitted by nuclear power design units, supporting the review of license application documents; for example, in the review of new passive reactor types such as the Hainan small modular reactor, this radioactive consequence analysis method can be used for rapid verification calculations; 2) Nuclear power design institutes: In the research and development and design optimization stage of new passive reactors, this radioactive consequence analysis method can be used to quickly assess the impact of different safety system design schemes on the consequences of severe accidents, thereby guiding more economical and efficient safety designs; 3) Nuclear power plant operating companies: This radioactive consequence analysis method can be used to verify and optimize operating procedures, introductions to severe accident management, and the formulation of emergency preparedness and response plans, and to assess the timeliness and effectiveness of different emergency intervention measures; 4) Research institutes and universities: This radioactive consequence analysis method can be used for teaching demonstrations and scientific research related to nuclear safety analysis, helping students and researchers understand the basic principles and methods of severe accident source term development, migration, and consequence assessment.

[0050] Based on the first embodiment of this application, in the second embodiment of this application, the content that is the same as or similar to the first embodiment described above can be referred to the above description, and will not be repeated hereafter. Based on this, please refer to... Figure 2 Step S10 includes steps S11 to S12: Step S11: Based on the state of the reactor operating at full power until the end of its equilibrium cycle life, calculate the rough estimated accumulation of core fission products. It should be noted that the rough estimate of the accumulated amount refers to the total radioactivity of fission products in the reactor core calculated under ideal conditions, without considering operational uncertainties, based on the reactor's physical design parameters and rated operating conditions. The end of the equilibrium cycle life refers to the moment when the production of fission products and their consumption through neutron capture and other means reach a dynamic equilibrium after the nuclear fuel has undergone one complete burn-up cycle in the reactor. At this point, the accumulated amount of fission products is usually at a high level.

[0051] In step S11, the computing terminal first needs to retrieve the reactor's design parameters, such as thermal power, fuel enrichment, and operating history. Then, using a pre-stored reactor physics calculation program or a pre-stored standard manual formula, it calculates the radioactivity of key fission product nuclides (such as Kr-85, Cs-137, I-131, Sr-90, etc.) in the reactor core when the reactor operates continuously at maximum full power and reaches the end of its equilibrium cycle life. This calculation result is the rough estimate of the accumulated amount.

[0052] Step S12: Multiply the estimated stock volume by the power uncertainty coefficient to obtain the core stock volume.

[0053] It should be noted that the power uncertainty factor is a numerical factor greater than 1, used to account for various uncertainties such as the actual operating power of the reactor potentially exceeding the design value, fuel manufacturing tolerances, and measurement errors. This ensures that the calculated core stock is sufficiently conservative to cover potential risks in the analysis. A typical value for this factor is 1.02.

[0054] In step S12, the rough estimate of the core stock calculated in step S11 is multiplied by a preset power uncertainty coefficient to obtain a larger and more enveloping final core stock. This final core stock will be used as input for all subsequent source terms and consequences calculations to ensure the relative conservatism of the analysis results.

[0055] In this embodiment, when obtaining the core stockpile, a rough estimate of the stockpile is first calculated based on the conservative state at the end of the reactor's full power balance cycle life. Then, it is multiplied by the power uncertainty coefficient, so that the final core stockpile considers both the upper limit of the reactor's design operation and a variety of uncertain factors. This provides a relatively conservative and reliable input basis for the entire radioactive consequences analysis, ensuring the envelope and credibility of the subsequent calculation results.

[0056] Based on the first embodiment of this application, in the third embodiment of this application, the content that is the same as or similar to that in the first embodiment can be referred to the above description, and will not be repeated hereafter. Based on this, please refer to... Figure 3 Step S40 includes step S41: Step S41: Based on the release rate, the radioactive decay constant of the nuclide, the natural removal coefficient set, and the containment leakage rate, the differential equations of the change of radioactivity within the containment over time are solved to obtain the environmental release source term.

[0057] It should be noted that the radioactive decay constant is a physical constant describing the reduction of a radioactive nuclide through spontaneous decay. Its value is inversely proportional to the nuclide's half-life, and its unit is per hour (h). -1 The system of differential equations describing the time-varying radioactivity within the containment is a first-order linear ordinary differential equation system. Based on the principle of mass balance, this system treats the containment and the environment as two connected nodes, describing the dynamic balance between the input (release) and output (decay, natural removal, leakage) of radionuclides within the containment. Solving this system of differential equations involves obtaining functions of the time-varying radioactivity within the containment and the release source terms to the environment, using analytical or numerical integration methods. The computational terminal pre-stores the decay constants of the radionuclides of interest. Furthermore, the computational terminal pre-stores a system of differential equations containing these radionuclides, where each term represents the contribution of the release rate, the reduction due to decay, the reduction due to natural removal, and the reduction due to leakage, respectively.

