Method for correcting protection setting value of nuclear reactor and computer readable storage medium

Abstract

The invention discloses a nuclear reactor protection setting value correction method and a computer readable storage medium. The method comprises the following steps: S1, establishing a nuclear reactor model and an intermediate range detector model; s2, a fuel assembly in a reactor core of the nuclear reactor model is selected for neutron transport simulation calculation, and an intermediate range response factor obtained after fission neutrons of the fuel assembly are transmitted to the position of the intermediate range detector model from a reactor core active area is obtained; s3, calculating an intermediate range theoretical current value at the beginning of the life period and an intermediate range theoretical current value at the current burnup point according to the intermediate range response factor, and obtaining a correction coefficient at the current burnup point; and S4, according to the correction coefficient under the current burn-up point, correcting the protection constant value of the middle range set at the beginning of the life period to obtain the protection constant value of the middle range under the current burn-up point. The method guarantees the accuracy of the protection constant value of the nuclear reactor, reduces the risk of mistaken shutdown of the nuclear reactor and the risk of protection failure of the reactor, and gives consideration to the safety of a unit.

Description

Technical Field

[0001] The present invention relates to the field of nuclear power technology, and in particular to a method for correcting a nuclear reactor protection constant and a computer-readable storage medium. Background Art

[0002] The nuclear instrument system of a nuclear power unit has three range detectors (power range, intermediate range and source range). Among them, the intermediate range (IRC: Intermidiate Range Channel) detector uses a gamma-boron compensated ionization chamber to measure the neutron flux signal of the nuclear reactor. The intermediate range detector of the nuclear instrument system measures the neutron flux leaking from the core to the outside of the reactor and converts it into a current value for display. The nuclear instrument system sets the following three protection signals according to the current value detected by the intermediate range: lockout and lifting power control rod protection signal (C1), emergency shutdown signal (RT) and expected transient protection signal (ATWT: Anticipated Transients without Trip) that failed to shut down the reactor.

[0003] The range of the mid-range detector is 10 -11 A to 10 -3 A, the intermediate range has reached the upper limit of the range before the unit reaches full power. The current setting values ​​corresponding to the three protection signals are usually calculated and calibrated at the 30% FP rated power level at the beginning of each cycle life. The intermediate range protection setting current value is in the order of 10 -4 The rated powers of A, C1, RT and ATWT are 20%FP, 25%FP and 30%FP respectively.

[0004] The intermediate range current value has reached the upper limit of the range before the unit is increased to full power. Because the current is saturated, the current value of the protection setting cannot be calibrated again during the normal operation of the unit. As the fuel consumption increases, the core power distribution, injection rate level, and neutron energy spectrum all change, causing the neutron injection rate transmitted from the core to the intermediate range position to change, which in turn affects the intermediate range current value, and the power level corresponding to the intermediate range setting will change. However, because the intermediate range has exceeded the upper limit of the range when operating at full power, the protection setting cannot be calibrated and calculated. The protection setting of the intermediate range set at the beginning of the life cycle is no longer applicable, and the reactor is at risk of accidental shutdown and reactor protection failure, affecting the safe operation of the unit. Summary of the invention

[0005] The technical problem to be solved by the present invention is to address at least one defect of the related technology mentioned in the above background technology: as the fuel consumption increases, the protection setting of the intermediate range set at the beginning of the life cycle is no longer applicable, and the reactor is at risk of accidental shutdown and reactor protection failure, affecting the safe operation of the unit.

[0006] The technical solution adopted by the present invention to solve the technical problem is: to provide a method for correcting a nuclear reactor protection setting, comprising the following steps:

[0007] S1: Establish a nuclear reactor model and an intermediate range detector model;

[0008] S2: Selecting a fuel assembly in the core of the nuclear reactor model to perform neutron transport simulation calculation, and obtaining an intermediate range response factor after the fission neutrons of the fuel assembly are transmitted from the core active region to the position of the intermediate range detector model;

[0009] S3: Calculate the intermediate range theoretical current value at the beginning of the life and the intermediate range theoretical current value at the current burnup point according to the intermediate range response factor, and obtain the correction coefficient at the current burnup point;

[0010] S4: Correcting the protection constant value of the intermediate range set at the beginning of the life cycle according to the correction coefficient at the current burnup point to obtain the protection constant value of the intermediate range at the current burnup point.

