Simplified off-site emergency management method based on the intrinsic safety of advanced reactors

Abstract

A simplified off-site emergency management method based on the intrinsic safety of advanced reactors, belonging to the nuclear power field, is proposed. This method includes the following steps: identifying accident source terms and calculating the total amount of radioactivity released into the environment under these source terms; conducting atmospheric dispersion analysis based on meteorological conditions to calculate and determine the diffusion path, concentration distribution, and deposition state of radioactive materials in the atmosphere; calculating radiation doses in the early, middle, and late stages of the accident; providing dose control criteria; and implementing off-site emergency protection based on the calculation results and dose control criteria. This method can quantitatively assess the off-site radioactive risk under accident conditions in third-generation advanced nuclear power plants, improve the safety management level of nuclear power plants, and reduce redundancy and waste.

Description

Technical Field

[0001] This invention belongs to the field of nuclear power safety, specifically relating to a simplified off-site emergency management method based on the intrinsic safety of advanced reactors. Background Technology

[0002] Off-site emergency response is a crucial measure to ensure public safety during nuclear power plant operation. However, the traditional design philosophy of off-site emergency response systems for nuclear power plants is relatively conservative, leading to excessive redundancy of emergency resources and the long-term idleness of a large number of off-site emergency facilities, resulting in resource waste. Furthermore, existing off-site emergency management methods lack scientific and reasonable calculation and assessment, and lack targeted designs adapted to the characteristics of third-generation nuclear power plant technologies. Therefore, providing a simplified off-site emergency management method based on the intrinsic safety of advanced reactors is of positive significance for improving the safety management level of nuclear power plants. Summary of the Invention

[0003] The purpose of this invention is to provide a simplified off-site emergency management method based on the intrinsic safety of advanced reactors, thereby improving the safety management level of nuclear power plants.

[0004] According to an embodiment of the present invention, a simplified off-site emergency management method based on the intrinsic safety of advanced reactors is provided, comprising the following steps:

[0005] Step a): Identify the accident source term and calculate the total amount of radioactivity released into the environment under the accident source term;

[0006] Step b): Perform atmospheric dispersion analysis based on meteorological conditions to calculate and determine the diffusion path, concentration distribution, and deposition state of radioactive materials in the atmosphere;

[0007] Step c): Calculate the radiation dose in the early, middle, and late stages of the accident, respectively;

[0008] Step d): Provide dose control guidelines and perform off-site emergency protection based on the calculation results of step c).

[0009] This method can form a simplified management method for off-site emergency response in nuclear power plants through quantitative calculations, improve the rationality and reliability of emergency management plans, optimize the level of safety emergency management in nuclear power plants, and reduce resource waste caused by redundancy of off-site emergency equipment.

[0010] Further, in some embodiments, in step a), the method for determining the accident source term is as follows: the accident source term is determined by deterministic methods and probabilistic safety assessment methods, and the accident source term is determined by a two-parameter threshold criterion, wherein the two-parameter threshold criterion is that the gaseous radioactive release exceeds 500 TBq. 131 The accident sequences and the cumulative frequency of accident sequences at equivalent doses of I are less than 10. -7 / deposit year, accident source items that satisfy the aforementioned two-parameter threshold criterion are excluded. Further, in some embodiments, the method for calculating the total amount of radioactive Q released into the environment in step a) is as follows:

[0011] ;

[0012] ;

[0013] Among them, Q i q represents the activity of nuclides in voxel i; i V represents the rate of radioactive release within volume element i; i F represents the volume or mass of volume element i; i λ is the inflow of volume element i; di λ is the decay constant of the nuclide in volume element i; ri C is the removal coefficient of energetic nuclides in voxel i; i (t) represents the intrinsic activity concentration of volume element i released into the environment at time t; F i (t) represents the mass release rate of volume element i into the environment; n represents the total number of volume elements in the accident analysis.

[0014] Furthermore, in some embodiments, in step b), the meteorological condition probability is a meteorological condition with a probability of 99.5%.

[0015] Furthermore, in some embodiments, in step b), the meteorological condition probability calculation takes into account atmospheric stability, wind speed, wind direction, and precipitation probability.

