An annular experimental device for simulating thermal-hydraulic behavior of fuel rod bending in a nuclear reactor core

Abstract

The application discloses a ring-shaped experimental device for simulating thermal-hydraulic behavior of fuel rod bending in a reactor core, which comprises an inner tube and an outer tube sleeved together, two ends of the inner tube are provided with inner tube electrodes for heating the inner tube, the inner tube is sealed with the outer tube through sealing sleeves at the upper and lower ends, the upper and lower ends of the outer tube are provided with outer tube electrodes for heating the outer tube, the side wall of the outer tube is connected with an inlet pipe and an outlet pipe, a working medium flows into the flow channel between the inner tube and the outer tube from the inlet pipe and finally flows out from the outlet pipe, the two ends of the outer tube are connected with connecting flanges, and a plurality of positioning grids are installed on the outer tube. Compared with the complex form of the rod bundle geometry, the application adopts a simple ring-shaped channel to simulate the thermal-hydraulic behavior of the fuel rod bending in the reactor core. Unlike a circular tube, the geometric form of the ring-shaped channel can consider the rod bundle effect, such as the gap size, the curvature and the existence of adjacent heating surfaces.

Description

Technical Field

[0001] This invention belongs to the field of nuclear reactor thermal-hydraulic experimental technology, specifically relating to a ring-shaped experimental device for simulating the bending thermal-hydraulic behavior of reactor core fuel rods. Background Technology

[0002] Modern pressurized water reactors generally employ rod-bundle nuclear fuel assemblies without a housing and with finger-shaped control components. The fuel elements within the assembly are arranged in a square pattern. A nuclear fuel assembly consists of fuel rods, guide tubes, positioning grids, and upper and lower mounting bases. In the Daya Bay Nuclear Power Plant, the fuel rods in the fuel assembly are arranged in a 17×17 square. The guide tubes connect to eight layers of grids and the upper and lower mounting bases, forming the basic structural framework of the fuel assembly. The fuel rods are supported and clamped within this framework, with the spacing between the rods remaining constant along the entire length of the assembly. Each assembly has 289 cells, with 24 guide tubes and one in-core flux measurement tube located within each cell, and the remaining 264 cells containing fuel rods. Based on the type of cell, core cells are divided into typical cells (all composed of fuel rods) and guide tube cells (composed of fuel rods and guide tubes or measurement tubes). During reactor operation, the thermohydraulic characteristics of guide tube cells differ significantly from those of typical cells due to the cold wall effect. Furthermore, at the end of the reactor core's lifespan, the release of fission gases and irradiation cause varying degrees of thermal expansion and swelling in the fuel rods and cladding, leading to more complex core thermal-hydraulic characteristics. Among these factors, the impact on the critical heat flux density is a crucial and non-negligible element in core thermal-hydraulic design.

[0003] Critical heat flux density is a crucial safety parameter in nuclear reactors, with reactor power limited within its range. Exceeding this density deteriorates heat transfer between the fuel cladding and coolant, leading to a rapid rise in cladding temperature and potential failure. The existence of critical heat flux density significantly challenges the safety and operational limits of nuclear reactors. Therefore, accurately predicting the critical heat flux density of the fuel cladding will help optimize power output and improve reactor operational safety. Furthermore, since the mechanisms by which cold wall effects and fuel bending affect critical heat flux density are not yet fully understood, further experimental research on the mechanisms of critical heat flux density is needed.

[0004] Traditional studies investigating the critical heat flux density of fuel rod bundles utilize 5×5 bundle channels, which have been proven effective in reflecting the thermal properties of the core's critical heat flux density. However, due to the high cost of bundle experiments and the inability to conduct large-scale experiments with varying degrees of bending, these methods are generally limited to engineering verification and cannot be used to study the mechanism by which rod bending affects the critical heat flux density. This hinders progress in critical heat flux density research and the development of novel fuel assemblies. Therefore, rapidly simulating the thermo-hydraulic behavior of fuel rod bending in a simple channel is an important method for investigating the mechanism of the critical heat flux density of fuel rod bundles. Summary of the Invention

[0005] To address the aforementioned problems and meet experimental requirements, the present invention aims to provide a ring-shaped experimental apparatus for simulating the bending thermal-hydraulic behavior of reactor core fuel rods. The apparatus is ring-shaped, designed based on the principle that the hydraulic equivalent diameter is the same as that of the cell under study. The inner and outer tubes can be heated separately or simultaneously to simulate typical cells and explore the mechanism of the cold wall effect. The outer tube is divided into four segments connected by clamps, and positioning grids with different eccentricities can be flexibly placed at three connection points to investigate the effects of different degrees of eccentricity, bending degree, and bending position on thermal-hydraulic characteristics. This invention overcomes the shortcomings of traditional large-scale fuel rod bundle critical heat flux density experimental apparatuses and provides new technical support for thermal-hydraulic experimental research within complex, irregularly shaped fuel rod bundle channels.

