A method for calculating the bending of a pressurized water reactor fuel assembly

Abstract

A kind of pressurized water reactor fuel assembly bending calculation method, by pressurized water reactor assembly calculation program, the fuel assembly of different water gap size is calculated, obtains the few-group constant of each type fuel assembly different water gap size;The few-group constant of each type fuel assembly is functioned, and the few-group constant library of fuel assembly of different water gap size is obtained by merging;Based on the measured data of pressurized water reactor fuel assembly bending, obtain the three-dimensional water gap distribution data of reactor core;Based on the three-dimensional water gap distribution data of reactor core, update each square block of reactor core to rectangular block, obtain the few-group constant of rectangular block by linear interpolation to water gap;Based on the conservation of interfacial neutron flux density and the conservation of interfacial neutron flux density of rectangular block, the staggered grid of rectangular block caused by the difference of water gap size is considered in reactor core calculation;The calculation of key physical quantity of reactor core is completed by pressurized water reactor core calculation program.The present application can provide accurate calculation method for pressurized water reactor fuel assembly bending, and fully guarantee the safety of reactor operation.

Description

Technical Field

[0001] This invention relates to the field of pressurized water reactor core physics calculation, and more specifically to a method for calculating the bending of pressurized water reactor fuel assemblies. Background Technology

[0002] When a pressurized water reactor is operating, fuel assemblies deform due to irradiation growth, hydraulic forces, and other factors, deviating from their normal vertical position and forming various curved shapes, such as "C" and "S" shapes. Excessive bending can affect the safe operation of the reactor, leading to problems such as difficulties in loading and unloading fuel, excessively high local power, difficulty in inserting control rods, and quadrant power tilting. Therefore, accurately simulating fuel assembly bending is crucial for reactor numerical simulation and safe operation.

[0003] From a physics perspective, fuel assembly bending directly alters the size of the water gap between fuel assemblies. From a neutronics perspective, increasing the water gap size around the fuel assembly enhances neutron moderation; conversely, decreasing the water gap size weakens neutron moderation, thus affecting physical parameters such as core power distribution and power peak factor. In pressurized water reactor (PWR) numerical simulations, commonly used software includes deterministic and probabilistic programs. Deterministic programs can be further divided into one-step and two-step methods. While it's feasible to directly simulate the core physics model after fuel assembly bending using either a one-step or probabilistic method, the modeling process is complex and computationally expensive, making it unsuitable for commercial PWRs. In a two-step method, the two-dimensional assembly is first non-uniformly calculated using a PWR assembly calculation program and then homogenized to generate a minority group constant. Next, the core is divided into sections using a PWR core calculation program, and three-dimensional neutron diffusion is calculated. Typically, a minority group constant is generated under standard water gap sizes, and the core sections are neatly arranged, with each section having a radial dimension equal to or one-quarter the size of a fuel assembly. Therefore, fuel assembly bending poses two challenges to the two-step procedure for pressurized water reactor numerical simulation: how to account for the effect of fuel assembly bending on the minority group constant in the assembly calculation; and how to account for the effect of bending on neutron diffusion calculation in the core calculation.

[0004] Simulations of fuel assembly bending based on a two-step method have been conducted both domestically and internationally. Westinghouse developed the ANCGAP program based on the ANC program for simulating fuel assembly bending. The PHOENIX-P program for pressurized water reactor (PWR) assembly calculations can generate minority group constants for different water gap sizes, which are then corrected for the actual volume in core calculations before use; however, it does not consider the impact of fuel assembly bending on three-dimensional neutron diffusion calculations. In China, the CGN Research Institute improved the PCM software package. The PINE program for PWR assembly calculations can generate minority group constants for different water gap sizes and obtain a multi-parameter table of neutron cross-sections. The COCO program for PWR core calculations interpolates the minority group constants of the nodes based on the water gap distribution and re-divides the nodes, but it only divides the nodes under the original fuel assembly geometry, making it unsuitable for situations with large bending amplitudes, such as when two bent assemblies are in direct contact. Summary of the Invention

[0005] To overcome the problems existing in the prior art, the present invention aims to provide a method for calculating the bending of pressurized water reactor fuel assemblies. This method involves obtaining a library of minority group constants for fuel assemblies with different water gap sizes using a pressurized water reactor assembly calculation program, obtaining three-dimensional water gap distribution data of the reactor core from measured data of fuel assembly bending, updating each square segment of the reactor core to a rectangular segment based on the three-dimensional water gap distribution data, and considering the interleaving of rectangular segment meshes caused by differences in water gap size in the core calculation by conserving the surface neutron flux density and surface neutron flux density between rectangular segments. Finally, the method completes the calculation of key physical quantities of the reactor core under the condition of fuel assembly bending using a pressurized water reactor core calculation program.

