Method and device for measuring activity concentration of primary loop of high-temperature gas cooled reactor

Abstract

The invention provides a high-temperature gas cooled reactor primary loop activity concentration measurement method and device, and the method comprises the steps: obtaining the first activity concentration of a reference object under the current working condition based on the working condition parameters of the current working condition and the calculation parameters of the reference object, and obtaining the first activity concentration of the reference object under the current working condition according to the working condition parameters of the current working condition and the calculation parameters of nuclide to be measured; obtaining the first activity concentration of the nuclide to be measured under the current working condition; acquiring a second activity concentration of the reference object under the current working condition; and obtaining the second activity concentration of the nuclide to be measured under the current working condition according to the first activity concentration of the nuclide to be measured under the current working condition, the first activity concentration of the reference object under the current working condition and the second activity concentration of the reference object under the current working condition. The device is used for executing the method. According to the method and the device for measuring the activity concentration of the primary loop of the high-temperature gas cooled reactor provided by the embodiment of the invention, the accuracy of measuring the activity concentration of the primary loop is improved.

Description

Technical Field

[0001] The invention relates to the technical field of nuclear energy application, and in particular to a method and a device for measuring the activity concentration of a primary loop of a high-temperature gas-cooled reactor. Background Art

[0002] An important support for the inherent safety of high temperature gas-cooled reactors is the use of fuel elements containing TRISO coated particles, which can retain most of the fission products under normal reactor conditions and accident situations. However, the damage of coated particles during the manufacturing process of fuel elements and uranium contamination in the process, natural uranium contamination in the matrix graphite material, irradiation damage of coated particles that may occur during reactor operation, and activation of impurities in the core graphite material and coolant will all cause the existence of a primary source term in the helium coolant of high temperature gas-cooled reactors under normal conditions.

[0003] The radioactivity level of the primary loop of a high-temperature gas-cooled reactor is crucial to the evaluation of the performance of the core fuel elements and the radiation safety of the reactor. On the one hand, the activity concentration of fission products in the primary loop, especially fission gas, directly reflects the state of the coating particles of the fuel elements in the core. On the other hand, it is also the basis for determining effluent emissions, source items of the secondary loop and other process systems, and accident source items. There are two main methods for determining the radioactivity of primary loop fission products: measurement of the radioactivity of the primary loop coolant and theoretical technology of the radioactivity of the primary loop coolant. In experimental research, it is mainly measured by online total γ activity concentration measurement of the primary loop pipeline or offline γ spectrum analysis of helium sampling in the primary loop. However, the above measurements are usually the results of stable operation for a period of time under typical working conditions. For radioactive nuclides with a short half-life, the experimentally measured primary loop activity concentration has a large error. In addition, due to the complexity of primary loop sampling, it is difficult to sample and measure the primary loop activity concentration of halogen nuclides. Therefore, how to propose a method for measuring the primary loop activity concentration and improve the accuracy and comprehensiveness of the primary loop activity concentration measurement has become an important issue to be solved in this field. Summary of the invention

[0004] In view of the problems in the prior art, an embodiment of the present invention provides a method and device for measuring the activity concentration of a primary loop of a high-temperature gas-cooled reactor, which can at least partially solve the problems in the prior art.

[0005] In a first aspect, the present invention provides a method for measuring the activity concentration of a primary loop of a high temperature gas-cooled reactor, comprising:

[0006] Based on the working condition parameters of the current working condition and the calculated parameters of the reference object, a first activity concentration of the reference object under the current working condition is obtained, and according to the working condition parameters of the current working condition and the calculated parameters of the nuclide to be measured, a first activity concentration of the nuclide to be measured under the current working condition is obtained;

[0007] Obtain the second activity concentration of the reference object under the current working condition; wherein, the second activity concentration of the reference object under the current working condition is obtained through the primary loop activity concentration experiment;

[0008] Obtain the second activity concentration of the nuclide to be measured under the current working condition according to the first activity concentration of the nuclide to be measured under the current working condition, the first activity concentration of the reference object under the current working condition, and the second activity concentration of the reference object under the current working condition.

[0009] In a second aspect, the present invention provides a measuring device for the primary loop activity concentration of a high-temperature gas-cooled reactor, comprising:

[0010] An obtaining unit, configured to obtain the first activity concentration of the reference object under the current working condition based on the working condition parameters of the current working condition and the calculation parameters of the reference object, and obtain the first activity concentration of the nuclide to be measured under the current working condition according to the working condition parameters of the current working condition and the calculation parameters of the nuclide to be measured;

[0011] An acquiring unit, configured to acquire the second activity concentration of the reference object under the current working condition; wherein, the second activity concentration of the reference object under the current working condition is obtained through the primary loop activity concentration experiment;

[0012] A measuring unit, configured to obtain the second activity concentration of the nuclide to be measured under the current working condition according to the first activity concentration of the nuclide to be measured under the current working condition, the first activity concentration of the reference object under the current working condition, and the second activity concentration of the reference object under the current working condition.

[0013] In a third aspect, the present invention provides a computer device, comprising a memory, a processor, and a computer program stored on the memory, and the processor executes the program to implement the method for measuring the primary loop activity concentration of a high-temperature gas-cooled reactor according to any one of the above embodiments.

[0014] In a fourth aspect, the present invention provides a computer-readable storage medium, which stores computer programs / instructions, and when the computer programs / instructions are executed by a processor, the method for measuring the primary loop activity concentration of a high-temperature gas-cooled reactor according to any one of the above embodiments is implemented.

[0015] In a fifth aspect, the present invention provides a computer program product, comprising computer programs / instructions, and when the computer programs / instructions are executed by a processor, the method for measuring the primary loop activity concentration of a high-temperature gas-cooled reactor according to any one of the above embodiments is implemented.

[0016] The measurement method and device for the primary circuit activity concentration of a high-temperature gas-cooled reactor provided by the embodiments of the present invention can obtain the first activity concentration of the reference object under the current working condition based on the working condition parameters of the current working condition and the calculation parameters of the reference object, and obtain the first activity concentration of the nuclide to be measured under the current working condition according to the working condition parameters of the current working condition and the calculation parameters of the nuclide to be measured; obtain the second activity concentration of the reference object under the current working condition; wherein, the second activity concentration of the reference object under the current working condition is obtained through experimental measurement of the primary circuit activity concentration; according to the first activity concentration of the nuclide to be measured under the current working condition, the first activity concentration of the reference object under the current working condition, and the second activity concentration of the reference object under the current working condition, obtain the second activity concentration of the nuclide to be measured under the current working condition, which can accurately measure the activity concentration of radionuclides with a short half-life, and improve the accuracy and comprehensiveness of the measurement of the primary circuit nuclide activity concentration. BRIEF DESCRIPTION OF THE DRAWINGS

[0017] In order to more clearly illustrate the technical solutions in the embodiments of the present invention or the prior art, the following will briefly introduce the drawings required for use in the description of the embodiments or the prior art. Obviously, the following drawings are only some embodiments of the present invention. For those of ordinary skill in the art, without creative efforts, other drawings can also be obtained based on these drawings. In the drawings:

[0018] Figure 1 is a schematic flowchart of the measurement method for the primary circuit activity concentration of a high-temperature gas-cooled reactor provided by the first embodiment of the present invention.

[0019] Figure 2a is a schematic structural diagram of Mode A of the decay chain provided by the second embodiment of the present invention.

[0020] Figure 2b is the nuclide provided by the second embodiment of the present invention 133m Xe and 133 is a schematic structural diagram of the decay chain of Xe.

[0021] Figure 3 is a schematic flowchart of the measurement method for the primary circuit activity concentration of a high-temperature gas-cooled reactor provided by the third embodiment of the present invention.

[0022] Figure 4 is a schematic flowchart of the measurement method for the primary circuit activity concentration of a high-temperature gas-cooled reactor provided by the fourth embodiment of the present invention.

[0023] Figure 5 is the experimental measurement value of the fission gas primary circuit activity concentration under different working conditions of HTR-10 provided by the fifth embodiment of the present invention.

[0024] Figure 6It is the theoretical calculated value of the activity concentration of gaseous fission nuclides in the primary circuit of HTR-10 under different operating conditions provided by the sixth embodiment of the present invention.

[0025] Figure 7 It is the primary circuit radioactive equilibrium diagram of HTR-10 at 10 MW provided by the seventh embodiment of the present invention.

[0026] Figure 8 It is the ratio of the theoretical value to the experimental value of the activity concentration of fission gases in the primary circuit of HTR-10 under different operating conditions provided by the eighth embodiment of the present invention.

[0027] Figure 9 It is the structural schematic diagram of the measuring device for the activity concentration of the primary circuit of a high-temperature gas-cooled reactor provided by the ninth embodiment of the present invention.

[0028] Figure 10 It is the structural schematic diagram of the measuring device for the activity concentration of the primary circuit of a high-temperature gas-cooled reactor provided by the tenth embodiment of the present invention.

[0029] Figure 11 It is the structural schematic diagram of the measuring device for the activity concentration of the primary circuit of a high-temperature gas-cooled reactor provided by the eleventh embodiment of the present invention.

[0030] Figure 12 It is the entity structural schematic diagram of the computer device provided by the twelfth embodiment of the present invention. Detailed implementation manners

[0031] To make the objectives, technical solutions, and advantages of the embodiments of the present invention clearer and more understandable, the following further elaborates on the embodiments of the present invention with reference to the accompanying drawings. Here, the illustrative embodiments of the present invention and their descriptions are used to explain the present invention, but do not limit the present invention. It should be noted that, without conflict, the embodiments and features in the embodiments in this application can be combined arbitrarily with each other. In the technical solutions of this application, the acquisition, storage, use, processing, etc. of data all comply with the relevant regulations of laws and regulations. The user information in the embodiments of this application is obtained through legal and compliant channels, and the acquisition, storage, use, processing, etc. of user information have obtained the authorization and consent of the customers.

[0032] To facilitate the understanding of the technical solutions provided in this application, the following first explains the relevant content of the technical solutions in this application.

[0033] Source term of high-temperature gas-cooled reactor: It refers to the type, quantity, release time, physical and chemical forms, and thermal energy of radioactive nuclides released from the reactor containment to the environment under normal operating conditions or assumed accident scenarios of the high-temperature gas-cooled reactor.

[0034] Primary circuit source term generally refers to the type, activity, etc. of radioactive substances generated in the primary circuit system of a nuclear reactor.

[0035] Core inventory: The total amount of radioactive substances contained in the core of a nuclear reactor.

