A nuclear reactor automatic modeling method based on subchannel calculation program

Abstract

The application discloses a kind of nuclear reactor automatic modeling methods based on sub-channel calculation program, main steps are as follows:1, read the three parts of input parameter, nuclear reactor operating condition parameter, core arrangement, power distribution of each component respectively;2, according to the arrangement mode of each component of nuclear reactor core obtained, the number of channels is counted;3, determine fuel assembly number and channel number;4, generate the position relationship of channel and channel;5, calculate the hydraulics parameters of each type of channel;6, generate the input file required by sub-channel calculation program, complete nuclear reactor automatic modeling function.The application has the full-core fast modeling function of realizing specific loading scheme, can realize the fast modeling of each type of fuel assembly, control rod assembly, isotope target according to specific loading scheme layout, so as to greatly improve the calculation efficiency of sub-channel calculation program.

Description

Technical Field

[0001] This invention belongs to the field of nuclear reactor thermal-hydraulic technology, specifically involving an automatic preprocessing modeling method based on a nuclear reactor subchannel calculation program. Background Technology

[0002] The single-channel model is commonly used for thermal-hydraulic calculations of nuclear reactors. In this model, the channel being calculated is considered isolated and closed, with no exchange of mass, momentum, or energy between it and other channels across the entire core height. This analytical model is unsuitable for open channels like those without cascaded assemblies. To make the calculations more realistic, the more advanced sub-channel model was developed. This model assumes that there is a lateral exchange of mass, momentum, and energy (i.e., lateral mixing) between coolants in adjacent channels during flow. Therefore, the coolant mass flow density in each channel will continuously change along the axial direction, and the enthalpy and temperature of the coolant in the hot channels will decrease. Correspondingly, the surface and core temperatures of the fuel elements will also decrease slightly. For large nuclear reactors, under constant thermal parameters, comparing the results calculated using the sub-channel model with those calculated using the single-channel model shows an increase of approximately 5% to 10% in the MDNBR value of the fuel element surface. It is evident that using the sub-channel model improves both the accuracy of thermal design and the economics of the reactor. However, unlike the single-channel model, which only considers a few hot channels and hot spots for calculation, the sub-channel model requires the analysis of a large number of channels. Therefore, the computational workload is very large.

[0003] Currently, there are many programs that use subchannel models for core thermal-hydraulic analysis, such as COBRA, RETRAN, and RELAP. For the COBRA program, since the actual arrangement of different reactor components is random, the core input cards require a large number of physical parameters to be entered according to the format specified by the subchannel program. The traditional approach is to manually fill in the cards, which is time-consuming, inefficient, and labor-intensive. Summary of the Invention

[0004] To overcome the problems existing in the prior art, the present invention aims to provide an automatic modeling method for nuclear reactors based on a subchannel calculation program. This method calculates hydraulic parameters and the positional relationship between channels and fuel by inputting nuclear reactor operating parameters, and finally outputs all parameters required for the subchannel calculation program. Input parameters include reactor operating condition parameters, core layout, and power distribution of each component. For each core component, hydraulic parameters such as flow area, wetted perimeter, and heating length are calculated based on its geometry. Output parameters are all parameters required by the subchannel calculation program, output in the format specified by the program.

[0005] To achieve the above objectives, the present invention adopts the following technical solution:

[0006] An automatic modeling method for nuclear reactors based on a subchannel calculation program is disclosed. The pre-processing automatic modeling method calculates hydraulic parameters and the positional relationship between channels and fuel by inputting nuclear reactor operating parameters, and finally outputs all parameters required for the subchannel calculation program. Input parameters include nuclear reactor operating condition parameters, core layout, and power distribution of each component. For each core component, hydraulic parameters such as flow area, wetted perimeter, and heating length are calculated based on its geometry. Output parameters are all parameters required by the subchannel calculation program, output in the format specified by the program. The method includes the following steps:

[0007] Step 1: Input the nuclear reactor operating condition parameters, core layout, and power distribution parameters of each component; the nuclear reactor operating condition parameters include the axial and radial node division of the control body, the geometric parameters of each component, the reactor coolant inlet temperature, the average mass flow rate of the coolant inlet, and the reactor system pressure; the core layout includes using the same number to represent the same type of component under a given core loading scheme, and using natural numbers 1-4 to mark the various types of components under specific loading; the power distribution parameters of each component include the line power data of all components in the core;

[0008] Step 2: Based on the input parameters from Step 1, obtain the geometric arrangement of each component in the core and count the number of channels of the same type; channels of adjacent different components are classified according to component type and the types of channels are counted.