[0006] To achieve the above objectives, the present invention adopts the following technical solution:

[0007] A method for calculating the bending of pressurized water reactor (PWR) fuel assemblies is characterized by: based on measured data of PWR fuel assembly bending, and using a two-step calculation strategy of assembly calculation and core calculation employed in PWRs, a library of minority group constants for fuel assemblies with different water gap sizes is created through the PWR assembly calculation program. Furthermore, the principles of conservation of surface neutron flux density and surface neutron flux density between rectangular nodes are introduced into the PWR core calculation program to complete the calculation of key core physical quantities under fuel assembly bending conditions, obtaining numerical simulation results of the key core physical quantities; including the following steps:

[0008] Step 1: Obtain the material and geometric information of the pressurized water reactor core, construct two-dimensional physical models of different types of fuel assemblies, and determine the state parameters of the calculation operating point, including fuel temperature, moderator temperature, boron concentration, and burnup depth.

[0009] Step 2: Use the pressurized water reactor assembly calculation program at a standard water gap size w iThe following calculations were performed on two-dimensional physical models of various types of fuel assemblies to obtain the standard water gap size w. i Minority group constant Σ(w) for various types of fuel assemblies i ), including the absorption section and the fission section; the water gap size of the two-dimensional physical model of the fuel assembly is changed to w respectively. j and w k And the water gap size was calculated to be w. j The minority group constant Σ(w) of various types of fuel components j ) and water gap size w k Minority group constant Σ(w) of various types of fuel components k );

[0010] Step 3: For the same water gap size w in Step 2 x The minority group constant Σ(w) under (x=i,j,k) x The function is used to obtain the water gap size w. x The corresponding fuel assembly minority group constant library; set the corresponding water gap size w for each fuel assembly minority group constant library. x The tags are merged to obtain a library of fuel assembly minority group constants with different water gap sizes;

[0011] Step 4: Based on the measured data of the pressurized water reactor fuel assembly bending, process the data to obtain the three-dimensional water gap distribution data of the core, which includes the water gap size of each layer on the four sides of the fuel assembly along the axial direction; during processing, the water gaps between fuel assemblies are evenly distributed to adjacent fuel assemblies, and the total water gap size is kept consistent with that before the pressurized water reactor fuel assembly bending.

[0012] Step 5: Based on the standard water gap size, the pressurized water reactor core calculation program divides the pressurized water reactor fuel assembly radially into square segments of one fuel assembly size or one-quarter of the fuel assembly size; firstly, based on the core three-dimensional water gap distribution data obtained in Step 4, each square segment is updated to a rectangular segment, and the water gap size w of the rectangular segment is obtained. real Secondly, based on the water gap size w of each rectangular segment... real Using the fuel assembly minority group constant library with different water gap sizes obtained in step 3, the minority group constant Σ(w) of each rectangular segment is determined by linear interpolation of the water gap. real The linear interpolation formula is shown in formula (1);

[0013]

[0014] In the formula:

[0015] c i —The water gap size is w i The interpolation coefficients of the corresponding fuel assembly minority group constant library;

[0016] c j—The water gap size is w j The interpolation coefficients of the corresponding fuel assembly minority group constant library;

[0017] c k —The water gap size is w k The interpolation coefficients of the corresponding fuel assembly minority group constant library;

[0018] Step 6: In order to consider the rectangular node mesh interlacing caused by updating each square node to a rectangular node in Step 5 in the pressurized water reactor core calculation program, the surface neutron flux density conservation and surface neutron flow density conservation between rectangular nodes are used as the boundary coupling conditions of the rectangular nodes, and the multi-group coarse mesh finite difference equation in the pressurized water reactor core calculation program is improved; the surface neutron flux density conservation and surface neutron flow density conservation of adjacent rectangular nodes k and k+1 are shown in Equation (2) and Equation (3), respectively;