[0036] In the calculation of the reactor source term, the radioactive source term specifically refers to the types and activities of radionuclides in a narrow sense, and is generally divided into the source term under normal operating conditions and the accident source term.

[0037] The primary circuit source term of a high-temperature gas-cooled reactor is divided into five categories from the perspective of radiation monitoring: tritium, carbon-14, radioactive dust, activation products, and fission products (including gaseous fission products and solid fission products).

[0038] The calculation model of the primary circuit activity concentration of short-lived gaseous fission products in a high-temperature gas-cooled reactor mainly consists of three parts: core inventory calculation, diffusion and release of nuclides in the core fuel elements, and the primary circuit nuclide migration model. The core inventory is the basis for source term calculation, and the total activity of radioactive nuclides in the core is mainly determined through the burnup equations of complex nuclide systems. However, due to the large number and complex properties of nuclides in the core, the generation and decay processes of fission products involve a large number of coupled nonlinear differential equations, making the solution process of the burnup equations extremely complex. The calculation of the diffusion and release of halogen nuclides and fission gases in HTGR fuel elements is based on the Booth model, which combines empirical derivation terms of temperature, burnup, thermal diffusion, and recoil mechanisms to varying degrees, as well as the distinction between core particles and matrix graphite. Currently, the primary circuit nuclide migration models for calculating the distribution and activity concentration of fission products in the primary circuit of a high-temperature gas-cooled reactor are divided into two types: zero-dimensional model and one-dimensional model. The zero-dimensional model does not consider the non-uniformity of the primary circuit physical parameters in the core, and the calculation is more conservative; while the one-dimensional model can consider the non-uniform physical parameters of the circuit, and the results are more accurate, but it increases the calculation parameters and the difficulty of obtaining engineering parameters. Moreover, the methods in the existing technologies cannot directly obtain the primary circuit activity concentration of each fission gas nuclide from the measured total activity concentration of the primary circuit.

[0039] Based on the existing theoretical calculation models, the present invention solves the limitations of existing models in practical applications by establishing a new calculation model for the gaseous fission products from the core to the primary circuit of a pebble-bed high-temperature gas-cooled reactor, and improves the theoretical framework for radioactive measurement of gaseous fission products in the primary circuit. And a theoretical calculation framework for the activity concentration of primary circuit nuclides applicable to high-temperature gas-cooled reactors is constructed, which combines the material property data, primary circuit operation parameters, and performance parameters of the core fuel elements of high-temperature gas-cooled reactors (such as the coating particle breakage rate and uranium contamination share) in the existing research to achieve accurate and comprehensive calculation of the activity concentration of primary circuit nuclides in the actual core.

[0040] Based on the physical relationships between short-lived inert gas fission nuclides and reactor operation parameters, as well as reactor power and burnup, this application established the IPGREGN model for rapidly calculating the inventory of fission nuclides in the reactor core to solve the problems of large, complex, and difficult-to-obtain parameters in conventional reactor core inventory calculation methods. Based on the high-temperature gas-cooled reactor, experimental measurements of primary circuit gas fission products were carried out at different power levels, a theoretical model for the diffusion and release of gas fission products from the reactor core fuel elements to the primary circuit was constructed, and a proportional method for source term analysis of short-lived iodine and fission gases in the primary circuit of the high-temperature gas-cooled reactor was proposed, that is, a measurement method for the activity concentration in the primary circuit of the high-temperature gas-cooled reactor, so as to be able to measure the activity concentration of gas fission products and halogen nuclides in the primary circuit.

[0041] Figure 1 is a schematic flow chart of the measurement method for the activity concentration in the primary circuit of the high-temperature gas-cooled reactor provided by the first embodiment of the present invention, as Figure 1 shown, the measurement method for the activity concentration in the primary circuit of the high-temperature gas-cooled reactor provided by the embodiment of the present invention includes:

[0042] S101. Based on the operating parameters of the current operating condition and the calculation parameters of the reference object, obtain the first activity concentration of the reference object under the current operating condition, and based on the operating parameters of the current operating condition and the calculation parameters of the nuclide to be measured, obtain the first activity concentration of the nuclide to be measured under the current operating condition;

[0043] Specifically, based on the constructed theoretical framework and calculation method for the activity concentration of radionuclides in the primary circuit of the high-temperature gas-cooled reactor, the theoretical activity concentration of the reference object under the current operating condition can be obtained based on the operating parameters of the current operating condition and the calculation parameters of the reference object, as the first activity concentration of the reference object under the current operating condition. And based on the operating parameters of the current operating condition and the calculation parameters of the nuclide to be measured, obtain the theoretical activity concentration of the nuclide to be measured under the current operating condition, as the first activity concentration of the nuclide to be measured under the current operating condition.

[0044] Among them, the current operating condition refers to the actual operating condition of the high-temperature gas-cooled reactor where the nuclide to be measured is located. The reference object refers to the reference nuclide in the primary circuit coolant or the set of fission gas nuclides in the primary circuit coolant. The reference nuclide can be a short-lived inert gas fission nuclide, such as 85m Kr, 87 Kr, 88 Kr or 135 Xe; the set of fission gas nuclides can be selected from the primary circuit fission gases, and the set of fission gas nuclides can include 85m Kr, 87 Kr, 88 Kr, 89 Kr, 133m Xe, 133Xe, 135m Xe, 135 Xe, 137 Xe and 138 Xe.

[0045] The operating condition parameters of the current operating condition include reactor characteristic parameters, reactor operating parameters, temperature, pressure, purification flow rate of the helium purification system, primary loop leakage removal coefficient, time for one cycle of helium circulation, time for helium to flow through the core per cycle, equivalent removal coefficient, equivalent release rate / generation rate of all fuel elements in the core, etc., which are selected according to actual needs and are not limited in the embodiments of the present invention. The calculation parameters of the reference object include the core inventory of the reference object, deposition rate of nuclides, neutron absorption cross-section of nuclides, etc.; the calculation parameters of the nuclide to be measured include the core inventory of the nuclide to be measured, deposition rate of the nuclide to be measured, neutron absorption cross-section of the nuclide to be measured, etc. The nuclide to be measured refers to gaseous fission products and halogen nuclides that need to be measured in the primary loop of a high-temperature gas-cooled reactor.

[0046] S102. Obtain the second activity concentration of the reference object under the current operating condition; wherein, the second activity concentration of the reference object under the current operating condition is obtained through on-line monitoring of the primary loop activity concentration experiment;

[0047] Specifically, the first activity concentration of the reference object under different operating conditions can be measured through the primary loop activity concentration experiment. In application, the second activity concentration of the reference object under the current operating condition can be obtained by query according to the current operating condition.

[0048] For example, the on-line monitoring of the primary loop activity concentration adopts a sampling method. The primary loop helium coolant containing radioactive nuclides enters the sampling tank through the intake pipeline of the sampling system after passing through the copper oxide bed. The primary loop helium coolant after sampling measurement can be discharged after being processed by the ventilation system through the outlet pipeline. After operating stably at a certain power for a period of time, a nuclide spectrogram is obtained by extracting the coolant from the primary loop. The activity concentration of radioactive nuclides in the primary loop coolant is obtained by analyzing the nuclide spectrogram of the primary loop, and then the activity concentration of radioactive nuclides in the primary loop coolant is corrected to obtain the second activity concentration of the reference object under the operating condition at the time of measurement. By changing the operating power of the high-temperature gas-cooled reactor and measuring the activity concentration of the reference object, the second activity concentration of the reference object under different operating conditions can be obtained.

[0049] S103. Obtain the second activity concentration of the nuclide to be measured under the current operating condition according to the first activity concentration of the nuclide to be measured under the current operating condition, the first activity concentration of the reference object under the current operating condition, and the second activity concentration of the reference object under the current operating condition.

[0050] Specifically, a calculation relationship can be pre-established among the first activity concentration of the nuclide to be measured under the current working condition, the first activity concentration of the reference object under the current working condition, the second activity concentration of the reference object under the current working condition, and the second activity concentration of the nuclide to be measured under the current working condition. After obtaining the specific values of the first activity concentration of the nuclide to be measured under the current working condition, the first activity concentration of the reference object under the current working condition, and the second activity concentration of the reference object under the current working condition, the second activity concentration of the nuclide to be measured under the current working condition can be obtained by combining the above calculation relationship.

[0051] The method for measuring the activity concentration of the primary loop of a high-temperature gas-cooled reactor provided by the embodiments of the present invention can obtain the first activity concentration of the reference object under the current working condition based on the working condition parameters of the current working condition and the calculation parameters of the reference object, and obtain the first activity concentration of the nuclide to be measured under the current working condition according to the working condition parameters of the current working condition and the calculation parameters of the nuclide to be measured; obtain the second activity concentration of the reference object under the current working condition; wherein, the second activity concentration of the reference object under the current working condition is obtained through experimental measurement of the primary loop activity concentration; according to the first activity concentration of the nuclide to be measured under the current working condition, the first activity concentration of the reference object under the current working condition, and the second activity concentration of the reference object under the current working condition, the second activity concentration of the nuclide to be measured under the current working condition can be obtained, which can accurately measure the activity concentration of radioactive nuclides with a short half-life, improve the accuracy of the measurement of the primary loop nuclide activity concentration, and can realize the measurement of the activity concentration of halogen nuclides, improving the comprehensiveness of the measurement of the primary loop nuclide activity concentration.

[0052] This application constructs a theoretical framework and calculation method for the activity concentration of radioactive nuclides in the primary loop of a high-temperature gas-cooled reactor, including the IPRFGN model for calculating the core inventory, the nuclide diffusion and release model, and the zero-dimensional primary loop nuclide migration model. Based on the above three models, the core inventory, release rate / generation rate, and equivalent removal coefficient can be determined respectively, which can be applied to the theoretical calculation of the activity concentration of primary loop nuclides.

