Neutron flux rate anomaly detection method and device, electronic equipment and storage medium

Abstract

The embodiment of the application provides a kind of neutron flux abnormality detection method and device, electronic equipment and storage medium, belong to nuclear power plant safety technical field.The method comprises: screening reference nuclear power reaction unit.Get the first reference counting rate detection module in multiple continuous preset time steps under the core neutron counting rate.Acquire concentration influence factor, original neutron source source intensity and reference neutron source source intensity.According to the first detection sensitivity, the first reference sensitivity, the core neutron counting rate, the concentration influence factor, the original neutron source source intensity and the reference neutron source source intensity, the predicted neutron flux of the core basket is obtained by calculation.The first loading of pressurized water reactor nuclear power reaction unit is carried out, and the target neutron counting rate is obtained.According to the target neutron counting rate and the first detection sensitivity, the target neutron flux is obtained by calculation, and the abnormality is detected based on the target neutron flux and the predicted neutron flux.The embodiment of the application can improve the efficiency of neutron flux abnormality detection.

Description

Technical Field

[0001] This application relates to the field of nuclear power plant safety technology, and in particular to a method and apparatus for detecting neutron flux anomalies, electronic equipment and storage medium. Background Technology

[0002] Currently, during the initial fuel loading of reactors in nuclear power plants, experts typically need to analyze the real-time neutron flux and, based on experience, determine whether the current neutron flux is abnormal. This results in low efficiency for detecting neutron flux anomalies. Therefore, improving the efficiency of neutron flux anomaly detection has become an urgent technical problem to be solved. Summary of the Invention

[0003] The main objective of this application is to provide a method, apparatus, electronic device, and storage medium for detecting neutron flux anomalies, aiming to improve the efficiency of neutron flux anomaly detection.

[0004] To achieve the above objectives, a first aspect of this application proposes a method for detecting anomalies in neutron flux rate, applied to a pressurized water reactor nuclear power unit. The pressurized water reactor nuclear power unit includes a pressure vessel, a core basket, and a first neutron count rate detection module. The core basket is used for nuclear reaction and is disposed within the pressure vessel. Coolant is injected into the gap between the core basket and the pressure vessel. The first neutron count rate detection module is disposed on the inner wall of the core basket. The method for detecting anomalies in neutron flux rate includes:

[0005] A benchmark nuclear reactor is selected from multiple preset reference nuclear reactors based on the reactor type of the pressurized water reactor.

[0006] The first reference count rate detection module obtains the core neutron count rate at multiple consecutive preset time steps; wherein, the first reference count rate detection module is the neutron count rate detection module of the reference nuclear power reactor, and the first reference count rate detection module is disposed on the inner wall of the core basket of the reference nuclear power reactor;

[0007] Obtain the concentration influence factor of boron concentration matched to the coolant, the original neutron source strength of the pressurized water reactor nuclear power unit under preset operating conditions, and the reference neutron source strength of the reference nuclear power reactor under the preset operating conditions.

[0008] The predicted neutron flux rate of the core basket at each preset time step is obtained by integrating and calculating the first detection sensitivity of the first neutron count rate detection module, the first reference sensitivity of the first reference count rate detection module, the core neutron count rate, the concentration influence factor, the original neutron source strength, and the reference neutron source strength.

[0009] The pressurized water reactor nuclear power unit is initially loaded with fuel, and the target neutron count rate is obtained through the first neutron count rate detection module;

[0010] The target neutron fluence rate is obtained by proportionally calculating the target neutron count rate and the first detection sensitivity.

[0011] Anomaly detection is performed based on the difference between the target neutron fluence rate and the predicted neutron fluence rate.

[0012] In some embodiments, the pressurized water reactor nuclear power unit further includes a probe well pipe and a second neutron count rate detection module. The probe well pipe is located outside the pressure vessel, and the second neutron count rate detection module is disposed inside the probe well pipe. The anomaly detection based on the difference between the target neutron flux rate and the predicted neutron flux rate includes:

[0013] The second neutron count rate of the second neutron count rate detection module is obtained at multiple consecutive preset time steps;

[0014] The second reference count rate detection module acquires the pipeline neutron count rate at multiple consecutive preset time steps; wherein, the second reference count rate detection module is the neutron count rate detection module of the reference nuclear power reactor, and the second reference count rate detection module is installed in the detection well pipeline of the reference nuclear power reactor;

[0015] The predicted injection rate of the pipeline at each preset time step is obtained by integrating and calculating the second detection sensitivity of the second neutron count rate detection module, the second reference sensitivity of the second reference count rate detection module, the pipeline neutron count rate, the concentration influence factor, the original neutron source strength, and the reference neutron source strength.

[0016] The pressurized water reactor nuclear power unit was initially loaded with fuel, and the detected neutron count rate was obtained through the second neutron count rate detection module;

[0017] The pipeline neutron injection rate is obtained by proportionally calculating the detected neutron count rate and the second detection sensitivity.

[0018] Anomaly detection is performed based on the difference between the target neutron fluence rate and the predicted neutron fluence rate, as well as the difference between the pipeline neutron fluence rate and the pipeline predicted fluence rate.

[0019] In some embodiments, the step of integrating and calculating the predicted injection rate of the pipeline at each preset time step within the probe well pipeline based on the second detection sensitivity of the second neutron count rate detection module, the second reference sensitivity of the second reference count rate detection module, the pipeline neutron count rate, the concentration influence factor, the original neutron source strength, and the reference neutron source strength includes:

[0020] Obtain a reference concentration factor, which is a concentration influence factor that matches the coolant boron concentration of the benchmark nuclear power reactor.

[0021] For the current preset time step, the predicted neutron count rate of the pipeline is calculated proportionally based on the concentration influence factor, the reference concentration factor, the second benchmark sensitivity, the second detection sensitivity, the original neutron source strength, the benchmark neutron source strength, the preset neutron lifetime time corresponding to the pressurized water reactor nuclear power unit, the preset neutron lifetime time corresponding to the benchmark nuclear power reactor unit, and the pipeline neutron count rate.

[0022] Return to the steps described above to obtain the reference concentration factor;

[0023] The predicted neutron count rate for each preset time step is obtained by calculating the ratio of the second neutron count rate of the second neutron count rate detection module at the first preset time step and the preset initial pipeline flux rate to each predicted neutron count rate of the pipeline.

[0024] In some embodiments, the step of proportionally calculating the predicted neutron count rate of the pipeline at the current preset time step based on the concentration influence factor, the reference concentration factor, the second baseline sensitivity, the second detection sensitivity, the original neutron source strength, the baseline neutron source strength, the preset neutron lifetime time corresponding to the pressurized water reactor nuclear power unit, the preset neutron lifetime time corresponding to the baseline nuclear power reactor, and the pipeline neutron count rate includes:

[0025] The original unit data is obtained by multiplying the concentration influence factor, the second detection sensitivity, the original neutron source strength, and the preset neutron lifetime time corresponding to the pressurized water reactor nuclear power unit.

[0026] Multiply the reference concentration factor, the second benchmark sensitivity, the benchmark neutron source strength, and the preset neutron lifetime time corresponding to the benchmark nuclear power reactor to obtain the benchmark reactor data;

[0027] The predicted neutron count rate of the pipeline is obtained by calculating the ratio of the pipeline neutron count rate, the reference unit data, and the original unit data at the current preset time step.

[0028] In some embodiments, the pressurized water reactor nuclear power unit further includes fuel assemblies disposed inside the core basket. The calculation of the pipeline neutron flux rate based on the detected neutron count rate and the second detection sensitivity includes:

[0029] Obtain the fuel influence factor matching the fuel assembly, the first container influence factor matching the core basket, the second container influence factor matching the pressure vessel, and the pipeline influence factor matching the exploration well pipeline;

[0030] The concentration influence factor, the second detection sensitivity, the fuel influence factor, the first container influence factor, the second container influence factor, and the pipeline influence factor are multiplied together to obtain the first intermediate data;

[0031] The pipeline neutron injection rate is obtained by calculating the ratio between the detected neutron count rate and the first intermediate data.

[0032] In some embodiments, the step of calculating the target neutron fluence rate by proportional calculation based on the target neutron count rate and the first detection sensitivity includes:

[0033] Multiply the fuel influence factor, the concentration influence factor, and the first detection sensitivity to obtain reference data;

[0034] The target neutron fluence rate is obtained by calculating the ratio between the target neutron count rate and the reference data.

[0035] In some embodiments, the step of calculating the target neutron fluence rate by comparing the target neutron count rate with the reference data includes:

[0036] Obtain the temperature influence factor that matches the temperature of the coolant;

[0037] The target neutron fluence rate is obtained by dividing the target neutron count rate by the product of the reference data and the temperature influence factor.