[0011] In some embodiments, step S2 includes:

[0012] S21: selecting i fuel assemblies on the periphery of the core of the nuclear reactor model, and dividing the i fuel assemblies into j segments in their axial heights, wherein i and j ≥ 1;

[0013] S22: performing neutron transport simulation calculation on the j-th segment of the i-th fuel assembly to obtain the reaction rate of the fission neutrons of the j-th segment of the i-th fuel assembly after being transmitted from the core active region to the position of the intermediate range detection model;

[0014] S23: performing normalized calculation on the reaction rate of the j-th segment of the i-th fuel assembly to obtain the intermediate range response factor of the j-th segment of the i-th fuel assembly.

[0015] In some embodiments, step S21 includes:

[0016] i fuel assemblies of the intermediate range detector model located at the periphery of the core of the nuclear reactor model and close to the 270° or 90° position are selected, and the i fuel assemblies are respectively divided into j segments in their axial height.

[0017] In some embodiments, step S22 includes:

[0018] The nuclear reactor model simulates the state of the jth segment of the i-th fuel assembly to obtain neutronics parameters;

[0019] Processing the neutronic parameters, constructing a core source term for Monte Carlo transport calculation, and simulating the neutron injection rate after the fission neutrons of the j-th segment of the i-th fuel assembly are transmitted from the core active region to the position of the intermediate range detection model;

[0020] According to the neutron injection rate and the coating sensitive material of the intermediate range detector model, the reaction rate of the fission neutrons of the j-th segment of the i-th fuel assembly after being transmitted from the core active area to the position of the intermediate range detection model is obtained.

[0021] In some embodiments, the intermediate-range theoretical current value of the intermediate-range detector model per unit time is proportional to the reaction rate of the fission neutrons of the node after being transmitted from the core active region to the position of the intermediate-range detection model.

[0022] In some embodiments, step S23 includes:

[0023] The reaction rates of all nodes of the i-th fuel assembly are summed to obtain the sum of the reaction rates of all nodes of the i-th fuel assembly;

[0024] The intermediate range response factor of the j-th section of the i-th fuel assembly is obtained by dividing the sum of the reactivity of all the sections of the i-th fuel assembly by the reactivity of the j-th section of the i-th fuel assembly.

[0025] In some embodiments, step S3 includes:

[0026] S31: calculating the initial life intermediate range theoretical current value of the j-th segment of the i-th fuel assembly according to the intermediate range response factor of the j-th segment of the i-th fuel assembly and the initial life power of the j-th segment of the i-th fuel assembly;

[0027] S32: calculating the intermediate range theoretical current value of the j-th segment of the i-th fuel assembly at the current burnup point according to the intermediate range response factor of the j-th segment of the i-th fuel assembly and the current burnup point power of the j-th segment of the i-th fuel assembly;

[0028] S33: Repeat step S31 until i fuel assemblies are traversed, sum up the theoretical current values ​​of the initial intermediate range of the life of all the j-th segments of the i-th fuel assembly, and obtain the theoretical current values ​​of the initial intermediate range of the life of all the segments of all the fuel assemblies;

[0029] S34: Repeat step S32 until all i fuel assemblies are traversed, and sum the intermediate range theoretical current values ​​of the j-th segment of all the i-th fuel assemblies at the current burnup point to obtain the intermediate range theoretical current values ​​of all the segments of all the fuel assemblies at the current burnup point;

[0030] S35: Divide the intermediate-range theoretical current value at the current burnup point of all the nodes of all the fuel assemblies by the intermediate-range theoretical current value at the beginning of the life of all the nodes of all the fuel assemblies to obtain the correction coefficient at the current burnup point.