[0019]

[0020]

[0021] In the formula:

[0022] —The discontinuity factor of the interface of the g-th energy group u in the positive direction of the rectangular nodal block k;

[0023] —The average surface neutron flux density at the interface of the g-th energy group u in the rectangular nodal k;

[0024] —The area of ​​the rectangular segment k on the interface in the positive u direction;

[0025] —The discontinuity factor of the negative direction interface of the g-th energy group u of the rectangular section k+1;

[0026] —The average surface neutron flux density at the interface of the g-th energy group u in the rectangular section k+1; —The area of ​​rectangular segment k+1 on the interface in the negative u direction;

[0027] —The average surface neutron flux density at the interface of the g-th energy group u in the rectangular nodal k; —The average surface neutron flux density at the interface of the g-th energy group u in the rectangular section k+1;

[0028] For rectangular node k, the improved multi-group coarse mesh finite difference equation is shown in formula (4);

[0029]

[0030] In the formula:

[0031] Δu k —The length of the rectangular segment k from the interface in the negative u direction to the interface in the positive u direction;

[0032] — Coarse-mesh finite-difference coupling factor of the g-th energy group u in the positive direction of the rectangular node k;

[0033] —Coupling correction factor in the positive direction of the g-th energy group u of the rectangular nodal k;

[0034] — Coarse-mesh finite-difference coupling factor of the g-th energy group u in the negative direction of the rectangular node k;

[0035] —Coupling correction factor in the negative direction of the g-th energy group u of the rectangular nodal k;

[0036] —The area of ​​the rectangular segment k on the interface in the negative u direction;

[0037] —The area of ​​rectangular segment k-1 on the interface in the positive u direction;

[0038] —The discontinuity factor of the interface in the negative direction of the g-th energy group u of the rectangular nodal block k;

[0039] —The discontinuity factor of the interface of the g-th energy group u in the positive direction of the rectangular nodal k-1;

[0040] —The average neutron flux density of the g-th energy group in the k+1 rectangular section;

[0041] —The average neutron flux density of the g-th energy group of the k-th rectangular nodal;

[0042] —The average neutron flux density of the g-th energy group in the k-1 rectangular section;

[0043] —The cross section of the g-th energy group of the rectangular nodal block k;

[0044] G—Total energy group number;

[0045] —The scattering cross section of rectangular nodal k from the h-th energy group to the g-th energy group;

[0046] —The average neutron flux density of the h-th energy group in the k-th rectangular nodal;

[0047] υ — the average number of neutrons produced per fission cycle;

[0048] χ g —The fission energy spectrum of the g-th energy group;

[0049] k eff —Effective multiplication coefficient;

[0050] —The fission cross section of the h-th energy group of the rectangular nodal k;

[0051] Step 7: Based on the pressurized water reactor core calculation program considering the rectangular segment grid in Step 6, and combined with the minority group constants of each rectangular segment obtained in Step 5, complete the three-dimensional neutron diffusion calculation of the pressurized water reactor fuel assembly bending, and obtain the simulation results of key physical quantities of the core, including critical boron concentration, temperature coefficient and control rod group integral value.

[0052] Compared with the prior art, the present invention has the following advantages:

[0053] 1. By using measured data of pressurized water reactor fuel assembly bending, the three-dimensional water gap distribution data of the reactor core is obtained, and the square segments divided in the pressurized water reactor core calculation program are updated to rectangular segments, so as to consider the impact of pressurized water reactor fuel assembly bending on the calculation of neutron diffusion in the reactor core;

[0054] 2. Based on the "two-step" calculation strategy of component calculation and core calculation commonly used in pressurized water reactors, a small group constant library of fuel assemblies with different water gap sizes is created through the pressurized water reactor component calculation program. In the pressurized water reactor core calculation program, the conservation of surface neutron flux density and surface neutron flow density between rectangular segments is introduced to consider the mesh stagger of rectangular segments caused by the difference in water gap size, so as to achieve accurate calculation. Attached Figure Description

[0055] Figure 1 Calculation process for pressurized water reactor fuel assembly bending;

[0056] Figure 2 A schematic diagram of a radial two-dimensional physical model of a 17×17 pressurized water reactor fuel assembly without gadolinium rods;

[0057] Figure 3 A schematic diagram of a radial two-dimensional physical model of a 17×17 pressurized water reactor fuel assembly containing 8 gadolinium rods;

[0058] Figure 4 A radial diagram illustrating the replacement of square segments with rectangular segments after bending a 3×3 fuel assembly.

