[0047]In order to make the objectives, technical solutions, and advantages of the embodiments of the present disclosure clearer, the specific implementation manners of the present disclosure will be described in detail below with reference to the accompanying drawings. It should be understood that the specific embodiments described herein are only used to illustrate and explain the present disclosure, and are not used to limit the present disclosure.
[0048]It should be noted that the terms "first" and "second" in the specification and claims of the present disclosure and the above-mentioned drawings are used to distinguish similar objects, and are not necessarily used to describe a specific sequence or sequence; Moreover, the embodiments in the present disclosure and the features in the embodiments can be combined with each other arbitrarily if there is no conflict.
[0049]Among them, the terms used in the embodiments of the present disclosure are only for the purpose of describing specific embodiments, and are not intended to limit the present disclosure. The singular forms of "a", "said" and "the" used in the embodiments of the present disclosure and the appended claims are also intended to include plural forms, unless the context clearly indicates other meanings.
[0050]Please refer tofigure 1 ,figure 1 This is a schematic flow chart of a calculation method for determining the environmental impact coefficient of the radioactive acceptance criteria for a severe accident of a nuclear power plant provided by an embodiment of the present disclosure. The method includes steps S101-S104.
[0051]In step S101, a number of radionuclides and their groupings for radioactive acceptance of a severe accident of a nuclear power plant are determined, and the representative nuclides of each radionuclide group are determined.
[0052]After a serious accident in a nuclear power plant, different radionuclides have different impacts on the environment. Therefore, this example first determines a number of radionuclides and their groupings. For example, 60 radionuclides are determined according to their physical and chemical properties. Divide them into 9 groups, and determine the representative radionuclide of each group. The groups of these 60 radionuclides and the representative nuclides of each group are shown in Table 1:
[0053]Table 1
[0054]
[0055]
[0056]Further, after step S101 and before step S102, the following steps are further included:
[0057]Calculate the core stock at the end of the fuel cycle life, where the core stock is the total amount of each radionuclide in the core;
[0058]Specifically, according to the physical design parameters and refueling parameters of the nuclear power plant core, the core stock at the end of the nuclear fuel cycle life is calculated.
[0059]In practical applications, the Origen program can be used to calculate the core storage at the end of the nuclear fuel cycle life based on the physical design parameters of the nuclear power plant core and the refueling plan, that is, the total amount of each radionuclide in the core, and then the The accident analysis program analyzes the reference release amount of each radionuclide to the environment after a serious accident of a nuclear power plant based on the total amount of each radionuclide in the core at the end of the fuel cycle life.
[0060]It is understandable that different nuclear power plants and different refueling plans have different core stocks, that is, the content of each radionuclide and isotope in the core is different.
[0061]In step S102, the reference release amount of each radionuclide to the environment after the occurrence of a serious accident in the nuclear power plant is analyzed, and the reference release amount of representative nuclides of each radionuclide group to the environment is obtained from it.
[0062]In this embodiment, in the severe accident analysis program, the reference release amount of each radionuclide to the environment after a severe accident of the nuclear power plant is analyzed based on the total amount of each radionuclide in the core at the end of the fuel cycle life.
[0063]Specifically, the severe accident analysis program is used to calculate the release process and results of radioactive source items to the environment after a typical severe accident; when calculating the release of radioactive source items in severe accidents, the analysis model used should truly reflect the design parameters of the power plant, and the analysis model The input data of radionuclides can be obtained by the above-mentioned Origen program, and when calculating the release of radioactive source items in severe accidents, in the severe accident analysis program, according to the chemical properties of the compounds formed by the radionuclides, they are divided into several compound groups. The program calculates the changes in the release share of each group of radionuclides to the environment over time.
[0064]Further, since the severe accident analysis program and the grouping of radionuclides used for radioactive acceptance of severe accidents of nuclear power plants may be different, this embodiment also needs to convert the release share of radionuclides to the environment. And according to the requirements of the radiological consequences analysis program, the release share change curve with time is appropriately simplified. Such asfigure 2 As shown, the curve represents the time-varying curve of the release share given by the severe accident analysis program. According to the trend of the curve, select a number of representative time points, as shown by the dots on the curve, that is, the release share changes with time. The appropriate simplification of the curve can be used in subsequent radiological consequences analysis procedures.
