A simulation experimental device for nuclear fuel behavior in loss of coolant accidents

Abstract

The invention relates to the technical field of nuclear fuel accident simulation, and discloses an experimental device for simulating the behavior of nuclear fuel in a loss of coolant accident. The device comprises a cladding, a pellet, an insulating member, a pressurized cooling pipe, and a cooling assembly. The pellet is arranged in the cladding, and the top wall and the bottom wall of the pellet are respectively connected to an external power supply via wires. An insulating member is arranged between the pellet and the cladding. A pressurized cooling pipe is arranged outside the cladding, and the pressurized cooling pipe is connected to the cooling assembly. The fission heat generation process of real nuclear fuel is simulated by energizing the pellet to generate self-heat. Combined with the synergistic effect of the pressurized cooling pipe and the cooling assembly, the thermal stress cracking of the pellet can be effectively restored, thereby realizing in-situ pellet cracking based on the same principle as real pellet cracking, and obtaining more realistic pellet cracking morphology and cladding deformation and rupture conditions, thereby improving the accuracy and reliability of experimental data.

Description

Technical Field

[0001] The invention relates to the technical field of nuclear fuel accident simulation, in particular to a nuclear fuel behavior simulation experimental device for a coolant loss accident. Background Art

[0002] Loss of coolant accident (LOCA) is a design basis accident for pressurized water reactors (PWRs). During a LOCA, the nuclear fuel rods undergo a complex series of behavioral changes, potentially threatening the structural integrity of the fuel elements and even the core. Therefore, research on fuel behavior and failure during LOCA is essential. The phenomena involved in LOCA accidents are complex, making numerical simulations extremely challenging. Experimental studies are necessary to gain a deeper understanding of the phenomena and mechanisms.

[0003] Currently, many experimental devices have been designed and established both domestically and internationally to investigate fuel behavior during LOCA accidents. Most existing experimental devices only consider the behavior of the cladding, not the pellets. These include direct heating of the cladding or the use of electric heating rods to simulate pellet heating. For example, patent publication number CN105070331B discloses an "experimental device for evaluating the performance of nuclear fuel cladding tubes under simulated LOCA conditions," which includes a mold cavity, a heating device, a water vapor circulation device, a vacuum pump, and a cooling device. The cladding is located within the mold cavity, and heating rods are placed within the cladding to simulate pellet heating. Existing devices that directly heat the cladding or use heating rods to heat the cladding ignore the mechanical effects on the cladding during pellet cracking and changes in heat transfer properties, potentially leading to experimental results that deviate from actual conditions.

[0004] In view of this, the existing technology still needs to be improved and developed. Summary of the Invention

[0005] In view of the above-mentioned deficiencies in the prior art, the purpose of the present invention is to provide a nuclear fuel behavior simulation experimental device for a loss of coolant accident, aiming to solve the problem that the LOCA accident simulation device in the prior art ignores the mechanical effects on the cladding and the changes in heat transfer characteristics during the cracking of the pellets, resulting in inaccurate experimental results.

[0006] The technical solutions adopted by the present invention to solve the technical problems are as follows:

[0007] A nuclear fuel behavior simulation experimental device for a loss of coolant accident, comprising:

[0008] cladding;

[0009] A core block is disposed inside the cladding; the top wall and the bottom wall of the core block are respectively connected to an external power source through wires;

[0010] an insulating member, disposed on the outer surface of the core block to isolate the core block from the cladding;

[0011] A pressurized cooling pipe is sleeved on the outside of the cladding and sealed with the cladding;

[0012] A cooling assembly is connected to the pressurized cooling pipe to form a coolant circulation path.

[0013] Furthermore, it also includes:

[0014] Two fixing flanges are respectively arranged at both ends of the pressurized cooling pipe; the pressurized cooling pipe is extended from both ends of the cladding, and the fixing flanges are respectively connected to the fixing flanges and the cladding.

[0015] Furthermore, an air filling port and an air releasing port are provided on the cladding outside the pressurized cooling pipe, and a first pressure gauge is provided inside the cladding.

[0016] Furthermore, it also includes:

[0017] Two springs are respectively arranged on the inner top wall and the inner bottom wall of the cladding; the two springs are respectively in contact with the top wall and the bottom wall of the core block.