[0038] In some embodiments, the anomaly detection based on the difference between the target neutron fluence rate and the predicted neutron fluence rate, and the difference between the pipeline neutron fluence rate and the predicted pipeline fluence rate, includes:

[0039] The difference between the target neutron fluence rate and the predicted neutron fluence rate is calculated to obtain the first neutron fluence rate difference.

[0040] The difference between the neutron flux rate in the pipeline and the predicted flux rate in the pipeline is calculated to obtain the second neutron flux rate difference.

[0041] If at least one of the differences between the first neutron fluence rate and the second neutron fluence rate is greater than a preset threshold, it is determined that the neutron fluence rate is abnormal.

[0042] In some embodiments, the step of integrating and calculating the predicted neutron flux rate of the core basket at each preset time step based on the first detection sensitivity of the first neutron count rate detection module, the first reference sensitivity of the first reference count rate detection module, the core neutron count rate, the concentration influence factor, the original neutron source strength, and the reference neutron source strength includes:

[0043] Obtain a reference concentration factor, which is a concentration influence factor that matches the coolant boron concentration of the benchmark nuclear power reactor.

[0044] For the current preset time step, the predicted core neutron count rate is obtained by integrating and calculating the concentration influence factor, the reference concentration factor, the first detection sensitivity, the first benchmark sensitivity, the original neutron source strength, the benchmark neutron source strength, the preset neutron lifetime time corresponding to the pressurized water reactor nuclear power unit, the preset neutron lifetime time corresponding to the benchmark nuclear power reactor unit, and the core neutron count rate.

[0045] Return to the step of obtaining the reference concentration factor, calculate the core predicted neutron count rate for the next preset time step, and so on until the core predicted neutron count rate for all preset time steps is obtained;

[0046] The predicted neutron flux rate for each preset time step is obtained by calculating the ratio of the first neutron count rate of the first neutron count rate detection module at the first preset time step and the preset initial core flux rate.

[0047] In some embodiments, the step of multiplying the concentration influence factor, the reference concentration factor, the first detection sensitivity, the first benchmark sensitivity, the original neutron source strength, the benchmark neutron source strength, the preset neutron lifetime time corresponding to the pressurized water reactor nuclear power unit, the preset neutron lifetime time corresponding to the benchmark nuclear power reactor, and the core neutron count rate to obtain the core predicted neutron count rate at the current preset time step includes...

[0048] The original core data are obtained by multiplying the concentration influence factor, the first detection sensitivity, the original neutron source strength, and the preset neutron lifetime time corresponding to the pressurized water reactor nuclear power unit.

[0049] The reference concentration factor, the first reference sensitivity, the reference neutron source strength, and the preset neutron lifetime time corresponding to the reference nuclear power reactor are multiplied together to obtain the reference core data.

[0050] The ratio of the core neutron count rate, the reference core data, and the original core data is calculated to obtain the predicted core neutron count rate at the current preset time step.

[0051] To achieve the above objectives, a second aspect of this application provides a neutron flux anomaly detection device, the device comprising:

[0052] The unit screening module is used to screen a reference nuclear power reactor that matches the unit type from a plurality of preset reference nuclear power reactors according to the unit type of the pressurized water reactor nuclear power reactor;

[0053] The first acquisition module is used to acquire the core neutron count rate of the first reference count rate detection module at multiple consecutive preset time steps; wherein, the first reference count rate detection module is the neutron count rate detection module of the reference nuclear power reactor, and the first reference count rate detection module is disposed on the inner wall of the core basket of the reference nuclear power reactor;

[0054] The second acquisition module is used to acquire the concentration influence factor of boron concentration matched with the coolant, the original neutron source strength of the pressurized water reactor nuclear power unit under preset operating conditions, and the reference neutron source strength of the reference nuclear power reactor unit under the preset operating conditions.

[0055] The prediction module is used to integrate and calculate the predicted neutron flux rate of the core basket at each preset time step based on the first detection sensitivity of the first neutron count rate detection module, the first reference sensitivity of the first reference count rate detection module, the core neutron count rate, the concentration influence factor, the original neutron source strength, and the reference neutron source strength.

[0056] The third acquisition module is used to perform the initial fuel loading of the pressurized water reactor nuclear power unit and to acquire the target neutron count rate through the first neutron count rate detection module.

[0057] The fluence rate calculation module is used to perform a proportional calculation based on the target neutron count rate and the first detection sensitivity to obtain the target neutron fluence rate.

[0058] An anomaly detection module is used to perform anomaly detection based on the difference between the target neutron fluence rate and the predicted neutron fluence rate.

[0059] To achieve the above objectives, a third aspect of this application provides an electronic device, which includes a memory and a processor. The memory stores a computer program, and the processor executes the computer program to implement the method described in the first aspect.

[0060] To achieve the above objectives, a fourth aspect of the present application provides a computer-readable storage medium storing a computer program that, when executed by a processor, implements the method described in the first aspect.

[0061] The neutron flux anomaly detection method, apparatus, electronic equipment, and storage medium proposed in this application select a reference nuclear power reactor that matches the reactor type of the pressurized water reactor (PWR). The reference reactor's first reference count rate detection module obtains the core neutron count rate, the neutron source strength of both the reference reactor and the PWR under the same operating conditions, and the concentration influence factor matching the coolant. Then, based on the above data, the first detection sensitivity of the first neutron count rate detection module, and the first reference count rate detection module's first reference sensitivity, a predicted neutron flux rate is calculated. Subsequently, the PWR is initially loaded, and the target neutron count rate for the actual reaction process is obtained using the first neutron count rate. The target neutron flux rate is then calculated based on this target neutron count rate and the first detection sensitivity. Finally, anomaly detection is performed based on the difference between the target neutron flux rate and the predicted neutron flux rate. The method in this application uses known data from a benchmark nuclear power reactor to calculate the predicted neutron flux of the current pressurized water reactor at multiple consecutive preset time steps. During the actual reaction process, it is only necessary to compare the actual target neutron flux obtained from the first neutron count rate with the predicted neutron flux to quickly determine if an anomaly exists. This method reduces reliance on human experience, significantly improves the efficiency of neutron flux anomaly detection, and avoids the shortcomings of traditional methods that rely on human experience. Through accurate prediction and comparative analysis, it can promptly detect anomalies and make adjustments, thereby improving safety during the initial fuel loading of the reactor. Attached Figure Description

[0062] Figure 1 This is a flowchart of the neutron flux anomaly detection method provided in the embodiments of this application;

[0063] Figure 2 This is a schematic diagram of the structure of a pressurized water reactor nuclear power unit provided in an embodiment of this application;

[0064] Figure 3 yes Figure 1 The flowchart of step S104 in the process;

[0065] Figure 4 yes Figure 3 The flowchart of step S302 in the document;

[0066] Figure 5 yes Figure 1 The flowchart of step S106 in the process;

[0067] Figure 6 yes Figure 1 The flowchart of step S107 in the process;

[0068] Figure 7 yes Figure 6 The flowchart of step S603 in the process;

[0069] Figure 8 yes Figure 7 The flowchart of step S702 in the process;

[0070] Figure 9 yes Figure 6 The flowchart of step S605 in the process;

[0071] Figure 10 yes Figure 6 The flowchart of step S606 in the process;

[0072] Figure 11 This is a schematic diagram of the neutron flux anomaly detection device provided in the embodiments of this application;

[0073] Figure 12 This is a schematic diagram of the hardware structure of the electronic device provided in the embodiments of this application. Detailed Implementation

[0074] To make the objectives, technical solutions, and advantages of this application clearer, the following detailed description is provided in conjunction with the accompanying drawings and embodiments. It should be understood that the specific embodiments described herein are merely illustrative and not intended to limit the scope of this application.

[0075] It should be noted that although functional modules are divided in the device schematic diagram and a logical order is shown in the flowchart, in some cases, the steps shown or described may be performed in a different order than the module division in the device or the order in the flowchart. The terms "first," "second," etc., in the specification, claims, and the aforementioned drawings are used to distinguish similar objects and are not necessarily used to describe a specific order or sequence.

[0076] Unless otherwise defined, all technical and scientific terms used herein have the same meaning as commonly understood by one of ordinary skill in the art to which this application belongs. The terminology used herein is for the purpose of describing embodiments of this application only and is not intended to limit this application.

[0077] First, let's analyze some of the terms used in this application:

[0078] Neutron fluence rate: a nuclear physics term, defined as the quotient of the number of neutrons entering a small sphere centered at a point in space per unit time divided by the maximum cross-sectional area of ​​the sphere, or the sum of the distances traveled by all neutrons per unit volume per unit time. The unit is n / (cm2·s), also known as neutron flux density.

