Method and device for calculating dynamic distribution of radioactivity of nuclear island plant under severe accident condition

Abstract

The application discloses a method and device for calculating dynamic distribution of radioactivity of a nuclear island plant under a severe accident condition, and belongs to the field of nuclear power. A nuclear island plant model is divided into multiple elements, and free volumes of each element are calculated; radioactive source items of a gas phase release of a containment and a liquid phase release of a pit are determined according to an accident sequence; a mass transfer path between associated elements is determined, and an air exchange flow matrix is established; equivalent migration flow between non-associated elements is calculated according to atmospheric dispersion parameters; dynamic differential equation groups of changes of radionuclide radioactivity in each element with time are established, evolution curves of radioactivity of each radionuclide are solved, and then instantaneous volume source strength of gamma rays is calculated. The method can effectively predict spatial distribution of radioactivity under a severe accident condition, and supports optimization of an accident decision.

Description

Technical Field

[0001] This invention belongs to the field of nuclear power, specifically relating to a method and apparatus for calculating the dynamic distribution of radioactivity in a nuclear island plant under severe accident conditions. Background Technology

[0002] In severe nuclear power plant accidents, such as core meltdown, a large number of radionuclides are released from damaged fuel into the containment cavity. As the accident evolves, if the containment fails through the reactor core or leaks from dedicated safety facilities outside the containment, the radionuclides will further diffuse into critical areas of the nuclear island, including the primary circuit auxiliary buildings, main control building, and fuel storage buildings. This not only leads to a significant increase in gamma radiation levels in localized areas, threatening the radiation safety of on-duty or standby personnel, but may also affect the accessibility of emergency facilities and the operating environment, thus hindering the effective implementation of accident mitigation measures. Therefore, accurate prediction of the dynamic distribution of radioactivity in the building under severe accident conditions is crucial for developing effective emergency response measures and improving the safety management level of nuclear power plants. Currently, existing analyses of radioactive migration behavior in building structures typically employ simplified modeling methods, treating radioactive release as a direct leakage process along a single path and in a single time period, which cannot effectively assess radiation levels in critical areas. Therefore, providing a more refined method for calculating the dynamic distribution of radioactivity in nuclear island buildings under severe accident conditions is of positive significance for improving the safety management capabilities of nuclear power plants. Summary of the Invention

[0003] The purpose of this invention is to provide a method for calculating the dynamic distribution of radioactivity in a nuclear island building under severe accident conditions, thereby improving the accuracy of radioactivity distribution prediction under such conditions. This invention also provides a device for calculating the dynamic distribution of radioactivity in a nuclear island building under severe accident conditions.

[0004] According to one embodiment of the present invention, a method for calculating the dynamic distribution of radioactivity in a nuclear island building under severe accident conditions is provided, the method comprising the following steps:

[0005] Step a): Provide a nuclear island building model and divide the nuclear island building model into multiple volumes. Set the spatial distribution gradient of radioactive materials within each volume to 0. Each volume is associated with at least one adjacent volume. Calculate the free volume of each volume after removing the space occupied by structural components and equipment.

[0006] Step b): Provide an accident sequence, and calculate the radioactive source terms for the containment gas phase release and the sump liquid phase release respectively based on the accident sequence;

[0007] Step c): Provide ventilation design and leakage paths for the nuclear island building; combine the nuclear island building model to determine the mass transfer paths between the associated volume elements and establish an air exchange flow matrix between the associated volume elements; provide atmospheric dispersion parameters and calculate the equivalent migration flow through atmospheric dispersion between non-associated volume elements.

[0008] Step d): For the volume elements other than those corresponding to the containment, establish a set of dynamic differential equations for the radioactivity of the nuclides in the volume elements as a function of time, based on the inflow from the associated volume elements, the outflow to other volume elements, and the effects of radioactive decay, wherein the coefficients of the dynamic differential equations change with time; calculate the evolution curve of the activity of each nuclide in each volume element over time based on the set of dynamic differential equations, and calculate the instantaneous volume source intensity of the gamma rays.

[0009] This method, by establishing a summation method of dual release sources, multi-volumetric system, multi-path migration and variable coefficient solution, can effectively solve the shortcomings of existing technologies such as poor spatial resolution, single nuclide release mechanism under investigation, and disconnect between internal and external models. It can effectively improve the accuracy and completeness of radiation safety assessment of severe accidents in nuclear power plants, and make the prediction results of simulation calculations more consistent with actual operating conditions.