[0018] Furthermore, the cooling assembly includes:

[0019] a first pipe, disposed on a side wall of the pressurized cooling pipe;

[0020] A first breach is provided on the first pipe;

[0021] a heat exchanger, disposed at an end of the first pipe away from the pressurized cooling pipe;

[0022] a second pipe disposed on the heat exchanger; one end of the second pipe being connected to the outlet of the heat exchanger, and the other end being disposed at the bottom of the pressurized cooling pipe and communicating with the interior of the pressurized cooling pipe;

[0023] a circulation pump, disposed on the second pipeline;

[0024] The second rupture is provided on the second pipe; by opening the first rupture and / or the second rupture, the processes of coolant spraying, flash evaporation, liquid level drop and evaporation can be simulated.

[0025] Furthermore, the cooling assembly further includes:

[0026] a reflooding water tank, disposed outside the pressurized cooling pipe; the reflooding water tank is connected to the interior of the pressurized cooling pipe through a pipe;

[0027] A water storage tank is arranged outside the pressurized cooling pipe; the water storage tank is connected to the second pipe through a connecting pipe, and the connection point between the connecting pipe and the second pipe is located on the side of the circulating pump away from the pressurized cooling pipe. The water storage tank cooperates with the reflooding water tank to simulate the refrigerant reflooding process.

[0028] Furthermore, it also includes:

[0029] A synchrotron radiation X-ray in-situ imaging device is provided on one side of the pressurized cooling pipe and is used to detect cracking of the core block and deformation of the cladding.

[0030] Furthermore, it also includes:

[0031] Two electrodes are respectively arranged on the top wall and the bottom wall of the core block; the electrodes penetrate the cladding and extend to the outside of the cladding, and the electrodes and the cladding are sealed.

[0032] Furthermore, there are multiple core blocks, and the multiple core blocks are stacked and located inside the pressurized cooling pipe.

[0033] Furthermore, an acoustic level gauge and a second pressure gauge are provided inside the pressurized cooling pipe.

[0034] Compared with the prior art, the present invention has the following beneficial effects:

[0035] In the present invention, a pellet is arranged in the cladding, the top wall and the bottom wall of the pellet are respectively connected to an external power supply through wires, an insulating member is provided between the pellet and the cladding, a pressurized cooling pipe is provided on the outside of the cladding, and the pressurized cooling pipe is connected to a cooling assembly; the fission heat generation process of real nuclear fuel is simulated by the self-heating of the pellet when it is energized, and the synergistic effect of the pressurized cooling pipe and the cooling assembly can effectively restore the thermal stress cracking of the pellet, thereby realizing in-situ pellet cracking based on the same principle as the real pellet cracking, and obtaining a more realistic pellet cracking morphology and the deformation and rupture of the cladding, thereby improving the accuracy and reliability of the experimental data. BRIEF DESCRIPTION OF THE DRAWINGS

[0036] Figure 1 It is a schematic diagram of the overall structure of the present invention.

[0037] The numbers in the figure are as follows: 1. cladding; 11. inflation port; 12. deflation port; 13. first pressure gauge; 14. spring; 2. core block; 21. electrode; 3. insulation; 4. pressurized cooling pipe; 5. cooling assembly; 51. first pipe; 52. first breach; 53. heat exchanger; 54. second pipe; 55. circulation pump; 56. second breach; 57. re-flooding tank; 58. water storage tank; 6. fixing flange; 7. synchrotron radiation X-ray in-situ imaging device; 8. acoustic level gauge; 9. second pressure gauge. DETAILED DESCRIPTION

[0038] In order to make the purpose, technical solution and effect of the present invention clearer and more specific, the present invention is further described in detail below with reference to the accompanying drawings and examples. It should be understood that the specific embodiments described herein are only used to explain the present invention and are not intended to limit the present invention.

[0039] In the description of the present invention, it should be understood that the terms "center", "longitudinal", "lateral", "up", "down", "front", "back", "left", "right", "vertical", "horizontal", "top", "bottom", "inside", "outside" and the like indicate orientations or positional relationships based on the orientations or positional relationships shown in the accompanying drawings, and are only for the convenience of describing the present invention and simplifying the description, rather than indicating or implying that the device or element referred to must have a specific orientation, be constructed and operated in a specific orientation, and therefore cannot be understood as limiting the present invention. In addition, the terms "first" and "second" are used for descriptive purposes only, and cannot be understood as indicating or implying relative importance or implicitly indicating the number of the indicated technical features. Therefore, the features defined as "first" and "second" may explicitly or implicitly include one or more of the features. In the description of the present invention, unless otherwise specified, "multiple" means two or more.

[0040] In the description of the present invention, it should be noted that, unless otherwise expressly specified or limited, the terms "mounted," "connected," and "connected" should be understood in a broad sense. For example, they may refer to fixed, detachable, or integral connections; mechanical or electrical connections; direct or indirect connections through an intermediate medium; and internal communication between two components. Those skilled in the art will understand the specific meanings of the above terms in the present invention based on the specific circumstances.