[0053] The inventory calculation method used in this application is based on the IPRFGN model, which treats short-lived halogen nuclides and fission gases ( 133 I, 135 I, 133 Xe, 133m Xe, 135 Xe, 135m Xe, 85m Kr, 89 Kr, 87 Kr, 88 Kr, 137 Xe and 138The decay chains of Xe etc. are summarized into two modes: A (divided into four sub - modes) and B. The IPRFGN model gives the analytical solution of its core inventory in the form of:

[0054] I i =∑ k c k A f,i,k (1)

[0055] Where, I i is the core inventory of nuclide i, c k is the coefficient of A f,i,k of nuclide k on the decay chain of nuclide i, k is the number of different nuclides on the decay chain of nuclide i, A f,i,k is the amount of nuclide i produced by a pre - selected fissile nuclide, and is a function of the final state power P l (in MW / tHM) and the average burnup B of the core fuel:

[0056] A f,i,k =∑ j Y j,k σ f,j φ 2 N j , j=U5,U8,Pu9,Pu1 (2)

[0057] Where, Y U5,k is the fission fraction of nuclide k produced by U - 235 fission, Y U8,k is the fission fraction of nuclide k produced by U - 238 fission, Y Pu9,k is the fission fraction of nuclide k produced by Pu - 239 fission, Y Pu1,k is the fission fraction of nuclide k produced by Pu - 241 fission; σ f,U5 is the fission cross - section of U - 235, σ f,U8 is the fission cross - section of U - 238, σ f,Pu9 is the fission cross - section of Pu - 239, σ f,Pu1 is the fission cross - section of Pu - 241; σ a,i is the spectrum - averaged neutron absorption cross - section of nuclide i; φ 2 is the reduced neutron flux in the core; N U5 is the number of atoms of U - 235, N U8 is the number of atoms of U - 238, N Pu9 is the number of atoms of Pu - 239, N Pu1 is the number of atoms of Pu - 241. The pre - selected fissile nuclide can be selected according to the actual fuel composition and experience, and is not limited in the embodiments of the present invention.

[0058] Next, taking the nuclide 133Taking the calculation process of the core inventory of Xe as an example for illustration.

[0059] Based on Figure 2a the structural schematic diagram of Mode A of the decay chain shown, when the nuclide is determined, the decay chains of nuclides 133m Xe and 133 Xe can be obtained, as shown in Figure 2b shown.

[0060] Based on Figure 2a , Figure 2b and formula (1), the analytical solution form of the core inventory of 133 Xe can be obtained as follows:

[0061]

[0062] where, I 01 represents the core inventory of nuclide 133 Xe, A f,1 represents the amount of nuclide 133 Xe produced by nuclide 133 Sb in decay chain No. 1 of 133 Xe, A f,21 represents the amount of nuclide 133 Xe produced by nuclide 133 Te + n in decay chain No. 21 of 133 Xe, A f,22 represents the amount of nuclide 133 Xe produced by nuclide 133m Te in decay chain No. 22 of 133 Xe, A f,3 represents the amount of nuclide 133 Xe produced by nuclide 133 I in decay chain No. 3 of 133 Xe, A f,02 represents the amount of nuclide 133 Xe produced by nuclide 133m Xe in decay chain No. 02 of 133 Xe, A f,01 represents the amount of nuclide 133 Xe produced by nuclide 133 Xe in decay chain No. 01 of 133 Xe; the coefficient of A 133 for nuclide 133 Sb on the decay chain of f,1 Xe, the coefficient of A 133 for nuclide 133 Te + n on the decay chain of f,21 Xe, the coefficient of A 133 for nuclide 133m Te on the decay chain of f,22Coefficient, nuclide 133 Nuclides on the Xe decay chain 133 A of I f,3 Coefficient, nuclide 133 Nuclides on the Xe decay chain 133m A of Xe f,02 Coefficient, nuclide 133 Nuclides on the Xe decay chain 133 A of Xe f,01 The coefficients are all equal, being λ 01 For nuclide 133 Decay constant of Xe, φ 2 For the reduced neutron flux in the core, σ a,01 For nuclide 133 Spectrum-averaged neutron absorption cross section of Xe. Reactor characteristic parameters include the reduced neutron flux φ 2 in the core and the numbers of atoms N U5 of the fissile nuclides U-235, U-238, Pu-239, and Pu-241 U8 、N Pu9 、N Pu1 ;Reactor operating parameters include the power P final at the final actual operating state of the core and the average burnup B of the core fuel. Select any P final and fix it as a reference, denoted as P ref . According to the reduced neutron flux φ ref in the core and the data of the average burnup B of different fuel elements under the reference reduced final-stage reactor power P 2 , the sub-parameters a Φ , b Φ and c Φ can be obtained by fitting. Through the operating condition parameters P final , B and P ref , and the following formula, the reduced neutron flux φ 2 in the core can be calculated.

[0063]

[0064] According to the numbers of atoms N U5 、N U8 、N Pu9 、N Pu1 of the fissile nuclides U-235, U-238, Pu-239, and Pu-241 and the average burnup B of the fuel elements, the sub-parameters a i , b i and c i of the numbers of atoms of the fissile nuclides can be obtained by fitting. Through the operating condition parameter B and the following formula, the number of atoms N j of the fissile nuclide j can be calculated.

[0065] N j = (a j B 2 + b j B + c j )(4)

[0066] where: j = U5, U8, Pu9, Pu1.

[0067] For the release rate / generation rate, the release of halogen nuclides and fission gas nuclides has three sources: intact coated particles, damaged coated particles, and uranium contamination in the graphite matrix material. Considering that the diffusion coefficient of fission gas nuclides in the SiC coating layer is very small and their half-lives are relatively short, the diffusion of fission gas nuclides through intact coated particles can be ignored. Both damaged coated particles and uranium contamination in the graphite matrix material can be described by equivalent spheres. At the same time, because the half-lives of fission gas nuclides are relatively short (compared with the diffusion time of nuclides in the fuel core), the time to reach equilibrium can be ignored, and the diffusion equation is solved under steady-state conditions. The three uranium-containing parts in the fuel element considered in this application that can generate and release halogen nuclides and fission gases into the primary loop include: damaged coated particles, matrix graphite grains, and amorphous carbon. Due to the presence of the SiC layer, intact coated particles have extremely strong retention ability for halogen nuclides and fission gases and are considered to completely retain halogen nuclides and fission gases. The UO 2 particles generated by process uranium contamination in matrix graphite are equivalently considered in the above three parts.

[0068] The equivalent release rate / generation rate of all fuel elements in the core is:

[0069]

[0070] where the breakage rate of the coated particles of the fuel element in the high-temperature gas-cooled reactor is F, and the uranium contamination share in the matrix graphite of the fuel element is C, and represent the release rate / generation rate of halogen nuclides and fission gases from the equivalent damaged coated particles and the release rate / generation rate of uranium contamination in the core fuel elements respectively, and are the release rate / generation rate of halogen nuclides and fission gases from the equivalent spheres of a single damaged coated particle and the equivalent spheres of uranium contamination respectively, and are the release rate / generation rate of halogen nuclides and fission gases from the equivalent spheres of a single matrix graphite grain and the release rate / generation rate of the equivalent spheres of amorphous carbon UO 2 particles.

[0071] At different temperatures, the diffusion coefficients of nuclides in different materials are different. When calculating the release rate / generation rate of halogen nuclides and fission gases in the equivalent damaged coated particles and uranium contamination of the reactor core fuel elements, the equivalent spheres of single damaged coated particles and uranium contamination, and single matrix graphite grains B and amorphous carbon UO 2 The temperature distribution of the fuel element needs to be considered for the release rate / generation rate of the equivalent spheres of the particles. The results are directly calculated at the corresponding temperatures and then weighted and averaged according to the temperature share.

[0072] The Booth equation is the basis of the equivalent sphere model for all short-lived inert gas fission products. This model has an analytical solution, expressed as:

[0073]

[0074] Among them, a is the radius of the equivalent sphere, R / B is the release rate / generation rate of nuclide release, D is the diffusion coefficient of the nuclide, λ eff = λ + σ a φ 1 is the equivalent decay constant of the nuclide considering decay and neutron absorption, λ is the decay constant of the nuclide, σ a is the spectral average neutron absorption cross-section of the nuclide, φ 1 is the equivalent neutron flux of the reactor core.

[0075] In addition, D' = D / a 2 is defined as the reduced diffusion coefficient of the nuclide and is expressed by the Arrhenius formula as:

[0076]

[0077] Among them, D 0 ′ is the frequency factor, e is the natural constant, Q is the activation energy, R g is the ideal gas constant, and T is the temperature of the diffusion material. The diffusion materials include but are not limited to uranium dioxide cores, grains of matrix graphite materials, amorphous carbon, etc.

[0078] The halogen nuclides and fission gases released into the primary coolant are transported from the reactor core region to the primary coolant system with the primary coolant. During the operation of the reactor, these radioactive nuclides are continuously removed by decay, purification by the helium purification system, deposition on the surfaces in contact with the primary coolant (fission gases can be considered to have no deposition phenomenon), a small amount of leakage during loop circulation, and absorption of neutrons when passing through the reactor core (mainly 135 Xe). The generation and removal of nuclides reach a dynamic equilibrium state after a certain period of time.

[0079] The theoretical value of the activity concentration A (Bq / L) at equilibrium is:

[0080]

[0081] λ eff = λ + σ a φ 1 (12)

[0082] where Re is the equivalent release rate of the nuclide from the core, I represents the core inventory of the nuclide, ε is the purification coefficient of the nuclide, Q is the purification flow rate of the helium purification system, V is the total volume of the primary circuit helium, δ is the deposition rate of the nuclide, t c is the time for one cycle of the helium circulation, ω is the primary circuit leakage removal coefficient, σ a is the spectral-averaged neutron absorption cross-section of the nuclide, φ 1 is the equivalent neutron flux of the core, t v is the time for the helium to flow through the core per cycle, τ is the equivalent removal coefficient, λ eff is the equivalent decay constant, is the equivalent release rate / generation rate of all fuel elements in the core, and λ is the decay constant of the nuclide.

[0083] Figure 3 is a schematic flow diagram of the method for measuring the activity concentration of the primary circuit of a high-temperature gas-cooled reactor provided by the third embodiment of the present invention. As Figure 3 shown, on the basis of the above embodiments, further, obtaining the first activity concentration of the target object under the current working condition based on the working condition parameters of the current working condition and the calculation parameters of the target object includes:

[0084] S301 Obtain the core inventory corresponding to the target object according to the amount of the target object generated by the pre-selected fissionable nuclide and the coefficient corresponding to the target object; wherein, the target object is a reference object or a nuclide to be measured;

[0085] Specifically, based on the amount of the target object generated by the pre-selected fissionable nuclide and the coefficient corresponding to the target object, the core inventory corresponding to the target object can be obtained. Wherein, the target object is a reference nuclide or a nuclide to be measured; the reference object includes a reference nuclide. If the target object is a reference nuclide, the core inventory of the reference nuclide can be calculated based on formula (1). The coefficient corresponding to the reference nuclide is the A f,i,k coefficient of nuclide k on the decay chain of the reference nuclide. When the reference nuclide is determined, the coefficient corresponding to the reference nuclide is also determined and can be obtained through calculation. The amount of the reference nuclide generated by the pre-selected fissionable nuclide can be calculated based on formula (2).

