CN112199811B - Method and device for determining reactor core parameters of nuclear thermal propulsion reactor - Google Patents
Method and device for determining reactor core parameters of nuclear thermal propulsion reactor Download PDFInfo
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- 229910052770 Uranium Inorganic materials 0.000 claims abstract description 15
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Abstract
The invention provides a method and a device for determining reactor core parameters of a nuclear heat propulsion reactor, relating to the technical field of nuclear heat propulsion, wherein the method comprises the following steps: acquiring a fluid-solid coupling heat transfer model of a nuclear thermal propulsion reactor; acquiring target power parameters meeting aircraft performance constraints, and performing heat transfer calculation on the fluid-solid coupling heat transfer model based on the target power parameters to obtain multiple groups of design parameters meeting thermal engineering constraints; each group of design parameters comprises geometric parameters and thermal parameters; performing neutron modeling and physical analysis based on multiple sets of design parameters to obtain target geometric parameters and target thermal parameters which meet neutron physical constraints; and taking the target power parameters, the target thermal parameters and the target geometric parameters as reactor core design parameters of the nuclear heat reactor. The method can improve the reliability of the reactor core parameter design in the low-enrichment-degree concentrated uranium nuclear heat propulsion reactor.
Description
Technical Field
The invention relates to the technical field of nuclear heat propulsion, in particular to a method and a device for determining reactor core parameters of a nuclear heat propulsion reactor.
Background
The thermonuclear propulsion means that nuclear fission energy is utilized to heat working media, and then the heated high-temperature high-pressure working media are directionally sprayed out to obtain thrust. Nuclear thermal propulsion reactors typically use high enriched (> 90%) enriched uranium as nuclear fuel from the fifties of the last century, and researchers are designing low enriched uranium (< 20%) nuclear thermal propulsion reactors in recent years in order to reduce the risk of nuclear diffusion and reduce the development cost. However, for the design of the low enrichment enriched uranium core parameters of the nuclear thermal propulsion reactor, researchers only make appropriate modifications on the basis of the high enrichment enriched uranium core design parameters, so that the existing core parameter design mode in the low enrichment enriched uranium nuclear thermal propulsion reactor has the problem of low reliability.
Disclosure of Invention
In view of this, the present invention provides a method and an apparatus for determining reactor core parameters of a nuclear thermal propulsion reactor, which can improve the reliability of the design of the reactor core parameters in the low enrichment enriched uranium nuclear thermal propulsion reactor.
In order to achieve the above purpose, the embodiment of the present invention adopts the following technical solutions:
in a first aspect, an embodiment of the present invention provides a method for determining a core parameter of a nuclear thermal propulsion reactor, including: acquiring a fluid-solid coupling heat transfer model of the nuclear thermal propulsion reactor; acquiring target power parameters meeting aircraft performance constraints, and performing heat transfer calculation on the fluid-solid coupling heat transfer model based on the target power parameters to obtain multiple groups of design parameters meeting thermal engineering constraints; each group of design parameters comprises geometric parameters and thermal parameters; performing neutron modeling and physical analysis based on the plurality of groups of design parameters to obtain target geometric parameters and target thermal parameters which meet neutron physical constraints; and taking the target dynamic parameter, the target thermal engineering parameter and the target geometric parameter as the core design parameters of the nuclear thermal reactor.
Further, embodiments of the present invention provide a first possible implementation manner of the first aspect, wherein the geometric parameters include a core height, a core radius, and a coolant flow channel diameter; the method comprises the following steps of obtaining a fluid-solid coupling heat transfer model of the nuclear thermal propulsion reactor, wherein the steps comprise: respectively determining a fuel assembly structure and a moderating assembly structure in the nuclear heat reactor based on the core height, the core radius and the coolant runner diameter; and obtaining a fluid-solid coupling heat transfer model of the nuclear heat reactor based on the fuel assembly structure, the moderating assembly structure and the flow direction of the coolant.
Further, embodiments of the present invention provide a second possible implementation manner of the first aspect, wherein the fuel assembly structure and the moderator structure are both prism structures, the height and cross-sectional shape of the fuel assembly structure and the moderator structure are the same, the fuel assembly structure includes a plurality of coolant flow channels, the moderator structure includes a second coolant flow channel, the coolant flows from the second coolant flow channel to a first coolant flow channel, and the first coolant flow channel is each coolant flow channel in the fuel assembly structure.
