A prediction method for the radial creep of pressure tubes in a heavy water reactor

Abstract

The present invention specifically relates to a method for predicting the radial creep of a pressure tube in a heavy water reactor, belonging to the technical field of nuclear reactor physics. The method includes: Step 1, calculating the radial creep rate at each fuel bundle of the pressure tube in each fuel channel of the heavy water reactor; Step 2, calculating the predicted value of the average inner diameter at each fuel bundle of the pressure tube in each fuel channel of the heavy water reactor; Step 3, taking the maximum value of the predicted values of the average inner diameters at all fuel bundles of the pressure tube in each fuel channel of the heavy water reactor as the predicted value of the inner diameter of the pressure tube in each fuel channel of the heavy water reactor. The device implements the steps of this method. The present invention obtains the relationship between the radial creep rate and the fast neutron flux at the fuel bundle of the pressure tube in the fuel channel of the heavy water reactor through linear fitting of the radial expansion measurement data of the pressure tube in the fuel channel of the heavy water reactor, and finally realizes the accurate prediction of the radial creep of the pressure tube after the long-term operation of the heavy water reactor.

Description

Technical Field

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

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

[0003] The purpose of the present invention is to provide a method for predicting the radial creep of a pressure tube in a heavy water reactor, 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 the thermal-hydraulic analysis and calculation after the aging of the unit, further ensuring the safe and economic operation of the heavy water reactor unit.

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

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

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

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

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

[0009] As one possible implementation, Step 1, calculating the radial creep rate at each fuel rod bundle of the pressure tube in each fuel channel of the heavy water reactor includes the following steps:

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

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

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

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

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

[0015] As one of the achievable ways, in Step 13, 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, to obtain the relational expression between the radial creep rate and the fast neutron flux at the fuel rod bundle in the fuel channel of the heavy water reactor, includes the following steps:

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

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

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

[0019] Taking the radial creep rate at the fuel rod bundle in the fuel channel of the heavy water reactor as the independent variable and the fast neutron flux at the fuel rod bundle in the fuel channel of the heavy water reactor as the dependent variable, perform a linear fit on the radial creep rate and the fast neutron flux measurement values at each fuel rod bundle under the measured equivalent full power operation time of the measured fuel channel pressure tube of the heavy water reactor, to obtain the relational expression between the radial creep rate and the fast neutron flux at the fuel rod bundle in the fuel channel of the heavy water reactor.

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

[0021] D 2 = D 1 +(t 2 - t 1 ) × r

[0022] wherein, t 1 is the equivalent full power operation time of the pressure tube in the fuel channel of the heavy water reactor, equal to 0; t 2 is the predicted equivalent full power operation time of the pressure tube in the fuel channel of the heavy water reactor; D 2 is the average inner diameter at the fuel rod bundle under the predicted equivalent full power operation time t 2 of the pressure tube in the fuel channel of the heavy water reactor; D 1 is the average inner diameter at the fuel rod bundle under the equivalent full power operation time t 1 of the pressure tube in the fuel channel of the heavy water reactor; r is the radial creep rate at the fuel rod bundle of the pressure tube in the fuel channel of the heavy water reactor.

[0023] Beneficial technical effects of the present invention:

[0024] The prediction method for the radial creep of the pressure tube of the heavy water reactor in the present invention obtains the relationship between the radial creep rate and the fast neutron flux at the fuel rod bundle of the pressure tube in the fuel channel of the heavy water reactor through linear fitting of the radial expansion measurement data of the pressure tube in the fuel channel of the heavy water reactor, and finally realizes the accurate prediction of the radial creep of the pressure tube after long-term operation of the heavy water reactor. The prediction result provides a basis for the thermal-hydraulic analysis calculation after the aging of the unit and the setting value of the overpower protection reactor trip for the calculation area, and further ensures the safe and economic operation of the heavy water reactor unit. Description of the drawings

[0025] Figure 1 is a schematic diagram of a fuel channel of the heavy water reactor;

[0026] Figure 2 is a schematic diagram of the irradiation creep process of the pressure tube of the heavy water reactor.