[0010] Furthermore, in some embodiments, the body element includes an external environment body element, a safety housing body element, a primary loop auxiliary system building body element, a main control building body element, a main steam pipe corridor body element, a fuel storage building body element, and a secondary control room and its connecting corridor body element.

[0011] Furthermore, in some embodiments, the method for calculating the free volume of the volume element includes: calculating the net building space size of each volume element according to the nuclear island plant model, and then deducting the equipment volume according to the equipment layout information within each volume element;

[0012] or,

[0013] The volume percentage method is used to estimate the equipment occupancy rate.

[0014] Furthermore, in some embodiments, in step b), the method for calculating the radioactive source term released from the containment gas phase is as follows:

[0015] ;

[0016] Among them, A 2,i (t) represents the activity of nucleon i in the containment nucleus at time t, Q i Let F be the initial activity of nuclide i in the reactor core, k be the time period of core release, and F be the initial activity of nuclide i in the reactor core. i,k ΔT represents the fraction of nuclide i released from the reactor core during time period k; t0 is the initial time of time period k, t1 is the final time of time period k, and ΔT is the fraction of nuclide i released from the reactor core during time period k. kLet λ be the time step of time interval k. i,d Let λ be the decay constant of nuclide i. i,e T is the removal factor for nuclide i. 2,3 (t) represents the containment leakage rate, and V2 represents the free volume of the containment.

[0017] Furthermore, in some embodiments, in step b), the method for calculating the radioactive source term released from the sump liquid phase is as follows:

[0018] The calculation assumes that all nuclides except inert gases completely enter the containment sump water and participate in the recirculation and re-leakage process of the dedicated safety facilities. The radioactive source term released from the sump liquid phase is calculated based on the vapor-water partitioning effect of nuclides from the liquid phase to the gas phase. The vapor-water partitioning coefficient of iodine nuclides is 0.1, and the vapor-water partitioning coefficient of alkali metals and other non-volatile aerosols is 0.01.

[0019] Furthermore, in some embodiments, in step c), the atmospheric dispersion parameters include atmospheric dispersion factors calculated based on historical meteorological statistics of the nuclear power plant location and the influence of building wakes under different wind directions, and the method for calculating the equivalent migration flow is as follows:

[0020] ;

[0021] Among them, T 2,j T represents the equivalent migration flow from the corresponding containment element to element j. 2,1 The flow rate from the containment to the external environment. T is the atmospheric dispersion factor from the corresponding containment element to element j. 1,j Let be the inhalation flow rate of volume element j.

[0022] Furthermore, in some embodiments, step c) further includes calculating the nucleus removal coefficient λ of the containment shell. e Steps:

[0023] ;

[0024] Where, λ e denoted as the first-order removal coefficient of aerosols and iodine by the spray droplets of the containment spray system, h as the droplet fall height, F as the spray flow rate, E as the collection efficiency, D as the droplet diameter, and V2 as the free volume of the containment.

[0025] Furthermore, in some embodiments, in step d), the dynamic differential equation system takes the form of:

[0026] ;

[0027] Among them, A j,i(t) represents the activity of element i in element j at time t. m,j (t) represents the migration flow from volume element m to volume element j at time t, V m This represents the volume of the volume element m.

[0028] Furthermore, in some embodiments, in step d), the method for calculating the dynamic differential equation system is as follows:

[0029] The entire accident process is discretized in the time domain, and the equations are matrixed. The matrix eigenvalues ​​are solved within each discretized time step. The final value of the previous time step is used as the initial value of the next time step. The calculation is carried out step by step until the numerical solution of the entire accident cycle is completed.

[0030] Furthermore, in some embodiments, the instantaneous volumetric source intensity of the gamma rays is calculated as follows:

[0031] ;

[0032] Among them, S γ,j,g (t) represents the instantaneous source intensity of γ-rays generated by the g-th energy group within volume element j at time t, Num represents the total number of nuclides, and E i γ,g Let g be the gamma-ray energy spectrum corresponding to the g-th energy group of nuclide i.

[0033] According to another embodiment of the present invention, a device for calculating the dynamic distribution of radioactivity in a nuclear island plant under severe accident conditions is provided, comprising a memory and a processor, wherein the memory stores a dynamic distribution calculation program for radioactivity, the dynamic distribution calculation program for radioactivity including a gamma energy group database, an input module, a calculation engine, and an output module.