Further, embodiments of the present invention provide a third possible implementation manner of the first aspect, wherein the thermal parameter includes a core outlet temperature; the step of carrying out heat transfer calculation on the fluid-solid coupling heat transfer model based on the target dynamic parameter to obtain a plurality of groups of design parameters meeting thermal engineering constraints comprises the following steps: performing axial heat transfer calculation and radial heat transfer calculation on the fluid-solid coupling heat transfer model based on the target power parameters to obtain the wall surface temperature and the fuel center temperature of the coolant flow channel, the mainstream temperature and the reactor core outlet temperature at any axial position and the geometric parameters corresponding to the reactor core outlet temperature; and repeatedly executing the axial heat transfer calculation and the radial heat transfer calculation until a first preset condition is reached to obtain a plurality of groups of design parameters.
Further, an embodiment of the present invention provides a fourth possible implementation manner of the first aspect, where the first preset condition includes: the fuel core temperature reaches a maximum fuel temperature, and/or the temperature of the moderator in the structure of the moderator reaches a maximum moderator temperature.
Further, embodiments of the present invention provide a fifth possible implementation manner of the first aspect, wherein the geometric parameters include a core height, a core radius, and a coolant flow channel diameter; the step of performing neutron modeling and physical analysis based on the plurality of sets of design parameters to obtain target geometric parameters and target thermal parameters meeting neutron physical constraints comprises the following steps: selecting any one group of design parameters from the multiple groups of design parameters to obtain first design parameters, and performing neutron modeling based on the geometric parameters in the first design parameters to obtain a target model; performing reactor physical analysis on the target model to obtain physical parameters of the target model; and judging whether the physical parameters meet a second preset condition, and if so, respectively taking the geometric parameters and the thermal parameters in the first design parameters as target geometric parameters and target thermal parameters.
Further, embodiments of the present invention provide a sixth possible implementation manner of the first aspect, wherein the physical parameter includes a core effective multiplication factor; the step of judging whether the physical parameter meets a second preset condition comprises the following steps: and judging whether the effective multiplication coefficient is larger than 1, and if so, determining that the physical parameter meets the second preset condition.
Further, an embodiment of the present invention provides a seventh possible implementation manner of the first aspect, where the method further includes: if the physical parameters do not meet the second preset condition, selecting any one group of design parameters except the first design parameters from the multiple groups of design parameters to obtain new first design parameters; and performing neutron modeling and physical analysis based on the geometric parameters in the new first design parameters until target geometric parameters and target thermal parameters are obtained.
In a second aspect, an embodiment of the present invention further provides a core parameter determination apparatus for a nuclear thermal propulsion reactor, including: the acquisition module is used for acquiring a fluid-solid coupling heat transfer model of the nuclear thermal propulsion reactor; the calculation module is used for acquiring target power parameters meeting aircraft performance constraints, and performing heat transfer calculation on the fluid-solid coupling heat transfer model based on the target power parameters to obtain multiple groups of design parameters meeting thermal engineering constraints; each group of design parameters comprises geometric parameters and thermal parameters; the analysis module is used for performing neutron modeling and physical analysis based on the plurality of groups of design parameters to obtain target geometric parameters and target thermal parameters meeting neutron physical constraints; and the parameter determining module is used for taking the target dynamic parameter, the target thermotechnical parameter and the target geometric parameter as the core design parameters of the nuclear heat reactor.
In a third aspect, the present invention provides a computer-readable storage medium, on which a computer program is stored, where the computer program is executed by a processor to perform the steps of the method described in any one of the above first aspects.
The embodiment of the invention provides a method and a device for determining reactor core parameters of a nuclear thermal propulsion reactor, wherein the method comprises the following steps: firstly, acquiring a fluid-solid coupling heat transfer model of a nuclear thermal propulsion reactor; acquiring target dynamic parameters meeting aircraft performance constraints, and performing heat transfer calculation on the fluid-solid coupling heat transfer model based on the target dynamic parameters to obtain multiple groups of design parameters meeting thermal engineering constraints (each group of design parameters comprises geometric parameters and thermal engineering parameters); performing neutron modeling and physical analysis based on multiple groups of design parameters to obtain target geometric parameters and target thermal parameters meeting neutron physical constraints; and taking the target power parameters, the target thermal parameters and the target geometric parameters as reactor core design parameters of the nuclear heat reactor. According to the method, the reactor core design parameters meeting the performance constraints of the aircraft, thermal constraints and neutron physical constraints can be obtained simultaneously by performing heat transfer calculation on the fluid-solid coupling heat transfer model of the nuclear thermal propulsion reactor based on the target power parameters meeting the performance constraints of the aircraft and performing neutron modeling and physical analysis, so that the effectiveness of the reactor core design parameters is ensured.
