Detection method and device for safe operation of pebble bed high temperature gas-cooled reactor

Abstract

The present application proposes a method and device for detecting the safe operation of a pebble bed high-temperature gas-cooled reactor, wherein the method comprises: performing core diffusion calculation on the pebble bed high-temperature gas-cooled reactor based on the nodal method to obtain the multi-group average neutron flux density of each nodal block of the pebble bed core; obtaining the multi-group neutron flux density and multi-group defect factor of each batch of fuel balls, as well as the multi-group neutron flux density and multi-group defect factor of the TRISO fuel particles in each batch of fuel balls; determining the power distribution of the TRISO fuel particles of the pebble bed high-temperature gas-cooled reactor based on the multi-group average neutron flux density of each nodal block of the pebble bed core, the fission energy generation cross section of the TRISO fuel particles, the multi-group defect factor of each batch of fuel balls and the TRISO fuel particles in each batch of fuel balls, and then judging whether the pebble bed high-temperature gas-cooled reactor is operating safely, thereby providing a more reliable data basis for the nuclear design and safety analysis of the pebble bed high-temperature gas-cooled reactor.

Description

Technical Field

[0001] The present application relates to the field of nuclear reactor engineering technology, and in particular to a method and device for detecting the safe operation of a pebble bed high-temperature gas-cooled reactor. Background Art

[0002] Due to its inherent safety and high-temperature multi-purpose properties, pebble-bed high-temperature gas-cooled reactors (HTGRs) have broad application prospects in power generation and heating. The nuclear fuel elements of HTGRs are all-ceramic spherical nuclear fuel elements, also known as fuel spheres. The fuel spheres contain TRISO fuel particles (coated fuel particles) dispersed in a graphite matrix. The in-core power distribution of HTGRs is particularly important in their nuclear design and safety analysis. Therefore, accurately determining the fuel power distribution of HTGRs has become a key research issue in the field of nuclear reactor engineering technology. Summary of the Invention

[0003] The present application aims to solve one of the technical problems in the related art at least to a certain extent.

[0004] To this end, the first aspect of the present application proposes a method for detecting the safe operation of a pebble bed high temperature gas-cooled reactor, comprising:

[0005] The core diffusion of a pebble bed high temperature gas-cooled reactor is calculated based on the three-dimensional cylindrical geometry segment method, and the multi-group average neutron flux density of each segment of the pebble bed core is obtained.

[0006] Obtaining multi-group neutron flux densities of each batch of fuel balls in the node according to multi-group homogenization group constants of the batch of fuel balls;

[0007] Obtaining a multi-group defect factor of each batch of fuel spheres according to the multi-group average neutron flux density of each segment of the pebble bed core and the multi-group neutron flux density of each batch of fuel spheres;

[0008] Obtaining multi-group neutron flux density of the TRISO fuel particles in each batch of fuel spheres according to the multi-group homogenization group constant of the TRISO fuel particles in each batch of fuel spheres;

[0009] Obtaining a multi-group defect factor of TRISO fuel particles in each batch of fuel spheres according to the multi-group neutron flux density of each batch of fuel spheres and the multi-group neutron flux density of TRISO fuel particles in each batch of fuel spheres;

[0010] Determining the power of a single TRISO fuel particle in each segment of the pebble bed core based on the multi-group average neutron flux density of each segment of the pebble bed core, the multi-group defect factor of each batch of fuel spheres, the multi-group defect factor of the TRISO fuel particles in each batch of fuel spheres, and the fission energy production cross section of the TRISO fuel particles in each batch of fuel spheres;

[0011] determining a TRISO fuel particle power distribution of the pebble bed core according to the power of a single TRISO fuel particle of each segment of the pebble bed core;

[0012] Based on the TRISO fuel particle power distribution of the pebble bed core, it is determined whether the pebble bed high temperature gas-cooled reactor is operating safely.

[0013] In some embodiments of the present application, the multi-group homogenization group constants of each batch of fuel balls include: the macroscopic transport cross section of each batch of fuel balls, the macroscopic scattering cross section of each batch of fuel balls, the probability of neutron production by fission of each batch of fuel balls, and the fission neutron production cross section of each batch of fuel balls; wherein, obtaining the multi-group neutron flux density of each batch of fuel balls according to the multi-group homogenization group constants of each batch of fuel balls in the node includes:

[0014] Obtaining the escape probability of each batch of fuel balls according to the macroscopic transport cross-section of each batch of fuel balls and the radius of each batch of fuel balls;

[0015] Obtaining the penetration probability of each batch of fuel balls according to the macroscopic transport cross-section of each batch of fuel balls and the radius of each batch of fuel balls;

[0016] Obtaining an average penetration probability of each batch of fuel balls according to the surface area fractions of each batch of fuel balls and the penetration probability of each batch of fuel balls;

[0017] Obtaining collision probabilities between fuel balls in each batch based on the escape probabilities of the fuel balls in each batch, the penetration probabilities of the fuel balls in each batch, the surface area fractions of the fuel balls in each batch, and the average penetration probabilities of the fuel balls in each batch within the node;

[0018] Establish the first collision probability equations;

[0019] The first collision probability equations are solved based on the collision probability between the batches of fuel balls, the volume fraction of the batches of fuel balls, the macroscopic transport cross-section of the batches of fuel balls, the macroscopic scattering cross-section of the batches of fuel balls, the probability of neutron production by fission of the batches of fuel balls, and the fission neutron production cross-section of the batches of fuel balls to obtain the multi-group neutron flux density of the batches of fuel balls.

[0020] In some embodiments of the present application, the multi-group defect factor of each batch of fuel balls is obtained by the following formula:

[0021]

[0022] Among them, S Pebble,i,g is the defect factor of the g-th energy group of the i-th batch of fuel balls, ΦPebble,i,g is the neutron flux density of the gth energy group of the i-th batch of fuel balls, Φ Nodal,g is the multi-group average neutron flux density of each segment of the pebble bed core.

[0023] In some embodiments of the present application, the multi-group homogenization group constants of the TRISO fuel particles in each batch of fuel spheres include: the macroscopic transport cross section of the TRISO fuel particles in each batch of fuel spheres, the macroscopic scattering cross section of the TRISO fuel particles in each batch of fuel spheres, the probability of neutron production by fission of the TRISO fuel particles in each batch of fuel spheres, and the fission neutron production cross section of the TRISO fuel particles in each batch of fuel spheres; wherein, obtaining the multi-group neutron flux density of the TRISO fuel particles in each batch of fuel spheres based on the multi-group homogenization group constants of the TRISO fuel particles in each batch of fuel spheres includes:

[0024] Obtaining the escape probability of the TRISO fuel particles in each batch of fuel spheres according to the macroscopic transport cross-section of the TRISO fuel particles in each batch of fuel spheres and the radius of the TRISO fuel particles in each batch of fuel spheres;

[0025] Obtaining the penetration probability of the TRISO fuel particles in each batch of fuel spheres according to the macroscopic transport cross-section of the TRISO fuel particles in each batch of fuel spheres and the radius of the TRISO fuel particles in each batch of fuel spheres;

[0026] Obtaining an average penetration probability of the TRISO fuel particles in each batch of fuel spheres according to the surface area fraction of the TRISO fuel particles in each batch of fuel spheres and the penetration probability of the TRISO fuel particles in each batch of fuel spheres;

[0027] Obtaining a collision probability between TRISO fuel particles in each batch of fuel spheres according to the escape probability of TRISO fuel particles in each batch of fuel spheres, the penetration probability of TRISO fuel particles in each batch of fuel spheres, the surface area fraction of TRISO fuel particles in each batch of fuel spheres, and the average penetration probability of TRISO fuel particles in each batch of fuel spheres;

[0028] Establish the second collision probability equations;

[0029] The second collision probability equations are solved based on the collision probability between TRISO fuel particles in each batch of fuel spheres, the volume fraction of TRISO fuel particles in each batch of fuel spheres, the macroscopic transport cross-section of TRISO fuel particles in each batch of fuel spheres, the macroscopic scattering cross-section of TRISO fuel particles in each batch of fuel spheres, the probability of neutron production by fission of TRISO fuel particles in each batch of fuel spheres, and the fission neutron production cross-section of TRISO fuel particles in each batch of fuel spheres to obtain the multi-group neutron flux density of TRISO fuel particles in each batch of fuel spheres.

[0030] In some embodiments of the present application, the multi-group defect factor of the TRISO fuel particles in each batch of fuel spheres is obtained by the following formula:

[0031]

[0032] Among them, S TRISO,i,g is the defect factor of the g-th energy group of TRISO fuel particles in the i-th batch of fuel balls, Φ TRISO,i,g is the neutron flux density of the g-th energy group of the TRISO fuel particles in the i-th batch of fuel balls, Φ Pebble,i,g is the neutron flux density of the gth energy group of the i-th batch of fuel balls.