[0059] Figure 5 This is a schematic diagram illustrating the conservation of surface neutron flux density and surface neutron flow density when rectangular nodes k and k+1 grids are interleaved. Detailed Implementation

[0060] The present invention will now be described in further detail with reference to the accompanying drawings and specific embodiments.

[0061] This invention, based on measured data of pressurized water reactor (PWR) fuel assembly bending, utilizes a two-step calculation strategy involving assembly and core calculations. It employs a PWR assembly calculation program to create a library of minority group constants for fuel assemblies with different water gap sizes. Furthermore, it introduces the principles of conservation of surface neutron flux density and surface neutron flux density between rectangular nodes into the PWR core calculation program. This allows for the calculation of key core physical quantities under fuel assembly bending conditions, yielding numerical simulation results for these key core physical quantities. Specific steps are as follows: Figure 1 As shown.

[0062] Step 1: Obtain the material and geometric information of the pressurized water reactor core, and construct two-dimensional physical models of different types of fuel assemblies, including enrichment levels and the number of gadolinium rods; such as Figure 2 The image shows a radial two-dimensional physical model of a 17×17 pressurized water reactor fuel assembly without gadolinium rods. The fuel assembly contains 265 fuel rods and 24 guide tubes, and has an outer layer of water gaps. Figure 3 The figure shows a two-dimensional physical model of a 17×17 pressurized water reactor fuel assembly containing 8 gadolinium rods. The fuel assembly includes 257 fuel rods, 8 gadolinium rods and 24 guide tubes, and has a water gap on the outside. The state parameters of the calculation operating point are determined, including fuel temperature, moderator temperature, boron concentration and burnup depth.

[0063] Step 2: Using the pressurized water reactor assembly calculation program LOCUST at a standard water gap size w i The following calculations were performed on two-dimensional physical models of various types of fuel assemblies to obtain the standard water gap size w. i Minority group constant Σ(w) for various types of fuel assemblies i ), including the absorption section and the fission section; the water gap size of the two-dimensional physical model of the fuel assembly is changed to w respectively. j and w k And the water gap size was calculated to be w. j The minority group constant Σ(w) of various types of fuel components j ) and water gap size w k Minority group constant Σ(w) of various types of fuel components k );

[0064] Step 3: Use the function-based program LtoS to process the same water gap size w from Step 2. x The minority group constant Σ(w) under (x=i,j,k) x The function is used to obtain the water gap size w. x The corresponding fuel assembly minority group constant library; set the corresponding water gap size w for each fuel assembly minority group constant library.x The tags are merged to obtain a library of fuel assembly minority group constants with different water gap sizes;

[0065] Step 4: Based on the measured data of the pressurized water reactor fuel assembly bending, specifically including the bending amount of the fuel assembly in the x and y directions at the grid position, process to obtain the core three-dimensional water gap distribution data including the water gap size of the four sides of each layer of the fuel assembly in the axial direction; during processing, the water gap between fuel assemblies is evenly distributed to adjacent fuel assemblies, and the total water gap size is kept consistent with that before the pressurized water reactor fuel assembly bending.