[0065]In step S103, based on the reference release amount of each radionuclide to the environment, the radioactive consequences caused by each radionuclide grouping after the occurrence of a serious accident in the nuclear power plant are analyzed.
[0066]Specifically, after the severe accident analysis program analyzes the reference release amount of each radionuclide, the radioactive consequences caused by each radionuclide grouping are analyzed in the severe accident consequence analysis program.
[0067]Further, the method further includes the following steps: obtaining environmental parameters of a nuclear power plant site, the environmental parameters of the nuclear power plant site at least including the following parameters: meteorological parameters of the site, population distribution parameters of the site, and dose estimation parameters of each exposure pathway.
[0068]The step S103 is specifically:
[0069]Input the environmental parameters of the nuclear power plant site and the reference release amount of all radionuclides to the environment in each radionuclide group after the occurrence of a serious nuclear power plant accident into the radiological consequence analysis program; and,
[0070]In the radiological consequence analysis program, based on the environmental parameters of the nuclear power plant site gas and the reference release amount of all nuclides in each radionuclide group to the environment, the consequences caused by the radioactive grouping of each radionuclide are obtained.
[0071]In this embodiment, the radioactive consequence analysis program is used to input the radionuclide release shares calculated by the severe accident analysis program over time into the radioactive consequence analysis program, such as MAACS, to calculate the radioactive consequences caused by the grouping of the i-th group of radionuclides. Di, In practical applications, by determining the distance x, the radiological consequences D at a certain distance x outside the nuclear power planti,x; In the radiological consequence analysis program, the total amount of radioactive material released to the environment (release amount) is input according to the calculation result of the serious accident analysis program. Among them, the above-mentioned distance x can be determined according to the distance requirements of the nuclear power plant to perform emergency protective actions, such as the boundary of the nuclear power plant site, the boundary of the non-residential area, and the boundary of the emergency plan area.
[0072]Further, in the calculation of the radiological consequences analysis program, each time it is assumed that one group of radionuclides is released, and the release amount of the remaining radionuclide groups is 0, so that the radioactivity caused by the group of radionuclides at a distance of x from the nuclear power plant is obtained. Consequence Di,x.
[0073]In step S104, the environmental impact coefficient of each representative nuclide is calculated based on the reference release amount of the representative nuclide of each radionuclide group to the environment and the radioactive consequences caused by each radionuclide group.
[0074]Further, the calculation method of step S104 is obtained according to the following calculation formula:
[0075]
[0076]In the formula, CiThe environmental impact coefficient of the representative nuclide grouping the i-th radionuclide, DiIs the radiological consequences caused by all nuclides in the i-th radionuclide group, R′iIt is the reference release amount of the representative nuclides of the i-th radionuclide group to the environment.
[0077]In practical applications, the radioactive consequences of radionuclides are usually related to distance, and the environmental impact coefficient of radionuclides can be expressed as:
[0078]
[0079]Understandable, Ci,x The equivalent environmental impact coefficient of the representative nuclides in the i-th radionuclide group at a distance x from the nuclear power plant, Di,x Is the radiological consequences caused by all nuclides in the i-th radionuclide group at a distance x from the nuclear power plant, R′iIt is the reference release amount of the representative nuclides of the i-th radionuclide group to the environment. As described above, the subscript x represents the distance. In combination with the following embodiments, when determining the dose reference value, it is also related to the distance x, which will not be repeated here.
[0080]In this embodiment, based on the reference release amount of the representative nuclides of the i-th group of radionuclides and the radiological consequences caused by all the nuclides in the i-th group of radionuclides at a distance x from the nuclear power plant, the environmental impact coefficient C is derivedi,x;among them Where R’iIs the reference release amount of the i-th representative nuclide, in TBq. E.g,
[0081]It should be noted that the environmental impact coefficient in this embodiment is the equivalent environmental impact coefficient of the representative nuclides of the radionuclide grouping, and the equivalent environmental impact is calculated by grouping the radionuclides and selecting the representative nuclides Coefficient, making the calculated environmental impact coefficient more representative.