[0027] In the figure, 1, inlet header; 2, outlet header; 3, pressure tube; 4, fuel channel; 5, fuel rod bundle; 6, reactor core. Specific embodiments

[0028] Unless otherwise defined, all technical and scientific terms used herein have the same meaning as commonly understood by one of ordinary skill in the technical field to which this application belongs; the terms used in the specification of this 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 drawings are intended to cover non-exclusive inclusion. The terms "first", "second", etc. in the specification and claims of this application or the above drawings are used to distinguish different objects and not to describe a specific order.

[0029] Reference to "embodiments" herein means that a particular feature, structure, or characteristic described in connection with the embodiments can be included in at least one embodiment of this application. The phrase appears in various places in the specification and does not necessarily refer to the same embodiment, nor is it an independent or alternative embodiment mutually exclusive with other embodiments. Those skilled in the art will explicitly and implicitly understand that the embodiments described herein can be combined with other embodiments.

[0030] The technical solutions of the present invention will be clearly and completely described below with reference to the accompanying drawings and specific embodiments.

[0031] As Figure 1 shown, the core 6 of the CANDU-6 heavy water reactor consists of a plurality of horizontal fuel channels 4. The fuel channel 4 consists of a pressure tube 3 and two end components; the pressure tube 3 contains coolant and 12 fuel rod bundles 5; the end components are the extended parts of the pressure tube 3 outside the core 6 and 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 bundles 5. As the unit operates, the pressure tube 3 in the core 6 will undergo radial creep, thus affecting the cooling of the fuel rod bundles 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 the pressure tube

[0033] As Figure 2 shown, 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 loops. The material of the pressure tube is zirconium alloy, and zirconium alloy is a crystalline material. In a high-temperature irradiation environment, vacancies in the crystal aggregate and collapse to form annular edge dislocation lines, that is, dislocation loops; 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 ways of dislocation movement; the third stage is the unstable stage.

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

[0035] The influence of the radial creep of the pressure tube on the cooling of the fuel rod bundle by the coolant occurs in the middle and late stages, that is, in the time 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 a linear law.

[0036] 2. Coolant Temperature and Coolant Pressure

[0037] There are 380 fuel channels in the core of the heavy water reactor, and each fuel channel has 12 fuel rod bundles. During the full-power operation of the heavy water reactor, the coolant temperatures at the inlets of the fuel channels in the core of the heavy water reactor are basically the same, and the temperature deviation at the outlets of the fuel channels in the core of the heavy water reactor is within 10°C.

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

[0039] Table 1 Coolant Temperature (°C) at the Same Fuel Rod Bundle in Different Fuel Channels of the Heavy Water Reactor

[0040]

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

[0042] Table 2 Coolant Pressure (MPa) at the Same Fuel Rod Bundle in Different Fuel Channels of the Heavy Water Reactor

[0043]

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

[0045]

[0046] where SD is the standard deviation of parameters at the same fuel rod bundle in different fuel channels; u is the average value of parameters at the same fuel rod bundle in different fuel channels; n is the number of fuel channels; x i is the parameter at the same fuel rod bundle in different fuel channels;

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

[0048]

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

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

[0051] 3. Fast neutron irradiation

[0052] Fast neutron irradiation is the main factor causing the radial creep of the pressure tube. Different fast neutron integral fluxes result in different magnitudes of radial creep. For the core of a long-term operating heavy water reactor, since the core is always refueled without shutting down and there is a liquid region control system constantly participating in controlling and eliminating power deviations to ensure that the power of each region of the core is within the designed power requirements, in the long run, the core of the heavy water reactor conforms to the time-average model. The fast neutron flux at each fuel rod bundle in all fuel channels can be calculated by a physical program, and thus the radial creep of each fuel rod bundle in all fuel channels can be calculated.

