Method for quantitatively evaluating metal foreign body activation time in core of nuclear power plant

Abstract

The application provides a method for quantitatively evaluating activation time of a metal foreign object in a core of a nuclear power plant, comprising: obtaining characteristic data of the foreign object, the characteristic data comprising a measured value of a gamma spectrum, a neutron flux, and measured gamma dose rates at different measurement distances; obtaining a measured radionuclide activity according to the measured value of the gamma spectrum; calculating theoretical radionuclide activities of the foreign object under different preset irradiation durations according to the neutron flux and the different preset irradiation durations; calculating theoretical gamma dose rates of the foreign object under the different preset irradiation durations according to the theoretical radionuclide activities corresponding to the different preset irradiation durations; comparing the measured gamma dose rates at the different measurement distances with the theoretical gamma dose rates corresponding to the different preset irradiation durations, and comparing the measured radionuclide activity with the theoretical radionuclide activities corresponding to the different preset irradiation durations, and determining an actual irradiation duration of the foreign object in the core according to a comparison result. The irradiation history of the metal foreign object in the core is quantitatively evaluated.

Description

Technical Field

[0001] This invention relates to the field of nuclear safety monitoring and radioactive source term analysis technology, and more specifically, to a method for quantitatively assessing the activation time of metallic foreign matter in nuclear power plant reactor cores. Background Technology

[0002] During refueling and overhauls at nuclear power plants, video inspections of the reactor core structure and fuel assemblies are standard procedures. If metallic foreign objects are found in the lower core grating or fuel assembly gaps, their source and potential risks must be immediately assessed. Prolonged retention of foreign objects in the core can lead to fuel cladding wear, flow channel blockage, or even mechanical jamming, seriously threatening reactor safety. Currently, routine procedures typically include visual and geometric measurements to confirm the morphology and location of the foreign object, material spectral analysis to determine its chemical composition (e.g., stainless steel, nickel-based alloys), and surface contamination and dose rate detection using portable instruments to measure dose rate and surface contamination levels. Existing methods rely heavily on experience and subjectivity. Judgments regarding the introduction time depend on personnel estimations based on experience, lacking scientific calculations based on physical models, resulting in low repeatability and accuracy. Furthermore, the inability to precisely pinpoint the introduction time window makes it difficult to trace specific operational steps (e.g., tool operations during a particular overhaul), leading to a lack of targeted improvements to foreign object prevention measures. Summary of the Invention

[0003] The technical problem to be solved by this invention is to provide a quantitative assessment method for the activation time of metal foreign objects in nuclear power plant reactor cores, addressing the shortcomings of existing technologies that rely on experience and are highly subjective.

[0004] The technical solution adopted by this invention to solve its technical problem is: constructing a method for quantitatively assessing the activation time of metallic foreign matter in a nuclear power plant core, comprising the following steps: Step S1: Obtain the characteristic data of the foreign object, including the measured value of the gamma spectrum, the neutron flux, and the measured gamma dose rate at different measurement distances; Step S2: Obtain the measured activity of the radionuclide based on the measured value of the γ energy spectrum; Step S3: Based on the neutron flux and at least two preset irradiation durations, calculate the theoretical radionuclide activity of the foreign object under each preset irradiation duration; Step S4: Calculate the theoretical γ dose rate of the foreign object under each preset irradiation duration based on the theoretical radionuclide activity corresponding to each preset irradiation duration; Step S5: Compare the measured γ dose rate at different measurement distances with the theoretical γ dose rate corresponding to each preset irradiation duration, compare the measured radionuclide activity with the theoretical radionuclide activity corresponding to each preset irradiation duration, and determine the actual irradiation duration of the foreign object in the reactor core based on the comparison results.

[0005] Furthermore, the characteristic data also includes the shutdown decay time interval. Step S2 includes: The measured values ​​of the γ-ray spectrum were analyzed to identify characteristic peaks, and multiple target nuclides were selected. The measured radioactivity of each target nuclide at the time of reactor shutdown was calculated. .

[0006] Furthermore, the feature data also includes the foreign object material type and the total mass of the foreign object; step S3 includes: Step S31: Obtain the decay constant and the corresponding microscopic reaction cross section of each target nuclide; Step S32: Based on the material type of the foreign object, obtain the number of target nuclei atoms for each target nuclide; Step S33: Based on the number of target nuclei atoms of each target nuclide, the neutron flux, the microscopic reaction cross section, the decay constant, and the preset irradiation duration... The neutron activation equation is used to calculate the theoretical radioactivity of each target nuclide at the time of reactor shutdown. Step S34: Based on the decay time interval, the theoretical radioactivity of each target nuclide at the time of shutdown is corrected for decay, and the theoretical radioactivity after decay correction is obtained as the theoretical radioactivity activity of the nuclide.

[0007] Further, step S32 includes: The elemental mass percentage of the target nucleus corresponding to each target nuclide is determined based on the material type of the foreign object. Calculate the number of target nuclei for each target nuclide using the following formula. : in, Let represent the elemental mass percentage of the target nucleus corresponding to the i-th target nuclide. The total mass of the foreign object. Let Avogadro's constant be 1. Let be the atomic weight of the target nucleus corresponding to the i-th target nuclide.

[0008] Further, step S33 includes: Using the neutron activation equation, calculate the theoretical radioactivity of each target nuclide at the end of the preset irradiation time t using the following formula: in, Let i be the number of target nuclei of the i-th target nuclide. The microscopic reaction cross section of the i-th target nuclide Let i be the neutron flux corresponding to the i-th target nuclide. Let be the decay constant of the i-th target nuclide. This is the preset irradiation duration.