[0058] In step S41, the release rate obtained in step S30, the corresponding coefficients in the predefined natural removal coefficient set, and the containment leakage rate are substituted into the differential equation. The computing terminal then solves the differential equation set according to the pre-stored program under given initial conditions (such as zero activity inside the containment at the start of the accident), thereby obtaining the environmental release source term A2(t) that varies with time. The purpose of solving the differential equation is to accurately quantify the containment's effect as a buffer container on the retention of radioactive materials, thereby accurately calculating the source term leaking into the environment.

[0059] In this embodiment, the problem is simplified to solving a system of first-order linear differential equations, and a concise mathematical model is used to quantitatively describe the dynamic changes in radioactivity within the containment. This embodiment clearly distinguishes three different reduction mechanisms: decay, natural removal, and leakage, and performs linear superposition processing on them. This greatly simplifies the calculation process while ensuring relatively conservative calculations, enabling the environmental release source term to be calculated quickly and accurately.

[0060] In one optional implementation, the differential equations relating the radioactivity within the containment to time are as follows: dA1 / dt = R - (λT + λS + L) * A1 dA2 / dt = L * A1 In the formula, A1 represents the radioactivity of the radionuclide within the containment, A2 represents the radioactivity of the radionuclide in the environment, R is the release rate, λT is the radioactive decay constant of the radionuclide, λS is the natural removal coefficient group, and L is the containment leakage rate.

[0061] It should be noted that the first differential equation describes the rate of change of the activity A1 of radionuclides within the containment over time, which is equal to the release rate R from the reactor core into the containment, minus the sum of the amount of radionuclides in the containment, A1 multiplied by the decay constant λT, the natural removal coefficient λS, and the leakage rate L. The second differential equation describes the rate of change of the cumulative activity A2 of radionuclides in the environment over time, which is equal to the activity A1 within the containment multiplied by the leakage rate L, i.e., the leakage rate.

[0062] In this implementation, the system can substitute the corresponding parameters R, λT, λS, and L for each radionuclide of concern into the aforementioned set of differential equations. Then, under the initial conditions A1(0)=0 and A2(0)=0, the system integrates over time t. The resulting series of A2(t) values ​​represents the environmental release source term accumulated by the radionuclide over time. This set of equations clearly reveals the function of the containment as a buffer: radioactive material enters at a rate R and is then removed from the containment at a rate (λT+λS+L), of which only the L*A1 portion leaks into the environment and poses a potential threat to the public. The introduction of the natural removal coefficient λS effectively characterizes the unique natural purification capacity within the containment of a passive nuclear power plant.

[0063] In one example, for cesium-137 in aerosol form, its radioactive decay constant is extremely small and can be approximated as zero over a short period of actual analysis (e.g., 7 days). Using the values ​​from the previous example, assume λS = 0.05 h. -1 L=0.01 h -1Within a certain computational time step, R is a constant of 1 × 10^15 Bq / h. By solving the differential equation, we can calculate that at the end of this time step, A1(t) = R / (λS+L) * (1 - exp(-(λS+L)*t)), and the increment of A2(t) is L multiplied by the average value of A1 and then multiplied by the time step. If there are multiple release stages in the accident process, each stage can be solved piecewise.

[0064] In this embodiment, by clearly defining the specific form of the differential equation, a clear and directly executable mathematical model is provided for this method. This model not only has a clear physical meaning, organically combining the three key aspects of core release, containment removal, and leakage, but also has a simple form, making it easy to solve quickly using conventional mathematical tools. This ensures the transparency and repeatability of the environmental release source term calculation, which is beneficial for the promotion and application of the technology and regulatory review.

[0065] Based on the first embodiment of this application, in the fourth embodiment of this application, the same or similar content as the first embodiment can be referred to the above description, and will not be repeated hereafter.

[0066] Based on this, the natural removal coefficient group includes the elemental iodine deposition removal coefficient and the radioactive aerosol natural removal coefficient.

[0067] It should be noted that the elemental iodine deposition removal coefficient is a rate constant specifically used to describe the removal of gaseous elemental iodine from the containment through physical processes such as adsorption and sedimentation, which adhere to the containment walls, equipment surfaces, or aerosol particles. This coefficient is calculated based on a wall deposition model. The radioactive aerosol natural removal coefficient is a comprehensive coefficient used to describe the rate constant of removal from the gas by small radioactive solid or liquid particles suspended in the containment gas through natural mechanisms such as gravitational settling, inertial impaction, diffusion, and thermophoresis.

[0068] In the pre-constructed simplified model of natural removal, the analysts decomposed the total natural removal coefficient λS into two main components: the elemental iodine deposition removal coefficient λ_iodine and the radioactive aerosol natural removal coefficient λ_aerosol. Different removal coefficients were selected for different forms of radioactive material. For example, for radioactive iodine existing in the form of elemental iodine, the removal mechanism is mainly wall deposition, and the natural removal coefficient λS is set to λ_iodine. For other radionuclides existing in the form of aerosols (such as oxides of cesium and ruthenium), the removal mechanism is mainly natural sedimentation, and λS is set to λ_aerosol. For inert gases, due to their inert chemical properties, they hardly react with or deposit with any substances, and their natural removal coefficient λS is set to 0. For organic iodine, its behavior is between that of elemental iodine and aerosols, but in this simplified model, to maintain simplicity, its natural removal coefficient is usually also set to 0 or a very small value.