[0016] Furthermore, in some embodiments, in step c), the calculation items for early-stage radiation dose include: direct external irradiation within the influence range of the radiation plume, external irradiation from ground deposition, external irradiation from body surface deposition, internal irradiation from inhalation, and internal irradiation from resuspension; the calculation items for mid-stage radiation dose include: external irradiation from ground deposition and internal irradiation from resuspension; and the calculation items for late-stage radiation dose include: the decay of long-lived radionuclides and the cumulative environmental effects.

[0017] Furthermore, in some embodiments, in step d), the dose control criteria include:

[0018] In the early stages of the accident, under 99.5% of weather conditions, if the radiation dose received by personnel over 7 days is less than 50 mSv, emergency evacuation is not required.

[0019] During the middle of the accident, under 99.5% weather conditions, if the radiation dose received by people in the first month is less than 30 mSv, and the monthly radiation dose received by people in the following 11 months is less than 10 mSv, then temporary evacuation is not required.

[0020] In the later stages of an accident, under 99.5% weather conditions, if the radiation dose received by people remaining at the original accident site for 70 years after the accident is less than 1 Sv, permanent relocation is not necessary.

[0021] Furthermore, in some embodiments, in the two-parameter threshold criterion, 131 The method for calculating the equivalent dose is as follows: ;

[0022] Among them, S DE.131 The equivalent of the total amount of radionuclides released into the environment 131 I activity; F inh,i For radionuclide i inhalation irradiation dose conversion factor; F img,i For the dose conversion factor of radionuclide i immersion irradiation; F gnd,i V is the dose conversion factor for radionuclide i ground deposition irradiation; i F' represents the deposition rate of nuclide i; inh for 131 I. Inhaled radiation dose conversion factor; F' img for 131 I immersion dose conversion factor; F' gnd for 131 I is the ground deposition irradiation dose conversion factor; V' is... 131 I deposition rate, S i The activity of nuclide i released into the environment. Attached Figure Description

[0023] Figure 1 This is a schematic diagram of simplified off-site emergency response guidelines in one embodiment;

[0024] Figure 2 This is a schematic diagram of a simplified off-site emergency management method based on the intrinsic safety of advanced reactors in one embodiment;

[0025] Figure 3 This is a schematic diagram of radioactive migration in one embodiment.

[0026] 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

[0027] The present invention will now be described in further detail with reference to specific embodiments and accompanying drawings.

[0028] 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.

[0029] In this article, "multiple" means at least two.

[0030] Currently, the design concept of traditional off-site emergency response systems for nuclear power plants is overly conservative, often delineating large-scale plume emergency planning zones, such as the 7km-10km plume emergency planning zone required by GB / T 17689.1-1999. Emergency actions require various measures, including evacuation, sheltering, and iodine protection. However, with the continuous development of nuclear power technology, especially the application of new technologies such as third-generation nuclear power and small modular reactors, off-site emergency management schemes based on second-generation reactor technology have led to the long-term idleness and waste of a large amount of emergency resources, such as vehicles and shelters, and are incompatible with the technical characteristics of third-generation nuclear power and small modular reactors. The "Safety Regulations for Nuclear Power Plant Design" (HAF102-2016) states that "nuclear power plant accident sequences that could lead to high radiation doses or large releases of radioactivity must be practically eliminated; high-frequency nuclear power plant accident sequences must be guaranteed to have no or only minor potential radioactive consequences. The basic goal of safety design is to technically achieve that off-site protective actions to mitigate radioactive consequences are limited or even negligible." However, in reality, there is currently a lack of quantifiable and simplified off-site emergency management methods for nuclear power plants.

[0031] To address the aforementioned issues, embodiments of the present invention provide a simplified off-site emergency management method based on the intrinsic safety of advanced reactors.

[0032] This method employs radioactive release control criteria and off-site emergency simplification criteria tailored to the intrinsic safety characteristics of advanced reactors.

[0033] The method for determining the criteria for controlling radioactive release is as follows:

[0034] The core of off-site emergency simplification lies in improving the inherent safety of nuclear energy utilization and effectively eliminating large-scale radioactive releases (referred to as "effective elimination"). This is achieved by strengthening reactor design, operational monitoring, and safety protection systems, constructing a multi-layered, in-depth defense mechanism to curb the uncontrolled release of radioactive materials at the source, thereby laying a solid safety foundation for off-site emergency simplification. The radioactive release control criteria adopt a two-parameter threshold criterion, namely, a dose equivalent to a gaseous radioactive release exceeding 500 TBq. 131 The cumulative frequency of I and the accident sequence is less than 10. -7 / years. The two-parameter threshold criterion compared to the IAEA SSG-2 "actual elimination" frequency threshold (10 -6 The standard for gaseous radioactive release is one order of magnitude more stringent than the standard for effective dose of 50 mSv (emergency evacuation intervention and level), providing an analytical basis for simplified off-site emergency management methods.