[0033] In some embodiments of the present application, determining the power of the TRISO fuel particles of the pebble bed high temperature gas-cooled reactor according to the multi-group average neutron flux density of each segment of the pebble bed core, the multi-group defect factor of each batch of fuel spheres, the multi-group defect factor of the TRISO fuel particles in each batch of fuel spheres, and the fission energy production cross section of the TRISO fuel particles in each batch of fuel spheres includes:

[0034] Obtaining a neutron flux distribution at a fuel position in the pebble bed core according to the multi-group average neutron flux density of each segment of the pebble bed core, the multi-group defect factor of each batch of fuel spheres, and the multi-group defect factor of TRISO fuel particles within each batch of fuel spheres;

[0035] The power of a single TRISO fuel particle in each segment of the pebble bed core is determined based on the neutron flux distribution at the fuel position of the pebble bed core and the fission energy generation cross section of the TRISO fuel particles in each batch of fuel spheres.

[0036] The second aspect of the present application provides a detection device for the safe operation of a pebble bed high temperature gas-cooled reactor, comprising:

[0037] The first acquisition module is used to perform core diffusion calculation on the pebble bed high temperature gas-cooled reactor based on the three-dimensional cylindrical geometry block method to obtain the multi-group average neutron flux density of each block in the pebble bed core.

[0038] The second acquisition module is used to obtain the multi-group neutron flux density of each batch of fuel balls in the block according to the multi-group homogenization group constants of the fuel balls in the block.

[0039] In some embodiments of the present application, the multi-group homogenization group constants of each batch of fuel balls include: the macroscopic transport cross section of each batch of fuel balls, the macroscopic scattering cross section of each batch of fuel balls, the probability of neutron production by fission of each batch of fuel balls, and the fission neutron production cross section of each batch of fuel balls; wherein the second acquisition module is specifically configured to:

[0040] Obtaining the escape probability of each batch of fuel balls according to the macroscopic transport cross-section of each batch of fuel balls and the radius of each batch of fuel balls;

[0041] Obtaining the penetration probability of each batch of fuel balls according to the macroscopic transport cross-section of each batch of fuel balls and the radius of each batch of fuel balls;

[0042] Obtaining an average penetration probability of each batch of fuel balls according to the surface area fractions of each batch of fuel balls and the penetration probability of each batch of fuel balls;

[0043] Obtaining collision probabilities between fuel balls in each batch based on the escape probabilities of the fuel balls in each batch, the penetration probabilities of the fuel balls in each batch, the surface area fractions of the fuel balls in each batch, and the average penetration probabilities of the fuel balls in each batch within the node;

[0044] Establish the first collision probability equations;

[0045] The first collision probability equations are solved based on the collision probability between the batches of fuel balls, the volume fraction of the batches of fuel balls, the macroscopic transport cross-section of the batches of fuel balls, the macroscopic scattering cross-section of the batches of fuel balls, the probability of neutron production by fission of the batches of fuel balls, and the fission neutron production cross-section of the batches of fuel balls to obtain the multi-group neutron flux density of the batches of fuel balls.

[0046] The third acquisition module is used to obtain the multi-group defect factors of each batch of fuel balls according to the multi-group average neutron flux density of each segment of the pebble bed core and the multi-group neutron flux density of each batch of fuel balls.

[0047] In some embodiments of the present application, the multi-group defect factor of each batch of fuel balls is obtained by the following formula:

[0048]

[0049] Among them, S Pebble,i,g is the defect factor of the g-th energy group of the i-th batch of fuel balls, Φ Pebble,i,gis the neutron flux density of the gth energy group of the i-th batch of fuel balls, Φ Nodal,g is the multi-group average neutron flux density of each segment of the pebble bed core.

[0050] The fourth acquisition module is used to obtain the multi-group neutron flux density of the TRISO fuel particles in each batch of fuel spheres according to the multi-group homogenization group constants of the TRISO fuel particles in each batch of fuel spheres.

[0051] In some embodiments of the present application, the multi-group homogenization group constants of the TRISO fuel particles in each batch of fuel spheres include: the macroscopic transport cross section of the TRISO fuel particles in each batch of fuel spheres, the macroscopic scattering cross section of the TRISO fuel particles in each batch of fuel spheres, the probability of neutron production by fission of the TRISO fuel particles in each batch of fuel spheres, and the fission neutron production cross section of the TRISO fuel particles in each batch of fuel spheres; wherein the fourth acquisition module is specifically configured to:

[0052] Obtaining the escape probability of the TRISO fuel particles in each batch of fuel spheres according to the macroscopic transport cross-section of the TRISO fuel particles in each batch of fuel spheres and the radius of the TRISO fuel particles in each batch of fuel spheres;

[0053] Obtaining the penetration probability of the TRISO fuel particles in each batch of fuel spheres according to the macroscopic transport cross-section of the TRISO fuel particles in each batch of fuel spheres and the radius of the TRISO fuel particles in each batch of fuel spheres;

[0054] Obtaining an average penetration probability of the TRISO fuel particles in each batch of fuel spheres according to the surface area fraction of the TRISO fuel particles in each batch of fuel spheres and the penetration probability of the TRISO fuel particles in each batch of fuel spheres;

[0055] Obtaining a collision probability between TRISO fuel particles in each batch of fuel spheres according to the escape probability of TRISO fuel particles in each batch of fuel spheres, the penetration probability of TRISO fuel particles in each batch of fuel spheres, the surface area fraction of TRISO fuel particles in each batch of fuel spheres, and the average penetration probability of TRISO fuel particles in each batch of fuel spheres;

[0056] Establish the second collision probability equations;

[0057] The second collision probability equations are solved based on the collision probability between TRISO fuel particles in each batch of fuel spheres, the volume fraction of TRISO fuel particles in each batch of fuel spheres, the macroscopic transport cross-section of TRISO fuel particles in each batch of fuel spheres, the macroscopic scattering cross-section of TRISO fuel particles in each batch of fuel spheres, the probability of neutron production by fission of TRISO fuel particles in each batch of fuel spheres, and the fission neutron production cross-section of TRISO fuel particles in each batch of fuel spheres to obtain the multi-group neutron flux density of TRISO fuel particles in each batch of fuel spheres.

[0058] The fifth acquisition module is used to obtain the multi-group defect factor of the TRISO fuel particles in each batch of fuel spheres according to the multi-group neutron flux density of each batch of fuel spheres and the multi-group neutron flux density of the TRISO fuel particles in each batch of fuel spheres.

[0059] In some embodiments of the present application, the multi-group defect factor of the TRISO fuel particles in each batch of fuel spheres is obtained by the following formula:

[0060]

[0061] Among them, D TRISO,i,g is the defect factor of the g-th energy group of TRISO fuel particles in the i-th batch of fuel balls, Φ TRISO,i,g is the neutron flux density of the g-th energy group of the TRISO fuel particles in the i-th batch of fuel balls, Φ Pebble,i,g is the neutron flux density of the gth energy group of the i-th batch of fuel balls.

[0062] The first determination module is used to determine the power of a single TRISO fuel particle in each block of the pebble bed core based on the multi-group average neutron flux density of each block of the pebble bed core, the multi-group defect factor of each batch of fuel balls, the multi-group defect factor of the TRISO fuel particles in each batch of fuel balls, and the fission energy production cross-section of the TRISO fuel particles in each batch of fuel balls.

[0063] In some embodiments of the present application, the first determining module is specifically configured to:

[0064] Obtaining a neutron flux distribution at a fuel position in the pebble bed core according to the multi-group average neutron flux density of each segment of the pebble bed core, the multi-group defect factor of each batch of fuel spheres, and the multi-group defect factor of TRISO fuel particles within each batch of fuel spheres;

[0065] The power of a single TRISO fuel particle in each segment of the pebble bed core is determined based on the neutron flux distribution at the fuel position of the pebble bed core and the fission energy generation cross section of the TRISO fuel particles in each batch of fuel spheres.

[0066] a second determining module, configured to determine the TRISO fuel particle power distribution of the pebble bed core according to the power of a single TRISO fuel particle of each segment of the pebble bed core;

[0067] The detection module is used to determine whether the pebble bed high temperature gas-cooled reactor is operating safely based on the TRISO fuel particle power distribution of the pebble bed core.

[0068] The third aspect of the present application proposes an electronic device, comprising: a processor; and a memory for storing instructions executable by the processor; wherein the instructions are executed by the processor so that the processor can execute the method described in the first aspect above.

[0069] The fourth aspect of the present application proposes a non-temporary computer-readable storage medium, characterized in that when the instructions in the storage medium are executed by a processor of an electronic device, the electronic device is enabled to execute the method described in the first aspect above.

[0070] According to the method for detecting the safe operation of a pebble-bed high-temperature gas-cooled reactor according to an embodiment of the present application, based on the calculation of the multi-group average neutron flux density of each node of the pebble-bed core using three-dimensional core diffusion, the multi-group neutron flux density of each batch of fuel spheres, the multi-group defect factor of each batch of fuel spheres, the multi-group neutron flux density of TRISO fuel particles within each batch of fuel spheres, and the multi-group defect factor of TRISO fuel particles within each batch of fuel spheres are further obtained, thereby determining the power of a single TRISO fuel particle in each node of the pebble-bed core and the TRISO fuel particle power distribution of the pebble-bed core. In this application, the dual defect factor (the multi-group defect factor of each batch of fuel spheres and the multi-group defect factor of TRISO fuel particles within each batch of fuel spheres) is used as the shape factor of the neutron flux density within the node of the pebble-bed core, taking into account the differences in the neutron flux of each batch of fuel spheres within the node and the differences in the neutron flux of TRISO fuel particles within each batch of fuel spheres. The TRISO fuel particle power distribution of the pebble-bed core determined in this way can more accurately reflect the "hot spots" of the fuel spheres. The TRISO fuel particle power distribution based on the pebble bed core can accurately detect the safe operating status of the pebble bed high-temperature gas-cooled reactor, providing a more reliable data basis for the nuclear design and safety analysis of the pebble bed high-temperature gas-cooled reactor.