[0069] In this embodiment, by explicitly including at least the elemental iodine deposition removal coefficient and the radioactive aerosol natural removal coefficient in the natural removal coefficient group, this method can model different removal mechanisms for radionuclides with different physicochemical forms. This distinction is more accurate than treating all radioactive materials "equally," and it is also simpler than performing fully coupled calculations on all processes in an integrated program, achieving a better balance between computational efficiency and physical accuracy.

[0070] In one optional implementation, the elemental iodine deposition removal coefficient is calculated using a wall deposition model, the calculation formula of which is: λ w = (A * K w ) / V Where, λ w is the elemental iodine deposition removal coefficient, A is the surface area wetted by the spray droplets inside the containment, V is the net free volume of the containment, and K is the iodine deposition removal coefficient. w is the mass transfer coefficient.

[0071] It should be noted that the wall deposition model is a simplified model based on mass transfer theory, used to estimate the rate at which gaseous elemental iodine contacts and is adsorbed and removed from the inner surface of the containment (especially surfaces that may be wetted by spray droplets or water films). In this model, λ w The unit is per hour (h) -1 Mass transfer coefficient K w The mass transfer efficiency, representing the transfer of iodine vapor to the wall, is an empirical or conservative value, measured in meters per hour (m / h). A conservative mass transfer coefficient K can be obtained by consulting relevant nuclear safety guidelines or standards (such as the Standards Review Guidelines (SRP) of the U.S. Nuclear Regulatory Commission).w The wetted surface area A is the total area of ​​the walls, equipment, and other surfaces inside the containment that are covered by liquid, measured in square meters (m²). 2 The wetted surface area A can be determined based on whether containment spraying is involved in the severe accident management strategy or whether liquid film formation due to condensation is considered. Net free volume V is the volume of gas available to occupy inside the containment after removing internal components, and is measured in cubic meters (m³). 3 ).

[0072] In this embodiment, a specific, mass transfer-based wall deposition model formula is used to calculate the elemental iodine removal coefficient, providing a concrete and quantifiable calculation method for the previously vague concept of natural removal. This formula, based on the physical dimensions of the containment and a standardized conservative mass transfer coefficient, ensures high repeatability and verifiability of the calculation results, making it easier for analysts and review agencies to accept and use, thus enhancing the engineering applicability of this method.

[0073] Alternatively, as an optional implementation, the natural removal coefficient of radioactive aerosols can be obtained based on a simplified model of aerosol removal through natural processes in the reactor containment, combined with the characteristics of domestic passive nuclear power plants.

[0074] It should be noted that the "simplified model method for aerosol removal in the reactor containment by natural processes" refers to the method disclosed in the US Nuclear Regulatory Commission's technical report NUREG / CR-6189. This method, through analysis of multiple sets of experimental and theoretical model results, extracts a set of simplified empirical formulas or recommended coefficients related to the accident process for natural removal mechanisms such as gravity settling, thermophoresis, and diffusion. The "characteristics of domestic passive nuclear power plants" mainly include the widespread use of single-layer steel containment structures, passive containment cooling systems, and their specific containment free volume, height, and internal structural design.

[0075] During the design process, analysts can first refer to the NUREG / CR-6189 report to obtain simplified models or baseline values ​​for the natural removal coefficients of radioactive aerosols for a typical pressurized water reactor containment under severe accident conditions. Then, these baseline coefficients are adaptively adjusted to the specific design characteristics of the domestic passive nuclear power plant being evaluated. For example, considering the typically large free volume of the containment in domestic passive nuclear power plants, this may reduce the aerosol concentration per unit volume, thus affecting gravity settling efficiency. Alternatively, considering that the temperature of the inner wall surface decreases after the passive containment cooling system is activated, it may enhance the thermophoresis effect. After these adjustments combined with specific design characteristics, a set of natural removal coefficients for radioactive aerosols applicable to the domestic passive nuclear power plant is finally derived. This set of coefficients may be a constant or an empirical function that varies over time.

[0076] In one example, analysts assessed the Guohe No. 1 nuclear power plant. They first obtained a recommended baseline aerosol natural removal factor of 0.02 h from the NUREG / CR-6189 report. -1 (Primarily considering gravity settling). Because Guohe-1 uses a large, passive steel containment vessel with a free volume V much larger than that of traditional power plants, the aerosol concentration is lower, resulting in reduced gravity settling efficiency. Therefore, the coefficient is adjusted to 0.015 h. -1 Meanwhile, considering that its passive containment cooling system can effectively maintain the low temperature of the steel shell wall after an accident, thus enhancing the thermophoretic effect, a 0.005 h [cooling factor] is added to this system. -1 The correction term makes the total natural removal coefficient 0.02 h. -1 Ultimately, a function (such as a piecewise function) can be constructed to correlate time with the natural removal coefficient. For example, in the initial stage of an accident (0-24 hours), the coefficient is relatively high at 0.025 h due to the significant thermophoretic effect. -1 After the thermophoretic effect weakens in the later stages, the coefficient drops to 0.015 h. -1 .