[0009] Further, step S34 includes: Obtain the time interval between heap decays ; The theoretical radioactivity of each target nuclide at the time of reactor shutdown is corrected for decay using the following formula: in, Preset irradiation duration Theoretical radioactivity at the end of the event. For the first The decay constant of the target nuclide, This is the time interval between reactor shutdown decays; The theoretical radioactivity after decay correction This refers to the theoretical activity of the radionuclide.

[0010] Furthermore, the feature data also includes the maximum linear dimension; step S4 includes: Step S41: Determine the relationship between each measured distance r and the maximum linear dimension. Relationship; Step S42: When At that time, a point source model was used to obtain the specific kerma rate constant of each target nuclide. The theoretical γ dose rate is calculated using the following formula: in, This represents the theoretical radioactivity after decay correction for the i-th target nuclide. For the first The specific kerma rate constant of the target nuclide; Step S43: When r < 5 × L_max, call the Monte Carlo simulation program to establish a three-dimensional geometric model of the foreign object, input the elemental composition and nuclide activity distribution, and simulate and calculate the theoretical γ dose rate H_theo(r) at the detection point.

[0011] Further, step S5 includes: Step S51: Calculate the dose rate deviation factor using the following formula. and activity deviation factor : in, H_meas is the measured gamma dose rate, H_theo is the theoretical gamma dose rate, A_meas is the measured radionuclide activity, and A_theo_final is the theoretical radionuclide activity when the foreign object was retrieved. Step S52: When both the dose rate deviation factor R_H and the activity deviation factor R_A approach 1, determine that the actual irradiation time of the foreign object is consistent with the corresponding preset irradiation time; otherwise, iteratively adjust the irradiation time. The theoretical radiation value is recalculated until convergence, thus obtaining the preliminary irradiation duration.

[0012] Furthermore, it also includes: Based on the initial irradiation duration, at least one short-half-life nuclide and at least one long-half-life nuclide are selected. When the short-half-life nuclide has reached saturation while the long-half-life nuclide has not, the irradiation duration is deduced from the growth curve of the long-half-life nuclide to obtain the deduced irradiation duration. The initial irradiation duration is then corrected using the deduced irradiation duration to obtain the final irradiation duration.

[0013] Furthermore, the step of inferring the irradiation duration from the growth curve of long-half-life nuclides to obtain the inferred irradiation duration includes: Measured radioactive nuclide activity of long half-life nuclides and its theoretical saturation activity N represents the number of target nuclei atoms. For the microscopic reaction cross section, Neutron flux The decay constant; Calculate the back-derived irradiation duration using the following formula : in, The decay constant of a long-half-life nuclide; The step of correcting the preliminary irradiation duration using the back-calculated irradiation duration to obtain the final irradiation duration includes: Determine whether the inverse irradiation duration is within the range of the initial irradiation duration; If so, the reversed irradiation duration will be output as the final irradiation duration; If not, select another combination of a short-half-life nuclide and a long-half-life nuclide, and repeat the above reverse calculation steps until the obtained reverse irradiation duration falls within the range of the preliminary irradiation duration. Then, take the reverse irradiation duration as the final irradiation duration.

[0014] The beneficial effects of implementing the present invention include: Step S1: acquiring characteristic data of the foreign object, the characteristic data including measured values ​​of the gamma spectrum, neutron flux, and measured gamma dose rates at different measurement distances; Step S2: obtaining the measured activity of radionuclides based on the measured values ​​of the gamma spectrum; Step S3: calculating the theoretical activity of radionuclides of the foreign object at each preset irradiation time based on the neutron flux and at least two preset irradiation times; Step S4: calculating the theoretical gamma dose rate of the foreign object at each preset irradiation time based on the theoretical activity of radionuclides corresponding to each preset irradiation time; Step S5: comparing the measured gamma dose rates at different measurement distances with the theoretical gamma dose rates corresponding to each preset irradiation time, comparing the measured radionuclide activities with the theoretical radionuclide activities corresponding to each preset irradiation time, and determining the actual irradiation time of the foreign object in the reactor core based on the comparison results. By comparing neutron activation theory with measured data on gamma dose rate and radionuclide activity in multiple dimensions, a quantitative assessment of the irradiation history of metallic foreign objects in the reactor core was achieved. Attached Figure Description

[0015] The present invention will be further described below with reference to the accompanying drawings and embodiments. In the accompanying drawings: Figure 1 This is a logic flowchart of a method for quantitatively assessing the activation time of metallic foreign matter in nuclear power plant reactor cores; Figure 2 This is a graph showing the activity of radionuclides in a foreign substance. Detailed Implementation

[0016] To provide a clearer understanding of the technical features, objectives, and effects of the present invention, specific embodiments of the invention are now described in detail with reference to the accompanying drawings. In the following description, specific details such as particular structures and techniques are set forth for illustrative purposes and not for limitation, so as to provide a thorough understanding of the embodiments of the invention. However, those skilled in the art will understand that the invention can also be implemented in other embodiments without these specific details. In other instances, detailed descriptions of well-known devices, circuits, and methods are omitted so as not to obscure the description of the invention with unnecessary detail.

[0017] like Figure 1 As shown, Figure 1 This is a logic flowchart of a method for quantitatively assessing the activation time of foreign metal objects in nuclear power plant reactor cores.