[0071] Additional aspects and advantages of the present application will be given in part in the description below, and in part will become apparent from the description below, or will be learned through practice of the present application. BRIEF DESCRIPTION OF THE DRAWINGS

[0072] The above and / or additional aspects and advantages of the present application will become apparent and easily understood from the following description of the embodiments in conjunction with the accompanying drawings, in which:

[0073] Figure 1 A flow chart of a method for detecting safe operation of a pebble bed high temperature gas-cooled reactor provided in an embodiment of the present application;

[0074] Figure 2 A flowchart of the implementation process of obtaining the multi-group neutron flux density of each batch of fuel spheres according to the multi-group homogenization group constants of each batch of fuel spheres in the node provided in the embodiment of the present application;

[0075] Figure 3 A flowchart of a process for obtaining the multi-group neutron flux density of TRISO fuel particles in each batch of fuel spheres based on the multi-group homogenization group constants of TRISO fuel particles in each batch of fuel spheres provided in an embodiment of the present application;

[0076] Figure 4 A schematic diagram of a detection device for the safe operation of a pebble bed high-temperature gas-cooled reactor provided in an embodiment of the present application. DETAILED DESCRIPTION

[0077] The following describes in detail embodiments of the present application, examples of which are shown in the accompanying drawings, wherein the same or similar reference numerals throughout represent the same or similar elements or elements having the same or similar functions. The embodiments described below with reference to the accompanying drawings are exemplary and are intended to be used to explain the present application, and should not be construed as limiting the present application.

[0078] The present application proposes a method and device for detecting the safe operation of a pebble bed type high temperature gas-cooled reactor. Specifically, the following describes the method and device for detecting the safe operation of a pebble bed type high temperature gas-cooled reactor according to an embodiment of the present application with reference to the accompanying drawings.

[0079] Figure 1 This is a flow chart of a method for detecting safe operation of a pebble bed high temperature gas-cooled reactor provided in an embodiment of the present application. Figure 1 As shown, the method for detecting the safe operation of the pebble bed high temperature gas-cooled reactor includes the following steps:

[0080] Step 101 , performing core diffusion calculation on a pebble bed high temperature gas-cooled reactor based on a three-dimensional cylindrical geometry nodal method, and obtaining a multi-group average neutron flux density of each nodal block in the pebble bed core.

[0081] As a possible implementation method, the following three-dimensional multi-group steady-state neutron diffusion equation can be established:

[0082] -D g (r) 2 Φ g (r)+Σ rg (r)Φ g (r) = Q g (r)

[0083]

[0084] Among them, r is the spatial position vector, Φ g is the neutron flux density of the g-th energy group, D g is the diffusion coefficient of the g-th energy group, v is the average number of neutrons produced in each fission, ∑ fh is the macroscopic fission cross section of the hth energy group, χ g is the fission energy spectrum of the gth energy group, k eff is an effective growth factor, ∑ s,h→g is the macroscopic scattering cross section of neutrons from the hth energy group to the gth energy group, ∑ rg is the macroscopic removal cross section of the g-th energy group (i.e. ∑ rg =∑ tg -∑ s,g→g , where ∑ tg is the total macroscopic cross section of the g-th energy group, ∑ s,g→g is the macroscopic self-scattering cross section of the g-th energy group. ).

[0085] Under a three-dimensional cylindrical geometry, the neutron diffusion equations are transversely integrated to obtain one-dimensional equations for the radial (r), circumferential (θ), and axial (z) directions. By expanding these equations in a one-dimensional polynomial and introducing intermediate variables for the neutron bias flux, the neutron bias flux response relationships for the three directions within the nodule are established. Finally, a nodular neutron balance equation is established, which is solved to obtain the multi-group average neutron flux density for each nodule in the pebble bed core.

[0086] Step 102 : obtaining the multi-group neutron flux density of each batch of fuel balls in the node according to the multi-group homogenization group constants of each batch of fuel balls.

[0087] It should be noted that the pebble bed core randomly mixes fuel spheres from different batches (where "different batches" refers to the different burnup and temperature of the TRISO fuel particles within the fuel spheres). Therefore, the number of fuel spheres from each batch within each node varies. To facilitate calculations, a Monte Carlo physics program can be used to pre-calculate the multi-group homogenization group constants for all individual batches of fuel spheres and create a group constant library. In subsequent applications, this group constant library is called based on the number of fuel spheres from each batch within a node to obtain the multi-group homogenization group constants for each batch within the node.

[0088] As an example, the multi-group homogenization group constants of each batch of fuel balls in a block may include: the macroscopic transport cross section of each batch of fuel balls, the macroscopic scattering cross section of each batch of fuel balls, the probability of each batch of fuel balls producing neutrons by fission, the fission neutron production cross section of each batch of fuel balls, etc.

[0089] As a possible implementation, the collision probability between fuel spheres in each batch can be calculated using the multi-group homogenized group constant and collision probability method. Based on the multi-group homogenized group constant and the collision probability between fuel spheres in each batch, the multi-group neutron flux density of each fuel sphere batch can be obtained.

[0090] Step 103 , obtaining the multi-group defect factors of each batch of fuel balls according to the multi-group average neutron flux density of each segment of the pebble bed core and the multi-group neutron flux density of each batch of fuel balls.

[0091] It should be noted that the defect factor reflects the differences in neutron flux. In some embodiments of the present application, the multi-group defect factors of each batch of fuel balls can be obtained by the following formula.

[0092]

[0093] Among them, S Pebble,i,g is the defect factor of the g-th energy group of the i-th batch of fuel balls, Φ Pebble,i,g is the neutron flux density of the gth energy group of the i-th batch of fuel balls, Φ Nodal,g is the multi-group average neutron flux density of each segment of the pebble bed core.

[0094] Step 104 : obtaining the multi-group neutron flux density of the TRISO fuel particles in each batch of fuel spheres based on the multi-group homogenized group constants of the TRISO fuel particles in each batch of fuel spheres.

[0095] It should be noted that the method for obtaining the multi-group neutron flux density of TRISO fuel particles within each batch of fuel spheres is basically similar to that for obtaining the multi-group neutron flux density of each batch of fuel spheres. A single fuel sphere can be equated to a single block, and the TRISO fuel particles within a single fuel sphere can be equated to fuel spheres within a single block. Furthermore, a Monte Carlo physics calculation program is used to calculate the multi-group homogenized group constants of different batches of TRISO fuel particles within the fuel sphere, and a group constant library is created.

[0096] As an example, the multi-group homogenization group constants of each batch of fuel balls in a block may include: the macroscopic transport cross section of each batch of fuel balls, the macroscopic scattering cross section of each batch of fuel balls, the probability of each batch of fuel balls producing neutrons by fission, the fission neutron production cross section of each batch of fuel balls, etc.

[0097] As a possible implementation, the collision probability between fuel spheres in each batch can be calculated using the multi-group homogenized group constant and collision probability method. Based on the multi-group homogenized group constant and the collision probability between fuel spheres in each batch, the multi-group neutron flux density of each fuel sphere batch can be obtained.

[0098] Step 105 , obtaining the multi-group defect factors of the TRISO fuel particles in each batch of fuel spheres based on the multi-group neutron flux density of each batch of fuel spheres and the multi-group neutron flux density of the TRISO fuel particles in each batch of fuel spheres.

[0099] In some embodiments of the present application, the multi-group defect factor S of TRISO fuel particles in each batch of fuel balls can be obtained by the following formula: TRISo,i,g .

[0100]

[0101] Among them, S TRISO,i,g is the defect factor of the g-th energy group of TRISO fuel particles in the i-th batch of fuel balls, Φ TRISO,i,g is the neutron flux density of the g-th energy group of the TRISO fuel particles in the i-th batch of fuel balls, Φ Pebble,i,g is the neutron flux density of the gth energy group of the i-th batch of fuel balls.

[0102] Step 106, determines the power of a single TRISO fuel particle in each segment of the pebble bed core based on the multi-group average neutron flux density of each segment of the pebble bed core, the multi-group defect factor of each batch of fuel balls, the multi-group defect factor of the TRISO fuel particles in each batch of fuel balls, and the fission energy generation cross section of the TRISO fuel particles in each batch of fuel balls.

