Fuel power calibration method for loop irradiation test

Abstract

The invention relates to the technical field of research of reactor fuel irradiation, in particular to a fuel power calibration method for a loop irradiation test, which comprises the following steps: carrying out an out-of-reactor hydraulic test, and establishing a high-precision CFD (computational fluid dynamics) model of a fuel loop irradiation device based on a hydraulic test result; calculating the gamma heat release rate ratio of each material and the total heat release in the test section, and preliminarily determining the nuclear power value of the fuel assembly; based on the heat conduction differential equation and the convective heat transfer equation, the gamma heat release rate of the material and the nuclear power value of the fuel assembly are corrected; repeating the calculation process of S3 until the nuclear power value of the fuel assembly converges, and obtaining the final theoretical calculation value of the gamma heat release rate of each material and the nuclear power of the fuel assembly; and further verifying the calculation result to finally obtain the nuclear power of the fuel core body, and obtaining the burnup of the fuel assembly in the irradiation test through time integration. The method aims to solve the technical problem that the nuclear power of the fuel assembly calculated based on thermotechnical measurement data in a loop irradiation test has deviation compared with an actual value.

Description

Technical Field

[0001] The invention relates to the technical field of research reactor fuel irradiation, and in particular to a fuel power calibration method for loop irradiation test. Background Art

[0002] A very important step in the design verification phase of fuel elements is to irradiate the fuel elements within the expected fuel life cycle and obtain the necessary test data in a timely manner during or after the irradiation test. Researchers can evaluate the comprehensive performance of the fuel elements in the research reactor and further verify the rationality of their structural design, the reliability of the manufacturing process and the feasibility of their application.

[0003] In high-throughput research reactors, loop irradiation tests or in-process instrumented irradiation tests are usually carried out for fuel element irradiation. The corresponding device that contains the fuel elements is called a loop irradiation device or an in-process instrumented irradiation device. The loop irradiation test device usually carries out irradiation tests on small-sized fuel assemblies, which can simulate the thermal hydraulics and water chemistry environment of the prototype reactor. The interior of the loop irradiation device is isolated from the reactor coolant, and the heat released by the fuel assemblies and structural materials in the device is taken away by the coolant in the irradiation test loop and exported to the secondary water loop through the heat exchanger.

[0004] Nuclear power and burnup are key indicators of fuel assembly irradiation test. Accurately determining the nuclear power and burnup of fuel assemblies during irradiation test can not only ensure the safety of fuel assemblies, but also determine whether the target burnup of irradiation test is achieved. During the test, two methods are usually used to track and calculate the fuel assembly burnup: neutron calculation and calculation based on thermal measurement data.

[0005] The burnup calculation method based on thermal measurement data first obtains the heat release power of the fuel assembly by the heat balance method through the thermal measurement data of the cooling water in the test loop, and then integrates the heat release power over time to obtain the fuel assembly burnup. However, the heat release power calculated based on the inlet and outlet temperatures and flow rates of the cooling water includes the nuclear fission power of the fuel assembly in the device and the gamma heat release power of the nuclear fuel and structural materials. Therefore, the nuclear power of the fuel assembly calculated based on the thermal measurement data deviates from the actual value, and the corresponding burnup value will also have errors. Summary of the invention

[0006] In view of this, the present invention provides a fuel power calibration method for loop irradiation test, which is used to solve the technical problem that the nuclear power of the fuel assembly calculated based on thermal measurement data in the existing loop irradiation test technology deviates from the actual value.