[0077] In this embodiment, by explicitly stating that the aerosol natural removal coefficient originates from a mature simplified model method and is customized for the characteristics of domestic passive nuclear power plants, the method is ensured to have both a reliable technical foundation and adaptability to the specific designs of different reactor types. This approach avoids directly using the complex mechanism model in the integrated program, which may have insufficient verification, and also avoids simply using a constant unrelated to the accident process. Thus, without significantly increasing computational complexity, the applicability and accuracy of the model for passive nuclear power plant applications are improved.

[0078] Based on any of the above embodiments of this application, in the fifth embodiment of this application, the content that is the same as or similar to any of the above embodiments can be referred to the above description, and will not be repeated hereafter. Based on this, please refer to... Figure 4 Step S50 includes step S51: Step S51: Based on the environmental release source term and atmospheric dispersion factor, the radiation dose that the public may receive at different time periods and distances is calculated using the radioactive dose assessment model; wherein, the distance is the distance between the public and the containment vessel.

[0079] It should be noted that the atmospheric dispersion factor is a physical quantity used to describe the concentration of radioactive material in the air at a specific location downwind after diffusion and dilution in the atmosphere. It takes into account the influence of meteorological conditions such as wind speed, wind direction, and atmospheric stability, and its unit is per cubic meter per second (s / m²). 3Radiation dose is a physical quantity that measures the potential health effects of ionizing radiation, usually expressed as effective dose (unit: sieverts, Sv) or organ equivalent dose. Different time periods refer to different time intervals after the start of an accident, such as 0-4 hours, 4-24 hours, 24-96 hours, etc.; different distances refer to different radial distances along the downwind direction or in various directions centered on the reactor containment vessel, such as 500 meters, 1 kilometer, 5 kilometers, etc.

[0080] In step S51, the computing terminal first obtains the conservative atmospheric dispersion factor of a specific plant site under specific meteorological conditions, or a set of atmospheric dispersion factors obtained from a probabilistic safety assessment based on multi-year meteorological data of the plant site. These atmospheric dispersion factors can be pre-existing in the computing terminal or manually input by the analyst before the calculation process. Then, the computing terminal combines the environmental release source term A2(t) obtained in step S40 with these atmospheric dispersion factors. The radioactive dose assessment model, based on a preset public exposure scenario (such as continuous stay in a certain location or evacuation within a certain time period), multiplies the environmental release source term in time segments (e.g., each 24-hour period) with the atmospheric dispersion factor for the corresponding time period, and then multiplies it by parameters such as respiration rate and dose conversion factor, thereby calculating the radiation dose that the public may receive at the corresponding distance within the corresponding time period.

[0081] As an optional implementation method, the radioactive dose evaluation model may include formulas for calculating the effective dose of external irradiation, formulas for calculating the effective dose of internal irradiation, and formulas for calculating the thyroid dose.

[0082] Assuming the environment is within a semi-infinite plume, the effective external radiation dose is calculated using the following formula:

[0083] In the formula, This is the effective dose for external irradiation; The effective dose conversion factor for external irradiation of radionuclide i; Let be the amount of radioactivity released by nuclide i during time period j. The atmospheric dispersion factor within time period j; The formula for calculating the effective dose of internal irradiation is as follows:

[0084] In the formula, This is the effective dose for internal irradiation; The internal irradiation effective dose conversion factor for radionuclide i; The amount of radioactivity released by nuclide i during time period j; The respiratory rate during time period j; denoted as the atmospheric dispersion factor (s / m³) within time period j.

[0085] The formula for calculating thyroid dosage is as follows:

[0086] In the formula, This is the thyroid dose; It is the thyroid dose-converting factor for radionuclide i; The amount of radioactivity released by nuclide i during time period j; The respiratory rate during time period j; denoted as the atmospheric dispersion factor (s / m³) within time period j.

[0087] In this embodiment, by combining the environmental release source term with the atmospheric dispersion factor and dose model, and explicitly calculating the public dose at different times and distances, the results of the source term calculation are directly transformed into the consequence indicators most important to nuclear safety decision-making. This spatiotemporal resolution calculation result can provide direct and quantitative technical basis for the delineation of emergency response areas, the selection of evacuation timing, and the formulation of iodine tablet distribution strategies, fully demonstrating the application value of this method in the field of nuclear power plant emergency preparedness and response.

[0088] Based on any of the above embodiments of this application, in the sixth embodiment of this application, the content that is the same as or similar to any of the above embodiments can be referred to the above description, and will not be repeated hereafter. Based on this, please refer to... Figure 5 Before step S30, the radioactive consequences analysis method further includes step S20: Step S20: Process the calculation results of multiple typical pressurized water reactor nuclear power plants and multiple typical accident sequences to obtain the standard source term framework.