[0086] For example, the calculation formula for obtaining the core inventory corresponding to the reference nuclide based on formula (1) is:

[0087]

[0088] Among them, I ref represents the in-core inventory of the reference nuclide ref, c k,ref represents the coefficient corresponding to the reference nuclide ref, A f,ref,k represents the amount of the reference nuclide ref produced by the pre-selected fissile nuclide, and k is the number of different nuclides in the decay chain of the nuclide ref.

[0089] If the target object is the nuclide to be measured, the in-core inventory of the nuclide to be measured can be calculated based on formula (1). The coefficient corresponding to the nuclide to be measured is the A of the nuclide k in the decay chain of the nuclide to be measured f,i,k The coefficient of. When the nuclide to be measured is determined, the coefficient corresponding to the nuclide to be measured is also determined and can be obtained through calculation. The amount of the nuclide to be measured produced by the pre-selected fissile nuclide can be calculated based on formula (2).

[0090] For example, the calculation formula for the in-core inventory corresponding to the nuclide to be measured obtained based on formula (1) is:

[0091]

[0092] Among them, I d represents the in-core inventory of the nuclide to be measured d, c k,d represents the coefficient corresponding to the nuclide to be measured d, A f,d,k represents the amount of the nuclide to be measured d produced by the pre-selected fissile nuclide, and k is the number of different nuclides in the decay chain of the nuclide d.

[0093] S302. Obtain the equivalent release rate / generation rate of all the fuel elements in the core according to the release rate / generation rate of the halogen nuclide and the fission gas in the equivalent damaged cladding particles of the core fuel elements and the release rate / generation rate of the uranium contamination;

[0094] Specifically, calculate the release rate / generation rate of the halogen nuclide and the fission gas in the equivalent damaged cladding particles of the core fuel elements and the sum of the halogen nuclide and the fission gas in the equivalent uranium contamination of the core fuel elements as the equivalent release rate / generation rate of all the fuel elements in the core.

[0095] Among them, the release rate / generation rate of the elemental nuclide and fission gas in the equivalent damaged cladding particles of the core fuel element can be obtained by calculating the product of the breakage rate of the coated particles of the high-temperature gas-cooled reactor fuel element and the release rate / generation rate of the elemental nuclide and fission gas in the equivalent sphere of a single damaged cladding particle. The release rate / generation rate of the equivalent uranium contamination of the elemental nuclide and fission gas in the core fuel element can be obtained by calculating the product of the uranium contamination fraction in the matrix graphite of the fuel element and the release rate / generation rate of the equivalent sphere of uranium contamination. The release rate / generation rate of the equivalent sphere of uranium contamination can be calculated based on formula (6).

[0096] S303. Obtain the equivalent removal coefficient of the target object according to the decay coefficient, purification coefficient, deposition rate, and neutron absorption cross-section of the target object, as well as the purification flow rate of the helium purification system, the total volume of helium in the primary circuit, the time for helium to circulate once, the primary circuit leakage removal coefficient, the equivalent neutron flux in the core, the time for helium to flow through the core per cycle, and the equivalent removal coefficient under the current working condition;

[0097] Specifically, the decay coefficient, purification coefficient, deposition rate, and neutron absorption cross-section of the target object are related to the target object and can be obtained by querying relevant materials according to the target object. The purification flow rate of the helium purification system, the total volume of helium in the primary circuit, the time for helium to circulate once, the primary circuit leakage removal coefficient, the equivalent neutron flux in the core, the time for helium to flow through the core per cycle, and the equivalent removal coefficient are related to the current working condition and are determined according to the current working condition. The decay coefficient, purification coefficient, deposition rate, and neutron absorption cross-section of the target object, as well as the purification flow rate of the helium purification system, the total volume of helium in the primary circuit, the time for helium to circulate once, the primary circuit leakage removal coefficient, the equivalent neutron flux in the core, the time for helium to flow through the core per cycle, and the equivalent removal coefficient under the current working condition can be used to obtain the equivalent removal coefficient of the target object.

[0098] For example, when the target object is a reference nuclide, the calculation formula for obtaining the equivalent removal coefficient of the reference nuclide based on formula (11) is:

[0099]

[0100] Among them, τ ref is the equivalent removal coefficient of the reference nuclide, λ ref is the decay constant of the reference nuclide, ε ref is the purification coefficient of the reference nuclide, δ ref is the deposition rate of the reference nuclide, σ a,ref is the spectral-averaged neutron absorption cross-section of the reference nuclide, Q is the purification flow rate of the helium purification system, V is the total volume of helium in the primary circuit, t c represents the time for helium to circulate once, ω is the primary circuit leakage removal coefficient, φ 1is the core equivalent neutron flux, t v is the time for helium to flow through the core per cycle.

[0101] For example, when the target object is the nuclide to be measured, based on formula (11), the calculation formula for the equivalent removal coefficient of the nuclide to be measured is:

[0102]

[0103] where τ d is the equivalent removal coefficient of the nuclide to be measured, λ d is the decay constant of the nuclide to be measured, ε d is the purification coefficient of the nuclide to be measured, δ d is the deposition rate of the nuclide to be measured, σ a,d is the spectrum-averaged neutron absorption cross-section of the nuclide to be measured, Q is the purification flow rate of the helium purification system, V is the total volume of helium in the primary circuit, t c represents the time for one cycle of helium circulation, ω is the primary circuit leakage removal coefficient, φ 1 is the core equivalent neutron flux, t v is the time for helium to flow through the core per cycle.

[0104] S304. Obtain the first activity concentration of the target object under the current working condition according to the core inventory corresponding to the target object, the equivalent release rate / generation rate of all fuel elements in the core, the equivalent removal coefficient of the target object, and the total volume of helium in the primary circuit.

[0105] Specifically, after obtaining the core inventory corresponding to the target object, the equivalent release rate / generation rate of all fuel elements in the core, and the equivalent removal coefficient of the target object, the first activity concentration of the target object under the current working condition can be obtained based on the core inventory corresponding to the target object, the equivalent release rate / generation rate of all fuel elements in the core, the equivalent removal coefficient of the target object, and the total volume of helium in the primary circuit.

[0106] On the basis of the above embodiments, further, the obtaining the first activity concentration of the target object under the current working condition according to the core inventory corresponding to the target object, the equivalent release rate / generation rate of all fuel elements in the core, the equivalent removal coefficient of the target object, and the total volume of helium in the primary circuit includes:

[0107] According to the formula calculate to obtain the first activity concentration A of the target object under the current working condition ref , λ eff represents the equivalent decay constant, I represents the core inventory corresponding to the target object, Represents the equivalent release rate / generation rate of all fuel elements in the core, τ represents the equivalent removal coefficient of the target object, V represents the total volume of helium in the primary circuit, λ eff = λ + σ a φ 1 , λ represents the decay constant of the target object, σ a represents the neutron absorption cross-section of the target object, φ 1 represents the equivalent neutron flux in the core.

[0108] Specifically, by the formula λ eff = λ + σ a φ 1 , after calculating the equivalent decay constant λ eff , substituting the equivalent decay constant λ eff , the core inventory I corresponding to the obtained target object, the equivalent release rate / generation rate of all fuel elements in the core the equivalent removal coefficient τ of the target object into the formula , the first activity concentration A of the target object under the current working condition can be calculated ref .

[0109] Based on the above embodiments, further, obtaining the equivalent removal coefficient of the target object according to the decay coefficient, purification coefficient, deposition rate, and neutron absorption cross-section of the target object, and the purification flow rate of the helium purification system, the total volume of helium in the primary circuit, the time for one cycle of helium circulation, the primary circuit leakage removal coefficient, the equivalent neutron flux in the core, the time for helium to flow through the core per cycle, and the equivalent removal coefficient includes:

[0110] According to the formula calculate the equivalent removal coefficient τ of the target object, λ represents the decay constant of the target object, ε represents the purification coefficient of the target object, Q represents the purification flow of the helium purification system, V represents the total volume of helium in the primary circuit, δ represents the deposition rate of the target object, t c represents the time for one cycle of helium circulation, ω represents the primary circuit leakage removal coefficient, σ a represents the neutron absorption cross-section of the target object, φ 1 represents the equivalent neutron flux in the core, t v represents the time for helium to flow through the core per cycle.

[0111] Based on the above embodiments, further, obtaining the equivalent release rate / generation rate of all fuel elements in the core according to the equivalent release rate / generation rate of damaged cladding particles and uranium pollution release rate / generation rate of halogen nuclides and fission gases in the core fuel elements includes:

[0112] According to the formula Calculate the equivalent release rate / generation rate of all fuel elements in the core F represents the breakage rate of the coated particles of the high-temperature gas-cooled reactor fuel elements, and C represents the uranium contamination fraction in the matrix graphite of the fuel elements represents the release rate / generation rate of halogen nuclides and fission gases in the equivalent sphere of a single broken coated particle represents the release rate / generation rate of halogen nuclides and fission gases in the equivalent sphere of uranium contamination

[0113] Based on the above embodiments, further, obtaining the core inventory corresponding to the target object according to the amount of the target object generated by the pre-selected fissionable nuclide and the coefficient corresponding to the target object includes:

[0114] According to formula I i =∑ k c k A f,i,k Calculate the core inventory of nuclide i, A f,i,k represents the amount of nuclide i generated by the pre-selected fissionable nuclide, and c k represents the A of nuclide k on the decay chain of nuclide i f,i,k coefficient, and k represents the serial numbers of different nuclides on the decay chain of nuclide i

[0115] Based on the above embodiments, further, the reference object is a set of fission gas nuclides. Correspondingly, obtaining the first activity concentration of the set of fission gas nuclides under the current working condition based on the working condition parameters of the current working condition and the calculation parameters of the set of fission gas nuclides includes:

[0116] According to the working condition parameters of the current working condition and the calculation parameters of each nuclide in the set of fission gas nuclides, calculate the first activity concentration of each nuclide in the set of fission gas nuclides under the current working condition;

[0117] Sum up the first activity concentrations of each nuclide in the set of fission gas nuclides under the current working condition to obtain the first activity concentration of the set of fission gas nuclides under the current working condition

[0118] Specifically, the set of fission gas nuclides includes multiple nuclides. For each nuclide in the set of fission gas nuclides, the first activity concentration of each nuclide in the set of fission gas nuclides under the current working condition can be calculated according to the working condition parameters of the current working condition and the calculation parameters of each nuclide. The specific calculation process of the first activity concentration of each nuclide in the set of fission gas nuclides under the current working condition is similar to the calculation process of the first activity concentration of the reference nuclide under the current working condition, and will not be elaborated here

[0119] After calculating the first activity concentration of each nuclide in the set of fission gas nuclides under the current working condition, calculate the sum of the first activity concentrations of each nuclide in the set of fission gas nuclides under the current working condition, and use the summation result as the first activity concentration of the set of fission gas nuclides under the current working condition.