[0018] The technical solution adopted by this invention to solve its technical problem is: constructing a method for quantitatively assessing the activation time of metallic foreign matter in a nuclear power plant core, comprising the following steps: Step S1: Obtain the characteristic data of the foreign object, including the measured value of the gamma spectrum, the neutron flux, and the measured gamma dose rate at different measurement distances; In this step, it should be noted that the characteristic data may also include, depending on the actual situation: the three-dimensional geometric dimensions, mass, elemental composition, and decay time interval of the foreign object during reactor shutdown. Neutron flux was obtained from the final safety analysis report (FSAR) of the power plant or measured axial / radial neutron flux distribution data in the reactor core; the measured gamma dose rate was measured using a calibrated portable gamma dose rate meter, with measurement distances including standard points such as 5cm and 10cm, and the maximum linear size of the foreign object was recorded. This is used for subsequent point source model determination.

[0019] Step S2: Obtain the measured activity of the radionuclide based on the measured values ​​of the γ-ray spectrum; In this step, a high-purity germanium (HPGe) detector is used to measure the gamma spectrum of the foreign matter. Characteristic peaks are identified using spectral analysis software, and key activating nuclides such as Cr-51, Co-58, Co-60, and Fe-59 are screened out. The activity is then calculated using the formula... Calculate the measured radioactivity of the nuclide, where N is the net count of the characteristic peak. To measure live time, To improve the overall peak detection efficiency, The branching ratio of gamma-ray emission. This is a correction factor; the calculation results need to be corrected to the time of reactor shutdown to eliminate the effects of radioactive decay during the period from shutdown to salvage and measurement.

[0020] Step S3: Calculate the theoretical radionuclide activity of the foreign object under each preset irradiation duration based on the neutron flux and at least two preset irradiation durations; In this step, it should be noted that the preset irradiation duration is set to scenarios such as short-term introduction (t≈0), single-cycle legacy (12-18 months), and multi-cycle legacy (n×fuel cycle duration); the mass ratio of target nuclei is determined based on the foreign material, and the theoretical activity is calculated by combining the neutron activation kinetic equation. The shutdown decay time is uniformly corrected to ensure consistency with the measured radionuclide activity time benchmark.

[0021] Step S4: Calculate the theoretical γ dose rate of the foreign object under each preset irradiation duration based on the theoretical radionuclide activity corresponding to each preset irradiation duration; In this step, it should be noted that the relationship between the measurement distance and the maximum linear size of the foreign object should be determined first, and a point source model or Monte Carlo simulation program should be selected for calculation. The point source model is suitable for scenarios that meet the point source approximation conditions, while the Monte Carlo simulation (MCNP, CIRPDose) is suitable for complex geometric scenarios without point source conditions. Finally, the theoretical γ dose rate at the preset distance is output.

[0022] Step S5: Compare the measured γ dose rate at different measurement distances with the theoretical γ dose rate corresponding to each preset irradiation duration, compare the measured radionuclide activity with the theoretical radionuclide activity corresponding to each preset irradiation duration, and determine the actual irradiation duration of the foreign object in the reactor core based on the comparison results.

[0023] In one embodiment, when the gamma dose rate measurement distance r of the foreign object is greater than or equal to 5 times the maximum geometric size L of the foreign object, the foreign object is approximated as a point source, and the theoretical gamma dose rate is calculated using a point source model; when the gamma dose rate measurement distance r of the foreign object is less than 5 times the maximum geometric size L of the foreign object, a three-dimensional model is established using a Monte Carlo simulation program (such as MCNP, CIRPDose), the elemental composition and nuclide activity distribution are input, and the dose rate at the detection point is simulated and calculated.

[0024] This invention achieves a quantitative assessment of the irradiation history of metallic foreign objects in the reactor core by comparing neutron activation theory with measured data of gamma dose rate and radionuclide activity in multiple dimensions. It can accurately distinguish whether the foreign object originates from long-term neutron activation or recent surface contamination, and precisely determine the introduction time and residence time of the foreign object within the core. This provides a scientific basis for foreign object source investigation, root cause analysis, and improvement of foreign object prevention measures. Furthermore, this invention combines dose rate calculation methods based on point source models or Monte Carlo simulations, making it applicable to engineering assessment scenarios for foreign objects of different sizes, with objective and accurate results.

[0025] Furthermore, the characteristic data also includes the time interval between reactor shutdown and decay. Step S2 includes: identifying characteristic peaks in the measured γ-ray spectrum, selecting multiple target nuclides, and calculating the measured radioactivity of each target nuclide at the time of reactor shutdown. .

[0026] like Figure 2As shown in one specific embodiment, during a major overhaul and refueling process at a pressurized water reactor nuclear power plant, a suspected foreign object was discovered near the lower end grid of a fuel assembly. This fuel assembly was being used for the first time in the C05 (5th fuel cycle) cycle, with its core location at K06, and was being used for the second time in the C06 (6th fuel cycle) cycle, with its core location at H01. After the foreign object was retrieved, measurements showed that the object was approximately 9.5mm × 4mm in size, approximately 0.4mm thick, and weighed approximately 0.0275g. The radiation level was measured using a calibrated portable gamma dose rate meter. The contact dose rate was approximately 9.34 mSv / h, the dose rate at a distance of 5 cm was 1.7 mSv / h, and the dose rate at a distance of 10 cm was 0.744 mSv / h. Laser-induced breakdown spectroscopy (LIBS) analysis revealed spectral lines for elements such as iron (Fe), chromium (Cr), manganese (Mn), nickel (Ni), and molybdenum (Mo). Considering both the spectral line intensity and the ionization energy of each element (as shown in Table 1), the material of the foreign object was determined to be 316 series stainless steel based on the number of elemental peaks and the signal intensity after background removal. Three gamma spectral analyses were performed on the foreign object. Based on the characteristic peak information of the foreign object's energy spectrum, the main radionuclides were identified as Cr-51, Co-58, and Co-60. Spectral analysis revealed that the activity of Cr-51 at the time of shutdown was 3.09E+8 Bq, the activity of Co-58 was 6.76E+7 Bq, and the activity of Co-60 was 2.08E+7 Bq. The shutdown activity calculations have taken into account the radioactive decay from shutdown to 7 days before retrieval.