[0035] In actual accident scenarios, a wide variety of radionuclides are released into the environment, and different radionuclides have different toxicities. To standardize the impact of radionuclides released into the environment under different accident conditions, from the perspective of radiological radiation dose effects, the radioactivity of all radionuclides is equivalently normalized to... 131 I. Quantitative analysis of the accident's impact is conducted using comparable physical quantities. Specifically:

[0036] ,

[0037] Among them, S DE.131 The equivalent of the total amount of radionuclides released into the environment 131 I activity; F inh,i For radionuclide i inhalation irradiation dose conversion factor; F img,i For the dose conversion factor of radionuclide i immersion irradiation; F gnd,i V is the dose conversion factor for radionuclide i ground deposition irradiation; i F' represents the deposition rate of nuclide i; inh for 131 I. Inhaled radiation dose conversion factor; F' img for 131 I immersion dose conversion factor; F' gnd for 131 I is the ground deposition irradiation dose conversion factor; V' is... 131 I deposition rate, S i The activity of nuclide i released into the environment.

[0038] S DE.131 Based on the existing model (which only considers inhalation irradiation), factors such as immersion irradiation and deposition irradiation are further introduced to better match the irradiation pathways involved in off-site emergency management, so as to reflect the differences in the comprehensive toxicity of different nuclides and further improve the accuracy of normalization treatment in evaluating the consequences of radioactive release.

[0039] Simplified guidelines for off-site emergency response, such as Figure 1 As shown, the method for determining it is as follows:

[0040] For accident categories that have been eliminated based on probabilistic safety analysis and practical engineering experience, their impact on the off-site radioactivity level is determined to be extremely low, supported by reliable theoretical evidence, and no additional off-site radioactivity consequence assessment is required. For accident categories that have not been eliminated, including potential design-baseline accidents and severe accident scenarios, the complexity and uncertainty of their potential risks should be fully considered, and in-depth specific analysis and evaluation should be conducted to ensure the effectiveness of risk control measures.

[0041] By establishing a comprehensive accident early warning, monitoring, intervention, and control system, unnecessary relocation of the public outside the site can be avoided. While ensuring radiation safety, necessary and beneficial protective actions, such as sheltering and taking iodine tablets, which have been proven effective in practice, should be retained. The top-level requirement for simplified off-site emergency response is: the public outside the plant site boundary should not need emergency evacuation during the early exposure phase, temporary relocation during the mid-exposure phase, or permanent relocation during the late exposure phase.

[0042] Wind direction, wind speed, and weather conditions directly affect the diffusion of radioactive materials in the atmosphere surrounding a nuclear power plant after an accident. Site meteorological conditions are a significant factor influencing personnel dose after an accident. Early off-site emergency response practices typically considered 95% of meteorological conditions for design basis accidents and 70% for severe accidents. This implementation plan, by adopting a high standard of 99.5% meteorological conditions across the entire accident spectrum, constructs a risk assessment system covering extreme weather scenarios. The envelope of the results is improved by 4-10 times, fundamentally eliminating blind spots in emergency decision-making caused by meteorological uncertainties.

[0043] Adapting to the aforementioned principles and aligning with the requirements of GB18871, and employing a conservative radiation dosimetry model, the following simplified and quantifiable phased off-site emergency response criteria are feasible:

[0044] Early exposure phase guidelines: For design basis accidents and severe accident conditions, under 99.5% weather conditions, ensure that the personnel dose after 7 days of continuous exposure is less than 50 mSv, and emergency evacuation is not required.

[0045] Mid-term exposure guidelines: Under severe accident conditions and 99.5% weather conditions, ensure that the personnel dose is less than 30 mSv in the first month after the early exposure phase, and less than 10 mSv per month for the following 11 months, without the need for temporary relocation.

[0046] Late-stage exposure guidelines: Under severe accident conditions and with 99.5% weather probability, ensure that the radiation dose to personnel remaining in their original place of residence for 70 years after the accident is less than 1 Sv, and that permanent relocation is not required.