[0009] Step 3: Based on the core layout parameters, determine the component numbers at each location in the core; channels are numbered in natural order from left to right and from top to bottom, and each channel can have a maximum of four adjacent channels, generating the positional relationship between channels;

[0010] Step 4: Calculate the hydraulic parameters of each component based on the geometric parameters of each component in the nuclear reactor operating condition parameters, including the flow area, wetted perimeter and heating length of each component;

[0011] Step 5: After calculating the hydraulic parameters of each channel, generate a corresponding number for each channel based on its location;

[0012] Step 6: Output all the calculated parameter results in the format required by the sub-channel calculation program, including the nuclear reactor operating condition parameters and power distribution parameters of each component input in Step 1, the channel types in Step 2, the positional relationship parameters between channels in Step 3, the hydraulic parameters of various components in Step 4, and the corresponding number parameters of various channels in Step 5, thus completing the automatic modeling of the nuclear reactor.

[0013] Compared with the prior art, the present invention has the following advantages:

[0014] This invention, for different nuclear reactors, only requires inputting the corresponding core loading method in the core layout, providing the power data of each component of the nuclear reactor, and the operating parameters of the nuclear reactor. Finally, it outputs all the parameters required by the sub-channel calculation program according to the format required by the sub-channel program, thereby realizing rapid modeling of different core layouts. This improves modeling efficiency while ensuring modeling accuracy, and facilitates the improvement of sub-channel thermal-hydraulic calculation efficiency. This invention is applicable to various core loading schemes of nuclear reactors and has wide applicability. Attached Figure Description

[0015] Figure 1 This is a flowchart of the automatic modeling method for nuclear reactors according to the present invention.

[0016] Figure 2 This is a schematic diagram of the operating parameters of a nuclear reactor.

[0017] Figure 3 This is a schematic diagram of the reactor core layout.

[0018] Figure 4 This is a schematic diagram of the power distribution of each component in the reactor core. Detailed Implementation

[0019] The present invention will now be further described with reference to examples and accompanying drawings:

[0020] like Figure 1 As shown, this invention provides an automatic modeling method for nuclear reactors based on a subchannel calculation program. Figure 2 This is a schematic diagram of the operating parameters of a nuclear reactor. The physical meaning of each parameter in each row is shown below:

[0021] Line 1: Analysis type, input parameter units, output parameter units, three-equation model.

[0022] Line 2: Closed channel, total number of coolant channels, total number of fuel rods, number of axial control body node divisions, number of channel types, number of positioning grid positions, type of positioning grid position, number of radial nodes of the pellet, number of fuel rod types.

[0023] Line 3: Axial control height.

[0024] Line 4: Input the number of control bodies NAXP.

[0025] Line 5: Pellet diameter, cladding thickness, fuel density, cladding density, fuel rod diameter, correction factor for heat source distribution in pellet.

[0026] Line 6: Chip thermal conductivity, chip specific heat capacity, cladding thermal conductivity, cladding specific heat capacity, gap heat transfer coefficient, fission power share directly entering the cladding.

[0027] Line 7: Turbulent mixing model selection, single-phase friction coefficient indicator, two-phase friction coefficient indicator, cavitation fraction indicator, inlet flow rate indicator, parameter selection indicator, iteration indicator, property indicator, heat transfer model, property pressure.

[0028] Line 8: Two-phase friction model.

[0029] Line 9: Selection of cavitation fraction model.

[0030] Line 10: Single-phase heat transfer coefficient model, subcooled boiling heat transfer coefficient model, saturated boiling heat transfer coefficient model, CHF point model, transition zone heat transfer coefficient model, film boiling heat transfer coefficient model.

[0031] Line 11: Crossflow drag coefficient, turbulent momentum factor, crossflow momentum equation parameters, angle between the channel and the numerical direction.

[0032] Line 12: Critical heat flux density formula selection, cold wall factor correction, axial non-uniform power correction, positioning grid pressure drop correction factor, critical power relationship, fuel rod section diameter.

[0033] Line 13: Conservation parameters: all default, pressure gradient solution.

[0034] Line 14: Inlet temperature options, reactor coolant inlet temperature, coolant inlet average mass flow rate, reactor system pressure, isobaric distribution flow rate, others are default values.

[0035] Line 15: Time data: steady state, so all values ​​are 0.

[0036] Line 16: Transient parameters: steady state, so all are 0.

[0037] Line 17: Output channel results + fuel rod results, the rest are default values.