[0120] For example, the first activity concentration of the p-th nuclide in the set of fission gas nuclides under the current working condition is A p , then the first activity concentration of the set of fission gas nuclides under the current working condition where p is a positive integer and p is less than or equal to z, and z is the total number of nuclides included in the set of fission gas nuclides.

[0121] On the basis of the above embodiments, further, obtaining the second activity concentration of the nuclide to be measured under the current working condition according to the first activity concentration of the nuclide to be measured under the current working condition, the first activity concentration of the reference object under the current working condition, and the second activity concentration of the reference object under the current working condition includes:

[0122] According to the formula calculate to obtain the second activity concentration a of the nuclide to be measured under the current working condition r,j , where A j represents the first activity concentration of the nuclide to be measured under the current working condition, A ref represents the first activity concentration of the reference object under the current working condition, and a ref represents the second activity concentration of the reference object under the current working condition.

[0123] Specifically, after obtaining the first activity concentration A of the nuclide to be measured under the current working condition j , the first activity concentration A of the reference object under the current working condition ref and the second activity concentration a of the reference object under the current working condition ref , through the formula the second activity concentration a of the nuclide to be measured under the current working condition can be calculated. r,j . The formula shows that by experimentally measuring the activity concentration value of the primary loop of nuclides with a shorter half-life or the total activity concentration value of the important fission gases in the primary loop, based on the proportional method, a more accurate activity concentration of the primary loop of nuclides with a longer half-life and halogen nuclides can be obtained.

[0124] When measuring the second activity concentration of a reference object under the current working condition through a primary loop activity concentration experiment, if the equipment used has a relatively high precision and can obtain the second activity concentration of the reference nuclide under the current working condition, the first activity concentration of the reference nuclide under the current working condition and the second activity concentration of the reference nuclide under the current working condition are used to calculate the second activity concentration of the nuclide to be measured under the current working condition. If the equipment used cannot obtain the first activity concentration of the reference nuclide under the current working condition, then the first activity concentration of the fission gas nuclide set under the current working condition and the second activity concentration of the fission gas nuclide set under the current working condition are used to calculate the second activity concentration of the nuclide to be measured under the current working condition.

[0125] Figure 4 is a schematic flow chart of the method for measuring the primary loop activity concentration provided by the fourth embodiment of the present invention. As Figure 4 described, on the basis of the above embodiments, further, obtaining the second activity concentration of the reference object through a primary loop activity concentration experiment measurement includes:

[0126] S401. Extract coolant from the primary loop under the current working condition and measure to obtain a nuclide spectrogram; wherein, the nuclides corresponding to the nuclide spectrogram include the reference object;

[0127] Specifically, for the primary loop activity concentration experiment measurement, sampling type on-line monitoring is adopted. The primary loop helium coolant containing radioactive nuclides enters the sampling tank through the intake pipeline of the sampling system after passing through the copper oxide bed. The primary loop helium coolant after sampling measurement can be discharged after being processed by the ventilation system through the outlet pipeline. After operating stably at a set power for a period of time, coolant is extracted from the primary loop to measure and obtain a nuclide spectrogram. Among them, the nuclides corresponding to the nuclide spectrogram include the reference object. The set power is selected according to actual experience, and the embodiments of the present invention do not make any limitations. The spectrogram can be a γ spectrogram, an x spectrogram, etc., and is selected according to actual needs, and the embodiments of the present invention do not make any limitations.

[0128] S402. Analyze the nuclide spectrogram to obtain the activity concentration of the reference object in the sample tank;

[0129] Specifically, analyze the primary loop nuclide spectrogram, identify the reference object in the sample tank, and then calculate the activity concentration of the reference object in the sample tank.

[0130] When the reference object is a reference nuclide, the analysis of the nuclide γ spectrum includes energy calibration, efficiency calibration, and nuclide identification of the nuclide γ spectrum. The γ rays of each nuclide have specific energies, and the energies corresponding to each nuclide can be used as a basis for distinguishing nuclides. When performing nuclide energy calibration, the peak energy identified in the nuclide γ spectrum is compared with the peak energy data corresponding to the known nuclides in the nuclide database to distinguish nuclides. Among them, the nuclide database is obtained in advance and includes the peak energy data corresponding to the known nuclides. For example, for the nuclide database, the built-in nuclear database in the professional software for γ spectrum analysis or the IAEA nuclear database can be used.

[0131] The peak energy data corresponding to the known nuclides include the value range of the peak energy. If the peak energy identified in the nuclide γ spectrum belongs to the value range of the peak energy corresponding to a certain known nuclide, it can be determined the nuclide corresponding to the peak energy identified in the nuclide γ spectrum. Since there may be a situation where the peak energy differences corresponding to multiple different nuclides in the nuclide γ spectrum are small and belong to the value range of the peak energy corresponding to the same known nuclide, it is necessary to compare the peak energy of the γ spectrum in the full spectrum energy range, considering the nuclides that may exist in the sampling tank, for further identification and judgment. When performing nuclide energy calibration, 88 Kr can be selected as the nuclide for energy calibration because 88 the content of 88 Kr in the primary coolant is usually large enough;

[0132] ref

[0133]

[0134] where a ref is the activity concentration of the reference nuclide in the sampling tank, C net,ref is the net peak area of the spectral peak of the reference nuclide, which is equal to the total peak area of the spectral peak minus the background, I γ is the emission probability of the energy peak of the reference nuclide, t is the time for measuring the spectral peak of the sample, V is the sample volume, ∈ represents the detection efficiency of the detector, and f 1 is the measurement time correction factor caused by the self-decay of the detected radionuclide during the detector measurement process. The detection efficiency ∈ can be obtained by performing efficiency calibration on the nuclide γ spectrum.

[0135] The measurement time correction factor f 1 can be calculated according to the formula and λ ref represents the decay constant of the reference nuclide.

[0136] Since there may be multiple energy peaks for the reference nuclide, in order to increase the credibility, the following methods are adopted: 1) Only the net peak area with an uncertainty lower than the threshold will be used to calculate the activity concentration of the reference nuclide; 2) The activity concentration of the reference nuclide is the weighted average of the activity concentrations corresponding to the energy peaks that meet 1), and the weight corresponding to the energy peak is obtained by normalizing the reciprocal of the square of the uncertainty of calculating the activity concentration for each energy peak. The threshold is set according to actual experience and is not limited in the embodiments of the present invention. For example, the threshold is taken as 12%.

[0137] Therefore, change the formula to where represents the activity concentration of the β-th energy peak of the reference nuclide, w β represents the corresponding weight, represents the net peak area of the spectral peak of the β-th energy peak of the reference nuclide, represents the emission probability of the β-th energy peak of the reference nuclide, ∈ represents the detection efficiency of the detector, t is the time for measuring the spectral peak of the sample, V is the sample volume, f 1 is the measurement time correction factor caused by the self-decay of the detected radionuclide during the detector measurement process, and m represents the total number of energy peaks corresponding to the reference nuclide.

[0138] When the reference object is a collection of fission gas nuclides, the analysis of the γ spectrum of the collection of fission gas nuclides includes energy calibration and efficiency calibration of the γ spectrum of the collection of fission gas nuclides. The γ rays of the collection of fission gas nuclides have specific energies. When performing nuclide energy calibration, the peak energies identified from the spectral diagram of the collection of fission gas nuclides will be used.

[0139] When performing nuclide energy calibration, 88 Kr can be selected as the nuclide for energy calibration because 88 the content of 88 Kr in the primary coolant is usually large enough;

[0140] The activity concentration of the collection of fission gas nuclides in the sample tank is calculated from the net peak area of the spectral peak:

[0141]

[0142] where a c is the activity concentration of the collection of fission gas nuclides in the sampling tank, C net,cis the net peak area of the collective spectral peak of fission gas nuclides, which is equal to the total peak area of the spectral peak minus the background. t is the time for measuring the spectral peak of the sample, V is the volume of the sample, and ∈ represents the detection efficiency of the detector. The detection efficiency ∈ can be obtained by calibrating the efficiency of the collective γ spectrum of fission gas nuclides.

[0143] Since there may be multiple energy peaks in the collective fission gas nuclides, in order to increase the credibility, the following methods are adopted: 1) Only the net peak areas with uncertainties lower than the threshold are used to calculate the activity concentration of the collective fission gas nuclides; 2) The activity concentration of the collective fission gas nuclides is the weighted average of the activity concentrations corresponding to the energy peaks that meet 1), and the weight corresponding to the energy peak is normalized by the reciprocal of the square of the uncertainty of calculating the activity concentration of each energy peak. The threshold is set according to actual experience and is not limited in the embodiments of the present invention. For example, the threshold is taken as 12%.

[0144] Therefore, the formula is changed to where represents the activity concentration of the θ-th energy peak of the collective fission gas nuclides, w θ represents the corresponding weight, represents the net peak area of the spectral peak of the θ-th energy peak of the collective fission gas nuclides, ∈ represents the detection efficiency of the detector, t is the time for measuring the spectral peak of the sample, V is the volume of the sample, and n represents the total number of energy peaks corresponding to the collective fission gas nuclides.

[0145] S403. Based on the sampling time, the density ratio of helium in the primary loop and the sampling tank, correct the activity concentration of the reference object in the sample tank to obtain the second activity concentration of the reference object under the current working condition.

[0146] Specifically, when the helium coolant enters the sampling tank from the primary loop, it needs to pass through a section of the primary loop system pipeline, including some pipelines and equipment of the helium purification system and the gas sampling pipeline, which takes a certain amount of time. It also takes a certain amount of time to fill the sample tank with the sampled gas. The above two parts of time are defined as the sampling time. The sampling time, the density of helium in the primary loop and the sample tank will all affect the activity concentration of the reference object in the sample tank. The activity concentration of the reference object in the sample tank can be corrected by the density ratio of the sampling time, the primary loop and the helium in the sampling tank to obtain the second activity concentration of the reference object under the current working condition. The sampling time is set according to actual needs and is not limited in the embodiments of the present invention.