[0066] Step 5: Based on the standard water gap size, the pressurized water reactor (PWR) core calculation program SPARK divides the PWR fuel assembly radially into square segments of one fuel assembly size or one-quarter of the fuel assembly size. To account for the influence of PWR fuel assembly bending on the minority group constant, based on the core three-dimensional water gap distribution data obtained in Step 4, each square segment is first updated to a rectangular segment, and the water gap size w of the rectangular segment is obtained. real ,like Figure 4 The diagram shows a radial representation of the replacement of square segments with rectangular segments after a 3×3 pressurized water reactor fuel assembly is bent. The radial direction of each segment is taken as the size of one fuel assembly. After the central fuel assembly bends upwards, the square segments of this fuel assembly and the adjacent fuel assemblies above and below it are replaced with rectangular segments. Due to differences in the water gap size, the rectangular segments intersect with the adjacent square segments on the left and right, forming a grid of intersecting segments. Furthermore, based on the water gap size w of each rectangular segment... real Using the fuel assembly minority group constant library with different water gap sizes obtained in step 3, the minority group constant Σ(w) of each rectangular segment is determined by linear interpolation of the water gap. real The linear interpolation formula is shown in formula (1);

[0067]

[0068] In the formula:

[0069] c i —The water gap size is w i The interpolation coefficients of the corresponding fuel assembly minority group constant library;

[0070] c j —The water gap size is w j The interpolation coefficients of the corresponding fuel assembly minority group constant library;

[0071] c k —The water gap size is w k The interpolation coefficients of the corresponding fuel assembly minority group constant library;

[0072] Step 6: To account for the rectangular node mesh interleaving caused by updating the square nodes to rectangular nodes in Step 5 in the pressurized water reactor core calculation program SPARK, the conservation of surface neutron flux density and surface neutron flux density between rectangular nodes is used as the node boundary coupling condition, such as... Figure 5 As shown, the surface neutron flux density and surface neutron flux density of the rectangular nodes k and k+1 with interlaced grids are conserved on adjacent surfaces. In the figure, the grids of the adjacent surfaces of the rectangular nodes k and k+1 in the positive u direction are interlaced. The surface neutron flux density and surface neutron flux density of the adjacent nodes k in the positive u direction are equal to the surface neutron flux density and surface neutron density of the node k+1 in the negative u direction. The multi-group coarse mesh finite difference equation in the pressurized water reactor core calculation program SPARK is improved. The surface neutron flux density and surface neutron flux density conservation of adjacent rectangular nodes k and k+1 are shown in Equation (2) and Equation (3), respectively.

[0073]

[0074]

[0075] In the formula:

[0076] —The discontinuity factor of the interface of the g-th energy group u in the positive direction of the rectangular nodal block k;

[0077] —The average surface neutron flux density at the interface of the g-th energy group u in the rectangular nodal k;

[0078] —The area of ​​the rectangular segment k on the interface in the positive u direction;

[0079] —The discontinuity factor of the negative direction interface of the g-th energy group u of the rectangular section k+1;

[0080] —The average surface neutron flux density at the interface of the g-th energy group u in the rectangular section k+1; —The area of ​​rectangular segment k+1 on the interface in the negative u direction;

[0081] —The average surface neutron flux density at the interface of the g-th energy group u in the rectangular nodal k;

[0082] —The average surface neutron flux density at the interface of the g-th energy group u in the rectangular section k+1;

[0083] For rectangular node k, the improved multi-group coarse mesh finite difference equation is shown in formula (4);

[0084]

[0085] In the formula:

[0086] Δu k —The length of the rectangular segment k from the interface in the negative u direction to the positive u direction;

[0087] — Coarse-mesh finite-difference coupling factor of the g-th energy group u in the positive direction of the rectangular node k;

[0088] —Coupling correction factor in the positive direction of the g-th energy group u of the rectangular nodal k;

[0089] — Coarse-mesh finite-difference coupling factor of the g-th energy group u in the negative direction of the rectangular node k;

[0090] —Coupling correction factor in the negative direction of the g-th energy group u of the rectangular nodal k;

[0091] —The area of ​​the rectangular segment k on the interface in the negative u direction;

[0092] —The area of ​​rectangular segment k-1 on the interface in the positive u direction;

[0093] —The discontinuity factor of the interface in the negative direction of the g-th energy group u of the rectangular nodal block k;

[0094] —The discontinuity factor of the interface of the g-th energy group u in the positive direction of the rectangular nodal k-1;

[0095] —The average neutron flux density of the g-th energy group in the k+1 rectangular section;

[0096] —The average neutron flux density of the g-th energy group of the k-th rectangular nodal;

[0097] —The average neutron flux density of the g-th energy group in the k-1 rectangular section;

