A method for modeling the dynamic power distribution of a nuclear reactor's thermal-hydraulics in a grid-based manner

Abstract

The present invention discloses a method for modeling the dynamic power distribution of a nuclear reactor's thermal-hydraulics in a grid pattern, including: dividing an aqueous solution-type nuclear reactor vessel into multiple concentric ring-shaped parallel channels; calculating the radial power distribution ratio within each ring-shaped channel; determining the initial liquid level height of the solution under steady-state conditions; performing axial grid division on each ring-shaped channel; establishing a corresponding simulation scheme for thermal components according to the determined nodalized grid model; determining the initial power distribution within each control volume of the solution; in the transient condition simulation, calculating the value of the solution liquid level in real time, judging the axial grid position where the liquid level is located, and calculating the power distribution of the solution in real time according to the liquid level height. This method achieves the purpose of accurately simulating the power distribution of an aqueous solution-type nuclear reactor using traditional thermal-hydraulics system programs, thus saving the cost and time for developing a system-level thermal-hydraulics program applicable to transient analysis of aqueous solution-type nuclear reactors, and achieving the effect of saving time and effort.

Description

Technical Field

[0001] The present invention relates to the technical field of nuclear reactor thermal-hydraulic analysis, and particularly relates to a method for modeling the dynamic power distribution of a nuclear reactor thermal-hydraulic grid. Background Art

[0002] Safety is the lifeline of nuclear energy development. The accident transient safety analysis of a nuclear reactor often relies on a thermal-hydraulic system program. First, a discrete modeling scheme for the target nuclear reactor is established, and then a discrete model is established and used for steady-state and transient safety analysis calculations. Due to the huge size and complexity of the nuclear reactor system, when using a system-level thermal-hydraulic program for discrete modeling, the grid division is relatively coarse, so it is difficult to simulate local fine phenomena, especially the refined three-dimensional core power distribution.

[0003] For an aqueous solution type nuclear reactor, it is a low-power nuclear reactor using uranyl nitrate aqueous solution as nuclear fuel, mainly used for generating medical isotopes. The nuclear fuel is stored in the reactor vessel in the form of a solution, and the fuel solution can naturally circulate and flow in the reactor vessel. The uranium fuel is evenly distributed in water and evenly mixed with the moderator, so that when the fuel temperature rises, it has a large negative temperature coefficient; at the same time, when the power increases, a negative reactivity will be introduced due to the bubble reactivity feedback. Due to the characteristics of the aqueous solution type nuclear reactor, it is extremely difficult to simulate the power distribution inside it in a thermal-hydraulic system program. Coupled with different types of reactivity feedback, it further increases the difficulty of accurately simulating the power distribution.

[0004] In addition, in the accident conditions of an aqueous solution type nuclear reactor, in order to protect it safely, it is necessary to carry out liquid extraction and shutdown protection for the solution reactor. During this process, the liquid level of the solution drops, and the thermal component used to simulate the heat source in the thermal-hydraulic system program cannot change dynamically, resulting in difficulty in using the system program to simulate the power distribution in the transient conditions of an aqueous solution type nuclear reactor. Therefore, it is necessary to develop a thermal-hydraulic grid dynamic power distribution modeling technology to be used for simulating the power distribution inside an aqueous solution type nuclear reactor. Summary of the Invention

[0005] The purpose of the present invention is to provide a method for modeling the dynamic power distribution of a nuclear reactor thermal-hydraulic grid to solve the problem that the power distribution of an aqueous solution type nuclear reactor cannot be accurately simulated using a traditional thermal-hydraulic system program. This method reasonably divides the nodal grid of the aqueous solution type nuclear reactor through sensitivity analysis, equivalently simulates the self-heating effect of the solution in the aqueous solution type nuclear reactor, and then dynamically calculates the power distribution in the solution based on the transient liquid level, achieving the purpose of accurately simulating the power distribution of the aqueous solution type nuclear reactor using a traditional thermal-hydraulic system program.

[0006] The present invention is achieved by the following technical solutions:

[0007] The present invention provides a method for modeling the dynamic power distribution of a nuclear reactor's thermal-hydraulics in a grid form, including:

[0008] Step 1: Divide the aqueous solution type nuclear reactor vessel into multiple concentric annular parallel channels;

[0009] Step 2: Calculate the radial power distribution ratio in each annular channel;

[0010] Step 3: Determine the initial liquid level height of the solution under steady-state conditions;

[0011] Step 4: Perform axial grid division on each annular channel;

[0012] Step 5: Establish a corresponding thermal component simulation scheme according to the determined nodalized grid model;

[0013] Step 6: Determine the initial power distribution in each control volume of the solution;

[0014] Step 7: In the transient condition simulation, calculate the value of the solution liquid level in real time, judge the axial grid position where the liquid level is located, and calculate the power distribution of the solution in real time according to the liquid level height.