[0089] It should be noted that a typical pressurized water reactor (PWR) nuclear power plant refers to a representative second- or third-generation PWR nuclear power plant with a large dry containment structure, such as AP1000 and Hualong One. A typical accident sequence refers to a severe accident scenario that dominates the risk contribution in probabilistic safety assessments, such as a loss-of-coolant accident combined with passive safety system failure or a plant-wide power outage. Processing these calculation results involves collecting and analyzing the time history data of core release fractions obtained after simulating these nuclear power plants and accident sequences using integrated programs such as MELCOR and MAAP, and then summarizing, statistically analyzing, and encapsulating this data.

[0090] Furthermore, it should be noted that step S20 is an offline construction step, completed before the actual radioactive consequence analysis. The standard source term framework is pre-stored in the computing terminal and can be directly invoked during online analysis, without needing to repeat step S20.

[0091] In step S20, as a preparatory step for implementing this method, analysts first collect detailed calculation results from multiple typical pressurized water reactor nuclear power plants for multiple typical severe accident sequences from publicly available literature, collaborative projects, or existing research results. Then, these results are systematically analyzed to extract typical value ranges and trends of the proportions of various radionuclide groups (such as inert gases, halogens, alkali metals, volatile and non-volatile fission products) released from the reactor core to the containment at different accident stages (such as gap release, instantaneous release during pressure vessel failure, and release from the reaction of molten material with concrete). Finally, based on the principles of conservatism or best estimation plus uncertainty, these data are summarized into a standardized, phased, and nuclide-specific source term release proportion timetable, forming this standard source term framework.

[0092] In this embodiment, a universally applicable standard source term framework is constructed by pre-processing and analyzing a wide range of computational data. This allows analysts to directly utilize this framework instead of performing complex and time-consuming integrated thermal-hydraulic and core meltdown simulations for specific nuclear power plants to generate source terms. This significantly reduces the computational threshold and time cost of each analysis, laying the foundation for the rapid application of this method.

[0093] Based on any of the above embodiments of this application, in the seventh embodiment of this application, the content that is the same as or similar to any of the above embodiments can be referred to the above description, and will not be repeated hereafter. Based on this, please refer to... Figure 6 Following step S50, the radioactive consequences analysis method further includes steps S60-S70: Step S60: Obtain key parameters; wherein, the key parameters include at least one of the source term release share, the containment leakage rate, the atmospheric dispersion factor, and the removal coefficient adjustment factor of the natural removal coefficient group; It should be noted that key parameters refer to input variables that have a significant impact on the final calculation results of radioactive consequences or involve considerable uncertainty. The source term release share is the release ratio at each stage as defined in the standard source term framework. The containment leakage rate L is a parameter characterizing the containment sealing performance. The atmospheric dispersion factor is a parameter characterizing atmospheric diffusion conditions. The removal coefficient adjustment factor is a multiplicative factor used to scale based on a predefined set of natural removal coefficients. For example, an adjustment coefficient K can be defined such that the actual natural removal coefficient used is λS_used = K * λS_base, and the effectiveness of natural removal can be simulated by adjusting the value of K.

[0094] In step S60, after the baseline radioactive consequences calculation is completed at the computing terminal, a set of key parameters needs to be identified and obtained for further analysis. These parameters may be the source term release fraction used in step S30, the containment leakage rate and atmospheric dispersion factor used in steps S40 and S50, or removal coefficient adjustment factors introduced for the natural removal coefficient set in step S40. The current baseline values ​​of these parameters are obtained, and their possible value ranges and distribution characteristics are determined.

[0095] The values ​​for the natural removal coefficient group are shown in the table below:

[0096] In one example, the calculation terminal obtains the elemental iodine release share from the source term release share. In the baseline calculation, the elemental iodine release share is 5%. Simultaneously, the containment leakage rate L, with a nominal value of 0.01 h⁻¹, is also considered. -1 To investigate the uncertainty of the natural removal effect of aerosols, an adjustment factor K is introduced, with a baseline value of 1. It is assumed that the value ranges from 0.5 to 2, representing that the natural removal effect may be lower or higher than the expected value.

[0097] Step S70: Adjust the key parameters and perform parameter sensitivity analysis and uncertainty analysis to assess the impact of different parameters on the radioactive consequences.

[0098] It should be noted that parameter sensitivity analysis refers to observing the impact of changing the value of a key parameter one by one (e.g., varying its value within its possible range) on the final radiological consequences (such as public dose) in order to identify which parameters are sensitive. Parameter uncertainty analysis, on the other hand, considers the distribution characteristics of multiple parameters simultaneously and uses random sampling methods such as Monte Carlo to quantify the range of uncertainty (e.g., 90% confidence interval) of the calculated radiological consequences.

