Method for predicting radial creep of heavy water reactor pressure tube

Abstract

The invention particularly relates to a method for predicting radial creep of a heavy water reactor pressure tube, and belongs to the technical field of nuclear reactor physics. The method comprises the following steps: step 1, calculating the radial creep rate of each fuel rod bundle of each fuel channel pressure pipe of the heavy water reactor; 2, calculating an average inner diameter predicted value of each fuel rod bundle of each fuel channel pressure pipe of the heavy water reactor; and step 3, taking the maximum value of the average inner diameter predicted values of all fuel rod bundles of each fuel channel pressure pipe of the heavy water reactor as the inner diameter predicted value of each fuel channel pressure pipe of the heavy water reactor. The device realizes the steps of the method. Through linear fitting of the radial expansion measurement data of the heavy water reactor fuel channel pressure tube, a relational expression between the radial creep rate and the fast neutron flux at the fuel rod bundle of the heavy water reactor fuel channel pressure tube is obtained, and finally, accurate prediction of the radial creep of the pressure tube after long-term operation of a heavy water reactor is realized.

Description

Technical Field

[0001] The invention relates to the technical field of nuclear reactor physics, and in particular to a method for predicting radial creep of a heavy water reactor pressure tube. Background Art

[0002] After long-term operation of a heavy water reactor unit, pressure tube creep is one of the main aging mechanisms of thermal-hydraulic degradation of the heavy water reactor, which is mainly manifested as: axial elongation and radial expansion of the pressure tube, wall thinning and sagging bending. Among them, the radial expansion caused by pressure tube creep has a greater impact on the thermal analysis and calculation of the heavy water reactor unit after aging. Summary of the invention

[0003] The purpose of the present invention is to provide a prediction method for the radial creep of a heavy water reactor pressure tube, to predict the radial creep of the pressure tube after long-term operation of the heavy water reactor, and the prediction result provides a basis for thermal analysis and calculation after aging of the unit, so as to further ensure the safe and economical operation of the heavy water reactor unit.

[0004] In order to achieve the above object, the present invention provides the following technical solutions:

[0005] A method for predicting radial creep of a heavy water reactor pressure tube comprises the following steps:

[0006] Step 1, calculating the radial creep rate of each fuel rod bundle at each fuel channel pressure tube of the heavy water reactor;

[0007] Step 2, calculating the predicted value of the average inner diameter of each fuel rod bundle at each fuel channel pressure tube of the heavy water reactor;

[0008] Step 3: Taking the maximum value of the average inner diameter prediction value at all fuel rod bundles of each fuel channel pressure tube of the heavy water reactor as the predicted value of the inner diameter of each fuel channel pressure tube of the heavy water reactor.

[0009] As one achievable method, step 1, calculating the radial creep rate at each fuel rod bundle of each fuel channel pressure tube of a heavy water reactor, comprises the following steps:

[0010] Step 11, obtaining a fast neutron flux measurement value at each fuel rod bundle of each fuel channel pressure tube of the heavy water reactor;

[0011] Step 12, obtaining in-service measurement data of radial expansion of the pressure tube of the fuel channel of a heavy water reactor and pre-service measurement data of radial expansion of the pressure tube of the fuel channel of a heavy water reactor;

[0012] Step 13, obtaining a relationship between the radial creep rate and the fast neutron flux at the fuel rod bundle of the fuel channel pressure tube of the heavy water reactor based on the in-service measurement data of the radial expansion of the fuel channel pressure tube of the heavy water reactor and the pre-service measurement data of the radial expansion of the fuel channel pressure tube of the heavy water reactor;

[0013] Step 14: Substitute the fast neutron flux measurement value at each fuel rod bundle of each fuel channel pressure tube of the heavy water reactor into the relationship between the radial creep rate and the fast neutron flux at the fuel rod bundle of the fuel channel of the heavy water reactor to obtain the radial creep rate at each fuel rod bundle of each fuel channel pressure tube of the heavy water reactor.

[0014] As one of the feasible methods, in step 12, the in-service measurement data of radial expansion of the fuel channel pressure tube of the heavy water reactor includes the average inner diameter measurement value at each fuel rod bundle of the measured equivalent full-power operation time of the fuel channel pressure tube of the heavy water reactor; the pre-service measurement data of radial expansion of the fuel channel pressure tube of the heavy water reactor includes the average inner diameter measurement value at each fuel rod bundle of the equivalent full-power operation time of each fuel channel pressure tube of the heavy water reactor is 0.