[0038] Figure 3 The core layout is for a specific loading scheme. The core layout includes the same type of components represented by the same number under a given core loading scheme. The various types of components under the specific loading are marked by natural numbers 1-4. In the figure, number a is the number of columns in the core layout, number b is the number of rows in the core layout, and number c is an example of the core layout. Figure 4 This is a schematic diagram of the power distribution of the reactor core components, containing the line power data for all components. Since the power of components varies at different locations, the power of different components is arranged in rows of six.

[0039] This invention possesses a rapid full-core modeling capability for specific loading schemes, enabling rapid modeling of fuel assemblies, control rod assemblies, and isotope targets based on specific loading scheme layout diagrams. It improves computational efficiency while ensuring modeling accuracy, facilitating the enhancement of subchannel thermal-hydraulic calculation efficiency.

[0040] The method includes the following steps:

[0041] Step 1: Input three sets of parameters, including nuclear reactor operating condition parameters, core layout, and power distribution parameters for each component. Nuclear reactor operating condition parameters include the axial and radial node division of the control body, geometric parameters of each component, reactor coolant inlet temperature, average mass flow rate of the coolant inlet, and reactor system pressure. The core layout is as follows: Figure 3 As shown, under a given core loading scheme, the same number represents the same type of component, and the natural numbers 1-4 are used to label the specific types of components under different loading schemes. The power distribution parameters of each component are as follows: Figure 4 As shown, it contains the line power data for all components of the reactor core;

[0042] Step 2: Based on the input parameters from Step 1, obtain the geometric arrangement of each component in the core and count the number of channels of the same type; channels of adjacent different components are classified according to component type and the types of channels are counted.

[0043] Step 3: Based on the core layout parameters, determine the component numbers at each location in the core; channels are numbered in natural order from left to right and from top to bottom, and each channel can have a maximum of four adjacent channels, generating the positional relationship between channels;

[0044] Step 4: Calculate the hydraulic parameters of each component based on the geometric parameters of each component in the nuclear reactor operating condition parameters, including the flow area, wetted perimeter, and heating length of each component;

[0045] Step 5: After calculating the hydraulic parameters of each channel, generate a corresponding number for each channel based on its location;

[0046] Step 6: Output all the calculated parameter results in the format required by the sub-channel calculation program, including the nuclear reactor operating condition parameters and power distribution parameters of each component input in Step 1, the channel types in Step 2, the positional relationship parameters between channels in Step 3, the hydraulic parameters of various components in Step 4, and the corresponding number parameters of various channels in Step 5, thus completing the automatic modeling of the nuclear reactor.

[0047] All contents not described in detail in this invention are common knowledge in the field.

Claims

1. An automatic modeling method for nuclear reactors based on a subchannel computation program, characterized in that: The automatic modeling method calculates hydraulic parameters and the positional relationship between channels and fuel by inputting nuclear reactor operating parameters, and finally outputs all parameters for the sub-channel calculation program. The input parameters include nuclear reactor operating condition parameters, core layout, and power distribution of each component. The hydraulic parameters of each core component, such as flow area, wetted perimeter, and heating length, are calculated based on its geometry. The output parameters are all the parameters required by the sub-channel calculation program, output in the format required by the sub-channel calculation program; the automatic modeling method includes the following steps: Step 1: Input the nuclear reactor operating condition parameters, core layout, and power distribution parameters of each component; the nuclear reactor operating condition parameters include the axial and radial node division of the control body, the geometric parameters of each component, the reactor coolant inlet temperature, the average mass flow rate of the coolant inlet, and the reactor system pressure; the core layout includes using the same number to represent the same type of component under a given core loading scheme, and using natural numbers 1-4 to mark the various types of components under specific loading; the power distribution parameters of each component include the line power data of all components in the core; Step 2: Based on the input parameters from Step 1, obtain the geometric arrangement of each component in the core and count the number of channels of the same type; channels of adjacent different components are classified according to component type and the types of channels are counted. Step 3: Based on the core layout parameters, determine the component numbers at each location in the core; channels are numbered in natural order from left to right and from top to bottom, and each channel can have a maximum of four adjacent channels, generating the positional relationship between channels; Step 4: Calculate the hydraulic parameters of each component based on the geometric parameters of each component in the nuclear reactor operating condition parameters, including the flow area, wetted perimeter and heating length of each component; Step 5: After calculating the hydraulic parameters of each channel, generate a corresponding number for each channel based on its location; Step 6: Output all the calculated parameter results in the format required by the sub-channel calculation program, including the nuclear reactor operating condition parameters and power distribution parameters of each component input in Step 1, the channel types in Step 2, the positional relationship parameters between channels in Step 3, the hydraulic parameters of various components in Step 4, and the corresponding number parameters of various channels in Step 5, thus completing the automatic modeling of the nuclear reactor.

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