[0031] In some embodiments, the initial life power of the jth segment of the i-th fuel assembly is: the initial life power of the jth segment of the i-th fuel assembly at the beginning of the simulated life;

[0032] The current burnup point power of the j-th segment of the i-th fuel assembly is: the current burnup point power of the j-th segment of the i-th fuel assembly under the simulated current burnup point.

[0033] In some embodiments, step S4 includes:

[0034] The protection set value of the intermediate range at the current burnup point is obtained by multiplying the correction coefficient at the current burnup point by the protection set value of the intermediate range set at the beginning of the life cycle.

[0035] The present invention also constructs a computer-readable storage medium on which a computer program is stored. When the computer program is executed by a processor, the method for correcting the reactor protection constant as described in any one of the above items is implemented.

[0036] By implementing the present invention, the following beneficial effects are achieved:

[0037] The present invention establishes a nuclear reactor model and an intermediate-range detector model, selects a fuel assembly in the core of the nuclear reactor model to perform neutron transport simulation calculation, obtains an intermediate-range response factor after the fission neutrons of the fuel assembly are transmitted from the core active region to the intermediate-range detector model position, then calculates the intermediate-range theoretical current value at the beginning of the life cycle and the intermediate-range theoretical current value at the current burnup point according to the intermediate-range response factor, and obtains a correction coefficient at the current burnup point, and finally corrects the intermediate-range protection setting set at the beginning of the life cycle according to the correction coefficient at the current burnup point to obtain the intermediate-range protection setting at the current burnup point, thereby realizing the correction of the intermediate-range protection setting set at the beginning of the life cycle, ensuring that the protection setting set at the beginning of the life cycle is applicable after theoretical correction, ensuring the accuracy of the nuclear reactor protection setting, reducing the risk of false shutdown of the nuclear reactor and the risk of reactor protection failure, and taking into account both the safety and economic benefits of the unit. BRIEF DESCRIPTION OF THE DRAWINGS

[0038] The present invention will be further described below with reference to the accompanying drawings and embodiments, in which:

[0039] Figure 1 It is a schematic flow chart of an embodiment of a method for correcting a nuclear reactor protection setting value of the present invention;

[0040] Figure 2 1 is a flow chart of step S2 of an embodiment of a method for correcting a nuclear reactor protection setting value of the present invention;

[0041] Figure 3 It is a schematic diagram of the planar arrangement of all fuel assemblies in a nuclear reactor model in one embodiment of the method for correcting the protection setting of a nuclear reactor of the present invention;

[0042] Figure 4 It is a schematic diagram of i fuel assemblies of an intermediate range detector model located at the periphery of a nuclear reactor model core and close to a position of 270° or 90° in an embodiment of a method for correcting a nuclear reactor protection setting value of the present invention;

[0043] Figure 5 It is a schematic diagram of a fuel assembly divided into j segments in its axial height in one embodiment of a method for correcting a nuclear reactor protection setting value of the present invention;

[0044] Figure 6 It is a flow chart of step S3 of an embodiment of the method for correcting the protection setting of a nuclear reactor of the present invention. DETAILED DESCRIPTION

[0045] In order to have a clearer understanding of the technical features, purposes and effects of the present invention, specific embodiments of the present invention are now described in detail with reference to the accompanying drawings.

[0046] It should be noted that the flowcharts shown in the accompanying drawings are only exemplary and do not necessarily include all the contents and operations / steps, nor do they have to be executed in the order described. For example, some operations / steps can be decomposed, while some operations / steps can be combined or partially combined, so the actual execution order may change according to actual conditions.

[0047] The block diagrams shown in the accompanying drawings are merely functional entities and do not necessarily correspond to physically independent entities. That is, these functional entities may be implemented in software form, or in one or more hardware modules or integrated circuits, or in different networks and / or processor devices and / or microcontroller devices.