Additional features and advantages of embodiments of the invention will be set forth in the description which follows, or in part may be learned by the practice of the embodiments of the invention or may be learned by the practice of the embodiments of the invention as set forth hereinafter.
In order to make the aforementioned and other objects, features and advantages of the present invention comprehensible, preferred embodiments accompanied with figures are described in detail below.
Drawings
In order to more clearly illustrate the embodiments of the present invention or the technical solutions in the prior art, the drawings used in the description of the embodiments or the prior art will be briefly described below, and it is obvious that the drawings in the following description are some embodiments of the present invention, and other drawings can be obtained by those skilled in the art without creative efforts.
FIG. 1 illustrates a flow chart of a method for determining core parameters of a nuclear thermal propulsion reactor according to an embodiment of the invention;
FIG. 2 is a top view of a fluid-solid coupling heat transfer model according to an embodiment of the present invention;
FIG. 3 is a side cut view of a fluid-solid coupling heat transfer model provided by an embodiment of the invention;
FIG. 4 illustrates a flow chart of a design of low enrichment uranium enrichment core parameters provided by an embodiment of the invention;
FIG. 5 is a schematic diagram illustrating a core parameter determination device of a nuclear thermal propulsion reactor according to an embodiment of the present invention;
fig. 6 shows a schematic structural diagram of an electronic device according to an embodiment of the present invention.
Icon:
011-a first coolant flow passage; 012-a second coolant flow channel; 013 — third coolant flow passages; 014-zirconium hydride material; 015-zirconium-4 material; 016-zirconium carbide material; 017-graphite material; 018-tungsten-uranium dioxide material; 021-core inlet; 022-core outlet.
Detailed Description
To make the objects, technical solutions and advantages of the embodiments of the present invention clearer, the technical solutions of the present invention will be described below with reference to the accompanying drawings, and it is apparent that the described embodiments are some, not all, embodiments of the present invention.
At present, in consideration of the problem of low reliability of the existing core parameter design mode of the nuclear thermal propulsion reactor, in order to improve the problem, the core parameter determination method and device of the nuclear thermal propulsion reactor provided by the embodiment of the invention can be applied to improving the reliability of the core parameter determination of the low-enrichment enriched uranium. The following describes embodiments of the present invention in detail.
The embodiment provides a method for determining reactor core parameters of a nuclear thermal propulsion reactor, which can be applied to parameter design of a low-enrichment enriched uranium nuclear thermal propulsion reactor, and is shown in a flow chart of a method for determining reactor core parameters of a nuclear thermal propulsion reactor shown in fig. 1, and the method mainly includes the following steps S102 to S106:
and S102, acquiring a fluid-solid coupling heat transfer model of the nuclear thermal propulsion reactor.
Due to the presence of two different components within the core of a nuclear heat-propelled reactor: the fuel assembly and the moderating assembly are used for establishing a fluid-solid coupling heat transfer model of the nuclear thermal propulsion reactor by determining the geometric parameters of the core (namely the size and the shape of the fuel assembly and the moderating assembly) and analyzing the flowing rule of the coolant in the nuclear thermal propulsion reactor. Through the fluid-solid coupling heat transfer model, the heat transfer process of the coolant, the fuel assemblies and the moderating assemblies in the core of the nuclear thermal propulsion reactor can be simulated and calculated.
And step S104, acquiring target power parameters meeting aircraft performance constraints, and performing heat transfer calculation on the fluid-solid coupling heat transfer model based on the target power parameters to obtain multiple groups of design parameters meeting thermal engineering constraints.
The power parameters comprise parameters such as specific impulse, thrust and the like, power parameter values meeting aircraft performance constraints are determined according to the value range of the power parameters to obtain target power parameters, the specific impulse value in a nuclear thermal propulsion system is as high as possible under normal conditions, and the value of the specific impulse (which can be called specific impulse and is the impulse generated by the unit propellant quantity) can be about 900 s.