[0015] As one achievable method, step 13, based on the in-service measurement data of the radial expansion of the heavy water reactor fuel channel pressure tube and the pre-service measurement data of the radial expansion of the heavy water reactor fuel channel pressure tube, obtains the relationship between the radial creep rate and the fast neutron flux at the fuel rod bundle of the heavy water reactor fuel channel pressure tube, including the following steps:

[0016] Calculate the radial creep rate of each fuel rod bundle at the measured equivalent full-power operation time of the fuel channel pressure tube of the heavy water reactor;

[0017] The radial creep rate of each fuel rod bundle at the equivalent full-power operation time of the fuel channel pressure tube measured in the heavy water reactor = the radial creep amount of each fuel rod bundle at the equivalent full-power operation time of the fuel channel pressure tube measured in the heavy water reactor / the average inner diameter measurement value of each fuel rod bundle at the equivalent full-power operation time of the fuel channel pressure tube measured in the heavy water reactor / the equivalent full-power operation time of the fuel channel pressure tube measured in the heavy water reactor;

[0018] The radial creep amount of each fuel rod bundle at the equivalent full-power operation time of the fuel channel pressure tube measured in the heavy water reactor = the average inner diameter measurement value of each fuel rod bundle at the equivalent full-power operation time of the fuel channel pressure tube measured in the heavy water reactor - the average inner diameter measurement value of each fuel rod bundle at the equivalent full-power operation time of the fuel channel pressure tube measured in the heavy water reactor is 0;

[0019] Taking the radial creep rate at the fuel rod bundle of the fuel channel pressure tube of a heavy water reactor as the independent variable and the fast neutron flux at the fuel rod bundle of the fuel channel pressure tube of a heavy water reactor as the dependent variable, a linear fitting was performed on the radial creep rate and fast neutron flux measured values ​​at each fuel rod bundle under the measured equivalent full-power operation time of the fuel channel pressure tube of a heavy water reactor, and the relationship between the radial creep rate and fast neutron flux at the fuel rod bundle of the fuel channel pressure tube of a heavy water reactor was obtained.

[0020] As one of the achievable methods, in step 2, the predicted value of the average inner diameter of each fuel rod bundle at each fuel channel pressure tube of the heavy water reactor is calculated according to the following formula:

[0021]

[0022] in, is the equivalent full-power operation time of the heavy water reactor fuel channel pressure tube, which is equal to 0; Predicting the equivalent full power operation time for heavy water reactor fuel channel pressure tubes; Predicting the equivalent full power operation time for heavy water reactor fuel channel pressure tubes The average inner diameter at the lower fuel bundle; Equivalent full-power operation time of the fuel channel pressure tube of the heavy water reactor The average inner diameter at the lower fuel bundle; is the radial creep rate of the fuel rod bundle in the pressure tube of the heavy water reactor fuel channel.

[0023] Beneficial technical effects of the present invention:

[0024] The method for predicting radial creep of a heavy water reactor pressure tube of the present invention obtains a relationship between radial creep rate and fast neutron flux at a fuel rod bundle of a fuel channel pressure tube of a heavy water reactor fuel by linear fitting of radial expansion measurement data of the fuel channel pressure tube of the heavy water reactor fuel, and finally realizes accurate prediction of radial creep of the pressure tube after long-term operation of the heavy water reactor. The prediction result provides a basis for thermal analysis calculation after aging of the unit and calculation of regional overpower protection shutdown setting value, and further ensures safe and economical operation of the heavy water reactor unit. BRIEF DESCRIPTION OF THE DRAWINGS

[0025] Figure 1 A schematic diagram of a fuel channel for a heavy water reactor;

[0026] Figure 2 Schematic diagram of the irradiation creep process of heavy water reactor pressure tubes.

[0027] In the figure, 1. Inlet header; 2. Outlet header; 3. Pressure tube; 4. Fuel channel; 5. Fuel rod bundle; 6. Core. DETAILED DESCRIPTION

[0028] Unless otherwise defined, all technical and scientific terms used herein have the same meaning as those commonly understood by technicians in the technical field of this application; the terms used in the specification of the application are only for the purpose of describing specific embodiments and are not intended to limit this application; the term "including" and its variations in the specification and claims of this application and the above-mentioned drawings are intended to cover non-exclusive inclusions. The terms "first", "second", etc. in the specification and claims of this application or the above-mentioned drawings are used to distinguish different objects, not to describe a specific order.

[0029] Reference to "embodiments" herein means that a particular feature, structure, or characteristic described in conjunction with the embodiments may be included in at least one embodiment of the present application. The appearance of the phrase in various locations in the specification does not necessarily refer to the same embodiment, nor is it an independent or alternative embodiment that is mutually exclusive with other embodiments. It is explicitly and implicitly understood by those skilled in the art that the embodiments described herein may be combined with other embodiments.