[0007] The embodiments of the present invention are implemented by the following technical solutions:

[0008] A fuel power calibration method for a loop irradiation test comprises the following steps:

[0009] S1. Conduct an out-of-core hydraulic test, establish a high-precision CFD model of the fuel loop irradiation device based on the hydraulic test results, analyze the CFD calculation results, and determine the coolant flow distribution in the square box of the loop irradiation device;

[0010] S2. Calculate the gamma heat release rate ratio of each material and the total heat release in the test section of the fuel loop irradiation device. Based on historical operating experience, preliminarily set the gamma heat release rate of the material to obtain the heat release of the pressure tube P1 and the heat release of the shunt tube P2, and then preliminarily determine the nuclear power value of the fuel assembly;

[0011] S3. Based on the heat conduction differential equation and the convection heat transfer equation, calculate the heat dissipation from the pressure tube to the reactor coolant and the heat transfer process of the coolant inside and outside the square box through the square box, so as to correct the gamma heat release rate of the material and preliminarily determine the nuclear power value of the fuel assembly;

[0012] S4, repeat the calculation process of S3 until the gamma heat release rate of the material and the nuclear power value of the fuel assembly converge, so as to obtain the final theoretical calculated value of the gamma heat release rate of each material and the nuclear power of the fuel assembly;

[0013] S5. The theoretical calculated values ​​are corrected by high-precision analytical calculation methods, and the calculation results are further verified by thermal measurement data, so as to finally obtain the nuclear power of the fuel core, and the burnup of the fuel assembly in the irradiation test is obtained by the time integral of the nuclear power.

[0014] Furthermore, in the fuel circuit irradiation device, thermocouples for measuring the coolant temperature are arranged on the outer wall of the upper end of the square box, the component inlet position and the component outlet position.

[0015] Furthermore, S1 also includes: establishing a CFD model of the test section where the fuel assembly is located and analyzing the CFD calculation results, and comparing the CFD results with the results of the out-of-core hydraulic test, so as to obtain a high-precision CFD model of the fuel loop irradiation device.

[0016] Furthermore, the S2 specifically includes the following steps:

[0017] S21. Obtaining the gamma heat release rate ratio of each material in the fuel loop irradiation device according to the neutronic calculation results;

[0018] S22. Calculate the total heat release in the test section of the fuel circuit irradiation device based on the irradiation test data and the heat balance method;

[0019] S23, preliminarily setting the gamma heat release rate of one of the materials, and obtaining the gamma heat release power of other materials according to the actual mass ratio of each material, and preliminarily obtaining the nuclear power value of the fuel assembly;

[0020] Furthermore, the materials include pressure tubes, square boxes, spacers, fuel cladding and fuel cores; and the irradiation test data include coolant flow and thermocouple measuring point temperature.

[0021] Furthermore, in S3, the heat absorbed by the coolant flowing through the fuel assembly area is balanced with the heat released in the device and the heat dissipated by the device to the reactor, that is:

[0022] m·(2-1)=P1+P2+P s -Q c -Q r (1)

[0023] Where m represents the coolant mass flow rate; h1 is the specific enthalpy of the coolant on the outer wall of the upper end of the box, h2 is the specific enthalpy of the coolant at the outlet of the fuel assembly, P1 is the heat released by the pressure tube, P2 is the heat released by the diverter tube, P s Heat release for fuel assembly, Q c is the heat dissipation caused by conduction, Q r is the amount of heat dissipated by radiation;

[0024] Energy conservation exists in the region inside and outside the shunt pipe, namely:

[0025] m·(2-0)=P s -P L (2)

[0026] m·(0-1)=P1+P2+P L -Q c -Q r (3)

[0027] Where h0 is the specific enthalpy of the coolant at the fuel assembly inlet, P L The cooling water in the shunt pipe releases heat to the cooling water outside the shunt pipe;

[0028] The inner and outer walls of the shunt pipe are used for cooling water convection heat dissipation. The shunt pipe itself is a radial heat conduction with an internal heat source. Based on the heat conduction differential equation and the convection heat transfer equation, we have:

[0029]

[0030] Among them, T f1 is the qualitative temperature of the coolant between the shunt pipe and the pressure pipe, T f2is the qualitative temperature of the coolant in the shunt pipe; r1 and r2 are the inner and outer wall diameters of the shunt pipe respectively; S1 and S2 are the inner and outer wall areas of the shunt pipe respectively; a1 and a2 are the convection heat transfer coefficients of the inner and outer flow channels of the shunt pipe respectively, which can be obtained by the heat transfer relationship; λ1 is the thermal conductivity of the shunt pipe; is the volume heat release rate of the manifold; l is the height of the area where the fuel assembly is located.