[0099] In step S70, analysts leverage the speed of this method to perform sensitivity and uncertainty analyses on the key parameters and their value ranges obtained in step S60. For example, in the sensitivity analysis, while keeping other parameters constant, the iodine release fraction can be adjusted sequentially from 5% to 10% and 20%, and the calculation process in steps S10, S30-S50 can be rerun to compare the final public dose under different release fractions, thereby determining the influence weight of the parameter. In the uncertainty analysis, a probability distribution (such as a normal distribution or log-normal distribution) can be defined for each key parameter (e.g., L, K, release fraction). Then, a computer program can automatically perform thousands of sampling calculations, and finally, statistical analysis of all calculation results can be performed to obtain the distribution range of radioactive consequences.

[0100] In one example, the computing terminal used this method to complete 500 calculations within minutes. During the calculations, the containment leakage rate L (following a log-normal distribution) and the removal coefficient adjustment factor K (following a uniform distribution) were simultaneously randomly sampled. The results showed that the final environmental release source term A2(t) fell within a specific range in 90% of cases. Furthermore, sensitivity analysis revealed that the containment leakage rate L was the most critical parameter affecting early (within 24 hours) public dose, while the natural removal coefficient (primarily adjusted by K) had a greater impact on late (after 24 hours) dose. This finding provides clear guidance for power plant optimization design (reducing leakage rates) and emergency response strategies (seizing the early window of opportunity).

[0101] In this embodiment, by leveraging the extremely high computational efficiency of this method, parameter sensitivity analysis and uncertainty analysis are explicitly proposed and implemented. This enables the computing terminal to go beyond simple point estimation calculations and gain a deeper understanding of the degree of influence of different factors on the consequences of an accident and the resulting overall risk range. This not only provides a powerful independent verification tool for nuclear safety regulatory review but also offers richer and more profound decision support information for nuclear power plant design optimization, the development of severe accident management guidelines, and the formulation of emergency plans, fully demonstrating the added value of this method as an advanced analytical tool.

[0102] This application also provides a device for analyzing the consequences of radioactivity; please refer to... Figure 7 The radioactive consequences analysis device includes: Module 10 is used to obtain the core accumulation amount; The determination module 30 is used to determine, based on the standard source term framework, the release rate from the core to the containment that varies over time corresponding to the core accumulation; The calculation module 40 is used to calculate the environmental release source term as a function of time based on the release rate and the containment leakage rate using a natural removal simplified model; wherein the natural removal simplified model includes a predefined set of natural removal coefficients. The consequences assessment module 50 is used to input the environmental release source term into the radioactivity dose assessment model and calculate the radioactive consequences.

[0103] The radioactive consequences analysis device provided in this application, employing the radioactive consequences analysis method described in the above embodiments, can solve the technical problem of high computational resource consumption in current methods for analyzing the radioactive consequences of passive nuclear power plant accidents. Compared with the prior art, the beneficial effects of the radioactive consequences analysis device provided in this application are the same as those of the radioactive consequences analysis method provided in the above embodiments, and other technical features in the radioactive consequences analysis device are the same as those disclosed in the methods of the above embodiments, and will not be repeated here.

[0104] This application provides a radioactive consequences analysis device, which includes: at least one processor; and a memory communicatively connected to the at least one processor; wherein the memory stores instructions executable by the at least one processor, which are executed by the at least one processor to enable the at least one processor to perform the radioactive consequences analysis method in Embodiment 1 above.

[0105] The following is for reference. Figure 8 The diagram illustrates a structural schematic of a radioactive consequence analysis device suitable for implementing embodiments of this application. The radioactive consequence analysis device in the embodiments of this application may include, but is not limited to, mobile terminals such as mobile phones, laptops, digital broadcast receivers, PDAs (Personal Digital Assistants), PADs (Portable Application Description), PMPs (Portable Media Players), in-vehicle terminals (e.g., in-vehicle navigation terminals), and fixed terminals such as digital TVs and desktop computers. Figure 8 The radioactive consequences analysis device shown is merely an example and should not be construed as limiting the functionality and scope of the embodiments of this application.

[0106] like Figure 8As shown, the radioactive consequences analysis device may include a processing unit 1001 (e.g., a central processing unit, a graphics processing unit, etc.) that can perform various appropriate actions and processes according to a program stored in a read-only memory (ROM) 1002 or a program loaded from a storage device 1003 into a random access memory (RAM) 1004. The RAM 1004 also stores various programs and data required for the operation of the radioactive consequences analysis device. The processing unit 1001, ROM 1002, and RAM 1004 are interconnected via a bus 1005. An input / output (I / O) interface 1006 is also connected to the bus. Typically, the following systems can be connected to the I / O interface 1006: input devices 1007 including, for example, a touchscreen, touchpad, keyboard, mouse, image sensor, microphone, accelerometer, gyroscope, etc.; output devices 1008 including, for example, a liquid crystal display (LCD), speaker, vibrator, etc.; storage devices 1003 including, for example, magnetic tape, hard disk, etc.; and communication devices 1009. Communication device 1009 allows the radioactive consequences analysis equipment to communicate wirelessly or wiredly with other equipment to exchange data. Although the figure shows a radioactive consequences analysis equipment with various systems, it should be understood that it is not required to implement or possess all of the systems shown. More or fewer systems may be implemented alternatively.