[0034] The input module is capable of receiving the data provided in steps a) to c) of the calculation method for the dynamic distribution of radioactivity in a nuclear island plant under severe accident conditions provided in any of the foregoing embodiments; the calculation engine includes a program for solving a system of differential equations with variable coefficients; and the output module is capable of individually outputting the activity curves of different nuclides at different spatial locations over time based on the calculation results of the calculation engine.

[0035] When the data provided in steps a) to c) of the method for calculating the dynamic distribution of radioactivity in a nuclear island building under severe accident conditions provided in any of the foregoing embodiments is input into the input module, and when the dynamic distribution of radioactivity calculation program is executed by the processor, the processor is able to calculate and output the instantaneous volume source strength of γ-rays in step d) of the method for calculating the dynamic distribution of radioactivity in a nuclear island building under severe accident conditions provided in any of the foregoing embodiments. Attached Figure Description

[0036] Figure 1 This is a flowchart of a method for calculating the dynamic distribution of radioactivity in a nuclear island building, as shown in one embodiment.

[0037] Figure 2 This is a schematic diagram of a containment leakage radionuclide migration model in one embodiment;

[0038] Figure 3 This is a schematic diagram of the migration of leaked radionuclides in an ESF recirculation system in one embodiment;

[0039] Figure 4 This is a curve showing the variation of the total source intensity of gamma rays in the main steam tunnel, calculated in one embodiment.

[0040] The purpose of the above figures is to provide a detailed description of the invention so that those skilled in the art can understand the technical concept of the invention, and not to limit the invention. Detailed Implementation

[0041] The present invention will now be described in further detail with reference to specific embodiments and accompanying drawings.

[0042] The term "embodiment" as used herein means that a particular feature, structure, or characteristic described in connection with an embodiment may be included in at least one embodiment herein. The phrase appearing in various places in the specification does not necessarily refer to the same embodiment, nor is it limited to mutually exclusive, independent, or alternative embodiments. Those skilled in the art will understand that the embodiments herein can be combined with other embodiments without causing structural conflicts.

[0043] In this article, "multiple" means at least two.

[0044] Currently, simplified models are commonly used to analyze and simulate radioactive migration behavior in serious accidents such as nuclear power plant core meltdowns. A typical approach is to treat radioactive release as a direct leakage process with a single path and a single time period, and to assess radioactivity levels based on treating all buildings around the containment as a uniformly mixed closed area. Existing analytical methods have the following shortcomings: Lack of spatial resolution; different buildings have different functions and varying degrees of physical isolation, and the analysis and simulation ignore structural separation, ventilation differences, and local high-dose hotspots between buildings, easily leading to distorted assessments of radiation levels in key areas; incomplete migration path modeling, considering only direct airflow paths between the containment building and adjacent buildings, ignoring indirect migration mechanisms through ventilation shafts, entrances, and exits; insufficient model coupling; traditional atmospheric dispersion models, such as Gaussian plume models, are mainly used for off-site environmental impact assessments and the habitability analysis of emergency facilities. While they can estimate the atmospheric diffusion concentration of radioactive materials around the plant area, they lack dynamic data interfaces with multiple systems within the plant, making it difficult to achieve closed-loop simulations of leakage-reabsorption-internal redistribution; and static migration models; most existing general-purpose or specialized commercial software uses steady-state or quasi-static assumptions, failing to reflect the time-varying characteristics of leakage rates, atmospheric dispersion factors, and other adoption factors during accidents, making it difficult to support dynamic risk assessments.

[0045] To overcome the aforementioned shortcomings of the prior art, one embodiment of the present invention provides a method for calculating the dynamic distribution of radioactivity in a nuclear island building under severe accident conditions. The method is as follows: Figure 1 As shown, it includes the following steps:

[0046] Step a): Construct a multi-volumetric model and define the free volume of each voxel.

[0047] A digital model of the nuclear island plant to be analyzed and simulated is provided. Based on the actual layout and ventilation structure characteristics of the nuclear island, the nuclear island plant model is divided into multiple volumes. The spatial distribution gradient of radioactive materials in each volume is regarded as 0 (i.e., according to the complete mixing assumption, it is assumed that the volume is uniformly distributed). Each volume is interconnected with at least one adjacent volume.