[0041] The devices in the prior art that heat the cladding 1 directly or through heating rods ignore the mechanical effects on the cladding 1 and the changes in heat transfer characteristics during the cracking of the pellets 2, which may cause the experimental results to deviate from the actual situation.

[0042] At high temperatures, nuclear fuel pellets 2 can crack due to factors such as thermal stress and radiation swelling, forming fragmented blocks or debris. Cracked pellets 2 lose their overall rigidity, and internal stresses are redistributed across the fractured surfaces, potentially exerting localized concentrated loads (such as compression and friction) on the inner wall of the cladding 1. Furthermore, while intact pellets 2 conduct heat through the contact surface with the cladding 1, cracked pellets 2 break into multiple pieces, creating numerous internal voids. This reduces heat transfer efficiency and may shift to a heat transfer mode dominated by radiation and convection of gases (such as fission gases).

[0043] In view of the shortcomings of the existing technology, this embodiment provides a nuclear fuel behavior simulation experimental device for a loss of coolant accident, which can be referred to as follows:

[0044] As attached Figure 1 As shown, a nuclear fuel behavior simulation experiment device for a loss of coolant accident includes a cladding 1, a pellet 2, an insulating member 3, a pressurized cooling pipe 4, and a cooling assembly 5; the pellet 2 is arranged inside the cladding 1, and its top and bottom walls are connected to an external power supply through wires; the insulating member 3 is arranged on the outer surface of the pellet 2 to isolate the pellet 2 from the cladding 1; the pressurized cooling pipe 4 is sealed and sleeved outside the cladding 1; and the cooling assembly 5 is connected to the pressurized cooling pipe 4 to form a coolant circulation path.

[0045] The pellet 2 is a block of ceramic material with the physical properties of nuclear fuel, specifically uranium oxide or cerium oxide, whose electrical resistance allows for self-heating when powered. The insulating element 3 is an insulating layer with both electrical insulation and high-temperature resistance, specifically an alumina ceramic coating. This ensures electrical isolation between the pellet 2 and the cladding 1 while allowing for thermal expansion and displacement. The pressurized cooling pipe 4 is a sealed container capable of withstanding internal pressure, connected by flanges to form a closed cooling space. The cooling assembly 5 is a circulation system capable of regulating the flow of the coolant, specifically a combination of a multi-stage pump and a heat exchanger 53, with valve control simulating coolant phase changes.

[0046] Specifically, the cladding 1 serves as a constraint structure for the fuel assembly, and the stacked fuel cores 2 inside generate Joule heat through electrical wires, accurately reproducing the heat generation mechanism during core operation. The ceramic insulation layer covering the surface of the core 2 can prevent current leakage. The sealed pressurized cooling pipe 4 wraps around the cladding 1 to form a pressure boundary, and the coolant is driven to flow in the pressurized cooling pipe 4 through the circulating pump 55. When simulating the coolant loss condition, the change process of the external cooling conditions of the cladding 1 can be accurately controlled by adjusting the coolant flow and pressure parameters. The crack propagation behavior generated by the core 2 during continuous heating is transmitted in real time through the spring 14 mechanism set in the cladding 1. The structural response of the fuel rod under accident conditions is fully reproduced.

[0047] Compared with the existing technology, the traditional experimental device uses an external heating rod to simulate the heat generation of the core block 2, which cannot produce a real radial temperature gradient, resulting in distorted calculation of the thermal stress distribution of the cladding 1. This solution uses the resistance of the core block 2 to generate heat, which not only reproduces the axial heat conduction characteristics of the fuel rod, but also more accurately simulates the dynamic process of the change of thermal resistance after the core block 2 is broken. The geometric shape of the prefabricated fragments in the existing technology is fixed, but this solution allows the core block 2 to crack naturally during its own heating process, producing irregular fragment shapes that meet actual working conditions. In terms of cooling system design, traditional single-loop systems find it difficult to simulate the transient process of coolant flash and re-flooding. This solution achieves precise control of coolant phase changes through a combination of multi-branch pipes and rupture devices.

[0048] Through the above technical solution, the present application can realistically simulate the cracking behavior of fuel pellets 2 under high temperature and high pressure conditions, and accurately capture the interaction mechanism between the pellets 2 and the pellet 2 fragments and the cladding 1. By directly using fuel materials for experiments, the physical property differences between the simulant and the prototype are eliminated. The controllable coolant circulation system can reproduce the entire process of coolant loss, spraying and re-flooding in LOCA accidents, providing reliable experimental data for studying fuel rod failure modes. The synergistic effect of the autonomous heat generation mode of the pellets 2 and the external cooling conditions makes the thermomechanical response analysis of the cladding 1 closer to actual working conditions.