[0030] The technical solution of the present invention is clearly and completely described below in conjunction with the accompanying drawings and specific embodiments.

[0031] like Figure 1 As shown, the core 6 of the CANDU 6 heavy water reactor is composed of a plurality of horizontal fuel channels 4. The fuel channel 4 is composed of a pressure tube 3 and two end parts; the pressure tube 3 is filled with coolant and 12 fuel rod bundles 5; the end parts are the extensions of the pressure tube 3 outside the core 6, which are used to connect the inlet header 1 and the outlet header 2 of the coolant. The coolant in the pressure tube 3 cools the fuel rod bundle 5. As the unit is running, the pressure tube 3 in the core 6 will produce radial creep, thereby affecting the cooling of the fuel rod bundle 5 by the coolant. The radial creep of the pressure tube 3 is mainly related to fast neutron irradiation, irradiation creep of the pressure tube 3, coolant temperature and coolant pressure.

[0032] 1. Irradiation creep of pressure tube

[0033] like Figure 2 As shown in the figure, the irradiation creep of the pressure tube is divided into three stages: the first stage is the transient stage, corresponding to the directional arrangement of dislocation rings. The material of the pressure tube is zirconium alloy, which is a crystalline material. When the pressure tube is exposed to high temperature irradiation, the vacancies in the crystal gather and collapse to form annular edge dislocation lines, that is, dislocation rings; the second stage is the steady-state stage, corresponding to dislocation slip and climb. Dislocation slip refers to the movement of dislocations along the slip plane in the crystal, and dislocation climb refers to the up and down movement of dislocations perpendicular to the slip plane in the crystal. Dislocation slip and dislocation climb are two modes of dislocation movement; the third stage is the unstable stage.

[0034] The first stage usually lasts less than 5 years, which is a very short time relative to the life of the pressure pipe, and its effect on the radial creep of the pressure pipe can be ignored; the third stage occurs outside the life of the pressure pipe and is not considered in the present invention; the irradiation creep of the pressure pipe is mainly based on dislocation slip, that is, the second stage is the main one. Since the external force is constant, the radial creep of the pressure pipe caused by dislocation slip conforms to the linear law.

[0035] The radial creep of the pressure tube affects the cooling of the fuel rod bundle by the coolant in the middle and late stages, that is, in the period after the first stage. During this period, the radial creep rate of the pressure tube is linearly related to the full-power operation time. Therefore, it can be considered that the relationship between the radial creep rate of the pressure tube and the full-power operation time satisfies the linear law.

[0036] 2. Coolant temperature and coolant pressure

[0037] The core of the heavy water reactor has 380 fuel channels, each with 12 fuel rod bundles. During the full power operation of the heavy water reactor, the inlet coolant temperature of each fuel channel in the heavy water reactor core is basically the same, and the outlet temperature deviation of each fuel channel in the heavy water reactor core is within 10°C.

[0038] After calculation, the relative standard deviation of the coolant temperature at the same fuel bundle in different fuel channels does not exceed 1%, see Table 1. Therefore, it is approximately assumed that the coolant temperature at the same fuel bundle in different fuel channels is constant.

[0039] Table 1 Coolant temperature at the same fuel bundle in different fuel channels of heavy water reactor (℃)

[0040]

[0041] The HWR outlet header pressure is a controlled parameter. During normal operation of the HWR unit, a fixed value control is adopted, and the set value is 9.89MPa. Although the pressure drop of each fuel channel is different, it is estimated that the relative standard deviation of the coolant pressure at the same fuel bundle in different fuel channels does not exceed 1.8%, as shown in Table 2. Therefore, it can be approximately considered that the coolant pressure at the same fuel bundle in different fuel channels is the same.

[0042] Table 2 Coolant pressure at the same fuel bundle in different fuel channels of heavy water reactor (MPa)

[0043]

[0044] Among them, the standard deviation of the parameters at the same fuel bundle in different fuel channels is:

[0045]

[0046] Among them, SD is the standard deviation of the parameters at the same fuel bundle in different fuel channels; is the average value of the parameters at the same fuel bundle in different fuel channels; is the number of fuel channels; It is the parameter of the same fuel bundle in different fuel channels;

[0047] The relative standard deviation of the parameters at the same fuel bundle in different fuel channels is:

[0048]

[0049] Where RSD is the relative standard deviation of the parameters at the same fuel bundle in different fuel channels; is the standard deviation of the parameters at the same fuel bundle in different fuel channels; is the average value of the parameters at the same fuel bundle in different fuel channels;

[0050] The parameters at the same fuel bundle in different fuel channels include the coolant pressure and the coolant temperature at the same fuel bundle in different fuel channels.