[0147] The activity concentration of the reference object in the sample tank can be corrected based on the formula a pc = a·f 2 ·f 3 to obtain the first activity concentration a of the reference object under the current working conditionpc , where f 2 represents the sampling time correction coefficient, and f 3 represents the density correction coefficient. λ represents the decay constant of the reference object, and t s represents the sampling time, and ρ pc (P, T) represents the average density of helium gas in the primary loop under the conditions of pressure P and temperature T, and ρ s (P, T) represents the average density of helium gas in the sampling tank under the conditions of pressure P and temperature T.

[0148] The following uses a specific embodiment to illustrate the application of the measurement method for the activity concentration in the primary loop of the high-temperature gas-cooled reactor provided by the embodiments of the present invention.

[0149] For the high-temperature gas-cooled reactor HTR-10, the activity concentration in the primary loop is measured under three operating powers of 3 MW, 7 MW, and 10 MW. The fission gas nuclides in the primary loop of HTR-10 are divided into 4 groups according to the half-life from large to small: Group 1, Group 2, Group 3, and Group 4, and the dividing lines are selected as 10, 1, and 0.1 hours. Among them, the nuclides in Group 1 are 133 Xe and 133m Xe, with half-lives of 5.25 days and 2.19 days respectively; the nuclides in Group 2 are 85m Kr, 87 Kr, 88 Kr, and 135 Xe, with half-lives of 4.48, 1.27, 2.84, and 9.11 hours respectively; the nuclides in Group 3 are 135m Xe and 138 Xe, with half-lives of 15.36 and 14.13 minutes respectively; the nuclides in Group 4 are 89 Kr and 137 Xe, with half-lives of 3.16 and 3.83 minutes respectively.

[0150] Figure 4 Shows the experimental measurement values of the activity concentrations of fission gas nuclides in the primary loop of HTR-10 at different power steps of 3, 7, and 10 MW. The activity concentrations of all nuclides are unified to the average activity concentration under standard conditions (temperature 25°C and pressure 1 atm), and the unit is Bq / Nm 3 . The standard volume of helium gas in the primary loop of HTR-10 is taken as 1301 Nm 3 . The results show that at 3, 7, and 10 MW, the experimental values of the total activity concentration of nuclides in the primary loop of HTR-10 are 3.93×10 6 , 1.47×10 7 , and 2.01×10 7 Bq / Nm 3, the experimental values of the activity concentrations of all fission gas nuclides are distributed between 3.65×10 4 and 4.82×10 6 Bq / Nm 3 . Since the half-lives of Kr-89 and Xe-137 are the shortest (3.83 and 3.16 minutes respectively), the data of these two nuclides are Figure 5 excluded to avoid large uncertainties. In addition, except that the experimental values of the primary loop activity concentrations measured for the nuclides in Group 1 are independent of the operating power, the experimental values of the primary loop activity concentrations measured for the nuclides in Groups 2, 3, and 4 are closely related to the operating power. The main reason is that the nuclides in Group 1 did not reach equilibrium during the experiment.

[0151] This application selects the experimental parameters of HTR-10 at operating powers of 3, 7, and 10 MW and the corresponding reactor design parameters, and calculates the activity concentrations of the primary loop gas fission nuclides in the equilibrium state under different operating conditions of HTR-10, as Figure 6 shown.

[0152] Figure 7 shows the equilibrium diagram of the primary loop gas fission nuclides of HTR-10 at 10 MW power. The corresponding situations at other operating powers (3 MW and 7 MW) are very similar to Figure 7 . In the equilibrium state, Figure 7 the total percentage column of the nuclides in it corresponds to the total amount of them entering the primary loop from the fuel elements, and the percentage columns of different colors correspond to the amounts of the primary loop nuclides under different removal paths. For gas fission nuclides, decay and purification are the most important removal methods, and the larger the decay constant (the smaller the half-life), the larger the proportion removed by decay and the smaller the proportion removed by purification. In the embodiments of the present invention, when HTR-10 has been operating for a preset duration, it is considered that HTR-10 is in an equilibrium state. The preset duration is selected according to actual experience and is not limited in the embodiments of the present invention. Usually, it is more than three times the half-life of the nuclide of concern.

[0153] Table 1 shows the ratios of the theoretical calculations of the activity concentrations of different radionuclides in Figure 8 to the reference nuclide 88 Kr in the primary loop and the experimental values of the activity concentrations of Figure 5 88 Kr in the primary loop, and the activity concentrations of short-lived fission gases determined by the proportional method are calculated through formula (9). The results show that at 3, 7, and 10 MW, the experimental values of the total activity concentrations of the primary loop nuclides of HTR-10 are 1.33×10 6 , 7.01×10 6 and 1.15×10 7 Bq / Nm 3 , and the experimental values of the activity concentrations of all fission gas nuclides are distributed between 6.58×10 3 ​to 1.96×10 6 Bq / Nm 3 between.

[0154] Table 1 Activity Concentration of Short-lived Fission Gases in the Primary Circuit

[0155]

[0156] Due to the obvious adsorption effect, a large amount of iodine is adsorbed on the core graphite, the cold section pipes of the steam generator and the walls of equipment, making it impossible to take samples for measurement. Therefore, the ratio method can be used as a radiation monitoring scheme to determine the activity of iodine nuclides in high-temperature gas-cooled reactors under operating conditions. 133 The theoretical ratios of I to the primary circuit activity concentration of Kr at 3MW, 7MW and 10MW are 0.027, 0.012 and 0.008 respectively. The corresponding primary circuit activity concentrations calculated by the ratio method are 5.21×10 88 、1.02×10 3 、1.06×10 4 and 1.06×10 4 Bq / Nm 3 ; 135 The theoretical ratios of I to the primary circuit activity concentration of Kr at 3MW, 7MW and 10MW are 0.047, 0.021 and 0.014 respectively. The corresponding primary circuit activity concentrations calculated by the ratio method are 9.07×10 88 、1.80×10 3 、1.86×10 4 and 1.86×10 4 Bq / Nm 3 . The measurement method for the primary circuit activity concentration of high-temperature gas-cooled reactors provided by the embodiments of the present invention is abbreviated as the ratio method.

[0157] This application constructs a theoretical model for the diffusion and release of gaseous fission products from the core fuel elements to the primary circuit, so as to accurately obtain the activity concentrations of gaseous fission products and halogen nuclides in the primary circuit. In the calculation of the core inventory, the present invention establishes the IPRFGN model. This model can quickly calculate the inventory of short-lived iodine and fission gas behavior characteristics, facilitating the accurate and rapid determination of the true core inventory during radiation monitoring. Combining the nuclide diffusion and release from the core fuel elements and the primary circuit nuclide migration model, F and C are determined through experiments to calculate the activity concentration ratios of each nuclide in the primary circuit to the reference nuclide. Finally, the activity concentrations of gaseous fission products and halogen nuclides in the primary circuit are determined by the ratio method. The present invention improves the radioactive measurement theory of fission gas nuclides in the primary circuit, and at the same time provides a scheme for the experimental measurement of iodine, providing important guidance for the radiation safety and protection work of high-temperature gas-cooled reactors.

[0158] Figure 9It is a schematic structural diagram of a measuring device for the activity concentration of the primary circuit of a high-temperature gas-cooled reactor provided by the ninth embodiment of the present invention. As Figure 9 shown, the measuring device for the activity concentration of the primary circuit of a high-temperature gas-cooled reactor provided by the embodiment of the present invention includes an obtaining unit 901, an acquiring unit 902, and a measuring unit 903, where:

[0159] The obtaining unit 901 is configured to obtain a first activity concentration of the reference object under the current working condition based on the working condition parameters of the current working condition and the calculation parameters of the reference object, and obtain a first activity concentration of the nuclide to be measured under the current working condition according to the working condition parameters of the current working condition and the calculation parameters of the nuclide to be measured; the acquiring unit 902 is configured to acquire a second activity concentration of the reference object under the current working condition; wherein, the second activity concentration of the reference object under the current working condition is obtained by measuring through a primary circuit activity concentration experiment; the measuring unit 903 is configured to obtain a second activity concentration of the nuclide to be measured under the current working condition according to the first activity concentration of the nuclide to be measured under the current working condition, the first activity concentration of the reference object under the current working condition, and the second activity concentration of the reference object under the current working condition.

[0160] Specifically, based on the constructed theoretical framework and calculation method for the activity concentration of radionuclides in the primary circuit of a high-temperature gas-cooled reactor, the obtaining unit 901 can obtain the theoretical activity concentration of the reference object under the current working condition based on the working condition parameters of the current working condition and the calculation parameters of the reference object, as the first activity concentration of the reference object under the current working condition. And based on the working condition parameters of the current working condition and the calculation parameters of the nuclide to be measured, obtain the theoretical activity concentration of the nuclide to be measured under the current working condition, as the first activity concentration of the nuclide to be measured under the current working condition.

[0161] The acquiring unit 902 can measure and obtain the first activity concentration of the reference object under different working conditions through a primary circuit activity concentration experiment. In application, the second activity concentration of the reference object under the current working condition can be obtained by querying according to the current working condition.

[0162] A calculation relationship can be established in advance among the first activity concentration of the nuclide to be measured under the current working condition, the first activity concentration of the reference object under the current working condition, the second activity concentration of the reference object under the current working condition, and the second activity concentration of the nuclide to be measured under the current working condition. After the measuring unit 903 obtains the specific values of the first activity concentration of the nuclide to be measured under the current working condition, the first activity concentration of the reference object under the current working condition, and the second activity concentration of the reference object under the current working condition, in combination with the above calculation relationship, obtain the second activity concentration of the nuclide to be measured under the current working condition.

[0163] The measurement device for the primary circuit activity concentration of a high-temperature gas-cooled reactor provided by an embodiment of the present invention can obtain the first activity concentration of a reference object under the current working condition based on the working condition parameters of the current working condition and the calculation parameters of the reference object, and obtain the first activity concentration of the nuclide to be measured under the current working condition according to the working condition parameters of the current working condition and the calculation parameters of the nuclide to be measured; obtain the second activity concentration of the reference object under the current working condition; wherein, the second activity concentration of the reference object under the current working condition is obtained through experimental measurement of the primary circuit activity concentration; according to the first activity concentration of the nuclide to be measured under the current working condition, the first activity concentration of the reference object under the current working condition, and the second activity concentration of the reference object under the current working condition, obtain the second activity concentration of the nuclide to be measured under the current working condition, which can accurately measure the activity concentration of radioactive nuclides with a short half-life, improve the accuracy of the measurement of the primary circuit nuclide activity concentration, and can realize the measurement of the activity concentration of halogen nuclides, improving the comprehensiveness of the measurement of the primary circuit nuclide activity concentration.