[0049] As attached Figure 1 As shown, the present application further proposes a coolant loss accident nuclear fuel behavior simulation experimental device including two fixed flanges 6, which are respectively arranged at both ends of the pressurized cooling pipe 4, and the pressurized cooling pipe 4 is extended from both ends of the cladding 1. The fixed flanges 6 are respectively connected to the pressurized cooling pipe 4 and the cladding 1 to seal and fix the pressurized cooling pipe 4 and the cladding 1.

[0050] The fixed flange 6 is an annular connector with a sealing structure that forms a rigid seal between the pressurized cooling duct 4 and the cladding 1. The extension of the pressurized cooling duct 4 from both ends of the cladding 1 means that the length of the cladding 1 exceeds the coverage of the duct. This can be achieved by adjusting the difference between the axial dimension of the cladding 1 and the inner length of the pressurized cooling duct 4, providing physical installation space for the external interface.

[0051] Specifically, the fixing flange 6 is bolted between the end of the pressurized cooling duct 4 and the outer wall of the cladding 1, forming a double seal. With the extended section of the cladding 1 exposed to the outside of the duct, the exposed area can be directly provided with the inflation port 11 and the deflation port 12, eliminating the need for openings in the pressurized cooling duct 4 itself.

[0052] Through the above-mentioned technical solution, the present application achieves a reliable sealed connection between the pressurized cooling pipe 4 and the cladding 1, while also providing an independent operating interface for monitoring and regulating the internal pressure of the cladding 1. The extended section structure allows components such as the inflation port 11, the deflation port 12, and the first pressure gauge 13 to be directly installed on the cladding 1 body, avoiding the potential for leakage caused by the opening of the pressurized cooling pipe 4 and ensuring precise control of pressure parameters during experiments.

[0053] As attached Figure 1 As shown, the present application further proposes that an air filling port 11 and an air release port 12 are provided on the cladding 1 located outside the pressurized cooling pipe 4 , and a first pressure gauge 13 is provided inside the cladding 1 .

[0054] The inflation port 11 is a gas injection port provided on the wall of the cladding 1, which can be implemented as a tubular structure with a sealing valve, and is used to inject gas into the interior of the cladding 1 to create a pressure environment. The deflation port 12 is a gas discharge port provided on the wall of the cladding 1, which can be implemented as an adjustable throttle valve structure, and is used to control the release rate of gas from the interior of the cladding 1. The first pressure gauge 13 is a pressure sensor installed in the cavity of the cladding 1, which can be implemented as a digital pressure transmitter, and is used to monitor the pressure value changes in the enclosed space in real time.

[0055] Specifically, during the experiment, the inflator 11 injects an inert gas (such as helium, which has excellent thermal conductivity) into the cladding 1 from an external gas source to increase the pressure. A first pressure gauge 13 continuously collects pressure data within the cladding 1. To simulate a sudden pressure drop caused by a coolant release, the vent 12 adjusts the gas discharge rate based on the feedback from the first pressure gauge 13, creating a closed-loop pressure control system. This coordinated operation of the inflator 11 and vent 12, combined with the real-time monitoring function of the first pressure gauge 13, allows for dynamic adjustment of the internal pressure of the cladding 1, accurately simulating the transient pressure loads experienced by the fuel rods during an accident.

[0056] Compared to existing technologies, traditional experimental devices typically use a sealed chamber with a fixed volume, which can only maintain a static pressure environment and cannot reproduce the pressure fluctuations during accident conditions. This solution overcomes the technical limitations of static pressure simulation by adding actively adjustable inflation port 11 and deflation port 12, combined with data feedback from a first pressure gauge 13. This allows the internal pressure of the cladding 1 to be dynamically controlled according to experimental requirements.

[0057] Through the above technical solution, the present application realizes the real-time monitoring and active regulation of the internal pressure of the cladding 1, effectively solving the defect that traditional devices cannot simulate the impact of transient pressure changes on the behavior of fuel rods, and provides precise experimental conditions for studying the cracking, deformation and interaction of fuel rods under pressure fluctuations in coolant loss accidents.

[0058] As attached Figure 1 As shown, the present application further proposes that the coolant loss accident nuclear fuel behavior simulation experimental device also includes two springs 14, the two springs 14 are respectively arranged on the inner top wall and the inner bottom wall of the cladding 1, and the two springs 14 are respectively in contact with the top wall and the bottom wall of the core block 2.

