Fault diagnosis method and system for nuclear power instrumentation and control system, and storage medium

Abstract

A fault diagnosis method and system for a nuclear power instrumentation and control system, and a storage medium. The method comprises: acquiring measured values of a sensor under test in a nuclear power instrumentation and control system (S100); on the basis of the measured values and parameters of the nuclear power instrumentation and control system, calculating reference standard values using multiple cross-validation methods (S200); on the basis of characteristics of the measured values, comparing the measured values with the reference standard values to obtain comparison results (S300); on the basis of the comparison results and preset criteria, determining an evaluation result of the sensor under test (S400); and on the basis of the evaluation result of the sensor under test, selecting a corresponding maintenance strategy (S500). The invention achieves fault diagnosis of instrumentation and control system equipment during online operation, improves the accuracy of fault diagnosis for nuclear power instrumentation and control systems, and enables effective maintenance and management of sensors.

Description

Fault diagnosis method and system for nuclear power instrument control system and storage medium

[0001] CROSS-REFERENCE TO RELATED APPLICATIONS

[0002] This application claims priority to Chinese Patent Application No. 202411034584.4, filed on July 30, 2024, the contents of which are incorporated herein by reference in their entirety. TECHNICAL FIELD

[0003] The present application belongs to the field of fault diagnosis, and particularly relates to a fault diagnosis method and system for a nuclear power instrument control system and a storage medium. BACKGROUND

[0004] The complex nuclear power instrument control system participates in the important control and protection functions of the reactor. During the power operation of the nuclear power unit, the instrument control system is in an online operation state and participates in the important control and protection functions. The instrument control system cannot be isolated for offline detection of the module. Therefore, how to realize fault diagnosis of the online running instrument control system device has been an engineering problem. SUMMARY

[0005] The present application provides a fault diagnosis method and system for a nuclear power instrument control system and a storage medium.

[0006] In a first aspect, the present application provides a fault diagnosis method for a nuclear power instrument control system, which comprises:

[0007] obtaining a measurement value of a to-be-tested sensor in the nuclear power instrument control system;

[0008] calculating a reference standard value based on a plurality of cross-validation according to the measurement value and parameters of the nuclear power instrument control system;

[0009] comparing the measurement value with the reference standard value according to a feature of the measurement value to obtain a comparison result;

[0010] determining an evaluation result of the to-be-tested sensor according to the comparison result and a preset criterion;

[0011] selecting a corresponding maintenance strategy according to the evaluation result of the to-be-tested sensor.

[0012] In some embodiments, the reference standard value includes a redundant measurement cross-validation reference standard, a test standard cross-validation reference standard, a real working condition cross-validation reference standard, and a historical same period cross-validation reference standard. Different reference standard values correspond to different working conditions. The reference standard value calculated based on a plurality of cross-validation according to the measurement value and the parameters of the nuclear power instrument control system includes:

[0013] determine a redundancy measurement cross-validation reference standard according to an average value of measurement values of non-sensor-under-test sensors;

[0014] determine a test standard cross-validation reference standard according to cross-computation of external parameters of the nuclear power instrument control system;

[0015] determine a real condition cross-validation reference standard according to cross-computation of an operating state of the nuclear power instrument control system;

[0016] determine a historical cross-validation reference standard according to cross-computation of historical data of the nuclear power instrument control system.

[0017] In some embodiments, the preset criteria include degradation criteria and failure criteria, and different working conditions correspond to different preset criteria, and the determining of the evaluation result of the sensor-under-test according to the comparison result and the preset criteria includes:

[0018] determine the degradation criteria and the failure criteria based on the nuclear power instrument control system measurement channel uncertainty construction;

[0019] when the difference between at least one of the measurement values and the corresponding reference standard value is greater than the degradation criteria, determine that the evaluation result of the sensor-under-test is degradation;

[0020] when the difference between at least one of the measurement values and the corresponding reference standard value is greater than the failure criteria, determine that the evaluation result of the sensor-under-test is failure.

[0021] In some embodiments, the nuclear power instrument control system measurement channel includes a process instrument layer, a process control layer, and an operation monitoring layer, and the nuclear power instrument control system measurement channel uncertainty construction includes:

[0022] respectively define the uncertainty of each of the channels to obtain process instrument layer uncertainty, process control layer uncertainty, and operation monitoring layer uncertainty;

[0023] determine sensor inherent uncertainty according to initial uncertainty of the sensor-under-test and initial uncertainty of the calibration instrument;

[0024] determine sensor technical uncertainty according to the position of the signal collected by the sensor-under-test and the uncertainty of each of the channels, the position of the collected signal being the position of each of the channels;

[0025] determine processing function tolerance according to the combined uncertainty obtained from the signals collected by the sensor-under-test in each of the channels;

[0026] determine sensor function tolerance according to a preset process function tolerance and the processing function tolerance.

[0027] In some embodiments, the measurement value is a value corresponding to a process quantity of the sensor to be measured, and the determination of the degradation criterion comprises:

[0028] determining a combined uncertainty of the non-sensor to be measured according to signals collected by the non-sensor to be measured in each channel under the same working condition, wherein a measurement value of the non-sensor to be measured is a value corresponding to a process quantity of the non-sensor to be measured, and the process quantity of the non-sensor to be measured is in one-to-one correspondence with the process quantity of the sensor to be measured;

[0029] determining a combined uncertainty of the sensor to be measured according to signals collected by the sensor to be measured in each channel under the same working condition;

[0030] determining a degradation criterion according to the combined uncertainty of the non-sensor to be measured and the combined uncertainty of the sensor to be measured.

[0031] In some embodiments, the determination of the failure criterion comprises:

[0032] determining a failure criterion according to a preset detection threshold coefficient and the degradation criterion.

[0033] In some embodiments, the method for determining the preset detection threshold coefficient comprises:

[0034] determining a first assumed detection threshold expression according to a detection threshold coefficient and the combined uncertainty of the sensor to be measured;

[0035] determining a second assumed detection threshold expression according to the sensor function tolerance, a difference between the detection threshold and the average value, and a ratio of the difference to the bias standard deviation;

[0036] determining the preset detection threshold coefficient according to the first assumed detection threshold expression and the second assumed detection threshold expression.

[0037] In a second aspect, the embodiments of the present application provide a fault diagnosis system of a nuclear power instrument control system, comprising:

[0038] a measurement value acquisition module configured to acquire a measurement value of a sensor to be measured in the nuclear power instrument control system;

[0039] a reference standard value calculation module configured to calculate a reference standard value based on a plurality of cross validations according to the measurement value and parameters of the nuclear power instrument control system;

[0040] a comparison module configured to compare the measurement value with the reference standard value to obtain a comparison result according to a feature of the measurement value;

[0041] an evaluation module configured to determine an evaluation result of the sensor to be measured according to the comparison result and a preset criterion.

[0042] a maintenance module configured to select a corresponding maintenance strategy according to the evaluation result of the to-be-tested sensor.

[0043] In a third aspect, an embodiment of the present application provides a terminal device, comprising a memory, a processor, and a computer program stored in the memory and executable on the processor, and the processor implements the method described above when executing the computer program.

[0044] In a fourth aspect, an embodiment of the present application provides a computer readable storage medium, which stores a computer program, and the computer program is executable on a processor to implement the method described above.

[0045] The present application compares the measurement value of the to-be-tested sensor with the reference standard value calculated based on multiple cross validations, evaluates the to-be-tested sensor according to the comparison result and the preset criterion, and selects a corresponding maintenance strategy, thereby realizing the fault diagnosis of the online running instrument control system device, improving the accuracy of the fault diagnosis of the nuclear power instrument control system, and effectively maintaining and managing the sensor. BRIEF DESCRIPTION OF DRAWINGS

[0046] In order to more clearly illustrate the technical solutions in the embodiments of the present application, the drawings needed in the embodiments or prior art description will be briefly introduced as follows. Obviously, the drawings in the following description are only some embodiments of the present application, and other drawings can be obtained by those skilled in the art without creative labor.