[0107] Specifically, according to the embodiments disclosed in this application, the processes described above with reference to the flowcharts can be implemented as computer software programs. For example, embodiments disclosed in this application include a computer program product comprising a computer program carried on a computer-readable medium, the computer program containing program code for performing the methods shown in the flowcharts. In such embodiments, the computer program can be downloaded and installed from a network via a communication device, or installed from storage device 1003, or installed from ROM 1002. When the computer program is executed by processing device 1001, it performs the functions defined in the methods of the embodiments disclosed in this application.

[0108] The radioactive consequences analysis equipment provided in this application, employing the radioactive consequences analysis method described in the above embodiments, can solve the technical problem of high computational resource consumption in current methods for analyzing the radioactive consequences of passive nuclear power plant accidents. Compared with the prior art, the beneficial effects of the radioactive consequences analysis equipment provided in this application are the same as those of the radioactive consequences analysis method provided in the above embodiments, and other technical features of this radioactive consequences analysis equipment are the same as those disclosed in the previous embodiment method, and will not be repeated here.

[0109] It should be understood that the various parts disclosed in this application can be implemented using hardware, software, firmware, or a combination thereof. In the description of the above embodiments, specific features, structures, materials, or characteristics can be combined in any suitable manner in one or more embodiments or examples.

[0110] The above description is merely a specific embodiment of this application, but the scope of protection of this application is not limited thereto. Any variations or substitutions that can be easily conceived by those skilled in the art within the scope of the technology disclosed in this application should be included within the scope of protection of this application. Therefore, the scope of protection of this application should be determined by the scope of the claims.

[0111] This application provides a computer-readable storage medium having computer-readable program instructions (i.e., a computer program) stored thereon, the computer-readable program instructions being used to execute the radioactive consequences analysis method in the above embodiments.

[0112] The computer-readable storage medium provided in this application may be, for example, a USB flash drive, but is not limited to, electrical, magnetic, optical, electromagnetic, infrared, or semiconductor systems, devices, or any combination thereof. More specific examples of computer-readable storage media may include, but are not limited to: electrical connections with one or more wires, portable computer disks, hard disks, random access memory (RAM), read-only memory (ROM), erasable programmable read-only memory (EPROM or flash memory), optical fibers, portable compact disk read-only memory (CD-ROM), optical storage devices, magnetic storage devices, or any suitable combination thereof. In this embodiment, the computer-readable storage medium may be any tangible medium containing or storing a program that can be used by or in conjunction with an instruction execution system, system, or device. The program code contained on the computer-readable storage medium may be transmitted using any suitable medium, including but not limited to: wires, optical cables, RF (Radio Frequency), etc., or any suitable combination thereof.

[0113] The aforementioned computer-readable storage medium may be included in the radioactive consequences analysis device; or it may exist independently and not be assembled into the radioactive consequences analysis device.

[0114] The aforementioned computer-readable storage medium carries one or more programs that, when executed by the radioactive consequences analysis device, cause the radioactive consequences analysis device to: acquire the core stockpile; determine, based on the standard source term framework and the core stockpile, the time-varying share of source terms released from the core into the containment to obtain the release rate; employ a natural removal simplified model to calculate, based on the release rate and the containment leakage rate, the time-varying environmental release source terms; wherein the natural removal simplified model includes a predefined set of natural removal coefficients; and input the environmental release source terms into a radioactivity dose assessment model to calculate the radioactive consequences.

[0115] Computer program code for performing the operations of this application can be written in one or more programming languages ​​or a combination thereof, including object-oriented programming languages ​​such as Java, Smalltalk, and C++, and conventional procedural programming languages ​​such as the "C" language or similar programming languages. The program code can be executed entirely on the user's computer, partially on the user's computer, as a standalone software package, partially on the user's computer and partially on a remote computer, or entirely on a remote computer or server. In cases involving remote computers, the remote computer can be connected to the user's computer via any type of network—including a Local Area Network (LAN) or a Wide Area Network (WAN)—or can be connected to an external computer (e.g., via the Internet using an Internet service provider).

[0116] The flowcharts and block diagrams in the accompanying drawings illustrate the architecture, functionality, and operation of possible implementations of systems, methods, and computer program products according to various embodiments of this application. In this regard, each block in a flowchart or block diagram may represent a module, segment, or portion of code containing one or more executable instructions for implementing a specified logical function. It should also be noted that in some alternative implementations, the functions indicated in the blocks may occur in a different order than those indicated in the drawings. For example, two consecutively indicated blocks may actually be executed substantially in parallel, and they may sometimes be executed in reverse order, depending on the functions involved. It should also be noted that each block in the block diagrams and / or flowcharts, and combinations of blocks in the block diagrams and / or flowcharts, can be implemented using a dedicated hardware-based system that performs the specified function or operation, or using a combination of dedicated hardware and computer instructions.