[0031] Furthermore, the qualitative temperature of the coolant between the shunt pipe and the pressure pipe is T f1 Calculated by the following formula:

[0032] T f1 =(T1+T0) / 2 (5);

[0033] Qualitative temperature of coolant in the shunt pipe T f2 Calculated by the following formula:

[0034] T f2 =(T0+T2) / 2 (6)

[0035] Among them, T0 is the inlet temperature of the fuel assembly, T1 is the coolant temperature of the outer wall of the upper end of the square box, and T2 is the outlet temperature of the fuel assembly.

[0036] Furthermore, in S4, the deviation of the nuclear power value of the fuel assembly before and after correction is compared. When the deviation is greater than the convergence range, the gamma heat release rate of the initially set material needs to be corrected, and the iterative cycle is repeated until the deviation of the nuclear power value of the fuel assembly before and after correction is within the convergence range.

[0037] Furthermore, the S5 comprises the following sub-steps:

[0038] S51. Using the gamma heat release rate of the material and the nuclear power value of the fuel assembly as the heat source input of the high-precision CFD model, and using the thermal measurement data as the inlet boundary conditions of the high-precision CFD model, perform CFD analysis and calculation;

[0039] S52, comparing the CFD calculation results with the actual irradiation test temperature measurement values, and based on the comparison results, further correcting the gamma heat release rate of each material and the nuclear power of the fuel assembly according to the pressure tube heat dissipation obtained by the high-precision CFD model;

[0040] S53, repeating step S51 again using the updated heat source to perform CFD analysis and calculation;

[0041] S54, repeat the above steps, iterate until the temperature calculation value and the measured value are consistent, and obtain the final material heat release rate and the nuclear power of the fuel core;

[0042] S55. Regularly update the gamma heat release rate ratio of each material, re-iterate the material gamma heat release rate and the nuclear power of the fuel assembly, and obtain the burnup of the fuel assembly in the irradiation test through the time integration of the nuclear power.

[0043] Furthermore, the thermal measurement data in S51 refers to the temperature measurement value of the outer wall at the upper end of the square box and the coolant flow rate.

[0044] The technical solution of the embodiment of the present invention has at least the following advantages and beneficial effects:

[0045] (1) Based on the heat balance method and supplemented by neutron calculation results, the thermal measurement data is fully utilized to achieve a more accurate real-time calculation of the fuel nuclear power in the loop irradiation test from a thermal perspective. This can effectively eliminate the uncertainty of factors such as material gamma heat release and heat loss on the nuclear power calculation, achieve real-time calculation of the nuclear power of the fuel assembly, and thus support the determination of fuel burnup during the irradiation test;

[0046] (2) Based on theoretical calculations, the heat transfer process in the fuel loop irradiation device is effectively analyzed to identify the amount of heat transferred from the coolant inside the box to the coolant outside the box and the amount of heat dissipated from the device to the research reactor coolant through the pressure tube, thereby reducing the deviation between the nuclear power of the fuel assembly calculated based on thermal measurement data and the actual value;

[0047] (3) Using the theoretically calculated material gamma heat release and fuel nuclear power as the heat source input of the CFD calculation model can further reduce the number of iterations of the CFD simulation process and accelerate the convergence speed;

[0048] (4) Taking the heat release rate of typical materials as the initial value of the iteration, through the dual iteration of theoretical calculation based on the thermal differential equation and high-precision simulation calculation based on CFD, the influence of factors such as material heat release and device heat loss on the uncertainty of nuclear power calculation can be effectively corrected. At the same time, the number of iterations can be greatly reduced, the analysis and calculation time can be saved, and the iteration efficiency of nuclear power calculation can be effectively improved;

