A method and device for calculating the effective fraction of photoneutron delay

Abstract

This invention discloses a method and apparatus for calculating the effective fraction of photo-delayed neutrons, belonging to the field of nuclear power. The method includes the following steps: providing a multi-group neutron-photon coupled transport database; performing transport calculations on the entire reactor core with a beryllium or heavy water reflector using a deterministic one-step method to obtain the neutron flux distribution, photon flux distribution, and their conjugate fluxes; based on the spatiotemporal dynamics equations of neutron-photon coupling, introducing a quasi-static approximation and the assumption of infinite photon velocity, calculating the average photo-delayed neutron production rate of the reactor core using the neutron flux and its conjugate flux, and calculating the average photon elution rate of the reactor core using the photon flux and its conjugate flux, thus calculating the effective fraction of photo-delayed neutrons. This invention avoids calculation errors caused by subtracting similar values, eliminates the need for coupling between deterministic and Monte Carlo programs, and improves calculation accuracy and efficiency.

Description

Technical Field

[0001] This invention belongs to the field of nuclear power, specifically relating to a method and apparatus for calculating the effective fraction of photo-delayed neutrons. Background Technology

[0002] Photons can react with certain nuclides to produce secondary particles, among which neutrons are called photoneutrons. Photonuclear reactions are threshold energy reactions, and photons in a reactor generally do not possess particularly high energies; typically, only beryllium and deuterium can undergo photonuclear reactions within the reactor core. Photons in the reactor core can be classified as transient photons and delayed photons. Transient photons include those produced by capture reactions and those directly generated during fission, while photons generated during the decay of fission fragments are called delayed photons. Since photonuclear reactions do not have a time effect, the time effect of photoneutrons originates from the photon's generation, and photoneutrons can be further divided into photoinduced transient neutrons and photoinduced delayed neutrons. When the reactor is operating in steady state, the cross-section of the photonuclear reaction is very small, resulting in a limited number of photoneutrons produced, with a negligible impact on the reactor core. However, when reactor operating conditions change, both photo-delayed neutrons and neutrons produced by the decay of fission fragments are delayed neutrons, which will affect the dynamic characteristics of the reactor core. Therefore, the proportion of photo-delayed neutrons, like that of delayed neutrons, is an important dynamic parameter.

[0003] In the research reactor, beryllium and heavy water, which can undergo photonuclear reactions, are usually located in the reflector layer. Due to the extremely small cross-section of the photonuclear reaction, the effect of photo-induced slowed neutron emission on steady-state operation is negligible, but it has a significant impact on dynamic characteristics when operating conditions change.

[0004] Currently, the methods used to calculate the effective fraction of photo-delayed neutrons can be mainly divided into two types: The first method uses the effective multiplication factor to calculate the effective share of photo-delayed neutrons. The main idea is to perform two transport calculations on the reactor core for which the effective share of photo-delayed neutrons needs to be calculated: calculating the effective multiplication factor of the core with and without considering photo-delayed neutrons, and then subtracting the two to give the total share of photo-delayed neutrons. Its main advantage is its simple calculation process and convenient numerical implementation. However, its disadvantage is that the cross-section of the photonuclear reaction is very small, and the effective multiplication factors obtained from the two calculations are very close, leading to a significant error when subtracting two similar values.

[0005] The second method uses a deterministic procedure to calculate the conjugate flux distribution of neutrons, obtaining the neutron value at each location. Then, a Monte Carlo procedure is used to calculate the photon flux distribution. Finally, the effective share of photoinduced delayed neutrons is obtained using the conjugate flux as weights. The advantage of this method is its relatively sound theoretical basis and fewer approximations. However, its disadvantages include a complex computational process, the difficulty of calculating conjugate fluxes using the Monte Carlo procedure, and high computational resource consumption.

[0006] Therefore, providing a method for optimizing the calculation of the effective fraction of photo-induced slowed neutrons by consuming computational resources is of positive significance for improving the accuracy of reactor core dynamics simulation. Summary of the Invention

[0007] The purpose of this invention is to provide a method for calculating the effective fraction of photo-delayed neutrons, optimizing resource consumption and improving calculation accuracy. This invention also provides a computing device for calculating the effective fraction of photo-delayed neutrons.