[0079] Neutron count rate: This refers to the count rate of a neutron detector and is one of the indicators for measuring the intensity of the neutron flux. A higher value indicates a higher reactor power, and a lower value indicates a lower reactor power.

[0080] Initial fuel loading: This refers to the process of initially loading nuclear fuel into the reactor core, typically a critical stage before the nuclear power plant begins operation. This process can also be understood as the reactor's first "start-up," where nuclear fuel is loaded into the reactor core, and the reactor is started to initiate a controlled nuclear fission reaction.

[0081] Currently, during the initial fuel loading of reactors in nuclear power plants, experts typically need to analyze the real-time neutron flux and, based on experience, determine whether the current neutron flux is abnormal. This results in low efficiency for detecting neutron flux anomalies. Therefore, improving the efficiency of neutron flux anomaly detection has become an urgent technical problem to be solved.

[0082] Based on this, embodiments of this application provide a method and apparatus for detecting neutron flux anomalies, an electronic device, and a storage medium, aiming to improve the efficiency of neutron flux anomaly detection.

[0083] The neutron flux anomaly detection method, apparatus, electronic device, and storage medium provided in this application are specifically described through the following embodiments. First, the neutron flux anomaly detection method in this application is described.

[0084] The neutron flux anomaly detection method provided in this application relates to the field of nuclear power plant safety technology. This method can be applied to a terminal, a server, or software running on either a terminal or a server. In some embodiments, the terminal can be a smartphone, tablet, laptop, desktop computer, etc.; the server can be configured as an independent physical server, a server cluster or distributed system composed of multiple physical servers, or a cloud server providing basic cloud computing services such as cloud services, cloud databases, cloud computing, cloud functions, cloud storage, network services, cloud communication, middleware services, domain name services, security services, CDN, and big data and artificial intelligence platforms; the software can be an application implementing the neutron flux anomaly detection method, but is not limited to the above forms.

[0085] This application can be used in a wide variety of general-purpose or special-purpose computer system environments or configurations. Examples include: personal computers, server computers, handheld or portable devices, tablet devices, multiprocessor systems, microprocessor-based systems, set-top boxes, programmable consumer electronics, network PCs, minicomputers, mainframe computers, and distributed computing environments including any of the above systems or devices. This application can be described in the general context of computer-executable instructions executed by a computer, such as program modules. Generally, program modules include routines, programs, objects, components, data structures, etc., that perform specific tasks or implement specific abstract data types. This application can also be practiced in distributed computing environments where tasks are performed by remote processing devices connected via a communication network. In distributed computing environments, program modules can reside in local and remote computer storage media, including storage devices.

[0086] Figure 1 This is an optional flowchart of the neutron flux anomaly detection method provided in this application embodiment. The method provided in this application embodiment is applied to a pressurized water reactor nuclear power unit. The pressurized water reactor nuclear power unit includes a pressure vessel, a core basket, and a first neutron count rate detection module. The core basket is used for nuclear reaction and is disposed inside the pressure vessel. The gap between the core basket and the pressure vessel is filled with coolant. The first neutron count rate detection module is disposed on the inner wall of the core basket. Figure 1 The method may include, but is not limited to, steps S101 to S107.

[0087] Step S101: Select a reference nuclear reactor from multiple preset reference nuclear reactors based on the reactor type of the pressurized water reactor.

[0088] Step S102: Obtain the core neutron count rate of the first reference count rate detection module at multiple consecutive preset time steps.

[0089] Step S103: Obtain the concentration influence factor of boron concentration matched to the coolant, the original neutron source strength of the pressurized water reactor nuclear power unit under preset operating conditions, and the reference neutron source strength of the reference nuclear power reactor under preset operating conditions.

[0090] Step S104: Based on the first detection sensitivity of the first neutron count rate detection module, the first reference sensitivity of the first reference count rate detection module, the core neutron count rate, the concentration influence factor, the original neutron source strength, and the reference neutron source strength, the predicted neutron flux rate of the core basket at each preset time step is obtained through integrated calculation.

[0091] Step S105: The pressurized water reactor nuclear power unit is initially loaded with fuel, and the target neutron count rate is obtained through the first neutron count rate detection module.

[0092] Step S106: Calculate the target neutron fluence rate by proportionally calculating the target neutron count rate and the first detection sensitivity.

[0093] Step S107: Anomaly detection is performed based on the difference between the target neutron fluence rate and the predicted neutron fluence rate.

[0094] Steps S101 to S107 of this embodiment involve selecting a reference nuclear reactor that matches the reactor type of the pressurized water reactor (PWR), and obtaining the core neutron count rate, the neutron source strength of the reference reactor and the PWR under the same operating conditions, and the concentration influence factor matching the coolant using the first reference count rate detection module of the reference reactor. Then, based on the above data, the first detection sensitivity of the first neutron count rate detection module, and the first reference sensitivity of the first reference count rate detection module, a predicted neutron flux rate is calculated. Subsequently, the PWR is initially loaded, and the target neutron count rate of the actual reaction process is obtained using the first neutron count rate. The target neutron flux rate is then calculated based on this target neutron count rate and the first detection sensitivity. Finally, anomaly detection is performed based on the difference between the target neutron flux rate and the predicted neutron flux rate. The method in this application uses known data from a benchmark nuclear power reactor to calculate the predicted neutron flux of the current pressurized water reactor at multiple consecutive preset time steps. During the actual reaction process, it is only necessary to compare the actual target neutron flux obtained from the first neutron count rate with the predicted neutron flux to quickly determine if an anomaly exists. This method reduces reliance on human experience, significantly improves the efficiency of neutron flux anomaly detection, and avoids the shortcomings of traditional methods that rely on human experience. Through accurate prediction and comparative analysis, it can promptly detect anomalies and make adjustments, thereby improving safety during the initial fuel loading of the reactor.

[0095] It should be noted that the neutron flux anomaly detection method provided in this application relies on the structure of a pressurized water reactor nuclear power unit. The structure of the pressurized water reactor nuclear power unit provided in this application is now described in detail. Please refer to... Figure 2 , Figure 2 This is a schematic diagram of the structure of a pressurized water reactor nuclear power unit provided in an embodiment of this application. The pressurized water reactor nuclear power unit provided in this embodiment includes a pressure vessel, a core basket, and a first neutron count rate detection module. The core basket is used for nuclear reaction and is located inside the pressure vessel. A primary coolant loop is filled in the gap between the core basket and the pressure vessel. The first neutron count rate detection module is located on the inner wall of the core basket and is used to detect the neutron count rate inside the core basket during initial fuel loading. Fuel assemblies and a neutron source assembly are also provided inside the core basket. The fuel assembly provides nuclear fuel, and its main function is to release energy through fission. The neutron source assembly provides a stable neutron source during initial fuel loading, helping to start the nuclear reactor and maintain the neutron environment inside the core, ensuring that the nuclear reaction can start smoothly and enter a stable operating state.

[0096] The pressurized water reactor nuclear power unit also includes a probe well pipe and a second neutron count rate detection module. The probe well pipe is located outside the pressure vessel, and the second neutron count rate detection module is set inside the probe well pipe. The second neutron count rate detection module can be a source-range neutron detector used to detect neutron count rate data at the probe well pipe.

[0097] In step S101 of some embodiments, based on the unit type of the pressurized water reactor nuclear power unit, a nuclear power reactor of the same unit type, with the same initial fuel loading sequence, initial criticality mode, and operating conditions is selected from the reference reactor units and used as the reference nuclear power reactor unit. The unit type can be, but is not limited to, this. The reference nuclear power reactor unit is a nuclear power reactor unit that has completed its initial fuel loading and has available historical count rate data (including neutron count rate and neutron flux rate).

[0098] In step S102 of some embodiments, the first reference count rate detection module is a neutron count rate detection module for a reference nuclear power reactor, and the first reference count rate detection module is disposed on the inner wall of the core basket of the reference nuclear power reactor. Preferably, the first reference count rate detection module and the first neutron count rate detection module are neutron detectors of the same type. The preset time step can be set according to actual detection requirements. In this embodiment, the preset time step is defined as the number of loading steps for the initial fuel loading of the reactor.

[0099] In some embodiments, the neutron count rate data is acquired in real time at each loading step by the first reference count rate detection module of the reference nuclear power reactor, and the neutron count rate change curve of the reference reactor is constructed.

[0100] In step S103 of some embodiments, the preset operating condition can be a normal temperature and pressure condition, but is not limited to this. The source strength of the original neutron source and the source strength of the reference neutron source are both preset values. The value of the source strength of the original neutron source can be 10.14, and the value of the source strength of the reference neutron source can be 28.98, but is not limited to this. Operators can obtain the corresponding values ​​from relevant design documents.