[0059] The spring 14 is an elastic support element whose elastic modulus is matched to the thermal expansion coefficient of the core block 2. The spring 14 absorbs the axial displacement of the core block 2 caused by heating through compression deformation, while also providing a buffer space in the event of core block 2 fragmentation. Abutment refers to the spring 14 maintaining contact with the end face of the core block 2 but not completely securing it. This can be achieved by presetting a preload, allowing the core block 2 to maintain electrical connection with the wires and transmit thermal expansion pressure axially during the experiment. This abutment method both constrains the displacement of the core block 2 and allows it to undergo limited deformation when heated.

[0060] Specifically, when the core block 2 expands axially, the springs 14 at both ends are compressed, and the disordered displacement of the core block 2 is limited by bidirectional elastic constraints; under the action of the double springs 14, the core block 2 can be located in the pressurized cooling pipe 4, so that when the core block 2 expands, the expansion force acts on the cladding 1 located in the pressurized cooling pipe 4, thereby accurately simulating the deformation and rupture of the cladding 1 under real conditions.

[0061] Compared to existing technologies, existing experimental setups often employ either rigid fixation or completely free placement of pellets 2. Rigid fixation hinders thermal expansion of the pellets 2, distorting the stress distribution in the cladding 1; free placement fails to reflect the dynamic contact behavior of the pellets 2 with the cladding 1 after fragmentation. This solution utilizes bidirectional elastic support, allowing the pellets 2 to expand naturally as they heat while preventing them from moving out of their pre-set position. Furthermore, the deformation of the spring 14 indirectly reflects the mechanical forces acting between the pellets 2 and the cladding 1, more realistically reproducing the interaction between the fuel rod's internal components under accident conditions.

[0062] Through the above-mentioned technical solution, the present application effectively solves the problem of displacement deviation caused by the unfixed pellet 2. By simulating the change in contact state between the pellet 2 and the cladding 1 after fragmentation through elastic constraints, it avoids direct impact damage to the end of the cladding 1 caused by the thermal expansion of the pellet 2. The elastic deformation characteristics of the spring 14 make the monitoring data of the pellet 2 behavior more closely resemble the physical response of real nuclear fuel rods under accident conditions, providing reliable experimental conditions for studying the mechanism linking the cracking of the pellet 2 and the deformation of the cladding 1.

[0063] As attached Figure 1As shown, the present application further proposes that the cooling assembly 5 includes a first pipe 51, a first break 52, a heat exchanger 53, a second pipe 54, a circulation pump 55 and a second break 56. The first pipe 51 is arranged on the side wall of the pressurized cooling pipe 4, the first break 52 is arranged on the first pipe 51, and the heat exchanger 53 is arranged at one end of the first pipe 51 away from the pressurized cooling pipe 4. The second pipe 54 is arranged on the heat exchanger 53, one end of which is connected to the outlet of the heat exchanger 53, and the other end is arranged at the bottom of the pressurized cooling pipe 4 and connected to the interior. The circulation pump 55 is arranged on the second pipe 54, and the second break 56 is arranged on the second pipe 54. By opening the first break 52 and / or the second break 56, the coolant spraying, flash evaporation, liquid level drop and evaporation process are simulated.

[0064] in, Figure 1 The arrow in the figure indicates the flow direction of the coolant; the first pipe 51 refers to the fluid transmission channel connecting the side wall of the pressurized cooling pipe 4 and the heat exchanger 53, which is used to establish a coolant circulation path. The first breach 52 refers to a controllable opening device provided on the first pipe 51, which is used to actively trigger the rapid leakage of the coolant. The heat exchanger 53 refers to a device with a heat exchange function, which is used to adjust the phase change of the coolant. The second pipe 54 refers to a reflux channel connecting the outlet of the heat exchanger 53 and the bottom of the pressurized cooling pipe 4, which is used to form a closed-loop circulation system. The circulation pump 55 refers to a mechanical device that drives the flow of the coolant, which can be specifically implemented by a centrifugal pump, and is used to maintain the circulation power of the coolant in the system. The second breach 56 refers to a pressure relief device provided on the second pipe 54, which is used to simulate the loss of coolant at different positions.

[0065] Specifically, when the coolant circulates in the pressurized cooling pipe 4, the first pipe 51 is connected through the side wall position to form an initial flow path. When the first breach 52 is opened, the coolant is quickly sprayed through the breach, simulating the sudden leakage process at the initial stage of the accident. The heat exchanger 53 performs heat exchange on the coolant after spraying, causing the liquid coolant to undergo a flash phase change. The second pipe 54 transports the coolant that has undergone heat exchange treatment to the bottom of the pressurized cooling pipe 4, and the circulating pump 55 controls the rate of drop of the liquid level in the channel by adjusting the flow rate. The opening of the second breach 56 can change the system pressure distribution, and cooperate with the combined operation of the first breach 52 to achieve different working conditions such as single-phase loss and two-phase mixed loss. By opening different breaches in stages, the evaporation process of the coolant in the channel can be gradually reproduced.