[0048] like Figure 1 As shown, some embodiments of the present invention disclose a method for correcting a nuclear reactor protection setting, comprising the following steps:

[0049] S1: Establish a nuclear reactor model and an intermediate range detector model;

[0050] S2: Select the fuel assembly in the core of the nuclear reactor model to perform neutron transport simulation calculation, and obtain the intermediate range response factor after the fission neutrons of the fuel assembly are transmitted from the core active area to the position of the intermediate range detector model;

[0051] S3: Calculate the theoretical current value of the intermediate range at the beginning of the life span and the theoretical current value of the intermediate range at the current burnup point according to the intermediate range response factor, and obtain the correction coefficient at the current burnup point;

[0052] S4: The protection constant of the intermediate range set at the beginning of the life cycle is corrected according to the correction coefficient at the current burnup point to obtain the protection constant of the intermediate range at the current burnup point.

[0053] This embodiment can realize the correction of the protection setting of the intermediate range set at the beginning of the service life, ensure that the protection setting set at the beginning of the service life is applicable after theoretical correction, ensure the accuracy of the nuclear reactor protection setting, reduce the risk of accidental shutdown of the nuclear reactor and the risk of reactor protection failure, and take into account the safety and economic benefits of the unit.

[0054] In some embodiments, Figure 2 As shown, step S2 includes:

[0055] S21: select i fuel assemblies on the periphery of the core of the nuclear reactor model, and divide the i fuel assemblies into j segments in their axial heights, where i and j ≥ 1;

[0056] S22: performing neutron transport simulation calculation on the j-th segment of the i-th fuel assembly to obtain the reaction rate of the fission neutrons of the j-th segment of the i-th fuel assembly after being transmitted from the core active region to the position of the intermediate range detection model;

[0057] S23: performing normalized calculation on the reaction rate of the j-th segment of the i-th fuel assembly to obtain the mid-range response factor of the j-th segment of the i-th fuel assembly.

[0058] In some embodiments, the planar layout of all fuel assemblies in the nuclear reactor model is as follows: Figure 3 As shown, step S21 includes:

[0059] Select i fuel assemblies of the intermediate range detector model located outside the core of the nuclear reactor model and close to the 270° or 90° position, that is, Figure 4 The gray squares shown in the figure divide the i fuel assembly into the following categories in terms of its axial height: Figure 5 The j segment block is shown.

[0060] Among them, the nuclear reactor structure in the nuclear reactor model is, from the core to the outside, the enclosure, radial plate, radial reflector water, basket, heat shield, downwelling water layer, RPV, cavity, insulation layer, cavity and concrete shielding wall. The main components of the top reflector include the fuel rod spring segment and upper end plug, top grid, upper tube seat and coolant. The main components of the bottom reflector include the fuel rod lower end plug, lower tube seat, lower grid plate and coolant. The off-core detector is located in the cavity between the insulation layer and the concrete shielding wall. Considering the symmetry of the core layout, only the intermediate range detector at the 270° or 90° position is simulated in the calculation.

[0061] In some embodiments, step S22 includes:

[0062] The nuclear reactor model simulates the state of the jth segment of the i-th fuel assembly to obtain neutronics parameters;

[0063] The neutronic parameters are processed to construct the core source term for Monte Carlo transport calculation, simulating the neutron injection rate after the fission neutrons of the jth segment of the i-th fuel assembly are transmitted from the core active area to the position of the intermediate range detection model;

[0064] According to the neutron injection rate and the coating sensitive material of the intermediate range detector model, the reaction rate of the fission neutrons of the jth segment of the i-th fuel assembly after being transmitted from the core active area to the position of the intermediate range detection model is obtained.

[0065] Among them, the reaction rate description is: for the fuel assembly arranged in the core, the response relationship between the current burnup point power at the fuel assembly position and the intermediate range theoretical current value under the current burnup point detected by the intermediate range detector model. In the nuclear reactor model, a uniform unit neutron source is defined in the grid element where the core fuel assembly block is located, and the neutron injection rate after the fission neutron is transported to the intermediate range detector model position is simulated and calculated.