[0006] The present invention is achieved using the following technical solution:

[0007] A ring-shaped experimental device for simulating the bending thermal-hydraulic behavior of reactor core fuel rods includes an inner tube and an outer tube fitted together. Inner tube electrodes are installed at both ends of the inner tube for heating the inner tube. The inner tube is sealed to the outer tube by sealing sleeves at both ends. Outer tube electrodes are installed at both ends of the outer tube for heating the outer tube. An inlet pipe and an outlet pipe are connected to the side wall of the outer tube. The working fluid flows in from the inlet pipe, passes through the flow channel between the inner and outer tubes, and finally flows out from the outlet pipe. Connecting flanges are connected to both ends of the outer tube, and several positioning grids are installed on the outer tube.

[0008] A further improvement of the present invention is that two nickel rods are welded to the upper and lower ends of the inner tube.

[0009] A further improvement of the present invention is that the sealing sleeve has a circular groove for placing an O-ring, thereby achieving a seal.

[0010] A further improvement of the present invention is that the inlet pipe and the outlet pipe are located at both ends of the outer pipe sidewall.

[0011] A further improvement of the present invention is that the connecting flange is divided into an inner connecting flange and an outer connecting flange, the outer pipe is welded to the inner connecting flange, and insulating gaskets are installed in the inner connecting flange and the outer connecting flange and tightened with bolts to ensure their sealing.

[0012] A further improvement of the present invention is that the inner tube passes through the positioning grid to achieve insulation and positioning between the inner tube and the outer tube.

[0013] A further improvement of the present invention is that a hexagonal groove is milled in the cross section of the outer tube connection to accommodate the positioning grid, and a circular groove is provided in the cross section of the connection to accommodate the O-ring.

[0014] A further improvement of the present invention is that the positioning grid adopts a hexagonal design to match the hexagonal groove, and three protrusions are welded inside the positioning grid to achieve positioning of the inner tube.

[0015] A further improvement of the present invention is that the simulated bending of the reactor core fuel rods is achieved by replacing the positioning grids at multiple connections of the outer tube.

[0016] A further improvement of the present invention is that the outer diameter of the inner tube is the same as the outer diameter of the fuel rod in the fuel assembly, and the inner diameter of the outer tube is calculated by satisfying the principle that the hydraulic diameter of a typical grid element is the same as the hydraulic diameter of the annular channel between the inner and outer tubes.

[0017] Compared with the prior art, the present invention has at least the following beneficial technical effects:

[0018] This invention provides a ring-shaped experimental apparatus for simulating the bending thermal-hydraulic behavior of fuel rods in a reactor core.

[0019] (1) Compared to the complex geometry of rod bundles, this invention uses a simple annular channel to simulate the thermal-hydraulic behavior of core fuel rod bending. Unlike circular tubes, the geometry of the annular channel can take into account rod bundle effects, such as gap size, curvature, and the presence of adjacent heated surfaces;

[0020] (2) The outer tube is divided into four sections and connected by clamps. Different eccentricity positioning grids can be flexibly placed at the three connection points to explore the influence of different degrees of eccentricity, bending degree and bending position on the thermal hydraulic characteristics. The design of the hexagonal positioning grid can ensure the consistency of the eccentricity direction of the three positioning grids;

[0021] (3) An insulating layer of polyetheretherketone (PEEK) material covers the positioning grid, which can ensure the insulation between the inner tube and the outer tube. The experimental setup provides a variety of different heating methods, such as heating only the inner tube, heating only the outer tube, and heating both the inner and outer tubes simultaneously, which can be used to simulate and explore the effects of fuel rod bending under different types of core rod bundle grids.

[0022] This invention enables rapid simulation of the bending thermo-hydraulic behavior of fuel rods within a rod bundle channel using a low-cost annular channel. It overcomes the shortcomings of traditional large-scale rod bundle critical heat flux density experimental devices, provides new technical support for thermo-hydraulic experimental research in complex irregular rod bundle channels, and has significant engineering practical value. Attached Figure Description

[0023] Figure 1 This is a schematic diagram of the structure of a ring-shaped experimental device for simulating the bending thermal-hydraulic behavior of fuel rods in a reactor core according to the present invention.