[0066] Through the above-mentioned technical solution, the present application can fully simulate the dynamic evolution of coolant during a coolant loss accident, from spraying, flash evaporation, liquid level drop, and finally evaporation. The combined operation of the first breach 52 and the second breach 56 can simulate leakage scenarios of varying locations and degrees. The synergistic effect of the heat exchanger 53 and the circulating pump 55 can precisely control the coolant phase and circulation state. The bottom connection design of the second pipe 54 links the liquid level changes with the coolant circulation, effectively reproducing the temporal and spatial distribution characteristics of the gradual loss of coolant in an actual accident.

[0067] As attached Figure 1 As shown, the present application further proposes that the cooling assembly 5 also includes a re-flooding water tank 57 and a water storage tank 58. The re-flooding water tank 57 is arranged outside the pressurized cooling pipe 4 and is connected to the inside of the pressurized cooling pipe 4 through a pipe. The water storage tank 58 is arranged outside the pressurized cooling pipe 4 and is connected to the second pipe 54 through a connecting pipe. The connection point of the connecting pipe and the second pipe 54 is located on the side of the circulating pump 55 away from the pressurized cooling pipe 4. The water storage tank 58 cooperates with the re-flooding water tank 57 to simulate the refrigerant re-flooding process.

[0068] The reflooding tank 57 is a container for storing emergency coolant, with a valve controlling fluid exchange with the pressurized cooling pipe 4. The water storage tank 58 is a container for storing the main circulation coolant, providing coolant replenishment to the second pipe 54 via a connecting pipe. The connection point between the connecting pipe and the second pipe 54 is located upstream of the circulation pump 55, specifically through a T-joint. This location utilizes the negative pressure generated by the circulation pump 55 to improve the efficiency of the water storage tank 58 in replenishing the main circulation system.

[0069] Specifically, when the experiment needs to simulate the coolant re-flooding condition, the connecting valve between the re-flooding water tank 57 and the pressurized cooling pipe 4 is opened, and the coolant is injected into the pipe under the action of gravity. The water storage tank 58 continuously replenishes the coolant to the second pipe 54 through the connecting pipe. Since the connection point is located on the inlet side of the circulating pump 55, the negative pressure generated when the circulating pump 55 is running accelerates the delivery speed of the coolant in the water storage tank 58. The coordinated operation of the two water tanks enables the main circulation system to simultaneously receive continuous supply from the water storage tank 58 and concentrated injection from the re-flooding water tank 57. By adjusting the liquid supply ratio of the two water tanks, re-flooding scenarios under different leakage rates can be simulated. The operating parameters of the circulating pump 55 form a dynamic balance with the water tank supply pressure, thereby realistically reproducing the fluid dynamic characteristics of the coolant re-flooding stage in the accident condition.

[0070] Through the above-mentioned technical solution, this application can precisely control the coolant replenishment rate and fluid mixing state during the reflooding process, effectively simulating the complex operating conditions of multiple coolant supply sources in a nuclear power plant accident. The independent liquid supply characteristics of the two water tanks enable the experimental setup to replicate the effects of different emergency water injection strategies on the reflooding of fuel rods, providing a realistic and reliable experimental environment for studying the thermodynamic behavior of fuel elements under dynamic reflooding conditions.

[0071] As attached Figure 1 As shown, the present application further proposes that a synchrotron radiation X-ray in-situ imaging device 7 is provided on one side of the pressurized cooling pipe 4 for detecting cracking of the pellets 2 and deformation of the cladding 1 .

[0072] The synchrotron radiation X-ray in-situ imaging device 7 uses high-energy X-rays to penetrate non-transparent materials and form real-time images. This device utilizes existing technology and adjusts X-ray energy parameters to accommodate different material thicknesses. By penetrating the pressurized cooling pipe 4 and the cladding 1, the device directly captures the crack growth process within the pellet 2.

[0073] Specifically, when simulating a loss-of-coolant accident, a synchrotron radiation X-ray in-situ imaging device 7 was positioned along the axis of the pressurized cooling pipe 4, with the X-ray beam penetrating the pipe wall, the cladding 1, and the pellet 2. When thermal stress generated by the temperature gradient in the pellet 2 causes crack initiation, changes in material density at the crack site cause differences in X-ray absorption, forming a real-time image with light and dark contrast on the detector. The local expansion or contraction deformation of the cladding 1 is quantified by the displacement of the cladding 1 contour in successive images. Three-dimensional deformation data of the cladding 1 is obtained after the image sequence is processed using a three-dimensional reconstruction algorithm. This process continues throughout the experiment, fully recording the natural formation of the pellet 2's fragmentation morphology.