[0066] This embodiment uses the Monte Carlo method program to establish a nuclear reactor model for neutron transport simulation calculation. First, a nuclear reactor model is established based on the SCIENCE refueling design software to simulate the core state and give the neutron parameters; then, the neutron parameters are processed to construct the core source term for Monte Carlo method calculation, simulating the neutron injection rate after the fission neutron is transmitted from the core active area of ​​the nuclear reactor model to the position of the intermediate range detector model. Afterwards, the intermediate range theoretical current value at the current burnup point is calculated by the reaction rate of the fission neutron and the coating sensitive material B-10 of the intermediate range detector model.

[0067] In some embodiments, the intermediate-range theoretical current value of the intermediate-range detector model per unit time is proportional to the reaction rate of the fission neutrons of the nodule after being transmitted from the core active region to the position of the intermediate-range detection model.

[0068] In some embodiments, step S23 includes:

[0069] The reaction rates of all nodes of the i-th fuel assembly are summed to obtain the sum of the reaction rates of all nodes of the i-th fuel assembly;

[0070] The intermediate range response factor of the j-th segment of the i-th fuel assembly is obtained by dividing the sum of the reactivity of all the segments of the i-th fuel assembly by the reactivity of the j-th segment of the i-th fuel assembly.

[0071] That is, according to formula (1): Calculate the formula (1), i represents the i-th fuel assembly, j represents the j-th segment of a fuel assembly, R i,j represents the reaction rate of the jth segment of the i-th fuel assembly, ∑ i, j R i,j represents the sum of the reaction rates of all nodes of the i-th fuel assembly, A i,j It represents the mid-range response factor of the j-th segment of the i-th fuel assembly.

[0072] In some embodiments, Figure 6 As shown, step S3 includes:

[0073] S31: calculating the initial life intermediate range theoretical current value of the j-th segment of the i-th fuel assembly according to the intermediate range response factor of the j-th segment of the i-th fuel assembly and the initial life power of the j-th segment of the i-th fuel assembly;

[0074] S32: calculating the intermediate range theoretical current value of the j-th segment of the i-th fuel assembly at the current burnup point according to the intermediate range response factor of the j-th segment of the i-th fuel assembly and the current burnup point power of the j-th segment of the i-th fuel assembly;

[0075] Among them, the intermediate range theoretical current value at the current burnup point detected by the intermediate range detector model is simulated by the Monte Carlo program. In the calculation, the intermediate range theoretical current value at the current burnup point of each fuel assembly at unit power is calculated by the Monte Carlo program. The intermediate range theoretical current value is proportional to the B-10 reaction rate of neutrons and the sensitive material coating of the intermediate range detector model.

[0076] S33: Repeat step S31 until all i fuel assemblies are traversed, and sum the theoretical current values ​​of the initial intermediate range of the life of all the j-th segments of the i-th fuel assembly to obtain the theoretical current values ​​of the initial intermediate range of the life of all the segments of all the fuel assemblies;

[0077] S34: Repeat step S32 until all i fuel assemblies are traversed, and sum the intermediate range theoretical current values ​​of the j-th segment of all the i-th fuel assemblies at the current burnup point to obtain the intermediate range theoretical current values ​​of all the segments of all the fuel assemblies at the current burnup point;

[0078] S35: Divide the intermediate-range theoretical current value at the current burnup point of all the nodes of all the fuel assemblies by the intermediate-range theoretical current value at the beginning of the life of all the nodes of all the fuel assemblies to obtain a correction coefficient at the current burnup point.

[0079] Wherein, step S33 can be based on the formula (2): I1 = C 3D ·∑ i,j A i,j ·P i,j Calculate the intermediate range theoretical current value of all nodes of all fuel assemblies at the beginning of their life. In formula (2), i represents the i-th fuel assembly, j represents the j-th node of a fuel assembly, and P i，j A represents the initial life power of the jth segment of the i-th fuel assembly, i，j represents the mid-range response factor of the jth segment of the i-th fuel assembly, ∑ i,j A i,j ·P i,j The sum of the products of the intermediate range response factors of all nodes of all fuel assemblies and the initial life power, C 3D is a constant, and I1 is the theoretical current value of the intermediate range at the beginning of the life of all nodes of all fuel assemblies.