[0047] FIG. 1 is a structural schematic diagram of a nuclear power unit steam generator measurement module provided by an embodiment of the present application;

[0048] FIG. 2 is a schematic diagram of a fault diagnosis method of a nuclear power instrument control system provided by an embodiment of the present application;

[0049] FIG. 3 is a specific process schematic diagram of S200 in FIG. 1 provided by an embodiment of the present application;

[0050] FIG. 4 is a specific process schematic diagram of S400 in FIG. 1 provided by an embodiment of the present application;

[0051] FIG. 5 is a schematic diagram of a two-level evaluation criterion of redundant cross validation degradation and failure provided by an embodiment of the present application;

[0052] FIG. 6 is a schematic diagram of a typical measurement channel configuration of a nuclear power plant instrument control system provided by an embodiment of the present application;

[0053] FIG. 7 is a schematic diagram of the correspondence between the measurement uncertainty of a nuclear power plant instrument control system and the process and channel configuration provided by an embodiment of the present application;

[0054] Fig. 8 is a schematic diagram of a redundancy cross-validation fault diagnosis evaluation standard based on uncertainty distribution according to an embodiment of the present application;

[0055] Fig. 9 is a schematic diagram of a distribution curve of a to-be-compared measurement and a reference value according to an embodiment of the present application;

[0056] Fig. 10 is a degradation evaluation model based on uncertainty propagation rate-failure probability distribution according to an embodiment of the present application;

[0057] Fig. 11 is a failure evaluation model based on uncertainty propagation rate-failure probability distribution according to an embodiment of the present application;

[0058] Fig. 12 is a schematic diagram of a multi-form fault diagnosis method based on cross-validation according to an embodiment of the present application;

[0059] Fig. 13 is a schematic diagram of a structure of a fault diagnosis system of a nuclear power instrument control system according to an embodiment of the present application;

[0060] Fig. 14 is a schematic diagram of a structure of a terminal device according to an embodiment of the present application. DETAILED DESCRIPTION

[0061] In the following description, specific details are set forth in order to provide a thorough understanding of embodiments of the application. However, persons of ordinary skill in the art will readily appreciate that embodiments of the application can be practiced without

[0062] It should be understood that the term "comprising" when used in this specification and the appended claims specifies the presence of stated features, integers, steps, operations, elements, and / or components, but does not preclude the presence or addition of one or more other features, integers, steps, operations, elements, components, and / or groups thereof.

[0063] It should also be understood that the term "and / or" when used in this specification and the appended claims indicates that one or more of the associated listed items can be present, and, further, that one or more of such associated listed items can be realized.

[0064] As used in this application and the appended claims, the term “if’ can be construed to mean “when” or “upon” or “in response to determining” or “in response to detecting,” depending on the context. Similarly, the phrase “if it is determined” or “if [a described condition or event] is detected” can be construed to mean “upon determining” or “in response to determining” or “upon detecting [the described condition or event]” or “in response to detecting [the described condition or event],” depending on the context.

[0065] In addition, in the description of the present application and the appended claims, the terms “first”, “second”, “third”, etc. are used only to distinguish descriptions, and cannot be understood as indicating or implying relative importance.

[0066] Reference in the specification to “one embodiment” or “some embodiments” means that a particular feature, structure, or characteristic described in connection with the embodiment is included in at least one embodiment of the application. The appearances of the phrase “in one embodiment” or “in some embodiments” in various places in the specification are not necessarily all referring to the same embodiment, although it can. The terms “comprising”, “including”, “having” and their variants, mean “including but not limited to”, unless otherwise expressly specified and are not limiting.

[0067] During the power operation of the nuclear power unit, the instrument control system is in an online operation state, participates in important control and protection functions, and cannot isolate the instrument control system to perform offline detection on the module. Therefore, how to realize fault diagnosis of the online running instrument control system has been an engineering problem for a long time.

[0068] In the related art, the temperature or temperature field is used as an index for fault diagnosis of the nuclear instrument control system, but due to the limitations of scattered distribution of equipment, limited temperature measuring points, weak correlation between fault mechanism and temperature, etc., the accuracy of fault judgment by the temperature field is low and the integrity is poor.

[0069] The complex nuclear instrument control system participates in the important control and protection functions of the reactor. The function of the reactor protection system is mainly to protect the integrity of the three nuclear safety barriers, including the fuel cladding, the primary loop pressure boundary and the containment. When the operating parameters reach or endanger the threshold of the integrity of the three barriers, the protection system acts to trigger the emergency shutdown of the reactor and start the special safety facilities. For example, taking the steam generator of the nuclear power unit as an example, FIG. 1 is a structural schematic diagram of a measurement module of a steam generator of a nuclear power unit according to an embodiment of the present application.

[0070] In the initial design of the nuclear power plant, the following design criteria are met:

[0071] (1)Redundancy: The design and installation of more components and units than are required for normal operation of the system, and a criterion that a certain combination of equipment will continue to perform its normal function in the presence of a single credible random failure anywhere in the system. All secondary failures resulting from the single failure are considered an integral part of the single failure and the criterion requires that the protection function of the system not be degraded by a single failure or single event.

[0072] (2)Diversity: Achieved through functional diversity and equipment diversity against common mode failures.

[0073] (3)Independence: The independence between redundant channels, i.e. electrical and physical independence.

[0074] The nuclear power unit steam generator is such a typical device. In order to effectively represent the operating parameters of the steam generator, the important transmitters of the steam generator are designed in accordance with the redundancy setting principle, as shown in Figure 1. The steam generator liquid level measurement adopts a combination of wide / narrow range measurement, wherein L SG1 , L SG2 , L SG3 , L SG4 are narrow range liquid level transmitters, the redundancy w = 4, and mainly participate in the evaporator water level control and protection; L SGw is a wide range liquid level transmitter, which is independently designed, and mainly participates in the evaporator water level indication. The steam pipe pressure measurement adopts a triple redundancy setting mode, wherein P ST1 , P ST2 , P ST3 are steam pipe pressure transmitters, the redundancy w = 3, and mainly participate in the steam pipe pressure protection. The steam generator main feed water flow adopts a combination of wide / narrow range measurement, wherein Q FW1 , Q FW2 are wide range flow transmitters, the redundancy w = 2, and mainly participate in the evaporator water level control; Q FWn is a narrow range flow transmitter, which is independently designed, and mainly participates in the expected transient protection under non-emergency shutdown. The steam flow measurement adopts a double redundancy setting mode, wherein Q ST1 , Q ST2 are steam flow transmitters, the redundancy w = 2, and mainly participate in the evaporator water level control.

[0075] The above nuclear power unit steam generator measurement redundancy setting is an example. The characteristics of the redundancy setting of the complex nuclear power instrument control system provide a good hardware basis for online failure. The fault diagnosis method of the nuclear power instrument control system provided in the application is applied to the above complex nuclear power instrument control system, can realize the fault diagnosis of the online running instrument control system equipment, improve the accuracy of the nuclear power instrument control system fault diagnosis, and effectively maintain and manage the sensor.

[0076] Fig. 2 is a schematic diagram of a fault diagnosis method of a nuclear power instrument control system according to an embodiment of the present application, the method comprising:

[0077] S100, obtaining a measurement value of a sensor to be tested in the nuclear power instrument control system.

[0078] The fault diagnosis of the nuclear power instrument control system in the embodiments of the present application is actually a diagnosis of the sensors of the nuclear power instrument control system. The sensor to be tested is diagnosed according to the data collected by the sensor to be tested by the method provided in the present application. First, the measurement value of the sensor to be tested in the nuclear power instrument control system is obtained. The measurement value is the value corresponding to the process quantity in the process system. The process quantity I rA , I rB , …, I rW is the physical signal of the nuclear power unit process system collected in real time, such as flow, pressure, liquid level, temperature, etc. The measurement value corresponding to the process quantity I rA , I rB , …, I rW is x rA , x rB , …, x rW , such as flow value, pressure value, liquid level value, temperature value, etc.