[0103] In some embodiments of the present application, the neutron flux density at the fuel position within the pebble bed core segment can be obtained based on the multi-group average neutron flux density of each pebble bed core segment, the multi-group defect factor of each batch of fuel spheres, and the multi-group defect factor of the TRISO fuel particles within each batch of fuel spheres. Furthermore, the power of a single TRISO fuel particle in each pebble bed core segment can be determined based on the neutron flux density at the fuel position within the pebble bed core segment and the fission energy generation cross section of the TRISO fuel particles within each batch of fuel spheres.

[0104] As an example, the calculation formula for the neutron flux density at the fuel position in the pebble bed core segment can be expressed as follows.

[0105]

[0106] Among them, Φ Fuel，g is the neutron flux density of the g-th energy group at the fuel position in the pebble bed core segment;

[0107] Φ Nodal，g is the multi-group average neutron flux density of each segment of the pebble bed core;

[0108] Φ Pebble，i，g is the neutron flux density of the gth energy group of the i-th batch of fuel balls;

[0109] ΦTRISO，i，g is the neutron flux density of the g-th energy group of the TRISO fuel particles in the i-th batch of fuel balls;

[0110] S Peb b le，i，g is the defect factor of the g-th energy group of the i-th batch of fuel balls;

[0111] S TRISO，i，g is the defect factor of the g-th energy group of TRISO fuel particles in the i-th batch of fuel balls.

[0112] Therefore, the neutron flux density at the fuel position within a pebble-bed core segment represents the multi-group neutron flux density of TRISO fuel particles at different locations and within different fuel sphere batches. Furthermore, based on the neutron flux density at the fuel position within the pebble-bed core segment and the fission energy generation cross-section of the TRISO fuel particles within each fuel sphere batch, the power of a single TRISO fuel particle in each pebble-bed core segment is determined. The formula for calculating the power of a single TRISO fuel particle in each pebble-bed core segment can be expressed as follows.

[0113]

[0114] Where P is the power of a single TRISO fuel particle in a pebble bed high temperature gas-cooled reactor segment;

[0115] (κ∑ f ) TRI，i，g is the fission energy generation cross section of the g-th energy group of TRISO fuel particles in the i-th batch of fuel spheres;

[0116] Φ Fuel，g is the neutron flux density of the g-th energy group at the fuel position in the pebble bed core segment;

[0117] Φ TRISO，i，g is the multi-group neutron flux density of TRISO fuel particles in the i-th batch of fuel balls.

[0118] Step 107 : determining the power distribution of the TRISO fuel particles in the pebble bed core according to the power of the individual TRISO fuel particles in each segment of the pebble bed core.

[0119] That is, all TRISO fuel particle powers of each node of the pebble bed core can be obtained based on step 106. Based on all TRISO fuel particle powers of each node of the pebble bed core, the TRISO fuel particle power distribution of the pebble bed core can be determined.

[0120] Step 108 : Based on the TRISO fuel particle power distribution of the pebble bed core, determine whether the pebble bed high temperature gas-cooled reactor is operating safely.

[0121] As a possible implementation, in some embodiments of the present application, after determining the TRISO fuel particle power distribution of a pebble bed core, this TRISO fuel particle power distribution can be substituted into subsequent thermal calculations to obtain a temperature distribution. The numerical results can be displayed graphically, visually reflecting the temperature distribution of the pebble bed high-temperature gas-cooled reactor and determining whether the maximum temperature exceeds a maximum temperature threshold. For example, if the temperature at a certain location exceeds the maximum temperature threshold, corresponding warning operations can be initiated, thereby detecting the safe operating status of the pebble bed high-temperature gas-cooled reactor.

[0122] According to the method for detecting the safe operation of a pebble-bed high-temperature gas-cooled reactor according to an embodiment of the present application, based on the calculation of the multi-group average neutron flux density of each node of the pebble-bed core using three-dimensional core diffusion, the multi-group neutron flux density of each batch of fuel spheres, the multi-group defect factor of each batch of fuel spheres, the multi-group neutron flux density of TRISO fuel particles within each batch of fuel spheres, and the multi-group defect factor of TRISO fuel particles within each batch of fuel spheres are further obtained, thereby determining the power of a single TRISO fuel particle in each node of the pebble-bed core and the TRISO fuel particle power distribution of the pebble-bed core. In this application, the dual defect factor (the multi-group defect factor of each batch of fuel spheres and the multi-group defect factor of TRISO fuel particles within each batch of fuel spheres) is used as the shape factor of the neutron flux density within the node of the pebble-bed core, taking into account the differences in the neutron flux of each batch of fuel spheres within the node and the differences in the neutron flux of TRISO fuel particles within each batch of fuel spheres. The TRISO fuel particle power distribution of the pebble-bed core determined in this way can more accurately reflect the "hot spots" of the fuel spheres. The TRISO fuel particle power distribution based on the pebble bed core can accurately detect the safe operating status of the pebble bed high-temperature gas-cooled reactor, providing a more reliable data basis for the nuclear design and safety analysis of the pebble bed high-temperature gas-cooled reactor.

[0123] In order to accurately obtain the multi-group neutron flux density of each batch of fuel balls, in some embodiments of the present application, the multi-group homogenization group constants of each batch of fuel balls may include the macroscopic transport cross section of each batch of fuel balls, the macroscopic scattering cross section of each batch of fuel balls, the probability of each batch of fuel balls producing neutrons by fission, and the fission neutron production cross section of each batch of fuel balls. The process of obtaining the multi-group neutron flux density of each batch of fuel balls according to the multi-group homogenization group constants of each batch of fuel balls in the block in step 102 can be as follows: Figure 2 As shown, the following steps are included:

[0124] Step 201 : Obtain the escape probability of each batch of fuel balls according to the macroscopic transport cross section and radius of each batch of fuel balls.

[0125] As an example, the calculation formula for the escape probability of each batch of fuel balls can be expressed as follows.

[0126]

[0127] Among them, E Peb,i,g is the escape probability of each batch of fuel balls, that is, the escape probability of the g-th energy group neutrons uniformly generated in the i-th batch of fuel balls escaping the fuel ball for the first time without collision; ∑ Peb,i,tr,g is the macroscopic transport cross section of the g-th energy group of the i-th batch of fuel balls; R Peb,i is the radius of the fuel ball of the i-th batch.

[0128] Step 202 : Obtain the penetration probability of each batch of fuel balls according to the macroscopic transport cross section and the radius of each batch of fuel balls.

[0129] As an example, the calculation formula for the penetration probability of each batch of fuel balls can be expressed as follows.

[0130]

[0131] Among them, T Peb,i,g The penetration probability of each batch of fuel balls is the probability that a neutron of energy group g that enters the fuel ball of the i-th batch isotropically penetrates the fuel ball without collision; ∑ Peb,i,tr,g is the macroscopic transport cross section of the g-th energy group of the i-th batch of fuel balls; R Peb,i is the radius of the fuel ball of the i-th batch.

[0132] Step 203 : obtaining the average penetration probability of each batch of fuel balls according to the surface area fraction of each batch of fuel balls and the penetration probability of each batch of fuel balls.

[0133] As an example, the average penetration probability of each batch of fuel balls is The calculation formula can be expressed as follows.

[0134]

[0135] in, A Peb,i is the surface area fraction of the fuel balls in the i-th batch; f Peb,i is the volume fraction of the i-th batch of fuel balls (i.e., the ratio of the volume of the i-th batch of fuel balls in the node to the volume of all batches of fuel balls in the node); T Peb,i,g is the probability that a neutron of energy group g that enters the i-th batch of fuel balls isotropically penetrates the fuel balls without collision.

[0136] Step 204 : Obtain the collision probability between each batch of fuel balls based on the escape probability of each batch of fuel balls, the penetration probability of each batch of fuel balls, the surface area share of each batch of fuel balls, and the average penetration probability of each batch of fuel balls in the node.

[0137] In some embodiments of the present application, the following collision process can be considered: a source neutron of the g-th energy group generated in a fuel ball in the i-th batch escapes the fuel ball and enters the next fuel ball isotropically. The probability that the fuel ball that enters belongs to the j-th batch is equal to the volume fraction f of the fuel ball in the j-th batch. Peb,j Similarly, the probability of a neutron of the g-th energy group colliding with the fuel ball is 1-T Peb,j,g If a neutron doesn't penetrate the fuel sphere, it will continue to the next fuel sphere, and the above process will repeat. Based on the physical meaning of collision probability, the collision probability between each batch of fuel spheres can be obtained. The calculation formula can be expressed as follows.

[0138]

[0139]

[0140] Where, P Peb,i→j,i≠j,h is the collision probability of a neutron of the g-th energy group escaping the i-th batch of fuel balls and entering the j-th batch of fuel balls, where the escaped i-th batch of fuel balls and the j-th batch of fuel balls are from different batches, that is, i≠j;

[0141] P Peb,i→j,i＝j,g is the collision probability of a neutron of the g-th energy group escaping the i-th batch of fuel balls and entering the j-th batch of fuel balls, where the escaped i-th batch of fuel balls and the j-th batch of fuel balls are from the same batch, that is, i=j;

[0142] E Peb,i,g is the probability that neutrons of the g-th energy group uniformly generated in the i-th batch of fuel spheres escape from the fuel sphere for the first time without collision;

[0143] A Peb,j is the surface area share of the jth batch of fuel balls;

[0144] T Peb,j,g is the probability that a neutron of energy group g that enters the j-th batch of fuel balls isotropically penetrates the fuel balls without collision;

[0145] is the average penetration probability of each batch of fuel balls.