[0047] Based on the above criteria, a simplified off-site emergency management method based on the intrinsic safety of advanced reactors, such as... Figure 2As shown, it includes the following steps:

[0048] Step a): Determine accident source terms. In determining design baseline accident source terms, from a deterministic analysis perspective, we fully draw upon various typical design baseline accident scenarios in safety analysis, and combine advanced system modeling and simulation technologies to reasonably define the characteristic parameters of accident source terms. For severe accident source terms, from a probabilistic perspective, we comprehensively consider reactor physical characteristics, system reliability, and other factors, and use advanced probabilistic safety evaluation methods to screen out representative meltdown sequences that may have a significant impact on the external environment, ensuring the reliability of source term determination.

[0049] Taking a design-baseline accident as an example, the migration process of radioactive materials is as follows: Figure 3 As shown, in the analysis and calculation process, each system or device with a clear boundary during radioactive migration is considered as a volume element. In a typical accident, a volume element can include the reactor core, containment vessel, primary loop system, secondary loop system, and other equipment. Radioactive material can flow interactively between volume elements. For any volume element and the nuclides involved, the following equilibrium equation holds:

[0050] ,

[0051] Among them, Q i q represents the activity of nuclides in voxel i; i V represents the rate of radioactive release within volume element i; i F represents the volume or mass of volume element i; i λ is the inflow of volume element i; di λ is the decay constant of the nuclide in volume element i; ri denoted as the removal coefficient of energetic nuclides in voxel i.

[0052] The above equations are used to describe all voxels involved in radioactive migration during the accident. Solving all the equations together yields the change in the activity concentration of any nuclide within each voxel over time. For the total amount released into the environment, integrating and summing the results over all voxels that released radioactivity into the environment gives the total amount released into the environment over any given time period.

[0053] ,

[0054] Among them, C i (t) represents the intrinsic activity concentration of volume element i released into the environment at time t; F i (t) represents the mass release rate of volume element i into the environment; n represents the total number of volume elements in the accident analysis.

[0055] Step b): Atmospheric dispersion analysis. Using professional atmospheric dispersion models such as the conservative Gaussian plume centerline model, combined with meteorological observation data from the nuclear power plant, a comprehensive consideration is given to the combined effects of atmospheric stability, wind speed, wind direction, precipitation, and other meteorological factors on the diffusion of radioactive materials. Through simulation and dynamic analysis of the diffusion path, concentration distribution, and deposition patterns of radioactive materials in the atmosphere, data is provided for subsequent assessments.

[0056] Step c): Radiation model determination. Adapting to the characteristics of different stages of accident development, radiation dose calculation models are established for the early, middle, and late stages, considering different exposure pathways:

[0057] In the early stages, a comprehensive consideration should be given to various irradiation pathways during or shortly after the plume passes, such as direct external irradiation, ground deposition external irradiation, body surface deposition external irradiation, inhalation internal irradiation, and resuspension inhalation internal irradiation.

[0058] In the medium-term stage, considering the migration and transformation characteristics of radioactive materials in the Earth's surface and environmental media after atmospheric release has basically ceased, the main irradiation pathways will be mainly considered, such as external irradiation through ground deposition and internal irradiation through resuspension inhalation.

[0059] In the late stage, the decay characteristics of long-lived radionuclides and the cumulative effects on the environment are taken into account, and the early and mid-stage irradiation pathways are considered comprehensively.

[0060] Step d): Off-site dose assessment. Based on meteorological observation data at the plant site, combined with the atmospheric dispersion analysis results and radiation model calculation parameters, advanced radiation dose calculation software and data analysis tools are used to calculate personnel doses at different stages. The calculation results are then rigorously compared and analyzed with the off-site emergency simplification criteria to determine whether the conditions for simplified off-site emergency protection actions are met.

[0061] In one specific embodiment, the simplified off-site emergency management method based on the intrinsic safety of advanced reactors is implemented as follows:

[0062] First, operating conditions were selected and source term analysis was performed. Based on NB / T 20444-2017RK, seven typical design basis accidents were selected, including the design basis loss-of-coolant accident (LOCA), main steam line rupture accident (MSLB), spring rod accident, steam generator heat transfer tube rupture accident (SGTR), main pump rotor / pump shaft breakage accident, fuel handling accident, and containment coolant-laden small pipe rupture accident. All selected accidents were representative and inclusive within their respective categories.