[0008] According to one embodiment of the present invention, a method for calculating the effective fraction of photo-delayed neutrons is provided, the method comprising the following steps: Step a): Provide a multi-group neutron-photon coupling transport database, which includes a neutron-to-photon scattering matrix, a photonuclear reaction scattering matrix from photon to neutron, and slowed photon grouping data; Step b): Based on the multi-group neutron-photon coupled transport database, a deterministic one-step method is used to perform transport calculations on the entire core of a nuclear reactor with a reflector layer to obtain the neutron flux distribution, photon flux distribution and their conjugate flux of the entire core of the nuclear reactor, wherein the reflector layer is beryllium or heavy water; Step c): Based on the spatiotemporal dynamics equations of neutron-photon coupling, the flux is decomposed into amplitude and shape functions using factor splitting. A quasi-static approximation and the assumption of infinite photon velocity are introduced. The average photo-induced delayed neutron production rate in the reactor core is calculated using the neutron flux distribution and its conjugate flux. The average photon removal rate in the reactor core is calculated using the photon flux distribution and its conjugate flux. Finally, the effective fraction of delayed photons is converted into the effective fraction of photo-induced delayed neutrons through photo-induced neutron reactivity. : , in, For the average slow photons in the reactor core; Reactivity induced by photo-delayed neutron emission; The average photo-induced delayed neutron production rate of the reactor core; The average photon evoked rate of the reactor core.

[0009] This method degenerates the photon equation into an algebraic relationship by assuming the infinite speed of photons, directly deriving the definition of the effective share of photo-delayed neutrons. It uses flux calculation instead of subtracting two similar effective multiplication factors, avoiding calculation errors caused by loss of effective numbers. At the same time, it adopts a deterministic one-step method to obtain the neutron and photon fluxes and conjugate fluxes in a single calculation, eliminating the need for coupling between deterministic and Monte Carlo programs, thus eliminating the process complexity caused by inter-program coupling and achieving high-precision direct calculation of the effective share of photo-delayed neutrons.

[0010] Furthermore, in some embodiments, , , in, For the first of Neutron conjugate angular flux of the group For the first yes The group's photon angular flux shape, The average neutron production rate of the reactor core. For the first yes Group photon conjugate angular flux For the first yes The total cross section of the group's photons The average photon generation rate of the reactor core. For the first yes Group The photon scattering cross section It is a spatial position vector. It is the azimuth vector. For time variables, For the first of neutron angular flux of the group For the first of Group The neutron scattering cross section.

[0011] Furthermore, in some embodiments, the slow-photon grouping data includes the grouping yield and grouping decay constant of slow-photon.

[0012] Furthermore, in some embodiments, the slow-emitting photon grouping data is obtained by: using the fission yield data of nuclides in the nuclear database, calculating the photon decay process of each nuclide through a burnup program and plotting the photon decay curve, and fitting the photon decay curve using a nonlinear least squares method to obtain the grouping yield and grouping decay constant of the slow-emitting photons.

[0013] Furthermore, in some embodiments, the neutron-to-photon scattering matrix is ​​fabricated by utilizing the characteristic that the ratio of the photon generation cross section to the neutron absorption cross section of a nuclide is fixed at a specific temperature.

[0014] Furthermore, in some embodiments, the deterministic one-step method employs a two-dimensional / one-dimensional coupled transport method, which transforms the three-dimensional particle transport equation into a two-dimensional transport equation and a one-dimensional transport equation through transverse integration, solves the two-dimensional transport equation and the one-dimensional transport equation in sequence, and couples them through leakage terms.

[0015] Furthermore, in some embodiments, a solid construction geometry method is used to perform integrated core modeling of the fuel assemblies, control assemblies, radial reflector, axial reflector, and structural materials of the nuclear reactor core.

[0016] Furthermore, in some embodiments, the conjugate flux is obtained by solving the adjoint transport equation, including the neutron conjugate flux of the neutron adjoint transport equation and the photon conjugate flux of the photon adjoint transport equation.