[0027] Table 1 Typical spectral lines and intensities of foreign materials Based on elemental analysis results and gamma-ray energy spectrum characteristics, target nuclides with high reaction cross-sections (>0.1b), long half-lives (>20d), and high emitted gamma-ray energies (>300keV) were screened. These include, but are not limited to: 58 Fe(n, γ) 59 Fe(T 1 / 2 =44.5d), 50 Cr(n, γ) 51 Cr(T 1 / 2 =27.7d), 58 Ni(n, p) 58 Co(T 1 / 2 =70.86d), 59 Co(n, γ) 60 Co(T 1 / 2 =5.27y). Distinguish between thermal neutron reactions and fast neutron reactions, and determine the neutron flux corresponding to each activation reaction ( ,, ).

[0028] Furthermore, the feature data also includes the foreign object material type and the total mass of the foreign object; step S3 includes: Step S31: Obtain the decay constant and the corresponding microscopic reaction cross section of each target nuclide; Step S32: Based on the material type of the foreign object, obtain the number of target nuclei atoms for each target nuclide; Further, step S32 includes: The elemental mass percentage of each target nuclide corresponding to the target nucleus is determined based on the material type of the foreign object. Calculate the number of target nuclei for each target nuclide using the following formula. : in, Let represent the elemental mass percentage of the target nucleus corresponding to the i-th target nuclide. The total mass of the foreign object. Let Avogadro's constant be 1. Let be the atomic weight of the target nucleus corresponding to the i-th target nuclide.

[0029] Specifically, based on the mass percentages of each element in 316 series stainless steel (reference range), the calculation conditions are: Cr 17%, Ni 12%, Fe 66%, and Co59 0.1%. The main calculation focuses on the dose rates contributed by the four radioactive nuclides Cr-51, Co-58, Co-60, and Fe-59 generated after activation by neutron irradiation in the reactor core.

[0030] Step S33: Based on the number of target nuclei, neutron flux, microscopic reaction cross section, decay constant, and preset irradiation duration for each target nuclide. The neutron activation equation is used to calculate the theoretical radioactivity of each target nuclide at the time of reactor shutdown. Further, step S33 includes: calling the neutron activation equation and calculating the theoretical radioactivity of each target nuclide at the end of the preset irradiation time t using the following formula: in, Let i be the number of target nuclei of the i-th target nuclide. The microscopic reaction cross section of the i-th target nuclide Let i be the neutron flux corresponding to the i-th target nuclide. Let be the decay constant of the i-th target nuclide. This is the preset irradiation duration.

[0031] Neutron flux data is input based on the typical neutron flux level during full-power operation in the Final Safety Analysis (FSAR) report of the reference power plant, or measured neutron flux distribution data at specific locations (axial / radial) in the reactor core. To cover the possibility of different foreign object introduction times, the following three hypothetical scenarios are constructed: Scenario A is a short-term introduction, assuming the foreign object falls in after the overhaul is opened, and the irradiation time is... Theoretical activation activity Scenario B is a single-cycle legacy, assuming the foreign object enters at the beginning of the previous refueling cycle, and the irradiation time t1 is typically 12 to 18 months; Scenario C is a multi-cycle legacy, assuming the foreign object exists for more than one refueling cycle, and the irradiation time... Where n is the number of periods, The duration of a single fuel cycle is given. The theoretical radioactivity under each scenario is calculated based on the neutron activation kinetics equation, where N is the number of target nuclei. For the microscopic reaction cross section, Neutron flux Let be the decay constant. Then, consider the time interval from the reactor shutdown time to the actual measurement time. The theoretical activity is decay-corrected to ensure that the time reference between the theoretical activity and the measured radionuclide activity is consistent.

[0032] In one specific embodiment, it is initially assumed that the foreign object remains in the reactor core at full power for 530 days (one fuel cycle). Based on the actual location of the foreign object in the reactor core (near the lower end grid of the fuel assembly), and referring to the typical neutron flux distribution data of FSAR at full power operation in Table 2, a neutron flux of 1E+13n / cm²·s is selected for neutron activation calculation. Based on the mass proportions of each element in 316 series stainless steel (Cr 17%, Ni 12%, Fe 66%, Co-59 0.1%), the number of target nuclei of each target nuclide is calculated, and the theoretical radioactivity is calculated by substituting them into the neutron activation equation. The decay during the period from reactor shutdown to retrieval measurement is also considered. =7 days) for correction. The calculation results are shown in Table 3: The theoretical calculated activity of Cr-51 is 3.14E+8 Bq, which is basically consistent with the energy dispersive spectroscopy (EDS) measured activity of 3.09E+8 Bq, and has reached the saturation activity of 3.14E+8 Bq; the theoretical calculated activity of Co-58 is 2.14E+7 Bq, which is on the same order of magnitude as the EDS measured activity of 6.76E+7 Bq, and has reached the saturation activity of 2.14E+7 Bq; the theoretical calculated activity of Co-60 is 1.81E+7 Bq, and the EDS measured activity is 2.08E+7 Bq, with a deviation of 14.9%, while its saturation activity is 1.04E+8 Bq, indicating that Co-60 is in an unsaturated growth stage.