[0048] In a preferred embodiment, the volume elements can be divided into: external environment volume element - volume element 1; containment volume element - volume element 2, including the containment gas space region (Contain) and the sump water region (Sump), where gas and liquid phase sources coexist; primary loop auxiliary system building volume elements - volumes 3 to 6, including areas at different elevations of the primary loop auxiliary system building (Aux-21, Aux-18, Aux-0, and Aux-B), with layered architecture reflecting vertical concentration differences; main control building (Croom) volume element - volume element 7; main steam pipe gallery (Sroom) volume element - volume element 8; fuel storage building (Froom) volume element - volume element 9; and auxiliary control room and its connecting corridor (Aroom) volume element - volume element 10. In other embodiments, depending on the nuclear island structure, the number of volume elements corresponding to the primary loop auxiliary system building can also be set to other numbers.

[0049] To accurately reflect the capacity and dilution effect of each region on radioactive materials, it is necessary to reasonably calculate the effective free volume of each volume element, that is, the available air volume after deducting the space occupied by structural components and equipment. In a preferred embodiment, the calculation method is as follows:

[0050] The building clearance dimensions of each volume element are extracted based on the 3D model of the nuclear island plant, and the volume occupied by equipment such as cabinets, pumps, storage tanks, and pipe racks is further deducted according to the equipment layout drawing. Alternatively, the "volume proportion method" can be used to estimate the equipment obstruction rate based on simulation or experience, with typical values ​​ranging from 5% to 25%. For example, in Croom, where equipment is dense, the obstruction rate is estimated to be 25%.

[0051] The multi-body model not only reflects the spatial isolation relationship of the nuclear island building, but also provides a basic topological structure for the subsequent establishment of migration equations.

[0052] Step b): Establish a dual-release source model.

[0053] To comprehensively describe the initial source of contamination, a full and accurate simulation of the radioactive source term is required, characterizing the initial inputs at different physical locations and release mechanisms.

[0054] The first type of source term is containment gas phase release.

[0055] During a severe accident, core meltdown leads to the release of a large number of radionuclides into the containment gas space (Volume 2). This area serves as the primary initial source of contamination, with its radioactivity dynamically evolving over time and continuously leaking into the external environment through gaps in the penetrations. Referring to the standard source terms recommended in the NUREG-1465 report published by the U.S. Nuclear Regulatory Commission (NRC), the severe accident process can be divided into four stages: interstitial release, early container release, external container release, and late container release. For each nuclide group, such as inert gases, iodine, alkali metals, tellurium, rare metals, ruthenium, barium, and strontium, each stage has a different release share and rate, while the initial total amount is determined based on the core radioactive accumulation at the end of the equilibrium cycle life. Specifically, iodine speciation is further subdivided into particulate iodine, elemental iodine, and organic iodine, each assigned a weighting percentage (e.g., 95%, 4.85%, 0.15%) to differentiate the varying responses of different iodine speciation in subsequent migration and removal, providing a basis for refined modeling. In one embodiment, a schematic diagram of the migration path of gas-phase released radionuclides from the containment structure (body element 2) to the various other body elements is shown below. Figure 2 As shown.

[0056] In a preferred embodiment, the activity of radionuclides released in the gas phase at a certain moment within the containment can be calculated as follows:

[0057] ;

[0058] Among them, A 2,i (t) represents the activity of nucleon i in the containment nucleus at time t, Q i Let F be the initial activity of nuclide i in the reactor core, k be the time period of core release, and F be the initial activity of nuclide i in the reactor core. i,k ΔT represents the fraction of nuclide i released from the reactor core during time period k; t0 is the initial time of time period k, t1 is the final time of time period k, and ΔT is the fraction of nuclide i released from the reactor core during time period k. k Let λ be the time step of time interval k. i,d Let λ be the decay constant of nuclide i. i,e T is the removal factor for nuclide i. 2,3 (t) represents the containment leakage rate, and V2 represents the free volume of the containment.

[0059] The above model comprehensively considers the generation, decay, removal, and leakage processes of nuclides, and can accurately reflect the dynamic changes of source terms within the containment. In other embodiments, a time-varying leakage physical mechanism of the containment can be further introduced, and the dynamically changing containment leakage rate can be calculated based on parameters such as internal and external pressure difference, temperature gradient, and material creep, thereby further improving the accuracy of the simulation calculation.