[0074] Through the above technical solution, the present application solves the technical bottleneck of being unable to observe the dynamic changes of the internal structure of the fuel rod in real time, provides direct experimental data support for analyzing the fragmentation mechanism of the pellet 2 and the mechanical response of the cladding 1 in the coolant loss accident, and avoids the problem of traditional destructive testing resulting in the non-reusability of experimental samples.

[0075] As attached Figure 1 As shown, the present application further proposes a coolant loss accident nuclear fuel behavior simulation experimental device including two electrodes 21, the two electrodes 21 are respectively arranged on the top wall and the bottom wall of the core block 2, the electrodes 21 pass through the cladding 1 and extend to the outside of the cladding 1, and the electrodes 21 are sealed with the cladding 1.

[0076] The electrode 21 passing through the cladding 1 means that the electrode 21 passes through the upper and lower walls of the cladding 1 in a penetrating manner, so as to facilitate the electrode 21 to extend into the interior of the cladding 1 and abut against the top wall or bottom wall of the core block 2, thereby ensuring that the electrode 21 maintains the structural integrity of the cladding 1 while conducting electricity.

[0077] If the wire is directly in contact with the core block 2, a wire with a larger diameter is required to provide conductive efficiency; the arrangement of the electrode 21 of the present application can prevent the wire from directly extending into the interior of the shell 1, and can also avoid the use of a wire with a larger diameter.

[0078] Specifically, the electrode 21 is led out from the top wall or bottom wall of the core block 2 and passes through the corresponding position of the cladding 1, extending to the outside of the cladding 1 and forming an electrical connection loop with the external power supply. The contact surface between the electrode 21 and the cladding 1 is sealed by a sealing structure.

[0079] As attached Figure 1 As shown, the present application further proposes that there are multiple core blocks 2 , and the multiple core blocks 2 are stacked and located inside the pressurized cooling pipe 4 .

[0080] The pellets 2 are independent units simulating the nuclear fuel core. Specifically, they can be made of ceramic or metal materials in cylindrical or annular structures. Stacked together to form an axially stacked fuel rod model, they can undergo relative displacement and interaction under thermal expansion or mechanical loads. The stacked arrangement of the pellets 2 along the axial direction is unconstrained, allowing them to undergo axial displacement and radial expansion under thermal stress, simulating the expansion and compression behavior of the pellets 2 in real fuel rods caused by temperature gradients.

[0081] Specifically, when multiple pellets 2 form a stacked structure inside the pressurized cooling pipe 4, the contact surfaces between the pellets 2 generate friction and contact stress under thermal load. When the coolant is lost and the temperature rises sharply, the pellets 2 undergo axial elongation and radial expansion due to thermal expansion, and the contact stress between adjacent pellets 2 increases, which may cause local cracking or dislocation. The stacked structure constrains the pellets 2 by adjacent pellets 2 during the expansion process, simulating the interaction force between the pellets 2 in real fuel rods. Under the action of continuous thermal stress, the stack of pellets 2 may undergo overall axial displacement or partial scattering, thereby dynamically reflecting the mechanical effect of the repositioning of the pellets 2 on the cladding 1 during the accident.

[0082] Through the above technical solution, the present application can accurately reproduce the dynamic mechanical effects between multiple core blocks 2 in a coolant loss accident, including axial extrusion caused by thermal expansion, stress concentration in the cladding 1 caused by radial expansion, and impact of scattered cracked core blocks 2 on the cladding 1. This solves the problem in the prior art that the experimental results deviate from the actual working conditions due to the single simulation method of the core blocks 2.

[0083] As attached Figure 1 As shown, the present application further proposes a technical solution of arranging an acoustic level meter 8 and a second pressure gauge 9 inside the pressurized cooling pipe 4 .

[0084] The acoustic level gauge 8 measures liquid level by emitting ultrasonic waves and receiving reflected signals. Specifically, this can be implemented using a pulsed ultrasonic sensor, which determines the liquid level by calculating the difference in acoustic wave propagation time. This device maintains measurement accuracy in high-temperature and high-pressure environments, resolving the response delay issue associated with traditional mechanical level gauges due to contact with the medium. The second pressure gauge 9 is an instrument used to detect fluid pressure within a confined space.