[0033] Table 2 Typical values ​​of neutron flux distribution and energy spectrum Table 3: Comparison of activities of different nuclides Step S34: Based on the decay time interval, the theoretical radioactivity of each target nuclide at the time of shutdown is corrected for decay, and the theoretical radioactivity after decay correction is obtained as the theoretical radionuclide activity.

[0034] Further, step S34 includes: Obtain the time interval between heap decays ; The theoretical radioactivity of each target nuclide at the time of reactor shutdown is corrected for decay using the following formula: in, Preset irradiation duration Theoretical radioactivity at the end of the event. For the first The decay constant of the target nuclide, This is the time interval between reactor shutdown decays; The theoretical radioactivity after decay correction As a theoretical measure of radionuclide activity.

[0035] Furthermore, the feature data also includes the maximum linear dimension; step S4 includes: Step S41: Determine the relationship between each measured distance r and the maximum linear dimension. Relationship; Step S42: When At that time, a point source model was used to obtain the specific kerma rate constant of each target nuclide. The theoretical γ dose rate is calculated using the following formula: in, This represents the theoretical radioactivity after decay correction for the i-th target nuclide. For the first The specific kerma rate constant of the target nuclide; Step S43: When r < 5 × L_max, call the Monte Carlo simulation program to establish a three-dimensional geometric model of the foreign object, input the elemental composition and nuclide activity distribution, and simulate and calculate the theoretical γ dose rate H_theo(r) at the detection point.

[0036] Specifically, in the point source model calculation, if the foreign object meets the point source condition (the measurement distance r is greater than or equal to 5 times the maximum geometric size L of the foreign object), the theoretical dose rate is calculated using the specific kerma rate constant Γ. The calculation formula is as follows: Where A_final,i is the theoretical radioactivity of the i-th target nuclide after decay correction. For the first The specific kerma rate constant of the target nuclide, For distance measurement. When modeling complex geometry, if the gamma dose rate measurement distance r of the foreign object is less than 5 times the maximum geometric size L of the foreign object, the point source model fails. In this case, a Monte Carlo simulation program (such as MCNP, CIRPDose) is used to establish a three-dimensional model, input the elemental composition and nuclide activity distribution, and simulate and calculate the dose rate at the detection point.

[0037] In one specific embodiment, the maximum geometric size of the foreign object is L≈9.5mm, and the measurement distances r are 5cm and 10cm, respectively, both satisfying r≥5×L (5×0.95cm=4.75cm). Therefore, a point source model is used for calculation. The theoretical radioactivity A_final,i after decay correction for each target nuclide and the corresponding kerma rate constant Γ_i (Γ=1.9×10−8(mSv·m²) / (Bq·h) for Cr−51, Γ=5.6×10−7(mSv·m²) / (Bq·h) for Co-58, and Γ=1.3×10−6(mSv·m²) / (Bq·h) for Co-60) are obtained and substituted into the point source model formula to further calculate the theoretical γ dose rate at 5cm and 10cm from the foreign object. The calculation results are shown in Table 4: When the neutron flux parameter is 1.00E+13n / cm²·s, the dose rate at 5cm is 1.145mSv / h and the dose rate at 10cm is 0.286mSv / h.

[0038] Table 4: Theoretical Calculation Results of Foreign Body Dose Rate Further, step S5 includes: Step S51: Calculate the dose rate deviation factor using the following formula. and activity deviation factor : in, H_meas is the measured gamma dose rate, H_theo is the theoretical gamma dose rate, A_meas is the measured radionuclide activity when the foreign object was retrieved, and A_theo_final is the theoretical radionuclide activity when the foreign object was retrieved. Step S52: When both the dose rate deviation factor R_H and the activity deviation factor R_A approach 1, determine that the actual irradiation time of the foreign object is consistent with the corresponding preset irradiation time; otherwise, iteratively adjust the irradiation time. The theoretical radiation value is recalculated until convergence, thus obtaining the preliminary irradiation duration.

[0039] Dose rate dimension comparison: Calculate the deviation factor R_H = H_meas / H_theo between the measured gamma dose rate H_meas and the theoretical gamma dose rate H_theo; Activity dimension comparison: Calculate the deviation factor between the measured nuclide activity A_meas and the theoretical activity A_theo. The judgment logic is as follows: if both R_H and R_A are close to 1, the foreign object is determined to have undergone neutron activation for the time corresponding to the preset scenario (e.g., scenario B corresponds to one refueling cycle); otherwise, the irradiation duration t needs to be iteratively increased or decreased, and the theoretical radiation value needs to be recalculated until convergence. Since there may be some error in the measurement of gamma dose rate, and the parameters introduced in the theoretical calculation of gamma dose rate may also have some error, in order to improve the accuracy of the analysis and judgment, it is also necessary to combine it with radioactivity for comprehensive analysis. A two-dimensional cross-validation method can effectively eliminate the error interference caused by a single measurement method or a single calculation parameter, significantly improving the confidence and reliability of the judgment results.