[0080] Step S34 can be performed according to the formula (3) I2 = C 3D ·∑ i,j A i,j ·P i,j Calculate the intermediate range theoretical current value of all nodes of all fuel assemblies at the current burnup point. In formula (3), i represents the i-th fuel assembly, j represents the j-th node of a fuel assembly, and P i，j represents the current burnup power of the jth segment of the i-th fuel assembly, A i，j represents the mid-range response factor of the jth segment of the i-th fuel assembly, ∑ i,j A i,j ·P i,j The sum of the products of the intermediate range response factors of all nodes of all fuel assemblies and the power at the current burnup point, C 3D is a constant, and I2 is the mid-range theoretical current value of the current burnup point of all nodes of all fuel assemblies.

[0081] For a specific intermediate-range detector model, the sensitivity of the coating sensitive material B-10 remains unchanged. The relationship between the current value and power is finally determined by collecting the actual power of the nuclear reactor model and the intermediate-range theoretical current value of the intermediate-range detector model, combined with the above-mentioned Monte Carlo method simulation results, and performing correction calculations.

[0082] When neutrons leak from the core of the nuclear reactor model to the outside of the nuclear reactor, the intermediate-range theoretical current value at the current burnup point of the nuclear instrument system is directly proportional to the reaction rate of the intermediate-range detector model coating sensitive material B-10 and fission neutrons. Therefore, the intermediate-range theoretical current value of the current burnup point of the intermediate-range detector can be characterized by simulating the reaction rate of the intermediate-range detector model coating sensitive material B-10 from the beginning of the life cycle to the current burnup point.

[0083] Based on the increasing burnup of the unit power operation, the neutronic parameters of the j-th segment of the ith fuel assembly of the nuclear reactor model at the current burnup point are calculated, and the intermediate-range theoretical current value of the nuclear instrument system is calculated and given. The intermediate-range theoretical current values ​​are normalized based on the intermediate-range theoretical current values ​​at the beginning of the life cycle, which further illustrates the trend of the current value detected by the extra-core intermediate-range detector model becoming larger between the middle and late stages of the cycle as the burnup increases, and the intermediate-range theoretical current value is quantitatively calculated.

[0084] The theoretical current value of the intermediate range at the beginning of the life can be simulated by theoretical simulation to simulate the deviation change of the theoretical data detected by the intermediate range detector model from the beginning of the life to the current burn-up point. The theoretical current value I1 of the intermediate range at the beginning of the life and the theoretical current value I2 of the intermediate range at the current burn-up point are calculated at the corresponding power. The correction coefficient is set by the ratio of the current values

[0085] In some embodiments, the initial life power of the jth segment of the i-th fuel assembly is: the initial life power of the jth segment of the i-th fuel assembly at the beginning of the simulated life;

[0086] The current burnup point power of the j-th segment of the ith fuel assembly is: the current burnup point power of the j-th segment of the ith fuel assembly under the simulated current burnup point.

[0087] Among them, the initial life power P of the jth segment of the i-th fuel assembly at the beginning of the life of the nuclear reactor model core periphery is simulated by software i,j ;

[0088] The current burnup point power P of the jth segment of the i-th fuel assembly at the current burnup point of the core periphery of the nuclear reactor model is simulated by software i,j .

[0089] In some embodiments, step S4 includes:

[0090] By multiplying the correction coefficient at the current burnup point by the protection setting value of the intermediate range set at the beginning of the life cycle, the protection setting value of the intermediate range at the current burnup point is obtained.

[0091] Among them, the actual data collected during the initial field test at the beginning of the service life is used to determine the protection setting of the intermediate range, that is, the current setting values ​​corresponding to C1, RT and ATWT. As the unit's operating fuel consumption increases, the current value corresponding to the same power changes. Therefore, the protection setting of the intermediate range needs to be theoretically corrected and calculated based on the current fuel consumption point. Among them, C1 is the protection signal for locking the power control rod, RT is the emergency shutdown signal, and ATWT is the expected transient protection signal for failure to shut down the reactor in an emergency.