[0051] 3. Fast neutron irradiation

[0052] Fast neutron irradiation is the main factor causing radial creep of pressure tubes. Different fast neutron integrated fluxes cause different radial creep amounts. For a long-term operating heavy water reactor core, because the core is always being refueled, and the liquid area control system is always involved in the control to eliminate power deviations to ensure that the power of each area of ​​the core is within the design requirements, in the long run, the heavy water reactor core conforms to the time-averaged model, and the fast neutron flux at each fuel rod bundle in all fuel channels can be calculated by the physical program, so the radial creep at each fuel rod bundle in all fuel channels can be calculated.

[0053] The present invention adopts a hierarchical fitting method, and divides each fuel channel into 12 layers according to the actual arrangement of the fuel rod bundle in the fuel channel, and eliminates the influence of the coolant temperature and pressure on the radial creep of the pressure tube. In this way, according to the mechanism of radial strain, the radial creep amount of the pressure tube at each fuel rod bundle in each fuel channel is proportional to the fast neutron integrated flux, and the fast neutron integrated flux is equal to the product of the fast neutron flux and the equivalent full-power operation time. According to the definition of radial creep amount, the radial creep amount is equal to the radial creep rate multiplied by the equivalent full-power operation time, so the radial creep rate is proportional to the fast neutron flux. According to this conclusion, when the measurement data of the inner diameter of the pressure tube of a limited number of fuel channels and the fast neutron flux at each fuel rod bundle of each fuel channel given by the physical program are known, the radial creep amount at each fuel rod bundle of all pressure tubes can be predicted.

[0054] The inner diameters of all fuel channel pressure tubes were measured before the heavy water reactor was put into commercial operation. After commercial operation, a certain number of fuel channels will be selected during some overhauls to measure the inner diameters of their pressure tubes to analyze the radial creep. Based on the measured pressure tube radial creep data, the radial creep rate at the pressure tube fuel rod bundle can be obtained through linear fitting, and the magnitude of the pressure tube radial creep can be predicted.

[0055] In actual engineering, the use of the maximum average inner diameter can effectively reduce the local fluctuations and errors caused by circumferential measurement. The maximum average inner diameter refers to the maximum value of the circumferential measurement average value when the inner diameter of a single pressure tube is measured along the axial direction. Specifically, the inner diameter of the pressure tube at the position of a fuel rod bundle is measured multiple times in the circumferential direction, and the average value of the multiple circumferential measurement values ​​is used as the average inner diameter of the pressure tube at the position of the fuel rod bundle; repeat the above steps until the average inner diameter of the pressure tube at 12 fuel rod bundles is obtained, and take the maximum value as the inner diameter measurement value of the pressure tube.

[0056] This embodiment provides a method for predicting radial creep of a heavy water reactor pressure tube, comprising the following steps:

[0057] Step 1, calculating the radial creep rate of each fuel rod bundle at each fuel channel pressure tube of the heavy water reactor;

[0058] Step 2, calculating the predicted value of the average inner diameter of each fuel rod bundle at each fuel channel pressure tube of the heavy water reactor;

[0059] Step 3: Taking the maximum value of the average inner diameter prediction value at all fuel rod bundles of each fuel channel pressure tube of the heavy water reactor as the predicted value of the inner diameter of each fuel channel pressure tube of the heavy water reactor.

[0060] In this embodiment, as one of the achievable methods, step 1, calculating the radial creep rate at each fuel rod bundle of each fuel channel pressure tube of the heavy water reactor, includes the following steps:

[0061] Step 11, obtaining a fast neutron flux measurement value at each fuel rod bundle of each fuel channel pressure tube of the heavy water reactor;

[0062] Step 12, obtaining in-service measurement data of radial expansion of the pressure tube of the fuel channel of a heavy water reactor and pre-service measurement data of radial expansion of the pressure tube of the fuel channel of a heavy water reactor;

[0063] Step 13, obtaining a relationship between the radial creep rate and the fast neutron flux at the fuel rod bundle of the fuel channel pressure tube of the heavy water reactor based on the in-service measurement data of the radial expansion of the fuel channel pressure tube of the heavy water reactor and the pre-service measurement data of the radial expansion of the fuel channel pressure tube of the heavy water reactor;

[0064] Step 14: Substitute the fast neutron flux measurement value at each fuel rod bundle of each fuel channel pressure tube of the heavy water reactor into the relationship between the radial creep rate and the fast neutron flux at the fuel rod bundle of the fuel channel of the heavy water reactor to obtain the radial creep rate at each fuel rod bundle of each fuel channel pressure tube of the heavy water reactor.