[0146] Step 205: Establish a first collision probability equation group.

[0147] For a certain block in the core of the bed, the following first collision probability equations can be established.

[0148]

[0149] Among them, f Peb,j is the volume fraction of the jth batch of fuel balls;

[0150] ∑Peb,j,tr,g is the macroscopic transport cross section of the g-th energy group of the j-th batch of fuel balls;

[0151] Φ Pebble,j,g is the neutron flux density of the gth energy group of the jth batch of fuel balls;

[0152] ∑ Peb,i,s,g′→g is the macroscopic scattering cross section of neutrons from the gˊth energy group of the i-th batch of fuel balls scattered to the g-th energy group;

[0153] Φ Pebble,i,g′ is the neutron flux density of the gˊth energy group of the i-th batch of fuel balls;

[0154] χ Peb，i，g is the probability of producing neutrons of energy group g during the fission of the i-th batch of fuel balls;

[0155] k eff It is an effective proliferation factor;

[0156] (v∑ f ) Peb，i，g′ is the fission neutron production cross section of the g′th energy group of the i-th batch of fuel balls;

[0157] P Peb，i→j，g is the collision probability of a neutron of the g-th energy group escaping from the i-th batch of fuel balls and entering the j-th batch of fuel balls.

[0158] Step 206, based on the collision probability between each batch of fuel balls, the volume fraction of each batch of fuel balls, the macroscopic transport cross section of each batch of fuel balls, the macroscopic scattering cross section of each batch of fuel balls, the probability of each batch of fuel balls producing neutrons by fission, and the fission neutron production cross section of each batch of fuel balls, solve the first collision probability equation group to obtain the multi-group neutron flux density of each batch of fuel balls.

[0159] That is, based on the collision probability P between each batch of fuel balls Peb，i→j，g , the volume fraction of each batch of fuel balls f Peb，j , Macroscopic transport cross section of each batch of fuel balls∑ Peb，j，tr，g , macroscopic scattering cross section of each batch of fuel balls ∑ Peb，i，s，g′→g , the probability of each batch of fuel balls producing neutrons by fission χ Peb，i，g , fission neutron production cross section of each batch of fuel balls (v∑ f )P eb，i，g′ , solve the first collision probability equations to obtain the multi-group neutron flux density Φ of each batch of fuel balls Pebble，j，g .

[0160] Thus, through steps 201 to 206 , the multi-group neutron flux density of each batch of fuel balls in the node can be accurately obtained based on the multi-group homogenization group constants of the fuel balls in the node.

[0161] In addition, in order to accurately obtain the multi-group neutron flux density of the TRISO fuel particles in each batch of fuel spheres, in some embodiments of the present application, the process of obtaining the multi-group neutron flux density of the TRISO fuel particles in each batch of fuel spheres according to the multi-group homogenization group constant of the TRISO fuel particles in each batch of fuel spheres in step 104 can be as follows: Figure 3 As shown, the following steps are included:

[0162] Step 301 : Obtain the escape probability of the TRISO fuel particles in each batch of fuel spheres according to the macroscopic transport cross section of the TRISO fuel particles in each batch of fuel spheres and the radius of the TRISO fuel particles in each batch of fuel spheres.

[0163] It should be noted that, in order to simplify the calculation, it can be assumed that all TRISO fuel particles in the same batch of fuel spheres are the same, that is, a single fuel sphere contains TRISO fuel particles from only one batch.

[0164] As an example, the calculation formula for the escape probability of TRISO fuel particles in each batch of fuel balls is expressed as follows.

[0165]

[0166] Among them, E TRI，i，g is the escape probability of a TRISO fuel particle in each batch of fuel spheres, that is, the escape probability of the g-th energy group neutrons uniformly generated by the TRISO fuel particles in the i-th batch of fuel spheres escaping the TRISO fuel particle for the first time without collision; ∑ TRI,i,tr,g is the macroscopic transport cross section of the g-th energy group of TRISO fuel particles in the i-th batch of fuel balls; R TRI,i is the radius of the TRISO fuel particle in the i-th batch of fuel spheres.

[0167] Step 302 : Obtain the penetration probability of the TRISO fuel particles in each batch of fuel spheres according to the macroscopic transport cross section of the TRISO fuel particles in each batch of fuel spheres and the radius of the TRISO fuel particles in each batch of fuel spheres.

[0168] As an example, the calculation formula for the penetration probability of TRISO fuel particles in each batch of fuel balls is expressed as follows.

[0169]

[0170] Among them, T TRI,i,g is the penetration probability of TRISO fuel particles in each batch of fuel spheres, that is, the probability that a neutron of the g-th energy group that enters a TRISO fuel particle in the i-th batch of fuel spheres isotropically penetrates the TRISO fuel particle without collision; ∑ TRI,i,tr,gis the macroscopic transport cross section of the g-th energy group of TRISO fuel particles in the i-th batch of fuel balls; R TRI,i is the radius of the TRISO fuel particle in the i-th batch of fuel spheres.

[0171] Step 303 : obtaining an average penetration probability of the TRISO fuel particles in each batch of fuel spheres according to the surface area fraction of the TRISO fuel particles in each batch of fuel spheres and the penetration probability of the TRISO fuel particles in each batch of fuel spheres.

[0172] As mentioned above, in this embodiment, it is assumed that all TRISO fuel particles in the same batch of fuel balls are the same, that is, a single fuel ball contains only TRISO fuel particles from one batch, so the volume fraction A of TRISO fuel particles in the fuel ball is TRI,i Always equal to 1.

[0173] As an example, the average penetration probability of TRISO fuel particles in each batch of fuel balls is The calculation formula can be expressed as follows.

[0174]

[0175] Among them, T TRI,i,g is the probability that a neutron of energy group g that enters a TRISO fuel particle in the i-th batch of fuel spheres isotropically will penetrate the TRISO fuel particle without collision. Therefore, the average penetration probability of TRISO fuel particles in each batch of fuel spheres is equal to the penetration probability of TRISO fuel particles in each batch of fuel spheres.

[0176] In step 304, the collision probability between the TRISO fuel particles in each batch of fuel spheres is obtained based on the escape probability of the TRISO fuel particles in each batch of fuel spheres, the penetration probability of the TRISO fuel particles in each batch of fuel spheres, the surface area share of the TRISO fuel particles in each batch of fuel spheres, and the average penetration probability of the TRISO fuel particles in each batch of fuel spheres.

[0177] According to the physical meaning of collision probability, the collision probability between TRISO fuel particles in each batch of fuel balls is obtained, and its calculation formula can be expressed as follows.

[0178]

[0179] Among them, P TRI，i→i，g is the collision probability of a neutron of the gth energy group escaping from a TRISO fuel particle and entering another TRISO fuel particle in the i-th batch of fuel balls; E TRI，i，g is the probability that the g-th energy group neutrons uniformly generated by the TRISO fuel particles in the i-th batch of fuel spheres escape from the TRISO fuel particles for the first time without collision; A TRI，iis the volume fraction of TRISO fuel particles in the fuel sphere; T TRI，i，g is the probability that a neutron of energy group g isotropically enters a TRISO fuel particle in the i-th batch of fuel spheres and penetrates the TRISO fuel particle without collision; is the average penetration probability of TRISO fuel particles in each batch of fuel balls.

[0180] Step 305: Establish a second collision probability equation group.

[0181] For a certain batch of fuel balls, the following second collision probability equations can be established.

[0182]

[0183] Further simplifying the above formula we can get:

[0184]

[0185] Among them, f TRI，i is the volume fraction of TRISO fuel particles in the i-th batch of fuel balls;

[0186] ∑ TRI，i，tr，g is the macroscopic transport cross section of the g-th energy group of TRISO fuel particles in the i-th batch of fuel balls;

[0187] Φ TRISO，i，g is the neutron flux density of the g-th energy group of the TRISO fuel particles in the i-th batch of fuel balls;

[0188] ∑ TRI，i，s，g′→g is the macroscopic scattering cross section of neutrons from the g′th energy group of the TRISO fuel particles in the i-th batch of fuel spheres scattered to the g-th energy group;

[0189] Φ TRISO，i，g′ is the neutron flux density of the g′th energy group of the TRISO fuel particles in the i-th batch of fuel balls;

[0190] χ Peb，i，g is the probability of producing neutrons of energy group g by fission of TRISO fuel particles in the i-th batch of fuel balls;

[0191] k eff It is an effective proliferation factor;

[0192] (v∑ f ) TRI，i，g′ is the fission neutron production cross section of the g′th energy group of the TRISO fuel particles in the i-th batch of fuel spheres;

[0193] P TRI，i→i，g is the collision probability of a neutron of the gth energy group escaping from a TRISO fuel particle and entering other TRISO fuel particles in the i-th batch of fuel balls.