[0015] In the above solution, by establishing a nodalized grid model of the container where the reactor solution is located, equivalently simulating the self-heating effect of the solution, determining the initial power distribution in each grid in the solution according to the divided grid scheme, calculating the value of the solution liquid level in real time in the transient condition simulation, judging the axial grid position where the liquid level is located, and calculating the power distribution of the solution in real time according to the liquid level height, the purpose of accurately simulating the power distribution of an aqueous solution type nuclear reactor using a traditional thermal-hydraulic system program is achieved.

[0016] In some embodiments, the number of nodes in the radial direction of the reactor vessel in Step 1 is determined by sensitivity analysis.

[0017] In some embodiments, Step 2 is to calculate the power distribution in the aqueous solution type nuclear reactor vessel according to three-dimensional neutron kinetics, and calculate the radial power distribution ratio in each annular channel according to the radial node division and the solution loading volume ratio.

[0018] In some embodiments, the initial liquid level height of the solution under steady-state conditions in Step 3 is calculated and determined according to the designed loading of the aqueous solution type nuclear reactor solution and the geometric structure of the reactor vessel.

[0019] In some embodiments, when performing axial grid division in Step 4, each annular channel is axially grid-divided at the same height, and the liquid level is located on the boundary line between two axial grids.

[0020] In some embodiments, when establishing the thermal component simulation scheme in step five, a thermal component with high thermal conductivity and low heat capacity is used to simulate the heat source of the solution, and the heat transfer coefficient between the thermal component and the solution is set to be greater than 10 8 Simulate the self-heating effect of the solution.

[0021] In some embodiments, in step six, the power distribution in each control volume of the solution is calculated according to the radial power distribution determined in step two and the axial grid divided in step four.

[0022] In some embodiments, in step seven, first calculate the liquid level height in each control volume by multiplying the liquid phase fraction in each control volume by the axial height of the control volume, and then calculate the total solution liquid level by weighted averaging the liquid level heights in all control volumes.

[0023] In some embodiments, in step seven, the volume distribution of the solution in each grid is calculated in real time according to the liquid level height, and then the power distribution of the solution is calculated in real time according to the volume distribution.

[0024] In some embodiments, in step seven, a judgment logic is used to set the logic of zeroing the power of the current control volume when the liquid level height in the axial control volume where the liquid level is located is lower than 1% of the current control volume height.

[0025] Compared with the prior art, the present invention has the following advantages and beneficial effects:

[0026] 1. The present invention proposes a thermal-hydraulic grid-based dynamic power distribution modeling technology for nuclear reactors. By sensitivity analysis, the nodalized grid of an aqueous solution nuclear reactor is reasonably divided, and a thermal component with high thermal conductivity and low heat capacity is used to equivalently simulate the self-heating effect of the solution in the aqueous solution nuclear reactor. Subsequently, based on the transient liquid level, the power distribution in the solution is dynamically calculated, achieving the purpose of accurately simulating the power distribution of the aqueous solution nuclear reactor using a traditional thermal-hydraulic system program;

[0027] 2. The thermal-hydraulic grid-based dynamic power distribution modeling technology proposed by the present invention can be used for the safety analysis of transient accident conditions of an aqueous solution nuclear reactor with liquid fuel. When there is a lack of a relevant transient analysis system-level thermal-hydraulic program, the traditional thermal-hydraulic system program is used to achieve an equivalent high-precision simulation effect. This modeling technology can greatly promote the accident safety analysis process of the aqueous solution nuclear reactor based on high-precision alternative simulation, saving the cost and time of developing a system-level thermal-hydraulic program suitable for transient analysis of the aqueous solution nuclear reactor, and there is no need to conduct additional tests for program verification, achieving the effect of saving time and effort. Description of the Drawings

[0028] To more clearly illustrate the technical solutions of the exemplary embodiments of the present invention, the accompanying drawings required for the embodiments will be briefly introduced below. It should be understood that the following drawings only show some embodiments of the present invention and should not be regarded as limiting the scope. For those of ordinary skill in the art, without creative efforts, other related drawings can also be obtained based on these drawings. In the drawings:

[0029] Figure 1 It is a flowchart of the method for modeling the thermal-hydraulic grid dynamic power distribution of the nuclear reactor in the present invention. Specific embodiments

[0030] To make the objectives, technical solutions and advantages of the present invention clearer and more understandable, the present invention will be further described in detail below in conjunction with the embodiments and the accompanying drawings. The illustrative embodiments and descriptions thereof of the present invention are only used to explain the present invention and do not limit the present invention.