[0117] The modules described in the embodiments of this application can be implemented in software or hardware. The names of the modules do not necessarily limit the functionality of the unit itself.

[0118] The readable storage medium provided in this application is a computer-readable storage medium that stores computer-readable program instructions (i.e., a computer program) for executing the above-described radioactive consequences analysis method. This solves the technical problem of high computational resource consumption in current methods for analyzing the radioactive consequences of passive nuclear power plant accidents. Compared with the prior art, the beneficial effects of the computer-readable storage medium provided in this application are the same as those of the radioactive consequences analysis method provided in the above embodiments, and will not be repeated here.

[0119] This application also provides a computer program product, including a computer program that, when executed by a processor, implements the steps of the radioactive consequences analysis method described above.

[0120] The computer program product provided in this application can solve the technical problem of high computational resource consumption in current methods for analyzing the radioactive consequences of passive nuclear power plant accidents. Compared with the prior art, the beneficial effects of the computer program product provided in this application are the same as those of the radioactive consequences analysis method provided in the above embodiments, and will not be repeated here.

[0121] The above description is only a part of the embodiments of this application and does not limit the patent scope of this application. All equivalent structural transformations made under the technical concept of this application and using the contents of the specification and drawings of this application, or direct / indirect applications in other related technical fields, are included in the patent protection scope of this application.

Claims

1. A method of radiological consequence analysis, characterized in that, Applied to nuclear power plant accidents, the method includes: Obtain the core accumulation; Based on the standard source term framework, and according to the core accumulation, the time-varying source term release fraction from the core into the containment is determined to obtain the release rate. Using a simplified natural removal model, the environmental release source term varying over time is calculated based on the release rate and containment leakage rate; wherein, the simplified natural removal model includes a predefined set of natural removal coefficients; The environmental release source term is input into the radioactivity dose assessment model to calculate the radioactive consequences.

2. The method as described in claim 1, characterized in that, The step of obtaining the core accumulation includes: Based on the state of the reactor operating at full power until the end of its equilibrium cycle life, a rough estimate of the accumulation of core fission products is calculated. The core stockpile is obtained by multiplying the rough estimate of the stockpile by the power uncertainty coefficient.

3. The method as described in claim 1, characterized in that, The step of calculating the environmental release source term over time using a simplified natural removal model, based on the release rate and containment leakage rate, includes: The environmental release source term is obtained by solving a set of differential equations on the change of radioactivity within the containment over time, based on the release rate, the radioactive decay constant of the nuclide, the natural removal coefficient set, and the containment leakage rate.

4. The method as described in claim 3, characterized in that, The differential equations governing the change of radioactivity within the containment over time are as follows: dA1 / dt = R - (λT + λS + L) * A1 dA2 / dt = L * A1 In the formula, A1 represents the radioactivity of the radionuclide within the containment, A2 represents the radioactivity of the radionuclide in the environment, R is the release rate, λT is the radioactive decay constant of the radionuclide, λS is the natural removal coefficient group, and L is the containment leakage rate.

5. The method as described in claim 1, characterized in that, The natural removal coefficient group includes the elemental iodine deposition removal coefficient and the radioactive aerosol natural removal coefficient.

6. The method according to any one of claims 1 to 5, characterized in that, The step of inputting the environmental release source term into the radioactivity dose assessment model and calculating the radioactive consequences includes: Based on the environmental release source term and atmospheric dispersion factor, the radiation dose that the public may receive at different times and distances is calculated using the radioactive dose assessment model; wherein, the distance is the distance between the public and the containment vessel.

7. A device for analyzing the consequences of radioactivity, characterized in that, The device includes: The acquisition module is used to obtain the core accumulation amount; The determination module is used to determine, based on the standard source term framework, the release rate from the core to the containment that varies over time corresponding to the core accumulation. The calculation module is used to calculate the environmental release source term as a function of time, based on the release rate and the containment leakage rate, using a natural removal simplified model; wherein the natural removal simplified model includes a predefined set of natural removal coefficients. The consequences assessment module is used to input the environmental release source terms into the radioactivity dose assessment model and calculate the radioactive consequences.

8. A device for analyzing the consequences of radioactivity, characterized in that, The device includes: a memory, a processor, and a computer program stored in the memory and executable on the processor, the computer program being configured to implement the steps of the radioactive consequences analysis method as described in any one of claims 1 to 6.

9. A storage medium, characterized in that, The storage medium is a computer-readable storage medium, and a computer program is stored on the storage medium. When the computer program is executed by a processor, it implements the steps of the radioactive consequences analysis method as described in any one of claims 1 to 6.

10. A computer program product, characterized in that, The computer program product includes a computer program that, when executed by a processor, implements the steps of the radioactive consequences analysis method as described in any one of claims 1 to 6.

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