[0040] In one specific embodiment, the measured dose rate at a distance of 5 cm from the foreign object was 1.7 mSv / h, and the measured dose rate at a distance of 10 cm from the foreign object was 0.744 mSv / h. The measured values ​​were compared with the theoretical calculation results in Table 4. The measured values ​​were close to the theoretical calculation results (1.145 mSv / h at 5 cm and 0.286 mSv / h at 10 cm) for an irradiation time of one fuel cycle (approximately 530 days). Combined with the Co-60 activity comparison results (the deviation factor R_A ≈ 1.149 between the measured radionuclide activity of 2.08E+7Bq and the theoretical activity of 1.81E+7Bq), and considering the comparison results of both dose rate and activity, it was preliminarily determined that the foreign object had been in the reactor core for at least one fuel cycle.

[0041] Furthermore, the method also includes: based on the initial irradiation duration, selecting at least one short-half-life nuclide and at least one long-half-life nuclide; when the short-half-life nuclide has reached saturation while the long-half-life nuclide has not, the irradiation duration is deduced from the growth curve of the long-half-life nuclide to obtain the deduced irradiation duration; the initial irradiation duration is corrected using the deduced irradiation duration to obtain the final irradiation duration.

[0042] Furthermore, by extrapolating the irradiation duration from the growth curves of long-half-life nuclides, the extrapolated irradiation duration includes: Measured radioactive nuclide activity of long half-life nuclides and its theoretical saturation activity N represents the number of target nuclei atoms. For the microscopic reaction cross section, Neutron flux The decay constant; Calculate the back-derived irradiation duration using the following formula : in, The decay constant of a long-half-life nuclide; The initial irradiation duration was corrected by reverse-calculating the irradiation duration, resulting in the final irradiation duration, which includes: Determine whether the retrospective irradiation duration is within the range of the initial irradiation duration; If present, the reverse irradiation duration will be used as the final irradiation duration output. If not, select another combination of a short-half-life nuclide and a long-half-life nuclide, and repeat the above reverse calculation steps until the obtained reverse irradiation duration falls within the range of the initial irradiation duration. Then, take the reverse irradiation duration as the final irradiation duration.

[0043] This step utilizes the difference in the time required for different radionuclides to reach saturation activity (t_sat≈6.65T1 / 2) for cross-validation, by analyzing short-lived nuclides (such as... 51 Cr) and long-lived nuclides (such as 60 The activity ratio of radioactive nuclides (Co) can be used to determine the irradiation time t. If short-lived nuclides are saturated while long-lived nuclides are not, the irradiation time t can be precisely calculated from the growth curve of the long-lived nuclide, achieving "precise timing." The expression for the radioactive nuclide growth curve (activity) is: .

[0044] Judgment logic and basis: Select characteristic nuclides with different half-lives as "time scales".

[0045] Short-lived nuclides (Cr-51, T1 / 2 = 27.7 days): saturation time is approximately 180 days. If its activity has reached saturation, it can only prove that the foreign object has remained in the reactor core for more than six months, but it cannot distinguish the specific duration.

[0046] Medium half-life nuclides (Co-58, T1 / 2 = 70.8 days): saturation time is approximately 460 days. Their state can be used to define a time window of six months to one and a half years.

[0047] Long half-life radionuclides (Co-60, T1 / 2 = 5.27 years): It takes about 15.8 years to reach 95% saturation, which is much longer than a single fuel cycle. Their radioactivity is extremely sensitive to irradiation time and is a key basis for precise timing.

[0048] Verification of the saturation state of short- and medium-half-lived nuclides: Cr-51 Validation: Energy Dispersive X-ray Spectroscopy (EDS) Activity Measurement With theoretical saturation activity The high degree of agreement indicates that the Cr-51 produced by the foreign object in the reactor core after neutron activation has reached saturation, confirming that the foreign object has been in the reactor core for a period of t > 180 days.

[0049] Co-58 Validation: Activity Measurement by Energy Dispersive Spectroscopy With theoretical saturation activity The activity was on the same order of magnitude, and the measured activity of the energy spectrum was higher than the theoretical saturation activity (the positive deviation may be caused by the local fast neutron flux in the core being too high), which ruled out the possibility that "Co-58 did not reach saturation and the irradiation time was less than 460 days". The lower limit of the irradiation time was further constrained to t>460 days, which matched the time range of a single fuel cycle (530 days).

[0050] Quantitative back-calculation of irradiation duration for long-half-life nuclides: Co-60 validation: Energy dispersive spectroscopy activity measured 2.08 × 10⁻⁸ Bq is less than its theoretical saturation activity of 1.04 × Bq indicates that Co-60 is in an unsaturated growth stage, and its activity is positively correlated with irradiation time, which can be used as a core indicator for quantitative determination. Assuming the foreign substance undergoes one fuel cycle (approximately 530 days), the theoretically calculated Co-60 activity is 1.81 × Bq, with an activity of 2.08 × 10⁻⁸ measured by energy dispersive spectroscopy. The Bq deviation was 14.9%.

[0051] Assuming the foreign substance undergoes two fuel cycles, the theoretically calculated Co-60 activity is 3.36 × Bq, with an activity of 2.08 × 10⁻⁸ measured by energy dispersive spectroscopy. The Bq deviation reached 61.5%.

[0052] Based on the operation records of the previous round of reactor overhaul inspections where no foreign objects were found, it was ultimately determined that the foreign object remained in the reactor core for one fuel cycle, a probability significantly higher than the dual fuel cycle hypothesis.