[0049] (5) A high-precision CFD model of the fuel loop irradiation device is established through CFD. High-precision simulation calculations are carried out using thermal measurement data consistent with theoretical calculations. The reliability of the CFD calculation results is verified using the temperature test values ​​of thermal measurements, thereby effectively ensuring the accuracy of the calculation results. BRIEF DESCRIPTION OF THE DRAWINGS

[0050] In order to more clearly illustrate the technical solutions of the embodiments of the present invention, the drawings required for use in the embodiments of the present invention are briefly introduced below. It should be understood that the following drawings only show certain embodiments of the present invention and therefore should not be regarded as limiting the scope. For ordinary technicians in this field, other related drawings can be obtained based on these drawings without creative work.

[0051] Figure 1 It is the principle diagram of the present invention;

[0052] Figure 2 It is a schematic diagram of the structure of the loop irradiation device of the present invention.

[0053] Figure 3 is a schematic diagram of a shunt pipe assembly;

[0054] Figure 4 A diagram of the first thermocouple arrangement in the fuel assembly area;

[0055] Figure 5 A diagram of the second thermocouple arrangement in the fuel assembly area;

[0056] Figure 6 This is a schematic diagram of the heat transfer process in the fuel assembly area of ​​the loop irradiation test device;

[0057] In the figure, 1-insulation pipe assembly; 2-pressure pipe assembly; 3-diverter pipe assembly; 4-transition pipe assembly; 5-lifting head assembly; 6-connecting pipe; 7-inner diverter pipe; 8-outer diverter pipe; 9-spring; 10-spring base; 11-square box; 12-square and round joint; 13-pressure pipe; 14-insulation pipe; 1#~8# are thermocouples. DETAILED DESCRIPTION

[0058] In order to make the purpose, technical solutions and advantages of the embodiments of the present invention clearer, the technical solutions in the embodiments of the present invention will be clearly and completely described below in conjunction with the drawings in the embodiments of the present invention. Obviously, the described embodiments are part of the embodiments of the present invention, not all of the embodiments. Generally, the components of the embodiments of the present invention described and shown in the drawings here can be arranged and designed in various different configurations.

[0059] like Figure 2As shown, the loop irradiation device of the present invention comprises an insulation pipe assembly 1, a pressure pipe assembly 2, a shunt pipe assembly 3 and a transition pipe assembly 4. The insulation pipe assembly 1 is located at the outer layer of the pressure pipe assembly 2, the upper flange is connected to the large flat top cover, and the lower joint is inserted into the grid plate. The lower flange end face of the transition pipe assembly 4 is connected to the insulation pipe flange, and the upper flange end face is connected to the middle flange of the pressure pipe assembly 2. Nitrogen is filled between the insulation pipe 14 and the pressure pipe 13 to detect whether the pressure pipe 13 is leaking. The outside of the insulation pipe 14 is the reactor coolant, and there is a nitrogen layer between the insulation pipe 14 and the pressure pipe 13. Therefore, a small amount of heat released in the device will also be lost to the reactor coolant. The shunt pipe assembly 3 is located inside the pressure pipe assembly 2, and the fuel assembly is placed in the square box of the shunt pipe assembly 3. The fuel assembly is pressed by the bracket of the lower head assembly of the pressure pipe assembly 2 and the spring base 10 of the shunt pipe assembly 3.

[0060] like Figure 3 As shown, the manifold assembly 3 is mainly composed of a lifting head assembly 5, a pipe 6, an inner manifold 7, an outer manifold 8, a square-round joint 12, a square box 11, a spring 9, a spring base 10, etc. The upper part of the manifold connects the lifting head assembly 5, the inner manifold 7, and the outer manifold 8 together through the pipe 6, and the lower part connects the inner manifold 7, the outer manifold 8, and the square box 11 together through the square-round joint 12. The fuel assembly is placed in the square box 11 of the manifold assembly 3, and is pressed together by the bracket of the lower head assembly of the pressure tube assembly 2 and the spring base 10 of the manifold assembly 3.