[0194] Step 306: Solve the second collision probability equations based on the collision probability between TRISO fuel particles in each batch of fuel spheres, the volume fraction of TRISO fuel particles in each batch of fuel spheres, the macroscopic transport cross-section of TRISO fuel particles in each batch of fuel spheres, the macroscopic scattering cross-section of TRISO fuel particles in each batch of fuel spheres, the probability of neutron production by fission of TRISO fuel particles in each batch of fuel spheres, and the fission neutron production cross-section of TRISO fuel particles in each batch of fuel spheres to obtain the multi-group neutron flux density of TRISO fuel particles in each batch of fuel spheres.

[0195] That is, according to the collision probability P between TRISO fuel particles in each batch of fuel balls TRI，i→i，g , the volume fraction f of TRISO fuel particles in each batch of fuel balls TRI，i , macroscopic transport cross section of TRISO fuel particles in each batch of fuel balls ∑ TRI，i，tr，g , the macroscopic scattering cross section of TRISO fuel particles in each batch of fuel balls ∑ TRI，i，s，g′→g , the probability of producing neutrons by fission of TRISO fuel particles in each batch of fuel balls χ Peb，i，g and the fission neutron production cross section (v∑ f ) TRI，i，g′ , solve the second collision probability equations to obtain the multi-group neutron flux density Φ of TRISO fuel particles in each batch of fuel balls TRISO，i，g .

[0196] Therefore, through steps 301 to 306 , the multi-group neutron flux density of the TRISO fuel particles in each batch of fuel spheres can be accurately obtained based on the multi-group homogenization group constants of the TRISO fuel particles in each batch of fuel spheres.

[0197] Figure 4 This is a schematic diagram of a detection device for the safe operation of a pebble bed type high temperature gas-cooled reactor provided in an embodiment of the present application. Figure 4 As shown, the detection device for safe operation of the pebble bed type high temperature gas-cooled reactor includes: a first acquisition module 401, a second acquisition module 402, a third acquisition module 403, a fourth acquisition module 404, a fifth acquisition module 405, a first determination module 406, a second determination module 407 and a detection module 408.

[0198] The first acquisition module 401 is used to perform core diffusion calculation on the pebble bed high temperature gas-cooled reactor based on the three-dimensional cylindrical geometry block method to obtain the multi-group average neutron flux density of each block in the pebble bed core.

[0199] The second acquisition module 402 is configured to obtain the multi-group neutron flux density of each batch of fuel balls according to the multi-group homogenized group constants of each batch of fuel balls in the node.

[0200] In some embodiments of the present application, the multi-group homogenization group constants of each batch of fuel balls include: the macroscopic transport cross-section of each batch of fuel balls, the macroscopic scattering cross-section of each batch of fuel balls, the probability of each batch of fuel balls producing neutrons by fission, and the fission neutron production cross-section of each batch of fuel balls.

[0201] Among them, the second acquisition module 402 is specifically used to: obtain the escape probability of each batch of fuel balls based on the macroscopic transport cross-section of each batch of fuel balls and the radius of each batch of fuel balls; obtain the penetration probability of each batch of fuel balls based on the macroscopic transport cross-section of each batch of fuel balls and the radius of each batch of fuel balls; obtain the average penetration probability of each batch of fuel balls based on the surface area share of each batch of fuel balls and the penetration probability of each batch of fuel balls; obtain the collision probability between each batch of fuel balls based on the escape probability of each batch of fuel balls, the penetration probability of each batch of fuel balls, the surface area share of each batch of fuel balls and the average penetration probability of each batch of fuel balls in the node; establish a first collision probability equation group; solve the first collision probability equation group based on the collision probability between each batch of fuel balls, the volume share of each batch of fuel balls, the macroscopic transport cross-section of each batch of fuel balls, the macroscopic scattering cross-section of each batch of fuel balls, the probability of each batch of fuel balls producing neutrons by fission, and the fission neutron production cross-section of each batch of fuel balls to obtain the multi-group neutron flux density of each batch of fuel balls.

[0202] The third acquisition module 403 is used to obtain the multi-group defect factor of each batch of fuel balls according to the multi-group average neutron flux density of each segment of the pebble bed core and the multi-group neutron flux density of each batch of fuel balls.

[0203] In some embodiments of the present application, the multi-group defect factor of each batch of fuel balls is obtained by the following formula:

[0204]

[0205] Among them, S Pebble,i,g is the defect factor of the g-th energy group of the i-th batch of fuel balls, Φ Pebble,i,g is the neutron flux density of the gth energy group of the i-th batch of fuel balls, Φ Nodal,g is the multi-group average neutron flux density of each segment of the pebble bed core.

[0206] The fourth acquisition module 404 is configured to obtain the multi-group neutron flux density of the TRISO fuel particles in each batch of fuel spheres according to the multi-group homogenization group constants of the TRISO fuel particles in each batch of fuel spheres.

[0207] In some embodiments of the present application, the multi-group homogenization group constants of the TRISO fuel particles in each batch of fuel spheres include: the macroscopic transport cross-section of the TRISO fuel particles in each batch of fuel spheres, the macroscopic scattering cross-section of the TRISO fuel particles in each batch of fuel spheres, the probability of neutron production by fission of the TRISO fuel particles in each batch of fuel spheres, and the fission neutron production cross-section of the TRISO fuel particles in each batch of fuel spheres.

[0208] Among them, the fourth acquisition module 404 is specifically used to: obtain the escape probability of TRISO fuel particles in each batch of fuel spheres according to the macroscopic transport cross-section of TRISO fuel particles in each batch of fuel spheres and the radius of TRISO fuel particles in each batch of fuel spheres; obtain the penetration probability of TRISO fuel particles in each batch of fuel spheres according to the macroscopic transport cross-section of TRISO fuel particles in each batch of fuel spheres and the radius of TRISO fuel particles in each batch of fuel spheres; obtain the average penetration probability of TRISO fuel particles in each batch of fuel spheres according to the surface area share of TRISO fuel particles in each batch of fuel spheres and the penetration probability of TRISO fuel particles in each batch of fuel spheres; obtain the average penetration probability of TRISO fuel particles in each batch of fuel spheres according to the escape probability of TRISO fuel particles in each batch of fuel spheres and the penetration probability of TRISO fuel particles in each batch of fuel spheres. , the surface area share of TRISO fuel particles in each batch of fuel spheres and the average penetration probability of TRISO fuel particles in each batch of fuel spheres, to obtain the collision probability between TRISO fuel particles in each batch of fuel spheres; establish a second collision probability equation group; according to the collision probability between TRISO fuel particles in each batch of fuel spheres, the volume share of TRISO fuel particles in each batch of fuel spheres, the macroscopic transport cross-section of TRISO fuel particles in each batch of fuel spheres, the macroscopic scattering cross-section of TRISO fuel particles in each batch of fuel spheres, the probability of fission neutron production of TRISO fuel particles in each batch of fuel spheres and the fission neutron production cross-section of TRISO fuel particles in each batch of fuel spheres, solve the second collision probability equation group to obtain the multi-group neutron flux density of TRISO fuel particles in each batch of fuel spheres.

[0209] The fifth acquisition module 405 is used to obtain the multi-group defect factors of the TRISO fuel particles in each batch of fuel spheres based on the multi-group neutron flux density of each batch of fuel spheres and the multi-group neutron flux density of the TRISO fuel particles in each batch of fuel spheres.

[0210] In some embodiments of the present application, the multi-group defect factor of TRISO fuel particles in each batch of fuel spheres is obtained by the following formula:

[0211]

[0212] Among them, S TRISo,i,g is the defect factor of the g-th energy group of TRISO fuel particles in the i-th batch of fuel balls, ΦTRISO,i,g is the neutron flux density of the g-th energy group of the TRISO fuel particles in the i-th batch of fuel balls, Φ Pebble,i,g is the neutron flux density of the gth energy group of the i-th batch of fuel balls.

[0213] The first determination module 406 is used to determine the power of a single TRISO fuel particle in each section of the pebble bed core based on the multi-group average neutron flux density of each section of the pebble bed core, the multi-group defect factor of each batch of fuel balls, the multi-group defect factor of the TRISO fuel particles in each batch of fuel balls, and the fission energy production cross section of the TRISO fuel particles in each batch of fuel balls.

[0214] In some embodiments of the present application, the first determination module 406 is specifically used to: obtain the neutron flux distribution of the fuel position of the pebble bed core based on the multi-group average neutron flux density of each node of the pebble bed core, the multi-group defect factor of each batch of fuel balls, and the multi-group defect factor of the TRISO fuel particles in each batch of fuel balls; determine the power of a single TRISO fuel particle of each node of the pebble bed core based on the neutron flux distribution of the fuel position of the pebble bed core and the fission energy generation cross-section of the TRISO fuel particles in each batch of fuel balls.

[0215] The second determining module 407 is used to determine the power distribution of the TRISO fuel particles in the pebble bed core according to the power of a single TRISO fuel particle in each segment of the pebble bed core.

[0216] The detection module 408 is used to determine whether the pebble bed high temperature gas-cooled reactor is operating safely based on the TRISO fuel particle power distribution of the pebble bed core.