[0079] S200, obtaining a reference standard value calculated based on multiple cross validations according to the measurement value and the parameters of the nuclear power instrument control system.

[0080] In addition to power operation conditions, the nuclear power unit also has grid peak shaving conditions, overhaul conditions, etc. The operation parameters of the instrument control system under different conditions are quite different. According to the design and installation characteristics of the complex nuclear power instrument control system and the deep integration of the unit operation conditions, multiple forms of redundant cross validation methods are researched to form a cross validation system integrating redundant measurement comparison, test standard checking, condition demand matching, and historical same period reference.

[0081] Through test verification, it is found that in order to ensure the accuracy of cross validation, the accuracy of the reference standard θ should not be lower than the accuracy of the device or measurement module to be verified when selected in engineering. In view of this, the fault method provided in the present application, which adopts the four verification forms of redundant measurement cross validation, test standard cross validation, real condition cross validation, and historical same period cross validation, can improve the reliability and accuracy of the monitoring results. Four reference standard values are calculated by the above four cross validation methods.

[0082] S300, comparing the measurement value with the reference standard value according to the characteristics of the measurement value to obtain a comparison result.

[0083] In the embodiments of the present application, the characteristics of the measurement values of the to-be-tested sensors are different due to the different types of the to-be-tested sensors, and the measurement values with different characteristics are compared with the reference standard values corresponding to the characteristics, for example, for a measurement value x rA with the redundant measurement cross-validation reference standard θ, the measurement value x rA is compared with the redundant measurement cross-validation reference standard θ to obtain a comparison result. It should be noted that different measurement values correspond to different characteristics, and the measurement value can include one or more characteristics. When multiple characteristics are included, the reference standard value corresponding to each characteristic is compared one by one to obtain the comparison result.

[0084] S400, determining an evaluation result of the to-be-tested sensor according to the comparison result and a preset criterion.

[0085] Nuclear power units have requirements for equipment availability, and unavailability of key equipment will directly lead to forced shutdown of the nuclear power unit. Therefore, how to discover existing fault signs in time before the equipment is unavailable and how to distinguish between important and urgent gradients of equipment evaluation are studied. In view of this, the present application proposes a two-level evaluation criterion of redundant cross-validation degradation and failure.

[0086] S500, selecting a corresponding maintenance strategy according to the evaluation result of the to-be-tested sensor.

[0087] In the embodiments of the present application, the corresponding maintenance strategy is formulated according to the monitoring and judgment result of S400, mainly including instrument control channel adjustment, instrument control deviation correction, instrument control whole replacement, aging trend judgment, etc.

[0088] Specifically, FIG. 3 is a specific process schematic diagram of S200 in FIG. 1 provided by the embodiments of the present application, and this part introduces how to determine four cross-validation reference standard values, including redundant measurement cross-validation reference standard, test standard cross-validation reference standard, real working condition cross-validation reference standard and historical same period cross-validation reference standard. Different reference standard values correspond to different working conditions, and the reference standard values calculated based on the measurement values and the parameters of the nuclear instrument control system include:

[0089] S201, determining a redundant measurement cross-validation reference standard according to an average value of the measurement values of the non-to-be-tested sensors.

[0090] The same measured object is provided with multiple measurement methods, such as multiple redundant measurements, wide / narrow range measurements, axial / radial measurements, etc. In the nuclear power instrument control system, multiple sensors are included, in addition to the to-be-measured sensor, there are non-to-be-measured sensors, and the redundant measurement cross verification algorithm is to remove the to-be-measured sensor, and use the measurement value or average value of other same, redundant or equivalent measurement object equipment, that is, use the measurement value or average value of the non-to-be-measured sensor as the reference standard θ R to evaluate the state of the to-be-measured sensor, which is defined as redundant measurement cross verification. The redundant measurement cross verification reference standard θ R The calculation formula is:

[0091] Where k is the number of redundant or equivalent measurement objects (i.e. non-to-be-measured sensors), x rA , x rB , …, x rW are measurement values.

[0092] S202, cross calculation according to the external parameters of the nuclear power instrument control system, to determine the test standard cross verification reference standard.

[0093] The test standard cross verification is a form of using external measurement test standard value as cross verification to evaluate the current equipment state. The test standard cross verification reference standard θ S is the external measurement test standard value, which is actually cross calculated according to the external parameters of the nuclear power instrument control system, such as the nuclear power unit heat balance flow. The heat balance flow is calculated through the nuclear power unit heat balance test, and the value is the feedwater flow calculated by the feedwater pressure, feedwater temperature and orifice differential pressure, which has high measurement accuracy. The calculation formula of the nuclear power unit heat balance flow is: θ S =Q e ;

[0094] Where C is the outflow coefficient (dimensionless), E is the progressive velocity coefficient (dimensionless), d is the orifice diameter under operating conditions, ε is the expansion factor of the fluid (dimensionless), and the incompressible fluid ε = 1, ρ is the fluid density upstream of the device, and ΔP is the pressure difference.

[0095] S203, cross calculation according to the operating state of the nuclear power instrument control system, to determine the real condition cross verification reference standard.

[0096] The real condition cross verification is a form of using the operating state of the unit process system to cross compare to evaluate the current equipment state. The real condition cross verification reference standard θ ZA certain physical reference value generated by the running state of the process system under certain working conditions, such as physical zero point check: checking the physical zero point under the nominal working condition of the measuring device, the reference value is a known physical value, such as zero flow, nominal flow, zero pressure, empty water tank liquid level, etc. S = a certain physical reference value;

[0097] S204, cross calculation is performed according to the historical data of the nuclear power instrument control system, and a historical same period cross verification reference standard is determined.

[0098] The historical same period cross verification is a form of cross comparison of historical data under the same running state to evaluate the current device state. The historical same period cross verification reference standard θ L is obtained by statistical analysis and calculation of historical data of the same type of device under the same device running state or the same system setting condition. In order to consider the deviation factors caused by system reasons, the reference standard θ L The concept of moving range is introduced, and the moving variance S L is one of the error factors, and the reference standard θ L and S L The calculation formulas are as follows:

[0099] Wherein, is the measurement value of different cycle periods and time periods, and j is the number of samples.

[0100] According to the comparison of the characteristics of the measurement value and the corresponding characteristic reference standard, the difference value is obtained, and the to-be-measured sensor is evaluated according to the difference value and the preset criterion. Specifically, FIG. 4 is a specific process diagram of S400 in FIG. 1 provided by the embodiment of the application. This part introduces how to evaluate the to-be-measured sensor according to the preset criterion. The preset criterion includes degradation criterion and failure criterion. Different working conditions correspond to different preset criteria. The evaluation result of the to-be-measured sensor is determined according to the comparison result and the preset criterion, including:

[0101] S401, based on the measurement channel uncertainty of the nuclear power instrument control system, the degradation criterion and the failure criterion are determined.

[0102] S402, when the difference between at least one of the measurement values and the corresponding reference standard value is greater than the degradation criterion, the evaluation result of the to-be-measured sensor is determined as degradation.

[0103] S403, when the difference between at least one of the measurement values and the corresponding reference standard value is greater than the failure criterion, the evaluation result of the to-be-measured sensor is determined as failure.

[0104] The value calculated by S300 in the embodiment of the present application is taken as a reference standard θ, and the device process quantity signal I rA 、 rB 、 …, rW The corresponding measurement value x rA 、 x rB 、 …, rW is compared with the reference standard θ, and the system or device running state is judged, and two-level evaluation criteria of degradation and failure are set, as shown in FIG. 5, which is a schematic diagram of two-level evaluation criteria of degradation and failure provided by the present application. In order to prevent the process quantity signal from causing interference or influence on the evaluation result due to process system fluctuation, the process quantity measurement value x rA 、 x rB 、 …, rW is the average value after the maximum value and the minimum value are removed in a certain period.