[0164] Figure 10 FIG. is a schematic structural diagram of the measurement device for the primary circuit activity concentration of a high-temperature gas-cooled reactor provided by the tenth embodiment of the present invention, as Figure 10 shown. On the basis of the above embodiments, further, the obtaining unit 901 includes a first obtaining subunit 9011, a second obtaining subunit 9012, a third obtaining subunit 9013, and a fourth obtaining subunit 9014, wherein:

[0165] The first obtaining subunit 9011 is configured to obtain the core inventory corresponding to the target object according to the amount of the target object generated by a preselected fissionable nuclide and the coefficient corresponding to the target object; wherein, the target object is a reference nuclide or a nuclide to be measured; the second obtaining subunit 9012 is configured to obtain the equivalent release rate / generation rate of all fuel elements in the core according to the release rate / generation rate of halogen nuclides and fission gases in the equivalent damaged cladding particles of the core fuel elements and the uranium pollution release rate / generation rate; the third obtaining subunit 9013 is configured to obtain the equivalent removal coefficient of the target object according to the decay coefficient, purification coefficient, deposition rate, and neutron absorption cross-section of the target object, and the purification flow rate of the helium purification system, the total volume of helium in the primary circuit, the time for helium to circulate once, the primary circuit leakage removal coefficient, the equivalent neutron flux of the core, the time for helium to flow through the core per cycle, and the equivalent removal coefficient under the current working condition; the fourth obtaining subunit 9014 is configured to obtain the first activity concentration of the target object under the current working condition according to the core inventory corresponding to the target object, the equivalent release rate / generation rate of all fuel elements in the core, the equivalent removal coefficient of the target object, and the total volume of helium in the primary circuit.

[0166] Based on the above embodiments, further, the fourth obtaining subunit 9014 is specifically configured to:

[0167] Calculate the first activity concentration A of the target object under the current working condition according to the formula where λ ref eff represents the equivalent decay constant, I represents the core inventory corresponding to the target object, represents the equivalent release rate / generation rate of all fuel elements in the core, τ represents the equivalent removal coefficient of the target object, V represents the total volume of helium in the primary loop, and λ eff = λ + σ a φ 1 where λ represents the decay constant of the target object, σ a represents the neutron absorption cross-section of the target object, and φ 1 represents the equivalent neutron flux in the core.

[0168] Based on the above embodiments, further, the third obtaining subunit 9013 is specifically configured to:

[0169] Calculate the equivalent removal coefficient τ of the target object according to the formula where λ represents the decay constant of the target object, ε represents the purification coefficient of the target object, Q represents the purification flow of the helium purification system, V represents the total volume of helium in the primary loop, δ represents the deposition rate of the target object, t c represents the time for one cycle of helium circulation, ω represents the primary loop leakage removal coefficient, σ a represents the neutron absorption cross-section of the target object, φ 1 represents the equivalent neutron flux in the core, and t v represents the time for helium to flow through the core in each cycle.

[0170] Based on the above embodiments, further, the second obtaining subunit 9012 is specifically configured to:

[0171] Calculate the equivalent release rate / generation rate of all fuel elements in the core according to the formula where F represents the breakage rate of the fuel element coating particles in the high-temperature gas-cooled reactor, C represents the uranium contamination fraction in the matrix graphite of the fuel element, represents the release rate / generation rate of halogen nuclides and fission gases in the equivalent sphere of a single broken coating particle, represents the release rate / generation rate of halogen nuclides and fission gases in the equivalent sphere of uranium contamination.

[0172] Based on the above embodiments, further, the first obtaining subunit 9011 is specifically configured to:

[0173] Calculate I according to the formula​​i = ∑ k c k A f,i,k Calculate the in-core inventory of nuclide i, A f,i,k represents the amount of nuclide i produced by a pre-selected fissile nuclide, c k represents A of nuclide k on the decay chain of nuclide i f,i,k coefficient, k represents the number of different nuclides on the decay chain of nuclide i

[0174] Based on the above embodiments, further, the reference object is a set of fission gas nuclides. Correspondingly, the obtaining unit 901 is specifically configured to:

[0175] According to the working condition parameters of the current working condition and the calculation parameters of each nuclide in the set of fission gas nuclides, calculate the first activity concentration corresponding to each nuclide in the set of fission gas nuclides; sum the first activity concentrations corresponding to each nuclide in the set of fission gas nuclides to obtain the first activity concentration of the set of fission gas nuclides under the current working condition

[0176] Based on the above embodiments, further, the measuring unit 903 is specifically configured to:

[0177] According to the formula Calculate to obtain the second activity concentration a of the nuclide to be measured under the current working condition r,j , where A j represents the first activity concentration of the nuclide to be measured under the current working condition, A ref represents the first activity concentration of the reference object under the current working condition, a ref represents the second activity concentration of the reference object under the current working condition

[0178] Figure 11 is a schematic structural diagram of a measuring device for the activity concentration of the primary circuit of a high-temperature gas-cooled reactor provided by the eleventh embodiment of the present invention. As Figure 11 shown, based on the above embodiments, further, the measuring device for the activity concentration of the primary circuit of a high-temperature gas-cooled reactor provided by the embodiments of the present invention further includes a spectrum unit 904, an activity concentration obtaining unit 905, and a correction unit 906, where:

[0179] The spectrum unit 904 is used to extract coolant from the primary loop under the current working condition and measure the nuclide spectrogram; wherein, the nuclides corresponding to the nuclide spectrogram include the reference object; the activity concentration obtaining unit 905 is used to analyze the nuclide spectrogram to obtain the activity concentration of the reference object in the sample tank; the correction unit 906 is used to correct the activity concentration of the reference object in the sample tank based on the sampling time, the density ratio of helium in the primary loop and the sampling tank, so as to obtain the first activity concentration of the reference object under the current working condition.

[0180] The embodiments of the device provided by the embodiments of the present invention can be specifically used to execute the processing flows of the above method embodiments, and their functions will not be elaborated here. For details, reference can be made to the detailed descriptions of the above method embodiments.

[0181] Figure 12 It is a schematic physical structure diagram of a computer device provided by the twelfth embodiment of the present invention. As Figure 12 shown, the computer device may include: a processor 1201, a communication interface 1202, a memory 1203, and a communication bus 1204. Among them, the processor 1201, the communication interface 1202, and the memory 1203 communicate with each other through the communication bus 1204. The processor 1201 can call the logical instructions in the memory 1203 to execute the methods provided by the above method embodiments, for example, including: obtaining the first activity concentration of the reference object under the current working condition based on the working condition parameters of the current working condition and the calculation parameters of the reference object, and obtaining the first activity concentration of the nuclide to be measured under the current working condition according to the working condition parameters of the current working condition and the calculation parameters of the nuclide to be measured; obtaining the second activity concentration of the reference object under the current working condition; wherein, the second activity concentration of the reference object under the current working condition is obtained by measuring through a primary loop activity concentration experiment; obtaining the second activity concentration of the nuclide to be measured under the current working condition according to the first activity concentration of the nuclide to be measured under the current working condition, the first activity concentration of the reference object under the current working condition, and the second activity concentration of the reference object under the current working condition.

[0182] In addition, when the logical instructions in the above-mentioned memory 1203 are implemented in the form of software functional units and sold or used as independent products, they can be stored in a computer-readable storage medium. Based on such an understanding, the technical solution of the present invention, in essence, or the part that contributes to the prior art, or a part of the technical solution, can be embodied in the form of a software product. The computer software product is stored in a storage medium and includes several instructions for causing a computer device (which may be a personal computer, a server, or a network device, etc.) to execute all or part of the steps of the methods described in the various embodiments of the present invention. The aforementioned storage medium includes: various media that can store program codes, such as USB flash drives, mobile hard disks, read-only memories (ROM, Read-Only Memory), random access memories (RAM, Random Access Memory), magnetic disks, or optical discs.

[0183] This embodiment discloses a computer program product. The computer program product includes computer programs / instructions. When the computer programs / instructions are executed by a processor, they can execute the methods provided in the above-mentioned method embodiments. For example, it includes: obtaining a first activity concentration of the reference object under the current working condition based on the working condition parameters of the current working condition and the calculation parameters of the reference object, and obtaining a first activity concentration of the nuclide to be measured under the current working condition according to the working condition parameters of the current working condition and the calculation parameters of the nuclide to be measured; obtaining a second activity concentration of the reference object under the current working condition; wherein, the second activity concentration of the reference object under the current working condition is obtained by measuring through a primary loop activity concentration experiment; obtaining a second activity concentration of the nuclide to be measured under the current working condition according to the first activity concentration of the nuclide to be measured under the current working condition, the first activity concentration of the reference object under the current working condition, and the second activity concentration of the reference object under the current working condition.

[0184] This embodiment provides a computer-readable storage medium. The computer-readable storage medium stores computer programs / instructions. When the computer programs / instructions are executed by a processor, the computer is caused to execute the methods provided in the above-described method embodiments. For example, it includes: obtaining a first activity concentration of the reference object under the current working condition based on the working condition parameters of the current working condition and the calculation parameters of the reference object, and obtaining a first activity concentration of the nuclide to be measured under the current working condition according to the working condition parameters of the current working condition and the calculation parameters of the nuclide to be measured; obtaining a second activity concentration of the reference object under the current working condition; wherein the second activity concentration of the reference object under the current working condition is obtained through a primary loop activity concentration experiment measurement; obtaining a second activity concentration of the nuclide to be measured under the current working condition according to the first activity concentration of the nuclide to be measured under the current working condition, the first activity concentration of the reference object under the current working condition, and the second activity concentration of the reference object under the current working condition.

[0185] Those skilled in the art should understand that the embodiments of the present invention can be provided as methods, systems, or computer program products. Therefore, the present invention can take the form of a completely hardware embodiment, a completely software embodiment, or an embodiment combining software and hardware aspects. Moreover, the present invention can take the form of a computer program product implemented on one or more computer-usable storage media (including but not limited to disk storage, CD-ROM, optical storage, etc.) that contain computer-usable program code.