[0017] Furthermore, in some embodiments, the assumption of infinite photon velocity is that the photon velocity is infinite during transport, so that the photon flux reaches a quasi-steady-state equilibrium with the photon source during any neutron transient process, thereby eliminating the time dependence of photons in the spatiotemporal dynamics equation.

[0018] According to another aspect of the present invention, a computing device is provided, the computing device including a processor and a memory, the memory storing a computer program, wherein the processor executes the computer program to implement the photoinduced delayed neutron effective fraction calculation method provided in any of the foregoing embodiments. Attached Figure Description

[0019] Figure 1 This is a schematic diagram of a method for calculating the effective fraction of photo-delayed neutrons in one embodiment; Figure 2 This is a geometric model of the reactor core for studying the heavy water reflector layer in one embodiment.

[0020] Meaning of the reference numerals in the attached figures: 1-Study the reactor core; 2-Heavy water reflector layer.

[0021] The purpose of the above-described drawings is to provide a detailed description of the present invention so that those skilled in the art can understand the technical concept of the invention, and is not intended to limit the invention. For the sake of brevity, the above-described drawings only schematically depict the structures related to the technical features of the present invention, and do not depict the complete structure and all details strictly according to actual scale. Detailed Implementation

[0022] The present invention will now be described in further detail with reference to specific embodiments and accompanying drawings.

[0023] The term "embodiment" as used herein means that a particular feature, structure, or characteristic described in connection with an embodiment may be included in at least one embodiment herein. The phrase appearing in various places in the specification does not necessarily refer to the same embodiment, nor is it limited to mutually exclusive, independent, or alternative embodiments. Those skilled in the art will understand that the embodiments herein can be combined with other embodiments without structural conflict. In the description herein, "a plurality of" means at least two.

[0024] Photons can react with certain nuclides to produce secondary particles, among which neutrons are called photoneutrons. Photonuclear reactions are threshold energy reactions, and photons in a reactor generally do not possess particularly high energies; typically, only beryllium and deuterium can undergo photonuclear reactions within the reactor core. Photons in the reactor core can be classified as transient photons and delayed photons. Transient photons include those produced by capture reactions and those directly generated during fission, while photons generated during the decay of fission fragments are called delayed photons. Since photonuclear reactions do not have a time effect, the time effect of photoneutrons originates from the photon's generation, and photoneutrons can be further divided into photoinduced transient neutrons and photoinduced delayed neutrons. When the reactor is operating in steady state, the cross-section of the photonuclear reaction is very small, resulting in a limited number of photoneutrons produced, with a negligible impact on the reactor core. However, when reactor operating conditions change, both photo-delayed neutrons and neutrons produced by the decay of fission fragments are delayed neutrons, which will affect the dynamic characteristics of the reactor core. Therefore, the proportion of photo-delayed neutrons, like that of delayed neutrons, is an important dynamic parameter.

[0025] In existing technologies, methods for calculating the effective fraction of photo-delayed neutrons using the effective multiplication factor suffer from significant errors due to the small cross-section of the photonuclear reaction. The effective multiplication factors obtained from two calculations are very close, and subtracting these two similar values ​​introduces a large error. Furthermore, methods using deterministic procedures to calculate the conjugate flux distribution of neutrons are computationally complex, particularly the Monte Carlo method, which struggles with calculating conjugate fluxes and consumes substantial computational resources. All existing methods fall short of meeting the requirements for simulating photo-delayed neutrons.

[0026] To overcome the aforementioned problems of the prior art, one embodiment of the present invention provides a method for calculating the effective fraction of photo-delayed neutrons, which can effectively improve calculation accuracy and reduce the consumption of computing resources.

[0027] Specifically, the method flow is as follows: Figure 1 As shown, it includes the following steps: Step a): Provide a multi-group neutron-photon coupling transport database. In different embodiments, an existing database can be used, or a database including neutron-to-photon scattering matrices, photonuclear reaction scattering matrices of photons to neutrons, and slow-release photon grouping data can be created based on existing data.