[0024] Figure 2 This is a cross-sectional view of a typical grid cell in a fuel assembly;

[0025] Figure 3 This invention provides a cross-sectional view of an annular channel used to simulate a typical gate element;

[0026] Figure 4 This is a cross-sectional view of the connection point of outer tube 2;

[0027] Figure 5 This is a schematic diagram of the positioning grid 4 structure.

[0028] Explanation of reference numerals in the attached figures:

[0029] 1-Inner tube, 2-Outer tube, 3-Outer tube electrode, 4-Positioning grid, 5-Inner connecting flange, 6-Outer connecting flange, 7-Sealing sleeve, 8-Inner tube electrode, 9-Inlet pipe, 10-Outlet pipe, 11-Circular groove, 12-Hexagonal groove, 13-Flow channel, 14-Insulation layer. Detailed Implementation

[0030] The technical solution of the annular experimental apparatus for simulating the bending thermal-hydraulic behavior of reactor core fuel rods will be clearly and completely described below with reference to the accompanying drawings in the embodiments of the present invention. All other embodiments obtained by those skilled in the art based on the embodiments of the present invention without inventive effort are within the protection scope of the present invention.

[0031] This invention provides a ring-shaped experimental apparatus for simulating the bending thermal-hydraulic behavior of reactor core fuel rods. Figure 1 The schematic diagram of the experimental apparatus provided for the example of the present invention is used to describe the main components and connection methods of the inventive apparatus.

[0032] This invention provides a ring-shaped experimental device for simulating the bending thermal-hydraulic behavior of reactor core fuel rods. The main body of the experimental device consists of an inner tube 1 and an outer tube 2, which can be made of stainless steel or Inco nickel. Two nickel rods are welded to the upper and lower ends of the inner tube 1. The nickel rods have very low resistance and generate almost no heat, ensuring that the heating length of the inner tube 1 is the same as that of the outer tube 2. Inner tube electrodes 8 are installed at both ends of the inner tube 1 for direct heating. The inner tube 1 is sealed by sealing sleeves 7 at the upper and lower ends of the experimental device. The sealing sleeves 7 have circular grooves for placing O-rings, thereby achieving a seal. Outer tube electrodes 3 are installed at the upper and lower ends of the outer tube 2 for direct heating. The outer tube 2 is connected to the working fluid inlet pipe 9 and the outlet pipe 10. The working fluid flows from the inlet pipe 9 through the flow channel 13 between the inner tube 1 and the outer tube 2, and finally flows out from the outlet pipe 10. The outer tube 2 is welded to the inner connecting flange 5. Insulating gaskets are installed in the inner connecting flange 5 and the outer connecting flange 6 and tightened with bolts to ensure a seal. The outer tube 4 is divided into four sections, and positioning grids 4 are installed at three connection points. The inner tube 1 passes through the positioning grids 4 to achieve insulation and positioning between the inner tube 1 and the outer tube 2.

[0033] Furthermore, Figure 2 This is a typical cross-sectional view of a grid element in a fuel assembly, where D is the outer diameter of the fuel rod and P is the grid pitch between the fuel rods. Figure 3 This invention provides a cross-sectional view of an annular channel used to simulate a typical cell element, where Di is the outer diameter of inner tube 1 and Do is the inner diameter of outer tube 2. The outer diameter of inner tube 1 is the same as the outer diameter of the fuel rod in the fuel assembly, and the inner diameter of outer tube 2 is calculated by satisfying the principle that the hydraulic diameter of the typical cell element is the same as the hydraulic diameter of the annular channel between inner tube 1 and outer tube 2 in this invention. This achieves similarity between the geometry and hydraulic conditions, thereby simulating a typical cell element of the fuel assembly. The electrode 8 of the inner tube and the electrode 3 of the outer tube can control the heating power of inner tube 1 and outer tube 2 respectively, enabling heating of only inner tube 1, heating of only outer tube 2, and simultaneous heating of both inner and outer tubes. Different heating methods can be used to explore the effects of cold walls.

[0034] Furthermore, Figure 4 This is a cross-sectional view of the connection point of outer tube 2. To allow for flexible replacement of the positioning grid 4, outer tube 2 is designed as a series of four interconnected tube segments. The outer diameter of outer tube 2 at the connection point is larger than at other locations, while its inner diameter is the same as at other locations (both are Do). A hexagonal groove 12 is milled into the cross-section at the connection point of outer tube 2 to accommodate the positioning grid 4. Unlike a circular design, the hexagonal design ensures consistency in the eccentricity direction during the eccentricity test of inner tube 1. Furthermore, a circular groove 11 is present in the cross-section at the connection point to accommodate an O-ring. A seal is achieved by clamping the O-ring between the two connected outer tube segments 2.