[0031] Unless otherwise defined, all technical and scientific terms used herein have the same meaning as commonly understood by those of ordinary skill in the technical field to which this application belongs; the terms used herein are only for the purpose of describing specific embodiments and are not intended to limit this application; the terms "including" and "having" and any variations thereof in the specification and claims of this application and the above drawings are intended to cover non-exclusive inclusion.

[0032] In the description of the embodiments of this application, technical terms such as "first" and "second" are only used to distinguish different objects and cannot be understood as indicating or implying relative importance or implicitly indicating the quantity, specific order or primary-secondary relationship of the indicated technical features.

[0033] Referring to "embodiments" herein means that the specific features, structures or characteristics described in connection with the embodiments can be included in at least one embodiment of this application. The phrase appears in various positions in the specification does not necessarily refer to the same embodiment, nor is it an independent or alternative embodiment mutually exclusive with other embodiments. Those skilled in the art explicitly and implicitly understand that the embodiments described herein can be combined with other embodiments.

[0034] In the description of the embodiments of this application, the term "and / or" is only a description of the association relationship of associated objects, indicating that three relationships can exist. For example, A and / or B can mean: there is A, there is both A and B, and there is B. In addition, the character " / " herein generally represents an "or" relationship between the associated objects before and after.

[0035] In the embodiments of the present application, the same reference numerals denote the same components, and for the sake of brevity, in different embodiments, the detailed description of the same components is omitted. It should be understood that the thickness, length, width, and other dimensions of various components in the embodiments of the present application shown in the drawings, as well as the overall thickness, length, width, etc. of the integrated device, are only for illustrative purposes and should not constitute any limitation to the present application.

[0036] In the description of the embodiments of the present application, the term "plurality" means two or more (including two). Similarly, "multiple groups" means two or more groups (including two groups), and "multiple pieces" means two or more pieces (including two pieces), unless otherwise specifically defined.

[0037] In the description of the embodiments of the present application, the technical terms "center", "longitudinal", "lateral", "length", "width", "thickness", "upper", "lower", "front", "rear", "left", "right", "vertical", "horizontal", "top", "bottom", "inner", "outer", "clockwise", "counterclockwise", "axial", "radial", "circumferential", etc. indicate the orientation or positional relationship based on the orientation or positional relationship shown in the drawings. It is only for the convenience of describing the embodiments of the present application 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 should not be construed as a limitation to the embodiments of the present application.

[0038] In the description of the embodiments of the present application, unless otherwise clearly specified and limited, technical terms such as "installation", "connection", "connection", "fixation", etc. should be understood in a broad sense. For example, it can be a fixed connection, a detachable connection, or integrated; it can also be a mechanical connection or an electrical connection; it can be directly connected or indirectly connected through an intermediate medium, and it can be the communication inside two elements or the interaction relationship between two elements. For those of ordinary skill in the art, the specific meanings of the above terms in the embodiments of the present application can be understood according to specific situations.

[0039] At present, the accident transient safety analysis of nuclear reactors often relies on the thermal-hydraulic system program, but this system program usually only performs steady-state and transient safety analysis calculations for solid-fuel nuclear reactors. For aqueous solution nuclear reactors, nuclear fuel is stored in the reactor vessel in the form of a solution, and the fuel solution can naturally circulate and flow in the reactor vessel. Due to the characteristics of aqueous solution nuclear reactors, it is extremely difficult to simulate the power distribution inside them in the thermal-hydraulic system program. Coupled with different types of reactivity feedback, it further increases the difficulty of accurately simulating the power distribution.

[0040] In the accident conditions of an aqueous solution type nuclear reactor, in order to protect it safely, it is necessary to perform liquid extraction shutdown protection on the solution reactor. During this process, the liquid level of the solution drops, and the thermal components used to simulate the heat source in the thermal-hydraulic system program cannot change dynamically, resulting in difficulty in using the system program to simulate the power distribution in the transient conditions of the aqueous solution type nuclear reactor.

[0041] Based on the above considerations, through in-depth research, the inventor proposed a method for modeling the thermal-hydraulic grid dynamic power distribution of a nuclear reactor. This method establishes a nodalized grid model of the container where the reactor solution is located, uses thermal components to simulate the self-heating effect of the solution, then determines the initial power distribution in each grid within the solution corresponding to the initial liquid level of the solution charge according to the divided grid scheme, calculates the value of the solution liquid level in real time during the transient condition simulation, determines the axial grid position where the liquid level is located, and calculates the power distribution of the solution in the transient condition according to the liquid level height. This technology achieves the purpose of accurately simulating the power distribution of an aqueous solution type nuclear reactor using the traditional thermal-hydraulic system program.