[0061] like Figure 4 and Figure 5 As shown, the fuel loop irradiation device arranges 8 thermocouples in the area near the fuel assembly, the 1# thermocouple is located on the outer wall of the upper end of the square box 11; the 3#, 4# and 6# thermocouples are located at the inlet of the assembly, wherein the 3# and 4# thermocouples are located on the positive side, and the 6# thermocouple is located on the negative side; the 2# and 5# thermocouples are located at the outlet of the assembly, wherein the 2# thermocouple is located on the positive side, and the 5# thermocouple is located on the negative side.

[0062] like Figure 1 and Figure 6 As shown, a fuel power calibration method for loop irradiation test, the specific implementation process includes the following steps:

[0063] S1. According to the structural design of the fuel loop irradiation device, the device processing and assembly are completed and the off-pillar cold hydraulic test is carried out. The CFD model of the test section where the fuel assembly is located is established and the calculation results are analyzed. The CFD results are compared with the off-pillar hydraulic test results to obtain a high-precision CFD model for the flow and heat transfer simulation of the subsequent test section, so as to obtain the flow distribution of the coolant between the square box 11 and the annular slit between the lower pipe seat and the square box 11.

[0064] S2. According to the neutronic calculation results, the gamma heat release rate ratio of each material in the fuel loop irradiation device (including the pressure tube 13, the square box 11, the spacer, the fuel cladding and the fuel core) is obtained, and it is considered that the ratio is constant within a certain irradiation time. According to the irradiation test data such as the coolant flow rate and the temperature of the thermocouple measuring point, the total heat release in the test section of the fuel loop irradiation device can be obtained based on the heat balance method. According to the historical operating experience, the gamma heat release rate of a certain structural material is initially set, and the gamma heat release rates of other materials are obtained in proportion. According to the actual mass of each material, its gamma heat release power can be obtained, and then the nuclear power value of the fuel assembly can be preliminarily obtained.

[0065] S3. Based on the coolant flow distribution scheme in the square box, through the heat conduction differential equation and the convection heat transfer equation, theoretical calculations are performed on the heat dissipation process of the pressure tube 13 to the reactor coolant through the nitrogen layer, the flow heat transfer process of the coolant flowing through the flow channel between the pressure tube 13 and the square box 11, the flow heat transfer process of the coolant flowing through the fuel assembly in the square box 11, and the heat transfer process of the coolant inside and outside the square box through the square box 11, and the heat transfer process of the coolant inside and outside the square box through the square box 11, so as to obtain the heat transfer between the unventilated areas, including the heat dissipation of the pressure tube 13 to the reactor coolant and the heat transfer process of the coolant inside and outside the square box through the square box 11, so as to correct the gamma heat release rate of the material and the preliminary fuel assembly nuclear power determined by the heat balance method. Specifically, it can be implemented according to the following process:

[0066] The heat absorbed by the coolant flowing through the fuel assembly area is balanced with the heat release in the device and the heat dissipation from the device to the reactor, that is:

[0067] m·(h2-h1)=P1+P2+P s -Q c -Q r (1)

[0068] Among them, m is the coolant mass flow rate, which can be measured in the loop; h1 and h2 are the specific enthalpies of the coolant at the inlet and outlet of the active area, respectively, and their temperatures can be measured by thermocouples.

[0069] Energy conservation also exists in the areas inside and outside the shunt pipe, that is:

[0070] m·(2-0)=P s -P L (2)

[0071] m·(0-1)=P1+P2+P L -Q c -Q r (3)

[0072] Equations (1), (2) and (3) can be derived from two of them to form the third equation.