[0105] It should be noted that the corresponding degradation criteria and failure criteria are different under different working conditions. The method of the present application can process signal data collected by different sensors under the same working condition, or can process signal data collected by the same sensor under different working conditions. At least one measurement value is collected, as shown in FIG. 5, C mv1 and C mv2 are different degradation criteria corresponding to different working conditions, respectively. fv1 and C fv2 are different failure criteria corresponding to different working conditions, respectively. m1 and x m2 are different measurement values of the to-be-measured sensor corresponding to different working conditions, respectively. mv and P2 are different working conditions, for example, power is 50% and power is 100%.

[0106] The degradation criteria C mv Evaluation: If the online running device is found to satisfy the formula at the comparison points P1 and P2 after diagnosis, it is judged that the device is degraded, but the deviation does not affect the running of the process system. The maintenance strategy is that the existing state does not affect the running of the device, and the inspection and treatment are arranged in a suitable window.

[0107] The criteria C mv | x m1 - θ1 | > C mv1 ∨ | x m2 - θ2 | > C mv2 ;

[0108] The failure criteria C fvEvaluation: If the online running equipment is diagnosed to find that the comparison points P1, P2 satisfy the following formula, it is judged that the equipment fails, the deviation affects the operation of the process system, and the emergency level is high. The maintenance strategy is to immediately carry out interventional maintenance to repair the defects.

[0109] C fv Criteria: |x m1 -θ1|>C fv1 ∨|x m2 -θ2|>C fv2 ;

[0110] Specifically, based on the uncertainty of the measurement channel of the nuclear power instrument control system, C mv and C fv are calculated, a total of three parts, the first part establishes the uncertainty of the measurement channel of the nuclear power instrument control system, the second part calculates C mv , and the third part calculates C fv .

[0111] This part introduces how to build the uncertainty of the measurement channel of the nuclear power instrument control system in the first part. Figure 6 is a schematic diagram of a typical measurement channel configuration of a nuclear power plant instrument control system provided by the embodiment of the application. The nuclear power instrument control system measurement channel includes a process instrument layer, a process control layer, and an operation monitoring layer. The uncertainty of the nuclear power instrument control system measurement channel is constructed, including:

[0112] A1: Define the uncertainty corresponding to each of the channels respectively to obtain the process instrument layer uncertainty, the process control layer uncertainty, and the operation monitoring layer uncertainty.

[0113] In the embodiment of the application, as shown in Figure 6, the uncertainty corresponding to each of the channels is defined respectively, the correspondence relationship between the channel uncertainty distribution and the measurement instrument channel configuration is established, the uncertainty of the process instrument layer (layer 0 in Figure 6) is defined as U0, which is mainly the uncertainty of the measurement device. The uncertainty of the process control layer (layer 1 in Figure 6) is defined as U1, which is mainly the uncertainty of the processing channel. Each channel component has uncertainty. The uncertainty of the operation monitoring layer (layer 2 in Figure 6) is defined as U2, which is mainly the uncertainty of the display, including the roundness of the result, the data refresh threshold setting, etc.

[0114] The correspondence between the measurement channel uncertainty and the process and channel configuration is established at the same time, as shown in FIG. 7. According to the distribution of the instrument control channel equipment, the corresponding variables are defined, respectively: sensor inherent uncertainty ε(i), sensor technical uncertainty ε(t), sensor functional tolerance δ(FS), processing functional tolerance δ(FT), and process functional tolerance δ(FP). The dashed line in FIG. 7 indicates that the position will change according to the selection of the actual signal acquisition point, the short dashed line indicates the calculation of the conformity uncertainty synthesis calculation formula, and the long dashed line indicates the calculation of the non-strict conformity uncertainty synthesis calculation formula.

[0115] A2: According to the initial uncertainty of the sensor to be tested and the initial uncertainty of the calibration instrument, the sensor inherent uncertainty is determined.

[0116] The sensor inherent uncertainty ε(i) calculation formula is: ε(i) = (ε 1-1 2 + ε 1-2 2 ) 1 / 2 ;

[0117] Wherein, ε 1-1 is the initial uncertainty of the sensor, which is generally the reference uncertainty provided by the manufacturer, and ε 1-2 is the initial uncertainty of the calibration instrument, for example: the uncertainty of the multimeter, pressure generator, and resistor.

[0118] Wherein, ε cal-i is the uncertainty introduced by the calibration.

[0119] Define the calibration criterion C(c), which is equal to the sensor inherent uncertainty ε(i) and approximately equal to the sensor initial uncertainty ε 1-1 . C(c) = ε(i) ≈ ε 1-1 ;

[0120] A3: According to the position of the signal collected by the sensor to be tested and the uncertainty of each channel, the sensor technical uncertainty is determined, and the position of the signal collected by the sensor to be tested is the position of each channel.

[0121] The sensor technical uncertainty depends on the signal acquisition point, and the sensor technical uncertainty of the same instrument measurement channel will be different if the position of the signal acquisition point is different. The sensor technical uncertainty includes the influence factors of the system, and the influence refers to all interference factors applied to the sensor under normal working conditions.

[0122] If the signal acquisition point is the process instrument layer, the sensor technical uncertainty calculation formula is:

[0123] Wherein, ε1-1 ε is initial uncertainty of sensor 1-2 ε is initial uncertainty of calibrator 1-3 ε is annual electrical deviation 1-4 ε is environmental temperature influence 1-5 ε is static pressure influence, etc.

[0124] If the signal acquisition point is the operation monitoring layer, the sensor technical uncertainty calculation formula is:

[0125] If the signal acquisition point is the process control layer, the sensor technical uncertainty calculation formula is:

[0126] Wherein, ε2(i) is the technical precision of each electronic module before and after the signal acquisition point, U0 is the uncertainty of the process instrument layer, U1 is the uncertainty definition of the process control layer, and U2 is the uncertainty of the operation monitoring layer.

[0127] A4: determining a processing function tolerance according to a combined uncertainty obtained from signals collected by the to-be-tested sensor in each channel;

[0128] The function allowable tolerance δ(FT) of the channel signal processing part is the root mean square of the technical uncertainty of each electronic module in the channel, which is combined uncertainty of the signal acquisition point of the process instrument layer, the process control layer and the operation monitoring layer in the embodiment of the application. According to A1, the uncertainty of the process instrument layer is defined as U0, the uncertainty of the process control layer is defined as U1, and the uncertainty of the operation monitoring layer is defined as U2. The combined uncertainty is the uncertainty obtained from the signals collected by the to-be-tested sensor in each channel. For example, the uncertainty of the to-be-tested sensor collected in the process instrument layer is U0, and the function allowable tolerance δ(FT) is the root mean square of U0. If the uncertainty of multiple channels is collected, for example, U0, U1 and U2, the combined uncertainty is U0+U1+U2, and the function allowable tolerance δ(FT) is the root mean square of U0+U1+U2.

[0129] The calculation formula is as follows:

[0130] Wherein, ε2(j) is the technical uncertainty of each electronic module after the signal acquisition point, the uncertainty of the process control layer is defined as U1, and ε2(i) is the technical uncertainty of each electronic module before the signal acquisition point.

[0131] A5: determining a sensor function tolerance according to a preset process function tolerance and the processing function tolerance.

[0132] Process function tolerance δ(FP): Nuclear power plant design documents specify the permissible error values ​​for the channel functions of these sensors. In one possible implementation, the process function tolerance is known, for example, the Final Safety Analysis Report (FSAR criterion) makes explicit requirements on the availability of reactor protection channels.