[0101] In step S104 of some embodiments, please refer to Figure 3 Step S104 may include, but is not limited to, steps S301 to S304:

[0102] Step S301: Obtain the reference concentration factor.

[0103] Step S302: For the current preset time step, the predicted neutron count rate of the core is obtained by integrating the concentration influence factor, reference concentration factor, first detection sensitivity, first benchmark sensitivity, original neutron source strength, benchmark neutron source strength, preset neutron lifetime time corresponding to the pressurized water reactor nuclear power unit, preset neutron lifetime time corresponding to the benchmark nuclear power reactor unit, and core neutron count rate.

[0104] Step S303: Return to the step of obtaining the reference concentration factor, calculate the core predicted neutron count rate for the next preset time step, until the core predicted neutron count rate for all preset time steps is obtained.

[0105] Step S304: Calculate the ratio of the predicted neutron count rate of each core to the first neutron count rate of the first neutron count rate detection module at the first preset time step and the preset initial core flux rate to obtain the predicted neutron flux rate for each preset time step.

[0106] In step S301 of some embodiments, the concentration influence factor is a parameter used to quantify the influence of the boron concentration of the coolant in a pressurized water reactor nuclear power unit on the change in the reactor core neutron count rate. The reference concentration factor is a concentration influence factor matched to the coolant boron concentration of a benchmark nuclear power reactor; both vary with the reactor reaction process.

[0107] It should be noted that, in addition to boron concentration, the thickness and temperature of the primary coolant can also affect the neutron count rate detection module. However, since the selected reference nuclear reactor has the same structure as the pressurized water reactor under test, the coolant thickness is the same for both, so the influence of thickness on the neutron count rate detection module can be ignored. Furthermore, during the initial fuel loading period, the system operates under deep subcritical, ambient temperature and pressure conditions, and the isothermal temperature coefficient of the primary coolant is a small positive value. When the primary coolant temperature varies within the ambient temperature range (generally 10℃ to 40℃), and when the average primary coolant temperature varies within ±2℃ of the reference temperature during the initial criticality (the isothermal temperature coefficient is approximately -3 pcm / ℃), the influence of temperature on the neutron detector count rate is negligible. However, in some other embodiments, temperature-related influencing factors can be considered.

[0108] In step S302 of some embodiments, please refer to Figure 4 Step S302 may include, but is not limited to, steps S401 to S403:

[0109] Step S401: Multiply the concentration influence factor, the first detection sensitivity, the original neutron source strength, and the preset neutron lifetime time corresponding to the pressurized water reactor nuclear power unit to obtain the original core data.

[0110] Step S402: Multiply the reference concentration factor, the first reference sensitivity, the reference neutron source strength, and the preset neutron lifetime time corresponding to the reference nuclear reactor to obtain the reference core data.

[0111] Step S403: Calculate the ratio of the core neutron count rate, the reference core data, and the original core data to obtain the predicted core neutron count rate at the current preset time step.

[0112] In step S401 of some embodiments, the first detection sensitivity refers to the response capability of the first neutron count rate detection module to neutron signals, specifically the change in count rate that the detection module can generate when the neutron source intensity changes by a unit. For example, if the first detection sensitivity of the first neutron count rate detection module is 5 cps / n·s, it means that for every increase of 1 neutron per second in neutron source intensity, the count rate of the detection module will increase by 5 counts per second. The preset neutron lifetime is the converted neutron generation time, i.e., the average neutron lifetime, which can be 27.99 seconds, 28.02 seconds, or is not limited to this. The calculation method for the original core data can refer to the following analytical formula:

[0113] k CB，A ×S A ×l 0，A ×Q A (1),

[0114] Among them, K CB，A S represents the concentration-influencing factor. A Indicates the first detection sensitivity. 0，A Indicates pressure

[0115] The preset neutron lifetime of a water-powered nuclear reactor. Q A This indicates the original neutron source strength. It should be noted that the subscript A in the formula represents a pressurized water reactor nuclear power unit, and the subscript B represents a reference nuclear power reactor unit.

[0116] In step S402 of some embodiments, the calculation method for the reference core data can refer to the following analytical expression:

[0117] k CB，B ×S B ×l 0，B ×Q B (2),

[0118] Where, k CB，B S represents the reference concentration factor. B Indicates the first reference sensitivity. 0，B This indicates the preset neutron lifetime of a benchmark nuclear power reactor. Q BThis indicates the source strength of the reference neutron source. It should be noted that the pressurized water reactor nuclear power unit operates under the same conditions as the reference nuclear power reactor.

[0119] In step S403 of some embodiments, the relationship between the predicted core neutron count rate, the core neutron count rate, the reference core data, and the original core data can be referred to the following analytical expression:

[0120]

[0121] Where, k CB，A ×S A ×l 0，A ×Q A This is the raw core data. CB，B ×S B ×l 0，B ×Q B This is the baseline core data. A (t) represents the predicted neutron count rate in the reactor core. C B (t) represents the core neutron count rate.

[0122] Steps S401 to S403 as shown in the embodiments of this application effectively improve the prediction accuracy of the neutron count rate at the core by using the ratio of the original core data to the reference core data. This ratio relationship also eliminates the influence of differences in coolant concentration, neutron source intensity and other factors between different units, thereby improving the prediction reliability of the neutron flux rate.

[0123] In step S303 of some embodiments, the process returns to step S301, continuously calculating the predicted value of the neutron count rate in the core basket (i.e., the predicted core neutron count rate) based on various parameters of the current preset time step, and progressively calculating up to all preset time steps until the prediction results for all time steps are completed. For example, assuming 150 loading steps need to be predicted, the predicted core neutron count rate corresponding to the first loading step is calculated according to steps S301 to S302. Then, the process returns to step S301 to obtain the reference concentration factor for the second loading step, and the calculation is repeated to obtain the predicted core neutron count rate corresponding to the second loading step. This process continues until the predicted core neutron count rate for the fifth time step is calculated.

[0124] In step S304 of some embodiments, the predicted neutron count rate and the predicted neutron fluence rate have a certain proportional relationship, as shown in the following analytical expression:

[0125]

[0126] Where t represents the time step number, C0 represents the first neutron count rate of the first neutron count rate detection module at the first preset time step, N0 represents the initial neutron fluence rate at the first preset time step, which is a known non-zero constant value, N(t) represents the predicted neutron fluence rate at the current prediction time step, and C(t) represents the predicted core neutron count rate at the current preset time step.

[0127] Steps S301 to S304 as shown in the embodiments of this application use known quantities such as concentration influence factor, reference concentration factor, first detection sensitivity, first reference sensitivity, original neutron source strength, reference neutron source strength, preset neutron lifetime time and core neutron count rate to calculate the result as the predicted neutron count rate, and maintain a one-to-one correspondence with the time step, so as to reduce the dependence on manual interpretation and improve the efficiency of anomaly detection in engineering implementation.

[0128] In step S105 of some embodiments, initial fuel loading refers to the initial loading of nuclear fuel into the pressurized water reactor nuclear power unit. During initial fuel loading, nuclear fuel assemblies are installed into the core basket, and the initial reaction process of the reactor is initiated. The nuclear fuel may be loaded into the core basket in batches according to a predetermined fuel loading schedule. The target neutron count rate refers to the neutron count rate data actually measured by the first neutron count rate detection module during initial fuel loading.

[0129] In step S106 of some embodiments, please refer to Figure 5 Step S106 may also include, but is not limited to, steps S501 to S502:

[0130] Step S501: Multiply the fuel influence factor, concentration influence factor and first detection sensitivity to obtain reference data.

[0131] Step S502: Calculate the target neutron fluence rate by performing a ratio calculation based on the target neutron count rate and reference data.

[0132] In step S501 of some embodiments, the fuel influence factor is a parameter used to quantify the degree of influence of fuel assemblies on the core neutron count rate in the reactor. The calculation method for the reference data can be referred to the following analytical expression:

[0133] k r ×k CB,A ×S TD (5),

[0134] Where, k CB，A k represents the concentration-influencing factor. r S represents the fuel impact factor. TD This indicates the first detection sensitivity.

[0135] In step S502 of some embodiments, the target neutron fluence rate can be obtained by dividing the target neutron count rate by the reference data.