[0082]Please refer toimage 3 ,image 3 This is a schematic flow chart of a method for determining radioactivity acceptance criteria for severe accidents in nuclear power plants according to an embodiment of the present disclosure. The method includes steps S301-S306.
[0083]In step S301, a number of radionuclides and their groupings for radioactive acceptance of severe accidents in nuclear power plants are determined, and the representative nuclides of each radionuclide group are determined.
[0084]In step S302, the reference release amount of each radionuclide to the environment after the occurrence of a serious accident in the nuclear power plant is analyzed, and the reference release amount of representative nuclides of each radionuclide group to the environment is obtained from it.
[0085]In step S303, the radiological consequences caused by each radionuclide grouping after a serious accident in the nuclear power plant are analyzed.
[0086]In step S304, the environmental impact coefficient of each representative nuclide is calculated based on the reference release amount of the representative nuclide of each radionuclide group to the environment and the radiological consequences caused by each radionuclide group.
[0087]It should be noted that steps S301-S304 in this embodiment correspond to steps S101-S104 in the previous embodiment, and will not be repeated here.
[0088]In step S305, a dose reference value for radioactivity acceptance of a severe accident of a nuclear power plant is determined.
[0089]Among them, step S305 is specifically:
[0090]Obtain the general optimized intervention level corresponding to the implementation of different emergency protective actions from the basic standard for ionizing radiation protection and radiation source safety GB18871; and,
[0091]The general optimized intervention level corresponding to the execution of different emergency protective actions is taken as the dose reference value for the radioactive acceptance of the severe accident of the nuclear power plant.
[0092]Specifically, in the basic standard for ionizing radiation protection and radiation source safety GB18871, look for the general optimized intervention level corresponding to different emergency protection actions in the "General optimized intervention level and action level under emergency exposure". It is understandable that the general intervention level is the dose level prescribed in GB18871 for determining the need to take emergency protective measures against the public under abnormal conditions.
[0093]In some embodiments, while determining the dose reference value for radioactive acceptance of severe accidents in nuclear power plants, it is also necessary to determine the distance requirements for radiological acceptance. Specifically, according to the design goals for severe accidents of nuclear power plants, determine the distance to perform emergency protective actions. Requirements, and the general optimized intervention level and distance requirements corresponding to the different emergency protective actions are used as the dose reference value and distance requirements for radioactive acceptance of severe accidents in nuclear power plants.
[0094]It is understandable that the distance requirements for different emergency protective actions are different, and the acceptance criteria formulas are also different. For example, if a nuclear power plant is designed to reach a range of 600m without evacuation, the distance requirement corresponding to the general optimized intervention level for evacuation is 600m.
[0095]In step S306, a radioactive acceptance criterion for a severe accident of a nuclear power plant is determined based on the environmental impact coefficient of each representative nuclide and the dose reference value.
[0096]Further, the determination of the radioactivity acceptance criteria for severe accidents of nuclear power plants based on the environmental impact coefficient of the representative nuclides of each radionuclide group and the dose reference value is obtained according to the following formula:
[0097]
[0098]In the formula, n represents the number of radionuclide groups, RiIt is the actual release amount of the representative nuclide of the i-th radionuclide group to the environment, the unit is TBq; CiIt is the environmental impact coefficient of the representative nuclides of the i-th radionuclide group, in Sv/TBq; criterion is the dose reference value, in Sv.
[0099]In combination with the above, the radioactive acceptance criteria for severe accidents of nuclear power plants under the condition of certain distances are obtained according to the following formula:
[0100]
[0101]Where Ci,x It is the equivalent environmental impact coefficient of the representative nuclides of the i-th radionuclide group at a distance of x from the nuclear power plant, in Sv/TBq; x is the distance requirement corresponding to the implementation of different emergency protective actions.
[0102]It should be noted that Ri is the actual release amount that a nuclear power plant may release to the environment after a serious accident. It can be obtained through analysis of the serious accident analysis program, or through other means such as nuclear power plant design parameters.