[0133] The process functional tolerance δ(FP) is the sum of the sensor functional tolerance δ(FS) and the processing functional tolerance δ(FT), and is calculated using the following formula: δ(FP) 2 =δ(FS) 2 +δ(FT) 2 ;

[0134] Sensor functional tolerance δ(FS): The maximum allowable uncertainty is obtained by attributing all the allowable uncertainty margin of the channel function to the sensor and part of the signal processing channel. The calculation formula is as follows: δ(FS)=(δ(FP)) 2 -δ(FT) 2 ) 1 / 2 ;

[0135] Figure 8 is a schematic diagram of the redundancy cross-validation fault diagnosis and evaluation standard based on uncertainty distribution provided in the embodiments of this application. As shown in Figure 8, the relationship between sensor inherent uncertainty ε(i), sensor technical uncertainty ε(t), sensor functional tolerance δ(FS), processing functional tolerance δ(FT), and process functional tolerance δ(FP) is shown. Among them, the sensor inherent accuracy ε(i) is the smallest, and the accuracy increases sequentially from bottom to top. The process functional tolerance δ(FP) is the sum of sensor functional tolerance δ(FS) and processing functional tolerance δ(FT).

[0136] Establish a degradation criterion based on the uncertainty propagation rate-failure probability distribution. Define degradation criterion C. mv C mv To calculate the maximum permissible deviation of the instrument's measurement channel from the perspective of uncertainty propagation rate. The main purpose of degradation judgment is to detect abnormal equipment deviations, but these deviations should not affect equipment availability. Assume the quantity to be compared is I. m Its uncertainty is u Im The measured value under specific working conditions is x Im There are also W measured / computed quantities [I] rA I rB …I rW The uncertainties for measuring the same physical parameter are [u], ... IrA u IrB …u IrW ], with I m The measured values ​​under the same working conditions are [x] IrA x IrB …x IrW] The above u IrA u IrB …u IrW ] and u Im are the combined uncertainties of the sensor at the measurement channel signal acquisition point.

[0137] The proportionality factor w of W measurement / calculations is calculated respectively i :

[0138] Wherein, u Iri is the combined uncertainty of the non-sensor under test.

[0139] Then the W most reliable values of measurement / calculations θ are obtained:

[0140] Wherein, x Iri is the measurement value of the non-sensor under test, and w i is the proportionality factor.

[0141] The value of θ is the reference value of cross-validation. The combined uncertainty of W measurement / calculations is calculated simultaneously

[0142] Wherein, x Iri is the measurement value of the non-sensor under test, w i is the proportionality factor, and u Iri is the combined uncertainty of the non-sensor under test.

[0143] Therefore, the measurement value I m under comparison obeys the normal distribution N(v m ,(u Im / 1.96) 2 ), and the reference value θ obeys the normal distribution That is:

[0144] X m and θ obey the probability density curve of the normal distribution f(x), as shown in FIG. 9, which is a distribution curve diagram of the measurement under comparison and the reference value provided by the embodiment of the present application. The middle dotted line is the reference value θ. The higher the accuracy, the smaller the range of the distribution curve deviation, and the smaller the deviation from the reference value.

[0145] This part introduces how to construct the degradation criterion based on the measurement channel uncertainty of the nuclear power instrument control system in the second part. The measurement value is the value corresponding to the process quantity of the sensor under test. The determination of the degradation criterion comprises:

[0146] B1: determining the combined uncertainty of the non-sensor-to-be-tested according to the signals collected by the non-sensor-to-be-tested in each channel under the same working condition, wherein the measured value of the non-sensor-to-be-tested is the value corresponding to the process quantity of the non-sensor-to-be-tested, and the process quantity of the non-sensor-to-be-tested is in one-to-one correspondence with the process quantity of the sensor-to-be-tested;

[0147] B2: determining the combined uncertainty of the sensor-to-be-tested according to the signals collected by the sensor-to-be-tested in each channel under the same working condition;

[0148] B3: determining the degradation criterion according to the combined uncertainty of the non-sensor-to-be-tested and the combined uncertainty of the sensor-to-be-tested.

[0149] In the embodiments of the present application, it is assumed that the process quantity of the sensor-to-be-tested is I m , the measured value under a specific working condition is x Im , and the combined uncertainty is u Im . According to the process quantities [I rA I rB …I rW ] collected by the other W non-sensor-to-be-tested under the same working condition, the same physical parameter is measured, the process quantity collected by the non-sensor-to-be-tested is the same as the process quantity collected by the sensor-to-be-tested, and I m The measured values under the same working condition are [x IrA x IrB …x IrW ], the uncertainties of the non-sensor-to-be-tested are [u IrA u IrB …u IrW ], and the average of the combined uncertainties of the multiple non-sensor-to-be-tested is As shown in FIG. 10, FIG. 10 is a degradation assessment model based on the uncertainty propagation rate-fault probability distribution provided by the embodiments of the present application, and the difference Δ between the measured value x Im under a specific working condition and the reference value θ satisfies the following relationship: Δ = |x Im - θ |.

[0150] The degradation assessment criterion C mv is calculated as follows:

[0151] wherein u Im is the combined uncertainty of the sensor-to-be-tested, and is the average of the combined uncertainties of the multiple non-sensor-to-be-tested.

[0152] If Δ ≤ C mv , it indicates that the measurement module is qualified, otherwise it indicates that the measurement module I m is degraded.

[0153] This part introduces how to determine the failure criterion based on the uncertainty of the measurement channel of the nuclear power instrument control system, and the determination of the failure criterion includes:

[0154] C1: determining a failure criterion according to a preset detection threshold coefficient and the degradation criterion.

[0155] Establishing a failure criterion based on the uncertainty propagation rate-failure probability distribution to define the failure criterion C fv , C fv is the maximum allowed deviation of the instrument measurement channel calculated in reverse from the functional allowable error, and the main purpose of the failure judgment is to find deviation anomalies that affect the availability of the equipment. For I m and θ described by the formula, a certain measurement value x m of I Im satisfies the following expression: x Im =v m +a m +e m ;

[0156] Where v m is the corresponding true physical value of I m , a m is a single measurement error value subject to normal distribution, and e m is the measurement deviation caused by the failure of the measurement module.

[0157] The difference Δ between the measurement value x Im and the reference value θ satisfies: Δ=v m +a m +e m -(v m +a θ )=a m +e m -a θ ;

[0158] Where v m is the corresponding true physical value of I m , a m is a single measurement error value subject to normal distribution, e m is the measurement deviation caused by the failure of the measurement module, and a θ is the true value of the reference value θ.

[0159] The failure criterion expression is constructed as: C fv =k t C mv ;

[0160] The criterion also introduces a preset detection threshold coefficient k tcharacterizing the tolerance to the accuracy of a certain measurement function, the coefficient is related to the failure probability of the detection. Nuclear power equipment has high reliability requirements, so the detection failure probability is set to be greater than or equal to 95%, that is, if a certain element has a failure, the C fv The probability of <Delta> is greater than 95%, which is conducive to improving the accuracy of measurement.

[0161] The preset detection threshold coefficient determination method comprises the following steps:

[0162] S1: determining a first assumed detection threshold expression according to the detection threshold coefficient and the combined uncertainty of the to-be-measured sensor;

[0163] In the embodiment of the application, the detection threshold is ST, and the first assumed detection threshold expression is established: S T = k t u Im ;

[0164] Wherein, u Im is the combined uncertainty of the to-be-measured sensor.

[0165] S2: determining a second assumed detection threshold expression according to the sensor function tolerance, the difference between the detection threshold and the average value and the bias standard deviation ratio. S T = delta(FS)-n b sigma d ;

[0166] Wherein, k t is the detection threshold coefficient, sigma d is the bias standard deviation, n b is the ratio of the difference between the detection threshold and the average value to the bias standard deviation, and delta(FS) is the sensor function tolerance. sigma d = u θ ;

[0167] S3: determining the preset detection threshold coefficient according to the first assumed detection threshold expression and the second assumed detection threshold expression.