[0136] In other embodiments, such as the embodiment described in step S301, the temperature influence factor of the coolant can be simultaneously considered in the calculation of the target neutron flux rate. Specifically, firstly, a temperature influence factor matching the temperature of the coolant is obtained, and the target neutron count rate is divided by the product of the reference data and the temperature influence factor to obtain the target neutron flux rate. The following analytical expression can be referred to:

[0137] C TD (t)=k r ×k CB,A ×k TM ×N(t)×S TD (6),

[0138] Where, k r ×k CB,A ×S TD For reference only. k TM This refers to the temperature-related factors.

[0139] Steps S501 to S502, as illustrated in the embodiments of this application, effectively improve the prediction accuracy of neutron flux rate by comprehensively considering fuel influence factors, concentration influence factors, and the first detection sensitivity. By calculating the ratio of these influence factors to the target neutron count rate, more accurate data support can be provided for reactor safety detection.

[0140] In step S107 of some embodiments, a threshold is set, and the deviation between the target neutron fluence rate and the predicted neutron fluence rate is calculated. When the deviation exceeds a set safety range, an automatic warning can be issued and an emergency response procedure can be initiated. For example, if the difference between the target neutron fluence rate and the predicted neutron fluence rate exceeds the allowable deviation range of 5%, it is determined to be an abnormal situation; if the deviation does not exceed the 5% deviation range, it is determined to be normal.

[0141] Please see Figure 6 In some embodiments, the pressurized water reactor nuclear power unit also includes a probe well pipe and a second neutron count rate detection module. The probe well pipe is located outside the pressure vessel, and the second neutron count rate detection module is disposed inside the probe well pipe. The neutron flux anomaly detection logic at the probe well pipe is consistent with the neutron flux anomaly detection logic at the core basket. Specifically, step S107 includes, but is not limited to, steps S601 to S606:

[0142] Step S601: Obtain the second neutron count rate of the second neutron count rate detection module at multiple consecutive preset time steps.

[0143] Step S602: Obtain the neutron count rate in the pipeline at multiple consecutive preset time steps by the second reference count rate detection module.

[0144] Step S603: Based on the second detection sensitivity of the second neutron count rate detection module, the second reference sensitivity of the second reference count rate detection module, the pipeline neutron count rate, the concentration influence factor, the original neutron source strength, and the reference neutron source strength, the pipeline predicted injection rate for each preset time step in the probe well pipeline is obtained through integrated calculation.

[0145] Step S604: The pressurized water reactor nuclear power unit is initially loaded with fuel, and the detection neutron count rate is obtained through the second neutron count rate detection module.

[0146] Step S605: Calculate the pipeline neutron flux rate by proportionally calculating the detection neutron count rate and the second detection sensitivity.

[0147] Step S606: Anomaly detection is performed based on the difference between the target neutron fluence rate and the predicted neutron fluence rate, and the difference between the pipeline neutron fluence rate and the pipeline predicted fluence rate.

[0148] In step S601 of some embodiments, the second neutron count rate detection module is a neutron detector installed in the detection well pipe of the pressurized water reactor nuclear power unit. The second neutron count rate is the reading of the second neutron count rate detection module.

[0149] In step S602 of some embodiments, the second reference count rate detection module is a neutron count rate detection module of a reference nuclear power reactor, and the second reference count rate detection module is installed in the detection well pipeline of the reference nuclear power reactor.

[0150] In step S603 of some embodiments, please refer to Figure 7 Step S603 may include, but is not limited to, steps S701 to S704:

[0151] Step S701: Obtain the reference concentration factor, which is the concentration influence factor of the coolant boron concentration matched to that of the benchmark nuclear power reactor.

[0152] Step S702: For the current preset time step, the predicted neutron count rate of the pipeline is calculated proportionally based on the concentration influence factor, reference concentration factor, second baseline sensitivity, second detection sensitivity, original neutron source strength, baseline neutron source strength, preset neutron lifetime time corresponding to the pressurized water reactor nuclear power unit, preset neutron lifetime time corresponding to the baseline nuclear power reactor unit, and pipeline neutron count rate.

[0153] Step S703: Return to the step of obtaining the reference concentration factor, calculate the pipeline predicted neutron count rate for the next preset time step, and so on until the pipeline predicted neutron count rate for all preset time steps is obtained.

[0154] Step S704: Calculate the ratio of the predicted neutron count rate of each pipeline based on the second neutron count rate of the second neutron count rate detection module at the first preset time step and the preset initial pipeline flux rate, to obtain the predicted pipeline flux rate for each preset time step.

[0155] In some embodiments, step S701 is consistent with the specific embodiment of step S301, and will not be described again here.

[0156] In step S702 of some embodiments, please refer to Figure 8 Step S702 may include, but is not limited to, steps S801 to S803:

[0157] Step S801: Multiply the concentration influence factor, the second detection sensitivity, the original neutron source strength, and the preset neutron lifetime time corresponding to the pressurized water reactor nuclear power unit to obtain the original unit data.

[0158] Step S802: Multiply the reference concentration factor, the second reference sensitivity, the reference neutron source strength, and the preset neutron lifetime time corresponding to the reference nuclear power reactor to obtain the reference reactor data.

[0159] Step S803: Calculate the ratio of the pipeline neutron count rate, the baseline unit data, and the original unit data to obtain the pipeline predicted neutron count rate at the current preset time step.

[0160] In step S801 of some embodiments, the second detection sensitivity refers to the response capability of the second neutron count rate detection module to neutron signals. The calculation method for the raw unit data can refer to the following analytical formula:

[0161] k CB，A ×S A，SRD ×l 0，A ×Q A (7),

[0162] Where, k CB，A S represents the concentration-influencing factor. A，SRD This indicates the second detection sensitivity. 0，A This indicates the preset neutron lifetime of the pressurized water reactor nuclear power unit. Q A This indicates the original neutron source strength. It should be noted that the subscript A in the formula represents a pressurized water reactor nuclear power unit, and the subscript B represents a reference nuclear power reactor unit.

[0163] In step S802 of some embodiments, the calculation method for the reference unit data can refer to the following analytical expression:

[0164] k CB，B ×S B，SRD ×l 0，B ×Q B (8),

[0165] Where, k CB，B S represents the reference concentration factor. B，SRD This indicates the second reference sensitivity. 0，B This indicates the preset neutron lifetime of a benchmark nuclear power reactor. Q B This indicates the strength of the benchmark neutron source.

[0166] In step S803 of some embodiments, the relationship between the predicted neutron count rate of the pipeline, the pipeline neutron count rate, the reference unit data, and the original unit data can be referred to the following analytical expression:

[0167]

[0168] Where, k CB，A ×S A，SRD ×l 0，A ×Q A This is the raw unit data. CB，B ×S B，SRD ×l 0，B ×Q B This is the baseline unit data. (C) A，SRD (t) represents the predicted neutron count rate of the pipeline. C B，SRD (t) represents the neutron count rate in the pipeline.

[0169] Steps S801 to S803 as shown in the embodiments of this application indicate that the method of this embodiment effectively improves the prediction efficiency of the neutron count rate of the pipeline by comprehensively calculating multiple relevant parameters of the well pipeline.

[0170] In some embodiments, the implementation logic of step S703 is consistent with that of the specific embodiment of step S303, and will not be described again here.

[0171] In some embodiments, step S704 has the same calculation logic as the specific embodiment of step S304. C0 is replaced with the second neutron count rate of the second neutron count rate detection module in the first preset time step, N0 is replaced with the preset initial injection rate of the pipeline (i.e., the neutron injection rate in the probe well pipe in the first preset time step), and C(t) is replaced with the predicted injection rate of the pipeline in the current preset time step. The specific calculation process will not be described in detail here.

[0172] Steps S701 to S704 shown in the embodiments of this application, by comprehensively considering factors such as concentration influence factor, second reference sensitivity, second detection sensitivity, original neutron source intensity, and reference neutron source intensity, can accurately predict the neutron flux rate at the pipeline.

[0173] In step S604 of some embodiments, the detected neutron count rate is the actual detection value of the second neutron count rate detection module during the initial loading.

[0174] In step S605 of some embodiments, please refer to Figure 9 Step S605 may include, but is not limited to, steps S901 to S903:

[0175] Step S901: Obtain the fuel influence factor matched to the fuel assembly, the first container influence factor matched to the core basket, the second container influence factor matched to the pressure vessel, and the pipeline influence factor matched to the exploration well pipeline.

[0176] Step S902: Multiply the concentration influence factor, the second detection sensitivity, the fuel influence factor, the first container influence factor, the second container influence factor, and the pipeline influence factor to obtain the first intermediate data.

[0177] Step S903: Calculate the pipeline neutron flux rate by comparing the detected neutron count rate with the first intermediate data.

[0178] In step S901 of some embodiments, the fuel influence factor matched to the fuel assembly, the first container influence factor matched to the core basket, the second container influence factor matched to the pressure vessel, and the pipeline influence factor matched to the exploration well pipeline are all preset values.