[0217] Regarding the apparatus of the above embodiment, the specific manner in which each module performs operations has been described in detail in the embodiment of the method, and will not be elaborated here.

[0218] According to the detection device for safe operation of a pebble-bed high-temperature gas-cooled reactor according to the embodiment of the present application, based on the calculation of the multi-group average neutron flux density of each node of the pebble-bed core using three-dimensional core diffusion, the multi-group neutron flux density of each batch of fuel spheres, the multi-group defect factor of each batch of fuel spheres, the multi-group neutron flux density of TRISO fuel particles within each batch of fuel spheres, and the multi-group defect factor of TRISO fuel particles within each batch of fuel spheres are further obtained, thereby determining the power of a single TRISO fuel particle in each node of the pebble-bed core and the TRISO fuel particle power distribution of the pebble-bed core. The present application uses the dual defect factor (the multi-group defect factor of each batch of fuel spheres and the multi-group defect factor of TRISO fuel particles within each batch of fuel spheres) as the shape factor of the neutron flux density within the node of the pebble-bed core, taking into account the neutron flux differences of each batch of fuel spheres within the node and the neutron flux differences of TRISO fuel particles within each batch of fuel spheres. The TRISO fuel particle power distribution of the pebble-bed core determined in this way can more accurately reflect the "hot spots" of the fuel spheres. The TRISO fuel particle power distribution based on the pebble bed core can accurately detect the safe operating status of the pebble bed high-temperature gas-cooled reactor, providing a more reliable data basis for the nuclear design and safety analysis of the pebble bed high-temperature gas-cooled reactor.

[0219] To implement the above embodiment, the present application further proposes an electronic device comprising: a processor and a memory for storing instructions executable by the processor, wherein the instructions are executed by the processor to enable the processor to perform the above-mentioned method for detecting safe operation of a pebble bed high-temperature gas-cooled reactor.

[0220] In order to implement the above embodiments, the present application also proposes a non-temporary computer-readable storage medium, which, when the instructions in the storage medium are executed by the processor of an electronic device, enables the electronic device to execute the aforementioned detection method for safe operation of a pebble bed high-temperature gas-cooled reactor.

[0221] In the description of this specification, the description with reference to the terms "one embodiment", "some embodiments", "example", "specific example", or "some examples" means that the specific features, structures, materials or characteristics described in conjunction with the embodiment or example are included in at least one embodiment or example of the present application. 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. In addition, those skilled in the art can combine and combine different embodiments or examples described in this specification and features of different embodiments or examples without contradiction.

[0222] Furthermore, the terms "first" and "second" are used for descriptive purposes only and should not be construed as indicating or implying relative importance or implicitly specifying the number of the technical features being referred to. Thus, a feature defined as "first" or "second" may explicitly or implicitly include at least one of such features. Throughout the description of this application, "plurality" means at least two, for example, two, three, etc., unless otherwise specifically defined.

[0223] Any process or method description in a flowchart or otherwise described herein may be understood to represent a module, segment or portion of code comprising one or more executable instructions for implementing the steps of a custom logical function or process, and the scope of the preferred embodiments of the present application includes alternative implementations in which functions may be performed out of the order shown or discussed, including performing functions in a substantially simultaneous manner or in the reverse order depending on the functions involved, which should be understood by those skilled in the art to which the embodiments of the present application belong.

[0224] It is understandable that the above embodiments are exemplary and should not be construed as limiting the present application. A person skilled in the art may change, modify, replace and modify the above embodiments within the scope of the present application.

Claims

1. A method for detecting safe operation of a pebble bed high temperature gas-cooled reactor, characterized in that: The following steps are involved: The core diffusion of a pebble bed high temperature gas-cooled reactor is calculated based on the three-dimensional cylindrical geometry segment method, and the multi-group average neutron flux density of each segment of the pebble bed core is obtained. Obtaining multi-group neutron flux densities of each batch of fuel balls in the node according to multi-group homogenization group constants of the batch of fuel balls; Obtaining a multi-group defect factor of each batch of fuel spheres according to the multi-group average neutron flux density of each segment of the pebble bed core and the multi-group neutron flux density of each batch of fuel spheres; Obtaining multi-group neutron flux density of the TRISO fuel particles in each batch of fuel spheres according to the multi-group homogenization group constant of the TRISO fuel particles in each batch of fuel spheres; Obtaining a multi-group defect factor of TRISO fuel particles in each batch of fuel spheres according to the multi-group neutron flux density of each batch of fuel spheres and the multi-group neutron flux density of TRISO fuel particles in each batch of fuel spheres; Determining the power of a single TRISO fuel particle in each segment of the pebble bed core based on the multi-group average neutron flux density of each segment of the pebble bed core, the multi-group defect factor of each batch of fuel spheres, the multi-group defect factor of the TRISO fuel particles in each batch of fuel spheres, and the fission energy production cross section of the TRISO fuel particles in each batch of fuel spheres; determining a TRISO fuel particle power distribution of the pebble bed core according to the power of a single TRISO fuel particle of each segment of the pebble bed core; Based on the TRISO fuel particle power distribution of the pebble bed core, it is determined whether the pebble bed high temperature gas-cooled reactor is operating safely.

2. The method according to claim 1, characterized in that The multi-group homogenization group constants of each batch of fuel balls include: the macroscopic transport cross section of each batch of fuel balls, the macroscopic scattering cross section of each batch of fuel balls, the probability of neutron production by fission of each batch of fuel balls, and the fission neutron production cross section of each batch of fuel balls; wherein, obtaining the multi-group neutron flux density of each batch of fuel balls according to the multi-group homogenization group constants of each batch of fuel balls in the node includes: Obtaining the escape probability of each batch of fuel balls according to the macroscopic transport cross-section of each batch of fuel balls and the radius of each batch of fuel balls; Obtaining the penetration probability of each batch of fuel balls according to the macroscopic transport cross-section of each batch of fuel balls and the radius of each batch of fuel balls; Obtaining an average penetration probability of each batch of fuel balls according to the surface area fractions of each batch of fuel balls and the penetration probability of each batch of fuel balls; Obtaining collision probabilities between fuel balls in each batch based on the escape probabilities of the fuel balls in each batch, the penetration probabilities of the fuel balls in each batch, the surface area fractions of the fuel balls in each batch, and the average penetration probabilities of the fuel balls in each batch within the node; Establish the first collision probability equations; The first collision probability equations are solved based on the collision probability between the batches of fuel balls, the volume fraction of the batches of fuel balls, the macroscopic transport cross-section of the batches of fuel balls, the macroscopic scattering cross-section of the batches of fuel balls, the probability of neutron production by fission of the batches of fuel balls, and the fission neutron production cross-section of the batches of fuel balls to obtain the multi-group neutron flux density of the batches of fuel balls.

3. The method according to claim 1, characterized in that The multi-group defect factor of each batch of fuel balls is obtained by the following formula: Among them, S Pebble,i,g is the defect factor of the g-th energy group of the i-th batch of fuel balls, Φ Pebble,i,g is the neutron flux density of the gth energy group of the i-th batch of fuel balls, Φ Nodal,g is the multi-group average neutron flux density of each segment of the pebble bed core.

4. The method according to claim 1, wherein The multi-group homogenized group constants of the TRISO fuel particles in each batch of fuel spheres include: a macroscopic transport cross section of the TRISO fuel particles in each batch of fuel spheres, a macroscopic scattering cross section of the TRISO fuel particles in each batch of fuel spheres, a probability of neutron production by fission of the TRISO fuel particles in each batch of fuel spheres, and a fission neutron production cross section of the TRISO fuel particles in each batch of fuel spheres; wherein, obtaining the multi-group neutron flux density of the TRISO fuel particles in each batch of fuel spheres based on the multi-group homogenized group constants of the TRISO fuel particles in each batch of fuel spheres includes: Obtaining the escape probability of the TRISO fuel particles in each batch of fuel spheres according to the macroscopic transport cross-section of the TRISO fuel particles in each batch of fuel spheres and the radius of the TRISO fuel particles in each batch of fuel spheres; Obtaining the penetration probability of the TRISO fuel particles in each batch of fuel spheres according to the macroscopic transport cross-section of the TRISO fuel particles in each batch of fuel spheres and the radius of the TRISO fuel particles in each batch of fuel spheres; Obtaining an average penetration probability of the TRISO fuel particles in each batch of fuel spheres according to the surface area fraction of the TRISO fuel particles in each batch of fuel spheres and the penetration probability of the TRISO fuel particles in each batch of fuel spheres; Obtaining a collision probability between TRISO fuel particles in each batch of fuel spheres according to the escape probability of TRISO fuel particles in each batch of fuel spheres, the penetration probability of TRISO fuel particles in each batch of fuel spheres, the surface area fraction of TRISO fuel particles in each batch of fuel spheres, and the average penetration probability of TRISO fuel particles in each batch of fuel spheres; Establish the second collision probability equations; The second collision probability equations are solved based on the collision probability between TRISO fuel particles in each batch of fuel spheres, the volume fraction of TRISO fuel particles in each batch of fuel spheres, the macroscopic transport cross-section of TRISO fuel particles in each batch of fuel spheres, the macroscopic scattering cross-section of TRISO fuel particles in each batch of fuel spheres, the probability of neutron production by fission of TRISO fuel particles in each batch of fuel spheres, and the fission neutron production cross-section of TRISO fuel particles in each batch of fuel spheres to obtain the multi-group neutron flux density of TRISO fuel particles in each batch of fuel spheres.