[0060] The second type of source term is the release of liquid phase from the sump.

[0061] Following the accident, cooling water containing highly radioactive nuclides accumulated in the containment sump, forming a localized liquid-phase radioactive source term. The nuclides in this water could enter the airspace through sealing defects in the Dedicated Safety Facility (ESF) recirculation system, constituting an independent secondary release source, a typical liquid-vapor phase transition migration process. In one embodiment, a schematic diagram of radioactive nuclide migration following an ESF recirculation system leak is shown below. Figure 3 As shown.

[0062] Specifically, based on the core radioactive release fraction defined in NUREG-1465, a conservative assumption is made that, except for inert gases, all other nuclides completely enter the containment sump water and participate in the ESF recirculation and leakage process. In the analysis of the preferred embodiment, the leakage rate of the recirculation system is conservatively taken as twice the maximum permissible leakage rate specified in the technical specifications to encompass uncertainties such as equipment aging and vibration. Under the assumption that no flash evaporation occurs in the leaked liquid, the vapor-water partitioning effect of nuclides from the liquid phase to the gas phase is further considered: for iodine nuclides, the vapor-water partitioning coefficient is taken as 0.1; for alkali metals and other non-volatile aerosols, the vapor-water partitioning coefficient is taken as 0.01.

[0063] The aforementioned dual-release source model describes the initial radioactive distribution of the containment gaseous source and the sump liquid source at different accident stages and spatial locations, ensuring the engineering rationality and universality of the source term input. Based on morphological subdivision and phase transition mechanisms, it effectively improves the physical authenticity of the source term release behavior. The key parameters are set using envelope settings, which can meet the conservative requirements of nuclear safety assessment.

[0064] Step c): Construct a multi-path migration mechanism.

[0065] First, the direct air exchange flow rate between volumetric elements is established. Based on the actual ventilation design, building layout, and leakage paths (i.e., mass transfer paths) of the nuclear island plant, the direct air exchange relationships between each volumetric element are established, and an air exchange flow rate matrix is ​​constructed to characterize the mass transfer capacity between adjacent areas. Types of mass transfer paths include normal ventilation duct supply and exhaust, leakage during door opening or closing, personnel / equipment gate leakage, cable penetration leakage, and pipe penetration leakage. The flow rates of all directly connected paths are expressed as volumetric flow rates (m³ / s). 3 The / h) calculation forms the basic part of the inter-regional mass transfer network and serves as the input to the convection term in the dynamic migration equation.

[0066] Next, the equivalent migration flow rate for non-directly connected volume elements is established. For areas without direct ventilation connection to the containment, such as Croom, Aroom, and Froom, the radioactive input mainly comes from the atmospheric dispersion of leaked materials from the containment through the external environment (volume element 1), entering the indoor space through building air inlets or doors and windows, which is a typical indirect migration path. To accurately simulate the indirect migration path, an equivalent migration flow rate modeling method is introduced to calculate atmospheric dispersion parameters: using an atmospheric dispersion analysis program (such as certified analysis programs like ARCON96), combined with historical meteorological statistics (including wind speed, wind direction, atmospheric stability distribution, etc.) of the simulated nuclear power plant location, the atmospheric dispersion factor from the containment to the receiving point of each volume element is calculated; considering the influence of building wake effect and prevailing wind direction, calculations are performed for different wind angles, and the largest dispersion factor is taken as the input value of the atmospheric dispersion parameter to ensure the conservatism of the assessment results. Based on the atmospheric dispersion factor between the containment and relevant volume elements, the equivalent migration flow rate of the containment leaking into the external environment, after atmospheric dispersion, and then entering other non-directly connected volume elements is calculated:

[0067] ;

[0068] Among them, T 2,j T represents the equivalent migration flow from the corresponding containment element to element j. 2,1 The flow rate from the containment to the external environment. T is the atmospheric dispersion factor from the corresponding containment element to element j. 1,j Let be the inhalation flow rate of volume element j. In other embodiments, a Lagrange particle tracking model can also be used to simulate the three-dimensional trajectory of radioactive materials in the atmosphere, adding physical processes such as building flow around, deposition effects, and turbulent diffusion, thereby improving the accuracy of atmospheric dispersion prediction.

[0069] The equivalent flow rate mentioned above, as an input to the indirect migration path, can effectively reflect the impact of external environmental migration on key areas such as Croom and Aroom. Since the atmospheric dispersion factor changes over time, the equivalent migration flow rate also changes dynamically.