[0168] The first assumed detection threshold expression and the second assumed detection threshold expression obtained according to S1 and S2 are combined to obtain the expression of kt:

[0169] Wherein, u Im is the combined uncertainty of the to-be-measured sensor, n b is the ratio of the difference between the detection threshold and the average value to the bias standard deviation, delta(FS) is the sensor function tolerance, and u θ is the average of the combined uncertainty of the W non-to-be-measured sensors.

[0170] Because of I m If it follows a Gaussian distribution, the probability P of being detected is:

[0171] Where, n b It is the ratio of the difference between the detection threshold and the mean to the standard deviation of the deviation.

[0172] Therefore, different n can be obtained based on different fault detection rates P. b If the probability of failure in detecting faults is set to ≥95%, then P is taken as 0.95, and n is calculated. b =1.645. At this point, the calculated value of kt is:

[0173] Where δ(FS) is the sensor functional tolerance, u Im Let u be the combined uncertainty of the sensor under test. θ Let be the mean of the combined uncertainty of the W non-test sensors.

[0174] Subsequently, the failure assessment criterion C was calculated. fv :

[0175] Among them, u θ Let δ(FS) be the mean of the combined uncertainty of W non-test sensors, and let u be the sensor functional tolerance. Iri This represents the combined uncertainty of each non-test sensor.

[0176] That is, if Δ≤C fv This indicates that the measuring module I m If it is qualified, otherwise it indicates that the measuring module I is qualified. m The failure, as shown in Figure 11, is based on a failure assessment model that considers the uncertainty propagation rate and the failure probability distribution.

[0177] Referring to FIG. 12, FIG. 12 is a schematic diagram of a multi-form fault diagnosis method based on cross verification provided in an embodiment of the present application. The present application provides a fault diagnosis method for a nuclear power instrument control system. First, a two-level evaluation of redundancy cross verification degradation and failure is established to increase the equipment evaluation importance and emergency gradient, which meets the abnormal response requirements of the nuclear power unit. At the same time, the fault diagnosis monitoring is implemented in the online state to maintain the initial reliability of the equipment. Second, the evaluation criteria are calculated based on the uncertainty propagation rate-fault probability distribution method, which covers the full working condition scene of 0-100% power without additional unit state setting, and has strong implementability. At the same time, the evaluation criteria are based on the channel uncertainty, and the evaluation results have high credibility. Finally, the redundancy cross verification form is established in combination with the nuclear power unit design and state to form four forms of fusion redundancy measurement comparison, test standard checking, working condition demand matching, and historical same period reference, and the multi-form verification improves the accuracy of the evaluation.

[0178] Referring to FIG. 13, a structural schematic diagram of a multi-form fault diagnosis system based on cross verification provided in an embodiment of the present application is shown. For ease of illustration, only the parts related to the embodiments of the present application are shown.

[0179] The multi-form fault diagnosis system based on cross verification 600 includes a processor, wherein the processor is configured to execute the following program modules stored in the memory:

[0180] The measurement value acquisition module 610 is configured to acquire a measurement value of a sensor to be measured in the nuclear power instrument control system.

[0181] The reference standard value calculation module 620 is configured to calculate a reference standard value based on multiple cross verifications according to the measurement value and parameters of the nuclear power instrument control system.

[0182] The comparison module 630 is configured to compare the measurement value with the reference standard value according to a feature of the measurement value to obtain a comparison result.

[0183] The evaluation module 640 is configured to determine an evaluation result of the sensor to be measured according to the comparison result and a preset criterion.

[0184] The repair module 650 is configured to select a corresponding repair strategy according to the evaluation result of the sensor to be measured.

[0185] The reference standard value calculation module 620, the reference standard value includes redundancy measurement cross verification reference standard, test standard cross verification reference standard, real working condition cross verification reference standard, and historical same period cross verification reference standard, different reference standard values correspond to different working conditions, and the reference standard value calculated based on multiple cross verifications according to the measurement value and the parameters of the nuclear power instrument control system includes the following units:

[0186] The redundancy measurement cross-validation unit 621 determines a redundancy measurement cross-validation reference standard according to an average value of the measurement values of the non-sensor-under-test;

[0187] The test standard cross-validation unit 622 determines a test standard cross-validation reference standard by cross-computing according to external parameters of the nuclear power plant I&C system;

[0188] The real condition cross-validation unit 623 determines a real condition cross-validation reference standard by cross-computing according to an operating state of the nuclear power plant I&C system;

[0189] The history synchronization cross-validation unit 624 determines a history synchronization cross-validation reference standard by cross-computing according to historical data of the nuclear power plant I&C system.

[0190] The evaluation module 640, the preset criteria include degradation criteria and failure criteria, different working conditions correspond to different preset criteria, the evaluation result of the sensor-under-test is determined according to the comparison result and the preset criteria, and includes the following units:

[0191] The criteria determination unit 641 determines the degradation criteria and the failure criteria based on the nuclear power plant I&C system measurement channel uncertainty construction;

[0192] The degradation unit 642 determines that the evaluation result of the sensor-under-test is degradation when the difference between at least one of the measurement values and the corresponding reference standard value is greater than the degradation criteria;

[0193] The failure unit 643 determines that the evaluation result of the sensor-under-test is failure when the difference between at least one of the measurement values and the corresponding reference standard value is greater than the failure criteria.

[0194] The criteria determination unit 641, the nuclear power plant I&C system measurement channel includes a process instrument layer, a process control layer, and an operation monitoring layer, and the nuclear power plant I&C system measurement channel uncertainty construction includes the following sub-units:

[0195] The uncertainty definition unit defines the uncertainty of each channel respectively to obtain process instrument layer uncertainty, process control layer uncertainty, and operation monitoring layer uncertainty;

[0196] The sensor inherent uncertainty unit determines the sensor inherent uncertainty according to the initial uncertainty of the sensor-under-test and the initial uncertainty of the calibration instrument;

[0197] The sensor technology uncertainty unit determines sensor technology uncertainty according to positions of signals collected by the to-be-tested sensor and uncertainty corresponding to each channel, and the positions of the signals collected are positions of each channel.

[0198] The processing function tolerance unit determines processing function tolerance according to combined uncertainty obtained from signals collected by the to-be-tested sensor in each channel.

[0199] The sensor function tolerance unit determines sensor function tolerance according to preset process function tolerance and the processing function tolerance.

[0200] The degradation unit 642, the measurement value is a value corresponding to a process quantity of the to-be-tested sensor, and the determination degradation criterion includes the following subunit:

[0201] The combined uncertainty unit of the non-to-be-tested sensor determines combined uncertainty of the non-to-be-tested sensor according to signals collected by the non-to-be-tested sensor in each channel under the same working condition, the measurement value of the non-to-be-tested sensor is a value corresponding to a process quantity of the non-to-be-tested sensor, and the process quantity of the non-to-be-tested sensor is one-to-one corresponding to the process quantity of the to-be-tested sensor.

[0202] The combined uncertainty unit of the to-be-tested sensor determines combined uncertainty of the to-be-tested sensor according to signals collected by the to-be-tested sensor in each channel under the same working condition.

[0203] The degradation criterion determination unit determines degradation criterion according to the combined uncertainty of the non-to-be-tested sensor and the combined uncertainty of the to-be-tested sensor.

[0204] The failure unit 643 is specifically configured to determine failure criterion according to a preset detection threshold coefficient and the degradation criterion, and includes the following subunit:

[0205] The detection threshold expression unit determines a first assumed detection threshold expression according to a detection threshold coefficient and the degradation criterion.

[0206] The detection threshold expression unit determines a second assumed detection threshold expression according to the sensor function tolerance, a difference between the detection threshold and the average value and a ratio of the difference to the bias standard deviation.

[0207] The preset detection threshold coefficient determination unit determines the preset detection threshold coefficient according to the first assumed detection threshold expression and the second assumed detection threshold expression.