[0065] In this embodiment, as one of the achievable methods, in step 12, the in-service measurement data of radial expansion of the fuel channel pressure tube of the heavy water reactor includes the average inner diameter measurement value of each fuel rod bundle at the equivalent full-power operation time of the fuel channel pressure tube of the heavy water reactor;

[0066] The pre-service measurement data of radial expansion of fuel channel pressure tubes of heavy water reactors include the average inner diameter measurement value of each fuel rod bundle at the equivalent full power operation time of each fuel channel pressure tube of the heavy water reactor is 0.

[0067] In this embodiment, as one of the achievable methods, step 13, based on the in-service measurement data of the radial expansion of the heavy water reactor fuel channel pressure tube and the pre-service measurement data of the radial expansion of the heavy water reactor fuel channel pressure tube, obtains the relationship between the radial creep rate and the fast neutron flux at the fuel rod bundle of the heavy water reactor fuel channel pressure tube, including the following steps:

[0068] Calculate the radial creep rate of each fuel rod bundle at the measured equivalent full-power operation time of the fuel channel pressure tube of the heavy water reactor;

[0069] The radial creep rate of each fuel rod bundle at the equivalent full-power operation time of the fuel channel pressure tube measured in the heavy water reactor = the radial creep amount of each fuel rod bundle at the equivalent full-power operation time of the fuel channel pressure tube measured in the heavy water reactor / the average inner diameter measurement value of each fuel rod bundle at the equivalent full-power operation time of the fuel channel pressure tube measured in the heavy water reactor / the equivalent full-power operation time of the fuel channel pressure tube measured in the heavy water reactor;

[0070] The radial creep amount of each fuel rod bundle at the equivalent full-power operation time of the fuel channel pressure tube measured in the heavy water reactor = the average inner diameter measurement value of each fuel rod bundle at the equivalent full-power operation time of the fuel channel pressure tube measured in the heavy water reactor - the average inner diameter measurement value of each fuel rod bundle at the equivalent full-power operation time of the fuel channel pressure tube measured in the heavy water reactor is 0;

[0071] Taking the radial creep rate at the fuel rod bundle of the fuel channel pressure tube of a heavy water reactor as the dependent variable and the fast neutron flux at the fuel rod bundle of the fuel channel pressure tube of a heavy water reactor as the independent variable, a linear fitting was performed on the radial creep rate and fast neutron flux measured values ​​at each fuel rod bundle under the measured equivalent full-power operation time of the fuel channel pressure tube of a heavy water reactor, and the relationship between the radial creep rate and fast neutron flux at the fuel rod bundle of the fuel channel pressure tube of a heavy water reactor was obtained.

[0072] The following example illustrates how to calculate the radial creep rate at each fuel rod bundle in each fuel channel pressure tube of a heavy water reactor.

[0073] As shown in Tables 3 to 7, the in-service measurement data of radial expansion of the fuel channel pressure tube of the heavy water reactor includes the average inner diameter measurement value of each fuel rod bundle at the measured equivalent full-power operation time of the fuel channel pressure tube of the heavy water reactor, and the pre-service measurement data of radial expansion of the fuel channel pressure tube of the heavy water reactor includes the average inner diameter measurement value of each fuel rod bundle at the equivalent full-power operation time of each fuel channel pressure tube of the heavy water reactor is 0;

[0074] According to the pre-service measurement data of radial expansion of the fuel channel pressure tube of the heavy water reactor and the in-service measurement data of radial expansion of the fuel channel pressure tube of the heavy water reactor, the radial creep rate and fast neutron flux at each fuel rod bundle under the measured equivalent full-power operation time of the fuel channel pressure tube of the heavy water reactor are calculated, and the measured equivalent full-power operation time is at least one of 22320h, 57400h and 97210h;

[0075] The average inner diameter of the fuel channel pressure tube of the heavy water reactor has been measured at the jth fuel rod bundle under the equivalent full power operation time. The unit is mm. The average inner diameter of the fuel channel pressure tube of the heavy water reactor at the jth fuel rod bundle under the equivalent full power operation time of 0 is The unit is mm. The fast neutron flux at the jth fuel bundle under the equivalent full power operation time of the fuel channel pressure tube of the heavy water reactor has been measured. Indicated in cm -2 ·s -1 ; The radial creep of the fuel channel pressure tube of the heavy water reactor has been measured at the jth fuel bundle under the equivalent full power operation time. Indicated in mm. ;