[0070] In a preferred embodiment, to more accurately reflect the decay behavior of radioactivity within the containment, a non-migratory loss mechanism should also be considered, namely, the first-order removal coefficient of the droplets sprayed by the containment sprinkler system for aerosols and iodine. Different removal coefficients are set for different chemical forms of the radionuclides: for iodine, the activity is set to decay exponentially after the sprinkler system is activated, and a decontamination factor threshold is set as the termination condition; for aerosols, a removal system model based on sprinkler flow rate, droplet diameter, collection efficiency, and containment volume is used, and different removal coefficients are set in stages. The removal coefficient λ e The specific calculation method is as follows:

[0071] ;

[0072] Where, λ e denoted as the first-order removal coefficient of aerosols and iodine by the spray droplets of the containment spray system, h as the droplet fall height, F as the spray flow rate, E as the collection efficiency, D as the droplet diameter, and V2 as the free volume of the containment.

[0073] In the above steps, different mass transfer paths are fully covered, ensuring that all potential input channels are identified and quantified; by taking the maximum value of the atmospheric dispersion factor and the envelope value of the leakage rate, the conservative assessment requirements of nuclear safety analysis are met; all flow rates and removal coefficients are output in parameterized form, which can be directly used as input terms for the next step of the dynamic migration differential equation system.

[0074] Step d): Construct and solve the dynamic migration mathematical model.

[0075] For all volume elements except for volume element 2 (corresponding to the containment building), the change in the radioactivity of the nuclides within them over time is determined by a system of differential equations consisting of production and disappearance terms. Taking into account the inflow from associated volume elements, the emission to other volume elements, and the effects of radioactive decay, a dynamic system of differential equations is established to determine the change in the radioactivity of nuclides within a volume element over time. The coefficients in this system of equations change with time. By solving this dynamic system of differential equations, the instantaneous volumetric source intensity of gamma rays within any volume element can be obtained, thus revealing the dynamic distribution of radioactivity within the nuclear island building.

[0076] Specifically, in a preferred embodiment, the dynamic differential equation system takes the form of:

[0077] ;

[0078] Among them, A j,i (t) represents the activity of element i in element j at time t. m,j (t) represents the migration flow from volume element m to volume element j at time t, V m This represents the volume of the volume element m.

[0079] The activities of each element in the above equation system influence each other, forming a strongly coupled system matrix; the migration flux between elements changes dynamically with the stage of the accident; there are other external source terms besides the elements.

[0080] Since the dynamic differential equation system is a non-homogeneous differential equation system and its coefficients change with time, it usually does not have an analytical solution. Therefore, in the preferred embodiment, it is solved by numerical methods. Specifically, the time-step eigenvalue method is used for numerical solution. First, time-domain discretization is performed to divide the entire accident process into multiple fine time steps, with the coefficients (flow rates) remaining unchanged in each step, achieving local constant coefficients. Next, in each time step, the activity variables of all volume elements are used to form a state constant, and the equation is written in matrix form: dA / dt=CA+B, where C represents the coefficient matrix, composed of flow rate, volume, and decay terms between volume elements; B represents the external inputs of other volume elements. In each time step, the characteristic equation is constructed through the characteristic polynomial of the coefficient matrix C, and the eigenvector A* is obtained. The similar upper triangular matrix of matrix C is further determined, and back substitution is performed to complete the analytical solution in that step. Finally, the final value of the previous time step is used as the initial value of the next time step, and the calculation is performed step by step to achieve continuous solution throughout the entire accident cycle. This computational method exhibits higher numerical stability and computational efficiency when dealing with high-dimensional coupled systems, and is particularly suitable for long-term accident evolution simulations. In other embodiments, to further improve computational speed, the equation system can also be reconstructed using a dimensionality-reduced matrix.

[0081] By solving the dynamic differential equations, the evolution curve of the activity of each nuclide in the voxel over time can be obtained. Based on this evolution curve, the instantaneous source intensity of gamma rays can be further calculated.

[0082] ;

[0083] Among them, S γ,j,g (t) represents the instantaneous source intensity of γ-rays generated by the g-th energy group within volume element j at time t, Num represents the total number of nuclides, and E i γ,g Let g be the gamma-ray energy spectrum corresponding to the g-th energy group of nuclide i.