[0073] The qualitative temperature T of the coolant between the shunt pipe and the pressure pipe 13f1 Calculated by the following formula:

[0074] T f1 =(T1+T0) / 2 (5)

[0075] Qualitative temperature of coolant in the shunt pipe T f2 Calculated by the following formula:

[0076] T f2 =(T0+T2) / 2 (6)

[0077] Among them, T0 is the inlet temperature of the fuel assembly, T1 is the coolant temperature of the outer wall of the upper end of the square box, and T2 is the outlet temperature of the fuel assembly.

[0078] The inner and outer walls of the shunt pipe are for cooling water convection heat transfer, and the shunt pipe itself is for radial heat conduction with an internal heat source. Based on the heat conduction differential equation and the convection heat transfer equation, we have:

[0079]

[0080] Among them, r1 and r2 are the inner and outer wall diameters of the shunt pipe respectively; S1 and S2 are the inner and outer wall areas of the shunt pipe respectively; a1 and a2 are the convective heat transfer coefficients of the inner and outer flow channels of the shunt pipe respectively, which can be obtained by the heat transfer relationship; λ1 is the thermal conductivity of the shunt pipe; It is the volume heat release rate of the shunt tube, which is closely related to the heat release P2 of the shunt tube.

[0081] By initially setting the material heat release rate, we can get the heat release of the pressure tube P1 and the heat release of the shunt tube P2, thus calculating the heat dissipation Q c And the radiation heat dissipation Q r Solve it and then get P through formula (1) s , thus solving the nuclear heat release of the fuel assembly; through (2) and P s , can be used for P L Solve; through (4), P L and P s The heat release rate of the shunt tube volume can be The solution is performed to obtain the heat release of the shunt pipe P2, and the heat release of the pressure pipe P1 is corrected at the same time.

[0082] S4, repeat the calculation process of S3 until the gamma heat release rate of the material and the nuclear power value of the fuel assembly converge, so as to obtain the final theoretical calculated value of the gamma heat release rate of each material and the nuclear power of the fuel assembly;

[0083] Specifically, the deviation of the nuclear power value of the fuel assembly before and after correction is compared. When the deviation is greater than the convergence range, the gamma heat release rate of the initially set material needs to be corrected, and the iteration cycle is repeated until the deviation of the nuclear power value of the fuel assembly before and after correction is within the convergence range. Here, the convergence range is 10-4 .

[0084] S5. Use high-precision analytical calculation methods to correct the theoretical calculation values, and further verify the calculation results through thermal measurement data, so as to finally obtain the nuclear power of the fuel core, and obtain the burnup of the fuel assembly in the irradiation test through the time integration of the nuclear power;

[0085] Specifically, in this embodiment, a high-precision analysis and calculation model of the test section of the fuel loop irradiation device is established based on CFD, the heat release rate of the structural material and the nuclear power of the fuel assembly are used as the heat source input of the CFD calculation model, and the temperature measurement value of the 1# thermocouple and the coolant flow rate are used as the inlet boundary conditions of the CFD calculation model. The CFD calculation results are compared with the actual irradiation test temperature measurement values ​​such as the component inlet temperature and the component outlet temperature. Based on the difference between the two and the heat dissipation of the pressure tube obtained by CFD, the gamma heat release rate of each structural material and the nuclear power of the fuel assembly are further corrected, and the updated heat source is used to perform CFD calculation again. Iterate in this way until the temperature calculation value is in good agreement with the measured value, thereby obtaining the final heat release rate of the structural material and the nuclear power of the fuel assembly.

[0086] The integral of the nuclear power of the fuel assembly over time is the burnup. For long-period loop irradiation tests, it is necessary to update the gamma heat release rate ratio of each material at regular intervals, and re-iterate the material heat release rate and the nuclear power of the fuel assembly according to the above-mentioned power calibration process, and accumulate the corresponding fuel burnup after segmented integration.

[0087] Specifically, in this embodiment, the neutron calculation results are updated when the core loading changes.

[0088] The basic concepts have been described above. Obviously, for those skilled in the art, the above detailed disclosure is only for example and does not constitute a limitation of this specification. Although not explicitly stated here, those skilled in the art may make various modifications, improvements and corrections to this specification. Such modifications, improvements and corrections are suggested in this specification, so such modifications, improvements and corrections still belong to the spirit and scope of the exemplary embodiments of this specification.