[0028] In a reactor, photons originate from various sources, including: radiative trapping reactions, where the target nucleus absorbs neutrons to form a composite nucleus, which, after being excited, returns to the ground state by emitting photons; inelastic scattering, (n, p) reactions, and (n, α) reactions, where the target nucleus absorbs neutrons to form a composite nucleus, emits other particles while remaining in an excited state, and then returns to the ground state by emitting photons; and photons directly emitted by fission reactions. In neutron-photon coupled transport calculations, the cross-sectional data of photon generation is represented by the scattering matrix from the neutron energy group to the photon energy group. In reactors, fission nuclides typically exhibit spatial self-shielding effects, and their cross-sections are usually not fixed. However, for a nuclide at a specific temperature, regardless of the presence of resonant self-shielding effects, the ratio of its photon generation cross-section to its neutron absorption cross-section is fixed.

[0029] The scattering cross section of a photon is usually associated with the interaction between the photon and an atom. The basic forms of light-atom interaction include coherent scattering, electron-pair reaction, photoelectric effect, and Compton scattering. The photoelectric effect is an absorption reaction and does not produce secondary photons; coherent scattering is elastic scattering, where the photon energy remains unchanged before and after the collision; electron-pair reaction produces two photons, each with an energy of 0.511 MeV, which can be considered an (γ, 2γ) reaction. When the energy of the incident photon exceeds the separation energy of a nucleus, a photonuclear reaction occurs. The photonuclear reaction is a threshold energy reaction; extremely high-energy photons do not exist in a reactor, and typically only nuclides are present. 9 Be and 2 H (usually derived from heavy water) can undergo photonuclear reactions. In a photonuclear reaction, a photon is first absorbed by the target nucleus, causing it to transition to an excited state. Subsequently, the target nucleus de-excites by emitting one or more secondary particles (including neutrons and photons). The neutron production cross section of a photonuclear reaction can also be represented as a scattering matrix from the photon energy group to the neutron energy group.

[0030] Calculating the effective fraction of photoinduced delayed neutrons in the reactor core requires data on delayed neutrons and photons from the core fission material. Delayed neutron data from the fission material can be obtained from databases, for example, some databases provide it directly in 6 or 8 groups. However, publicly available databases rarely provide data on delayed photons. Therefore, in some embodiments, delayed photon data can be obtained by performing calculations based on the delayed neutron data from the fission material. Specifically, in actual fission fragments, there may be hundreds or thousands of precursor nuclei for delayed photons, and these may also include secondary decays. Obviously, fission product data cannot be directly applied to core calculations, so a method that can concisely and accurately describe delayed photon data is needed. In conventional publicly available databases, delayed neutrons are usually grouped into 6 or 8 groups according to their decay constants, which is much smaller than the number of groups of decaying neutrons in fission fragments during actual fission. Therefore, the yield and decay constant of each group are the result of the recombination of multiple precursor nuclei in actual fission. A small amount of fissile material was irradiated in an experimental setup, and the number of neutrons emitted by the fissile material after irradiation was determined using a counter over time. Finally, decay curves were plotted based on the curves showing the number of emitted neutrons over time, and a nonlinear least-squares method was used for fitting, ultimately providing the delayed neutron groupings and related data for several fissile nuclides. For delayed photon data, the groupings and related data could be obtained using a similar method to that used for delayed neutron data. Since the fission yield and photon decay data for each nuclide are relatively complete in current nuclear databases, irradiation experiments are unnecessary. The fission yield of the nuclide can be used from the nuclear database, followed by a burnup procedure to determine the photon decay process, plotting the photon decay curves for each nuclide, and finally fitting using a nonlinear least-squares method to obtain the groupings and related data for the decayed photons. In other embodiments, experimental data can also be directly used to obtain the relevant data for the grouped decayed photons.

[0031] Step b): Based on the multi-group neutron-photon coupled transport database, the transport calculation of the entire core of a nuclear reactor with a beryllium or heavy water reflector is performed using a deterministic one-step method to obtain the neutron flux distribution, photon flux distribution and conjugate flux of the entire core of the nuclear reactor.

[0032] The deterministic one-step method is based on massive parallel computing technology and performs integrated modeling and calculation of the entire core directly, completing particle transport calculations across the entire core.