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

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

[0055] In actual engineering, using 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 among the circumferential measurement averages when measuring the inner diameter of a single pressure tube along the axial direction. Specifically, the inner diameter of the pressure tube at a fuel rod bundle position is measured circumferentially multiple times, and the average value of the multiple circumferential measurement values is used as the average inner diameter of the pressure tube at this fuel rod bundle position; repeat the above steps until the average inner diameters of the pressure tube at 12 fuel rod bundles are obtained, and then take the maximum value among them as the inner diameter measurement value of the pressure tube.

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

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

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

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

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

[0061] Step 11: Obtain the fast neutron flux measurement values at each fuel rod bundle of the pressure tube in each fuel channel of the heavy water reactor;

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

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

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

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

[0066] The pre-service measurement data of the radial expansion of the fuel channel pressure tube of the heavy water reactor includes the average inner diameter measurement values at each fuel rod bundle when 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 possible implementation, in Step 13, 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, obtaining the relational expression between the radial creep rate and the fast neutron flux at the fuel rod bundle in the fuel channel of the heavy water reactor includes the following steps:

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

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

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

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

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

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

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

[0075] The measured average inner diameter value at the j-th fuel rod bundle under the measured equivalent full power operation time of the measured pressure tube in the heavy water reactor fuel channel is denoted as B j in mm; the average inner diameter at the j-th fuel rod bundle when the equivalent full power operation time of the measured pressure tube in the heavy water reactor fuel channel is 0 is denoted as B 0 in mm; the fast neutron flux at the j-th fuel rod bundle under the measured equivalent full power operation time of the measured pressure tube in the heavy water reactor fuel channel is denoted as BFF j in cm -2 ·s -1 ; the radial creep amount at the j-th fuel rod bundle under the measured equivalent full power operation time of the measured pressure tube in the heavy water reactor fuel channel is denoted as CPR j in mm, and CPR j =B j -B 0 ;

[0076] The measured equivalent full power operation time of the measured pressure tube in the heavy water reactor fuel channel is denoted as EFPH in hours; the radial creep rate at the j-th fuel rod bundle under the measured equivalent full power operation time of the measured pressure tube in the heavy water reactor fuel channel is denoted as CPRj Indicates, unit is mm / h, CPR j =(CP j / B j ) / 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 pressure tube in the fuel channel of the heavy water reactor as the dependent variable and the fast neutron flux at the fuel rod bundle of the pressure tube in the fuel channel of the heavy water reactor as the independent variable, perform a linear fit on the measured radial creep rate and fast neutron flux measurement values at each fuel rod bundle under the measured equivalent full power operation time of the pressure tube in the fuel channel of the heavy water reactor, and obtain the relationship between the radial creep rate and the fast neutron flux at the fuel rod bundle of the pressure tube in the fuel channel of the heavy water reactor;

[0078] Substitute the measured fast neutron flux values at each fuel rod bundle of each pressure tube in the fuel channel 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, and obtain the radial creep rate at each fuel rod bundle of each pressure tube in the fuel channel of the heavy water reactor.

[0079] Table 3 Measured average inner diameter measurement values (mm) at each fuel rod bundle under the measured equivalent full power operation time of the pressure tube in the fuel channel of the heavy water reactor

[0080]

[0081]