[0179] In step S902 of some embodiments, the calculation method of the first intermediate data can refer to the following analytical expression:

[0180] k r ×k c1 ×k c2 ×k g ×k CB，A ×S SRD (10),

[0181] Where, k c1 This represents the influence factor of the first container. k c2 This represents the influence factor of the second container. k g S represents the pipeline impact factor. SRD This indicates the second detection sensitivity. k CB，A k represents the concentration-influencing factor. r This indicates the fuel impact factor.

[0182] In step S903 of some embodiments, the calculation process for the ratio can refer to the following analytical expression:

[0183] C SRC (t)=k r ×k c1 ×k c2 ×k g ×k CB，A ×N(t)×S SRD (11),

[0184] Among them, C SRC (t) represents the detected neutron count rate, where N(t) is the neutron injection rate in the pipeline.

[0185] Steps S901 to S903 as shown in the embodiments of this application improve the prediction efficiency of neutron flux rate in pipelines by introducing comprehensive calculations of multiple influencing factors such as fuel assemblies, core baskets, pressure vessels, and probe well pipelines, thereby enabling a comprehensive reflection of the operating status of each part of the reactor.

[0186] In step S606 of some embodiments, please refer to Figure 10 Step S606 may include, but is not limited to, steps S1001 to S1003:

[0187] Step S1001: Calculate the difference between the target neutron fluence rate and the predicted neutron fluence rate to obtain the first neutron fluence rate difference.

[0188] Step S1002: Calculate the difference between the pipeline neutron flux rate and the pipeline predicted flux rate to obtain the second neutron flux rate difference.

[0189] Step S1003: If at least one of the first neutron fluence rate difference and the second neutron fluence rate difference is greater than a preset threshold, it is determined that the neutron fluence rate is abnormal.

[0190] In step S1003 of some embodiments, if the difference between the first neutron fluence rate and the second neutron fluence rate are both less than or equal to a preset threshold, it is determined that the neutron fluence rate has not been abnormal.

[0191] Steps S1001 to S1003 as illustrated in the embodiments of this application can detect potential abnormal changes in real time by comparing the differences between the target neutron flux rate and the predicted neutron flux rate, and between the pipeline neutron flux rate and the pipeline predicted flux rate.

[0192] Steps S601 to S606 of this embodiment, by adding a second neutron count rate detection module at the detection well pipeline, achieve unification and efficiency in the detection of neutron flux anomalies throughout the pressurized water reactor nuclear power unit. The method of this embodiment not only reduces reliance on manual judgment but also significantly improves the accuracy of neutron flux anomaly detection, further enhancing the safety and stability of the nuclear power reactor.

[0193] Please see Figure 11 This application also provides a neutron flux anomaly detection device, which can implement the above-mentioned neutron flux anomaly detection method. The device includes:

[0194] The unit screening module 1101 is used to screen a benchmark nuclear power reactor that matches the unit type from multiple preset reference nuclear power reactors according to the unit type of the pressurized water reactor nuclear power reactor.

[0195] The first acquisition module 1102 is used to acquire the core neutron count rate of the first reference count rate detection module at multiple consecutive preset time steps; wherein, the first reference count rate detection module is the neutron count rate detection module of the reference nuclear power reactor, and the first reference count rate detection module is disposed on the inner wall of the core basket of the reference nuclear power reactor.

[0196] The second acquisition module 1103 is used to acquire the concentration influence factor of boron concentration matched with coolant, the original neutron source strength of pressurized water reactor nuclear power unit under preset operating conditions, and the reference neutron source strength of reference nuclear power reactor under preset operating conditions.

[0197] Prediction module 1104 is used to integrate and calculate the predicted neutron flux rate of the core basket at each preset time step based on the first detection sensitivity of the first neutron count rate detection module, the first reference sensitivity of the first reference count rate detection module, the core neutron count rate, the concentration influence factor, the original neutron source strength, and the reference neutron source strength.

[0198] The third acquisition module 1105 is used to perform the initial fuel loading of the pressurized water reactor nuclear power unit and to acquire the target neutron count rate through the first neutron count rate detection module.

[0199] Flux rate calculation module 1106 is used to perform proportional calculation based on the target neutron count rate and the first detection sensitivity to obtain the target neutron flux rate;

[0200] The anomaly detection module 1107 is used to detect anomalies based on the difference between the target neutron fluence rate and the predicted neutron fluence rate.

[0201] The specific implementation of this neutron flux anomaly detection device is basically the same as the specific embodiment of the neutron flux anomaly detection method described above, and will not be repeated here.

[0202] This application also provides an electronic device, which includes a memory and a processor. The memory stores a computer program, and the processor executes the computer program to implement the above-described neutron flux rate anomaly detection method. This electronic device can be any smart terminal, including tablet computers, in-vehicle computers, etc.

[0203] Please see Figure 12 , Figure 12 The hardware structure of an electronic device according to another embodiment is illustrated. The electronic device includes:

[0204] The processor 1201 can be implemented using a general-purpose CPU (Central Processing Unit), microprocessor, application-specific integrated circuit (ASIC), or one or more integrated circuits, and is used to execute relevant programs to implement the technical solutions provided in the embodiments of this application.

[0205] The memory 1202 can be implemented as a read-only memory (ROM), static storage device, dynamic storage device, or random access memory (RAM). The memory 1202 can store the operating system and other application programs. When the technical solutions provided in the embodiments of this specification are implemented through software or firmware, the relevant program code is stored in the memory 1202 and is called and executed by the processor 1201 to execute the neutron fluence rate anomaly detection method of the embodiments of this application.

[0206] The input / output interface 1203 is used to implement information input and output;

[0207] The communication interface 1204 is used to enable communication and interaction between this device and other devices. Communication can be achieved through wired means (such as USB, network cable, etc.) or wireless means (such as mobile network, WIFI, Bluetooth, etc.).

[0208] Bus 1205 transmits information between various components of the device (e.g., processor 1201, memory 1202, input / output interface 1203, and communication interface 1204);

[0209] The processor 1201, memory 1202, input / output interface 1203 and communication interface 1204 are connected to each other within the device via bus 1205.

[0210] This application also provides a computer-readable storage medium storing a computer program that, when executed by a processor, implements the above-described neutron fluence rate anomaly detection method.

[0211] Memory, as a non-transitory computer-readable storage medium, can be used to store non-transitory software programs and non-transitory computer-executable programs. Furthermore, memory may include high-speed random access memory, and may also include non-transitory memory, such as at least one disk storage device, flash memory device, or other non-transitory solid-state storage device. In some embodiments, memory may optionally include memory remotely located relative to the processor, and these remote memories can be connected to the processor via a network. Examples of such networks include, but are not limited to, the Internet, intranets, local area networks, mobile communication networks, and combinations thereof.

[0212] The neutron flux anomaly detection method, device, electronic equipment, and storage medium provided in this application embodiment select a reference nuclear power reactor that matches the reactor type of the pressurized water reactor (PWR). The reference reactor's first reference count rate detection module obtains the core neutron count rate, the neutron source strength of the reference reactor and the PWR under the same operating conditions, and the concentration influence factor matching the coolant. Then, based on the above data, the first detection sensitivity of the first neutron count rate detection module, and the first reference count rate detection module's first reference sensitivity, a predicted neutron flux rate is calculated. Subsequently, the PWR is initially loaded, and the target neutron count rate for the actual reaction process is obtained using the first neutron count rate. The target neutron flux rate is then calculated based on this target neutron count rate and the first detection sensitivity. Finally, anomaly detection is performed based on the difference between the target neutron flux rate and the predicted neutron flux rate. The method in this application uses known data from a benchmark nuclear power reactor to calculate the predicted neutron flux of the current pressurized water reactor at multiple consecutive preset time steps. During the actual reaction process, it is only necessary to compare the actual target neutron flux obtained from the first neutron count rate with the predicted neutron flux to quickly determine if an anomaly exists. This method reduces reliance on human experience, significantly improves the efficiency of neutron flux anomaly detection, and avoids the shortcomings of traditional methods that rely on human experience. Through accurate prediction and comparative analysis, it can promptly detect anomalies and make adjustments, thereby improving safety during the initial fuel loading of the reactor.

[0213] The embodiments described in this application are for the purpose of more clearly illustrating the technical solutions of the embodiments of this application, and do not constitute a limitation on the technical solutions provided by the embodiments of this application. As those skilled in the art will know, with the evolution of technology and the emergence of new application scenarios, the technical solutions provided by the embodiments of this application are also applicable to similar technical problems.