[0033] Specifically, firstly, a deterministic one-step high-precision model of the entire reactor core is performed, using a solid construction geometry method to establish a complete model of the reactor. High-precision modeling is then performed on the fuel assemblies, control assemblies, radial and axial reflectors, structural materials, etc., of the reactor core. In a specific embodiment, the complete model of the reactor core is as follows: Figure 2As shown, this includes studying the reactor core 1 and the heavy water reflector layer 2. Next, based on the multi-group neutron-photon coupled transport database obtained in step a), the three-dimensional particle transport equation is transformed into two-dimensional and one-dimensional equations through transverse integration. The two-dimensional and one-dimensional equations are solved sequentially, and the two are coupled through a leakage term to obtain the particle flux distribution. Deterministic one-step transport calculations are then performed on the entire reactor core to obtain the neutron and photon flux distribution of the entire reactor core.

[0034] Step c): Based on a portion of the transport calculation results, further calculate the effective fraction of photo-delayed neutrons.

[0035] Specifically, according to the definition of photo-delayed neutrons, we can obtain: , in, For photo-delayed neutron fraction, For slow photon flux, It is a spatial position vector. E For energy variables, For the cross section of photonuclear reaction, Photon flux Neutron flux The number of neutrons produced in each fission. The cross section is the fission section.

[0036] In a real reactor, beryllium and heavy water, which can undergo photonuclear reactions, are typically located in the reflector layer. However, due to the presence of structural materials and coolants, photo-delayed neutrons generated in the reflector layer also reside there, resulting in a spatial effect. Therefore, the concept of the effective fraction of photo-delayed neutrons is introduced. The effective fraction of photo-delayed neutrons takes into account the spatial effect and can be used to truly measure the contribution of photo-delayed neutrons to the reactor core.

[0037] The neutron-related part of the spacetime dynamics equations for neutron-photon coupling is: , in, It is the azimuth vector. For time variables, For the first of Group neutron velocity, For the first of neutron angular flux of the group For the first of The total neutron cross section of the group, For the first of Group Neutron scattering cross section For the first of The instantaneous neutron fission energy spectrum of the group, To delay the issuance of the total neutron quota, This represents the number of delayed-emission neutron precursor nuclei during the fission process. For the first of Energy Group i Group of slow-emission neutron fission spectra For the first i Group of slow-emitting neutron precursor nuclear decay constants For the first i Concentration of slow-emitting neutron precursor nuclei. For the first of Energy group transient neutron fission spectrum, For the first The fission cross section of the group To delay the issuance of neutron quotas, For the first yes Group photon angular flux, For the first yes Group The cross section of the light nucleus.

[0038] Solving the equations typically employs a quasi-static method. First, the flux in the equations is decomposed into a strongly time-dependent amplitude function and a weakly time-dependent shape function using the concept of factorization. Based on this, a quasi-static approximation is introduced. Similarly, the same assumptions can be applied to photons. The decomposition of the equations is not unique, requiring the introduction of normalization conditions. For the neutron shape function, the conjugate flux with the neutron is conserved over the integrals of space, angle, and energy. For the photon shape function, the conjugate flux with the photon is also conserved over the integrals of space, angle, and energy.

[0039] Based on the quasi-static assumption, neutron conjugate flux and photon conjugate flux are introduced as weights, that is, the two ends of the simplified neutron-photon coupling spacetime dynamics are multiplied by the conjugate flux of the corresponding particles. Partial core-averaged parameters are also defined, and the precise point-pile dynamics equations for neutron-photon coupling are finally obtained as follows: , , in, This represents the core neutron flux amplitude. For core reactivity, This represents the average total proportion of delayed neutron emission in the reactor core. For the average first core i Group of slow-emitting neutron precursor nuclear decay constants For the first i Concentration of slow-emitting neutron precursor nuclei. This represents the average lifetime of neutrons in the reactor core. For delayed neutron group identification, , The average photo-induced delayed neutron production rate of the reactor core. This represents the core photon flux amplitude. This represents the average total proportion of slow-release photons in the reactor core. This represents the average lifetime of neutrons in the reactor core. The average photon evoked rate of the reactor core. For slow-emitting electron grouping identifiers, , For the average first core j Group of slow-photon precursor nuclear decay constants For the first i Concentration of slow-emitting neutron precursor nuclei.