[0076] The measured equivalent full power operation time of the fuel channel pressure tube of the heavy water reactor is expressed as EFPH, in hours; the radial creep rate at the jth fuel rod bundle under the measured equivalent full power operation time of the fuel channel pressure tube of the heavy water reactor is expressed as Indicated in mm / h. / EPFH; j = 1, 2, ..., m, for CANDU 6 heavy water reactor, m = 12;

[0077] Taking the radial creep rate at the fuel rod bundle of the fuel channel pressure tube of the heavy water reactor as the dependent variable and the fast neutron flux at the fuel rod bundle of the fuel channel pressure tube of the heavy water reactor as the independent variable, a linear fitting is performed on the radial creep rate and fast neutron flux measured values ​​at each fuel rod bundle under the measured equivalent full power operation time of the fuel channel pressure tube of the heavy water reactor, and the relationship between the radial creep rate and fast neutron flux at the fuel rod bundle of the fuel channel pressure tube of the heavy water reactor is obtained;

[0078] The fast neutron flux measurement value at each fuel rod bundle of each fuel channel pressure tube of the heavy water reactor is substituted into the relationship between the radial creep rate and the fast neutron flux at the fuel rod bundle of the fuel channel of the heavy water reactor, and the radial creep rate at each fuel rod bundle of each fuel channel pressure tube of the heavy water reactor is obtained.

[0079] Table 3 Average inner diameter of each fuel rod bundle at the equivalent full power operation time of the fuel channel pressure tube of the heavy water reactor (mm)

[0080]

[0081]

[0082] Table 4 Measured fast neutron flux values ​​at the fuel rod bundle of the fuel channel pressure tube of the heavy water reactor at the equivalent full power operation time (cm -2 ·s -1 )

[0083]

[0084] Table 5 The measured fast neutron flux values ​​at the fuel rod bundle of the fuel channel pressure tube of the heavy water reactor at the equivalent full power operation time (cm -2 ·s -1 )

[0085]

[0086] Table 6 Radial creep of each fuel rod bundle at equivalent full power operation time of fuel channel pressure tube of heavy water reactor (mm)

[0087]

[0088] Table 7 Radial creep rate of each fuel rod bundle at the equivalent full power operation time of the fuel channel pressure tube of the heavy water reactor (mm / h)

[0089]

[0090] In this embodiment, as one of the achievable methods, in step 2, the predicted value of the average inner diameter of each fuel rod bundle at each fuel channel pressure tube of the heavy water reactor is calculated according to the following formula:

[0091]

[0092] in, is the equivalent full-power operation time of the heavy water reactor fuel channel pressure tube, which is equal to 0; Predicting the equivalent full power operation time for heavy water reactor fuel channel pressure tubes; Predicting the equivalent full power operation time for heavy water reactor fuel channel pressure tubes The average inner diameter at the lower fuel bundle; Equivalent full-power operation time of the fuel channel pressure tube of the heavy water reactor The average inner diameter at the lower fuel bundle; is the radial creep rate of the fuel rod bundle in the pressure tube of the heavy water reactor fuel channel.

[0093] The following example illustrates how to calculate the pressure tubes of a heavy water reactor fuel channel.

[0094] Table 8 shows the fast neutron flux measurement values ​​at the fuel rod bundle of the fuel channel pressure tube of the heavy water reactor, expressed as BFF j It represents the fast neutron flux measurement value at the jth fuel bundle of the heavy water reactor fuel channel pressure tube, in cm -2 ·s -1 , j = 1, 2, ..., m, for CANDU 6 heavy water reactor, m = 12;

[0095] Table 9 shows the radial creep rate of the fuel rod bundle at the fuel channel pressure tube of the heavy water reactor, expressed as CPR j represents the radial creep rate at the jth fuel rod bundle of the fuel channel pressure tube of a heavy water reactor, in units of mm / h; j=1, 2, ..., m; for the CANDU 6 heavy water reactor, m=12; the radial creep rate at the fuel rod bundle of the fuel channel pressure tube of a heavy water reactor is obtained by substituting the fast neutron flux measurement value at the fuel rod bundle of the fuel channel pressure tube of the heavy water reactor into the relationship between the radial creep rate and the fast neutron flux at the fuel rod bundle of the fuel channel pressure tube of the heavy water reactor;

[0096] Table 10 shows the predicted average inner diameter of the fuel rod bundle at the fuel channel pressure tube of the heavy water reactor under the predicted equivalent full power operation time, with D jrepresents the predicted value of the average inner diameter of the fuel channel pressure tube at the jth fuel rod bundle under the predicted equivalent full-power operation time of the heavy water reactor fuel channel, in mm; j=1, 2, ..., m; for the CANDU 6 heavy water reactor, m=12; the predicted value of the average inner diameter of the fuel rod bundle under the predicted equivalent full-power operation time of the heavy water reactor fuel channel pressure tube is calculated according to the following formula:

[0097]

[0098] in, is the equivalent full-power operation time of the heavy water reactor fuel channel pressure tube, which is equal to 0; Predict the equivalent full-power operation time for the heavy water reactor fuel channel pressure tube, in hours; Predicting the equivalent full power operation time for heavy water reactor fuel channel pressure tubes The average inner diameter of the lower fuel bundle, in mm; Equivalent full-power operation time of the fuel channel pressure tube of the heavy water reactor The average inner diameter of the lower fuel bundle, in mm; is the radial creep rate of the fuel rod bundle at the pressure tube of the heavy water reactor fuel channel, in mm / h;

[0099] Table 11 shows the predicted inner diameter of the heavy water reactor fuel channel pressure tube under the predicted equivalent full-power operation time; the maximum value of the average predicted inner diameter at all fuel rod bundles under the predicted equivalent full-power operation time of the heavy water reactor fuel channel pressure tube is taken as the predicted inner diameter of the heavy water reactor fuel channel pressure tube under the predicted equivalent full-power operation time.

[0100] The average inner diameter measurement value of each fuel rod bundle at the predicted equivalent full-power operation time of the heavy water reactor fuel channel pressure tube is measured, and the maximum value of the average inner diameter measurement value of all fuel rod bundles at the predicted equivalent full-power operation time of the heavy water reactor fuel channel pressure tube is taken as the measured inner diameter of the heavy water reactor fuel channel pressure tube at the predicted equivalent full-power operation time.

[0101] Compared with the measured inner diameter of the heavy water reactor fuel channel pressure tube under the predicted equivalent full-power operating time, the predicted inner diameter error of the heavy water reactor fuel channel pressure tube under the predicted equivalent full-power operating time does not exceed 0.5%. The present invention can accurately predict the inner diameter of the heavy water reactor pressure tube.

[0102] Table 8 Fast neutron flux measurements at the fuel rod bundle of the heavy water reactor fuel channel pressure tube (cm -2 ·s -1 )

[0103]

[0104] Table 9 Radial creep rate at the fuel rod bundle of the heavy water reactor fuel channel pressure tube (mm / h)

[0105]

[0106] Table 10 Predicted values ​​of average inner diameter of fuel rod bundle at equivalent full power operation time of heavy water reactor fuel channel pressure tube (mm)

[0107]

[0108] Table 11 Predicted inner diameter of heavy water reactor fuel channel pressure tube at equivalent full power operation time (mm)

[0109]

[0110] As an implementation of the above method, the present invention provides an embodiment of a device for predicting radial creep of a heavy water reactor pressure tube. The device embodiment corresponds to the embodiment of the above method, and the device can be specifically applied to various electronic devices.

[0111] The device for predicting radial creep of a heavy water reactor pressure tube according to this embodiment includes:

[0112] The first calculation module is used to calculate the radial creep rate of each fuel rod bundle at each fuel channel pressure tube of the heavy water reactor;

[0113] The second calculation module is used to calculate the average inner diameter prediction value of each fuel rod bundle at each fuel channel pressure tube of the heavy water reactor;

[0114] The prediction module is used to take the maximum value of the average inner diameter prediction value of all fuel rod bundles of each fuel channel pressure tube of the heavy water reactor as the predicted value of the inner diameter of each fuel channel pressure tube of the heavy water reactor.

[0115] The above-mentioned inner diameter refers to the diameter.

[0116] The above-mentioned embodiments only express several implementation methods of the present invention, and the description thereof is relatively specific and detailed, but it cannot be understood as limiting the scope of the patent of the present invention. It should be pointed out that, for ordinary technicians in this field, several variations and improvements can be made without departing from the concept of the present invention, which all belong to the protection scope of the present invention. Therefore, the protection scope of the patent of the present invention shall be subject to the attached claims.