[0042] Please refer to Figure 1 , a method for modeling the thermal-hydraulic grid dynamic power distribution of a nuclear reactor provided in the embodiment of the present application specifically includes the following steps:

[0043] Step 1: Determine the geometric design of the target aqueous solution type nuclear reactor, divide the reactor container into multiple concentric ring-shaped parallel channels, and the number of nodes in the radial direction is determined by sensitivity analysis.

[0044] Taking a common cylindrical reactor container as an example, multiple concentric cylindrical surfaces are used to divide the internal space of the container, thereby forming multiple ring-shaped channels, where the number of cylindrical surfaces is the number of nodes. It should be noted that in the radial direction, the multiple concentric cylindrical surfaces can be evenly distributed or non-uniformly distributed.

[0045] Step 2: Calculate the power distribution in the aqueous solution type nuclear reactor container according to three-dimensional neutron kinetics, and calculate the radial power distribution ratio in each ring-shaped channel according to the radial node division and the solution charge volume ratio.

[0046] Step 3: Calculate and determine the initial liquid level height of the solution under steady-state conditions according to the designed filling volume of the aqueous solution type nuclear reactor solution and the geometric structure of the reactor container.

[0047] Step 4: Divide each ring-shaped channel into axial grids at the same height, and make the initial liquid level on the boundary line between two axial grids.

[0048] After the division, multiple axial grids (control volumes or nodes) are formed inside the container. It should be noted that in the axial direction, the axial grids can be divided in an equal-height manner or an unequal-height manner.

[0049] Step 5: After determining the hydrodynamic nodalization grid model of the container where the solution is located, establish a corresponding thermal component simulation scheme. Use a thermal component with high thermal conductivity and low heat capacity to simulate the heat source of the solution, and set the heat transfer coefficient between the thermal component and the solution to be greater than 10 8 Simulate the self-heating effect of the solution.

[0050] Step 6: According to the radial power distribution determined in Step 2 and the axial grids divided in Step 4, calculate the power distribution in each control volume of the solution, and use the method in Step 5 to simulate the self-heating effect of the solution in different control volumes.

[0051] Step 7: Write a program to calculate variables for calculating the liquid level of the solution: First, calculate the liquid level height in each control volume by multiplying the liquid phase fraction in each control volume by the axial height of the control volume, and then calculate the total liquid level of the solution by weighted averaging the liquid level heights in all control volumes.

[0052] Step 8: In the transient condition simulation, calculate the value of the liquid level of the solution in real time, judge the axial grid position where the liquid level is located, calculate the volume distribution of the solution in each grid in real time according to the liquid level height, and then calculate the power distribution of the solution in real time according to the volume distribution. Use a judgment logic to set the logic of zeroing the power of the current control volume when the liquid level height in the axial control volume where the liquid level is located is lower than 1% of the current control volume height, so as to achieve the goal of dynamically simulating the power distribution of the aqueous solution type nuclear reactor in transient conditions.

[0053] Specifically, judge whether the liquid level height in the axial control volume where the liquid level is located is lower than 1% of the current control volume height. If so, zero the power of the current control volume and calculate the axial power distribution according to the liquid level heights in the axial control volumes below the liquid level; if not, calculate the axial power distribution according to the liquid level heights in each axial control volume.

[0054] In the nuclear reactor thermal-hydraulic grid-based dynamic power distribution modeling technology proposed in the embodiments of the present application, the nodalization grid of the aqueous solution type nuclear reactor is reasonably divided through sensitivity analysis, and a thermal component with high thermal conductivity and low heat capacity is used to equivalently simulate the self-heating effect of the solution of the aqueous solution type nuclear reactor. Subsequently, the power distribution in the solution is dynamically calculated based on the transient liquid level, achieving the purpose of accurately simulating the power distribution of the aqueous solution type nuclear reactor using a traditional thermal-hydraulic system program.

[0055] The thermal-hydraulic grid-based dynamic power distribution modeling technology proposed in the embodiments of the present application can be used for the safety analysis of transient accident conditions of an aqueous solution-type nuclear reactor with liquid nuclear fuel. When there is a lack of relevant transient analysis system-level thermal-hydraulic programs, it can achieve an equivalent high-precision simulation effect using traditional thermal-hydraulic system programs. This modeling technology can greatly promote the accident safety analysis process of aqueous solution-type nuclear reactors based on high-precision surrogate simulations, save the cost and time of developing system-level thermal-hydraulic programs suitable for transient analysis of aqueous solution-type nuclear reactors, and there is no need to conduct additional tests for program verification, achieving the effect of saving time and effort.