5. The method according to claim 1, wherein The multi-group defect factor of the TRISO fuel particles in each batch of fuel spheres is obtained by the following formula: Among them, S TRISO,i,g is the defect factor of the g-th energy group of TRISO fuel particles in the i-th batch of fuel balls, Φ TRISO,i,g is the neutron flux density of the g-th energy group of the TRISO fuel particles in the i-th batch of fuel balls, Φ Pebble,i,g is the neutron flux density of the gth energy group of the i-th batch of fuel balls.

6. The method according to claim 1, characterized in that The method of determining the power of the TRISO fuel particles of the pebble bed high temperature gas-cooled reactor according to the multi-group average neutron flux density of each segment of the pebble bed core, the multi-group defect factor of each batch of fuel spheres, the multi-group defect factor of the TRISO fuel particles in each batch of fuel spheres, and the fission energy generation cross section of the TRISO fuel particles in each batch of fuel spheres comprises: Obtaining a neutron flux distribution at a fuel position in the pebble bed core according to the multi-group average neutron flux density of each segment of the pebble bed core, the multi-group defect factor of each batch of fuel spheres, and the multi-group defect factor of TRISO fuel particles within each batch of fuel spheres; The power of a single TRISO fuel particle in each segment of the pebble bed core is determined based on the neutron flux distribution at the fuel position of the pebble bed core and the fission energy generation cross section of the TRISO fuel particles in each batch of fuel spheres.

7. A detection device for safe operation of a pebble bed high temperature gas-cooled reactor, characterized in that: include: The first acquisition module is used to perform core diffusion calculation for a pebble bed high temperature gas-cooled reactor based on a three-dimensional cylindrical geometry block method to obtain a multi-group average neutron flux density of each block in the pebble bed core; A second acquisition module is configured to obtain multi-group neutron flux densities of each batch of fuel balls in the node according to multi-group homogenization group constants of each batch of fuel balls in the node; a third acquisition module, configured to obtain a multi-group defect factor of each batch of fuel spheres according to the multi-group average neutron flux density of each segment of the pebble bed core and the multi-group neutron flux density of each batch of fuel spheres; a fourth acquisition module, configured to obtain the multi-group neutron flux density of the TRISO fuel particles in each batch of fuel spheres according to the multi-group homogenization group constants of the TRISO fuel particles in each batch of fuel spheres; a fifth acquisition module, configured to obtain a multi-group defect factor of TRISO fuel particles in each batch of fuel spheres based on the multi-group neutron flux density of each batch of fuel spheres and the multi-group neutron flux density of TRISO fuel particles in each batch of fuel spheres; a first determination module, configured to determine the power of a single TRISO fuel particle in each segment of the pebble bed core based on the multi-group average neutron flux density of each segment of the pebble bed core, the multi-group defect factor of each batch of fuel spheres, the multi-group defect factor of the TRISO fuel particles in each batch of fuel spheres, and the fission energy production cross section of the TRISO fuel particles in each batch of fuel spheres; a second determining module, configured to determine the TRISO fuel particle power distribution of the pebble bed core according to the power of a single TRISO fuel particle of each segment of the pebble bed core; The detection module is used to determine whether the pebble bed high temperature gas-cooled reactor is operating safely based on the TRISO fuel particle power distribution of the pebble bed core.

8. The device according to claim 7, characterized in that The multi-group homogenization group constants of each batch of fuel balls include: the macroscopic transport cross section of each batch of fuel balls, the macroscopic scattering cross section of each batch of fuel balls, the probability of neutron production by fission of each batch of fuel balls, and the fission neutron production cross section of each batch of fuel balls; wherein the second acquisition module is specifically used to: Obtaining the escape probability of each batch of fuel balls according to the macroscopic transport cross-section of each batch of fuel balls and the radius of each batch of fuel balls; Obtaining the penetration probability of each batch of fuel balls according to the macroscopic transport cross-section of each batch of fuel balls and the radius of each batch of fuel balls; Obtaining an average penetration probability of each batch of fuel balls according to the surface area fractions of each batch of fuel balls and the penetration probability of each batch of fuel balls; Obtaining collision probabilities between fuel balls in each batch based on the escape probabilities of the fuel balls in each batch, the penetration probabilities of the fuel balls in each batch, the surface area fractions of the fuel balls in each batch, and the average penetration probabilities of the fuel balls in each batch within the node; Establish the first collision probability equations; The first collision probability equations are solved based on the collision probability between the batches of fuel balls, the volume fraction of the batches of fuel balls, the macroscopic transport cross-section of the batches of fuel balls, the macroscopic scattering cross-section of the batches of fuel balls, the probability of neutron production by fission of the batches of fuel balls, and the fission neutron production cross-section of the batches of fuel balls to obtain the multi-group neutron flux density of the batches of fuel balls.

9. The device according to claim 7, characterized in that The multi-group defect factor of each batch of fuel balls is obtained by the following formula: Among them, S Pebble,i,g is the defect factor of the g-th energy group of the i-th batch of fuel balls, Φ Pebble,i,g is the neutron flux density of the gth energy group of the i-th batch of fuel balls, Φ Nodal,g is the multi-group average neutron flux density of each segment of the pebble bed core.

10. The device according to claim 7, characterized in that The multi-group homogenization group constants of the TRISO fuel particles in each batch of fuel spheres include: the macroscopic transport cross section of the TRISO fuel particles in each batch of fuel spheres, the macroscopic scattering cross section of the TRISO fuel particles in each batch of fuel spheres, the probability of neutron production by fission of the TRISO fuel particles in each batch of fuel spheres, and the fission neutron production cross section of the TRISO fuel particles in each batch of fuel spheres; wherein the fourth acquisition module is specifically used to: Obtaining the escape probability of the TRISO fuel particles in each batch of fuel spheres according to the macroscopic transport cross-section of the TRISO fuel particles in each batch of fuel spheres and the radius of the TRISO fuel particles in each batch of fuel spheres; Obtaining the penetration probability of the TRISO fuel particles in each batch of fuel spheres according to the macroscopic transport cross-section of the TRISO fuel particles in each batch of fuel spheres and the radius of the TRISO fuel particles in each batch of fuel spheres; Obtaining an average penetration probability of the TRISO fuel particles in each batch of fuel spheres according to the surface area fraction of the TRISO fuel particles in each batch of fuel spheres and the penetration probability of the TRISO fuel particles in each batch of fuel spheres; Obtaining a collision probability between TRISO fuel particles in each batch of fuel spheres according to the escape probability of TRISO fuel particles in each batch of fuel spheres, the penetration probability of TRISO fuel particles in each batch of fuel spheres, the surface area fraction of TRISO fuel particles in each batch of fuel spheres, and the average penetration probability of TRISO fuel particles in each batch of fuel spheres; Establish the second collision probability equations; The second collision probability equations are solved based on the collision probability between TRISO fuel particles in each batch of fuel spheres, the volume fraction of TRISO fuel particles in each batch of fuel spheres, the macroscopic transport cross-section of TRISO fuel particles in each batch of fuel spheres, the macroscopic scattering cross-section of TRISO fuel particles in each batch of fuel spheres, the probability of neutron production by fission of TRISO fuel particles in each batch of fuel spheres, and the fission neutron production cross-section of TRISO fuel particles in each batch of fuel spheres to obtain the multi-group neutron flux density of TRISO fuel particles in each batch of fuel spheres.

11. The device according to claim 7, characterized in that The multi-group defect factor of the TRISO fuel particles in each batch of fuel spheres is obtained by the following formula: Among them, S TRISo,i,g is the defect factor of the g-th energy group of TRISO fuel particles in the i-th batch of fuel balls, Φ TRISO,i,g is the neutron flux density of the g-th energy group of the TRISO fuel particles in the i-th batch of fuel balls, Φ Pebble,i,g is the neutron flux density of the gth energy group of the i-th batch of fuel balls.

12. The device according to claim 7, characterized in that The determining module is specifically configured to: Obtaining a neutron flux distribution at a fuel position in the pebble bed core according to the multi-group average neutron flux density of each segment of the pebble bed core, the multi-group defect factor of each batch of fuel spheres, and the multi-group defect factor of TRISO fuel particles within each batch of fuel spheres; The power of a single TRISO fuel particle in each segment of the pebble bed core is determined based on the neutron flux distribution at the fuel position of the pebble bed core and the fission energy generation cross section of the TRISO fuel particles in each batch of fuel spheres.

13. An electronic device, characterized in that: include: processor; A memory for storing instructions executable by the processor; wherein the instructions are executed by the processor to enable the processor to perform the method according to any one of claims 1 to 6.

14. A non-transitory computer-readable storage medium, characterized in that When the instructions in the storage medium are executed by a processor of an electronic device, the electronic device is enabled to execute the method according to any one of claims 1 to 6.

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