[0208] The fault diagnosis system of the nuclear power instrument control system provided by the embodiments of the present application can be applied in the foregoing method embodiments, and details are referred to the description of the method embodiments, which will not be repeated here.

[0209] FIG. 14 is a structural schematic diagram of a terminal device provided by an embodiment of the present application. As shown in FIG. 14, the terminal device 700 of this embodiment comprises at least one processor 710 (only one processor is shown in FIG. 14), a memory 720, and a computer program 721 stored in the memory 720 and capable of running on the at least one processor 710, wherein the processor 710 implements the steps in the method embodiments described above when running the computer program 721.

[0210] The terminal device 700 can be a desktop computer, a notebook computer, a palm computer, a cloud server, or the like. The terminal device can include, but is not limited to, the processor 710 and the memory 720. Those skilled in the art can understand that FIG. 14 is only an example of the terminal device 700, and does not limit the terminal device 700, which can include more or fewer components than those shown, or combine some components, or different components, for example, can also include an input / output device, a network access device, and the like.

[0211] The processor 710 can be a central processing unit (CPU), and can also be other general-purpose processors, digital signal processors (DSP), application specific integrated circuits (ASIC), field-programmable gate arrays (FPGA) or other programmable logic devices, discrete gates or transistor logic components, discrete hardware components, or the like. The general-purpose processor can be a microprocessor or can also be any conventional processor.

[0212] The memory 720 can be an internal storage unit of the terminal device 700 in some embodiments, for example, a hard disk or a memory of the terminal device 700. The memory 720 can also be an external storage device of the terminal device 700 in other embodiments, for example, a plug-in hard disk, a smart media card (SMC), a secure digital (SD) card, a flash card, or the like. Further, the memory 720 can include both the internal storage unit and the external storage device of the terminal device 700. The memory 720 is used to store an operating system, application programs, a boot loader, data, and other programs, for example, program codes of the computer program, and the like. The memory 720 can also be used to temporarily store data that has been output or will be output.

[0213] Those skilled in the art can clearly understand that, for the convenience and brevity of description, only the above-mentioned division of each functional unit and module is exemplified, and in actual application, the above-mentioned functions can be completed by different functional units and modules according to needs, that is, the internal structure of the device is divided into different functional units or modules to complete all or part of the functions described above. Each functional unit and module in the embodiment can be integrated in one processing unit, or each unit can be physically present separately, or two or more units can be integrated in one unit. The above-mentioned integrated unit can be realized in the form of hardware or software function unit. In addition, the specific name of each functional unit and module is only for the convenience of mutual distinction, and does not limit the protection scope of the present application. The specific working process of the unit and module in the above system can refer to the corresponding process in the foregoing method embodiment, which will not be described here.

[0214] In the above embodiments, the description of each embodiment has its own emphasis, and the parts not described or recorded in detail in a certain embodiment can be referred to the related description of other embodiments.

[0215] Those skilled in the art can realize that the units and algorithm steps of each example described in combination with the embodiments disclosed in the present application can be realized by electronic hardware or a combination of computer software and electronic hardware. Whether the functions are executed in hardware or software depends on the specific application and design constraints of the technical solution. Professional technicians can use different methods for each specific application to implement the described functions, but such implementation should not be considered beyond the scope of the present application.

[0216] In the embodiments provided in the present application, it should be understood that the disclosed apparatus / terminal device and method can be implemented by other means. For example, the apparatus / terminal device embodiments described above are only schematic, for example, the division of the modules or units is only a logical function division, and actual implementation can have another division manner, for example, a plurality of units or components can be combined or integrated into another system, or some features can be ignored or not executed. In addition, the coupling or direct coupling or communication connection between the displayed or discussed each other can be through some interface, indirect coupling or communication connection between the units or devices, which can be electrical, mechanical or other forms.

[0217] The units described as separate components can or can not be physically separated, and the components shown as units can or can not be physical units, that is, they can be located in one place, or they can be distributed on multiple network units. Part or all of the units can be selected according to actual needs to achieve the purpose of the embodiment.

[0218] In addition, each functional unit in each embodiment of the present application can be integrated in one processing unit, or each unit can exist physically, or two or more units can be integrated in one unit. The integrated unit can be realized in the form of hardware or in the form of a software functional unit.

[0219] The integrated module / unit, if realized in the form of a software functional unit and sold or used as an independent product, can be stored in a computer readable storage medium. Based on such understanding, all or part of the processes in the above-mentioned embodiment methods can also be completed by a computer program instructing related hardware, and the computer program can be stored in a computer readable storage medium. The computer program can implement the steps of each method embodiment when executed by a processor. The computer program includes computer program code, which can be in the form of source code, object code, executable files or some intermediate forms. The computer readable medium can include any entity or device capable of carrying the computer program code, recording medium, U disk, mobile hard disk, magnetic disk, optical disk, computer memory, read-only memory (ROM), random access memory (RAM), electrical carrier signal, telecommunication signal and software distribution medium, etc. It should be noted that the computer readable medium can include appropriate contents according to the requirements of legislation and patent practice in the jurisdiction, for example, in some jurisdictions, according to legislation and patent practice, the computer readable medium does not include electrical carrier signals and telecommunication signals.

[0220] The above-mentioned embodiment methods can also be completed by a computer program product, which, when running on a terminal device, enables the terminal device to execute the steps in each method embodiment.

[0221] The above-mentioned embodiments are only used to illustrate the technical solutions of the present application, rather than limit them. Although the present application has been described in detail with reference to the foregoing embodiments, those skilled in the art should understand that the technical solutions recorded in the foregoing embodiments can be modified, or some technical features can be replaced by equivalent ones; and these modifications or replacements do not make the essence of the corresponding technical solutions deviate from the spirit and scope of the technical solutions of the embodiments of the present application, and should be included in the protection scope of the present application.

Claims

1. A method for diagnosing a fault of a nuclear power plant instrumentation and control system, wherein, The method comprises: acquiring a measurement value of a to-be-tested sensor in a nuclear power instrument control system; calculating a reference standard value based on multiple cross validations according to the measurement value and parameters of the nuclear power instrument control system; comparing the measurement value with the reference standard value to obtain a comparison result according to a feature of the measurement value; determining an evaluation result of the to-be-tested sensor according to the comparison result and a preset criterion; selecting a corresponding maintenance strategy according to the evaluation result of the to-be-tested sensor.

2. The method of claim 1, wherein, The reference standard value comprises a redundant measurement cross validation reference standard, a test standard cross validation reference standard, a real working condition cross validation reference standard and a historical same period cross validation reference standard, different reference standard values correspond to different working conditions, and the reference standard value calculated based on multiple cross validations according to the measurement value and the parameters of the nuclear power instrument control system comprises: determining a redundant measurement cross validation reference standard according to an average value of measurement values of non-to-be-tested sensors; determining a test standard cross validation reference standard by cross calculation according to external parameters of the nuclear power instrument control system; determining a real working condition cross validation reference standard by cross calculation according to an operating state of the nuclear power instrument control system; determining a historical same period cross validation reference standard by cross calculation according to historical data of the nuclear power instrument control system.

3. The method of claim 1, wherein, The preset criterion comprises a degradation criterion and a failure criterion, different preset criteria correspond to different working conditions, and the evaluation result of the to-be-tested sensor is determined according to the comparison result and the preset criterion, which comprises: determining the degradation criterion and the failure criterion based on a nuclear power instrument control system measurement channel uncertainty construction; when a difference between at least one measurement value and a corresponding reference standard value is greater than the degradation criterion, determining that the evaluation result of the to-be-tested sensor is degradation; when a difference between at least one measurement value and a corresponding reference standard value is greater than the failure criterion, determining that the evaluation result of the to-be-tested sensor is failure.