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

[0083] Fuel channel EPFH BFF1 BFF2 BFF3 BFF4 BFF5 BFF6 B15 22320 3.281E+12 8.785E+12 1.462E+13 1.922E+13 2.22E+13 2.366E+13 B15 57400 3.281E+12 8.785E+12 1.462E+13 1.922E+13 2.22E+13 2.366E+13 C09 22320 5.017E+12 1.289E+13 1.962E+13 2.496E+13 2.872E+13 3.065E+13 E20 22320 2.327E+12 6.373E+12 1.157E+13 1.574E+13 1.859E+13 1.984E+13 F03 57400 3.269E+12 8.68E+12 1.421E+13 1.882E+13 2.219E+13 2.362E+13 F03 97210 3.269E+12 8.68E+12 1.421E+13 1.882E+13 2.219E+13 2.362E+13 G18 97210 5.977E+12 1.489E+13 2.078E+13 2.477E+13 3.133E+13 3.384E+13 H09 22320 7.799E+12 1.904E+13 2.669E+13 3.056E+13 3.101E+13 3.277E+13 H09 57400 7.799E+12 1.904E+13 2.669E+13 3.056E+13 3.101E+13 3.277E+13 H09 97210 7.799E+12 1.904E+13 2.669E+13 3.056E+13 3.101E+13 3.277E+13 H12 22320 7.919E+12 1.915E+13 2.625E+13 2.972E+13 3.126E+13 3.294E+13 H12 57400 7.919E+12 1.915E+13 2.625E+13 2.972E+13 3.126E+13 3.294E+13 H12 97210 7.919E+12 1.915E+13 2.625E+13 2.972E+13 3.126E+13 3.294E+13 K04 22320 6.301E+12 1.582E+13 2.301E+13 2.833E+13 3.291E+13 3.507E+13 K04 57400 6.301E+12 1.582E+13 2.301E+13 2.833E+13 3.291E+13 3.507E+13 K17 22320 7.444E+12 1.821E+13 2.485E+13 2.801E+13 3.237E+13 3.491E+13 K17 57400 7.444E+12 1.821E+13 2.485E+13 2.801E+13 3.237E+13 3.491E+13 N04 22320 6.299E+12 1.594E+13 2.367E+13 2.95E+13 3.35E+13 3.535E+13 N04 57400 6.299E+12 1.594E+13 2.367E+13 2.95E+13 3.35E+13 3.535E+13 N04 97210 6.299E+12 1.594E+13 2.367E+13 2.95E+13 3.35E+13 3.535E+13 N11 22320 8.158E+12 1.951E+13 2.579E+13 2.775E+13 3.053E+13 3.248E+13 N12 97210 8.162E+12 1.952E+13 2.58E+13 2.777E+13 3.055E+13 3.251E+13 N18 22320 7.012E+12 1.75E+13 2.533E+13 3.05E+13 3.391E+13 3.601E+13 N21 97210 4.112E+12 1.086E+13 1.735E+13 2.258E+13 2.601E+13 2.751E+13 O08 97210 7.835E+12 1.919E+13 2.707E+13 3.126E+13 3.209E+13 3.383E+13 Q07 22320 7.029E+12 1.734E+13 2.439E+13 2.937E+13 3.324E+13 3.556E+13 Q14 22320 7.528E+12 1.85E+13 2.615E+13 3.048E+13 3.151E+13 3.336E+13 Q14 57400 7.528E+12 1.85E+13 2.615E+13 3.048E+13 3.151E+13 3.336E+13 Q14 97210 7.528E+12 1.85E+13 2.615E+13 3.048E+13 3.151E+13 3.336E+13 R06 57400 5.979E+12 1.483E+13 2.052E+13 2.443E+13 3.086E+13 3.349E+13