[0214] Those skilled in the art will understand that the technical solutions shown in the figures do not constitute a limitation on the embodiments of this application, and may include more or fewer steps than shown, or combine certain steps, or different steps.

[0215] The device embodiments described above are merely illustrative. The units described as separate components may or may not be physically separate; that is, they may be located in one place or distributed across multiple network units. Some or all of the modules can be selected to achieve the purpose of this embodiment according to actual needs.

[0216] Those skilled in the art will understand that all or some of the steps in the methods disclosed above, as well as the functional modules / units in the systems and devices, can be implemented as software, firmware, hardware, or suitable combinations thereof.

[0217] The terms “first,” “second,” “third,” “fourth,” etc. (if present) in the specification and accompanying drawings of this application are used to distinguish similar objects and are not necessarily used to describe a specific order or sequence. It should be understood that such data can be interchanged where appropriate so that the embodiments of this application described herein can be implemented in orders other than those illustrated or described herein. Furthermore, the terms “comprising” and “having,” and any variations thereof, are intended to cover non-exclusive inclusion; for example, a process, method, system, product, or apparatus that comprises a series of steps or units is not necessarily limited to those steps or units explicitly listed, but may include other steps or units not explicitly listed or inherent to such processes, methods, products, or apparatus.

[0218] It should be understood that in this application, "at least one (item)" means one or more, and "more than" means two or more. "And / or" is used to describe the relationship between related objects, indicating that three relationships can exist. For example, "A and / or B" can represent three cases: only A exists, only B exists, and both A and B exist simultaneously, where A and B can be singular or plural. The character " / " generally indicates that the preceding and following related objects are in an "or" relationship. "At least one (item) of the following" or similar expressions refer to any combination of these items, including any combination of single or plural items. For example, at least one (item) of a, b, or c can represent: a, b, c, "a and b", "a and c", "b and c", or "a and b and c", where a, b, and c can be single or multiple.

[0219] In the several embodiments provided in this application, it should be understood that the disclosed apparatus and methods can be implemented in other ways. For example, the apparatus embodiments described above are merely illustrative; for instance, the division of the units described above is only a logical functional division, and in actual implementation, there may be other division methods. For example, multiple units or components may be combined or integrated into another system, or some features may be ignored or not executed. Furthermore, the coupling or direct coupling or communication connection shown or discussed may be through some interfaces; the indirect coupling or communication connection between apparatuses or units may be electrical, mechanical, or other forms.

[0220] The units described above as separate components may or may not be physically separate. The components shown as units may or may not be physical units; that is, they may be located in one place or distributed across multiple network units. Some or all of the units can be selected to achieve the purpose of this embodiment according to actual needs.

[0221] Furthermore, the functional units in the various embodiments of this application can be integrated into one processing unit, or each unit can exist physically separately, or two or more units can be integrated into one unit. The integrated unit can be implemented in hardware or as a software functional unit.

[0222] If the integrated unit is implemented as a software functional unit and sold or used as an independent product, it can be stored in a computer-readable storage medium. Based on this understanding, the technical solution of this application, in essence, or the part that contributes to the prior art, or all or part of the technical solution, can be embodied in the form of a software product. This computer software product is stored in a storage medium and includes multiple instructions to cause a computer device (which may be a personal computer, server, or network device, etc.) to execute all or part of the steps of the methods of the various embodiments of this application. The aforementioned storage medium includes various media capable of storing programs, such as USB flash drives, portable hard drives, read-only memory (ROM), random access memory (RAM), magnetic disks, or optical disks.

[0223] The preferred embodiments of the present application have been described above with reference to the accompanying drawings, but this does not limit the scope of the claims of the present application. Any modifications, equivalent substitutions, and improvements made by those skilled in the art without departing from the scope and substance of the embodiments of the present application shall be within the scope of the claims of the present application.

Claims

1. A method for detecting anomalies in neutron flux rate, characterized in that, This invention relates to a pressurized water reactor (PWR) nuclear power unit, which includes a pressure vessel, a core basket, and a first neutron count rate detection module. The core basket is used for nuclear reactions and is located inside the pressure vessel. Coolant is injected into the gap between the core basket and the pressure vessel. The first neutron count rate detection module is located on the inner wall of the core basket. The neutron flux anomaly detection method includes: A benchmark nuclear reactor is selected from multiple preset reference nuclear reactors based on the reactor type of the pressurized water reactor. The first reference count rate detection module obtains the core neutron count rate at multiple consecutive preset time steps; wherein, the first reference count rate detection module is the neutron count rate detection module of the reference nuclear power reactor, and the first reference count rate detection module is disposed on the inner wall of the core basket of the reference nuclear power reactor; Obtain the concentration influence factor of boron concentration matched to the coolant, the original neutron source strength of the pressurized water reactor nuclear power unit under preset operating conditions, and the reference neutron source strength of the reference nuclear power reactor under the preset operating conditions. The predicted neutron flux rate of the core basket at each preset time step is obtained by integrating and calculating the first detection sensitivity of the first neutron count rate detection module, the first reference sensitivity of the first reference count rate detection module, the core neutron count rate, the concentration influence factor, the original neutron source strength, and the reference neutron source strength. The pressurized water reactor nuclear power unit is initially loaded with fuel, and the target neutron count rate is obtained through the first neutron count rate detection module; The target neutron fluence rate is obtained by proportionally calculating the target neutron count rate and the first detection sensitivity. Anomaly detection is performed based on the difference between the target neutron flux rate and the predicted neutron flux rate. The pressurized water reactor nuclear power unit further includes a probe well pipe and a second neutron count rate detection module. The probe well pipe is located outside the pressure vessel, and the second neutron count rate detection module is disposed within the probe well pipe. Anomaly detection based on the difference between the target neutron flux rate and the predicted neutron flux rate includes: acquiring the second neutron count rate of the second neutron count rate detection module at multiple consecutive preset time steps; acquiring the pipe neutron count rate of the second reference count rate detection module at multiple consecutive preset time steps; wherein the second reference count rate detection module is the neutron count rate detection module of the reference nuclear power reactor unit, and the second reference count rate detection module is disposed within the reference nuclear power reactor unit. Inside the probe well of the quasi-nuclear power reactor; based on the second detection sensitivity of the second neutron count rate detection module, the second reference sensitivity of the second reference count rate detection module, the neutron count rate of the pipeline, the concentration influence factor, the original neutron source strength, and the reference neutron source strength, the predicted flux rate of the pipeline at each preset time step is obtained; the pressurized water reactor nuclear power reactor is initially loaded, and the detected neutron count rate is obtained through the second neutron count rate detection module; the pipeline neutron flux rate is obtained by proportional calculation based on the detected neutron count rate and the second detection sensitivity; anomaly detection is performed based on the difference between the target neutron flux rate and the predicted neutron flux rate, and the difference between the pipeline neutron flux rate and the predicted pipeline flux rate; The pressurized water reactor nuclear power unit further includes fuel assemblies, which are disposed inside the core basket. The process of calculating the pipeline neutron flux rate based on the detected neutron count rate and the second detection sensitivity includes: obtaining a fuel influence factor matched to the fuel assembly, a first container influence factor matched to the core basket, a second container influence factor matched to the pressure vessel, and a pipeline influence factor matched to the probe well pipeline; multiplying the concentration influence factor, the second detection sensitivity, the fuel influence factor, the first container influence factor, the second container influence factor, and the pipeline influence factor to obtain first intermediate data; and calculating the pipeline neutron flux rate by ratio between the detected neutron count rate and the first intermediate data.

2. The method according to claim 1, characterized in that, The step of integrating and calculating the predicted injection rate of the pipeline at each preset time step within the probe well pipeline based on the second detection sensitivity of the second neutron count rate detection module, the second reference sensitivity of the second reference count rate detection module, the pipeline neutron count rate, the concentration influence factor, the original neutron source strength, and the reference neutron source strength includes: Obtain a reference concentration factor, which is a concentration influence factor that matches the coolant boron concentration of the benchmark nuclear power reactor. For the current preset time step, the predicted neutron count rate of the pipeline is calculated proportionally based on the concentration influence factor, the reference concentration factor, the second benchmark sensitivity, the second detection sensitivity, the original neutron source strength, the benchmark neutron source strength, the preset neutron lifetime time corresponding to the pressurized water reactor nuclear power unit, the preset neutron lifetime time corresponding to the benchmark nuclear power reactor unit, and the pipeline neutron count rate. Return to the steps described above to obtain the reference concentration factor; The predicted neutron count rate for each preset time step is obtained by calculating the ratio of the second neutron count rate of the second neutron count rate detection module at the first preset time step and the preset initial pipeline flux rate to each predicted neutron count rate of the pipeline.