[0089] Similarly, it should be noted that in order to simplify the description disclosed in this specification and thus help understand one or more embodiments of the invention, in the above description of the embodiments of this specification, multiple features are sometimes combined into one embodiment, figure or description thereof. However, this disclosure method does not mean that the features required by the subject matter of this specification are more than the features mentioned in the claims. In fact, the features of the embodiments are less than all the features of the single embodiment disclosed above.

[0090] Finally, it should be understood that the embodiments described in this specification are only used to illustrate the principles of the embodiments of this specification. Other variations may also fall within the scope of this specification. Therefore, as an example and not a limitation, alternative configurations of the embodiments of this specification may be considered consistent with the teachings of this specification. Accordingly, the embodiments of this specification are not limited to the embodiments explicitly introduced and described in this specification.

Claims

1. A fuel power calibration method for loop irradiation test, characterized in that: The following steps are involved: S1. Conduct an out-of-core hydraulic test, establish a high-precision CFD model of the fuel loop irradiation device based on the hydraulic test results, analyze the CFD calculation results, and determine the coolant flow distribution in the square box of the loop irradiation device; S2. Calculate the gamma heat release rate ratio of each material and the total heat release in the test section of the fuel loop irradiation device. Based on historical operating experience, preliminarily set the gamma heat release rate of the material to obtain the heat release of the pressure tube P1 and the heat release of the shunt tube P2, and then preliminarily determine the nuclear power value of the fuel assembly; S3. Based on the heat conduction differential equation and the convection heat transfer equation, calculate the heat dissipation from the pressure tube to the reactor coolant and the heat transfer process of the coolant inside and outside the square box through the square box, so as to correct the gamma heat release rate of the material and preliminarily determine the nuclear power value of the fuel assembly; S4, repeat the calculation process of S3 until the gamma heat release rate of the material and the nuclear power value of the fuel assembly converge, so as to obtain the final theoretical calculated value of the gamma heat release rate of each material and the nuclear power of the fuel assembly; S5. The theoretical calculated values ​​are corrected by high-precision analytical calculation methods, and the calculation results are further verified by thermal measurement data, so as to finally obtain the nuclear power of the fuel core, and the burnup of the fuel assembly in the irradiation test is obtained by the time integral of the nuclear power.

2. A fuel power calibration method for loop irradiation test according to claim 1, characterized in that: In the fuel circuit irradiation device, thermocouples for measuring the coolant temperature are arranged on the outer wall of the upper end of the square box, the component inlet position and the component outlet position.

3. A fuel power calibration method for loop irradiation test according to claim 1, characterized in that: The S1 also includes: establishing a CFD model of the test section where the fuel assembly is located and analyzing the CFD calculation results, and comparing the CFD results with the results of the out-of-core hydraulic test, so as to obtain a high-precision CFD model of the fuel loop irradiation device.

4. A fuel power calibration method for loop irradiation test according to claim 1, characterized in that: The S2 specifically includes the following steps: S21. Obtaining the gamma heat release rate ratio of each material in the fuel loop irradiation device according to the neutronic calculation results; S22. Calculate the total heat release in the test section of the fuel circuit irradiation device based on the irradiation test data and the heat balance method; S23. Preliminarily set the gamma heat release rate of one of the materials, and obtain the gamma heat release power of other materials based on the actual mass ratio of each material, and preliminarily obtain the nuclear power value of the fuel assembly.

5. A fuel power calibration method for loop irradiation test as claimed in claim 4, characterized in that: The materials include pressure tubes, square boxes, spacers, fuel cladding and fuel cores; the irradiation test data include coolant flow and thermocouple measuring point temperature.