[0084] Instantaneous gamma-ray source intensity data can serve as the basic input for subsequent radiation field simulation and dose assessment, used to delineate post-accident radiation zones, assess the short-term accessibility of personnel in critical passages, calculate the cumulative dose of emergency personnel, and optimize ventilation and protection strategies.

[0085] The calculation steps in the method for calculating the dynamic distribution of radioactivity in a nuclear island building under severe accident conditions provided in the above embodiments can be completed by a computing device provided in another embodiment of the present invention. Specifically, the computing device includes a processor and a memory. The memory stores a program for calculating the dynamic distribution of radioactivity, which includes a gamma energy group database, an input module, a computing engine, and an output module. The gamma energy group database stores 18 groups of gamma energy spectrum data to support the final calculation of the instantaneous source intensity of gamma rays. The input module can support the input and calculation of various input data in steps a) to c), such as core source terms, volumetric parameters, connection flow, leakage rate, atmospheric dispersion factor, and removal coefficient. The computing engine can solve the system of differential equations with varying coefficients and call the data from the gamma energy group database to automatically complete the source intensity calculation. The output module can output the calculation results of the computing engine and output the activity curves of different nuclides in volumetric elements at different spatial locations over time.

[0086] In one embodiment, the calculated output curve of the total gamma-ray source intensity variation within the Sroom is as follows: Figure 4 As shown. The calculated data can be directly imported into programs such as MCNP and JMCT-S for subsequent shielding design.

[0087] The radioactive dynamic distribution calculation program provided in the above embodiments can be implemented by a self-written computer program without relying on general platforms such as MATLAB and COMSOL.

[0088] The purpose of the above embodiments is to provide a further detailed description of the present invention 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, optimization or equivalent substitution of the technical features involved, as well as combination of implementation methods in different embodiments without causing structural and principle conflicts, all fall within the protection scope of the present invention.

Claims

1. A method for calculating the dynamic distribution of radioactivity in a nuclear island building under severe accident conditions, characterized in that, Includes the following steps: Step a): Provide a nuclear island building model and divide the nuclear island building model into multiple volumes. Set the spatial distribution gradient of radioactive materials within each volume to 0. Each volume is associated with at least one adjacent volume. Calculate the free volume of each volume after removing the space occupied by structural components and equipment. Step b): Provide an accident sequence, and calculate the radioactive source terms for the containment gas phase release and the sump liquid phase release respectively based on the accident sequence; Step c): Provide ventilation design and leakage paths for the nuclear island building; combine the nuclear island building model to determine the mass transfer paths between the associated volume elements and establish an air exchange flow matrix between the associated volume elements; provide atmospheric dispersion parameters and calculate the equivalent migration flow through atmospheric dispersion between non-associated volume elements. Step d): For the units other than the corresponding containment unit, establish a set of dynamic differential equations for the time-varying radioactivity of the nuclides in the units, based on the effects of inflow from associated units, outflow to other units, and radioactive decay: ， The coefficients of the dynamic differential equation system change with time, A j,i (t) represents the activity of element i in element j at time t. m,j (t) represents the migration flow from volume element m to volume element j at time t, V m The volume of volume element m is represented; the evolution curve of the activity of each nuclide in each volume element over time is calculated according to the dynamic differential equation system, and the instantaneous volume source intensity of γ-rays is calculated.

2. The method for calculating the dynamic distribution of radioactivity in a nuclear island building under severe accident conditions according to claim 1, characterized in that, The components include the external environment component, the safety housing component, the primary loop auxiliary system building component, the main control building component, the main steam pipe corridor component, the fuel storage building component, and the secondary control room and its connecting corridor component.

3. The method for calculating the dynamic distribution of radioactivity in a nuclear island building under severe accident conditions according to claim 1 or 2, characterized in that, The method for calculating the free volume of the volume element includes: calculating the net building space size of each volume element based on the nuclear island plant model, and then deducting the equipment volume based on the equipment layout information within each volume element; or, The volume percentage method is used to estimate the equipment occupancy rate.