[0063] For the aforementioned design baseline accidents, attention should be paid to assessing the activity of radionuclides released into the atmosphere in segments within 7 days after the accident.

[0064] Next, atmospheric dispersion analysis was conducted. Using local meteorological observation data of the nuclear power plant, the atmospheric dispersion factor at the site boundary corresponding to 99.5% of meteorological conditions was evaluated. The maximum values ​​for the time periods of 0-2h, 2-8h, 8-24h, 24-96h, and 96-168h were selected from the obtained atmospheric dispersion factor data pool of 16 standard azimuths (each azimuth is 22.5° apart) to provide data for subsequent assessment.

[0065] Subsequently, a radiation irradiation simulation was conducted. After the radioactive nuclide was released into the environment, it caused immersion external radiation and inhalation internal radiation to personnel during the dispersal of the plume, and the deposition effect of radioactivity in the ambient atmosphere was conservatively ignored.

[0066] Finally, the universality of the simplified off-site emergency response system was verified based on meteorological data from typical coastal plant sites in the Yellow Sea, East China Sea, and South China Sea, ensuring technical feasibility. Evaluation showed that during the early irradiation phase, the continuous irradiation dose to personnel at the plant site boundaries caused by the seven design baseline accidents was significantly less than 50 mSv over seven days, eliminating the need for emergency evacuation.

[0067] In another embodiment, the implementation process of the simplified off-site emergency response method based on the intrinsic safety of advanced reactors under severe accident conditions is as follows:

[0068] First, operating conditions are selected and source term analysis is performed. Based on the behavior of the containment, severe accident operating conditions are usually divided into six release categories: containment integrity, containment bypass, containment isolation failure, early containment failure, mid-containment failure, and late containment failure.

[0069] Based on the PSA analysis results, factors such as containment failure and bypass that have a cumulative frequency of less than 10 were excluded. -7 Release types that meet the actual target of eliminating a large number of radioactive releases per reactor year are selected, and those with intact containment (core meltdown) are screened for accident source term evaluation, with a focus on determining the proportion of radionuclides released into the environment after the accident relative to the core accumulation.

[0070] Next, atmospheric dispersion analysis is conducted. Based on the hourly meteorological conditions of the nuclear power plant (including hourly wind speed, wind direction, precipitation, and atmospheric stability throughout the year), the classic Gaussian plume equation is used to simulate atmospheric dispersion during a reactor accident. Furthermore, the decay effects of radionuclides in the ambient atmosphere, such as dry deposition and wet deposition, are considered.

[0071] The next step is to evaluate the exposure methods at different time periods. This will cover the impact of the early, middle, and late exposure phases following the accident. Specifically: the early exposure phase will evaluate the impact in the first 7 days after the accident; the middle exposure phase will evaluate the impact in the first and second months following the early exposure phase. As time progresses, due to radioactive decay and natural processes (e.g., erosion, weathering), the monthly dose generally tends to decrease; therefore, the radiation impact in the second month can encompass the results of each month from the second to the twelfth month; the late exposure phase will consider a broad "late exposure period," conservatively superimposing the combined effects of the early, middle, and late phases to evaluate the impact on personnel over an indefinite period from the accident.

[0072] The early irradiation phase mainly considers irradiation during or shortly after the plume passes through. This phase considers five irradiation methods: direct external irradiation from radionuclides in the passing plume, external irradiation from radionuclides deposited on the ground, external irradiation from radionuclides deposited on clothing or skin, internal irradiation from inhalation of radionuclides in the plume, and internal irradiation from inhalation of resuspended radionuclides.

[0073] The late exposure phase is closely related to the presence of long-lived radionuclides, and the exposure pathways in this phase are the same as those in the intermediate phase. However, as mentioned above, in order to further assess the lifetime dose to personnel, the broad late exposure phase, which spans from the occurrence of the accident to an indefinite period of time, is also included in the exposure pathways of the early exposure phase.

[0074] Finally, a multi-site practicality evaluation was conducted.

[0075] Evaluation showed that during the early irradiation phase, based on 99.5% meteorological conditions, the personnel dose at the plant site boundary for 7 consecutive days met the limit requirement of 50 mSv, and no emergency evacuation was required. During the mid-term irradiation phase, the personnel dose at the plant site boundary met the limits of 30 mSv and 10 mSv for the first month and the following 11 months, respectively, and no temporary relocation was required. During the late irradiation phase, the lifetime dose for personnel at the plant site boundary met the limit requirement of 1 Sv, and no permanent relocation was required.