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

[0085] Fuel channel EPFH BFF7 BFF8 BFF9 BFF10 BFF11 BFF12 B15 22320 2.366E+13 2.219E+13 1.911E+13 1.441E+13 8.629E+12 3.301E+12 B15 57400 2.366E+13 2.219E+13 1.911E+13 1.441E+13 8.629E+12 3.301E+12 C09 22320 3.064E+13 2.87E+13 2.483E+13 1.934E+13 1.267E+13 5.069E+12 E20 22320 1.987E+13 1.861E+13 1.566E+13 1.143E+13 6.267E+12 2.334E+12 F03 57400 2.372E+13 2.224E+13 1.876E+13 1.404E+13 8.541E+12 3.291E+12 F03 97210 2.372E+13 2.224E+13 1.876E+13 1.404E+13 8.541E+12 3.291E+12 G18 97210 3.403E+13 3.139E+13 2.47E+13 2.05E+13 1.465E+13 6.079E+12 H09 22320 3.28E+13 3.103E+13 3.028E+13 2.595E+13 1.836E+13 7.812E+12 H09 57400 3.28E+13 3.103E+13 3.028E+13 2.595E+13 1.836E+13 7.812E+12 H09 97210 3.28E+13 3.103E+13 3.028E+13 2.595E+13 1.836E+13 7.812E+12 H12 22320 3.288E+13 3.12E+13 2.943E+13 2.551E+13 1.848E+13 7.949E+12 H12 57400 3.288E+13 3.12E+13 2.943E+13 2.551E+13 1.848E+13 7.949E+12 H12 97210 3.288E+13 3.12E+13 2.943E+13 2.551E+13 1.848E+13 7.949E+12 K04 22320 3.505E+13 3.289E+13 2.82E+13 2.268E+13 1.555E+13 6.385E+12 K04 57400 3.505E+13 3.289E+13 2.82E+13 2.268E+13 1.555E+13 6.385E+12 K17 22320 3.493E+13 3.238E+13 2.78E+13 2.423E+13 1.765E+13 7.505E+12 K17 57400 3.493E+13 3.238E+13 2.78E+13 2.423E+13 1.765E+13 7.505E+12 N04 22320 3.531E+13 3.347E+13 2.935E+13 2.332E+13 1.566E+13 6.372E+12 N04 57400 3.531E+13 3.347E+13 2.935E+13 2.332E+13 1.566E+13 6.372E+12 N04 97210 3.531E+13 3.347E+13 2.935E+13 2.332E+13 1.566E+13 6.372E+12 N11 22320 3.253E+13 3.056E+13 2.755E+13 2.509E+13 1.887E+13 8.242E+12 N12 97210 3.25E+13 3.054E+13 2.754E+13 2.508E+13 1.886E+13 8.238E+12 N18 22320 3.598E+13 3.388E+13 3.033E+13 2.489E+13 1.713E+13 7.089E+12 N21 97210 2.754E+13 2.602E+13 2.249E+13 1.715E+13 1.07E+13 4.151E+12 O08 97210 3.386E+13 3.212E+13 3.102E+13 2.641E+13 1.859E+13 7.878E+12 Q07 22320 3.539E+13 3.315E+13 2.909E+13 2.38E+13 1.682E+13 7.044E+12 Q14 22320 3.35E+13 3.157E+13 3.024E+13 2.549E+13 1.791E+13 7.555E+12 Q14 57400 3.35E+13 3.157E+13 3.024E+13 2.549E+13 1.791E+13 7.555E+12 Q14 97210 3.35E+13 3.157E+13 3.024E+13 2.549E+13 1.791E+13 7.555E+12 R06 57400 3.329E+13 3.077E+13 2.429E+13 2.019E+13 1.455E+13 6.062E+12

[0086] Table 6 Measured radial creep variables (mm) at each fuel rod bundle under the measured equivalent full power operation time of the pressure tube in the fuel channel of the heavy water reactor

[0087]

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

[0089]

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

[0091] D 2 = D 1 +(t 2 - t 1 )×r

[0092] where t 1 is the equivalent full power operation time of the pressure tube in the fuel channel of the heavy water reactor, equal to 0; t 2 is the predicted equivalent full power operation time of the pressure tube in the fuel channel of the heavy water reactor; D 2 is the average inner diameter at the fuel rod bundle under the predicted equivalent full power operation time t 2 of the pressure tube in the fuel channel of the heavy water reactor; D 1 is the average inner diameter at the fuel rod bundle under the equivalent full power operation time t 1 of the pressure tube in the fuel channel of the heavy water reactor; r is the radial creep rate at the fuel rod bundle of the pressure tube in the fuel channel of the heavy water reactor.

[0093] The following is an example to illustrate how to calculate the pressure tube in the fuel channel of the heavy water reactor.