[0098] —The cross section of the g-th energy group of the rectangular nodal block k;

[0099] G—Total energy group number;

[0100] —The scattering cross section of rectangular nodal k from the h-th energy group to the g-th energy group;

[0101] —The average neutron flux density of the h-th energy group in the k-th rectangular nodal;

[0102] υ — the average number of neutrons produced per fission cycle;

[0103] χ g —The fission energy spectrum of the g-th energy group;

[0104] k eff —Effective multiplication coefficient;

[0105] —The fission cross section of the h-th energy group of the rectangular nodal k;

[0106] Step 7: Based on the SPARK core calculation program for pressurized water reactors (PWRs) considering the interlaced rectangular segment meshes in Step 6, and combined with the minority group constants of each rectangular segment obtained in Step 5, the three-dimensional neutron diffusion calculation for the bending of the PWR fuel assembly is completed, and the simulation results of key physical quantities of the core are obtained. These key physical quantities include critical boron concentration, temperature coefficient, and control rod assembly integral value. After applying the above steps to the calculations at the Taishan Nuclear Power Plant using the LOCUST PWR assembly calculation program and the SPARK PWR core calculation program, the calculation results and errors of the critical boron concentration and temperature coefficient for a certain cycle of startup physical test of a certain unit at the Taishan Nuclear Power Plant are shown in Table 1, and the control rod assembly integral value and relative error for a certain cycle of startup physical test of a certain unit at the Taishan Nuclear Power Plant are shown in Table 2.

[0107] Table 1

[0108] Test content Calculated value error Error limit requirements Critical boron concentration 958.7ppm 31.5ppm ±50ppm Isothermal temperature coefficient -33.1 pcm / K -1.1pcm / K ±3.6pcm / K Temperature coefficient of moderator -29.9pcm / K -1.1pcm / K ±3.6pcm / K

[0109] Table 2

[0110]

[0111] Numerical results show that the critical boron concentration error is less than the limit requirement of ±50 ppm, the isothermal temperature coefficient and moderator temperature coefficient errors are both less than the limit requirement of ±3.6 pcm / K, and the relative error of the integral value of the control rod assembly is much lower than the limit requirement of ±10%. Therefore, the method proposed in this invention can achieve accurate calculation of pressurized water reactor fuel assembly bending and has industrial application value.