[0076] 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 simplified off-site emergency management method based on the intrinsic safety of advanced reactors, characterized in that, Includes the following steps: Step a): Identify the accident source term and calculate the total amount of radioactivity released into the environment under the accident source term; The method for determining the accident source terms is as follows: Accident source terms are determined using deterministic methods and probabilistic safety assessment methods. The accident source terms are judged using a two-parameter threshold criterion, where the two-parameter threshold criterion is a gaseous radioactive release exceeding 500 TBq. 131 The accident sequences and the cumulative frequency of accident sequences at equivalent doses of I are less than 10. -7 / year, for accident source items that satisfy the two-parameter threshold criterion, they are excluded; in the two-parameter threshold criterion... 131 The method for calculating the equivalent dose is as follows: ； Among them, S DE.131 The equivalent of the total amount of radionuclides released into the environment 131 I activity; F inh,i For radionuclide i inhalation irradiation dose conversion factor; F img,i For the dose conversion factor of radionuclide i immersion irradiation; F gnd,i V is the dose conversion factor for radionuclide i ground deposition irradiation; i F' represents the deposition rate of nuclide i; inh for 131 I. Inhaled radiation dose conversion factor; F' img for 131 I immersion dose conversion factor; F' gnd for 131 I is the ground deposition irradiation dose conversion factor; V' is... 131 I deposition rate, S i The activity of nuclide i released into the environment Step b): Perform atmospheric dispersion analysis based on meteorological conditions to calculate and determine the diffusion path, concentration distribution, and deposition state of radioactive materials in the atmosphere; Step c): Calculate the radiation dose in the early, middle, and late stages of the accident, respectively; Step d): Provide dose control guidelines and perform off-site emergency protection based on the calculation results of step c).

2. The simplified off-site emergency management method based on the intrinsic safety of advanced reactors according to claim 1, characterized in that, In step a), the method for calculating the total amount of radioactivity Q released into the environment is as follows: ； ； Among them, Q i q represents the activity of nuclides in voxel i; i V represents the rate of radioactive release within volume element i; i F represents the volume or mass of volume element i; i λ is the inflow of volume element i; di λ is the decay constant of the nuclide in volume element i; ri C is the removal coefficient of energetic nuclides in voxel i; i (t) represents the intrinsic activity concentration of volume element i released into the environment at time t; F i (t) represents the mass release rate of volume element i into the environment; n represents the total number of volume elements in the accident analysis.

3. The simplified off-site emergency management method based on the intrinsic safety of advanced reactors according to claim 1, characterized in that, In step b), the meteorological condition probability is a meteorological condition with a probability of 99.5%.

4. The simplified off-site emergency management method based on the intrinsic safety of advanced reactors according to claim 1 or 3, characterized in that, In step b), the meteorological condition probability calculation takes into account atmospheric stability, wind speed, wind direction, and precipitation probability.

5. The simplified off-site emergency management method based on the intrinsic safety of advanced reactors according to claim 1, characterized in that, In step c), the calculation items for early-stage radiation dose include: direct external irradiation within the influence range of the radiation plume, external irradiation from ground deposition, external irradiation from body surface deposition, internal irradiation from inhalation, and internal irradiation from resuspension; the calculation items for mid-stage radiation dose include: external irradiation from ground deposition and internal irradiation from resuspension; and the calculation items for late-stage radiation dose include: decay of long-lived radionuclides and cumulative environmental effects.

6. The simplified off-site emergency management method based on the intrinsic safety of advanced reactors according to claim 1, characterized in that, In step d), the dose control criteria include: In the early stages of the accident, under 99.5% of weather conditions, if the radiation dose received by personnel over 7 days is less than 50 mSv, emergency evacuation is not required. During the middle of the accident, under 99.5% weather conditions, if the radiation dose received by people in the first month is less than 30 mSv, and the monthly radiation dose received by people in the following 11 months is less than 10 mSv, then temporary evacuation is not required. In the later stages of an accident, under 99.5% weather conditions, if the radiation dose received by people remaining at the original accident site for 70 years after the accident is less than 1 Sv, permanent relocation is not necessary.

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