[0085] Specifically, an acoustic level gauge 8 is installed at the top of the pressurized cooling pipe 4 and transmits an ultrasonic signal into the pipe. This signal is reflected by the liquid surface and captured by a receiver, which uses the time difference to calculate the real-time liquid level. A second pressure gauge 9 is embedded in the sidewall of the pipe to monitor the pressure within the pressurized cooling pipe 4.

[0086] Through the above technical solution, the present application realizes real-time synchronous monitoring of the coolant liquid level and system pressure, and can accurately capture the dynamic parameter changes of the coolant spraying, flashing and evaporation stages in the LOCA accident simulation. It solves the problem of inaccurate analysis of the correlation between the cracking behavior of the pellet 2 and the deformation of the cladding 1 due to measurement lag in the existing device, and provides a reliable data basis for studying the failure mechanism of the fuel element during the coolant loss process.

[0087] Other embodiments of the present invention will readily occur to those skilled in the art after considering the specification and practicing the embodiments disclosed herein. The present invention is intended to cover any variations, uses, or adaptations of the present invention that follow the general principles of the invention and include common knowledge or customary techniques in the art not disclosed herein. The description and examples are to be considered as exemplary only, with the true scope and spirit of the invention being indicated by the claims.

Claims

1. A nuclear fuel behavior simulation experimental device for a loss of coolant accident, characterized in that: include: cladding; A core block is disposed inside the cladding; the top wall and the bottom wall of the core block are respectively connected to an external power source through wires; an insulating member, disposed on the outer surface of the core block to isolate the core block from the cladding; A pressurized cooling pipe is sleeved on the outside of the cladding and sealed with the cladding; a cooling assembly connected to the pressurized cooling pipe to form a coolant circulation flow path; The cooling assembly comprises: a first pipe, disposed on a side wall of the pressurized cooling pipe; A first breach is provided on the first pipe; a heat exchanger, disposed at an end of the first pipe away from the pressurized cooling pipe; a second pipe disposed on the heat exchanger; one end of the second pipe being connected to the outlet of the heat exchanger, and the other end being disposed at the bottom of the pressurized cooling pipe and communicating with the interior of the pressurized cooling pipe; a circulation pump, disposed on the second pipeline; A second rupture is provided on the second pipe; by opening the first rupture and / or the second rupture, the process of coolant spraying, flash evaporation, liquid level drop and evaporation can be simulated; a reflooding water tank, disposed outside the pressurized cooling pipe; the reflooding water tank is connected to the interior of the pressurized cooling pipe through a pipe; a water storage tank disposed outside the pressurized cooling pipe; the water storage tank is connected to the second pipe via a connecting pipe, the connection point between the connecting pipe and the second pipe being located on a side of the circulating pump away from the pressurized cooling pipe and upstream of the circulating pump; the water storage tank cooperates with the reflooding water tank to simulate the refrigerant reflooding process; An acoustic level gauge and a second pressure gauge are provided inside the pressurized cooling pipe.

2. The apparatus for simulating nuclear fuel behavior in a loss of coolant accident according to claim 1, characterized in that: It also includes: Two fixing flanges are respectively arranged at both ends of the pressurized cooling pipe; the pressurized cooling pipe is extended from both ends of the cladding, and the fixing flanges are respectively connected to the fixing flanges and the cladding.

3. The apparatus for simulating nuclear fuel behavior during a loss of coolant accident according to claim 2, characterized in that: An air filling port and an air releasing port are provided on the cladding outside the pressurized cooling pipe, and a first pressure gauge is provided inside the cladding.

4. The apparatus for simulating nuclear fuel behavior in a loss of coolant accident according to claim 2, characterized in that: It also includes: Two springs are respectively arranged on the inner top wall and the inner bottom wall of the cladding; the two springs are respectively in contact with the top wall and the bottom wall of the core block.

5. The apparatus for simulating nuclear fuel behavior in a loss of coolant accident according to claim 1, characterized in that: It also includes: A synchrotron radiation X-ray in-situ imaging device is provided on one side of the pressurized cooling pipe and is used to detect cracking of the core block and deformation of the cladding.

6. The apparatus for simulating nuclear fuel behavior in a loss of coolant accident according to claim 1, characterized in that: It also includes: Two electrodes are respectively arranged on the top wall and the bottom wall of the core block; the electrodes penetrate the cladding and extend to the outside of the cladding, and the electrodes and the cladding are sealed.

7. The apparatus for simulating nuclear fuel behavior in a loss of coolant accident according to claim 1, characterized in that: There are multiple core blocks, and the multiple core blocks are stacked and located inside the pressurized cooling pipe.

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