4. The method of claim 3, wherein, The nuclear power instrument control system measurement channel comprises a process instrument layer, a process control layer and an operation monitoring layer, and the nuclear power instrument control system measurement channel uncertainty construction comprises: respectively defining an uncertainty of each channel to obtain a process instrument layer uncertainty, a process control layer uncertainty and an operation monitoring layer uncertainty; determining a sensor inherent uncertainty according to an initial uncertainty of the to-be-tested sensor and an initial uncertainty of a calibration instrument; determining a sensor technical uncertainty according to a position of a signal collected by the to-be-tested sensor and the uncertainty of each channel, the position of the collected signal being a position of each channel; determining a processing function tolerance according to a combined uncertainty obtained from signals collected by the to-be-tested sensor in each channel; determining a sensor function tolerance according to a preset process function tolerance and the processing function tolerance.

5. The method of claim 3, wherein, The measurement value is a value corresponding to a process quantity of the to-be-tested sensor, and the determination of the degradation criterion comprises: Determine a combined uncertainty of the non-sensor-to-be-tested according to signals collected by each of the channels under the same working condition, wherein a measurement value of the non-sensor-to-be-tested is a value corresponding to a process quantity of the non-sensor-to-be-tested, and the process quantity of the non-sensor-to-be-tested corresponds to the process quantity of the sensor-to-be-tested one by one; Determine a combined uncertainty of the sensor-to-be-tested according to signals collected by the sensor-to-be-tested under the same working condition; Determine a degradation criterion according to the combined uncertainty of the non-sensor-to-be-tested and the combined uncertainty of the sensor-to-be-tested.

6. The method of claim 3, wherein, The determination of the failure criterion comprises: Determine a failure criterion according to a preset detection threshold coefficient and the degradation criterion.

7. The method of claim 6, wherein, The preset detection threshold coefficient determination method comprises: Determine a first assumed detection threshold expression according to a detection threshold coefficient and the combined uncertainty of the sensor-to-be-tested; Determine a second assumed detection threshold expression according to a sensor function tolerance, a bias standard deviation, and a ratio of a difference between a detection threshold and an average value to the bias standard deviation; Determine the preset detection threshold coefficient according to the first assumed detection threshold expression and the second assumed detection threshold expression.

8. A fault diagnosis system for a nuclear power plant I&C system, wherein, Comprise: A measurement value acquisition module configured to acquire a measurement value of a sensor-to-be-tested in a nuclear power instrument control system; A reference standard value calculation module configured to calculate a reference standard value based on a plurality of cross validations according to the measurement value and parameters of the nuclear power instrument control system; A comparison module configured to compare the measurement value with the reference standard value according to a feature of the measurement value to obtain a comparison result; An evaluation module configured to determine an evaluation result of the sensor-to-be-tested according to the comparison result and a preset criterion; A repair module configured to select a corresponding repair strategy according to the evaluation result of the sensor-to-be-tested.

9. The fault diagnostic system of a nuclear power instrumentation and control system according to claim 8, wherein, The reference standard value comprises a redundant measurement cross validation reference standard, a test standard cross validation reference standard, a real working condition cross validation reference standard, and a historical same period cross validation reference standard, and different working conditions correspond to different reference standard values; the reference standard value calculation module comprises a redundant measurement cross validation unit, a test standard cross validation unit, a real working condition cross validation unit, and a historical same period cross validation unit; the redundant measurement cross validation unit is configured to determine a redundant measurement cross validation reference standard according to an average value of a measurement value of a non-sensor-to-be-tested; the test standard cross validation unit is configured to cross calculate according to external parameters of the nuclear power instrument control system to determine a test standard cross validation reference standard; the real working condition cross validation unit is configured to cross calculate according to an operating state of the nuclear power instrument control system to determine a real working condition cross validation reference standard; and the historical same period cross validation unit is configured to cross calculate according to historical data of the nuclear power instrument control system to determine a historical same period cross validation reference standard.

10. The fault diagnostic system of a nuclear power plant I&C system according to claim 8, wherein, The preset criterion comprises a degradation criterion and a failure criterion, and different working conditions correspond to different preset criteria; and the evaluation module comprises a criterion determination unit, a degradation unit, and a failure unit. The criterion determination unit is configured to determine the degradation criterion and the failure criterion based on the nuclear power instrument control system measurement channel uncertainty construction; The degradation unit is configured to determine that the evaluation result of the to-be-measured sensor is degradation when the difference between at least one of the measurement values and the corresponding reference standard value is greater than the degradation criterion; The failure unit is configured to determine that the evaluation result of the to-be-measured sensor is failure when the difference between at least one of the measurement values and the corresponding reference standard value is greater than the failure criterion.

11. The fault diagnostic system of a nuclear power instrumentation and control system according to claim 10, wherein, The nuclear power instrument control system measurement channel comprises a process instrument layer, a process control layer and an operation monitoring layer; the criterion determination unit comprises an uncertainty definition unit, a sensor inherent uncertainty unit, a sensor technical uncertainty unit, a processing function tolerance unit and a sensor function tolerance unit; the uncertainty definition unit is configured to define the uncertainty corresponding to each of the channels respectively to obtain process instrument layer uncertainty, process control layer uncertainty and operation monitoring layer uncertainty; the sensor inherent uncertainty unit is configured to determine sensor inherent uncertainty according to initial uncertainty of the to-be-measured sensor and initial uncertainty of a calibration instrument; the sensor technical uncertainty unit is configured to determine sensor technical uncertainty according to a position of a signal collected by the to-be-measured sensor in each of the channels and the uncertainty corresponding to each of the channels; The processing function tolerance unit is configured to determine processing function tolerance according to combined uncertainty obtained from signals collected by the to-be-measured sensor in each of the channels; and the sensor function tolerance unit is configured to determine sensor function tolerance according to a preset process function tolerance and the processing function tolerance.

12. The fault diagnostic system of a nuclear power instrumentation and control system according to claim 10, wherein, The measurement value is a value corresponding to a process quantity of the to-be-measured sensor; the degradation unit comprises a non-to-be-measured sensor combined uncertainty unit, a to-be-measured sensor combined uncertainty unit and a degradation criterion determination unit; the non-to-be-measured sensor combined uncertainty unit is configured to determine combined uncertainty of the non-to-be-measured sensor according to signals collected by the non-to-be-measured sensor in each of the channels under the same working condition, the measurement value of the non-to-be-measured sensor being a value corresponding to a process quantity of the non-to-be-measured sensor, the process quantity of the non-to-be-measured sensor corresponding to the process quantity of the to-be-measured sensor one by one; the to-be-measured sensor combined uncertainty unit is configured to determine combined uncertainty of the to-be-measured sensor according to signals collected by the to-be-measured sensor in each of the channels under the same working condition; The degradation criterion determination unit is configured to determine the degradation criterion according to the combined uncertainty of the non-to-be-measured sensor and the combined uncertainty of the to-be-measured sensor; The failure unit is configured to determine the failure criterion according to a preset detection threshold coefficient and the degradation criterion.

13. The fault diagnostic system of a nuclear power instrumentation and control system according to claim 12, wherein, The failure unit comprises a detection threshold expression unit and a preset detection threshold coefficient determination unit; the detection threshold expression unit is configured to determine a first assumed detection threshold expression according to a detection threshold coefficient and a combined uncertainty of the sensor to be measured, and determine a second assumed detection threshold expression according to a sensor function tolerance, a bias amount standard deviation, and a ratio of a difference between a detection threshold and an average value to the bias amount standard deviation; The preset detection threshold coefficient determination unit is configured to determine the preset detection threshold coefficient according to the first assumed detection threshold expression and the second assumed detection threshold expression.

14. A terminal device comprising a memory, a processor, and a computer program stored in the memory and executable on the processor, wherein, The processor implements the method of any one of claims 1 to 7 when executing the computer program.

15. A computer-readable storage medium storing a computer program, wherein, The computer program, when executed by the processor, implements the method of any one of claims 1 to 7.

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