[0103]In order to facilitate the understanding of this embodiment, the following takes a nuclear power plant as an example, combined withFigure 4 As shown, the specific application of the above method (considering the distance x):
[0104]a. Obtain the design parameters of nuclear power plants;
[0105]b. Use the Origen program to calculate the core physical design parameters and refueling plan of a nuclear power plant, and calculate the core stock at the end of the fuel cycle life. The obtained core stock is the total amount of each nuclide in the core. Activity expression;
[0106]c. Input the core accumulation data and other data into the serious accident analysis program to analyze the release result of the radioactive source item;
[0107]d. Use the integrated accident analysis program for severe accidents (ie, severe accident analysis program) to calculate the release process and release results of radioactive source items to the environment after a typical severe accident. The integrated analysis program for severe accidents used here can be the MAAP program;
[0108]e. The share of the total amount released into the environment by different radioactive material groups (in the form of compounds) in the core stock is given. The details are shown in Table 2 below:
[0109]Table 2
[0110] serial number MAAP grouping 2 days release share Free share within 7 days 1 Kr,Xe 4.47E-03 1.83E-02 2 CsI 1.18E-05 1.18E-05 3 TeO2 1.10E-05 1.10E-05 4 SrO 3.90E-07 3.90E-07 5 MoO2 2.22E-06 2.22E-06 6 CsOH 1.05E-05 1.06E-05 7 BaO 5.35E-07 5.35E-07 8 La2O3 8.87E-09 8.87E-09 9 CeO2 4.72E-08 4.72E-08 10 Sb 6.77E-06 6.84E-06 11 Te2 0.00E+00 0.00E+00 12 UO2 0.00E+00 0.00E+00
[0111]Among them, the conversion of radioactive release results: Because the severe accident analysis procedures and the radionuclides used for radioactive acceptance of severe accidents in nuclear power plants are different, it is necessary to convert the release share of radionuclides to the environment. And according to the requirements of the radiological consequences analysis program, the release share change curve with time is appropriately simplified.
[0112]In the analysis of radiological consequences, the MACCS program is used. The program only focuses on the release of the following radionuclides. Therefore, it is necessary to convert the release share of the MAAP program to the release share of the MACCS program, and convert the share to Activity, that is, the reference release amount of each representative nuclide is obtained, as shown in Table 3 below:
[0113]table 3
[0114]
[0115]In the above table, there are a total of 9 representative nuclides. These 9 representative nuclides represent 9 groups of different types of radionuclides (combined with Table 1).
[0116]f. Obtain the meteorological condition data of the plant site (ie, the environmental parameters of the nuclear power plant site), including the meteorological parameters of the plant site, the population distribution parameters of the plant site, and the dose estimation parameters of various exposure pathways.
[0117]g. Using the radiological consequence analysis program, respectively input the radionuclide release share of each group calculated by the severe accident analysis program over time and the meteorological conditions of the site gas into the radiological consequence analysis program.
[0118]h. In the radiological consequences analysis program, calculate the radiological consequences Di,x, in units of Sv, caused by all nuclides in the i-th radionuclide group at a certain distance x outside the nuclear power plant.
[0119]Using the MACCS program, each time it is assumed that a radionuclide group in the above table is released, and the corresponding radiological consequences Di,x are obtained. Among them, the calculation of Di, x is related to the two factors of time and distance. Time is the time after the accident, and distance is the distance from the center of the reactor. When calculating Di,x, the time should be determined according to the general optimized intervention level of the corresponding emergency protective action, and the distance should be determined according to the distance requirement for the emergency protective action.
[0120]i. Based on the reference release amount of the i-th representative nuclide and the radiological consequences caused by all nuclides in the radionuclide group where it is located, the environmental impact coefficient Ci,x is derived; Where R′iIs the reference release amount of the i-th representative nuclide, in TBq.
[0121]j. Determine the design goals of serious accidents, according to the requirement that “the basic goal of safety design is to technically realize that off-site protective actions to mitigate the consequences of radioactivity are limited or even can be eliminated” in the "Safety Regulations for Nuclear Power Plant Design" , The design goals for serious accidents are determined as: 1) No temporary evacuation outside the non-residential area; 2) No concealment outside the emergency plan area.