Claims

1. A method for calculating the bending of pressurized water reactor fuel assemblies, characterized in that: According to the measured data of the PWR fuel assembly bending, based on the two-step calculation strategy of assembly calculation and core calculation used by PWR, through the PWR assembly calculation program to make the fuel assembly few-group constant library of different water gap size and introduce the principle of face neutron flux density conservation and face neutron flow density conservation between rectangular blocks in the PWR core calculation program, the calculation of the key physical quantities of the core under the condition of fuel assembly bending is completed, and the numerical simulation results of the key physical quantities of the core are obtained; including the following steps: Step 1: Obtain the material and geometric information of the PWR core, construct the two-dimensional physical model of different types of fuel assemblies, and determine the state parameters of the calculation working point, including fuel temperature, moderator temperature, boron concentration and burnup depth; Step 2: Use the pressurized water reactor assembly calculation program at a standard water gap size w i The following calculations were performed on two-dimensional physical models of various types of fuel assemblies to obtain the standard water gap size w. i Minority group constant Σ(w) for various types of fuel assemblies i ), including the absorption section and the fission section; the water gap size of the two-dimensional physical model of the fuel assembly is changed to w respectively. j and w k And the water gap size was calculated to be w. j Minority group constant Σ(w) of various types of fuel components j ) and water gap size w k Minority group constant Σ(w) of various types of fuel components k ); Step 3: For the same water gap size w in Step 2 λ The minority group constant Σ(w) λ The function is used to obtain the water gap size w. λ The corresponding fuel assembly minority group constant function; set the corresponding water gap size w for each fuel assembly minority group constant function. λ The tags are merged to obtain a library of fuel assembly minority group constants with different water gap sizes; where λ=i,j,k; Step 4: Based on the measured data of the PWR fuel assembly bending, process the obtained three-dimensional water gap distribution data of the core containing the four-side water gap size of each layer of the fuel assembly in the axial direction; when processing, the water gap between the fuel assemblies is evenly distributed to the adjacent fuel assemblies, and the total water gap size is kept consistent before the PWR fuel assembly bending; Step 5: Based on the standard water gap size, the PWR core calculation program divides the PWR fuel assembly radially into square segments of one fuel assembly size or one quarter fuel assembly size; first, based on the three-dimensional water gap distribution data obtained in Step 4, update each square segment to a rectangular segment and obtain the water gap size w of the rectangular segment real ; second, based on the water gap size w of each rectangular segment real and the fuel assembly few-group constant library of different water gap sizes obtained in Step 3, determine the few-group constant Σ(w real ) of each rectangular segment by linear interpolation of the water gap, and the linear interpolation formula is shown in formula (1) Σ(w) = Σ(w0) + (Σ(w1) - Σ(w0)) * (w - w0) / (w1 - w0) (1) In the formula: c i - water gap size w i corresponding fuel assembly few-group constant library interpolation coefficients; c j - water gap size w j corresponding fuel assembly few-group constant library interpolation coefficients; c k — water gap size is w k corresponding fuel assembly few-group constant library interpolation coefficients; Step 6: In order to consider the grid interlacing of the rectangular block caused by updating each square block to rectangular block in step 5 in the PWR core calculation program, the face neutron flux density conservation and face neutron flow density conservation between rectangular blocks are taken as the coupling conditions of the rectangular block boundary, and the multi-group coarse mesh finite difference equation in the PWR core calculation program is improved; the face neutron flux density conservation and face neutron flow density conservation of adjacent rectangular blocks k and k+1 are shown in formula (2) and formula (3) respectively; In the formula: - discontinuity factor of the positive direction interface of the rectangular bin k, the gth energy group, and the u direction; — average fission neutron flux density in the face of the positive u-direction interface of rectangular cell k, gth energy group, u; — area of the positive u-direction interface of rectangular cell k; - discontinuity factor of the negative direction interface of the rectangular bin k+1st g energy group u; — average fission neutron flux density in the negative u-direction interface of rectangular cell k+1, gth energy group, u; — area of the negative u-direction interface of rectangular cell k+1, gth energy group, u; — average surface neutron flux density in the positive direction interface of the rectangular cell k, gth energy group, u; — average surface neutron flux density in the negative direction interface of the rectangular cell k+1, gth energy group, u; For rectangular block k, the improved multi-group coarse mesh finite difference equation is shown in formula (4); In the formula: Δu k — Length of rectangular bin k from u-negative direction interface to u-positive direction interface; - coarse mesh finite difference coupling factor for the positive direction of the u-th energy group of the k-th rectangular block; - coupling correction factor for the positive direction of the u-th energy group of the k-th rectangular block; - coarse mesh finite difference coupling factor for the negative direction of the u-th energy group in the k-th rectangular block; - coupling correction factor for the negative direction of the u-th energy group in the k-th rectangular block; - Area of rectangular bin k at the interface in the negative u direction; - Area of rectangular bin k-1 at the u positive direction interface; - discontinuity factor of the negative direction interface of the rectangular bin k, the gth energy group, and the u direction; - discontinuity factor of the positive direction interface of the rectangular bin k-1st g energy group u; - average neutron flux density of the rectangular bin k+1, gth energy group; - average neutron flux density of the rectangular bin k, gth energy group; - average neutron flux density of the rectangular cell k-1st g energy group; - the removed section of the rectangular block kthg energy group; G——total energy group number; - the scattering cross section of the rectangular cell k from the hth energy group to the gth energy group; - average neutron flux density of the hth energy group in the rectangular cell k; υ——average number of neutrons produced per fission; χ g - fission energy spectrum of the gth group of energies; k eff — effective multiplication factor; - fission cross section of the rectangular bin k for the hth energy group; Step 7: Based on the PWR core calculation program considering the grid interlacing of the rectangular block in step 6, combined with the few-group constants of each rectangular block obtained in step 5, the three-dimensional neutron diffusion calculation of the PWR fuel assembly bending is completed, and the simulation results of the key physical quantities of the core are obtained, including the critical boron concentration, temperature coefficient and control rod group integral value.

Images