[0092] The protection setting values ​​of the intermediate range set at the beginning of the life cycle, that is, the current setting values ​​corresponding to C1, RT, and ATWT, are theoretically corrected as follows:

[0093] C1 (current burnup point) = C1 (beginning of life) × α;

[0094] RT (current burnup point) = RT (beginning of life) × α;

[0095] ATWT (current burnup point) = ATWT (beginning of life) × α.

[0096] Some embodiments of the present invention disclose a computer-readable storage medium having a computer program stored thereon. When the computer program is executed by a processor, a method for correcting a reactor protection constant as in any of the above embodiments is implemented, which will not be described in detail herein.

[0097] By implementing the present invention, the following beneficial effects are achieved:

[0098] The present invention establishes a nuclear reactor model and an intermediate-range detector model, selects a fuel assembly in the core of the nuclear reactor model to perform neutron transport simulation calculation, obtains an intermediate-range response factor after the fission neutrons of the fuel assembly are transmitted from the core active region to the intermediate-range detector model position, then calculates the intermediate-range theoretical current value at the beginning of the life cycle and the intermediate-range theoretical current value at the current burnup point according to the intermediate-range response factor, and obtains a correction coefficient at the current burnup point, and finally corrects the intermediate-range protection setting set at the beginning of the life cycle according to the correction coefficient at the current burnup point to obtain the intermediate-range protection setting at the current burnup point, thereby realizing the correction of the intermediate-range protection setting set at the beginning of the life cycle, ensuring that the protection setting set at the beginning of the life cycle is applicable after theoretical correction, ensuring the accuracy of the nuclear reactor protection setting, reducing the risk of false shutdown of the nuclear reactor and the risk of reactor protection failure, and taking into account both the safety and economic benefits of the unit.

[0099] It can be understood that the above embodiments only express some implementation methods of the present invention, and the description thereof is relatively specific and detailed, but it cannot be understood as limiting the patent scope of the present invention. It should be pointed out that for ordinary technicians in this field, without departing from the concept of the present invention, the above embodiments or technical features can be freely combined, and several deformations and improvements can be made, which all belong to the protection scope of the present invention, that is, the embodiments described in "some embodiments" can be freely combined with any of the above and below embodiments; therefore, all equivalent changes and modifications made to the scope of the claims of the present invention should fall within the scope of the claims of the present invention.

Claims

1. A method for correcting a nuclear reactor protection setting, characterized in that: The method comprises the following steps: S1: Establish a nuclear reactor model and an intermediate range detector model; S2: Selecting a fuel assembly in the core of the nuclear reactor model to perform neutron transport simulation calculation, and obtaining an intermediate range response factor after the fission neutrons of the fuel assembly are transmitted from the core active region to the position of the intermediate range detector model; S3: Calculate the intermediate range theoretical current value at the beginning of the life and the intermediate range theoretical current value at the current burnup point according to the intermediate range response factor, and obtain the correction coefficient at the current burnup point; S4: Correcting the protection constant value of the intermediate range set at the beginning of the life cycle according to the correction coefficient at the current burnup point to obtain the protection constant value of the intermediate range at the current burnup point.

2. The method for correcting the nuclear reactor protection setting value according to claim 1, characterized in that: Step S2 includes: S21: selecting i fuel assemblies on the periphery of the core of the nuclear reactor model, and dividing the i fuel assemblies into j segments in their axial heights, wherein i and j ≥ 1; S22: performing neutron transport simulation calculation on the j-th segment of the i-th fuel assembly to obtain the reaction rate of the fission neutrons of the j-th segment of the i-th fuel assembly after being transmitted from the core active region to the position of the intermediate range detection model; S23: performing normalized calculation on the reaction rate of the j-th segment of the i-th fuel assembly to obtain the intermediate range response factor of the j-th segment of the i-th fuel assembly.