[0035] Furthermore, Figure 5 The diagram shows the structure of the positioning grid 4, which also adopts a hexagonal design to match the hexagonal groove 12. Three protrusions are welded inside the positioning grid 4 to position the inner tube 1. Positioning grids 4 can be manufactured with different eccentricities according to requirements. An insulating layer 14 made of polyetheretherketone (PEEK) covers the positioning grid 4, ensuring insulation between the inner tube 1 and the outer tube 2.

[0036] Simulated core fuel rod bending is achieved by replacing the positioning grids 4 at the three joints of the outer tube 2 of the annular channel experimental device. For example, to simulate bending at the outlet position, the positioning grids 4 at the upstream and middle positions of the flow channel 13 are concentric grids, while the positioning grid 4 at the outlet position is replaced with an eccentric grid. To simulate eccentricity of the core fuel rods, the three positioning grids 4 are replaced with grids of the same eccentricity. The degree of bending and eccentricity is determined by the eccentricity of the positioning grids 4, and grids with different eccentricities can be manufactured according to requirements.

[0037] This invention provides a simple annular channel experimental setup that simulates the thermal-hydraulic behavior of fuel rod bending in a reactor core. This overcomes the shortcomings of traditional large-scale rod bundle critical heat flux density experimental setups and provides new technical support for thermal-hydraulic experimental research under complex conditions.

[0038] The above description is merely a specific example of the present invention, but the scope of protection of the present invention is not limited thereto. Any variations or substitutions that can be easily conceived by those skilled in the art within the scope of the technology disclosed in the present invention should be included within the scope of protection of the present invention. Therefore, the scope of protection of the present invention should be determined by the scope of the claims.

Claims

1. A ring-shaped experimental apparatus for simulating the bending thermal-hydraulic behavior of reactor core fuel rods, characterized in that, The inner tube (1) and outer tube (2) are assembled together. The inner tube (1) is equipped with inner tube electrodes (8) at both ends for heating the inner tube (1). The inner tube (1) is sealed to the outer tube (2) by sealing sleeves (7) at both ends. The outer tube (2) is equipped with outer tube electrodes (3) at both ends for heating the outer tube (2). The side wall of the outer tube (2) is connected to an inlet pipe (9) and an outlet pipe (10). The working fluid flows from the inlet pipe (9) through the flow channel (13) between the inner tube (1) and the outer tube (2) and finally flows out from the outlet pipe (10). The two ends of the outer tube (2) are connected with connecting flanges. Several positioning grids (4) are installed on the outer tube (2). Two nickel rods are welded to the upper and lower ends of the inner tube (1); the sealing sleeve (7) has a circular groove for placing an O-ring, thereby achieving a seal; The inner tube (1) passes through the positioning grid (4) to achieve insulation and positioning between the inner tube (1) and the outer tube (2); a hexagonal groove (12) is milled at the cross section of the connection of the outer tube (2) to place the positioning grid (4), and a circular groove (11) is placed at the cross section of the connection to place the O-ring; the positioning grid (4) adopts a hexagonal design to match the hexagonal groove (12), and three protrusions are welded inside the positioning grid (4) to achieve positioning of the inner tube (1); the simulated core fuel rod bending is achieved by replacing the positioning grid (4) at multiple connections of the outer tube (2).

2. The annular experimental apparatus for simulating the bending thermal-hydraulic behavior of reactor core fuel rods according to claim 1, characterized in that, The inlet pipe (9) and the outlet pipe (10) are located at both ends of the side wall of the outer pipe (2).

3. The annular experimental apparatus for simulating the bending thermal-hydraulic behavior of reactor core fuel rods according to claim 1, characterized in that, The connecting flange is divided into an inner connecting flange (5) and an outer connecting flange (6). The outer pipe (2) is welded to the inner connecting flange (5). Insulating gaskets are installed in the inner connecting flange (5) and the outer connecting flange (6) and tightened with bolts to ensure their sealing.

4. The annular experimental apparatus for simulating the bending thermal-hydraulic behavior of reactor core fuel rods according to claim 1, characterized in that, The outer diameter of the inner tube (1) is the same as the outer diameter of the fuel rod in the fuel assembly. The inner diameter of the outer tube (2) is calculated by satisfying the principle that the hydraulic diameter of a typical grid element is the same as the hydraulic diameter of the annular channel between the inner tube (1) and the outer tube (2).

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