[0053] In one embodiment, it should be noted that the preliminary irradiation duration is an estimated value obtained by iterative calculation after convergence, based on the comparison results of the dose rate deviation factor R_H and the activity deviation factor R_A. The final irradiation duration is obtained by using the saturation state of short-half-life nuclides as a time range constraint on the preliminary irradiation duration, and then calculating it backwards from the unsaturated growth curve of long-half-life nuclides. After cross-validation by multiple nuclides, the final irradiation duration is corrected. This value satisfies the following: the measured radioactive nuclide activity of short-half-life nuclides (such as ^51Cr) is basically consistent with the theoretical saturation activity (saturation has been reached), and the deviation between the back-calculated irradiation duration of long-half-life nuclides (such as ^60Co) and the preliminary judgment result is within an acceptable range (such as <20%).

[0054] This invention has the following advantages and beneficial effects: First, it is scientifically quantified, applying for the first time to the quantitative assessment of the introduction time of foreign objects into the reactor core of a nuclear power plant, transforming traditional qualitative experience-based judgments into quantitative physical calculations, fundamentally overcoming the subjectivity and ambiguity of traditional methods that rely on human experience for estimation; Second, it employs a dual verification mechanism, uniquely creating… The "dose rate + radionuclide activity" two-dimensional comparison algorithm compares actual measurements. Dose rate and theoretical Cross-comparison of dose rate, measured nuclide activity, and theoretical nuclide activity effectively eliminates the error interference caused by a single measurement method or a single calculation parameter, significantly improving the confidence and reliability of the judgment results. Third, precise time judgment: by utilizing the differences in growth-decay characteristics of nuclides with different half-lives, the time range is constrained by the saturation state of short-half-lived nuclides (such as ^51Cr), and quantitative back-inference is performed by the unsaturated growth curve of long-half-lived nuclides (such as ^60Co). This not only accurately distinguishes between "new foreign objects" and "old foreign objects," but also accurately estimates the specific residence time of foreign objects in the reactor core. Fourth, strong engineering applicability: the neutron flux data, material composition data, and γ dose rate data required by this method are all routine monitoring data of nuclear power plants, requiring no additional complex measurements. The calculation model can be solidified into a software module, facilitating rapid on-site implementation and promotion. Fifth, safety decision support: this method can provide conclusive physical evidence for root cause analysis (RCA) of foreign object events, directly guiding targeted improvements to foreign object prevention procedures (FME), and effectively improving the level of nuclear power plant operation safety management.

[0055] In terms of power plant applications, this invention can be directly embedded into the foreign matter handling process during nuclear power plant overhauls. It serves as a standard technical tool for identifying the source of foreign matter and determining the introduction time. It is applicable to various reactor types, such as pressurized water reactors (PWRs), and provides nuclear power plants with full-process technical support from foreign matter discovery, activation analysis, time determination to conclusion output.

[0056] The contribution of this invention to power plant productivity lies in the following: By rapidly and accurately determining the introduction time and source of foreign objects (FOOs) in the reactor core, this invention significantly reduces the extended overhaul and maintenance time caused by FEO detection, optimizes overhaul cycles and resource allocation, and reduces maintenance costs and resource waste. Simultaneously, through precise source tracing analysis of FEO sources, this method can provide a scientific basis for root cause analysis and safety management of FEO incidents in nuclear power plants, effectively preventing the recurrence of similar FEO incidents, thereby improving the operational reliability, economy, and safety of the power plant.

[0057] It is understood that the above embodiments only illustrate preferred embodiments of the present invention, and their descriptions are relatively specific and detailed, but they should not be construed as limiting the scope of the present invention. It should be noted that those skilled in the art can freely combine the above technical features without departing from the concept of the present invention, and can also make several modifications and improvements, all of which fall within the protection scope of the present invention. Therefore, all equivalent transformations and modifications made with respect to the scope of the claims of the present invention should fall within the scope of the claims of the present invention.

Claims

1. A method for quantitatively assessing the activation time of metallic foreign matter in a nuclear power plant reactor core, characterized in that, Includes the following steps: Step S1: Obtain the characteristic data of the foreign object, including the measured value of the gamma spectrum, the neutron flux, and the measured gamma dose rate at different measurement distances; Step S2: Obtain the measured activity of the radionuclide based on the measured value of the γ energy spectrum; Step S3: Based on the neutron flux and at least two preset irradiation durations, calculate the theoretical radionuclide activity of the foreign object under each preset irradiation duration; Step S4: Calculate the theoretical γ dose rate of the foreign object under each preset irradiation duration based on the theoretical radionuclide activity corresponding to each preset irradiation duration; Step S5: Compare the measured γ dose rate at different measurement distances with the theoretical γ dose rate corresponding to each preset irradiation duration, compare the measured radionuclide activity with the theoretical radionuclide activity corresponding to each preset irradiation duration, and determine the actual irradiation duration of the foreign object in the reactor core based on the comparison results.

2. The method for quantitatively assessing the activation time of metallic foreign matter in a nuclear power plant core according to claim 1, characterized in that, The measured values ​​of the γ-ray spectrum were analyzed to identify characteristic peaks, and multiple target nuclides were selected. The measured radioactivity of each target nuclide after the metal foreign object was retrieved was calculated. .