[0122]k. Determine the distance requirements and dose benchmarks for different emergency protective actions. According to the design of the nuclear power plant, the boundary of the non-residential area is 600m, and the boundary of the emergency plan area is 3km, that is, serious accident requirements 1) No temporary evacuation is required outside the non-residential area (600m from the reactor); 2) No concealment is required beyond 3km from the reactor. Determine the value of the criterion in the expression according to the expected dose level of intervention under any circumstances specified in GB18871 and the intervention level and action level of emergency exposure situations. As mentioned above, in view of the design goal of “no need to evacuate outside the non-residential area (600m from the reactor)”, according to GB18871 Appendix E2.1.2 of the general optimization intervention level for temporary evacuation, the cumulative environmental dose within 7 days after the accident is required to be less than 50mSv. Aiming at the goal of “no concealment is required beyond 3km from the reactor”, according to the general optimization intervention level of concealment in GB18871 Appendix E2.1.1, the environmental cumulative dose level is required to be less than 10mSv in the two days after the accident.
[0123]1. Determine the final acceptance criterion expression. According to the above steps, the expression forms of the above two acceptance criteria are finally determined as follows:
[0124]Acceptance criteria 1: No temporary evacuation is required outside the non-residential area (600m from the reactor)
[0125]
[0126]Acceptance criterion 2: No concealment is required beyond 3km from the reactor
[0127]
[0128]Environmental impact coefficient C in the acceptance criteriai,600m , Ci,3000m The derivation results are shown in the table below
[0129] Representative nuclide C i,600m (Acceptance Criteria 1)
[0130]Please refer toFigure 5 ,Figure 5 This is a schematic flow chart of a method for radioactive acceptance of a severe accident in a nuclear power plant provided by an embodiment of the present disclosure. The method includes step S501 and step S502.
[0131]In step S501, it is judged whether the design of the relevant nuclear power plant meets the radioactivity acceptance criteria for severe accidents of nuclear power plants as determined by the method for determining radioactivity acceptance criteria for severe accidents of nuclear power plants;
[0132]In step S502, if the nuclear power plant meets the radioactivity acceptance criteria for a serious accident of the nuclear power plant, it is determined that the nuclear power plant meets the design target for a serious accident.
[0133]Correspondingly, if the design of a nuclear power plant does not meet the radiological acceptance criteria for severe accidents, it means that the nuclear power plant does not meet the design goals for severe accidents, and relevant improvements should be made to the design of the nuclear power plant to ensure that the nuclear power plant meets the " In the Nuclear Power Plant Design Safety Regulations, the safety goal of "technically realizing off-site protective actions to mitigate the consequences of radioactivity is limited or even can be eliminated".
[0134]In summary, the method for calculating the environmental impact coefficient of the radioactive acceptance criteria for severe accidents of nuclear power plants, the method for determining radioactive acceptance criteria for severe accidents of nuclear power plants, and the methods for radioactive acceptance of severe accidents of nuclear power plants provided by the embodiments of the present disclosure are analyzed by analyzing the radioactivity of nuclear power plants. The reference release amount of nuclides and the radioactive consequences caused by the radionuclide group, determine the equivalent environmental impact coefficients of representative nuclides in nuclear power plants, and then determine the acceptance criteria for severe accidents in nuclear power plant design, so as to at least solve the current lack of nuclear power plant design The acceptance criteria for the radiological consequences of severe accidents to the environment improve the scientific nature of the design of nuclear power plants, which can be simple, effective and widely applicable to the acceptance criteria for serious accidents of various nuclear power plants.
[0135]Finally, it should be noted that the above embodiments are only used to illustrate the technical solutions of the present disclosure, not to limit them; although the present disclosure has been described in detail with reference to the foregoing embodiments, those of ordinary skill in the art should understand that: The technical solutions recorded in the foregoing embodiments can still be modified, or some or all of the technical features can be equivalently replaced; and these modifications or replacements do not cause the essence of the corresponding technical solutions to deviate from the technical solutions of the embodiments of the present disclosure. range.