[0040] Next, based on the assumption of infinite photon velocity, the definition of photo-delayed neutron is derived. The aforementioned exact point-pile equations are non-autonomous, and the neutron and photon terms are strongly coupled, making direct solution complex. Furthermore, it cannot directly define the effective fraction of photo-delayed neutrons, so further simplification of the equations is necessary. To address this issue, the assumption of infinite photon velocity is introduced. This assumption states that the speed of photons during transport is infinite, i.e., in the spacetime dynamics equations... The physical implication of this assumption is that when the photon source changes, photons are transported instantaneously, with the transport time approaching zero. The photon flux follows a dynamic equilibrium of source, removal, and leakage, thus the photon flux in the reactor core remains in a quasi-steady state. This assumption eliminates the photon dependence on events in the equations, reducing the strong coupling to some extent. Since the speed of photons in the reactor core is typically one or several orders of magnitude higher than that of neutrons, the photon flux almost immediately reaches equilibrium with the photon source during any reactor transient. Therefore, this assumption can be considered to hold true in the neutron-photon spacetime dynamics equations.

[0041] The equations after introducing approximations are essentially identical in form to the traditional point pile dynamics equations that do not consider photons; the only difference lies in the number of delayed neutron groups. i The group became i+j Group, and this extra j The delayed neutrons are precisely the photo-induced delayed neutrons produced by delayed photons. Therefore, based on the above derivation, we can directly... Defined as the effective share of photo-delayed neutrons.

[0042] The specific calculation method is as follows: , in, For the average slow photons in the reactor core, Reactivity induced by photo-delayed neutrons The average photo-induced delayed neutron production rate of the reactor core. The average photon evoked rate of the reactor core.

[0043] The average photo-induced delayed neutron production rate in the reactor core is defined as follows: , in, For the first of Neutron conjugate angular flux of the group For the first yes The group's photon angular flux shape, This represents the average neutron production rate of the reactor core.

[0044] The average photon removal rate of the core is defined as follows: , in, For the first yes Group photon conjugate angular flux For the first yes The total cross section of the group's photons The average photon generation rate of the reactor core. For the first yes Group The photon scattering cross section.

[0045] In the above process, apart from the quasi-static approximation, the only introduced approximation is the assumption of infinite speed of photons. In reality, the approximation of this assumption is also very small, so the above derivation process is valid. The effective share of photo-delayed neutrons is mainly achieved by introducing photo-delayed neutron reactivity, which can directly reflect the contribution of photon reactor reactivity. Based on this, the effective share of delayed photons is transformed into the effective share of photo-delayed neutrons, and the delayed photon precursor nucleus is transformed into the photo-delayed neutron precursor nucleus.

[0046] The method for calculating the effective fraction of photoinduced delayed neutrons provided in the above embodiments uses flux for calculation. Compared to methods using effective multiplication factors, it eliminates the need to subtract similar numbers, effectively reducing calculation errors. Compared to methods that use deterministic procedures to calculate the conjugate flux distribution of neutrons and then use Monte Carlo procedures to calculate the photon flux distribution, the calculation method provided in the above embodiments does not require coupling between different procedures, has simpler program settings, does not require the intervention of Monte Carlo methods, and has higher computational efficiency.

[0047] The above method can be implemented by a computing device provided in another aspect of the present invention. This computing device includes a processor and a memory, in which a calculation program is stored. When the calculation program is executed by the processor, the computing device can implement the photo-induced delayed neutron effective fraction calculation method provided in the foregoing embodiments. In different embodiments, the computing device can take different forms; it can be a general-purpose computer, a dedicated computing device, a virtual machine, or a cloud computing device.

[0048] The purpose of the above embodiments is to provide a further detailed description of the present invention in conjunction with the accompanying drawings, so that those skilled in the art can understand the technical concept of the present invention. Within the scope of the present invention, optimization or equivalent substitution of the technical features involved, as well as combination of implementation methods in different embodiments without causing structural and principle conflicts, all fall within the protection scope of the present invention.