[0186] The present invention is described with reference to the flowcharts and / or block diagrams of methods, apparatuses (systems), and computer program products according to embodiments of the present invention. It should be understood that each flow and / or block in the flowchart and / or block diagram can be implemented by computer program instructions, and the combination of flows and / or blocks in the flowchart and / or block diagram can also be implemented by computer program instructions. These computer program instructions can be provided to the processor of a general-purpose computer, a special-purpose computer, an embedded processor, or other programmable data processing devices to generate a machine, so that the instructions executed by the processor of the computer or other programmable data processing devices generate means for implementing the functions specified in Figure 1 one flow or multiple flows and / or blocks Figure 1 one block or multiple blocks.

[0187] These computer program instructions can also be stored in a computer-readable memory that can direct a computer or other programmable data processing device to work in a specific manner, so that the instructions stored in the computer-readable memory generate a manufactured article including instruction means, and the instruction means implements the functions specified in Figure 1 one flow or multiple flows and / or blocks Figure 1 one block or multiple blocks.

[0188] These computer program instructions can also be loaded onto a computer or other programmable data processing apparatus, so that a series of operation steps are performed on the computer or other programmable apparatus to produce a computer-implemented process, thereby providing instructions for implementing the functions specified in one process or a plurality of processes and / or boxes Figure 1 One process or a plurality of processes and / or boxes Figure 1 Steps for implementing the functions specified in one box or a plurality of boxes

[0189] In the description of this specification, descriptions with reference to terms such as "one embodiment", "a specific embodiment", "some embodiments", "for example", "example", "specific example", or "some examples", etc. mean that the specific features, structures, materials, or characteristics described in connection with the embodiment or example are included in at least one embodiment or example of the present invention. In this specification, the schematic representations of the above terms do not necessarily refer to the same embodiment or example. Moreover, the specific features, structures, materials, or characteristics described can be combined in any one or more embodiments or examples in a suitable manner

[0190] The specific embodiments described above have further elaborated on the objectives, technical solutions, and beneficial effects of the present invention. It should be understood that the above are only specific embodiments of the present invention and are not used to limit the protection scope of the present invention. Any modifications, equivalent replacements, improvements, etc. made within the spirit and principles of the present invention shall be included in the protection scope of the present invention

Claims

1. A method for measuring the activity concentration of a high temperature gas-cooled reactor primary loop, characterized in that: include: Based on the working condition parameters of the current working condition and the calculated parameters of the reference object, a first activity concentration of the reference object under the current working condition is obtained, and according to the working condition parameters of the current working condition and the calculated parameters of the nuclide to be measured, a first activity concentration of the nuclide to be measured under the current working condition is obtained; Acquiring a second activity concentration of the reference object under the current working condition; wherein the second activity concentration of the reference object under the current working condition is obtained by measuring a primary loop activity concentration experiment; The second activity concentration of the nuclide to be measured under the current working condition is obtained according to the first activity concentration of the nuclide to be measured under the current working condition, the first activity concentration of the reference object under the current working condition and the second activity concentration of the reference object under the current working condition.

2. The method according to claim 1, characterized in that Based on the operating condition parameters of the current operating condition and the calculation parameters of the target object, obtaining the first activity concentration of the target object under the current operating condition includes: According to the amount of the target object produced by the pre-selected fissile nuclide and the coefficient corresponding to the target object, the core inventory corresponding to the target object is obtained; wherein the target object is a reference nuclide or a nuclide to be measured; According to the equivalent release rate / generation rate of damaged coated particles and uranium contamination of halogen nuclides and fission gases in the core fuel elements, the equivalent release rate / generation rate of all fuel elements in the core is obtained; According to the decay coefficient, purification coefficient, deposition rate and neutron absorption cross section of the target object, as well as the purification flow rate of the helium purification system under the current working condition, the total volume of helium in the primary circuit, the time for helium to circulate for one cycle, the primary circuit leakage removal coefficient, the core equivalent neutron flux, the time for helium to flow through the core per cycle and the equivalent removal coefficient, the equivalent removal coefficient of the target object is obtained; According to the core inventory corresponding to the target object, the equivalent release rate / generation rate of all fuel elements in the core, the equivalent removal coefficient of the target object and the total volume of helium in a primary loop, the first activity concentration of the target object under the current operating conditions is obtained.

3. The method according to claim 2, characterized in that The obtaining, according to the core inventory corresponding to the target object, the equivalent release rate / generation rate of all fuel elements in the core, the equivalent removal coefficient of the target object, and the total volume of primary helium, of the first activity concentration of the target object under the current operating condition comprises: According to the formula Calculate and obtain the first activity concentration A of the target object under the current working conditions ref ,λ eff represents the equivalent decay constant, I represents the core inventory corresponding to the target object, represents the equivalent release rate / generation rate of all fuel elements in the core, τ represents the equivalent removal coefficient of the target object, V represents the total volume of helium in the first loop, and λ eff =λ+σ a φ1,λ represents the decay constant of the target object,σ a represents the neutron absorption cross section of the target object, and φ1 represents the core equivalent neutron flux.

4. The method according to claim 2, characterized in that: The obtaining of the equivalent removal coefficient of the target object according to the decay coefficient, purification coefficient, deposition rate and neutron absorption cross section of the target object, as well as the purification flow rate of the helium purification system under the current working condition, the total volume of helium in the primary circuit, the time for helium to circulate for one cycle, the primary circuit leakage removal coefficient, the core equivalent neutron flux, the time for helium to flow through the core per cycle and the equivalent removal coefficient comprises: According to the formula The equivalent removal coefficient τ of the target object is obtained by calculation, λ represents the decay constant of the target object, ε represents the purification coefficient of the target object, Q represents the purification flow of the helium purification system, V represents the total volume of helium in a loop, δ represents the deposition rate of the target object, t c represents the time for helium to circulate for one cycle, ω represents the leakage removal coefficient of one circuit, σ a represents the neutron absorption cross section of the target object, φ1 represents the core equivalent neutron flux, t v Indicates the time it takes for helium to flow through the core per cycle.

5. The method according to claim 2, characterized in that: The method of obtaining the equivalent release rate / generation rate of all fuel elements in the core according to the equivalent release rate / generation rate of damaged coated particles and the uranium contamination release rate / generation rate of halogen nuclides and fission gases in the core fuel elements includes: According to the formula Calculate the equivalent release rate / generation rate of all fuel elements in the core F represents the damage rate of the coating particles of the high-temperature gas-cooled reactor fuel element, C represents the uranium contamination ratio in the graphite of the fuel element matrix, It represents the release rate / generation rate of halogen nuclides and fission gases in the equivalent sphere of a single damaged coated particle. It represents the release rate / generation rate of halogen nuclides and fission gases in a uranium-contaminated equivalent sphere.

6. The method according to claim 2, characterized in that The obtaining of the core inventory corresponding to the target object according to the amount of the target object generated by the pre-selected fissile nuclides and the coefficient corresponding to the target object comprises: According to formula I i =∑ k c k A f,i,k Calculate the core inventory of nuclide i, A f,i,k represents the amount of nuclide i produced by the pre-selected fissile nuclide, c k A represents the decay chain of nuclide k in nuclide i f,i,k The coefficient of k represents the number of different nuclides in the decay chain of nuclide i.

7. The method according to claim 1, characterized in that The reference object is a fission gas nuclide set. Accordingly, based on the operating condition parameters of the current operating condition and the calculation parameters of the fission gas nuclide set, obtaining the first activity concentration of the fission gas nuclide set under the current operating condition includes: Calculating and obtaining a first activity concentration corresponding to each nuclide in the fission gas nuclide set according to the operating condition parameters of the current operating condition and the calculation parameters of each nuclide in the fission gas nuclide set; The first activity concentration corresponding to each nuclide in the fission gas nuclide set is summed to obtain the first activity concentration of the fission gas nuclide set under the current working condition.

8. The method according to claim 1, characterized in that The obtaining, according to the first activity concentration of the nuclide to be measured under the current working condition, the first activity concentration of the reference object under the current working condition, and the second activity concentration of the reference object under the current working condition, the second activity concentration of the nuclide to be measured under the current working condition comprises: According to the formula Calculate and obtain the second activity concentration a of the nuclide to be measured under the current working conditions r,j , where A j represents the first activity concentration of the nuclide to be measured under the current working conditions, A ref represents the first activity concentration of the reference object under the current working condition, a ref Represents the second activity concentration of the reference object under the current operating conditions.

9. The method according to any one of claims 1 to 8, characterized in that: The second activity concentration of the reference object under the current working condition obtained by experimental measurement includes: Under the current working condition, coolant is extracted from a primary circuit, and a nuclide spectrum is measured to obtain the nuclide spectrum; wherein the nuclide corresponding to the nuclide spectrum includes the reference object; Analyzing the nuclide spectrum to obtain the activity concentration of the reference object in the sample can; The activity concentration of the reference object in the sample tank is corrected based on the sampling time, the density ratio of the primary loop and the helium in the sampling tank, and a second activity concentration of the reference object under the current working condition is obtained.

10. A device for measuring the activity concentration of a primary circuit of a high temperature gas-cooled reactor, characterized in that: include: an obtaining unit, configured to obtain a first activity concentration of the reference object under the current working condition based on the working condition parameters of the current working condition and the calculated parameters of the reference object, and to obtain a first activity concentration of the nuclide to be measured under the current working condition according to the working condition parameters of the current working condition and the calculated parameters of the nuclide to be measured; An acquisition unit, used to acquire a second activity concentration of the reference object under the current working condition; wherein the second activity concentration of the reference object under the current working condition is obtained by measuring a primary loop activity concentration experiment; The measuring unit is used to obtain the second activity concentration of the nuclide to be measured under the current working condition according to the first activity concentration of the nuclide to be measured under the current working condition, the first activity concentration of the reference object under the current working condition and the second activity concentration of the reference object under the current working condition.

11. A computer device comprising a memory, a processor and a computer program stored in the memory, characterized in that: The processor executes the computer program to implement the method according to any one of claims 1 to 9.

12. A computer-readable storage medium, characterized in that: The computer-readable storage medium stores a computer program / instruction, and when the computer program / instruction is executed by a processor, the method according to any one of claims 1 to 9 is implemented.

13. A computer program product comprising a computer program / instructions, characterized in that When the computer program / instructions are executed by a processor, the method according to any one of claims 1 to 9 is implemented.

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