6. A fuel power calibration method for loop irradiation test according to claim 1, characterized in that: In S3, the heat absorbed by the coolant flowing through the fuel assembly area is balanced with the heat released in the device and the heat dissipated by the device to the reactor, that is: m·(2-1)=P1+P2+P s -Q c -Q r (1) Where m represents the coolant mass flow rate; h1 is the specific enthalpy of the coolant on the outer wall of the upper end of the box, h2 is the specific enthalpy of the coolant at the outlet of the fuel assembly, P1 is the heat released by the pressure tube, P2 is the heat released by the diverter tube, P s Heat release for fuel assembly, Q c is the heat dissipation by conduction, Q r is the amount of heat dissipated by radiation; Energy conservation exists in the region inside and outside the shunt pipe, namely: m·(2-0)=P s -P L (2) m·(0-1)=P1+P2+P L -Q c -Q r (3) Where h0 is the specific enthalpy of the coolant at the fuel assembly inlet, P L The cooling water in the shunt pipe releases heat to the cooling water outside the shunt pipe; The inner and outer walls of the shunt pipe are used for cooling water convection heat dissipation. The shunt pipe itself is a radial heat conduction with an internal heat source. Based on the heat conduction differential equation and the convection heat transfer equation, we have: Among them, T f1 is the qualitative temperature of the coolant between the shunt pipe and the pressure pipe, T f2 is the qualitative temperature of the coolant in the shunt pipe; r1 and r2 are the inner and outer wall diameters of the shunt pipe respectively; S1 and S2 are the inner and outer wall areas of the shunt pipe respectively; a1 and a2 are the convection heat transfer coefficients of the inner and outer flow channels of the shunt pipe respectively, which can be obtained by the heat transfer relationship; λ1 is the thermal conductivity of the shunt pipe; is the volume heat release rate of the manifold; l is the height of the area where the fuel assembly is located.

7. A fuel power calibration method for loop irradiation test according to claim 6, characterized in that: Qualitative temperature of coolant between shunt pipe and pressure pipe T f1 Calculated by the following formula: T f1 =(T1+T0) / 2 (5); Qualitative temperature of coolant in the shunt pipe T f2 Calculated by the following formula: T f2 =(T0+T2) / 2 (6) Among them, T0 is the inlet temperature of the fuel assembly, T1 is the coolant temperature of the outer wall of the upper end of the square box, and T2 is the outlet temperature of the fuel assembly.

8. A fuel power calibration method for loop irradiation test according to claim 1, characterized in that: In S4, the deviation of the nuclear power value of the fuel assembly before and after correction is compared. When the deviation is greater than the convergence range, the gamma heat release rate of the initially set material needs to be corrected, and the iterative cycle is repeated until the deviation of the nuclear power value of the fuel assembly before and after correction is within the convergence range.

9. A fuel power calibration method for loop irradiation test according to claim 1, characterized in that: The S5 comprises the following sub-steps: S51. Using the gamma heat release rate of the material and the nuclear power value of the fuel assembly as the heat source input of the high-precision CFD model, and using the thermal measurement data as the inlet boundary conditions of the high-precision CFD model, perform CFD analysis and calculation; S52, comparing the CFD calculation results with the actual irradiation test temperature measurement values, and based on the comparison results, further correcting the gamma heat release rate of each material and the nuclear power of the fuel assembly according to the pressure tube heat dissipation obtained by the high-precision CFD model; S53, repeating step S51 again using the updated heat source to perform CFD analysis and calculation; S54, repeat the above steps, iterate until the temperature calculation value and the measured value are consistent, and obtain the final material heat release rate and the nuclear power of the fuel core; S55. Regularly update the gamma heat release rate ratio of each material, re-iterate the material gamma heat release rate and the nuclear power of the fuel assembly, and obtain the burnup of the fuel assembly in the irradiation test through the time integration of the nuclear power.

10. A fuel power calibration method for loop irradiation test according to claim 9, characterized in that: The thermal measurement data in S51 refers to the temperature measurement value of the outer wall of the upper end of the square box and the coolant flow rate.

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