Claims

1. A method for predicting radial creep of a heavy water reactor pressure tube, characterized in that: The steps include: Step 1, calculating the radial creep rate of each fuel rod bundle at each fuel channel pressure tube of the heavy water reactor; Step 2, calculating the predicted value of the average inner diameter of each fuel rod bundle at each fuel channel pressure tube of the heavy water reactor; Step 3, taking the maximum value of the average inner diameter prediction value of all fuel rod bundles of each fuel channel pressure tube of the heavy water reactor as the predicted value of the inner diameter of each fuel channel pressure tube of the heavy water reactor; Wherein, the step 1, calculating the radial creep rate at each fuel rod bundle of each fuel channel pressure tube of the heavy water reactor, comprises the following steps: Step 11, obtaining a fast neutron flux measurement value at each fuel rod bundle of each fuel channel pressure tube of the heavy water reactor; Step 12, obtaining in-service measurement data of radial expansion of the pressure tube of the fuel channel of a heavy water reactor and pre-service measurement data of radial expansion of the pressure tube of the fuel channel of a heavy water reactor; Step 13, obtaining a relationship between the radial creep rate and the fast neutron flux at the fuel rod bundle of the fuel channel pressure tube of the heavy water reactor based on the in-service measurement data of the radial expansion of the fuel channel pressure tube of the heavy water reactor and the pre-service measurement data of the radial expansion of the fuel channel pressure tube of the heavy water reactor; Step 14: Substitute the fast neutron flux measurement value at each fuel rod bundle of each fuel channel pressure tube of the heavy water reactor into the relationship between the radial creep rate and the fast neutron flux at the fuel rod bundle of the fuel channel of the heavy water reactor to obtain the radial creep rate at each fuel rod bundle of each fuel channel pressure tube of the heavy water reactor.

2. The method for predicting radial creep of a heavy water reactor pressure tube according to claim 1, characterized in that: In step 12, the in-service measurement data of the radial expansion of the heavy water reactor fuel channel pressure tube includes the average inner diameter measurement value at each fuel rod bundle of the measured equivalent full-power operation time of the heavy water reactor fuel channel pressure tube; the pre-service measurement data of the radial expansion of the heavy water reactor fuel channel pressure tube includes the average inner diameter measurement value at each fuel rod bundle of each fuel channel pressure tube of the heavy water reactor when the equivalent full-power operation time is 0.

3. The method for predicting radial creep of a heavy water reactor pressure tube according to claim 1, characterized in that: Step 13, obtaining a relationship between the radial creep rate and the fast neutron flux at the fuel rod bundle of the fuel channel pressure tube of the heavy water reactor according to the in-service measurement data of the radial expansion of the fuel channel pressure tube of the heavy water reactor and the pre-service measurement data of the radial expansion of the fuel channel pressure tube of the heavy water reactor, comprising the following steps: Calculate the radial creep rate of each fuel rod bundle at the measured equivalent full-power operation time of the fuel channel pressure tube of the heavy water reactor; The radial creep rate of each fuel rod bundle at the equivalent full-power operation time of the fuel channel pressure tube measured in the heavy water reactor = the radial creep amount of each fuel rod bundle at the equivalent full-power operation time of the fuel channel pressure tube measured in the heavy water reactor / the average inner diameter measurement value of each fuel rod bundle at the equivalent full-power operation time of the fuel channel pressure tube measured in the heavy water reactor / the equivalent full-power operation time of the fuel channel pressure tube measured in the heavy water reactor; The radial creep amount of each fuel rod bundle at the equivalent full-power operation time of the fuel channel pressure tube measured in the heavy water reactor = the average inner diameter measurement value of each fuel rod bundle at the equivalent full-power operation time of the fuel channel pressure tube measured in the heavy water reactor - the average inner diameter measurement value of each fuel rod bundle at the equivalent full-power operation time of the fuel channel pressure tube measured in the heavy water reactor is 0; Taking the radial creep rate at the fuel rod bundle of the fuel channel pressure tube of a heavy water reactor as the independent variable and the fast neutron flux at the fuel rod bundle of the fuel channel pressure tube of a heavy water reactor as the dependent variable, a linear fitting was performed on the radial creep rate and fast neutron flux measured values ​​at each fuel rod bundle under the measured equivalent full-power operation time of the fuel channel pressure tube of a heavy water reactor, and the relationship between the radial creep rate and fast neutron flux at the fuel rod bundle of the fuel channel pressure tube of a heavy water reactor was obtained.

4. The method for predicting radial creep of a heavy water reactor pressure tube according to claim 1, characterized in that: Step 2: Calculate the predicted value of the average inner diameter of each fuel rod bundle in each fuel channel pressure tube of the heavy water reactor according to the following formula: in, is the equivalent full-power operation time of the heavy water reactor fuel channel pressure tube, which is equal to 0; Predicting the equivalent full power operation time for heavy water reactor fuel channel pressure tubes; Predicting the equivalent full power operation time for heavy water reactor fuel channel pressure tubes The average inner diameter at the lower fuel bundle; Equivalent full-power operation time of the fuel channel pressure tube of the heavy water reactor The average inner diameter at the lower fuel bundle; is the radial creep rate of the fuel rod bundle in the pressure tube of the heavy water reactor fuel channel.

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