[0094] Table 8 shows the measured fast neutron flux values at the fuel rod bundles of the pressure tube in the fuel channel of the heavy water reactor. Taking BFF j to represent the measured fast neutron flux value at the j-th fuel rod bundle of the pressure tube in the fuel channel of the heavy water reactor, with the unit of cm -2 ·s -1 , j = 1, 2,..., m, for the CANDU-6 type heavy water reactor, m = 12;

[0095] Table 9 shows the radial creep rate at the fuel rod bundles of the pressure tube in the fuel channel of the heavy water reactor. Taking CPR j to represent the radial creep rate at the j-th fuel rod bundle of the pressure tube in the fuel channel of the heavy water reactor, with the unit of mm / h; j = 1, 2,..., m; for the CANDU-6 type heavy water reactor, m = 12; the radial creep rate at the fuel rod bundles of the pressure tube in the fuel channel of the heavy water reactor is obtained by substituting the measured fast neutron flux value at the fuel rod bundles of the pressure tube in the fuel channel of the heavy water reactor into the relational expression between the radial creep rate and the fast neutron flux at the fuel rod bundles of the pressure tube in the fuel channel of the heavy water reactor;

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

[0097] D 2 = D 1 +(t 2 - t 1 )×r

[0098] where, t 1 is the equivalent full power operation time of the pressure tube in the heavy water reactor fuel channel, equal to 0; t 2 is the predicted equivalent full power operation time of the pressure tube in the heavy water reactor fuel channel, with the unit of h; D 2 is the average inner diameter at the fuel rod bundle of the pressure tube in the heavy water reactor fuel channel under the predicted equivalent full power operation time t 2 , with the unit of mm; D 1 is the average inner diameter at the fuel rod bundle of the pressure tube in the heavy water reactor fuel channel under the equivalent full power operation time t 1 , with the unit of mm; r is the radial creep rate at the fuel rod bundle of the pressure tube in the heavy water reactor fuel channel, with the unit of mm / h;

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

[0100] By measuring the measured average inner diameter at each fuel rod bundle of the pressure tube in the heavy water reactor fuel channel under the predicted equivalent full power operation time, and taking the maximum value of the measured average inner diameter at all fuel rod bundles of the pressure tube in the heavy water reactor fuel channel under the predicted equivalent full power operation time, as the measured inner diameter of the pressure tube in the heavy water reactor fuel channel under the predicted equivalent full power operation time.

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

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

[0103]

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

[0105]

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

[0107]

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

[0109]

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

[0111] The device for predicting the radial creep of the heavy water reactor pressure tube described in this embodiment includes:

[0112] A first calculation module for calculating the radial creep rate at each fuel rod bundle of each pressure tube in each fuel channel of the heavy water reactor;

[0113] A second calculation module for calculating the predicted average inner diameter value at each fuel rod bundle of each pressure tube in each fuel channel of the heavy water reactor;

[0114] A prediction module for using the maximum value of the predicted average inner diameter values at all fuel rod bundles of each pressure tube in each fuel channel of the heavy water reactor as the predicted inner diameter value of each pressure tube in each fuel channel of the heavy water reactor.

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

[0116] The above embodiments only represent several implementation manners of the present invention. Their descriptions are relatively specific and detailed, but they should not be construed as limiting the scope of the patent of the present invention. It should be noted that for those of ordinary skill in the art, without departing from the concept of the present invention, several deformations and improvements can still be made, and these all belong to the protection scope of the present invention. Therefore, the protection scope of the patent of the present invention should be subject to the appended 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 by 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 by 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 by the heavy water reactor) / the equivalent full-power operation time of the fuel channel pressure tube measured by 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: D2=D1+(t2-t1)×r Among them, t1 is the equivalent full-power operation time of the heavy water reactor fuel channel pressure tube, which is equal to 0; t2 is the predicted equivalent full-power operation time of the heavy water reactor fuel channel pressure tube; D2 is the average inner diameter of the fuel rod bundle at the predicted equivalent full-power operation time t2 of the heavy water reactor fuel channel pressure tube; D1 is the average inner diameter of the fuel rod bundle at the equivalent full-power operation time t1 of the heavy water reactor fuel channel pressure tube; r is the radial creep rate at the fuel rod bundle of the heavy water reactor fuel channel pressure tube.

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