Claims

1. A method for calculating the effective fraction of photo-delayed neutrons, characterized in that, include: Step a): Provide a multi-group neutron-photon coupling transport database, which includes a neutron-to-photon scattering matrix, a photonuclear reaction scattering matrix from photon to neutron, and slowed photon grouping data; Step b): Based on the multi-group neutron-photon coupled transport database, a deterministic one-step method is used to perform transport calculations on the entire core of a nuclear reactor with a reflector layer to obtain the neutron flux distribution, photon flux distribution and their conjugate flux of the entire core of the nuclear reactor, wherein the reflector layer is beryllium or heavy water; Step c): Based on the spatiotemporal dynamics equations of neutron-photon coupling, the flux is decomposed into amplitude and shape functions using factor splitting. A quasi-static approximation and the assumption of infinite photon velocity are introduced. The average photo-induced delayed neutron production rate in the reactor core is calculated using the neutron flux distribution and its conjugate flux. The average photon removal rate in the reactor core is calculated using the photon flux distribution and its conjugate flux. Finally, the effective fraction of delayed photons is converted into the effective fraction of photo-induced delayed neutrons through photo-induced neutron reactivity. : ， in, For the average slow photons in the reactor core; Reactivity induced by photo-delayed neutron emission; The average photo-induced delayed neutron production rate of the reactor core; The average photon evoked rate of the reactor core.

2. The method for calculating the effective fraction of photo-delayed neutrons according to claim 1, characterized in that, ， ， in, For the first ng Neutron conjugate angular flux of the group For the first pg The group's photon angular flux shape, The average neutron production rate of the reactor core. For the first pg Group photon conjugate angular flux For the first pg The total cross section of the group's photons The average photon generation rate of the reactor core. For the first pg Group The photon scattering cross section It is a spatial position vector. It is the azimuth vector. For time variables, For the first ng neutron angular flux of the group For the first ng Group The neutron scattering cross section.

3. The method for calculating the effective fraction of photo-delayed neutrons according to claim 1, characterized in that, The slow-luminescent electron grouping data includes the grouping yield and grouping decay constant of slow-luminescent electrons.

4. The method for calculating the effective fraction of photo-delayed neutrons according to claim 1, characterized in that, The slow-emitting photon grouping data is obtained by using the fission yield data of nuclides in the nuclear database, calculating the photon decay process of each nuclide through a burnup program and plotting the photon decay curve, and fitting the photon decay curve using the nonlinear least squares method to obtain the grouping yield and grouping decay constant of the slow-emitting photons.

5. The method for calculating the effective fraction of photo-delayed neutrons according to claim 1, characterized in that, The neutron-to-photon scattering matrix is ​​constructed by utilizing the property that the ratio of the photon generation cross section to the neutron absorption cross section of a nuclide is fixed at a specific temperature.

6. The method for calculating the effective fraction of photo-delayed neutrons according to claim 1, characterized in that, The deterministic one-step method employs a two-dimensional / one-dimensional coupled transport method, which transforms the three-dimensional particle transport equation into a two-dimensional transport equation and a one-dimensional transport equation through transverse integration. The two-dimensional transport equation and the one-dimensional transport equation are solved sequentially, and coupling is achieved through leakage terms.

7. The method for calculating the effective fraction of photo-delayed neutrons according to claim 6, characterized in that, A solid construction geometry method is used to perform integrated core modeling of the nuclear reactor core, including fuel assemblies, control assemblies, radial reflector, axial reflector, and structural materials.

8. The method for calculating the effective fraction of photo-delayed neutrons according to claim 1, characterized in that, The conjugate flux is obtained by solving the adjoint transport equation, including the neutron conjugate flux of the neutron adjoint transport equation and the photon conjugate flux of the photon adjoint transport equation.

9. The method for calculating the effective fraction of photo-delayed neutrons according to claim 1, characterized in that, The assumption of infinite photon velocity is that the photon velocity is infinite during transport, so that the photon flux reaches a quasi-steady-state equilibrium with the photon source during any neutron transient process, thereby eliminating the time dependence of photons in the spatiotemporal dynamics equation.

10. A computing device, characterized in that, It includes a processor and a memory, wherein the memory stores a computer program, and the processor executes the computer program to implement the photoinduced delayed neutron effective fraction calculation method as described in any one of claims 1 to 9.

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