4. The method for calculating the dynamic distribution of radioactivity in a nuclear island building under severe accident conditions according to claim 1, characterized in that, In step b), the method for calculating the radioactive source term released from the containment gas phase is as follows: ； Among them, A 2,i (t) represents the activity of nucleon i in the containment nucleus at time t, Q i Let F be the initial activity of nuclide i in the reactor core, k be the time period of core release, and F be the initial activity of nuclide i in the reactor core. i,k ΔT represents the fraction of nuclide i released from the reactor core during time period k; t0 is the initial time of time period k, t1 is the final time of time period k, and ΔT is the fraction of nuclide i released from the reactor core during time period k. k Let λ be the time step of time interval k. i,d λ is the decay constant of nuclide i. i,e T is the removal factor for nuclide i. 2,3 (t) represents the containment leakage rate, and V2 represents the free volume of the containment.

5. The method for calculating the dynamic distribution of radioactivity in a nuclear island building under severe accident conditions according to claim 1, characterized in that, In step b), the method for calculating the radioactive source term released from the sump liquid phase is as follows: The calculation assumes that all nuclides except inert gases completely enter the containment sump water and participate in the recirculation and re-leakage process of the dedicated safety facilities. The radioactive source term released from the sump liquid phase is calculated based on the vapor-water partitioning effect of nuclides from the liquid phase to the gas phase. The vapor-water partitioning coefficient of iodine nuclides is 0.1, and the vapor-water partitioning coefficient of alkali metals and other non-volatile aerosols is 0.

01.

6. The method for calculating the dynamic distribution of radioactivity in a nuclear island building under severe accident conditions according to claim 1, characterized in that, In step c), the atmospheric dispersion parameters include atmospheric dispersion factors calculated based on historical meteorological statistics of the nuclear power plant location and the influence of building wakes under different wind directions. The method for calculating the equivalent migration flow is as follows: ； Among them, T 2,j T represents the equivalent migration flow from the corresponding containment element to element j. 2,1 The flow rate from the containment to the external environment. T is the atmospheric dispersion factor from the corresponding containment element to element j. 1,j Let be the inhalation flow rate of volume element j.

7. The method for calculating the dynamic distribution of radioactivity in a nuclear island building under severe accident conditions according to claim 1 or 6, characterized in that, Step c) also includes calculating the containment shell resorption coefficient λ. e Steps: ； Where, λ e denoted as the first-order removal coefficient of aerosols and iodine by the spray droplets of the containment spray system, h as the droplet fall height, F as the spray flow rate, E as the collection efficiency, D as the droplet diameter, and V2 as the free volume of the containment.

8. The method for calculating the dynamic distribution of radioactivity in a nuclear island building under severe accident conditions according to claim 1, characterized in that, In step d), the calculation method for the dynamic differential equation system is as follows: The entire accident process is discretized in the time domain, and the dynamic differential equation system is matrixed. The matrix eigenvalues ​​are solved within each discretized time step. The final value of the previous time step is used as the initial value of the next time step. The calculation is carried out step by step until the numerical solution of the entire accident cycle is completed.

9. The method for calculating the dynamic distribution of radioactivity in a nuclear island building under severe accident conditions according to claim 8, characterized in that, The method for calculating the instantaneous volumetric source strength of the gamma rays is as follows: ； Among them, S γ,j,g (t) represents the instantaneous source intensity of γ-rays generated by the g-th energy group within volume element j at time t, Num represents the total number of nuclides, and E i γ,g Let g be the gamma-ray energy spectrum corresponding to the g-th energy group of nuclide i.

10. A calculation device for the dynamic distribution of radioactivity in a nuclear island building under severe accident conditions, comprising a memory and a processor, characterized in that, The memory stores a radioactivity dynamic distribution calculation program, which includes a gamma energy group database, an input module, a calculation engine, and an output module. The input module is capable of receiving the data provided in steps a) to c) of the method for calculating the dynamic distribution of radioactivity in a nuclear island plant under severe accident conditions as described in any one of claims 1 to 9; the calculation engine includes a program for solving a system of differential equations with variable coefficients; and the output module is capable of individually outputting the activity curves of different nuclides at different spatial locations over time based on the calculation results of the calculation engine. When the data provided in steps a) to c) of the method for calculating the dynamic distribution of radioactivity in a nuclear island building under severe accident conditions as described in any one of claims 1 to 9 is input into the input module, and when the dynamic distribution of radioactivity calculation program is executed by the processor, the processor is able to calculate and output the instantaneous volume source intensity of γ-rays in step d) of the method for calculating the dynamic distribution of radioactivity in a nuclear island building under severe accident conditions as described in any one of claims 1 to 9.

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