3. The method according to claim 2, characterized in that, The step of proportionally calculating the predicted neutron count rate of the pipeline at the current preset time step based on the concentration influence factor, the reference concentration factor, the second benchmark sensitivity, the second detection sensitivity, the original neutron source strength, the benchmark neutron source strength, the preset neutron lifetime time corresponding to the pressurized water reactor nuclear power unit, the preset neutron lifetime time corresponding to the benchmark nuclear power reactor, and the pipeline neutron count rate includes: The original unit data is obtained by multiplying the concentration influence factor, the second detection sensitivity, the original neutron source strength, and the preset neutron lifetime time corresponding to the pressurized water reactor nuclear power unit. Multiply the reference concentration factor, the second benchmark sensitivity, the benchmark neutron source strength, and the preset neutron lifetime time corresponding to the benchmark nuclear power reactor to obtain the benchmark reactor data; The predicted neutron count rate of the pipeline is obtained by calculating the ratio of the pipeline neutron count rate, the reference unit data, and the original unit data at the current preset time step.

4. The method according to claim 1, characterized in that, The step of calculating the target neutron fluence rate by proportionally calculating the target neutron count rate and the first detection sensitivity includes: Multiply the fuel influence factor, the concentration influence factor, and the first detection sensitivity to obtain reference data; The target neutron fluence rate is obtained by calculating the ratio between the target neutron count rate and the reference data.

5. The method according to claim 4, characterized in that, The step of calculating the target neutron fluence rate by comparing the target neutron count rate with the reference data includes: Obtain the temperature influence factor that matches the temperature of the coolant; The target neutron fluence rate is obtained by dividing the target neutron count rate by the product of the reference data and the temperature influence factor.

6. The method according to claim 1, characterized in that, The anomaly detection based on the difference between the target neutron fluence rate and the predicted neutron fluence rate, and the difference between the pipeline neutron fluence rate and the predicted pipeline fluence rate, includes: The difference between the target neutron fluence rate and the predicted neutron fluence rate is calculated to obtain the first neutron fluence rate difference. The difference between the neutron flux rate in the pipeline and the predicted flux rate in the pipeline is calculated to obtain the second neutron flux rate difference. If at least one of the differences between the first neutron fluence rate and the second neutron fluence rate is greater than a preset threshold, it is determined that the neutron fluence rate is abnormal.

7. The method according to claim 1, characterized in that, The step of integrating and calculating the predicted neutron flux rate of the core basket at each preset time step based on the first detection sensitivity of the first neutron count rate detection module, the first reference sensitivity of the first reference count rate detection module, the core neutron count rate, the concentration influence factor, the original neutron source strength, and the reference neutron source strength includes: Obtain a reference concentration factor, which is a concentration influence factor that matches the coolant boron concentration of the benchmark nuclear power reactor. For the current preset time step, the predicted core neutron count rate is obtained by integrating and calculating the concentration influence factor, the reference concentration factor, the first detection sensitivity, the first benchmark sensitivity, the original neutron source strength, the benchmark neutron source strength, the preset neutron lifetime time corresponding to the pressurized water reactor nuclear power unit, the preset neutron lifetime time corresponding to the benchmark nuclear power reactor unit, and the core neutron count rate. Return to the step of obtaining the reference concentration factor, calculate the core predicted neutron count rate for the next preset time step, and so on until the core predicted neutron count rate for all preset time steps is obtained; The predicted neutron flux rate for each preset time step is obtained by calculating the ratio of the first neutron count rate of the first neutron count rate detection module at the first preset time step and the preset initial core flux rate.

8. The method according to claim 7, characterized in that, The step of integrating and calculating the predicted neutron count rate of the core at the current preset time step based on the concentration influence factor, the reference concentration factor, the first detection sensitivity, the first benchmark sensitivity, the original neutron source strength, the benchmark neutron source strength, the preset neutron lifetime time corresponding to the pressurized water reactor nuclear power unit, the preset neutron lifetime time corresponding to the benchmark nuclear power reactor, and the core neutron count rate includes: The original core data are obtained by multiplying the concentration influence factor, the first detection sensitivity, the original neutron source strength, and the preset neutron lifetime time corresponding to the pressurized water reactor nuclear power unit. The reference concentration factor, the first reference sensitivity, the reference neutron source strength, and the preset neutron lifetime time corresponding to the reference nuclear power reactor are multiplied together to obtain the reference core data. The ratio of the core neutron count rate, the reference core data, and the original core data is calculated to obtain the predicted core neutron count rate at the current preset time step.

9. A neutron flux anomaly detection device, characterized in that, This device is applied to a pressurized water reactor (PWR) nuclear power unit. The PWR includes a pressure vessel, a core basket, and a first neutron count rate detection module. The core basket is used for nuclear reactions and is located inside the pressure vessel. Coolant is injected into the gap between the core basket and the pressure vessel. The first neutron count rate detection module is located on the inner wall of the core basket. The device includes: The unit screening module is used to screen a reference nuclear power reactor that matches the unit type from a plurality of preset reference nuclear power reactors according to the unit type of the pressurized water reactor nuclear power reactor; The first acquisition module is used to acquire the core neutron count rate of the first reference count rate detection module at multiple consecutive preset time steps; wherein, the first reference count rate detection module is the neutron count rate detection module of the reference nuclear power reactor, and the first reference count rate detection module is disposed on the inner wall of the core basket of the reference nuclear power reactor; The second acquisition module is used to acquire the concentration influence factor of boron concentration matched with the coolant, the original neutron source strength of the pressurized water reactor nuclear power unit under preset operating conditions, and the reference neutron source strength of the reference nuclear power reactor unit under the preset operating conditions. The prediction module is used to integrate and calculate the predicted neutron flux rate of the core basket at each preset time step based on the first detection sensitivity of the first neutron count rate detection module, the first reference sensitivity of the first reference count rate detection module, the core neutron count rate, the concentration influence factor, the original neutron source strength, and the reference neutron source strength. The third acquisition module is used to perform the initial fuel loading of the pressurized water reactor nuclear power unit and to acquire the target neutron count rate through the first neutron count rate detection module. The fluence rate calculation module is used to perform a proportional calculation based on the target neutron count rate and the first detection sensitivity to obtain the target neutron fluence rate. An anomaly detection module is used to detect anomalies based on the difference between the target neutron flux rate and the predicted neutron flux rate. The pressurized water reactor nuclear power unit further includes a probe well pipeline and a second neutron count rate detection module. The probe well pipeline is located outside the pressure vessel, and the second neutron count rate detection module is disposed within the probe well pipeline. The anomaly detection based on the difference between the target neutron flux rate and the predicted neutron flux rate includes: acquiring the second neutron count rate of the second neutron count rate detection module at multiple consecutive preset time steps; acquiring the pipeline neutron count rate of the second reference count rate detection module at multiple consecutive preset time steps; wherein the second reference count rate detection module is the neutron count rate detection module of the reference nuclear power reactor unit, and the second reference count rate detection module is configured... Within the probe well pipeline of the reference nuclear reactor unit; based on the second detection sensitivity of the second neutron count rate detection module, the second reference sensitivity of the second reference count rate detection module, the pipeline neutron count rate, the concentration influence factor, the original neutron source strength, and the reference neutron source strength, the predicted pipeline flux rate for each preset time step within the probe well pipeline is calculated; the pressurized water reactor nuclear reactor unit is initially loaded, and the detected neutron count rate is obtained through the second neutron count rate detection module; the pipeline neutron flux rate is obtained by proportional calculation based on the detected neutron count rate and the second detection sensitivity; anomaly detection is performed based on the difference between the target neutron flux rate and the predicted neutron flux rate, and the difference between the pipeline neutron flux rate and the predicted pipeline flux rate; The pressurized water reactor nuclear power unit further includes fuel assemblies, which are disposed inside the core basket. The process of calculating the pipeline neutron flux rate based on the detected neutron count rate and the second detection sensitivity includes: obtaining a fuel influence factor matched to the fuel assembly, a first container influence factor matched to the core basket, a second container influence factor matched to the pressure vessel, and a pipeline influence factor matched to the probe well pipeline; multiplying the concentration influence factor, the second detection sensitivity, the fuel influence factor, the first container influence factor, the second container influence factor, and the pipeline influence factor to obtain first intermediate data; and calculating the pipeline neutron flux rate by ratio between the detected neutron count rate and the first intermediate data.

10. An electronic device, characterized in that, The electronic device includes a memory and a processor, the memory storing a computer program, and the processor executing the computer program to implement the method according to any one of claims 1 to 8.

11. A computer-readable storage medium storing a computer program, characterized in that, When the computer program is executed by a processor, it implements the method of any one of claims 1 to 8.

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