3. The method for correcting the nuclear reactor protection setting value according to claim 2, characterized in that: Step S21 includes: i fuel assemblies of the intermediate range detector model located at the periphery of the core of the nuclear reactor model and close to the 270° or 90° position are selected, and the i fuel assemblies are respectively divided into j segments in their axial height.

4. The method for correcting the nuclear reactor protection setting value according to claim 2, characterized in that: Step S22 includes: The nuclear reactor model simulates the state of the jth segment of the i-th fuel assembly to obtain neutronics parameters; Processing the neutronic parameters, constructing a core source term for Monte Carlo transport calculation, and simulating the neutron injection rate after the fission neutrons of the j-th segment of the i-th fuel assembly are transmitted from the core active region to the position of the intermediate range detection model; According to the neutron injection rate and the coating sensitive material of the intermediate range detector model, the reaction rate of the fission neutrons of the j-th segment of the i-th fuel assembly after being transmitted from the core active area to the position of the intermediate range detection model is obtained.

5. The method for correcting the nuclear reactor protection setting value according to claim 2, characterized in that: The intermediate range theoretical current value of the intermediate range detector model per unit time is proportional to the reaction rate of the fission neutrons of the node after being transmitted from the core active region to the position of the intermediate range detection model.

6. The method for correcting the nuclear reactor protection setting value according to claim 2, characterized in that: Step S23 includes: The reaction rates of all nodes of the i-th fuel assembly are summed to obtain the sum of the reaction rates of all nodes of the i-th fuel assembly; The intermediate range response factor of the j-th section of the i-th fuel assembly is obtained by dividing the sum of the reactivity of all the sections of the i-th fuel assembly by the reactivity of the j-th section of the i-th fuel assembly.

7. The method for correcting the nuclear reactor protection setting value according to claim 2, characterized in that: Step S3 includes: S31: calculating the initial life intermediate range theoretical current value of the j-th segment of the i-th fuel assembly according to the intermediate range response factor of the j-th segment of the i-th fuel assembly and the initial life power of the j-th segment of the i-th fuel assembly; S32: calculating the intermediate range theoretical current value of the j-th segment of the i-th fuel assembly at the current burnup point according to the intermediate range response factor of the j-th segment of the i-th fuel assembly and the current burnup point power of the j-th segment of the i-th fuel assembly; S33: Repeat step S31 until i fuel assemblies are traversed, sum up the theoretical current values ​​of the initial intermediate range of the life of all the j-th segments of the i-th fuel assembly, and obtain the theoretical current values ​​of the initial intermediate range of the life of all the segments of all the fuel assemblies; S34: Repeat step S32 until all i fuel assemblies are traversed, and sum the intermediate range theoretical current values ​​of the j-th segment of all the i-th fuel assemblies at the current burnup point to obtain the intermediate range theoretical current values ​​of all the segments of all the fuel assemblies at the current burnup point; S35: Divide the intermediate-range theoretical current value at the current burnup point of all the nodes of all the fuel assemblies by the intermediate-range theoretical current value at the beginning of the life of all the nodes of all the fuel assemblies to obtain the correction coefficient at the current burnup point.

8. The method for correcting the nuclear reactor protection setting value according to claim 7, characterized in that: The initial life power of the jth segment of the i-th fuel assembly is: the initial life power of the jth segment of the i-th fuel assembly at the beginning of the simulated life; The current burnup point power of the j-th segment of the i-th fuel assembly is: the current burnup point power of the j-th segment of the i-th fuel assembly under the simulated current burnup point.

9. The method for correcting the nuclear reactor protection setting value according to claim 1, characterized in that: Step S4 includes: The protection set value of the intermediate range at the current burnup point is obtained by multiplying the correction coefficient at the current burnup point by the protection set value of the intermediate range set at the beginning of the life cycle.

10. A computer-readable storage medium having a computer program stored thereon, characterized in that: When the computer program is executed by a processor, the method for correcting the nuclear reactor protection constant as described in any one of claims 1 to 9 is implemented.

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