[0056] The specific embodiments described above further elaborate on the purpose, technical solutions, and beneficial effects of the present invention. It should be understood that the above are only specific embodiments of the present invention and are not used to limit the protection scope of the present invention. Any modifications, equivalent replacements, improvements, etc. made within the spirit and principles of the present invention shall be included in the protection scope of the present invention.

Claims

1. A method for modeling the thermal-hydraulic grid-based dynamic power distribution in a nuclear reactor, characterized in that, Including: Step 1: Divide the aqueous solution type nuclear reactor vessel into multiple concentric ring-shaped parallel channels; Step 2: Calculate the radial power distribution ratio in each of the concentric ring-shaped parallel channels; Step 3: Determine the initial liquid level height of the solution under steady-state conditions; Step 4: Conduct axial grid division for each of the concentric ring-shaped parallel channels. After division, multiple axial grids are formed inside the vessel, and each axial grid is a control volume. The multiple axial grids constitute a nodalized grid model; Step 5: Establish a corresponding thermal component simulation scheme according to the determined nodalized grid model; Step 6: Determine the initial power distribution in each control volume of the solution; Step 7: During the transient condition simulation, calculate the value of the solution liquid level in real time, judge the axial grid position where the liquid level is located, and calculate the power distribution of the solution in real time according to the liquid level height.

2. The method for modeling the dynamic power distribution of a nuclear reactor's thermal-hydraulics in a grid form according to claim 1, wherein In Step 1, the number of nodes in the radial direction of the reactor vessel is determined by sensitivity analysis, and the number of nodes is the number of concentric cylindrical surfaces dividing the internal space of the vessel.

3. The method for modeling the thermal-hydraulic grid dynamic power distribution of a nuclear reactor according to claim 1, characterized in that, Step 2 is to calculate the power distribution in the aqueous solution type nuclear reactor vessel according to three-dimensional neutron kinetics, and calculate the radial power distribution ratio in each of the concentric ring-shaped parallel channels according to the radial node division and the solution loading volume ratio. The nodes are the concentric cylindrical surfaces dividing the internal space of the vessel.

4. The method for modeling the dynamic power distribution of a nuclear reactor's thermal-hydraulics in a grid format according to claim 1, characterized in that, In Step 3, the initial liquid level height of the solution under steady-state conditions is calculated and determined according to the designed loading of the aqueous solution type nuclear reactor solution and the geometric structure of the reactor vessel.

5. The method for modeling the dynamic power distribution of a nuclear reactor's thermal-hydraulics in a grid pattern according to claim 1, characterized in that, When dividing the axial grid in Step 4, each of the concentric ring-shaped parallel channels is axially grid-divided at the same height, and the liquid level is made to be on the boundary line between two axial grids.

6. The method for modeling the thermal-hydraulic grid dynamic power distribution of a nuclear reactor according to claim 1, wherein When establishing the thermal component simulation scheme in step five, use a thermal component with high thermal conductivity and low heat capacity to simulate the heat source of the solution, and set the heat transfer coefficient between the thermal component and the solution to be greater than 10 8 Simulate the self-heating effect of the solution.

7. The method for modeling the dynamic power distribution of a nuclear reactor's thermal-hydraulics in a grid pattern according to claim 1, characterized in that, In Step 6, the power distribution in each control volume of the solution is calculated according to the radial power distribution ratio determined in Step 2 and the axial grid divided in Step 4.

8. The method for modeling the thermal-hydraulic grid dynamic power distribution of a nuclear reactor according to claim 1, characterized in that, In Step 7, first calculate the liquid level height in each control volume by multiplying the liquid phase fraction in each control volume by the axial height of the control volume, and then calculate the total solution liquid level by performing weighted averaging on the liquid level heights in all control volumes.

9. The method for modeling the dynamic power distribution of a nuclear reactor's thermal-hydraulics in a grid form according to claim 8, characterized in that, In Step 7, calculate the volume distribution of the solution in each grid in real time according to the liquid level height, and then calculate the power distribution of the solution in real time according to the volume distribution.

10. The method for modeling the thermal-hydraulic grid-based dynamic power distribution of a nuclear reactor according to claim 9, characterized in that, In Step 7, set the judgment logic that when the liquid level height in the axial control volume where the liquid level is located is lower than 1% of the current control volume height, the power of the current control volume is set to zero.

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