3. The method for quantitatively assessing the activation time of metallic foreign matter in a nuclear power plant core according to claim 2, characterized in that, The feature data also includes the foreign object material type and the total mass of the foreign object; step S3 includes: Step S31: Obtain the decay constant and the corresponding microscopic reaction cross section of each target nuclide; Step S32: Based on the material type of the foreign object, obtain the number of target nuclei atoms for each target nuclide; Step S33: Based on the number of target nuclei atoms of each target nuclide, the neutron flux, the microscopic reaction cross section, the decay constant, and the preset irradiation duration... The neutron activation equation is used to calculate the theoretical radioactivity of each target nuclide at the time of reactor shutdown. Step S34: Based on the decay time interval, the theoretical radioactivity of each target nuclide at the time of shutdown is corrected for decay, and the theoretical radioactivity after decay correction is obtained as the theoretical radioactivity activity of the nuclide.

4. The method for quantitatively assessing the activation time of metallic foreign matter in a nuclear power plant core according to claim 3, characterized in that, Step S32 includes: The elemental mass percentage of the target nucleus corresponding to each target nuclide is determined based on the material type of the foreign object. Calculate the number of target nuclei for each target nuclide using the following formula. : in, Let represent the elemental mass percentage of the target nucleus corresponding to the i-th target nuclide. The total mass of the foreign object. Let Avogadro's constant be 1. Let be the atomic weight of the target nucleus corresponding to the i-th target nuclide.

5. The method for quantitatively assessing the activation time of metallic foreign matter in a nuclear power plant core according to claim 3, characterized in that, Step S33 includes: Using the neutron activation equation, calculate the theoretical radioactivity of each target nuclide at the end of the preset irradiation time t using the following formula: in, Let i be the number of target nuclei of the i-th target nuclide. The microscopic reaction cross section of the i-th target nuclide Let i be the neutron flux corresponding to the i-th target nuclide. Let be the decay constant of the i-th target nuclide. This is the preset irradiation duration.

6. The method for quantitatively assessing the activation time of metallic foreign matter in a nuclear power plant core according to claim 3, characterized in that, Step S34 includes: Obtain the decay time interval from the time the foreign object was stopped to the time it was retrieved. ; The theoretical radioactivity of each target nuclide at the time of reactor shutdown is corrected for decay using the following formula: in, Preset irradiation duration Theoretical radioactivity at the end of the event, For the first The decay constant of the target nuclide, The decay time interval from when the foreign object was stopped from being retrieved; The theoretical radioactivity after decay correction Theoretical radionuclide activity when retrieved as a foreign object.

7. The method for quantitatively assessing the activation time of metallic foreign matter in a nuclear power plant core according to claim 1, characterized in that, The feature data also includes the maximum linear dimension; step S4 includes: Step S41: Determine the relationship between each measured distance r and the maximum linear dimension. Relationship; Step S42: When At that time, a point source model was used to obtain the specific kerma rate constant of each target nuclide. The theoretical γ dose rate is calculated using the following formula: in, This represents the theoretical radioactivity after decay correction for the i-th target nuclide. For the first The specific kerma rate constant of the target nuclide; Step S43: When r < 5 × L_max, call the Monte Carlo simulation program to establish a three-dimensional geometric model of the foreign object, input the elemental composition and nuclide activity distribution, and simulate and calculate the theoretical γ dose rate H_theo(r) at the detection point.

8. The method for quantitatively assessing the activation time of metallic foreign matter in a nuclear power plant core according to claim 1, characterized in that, Step S5 includes: Step S51: Calculate the dose rate deviation factor using the following formula. and activity deviation factor : in, H_meas is the measured gamma dose rate, H_theo is the theoretical gamma dose rate, A_meas is the measured radionuclide activity, and A_theo_final is the theoretical radionuclide activity when the foreign object was retrieved. Step S52: When both the dose rate deviation factor R_H and the activity deviation factor R_A approach 1, determine that the actual irradiation time of the foreign object is consistent with the corresponding preset irradiation time; otherwise, iteratively adjust the irradiation time. The theoretical radiation value is recalculated until convergence, thus obtaining the preliminary irradiation duration.

9. The method for quantitatively assessing the activation time of metallic foreign matter in a nuclear power plant core according to claim 8, characterized in that, Also includes: Based on the initial irradiation duration, at least one short-half-life nuclide and at least one long-half-life nuclide are selected. When the short-half-life nuclide has reached saturation while the long-half-life nuclide has not, the irradiation duration is deduced from the growth curve of the long-half-life nuclide to obtain the deduced irradiation duration. The initial irradiation duration is then corrected using the deduced irradiation duration to obtain the final irradiation duration.

10. The method for quantitatively assessing the activation time of metallic foreign matter in a nuclear power plant core according to claim 9, characterized in that, The process of inferring the irradiation duration from the growth curve of long-half-life nuclides, and obtaining the inferred irradiation duration, includes: Measured radioactive nuclide activity of long half-life nuclides and its theoretical saturation activity N represents the number of target nuclei atoms. For the microscopic reaction cross section, Neutron flux The decay constant; Calculate the back-derived irradiation duration using the following formula : in, The decay constant of a long-half-life nuclide; The step of correcting the preliminary irradiation duration using the back-calculated irradiation duration to obtain the final irradiation duration includes: Determine whether the inverse irradiation duration is within the range of the initial irradiation duration; If so, the reversed irradiation duration will be output as the final irradiation duration; If not, select another combination of a short-half-life nuclide and a long-half-life nuclide, and repeat the above reverse calculation steps until the obtained reverse irradiation duration falls within the range of the preliminary irradiation duration. Then, take the reverse irradiation duration as the final irradiation duration.

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