Each set of design parameters includes a geometric parameter and a thermal parameter, the thermal parameter may be a core outlet temperature, and the geometric parameter may be a core height, a radius, a coolant flow channel diameter, and the like. In the process of carrying out heat transfer calculation on the fluid-solid coupling heat transfer model, each iteration calculation can obtain a group of design parameter values (reactor core outlet temperature and corresponding geometric parameter values), and when the iteration calculation meets the thermal convergence condition, a plurality of groups of design parameters meeting thermal constraints can be obtained.
And S106, performing neutron modeling and physical analysis based on multiple sets of design parameters to obtain target geometric parameters and target thermal parameters meeting neutron physical constraints.
And (4) obtaining any one group of design parameters from the plurality of groups of design parameters obtained in the step (S104) to perform middle-school modeling, and performing physical analysis on the obtained model, so as to screen out geometric parameters and thermal parameters meeting neutron physical constraint conditions from the plurality of groups of design parameters to obtain target geometric parameters and target thermal parameters.
And S108, taking the target power parameter, the target thermal parameter and the target geometric parameter as reactor core design parameters of the nuclear thermal reactor.
The target dynamic parameter, the target thermal engineering parameter and the target geometric parameter obtained by the calculation are design parameters meeting aircraft performance constraints, thermal engineering constraints and neutron physical constraints, so that the values of the target dynamic parameter, the target thermal engineering parameter and the target geometric parameter meeting the aircraft performance constraints, the thermal engineering constraints and the neutron physical constraints at the same time can be used as reactor core design parameters of the nuclear thermal propulsion reactor.
According to the method for determining the reactor core parameters of the nuclear thermal propulsion reactor, the reactor core design parameters meeting the aircraft performance constraint, the thermal engineering constraint and the neutron physical constraint can be obtained by performing heat transfer calculation on the fluid-solid coupling heat transfer model of the nuclear thermal propulsion reactor based on the target power parameters meeting the aircraft performance constraint and performing neutron modeling and physical analysis, so that the effectiveness of the reactor core design parameters is ensured.
In order to accurately obtain the fluid-solid coupled heat transfer model, this embodiment provides a specific implementation manner of obtaining the fluid-solid coupled heat transfer model of the nuclear thermal propulsion reactor: and respectively determining the structure of the fuel assembly and the structure of the slowing-down assembly in the nuclear heat reactor based on the height of the reactor core, the radius of the reactor core and the diameter of the coolant flow channel. And obtaining a fluid-solid coupling heat transfer model of the nuclear heat reactor based on the fuel assembly structure, the moderating assembly structure and the flow direction of the coolant. The core height, the core radius and the coolant flow channel diameter are geometric parameters of the nuclear heat propulsion reactor core, the core comprises fuel assemblies and a slowing assembly, the core height comprises the height of the fuel assemblies and the height of the slowing assembly, the core radius comprises the radius of the cross section of the fuel assemblies and the radius of the cross section of the slowing assembly, and the coolant flow channel diameter comprises the coolant flow channel diameter in the fuel assembly structure and the coolant flow channel diameter in the slowing assembly structure.
In a specific embodiment, the fuel assembly structure and the moderator structure are both prism structures, the height and cross-sectional shape of the fuel assembly structure and the moderator structure are the same, see the top view of the fluid-solid coupled heat transfer model shown in fig. 2, the left hexagonal prism in fig. 2 is the fuel assembly structure, the right hexagonal prism is the moderator structure, as shown in fig. 2, the fuel assembly structure is tungsten-uranium dioxide material 018 in the shape of a hexagonal prism, a plurality of cylindrical coolant channels (the cross-sections of the channels are shown by a plurality of circles in fig. 2) are uniformly distributed in the prism structure, and can be represented by a first coolant channel 011, each of the channels in the fuel assembly structure can be used as the first coolant channel 011, the moderator structure comprises a second coolant channel 012 and a third coolant channel 013, the third coolant channel 013 is a cylindrical channel, and is located in the center of the moderator structure, the second coolant channel 012 is an annular channel, the second coolant channel 012 and the third coolant channel 013 are filled with zirconium hydride material 014 and zirconium-4 material, the second coolant channel 015 and the zirconium carbide material are sequentially filled with the actual values according to the actual coolant material. Referring to the cut-away side view of the fluid-solid coupling heat transfer model shown in fig. 3, the coolant flow exchanges heat from the core inlet 021 into the third coolant flow channel 013 (also referred to as flow channel three), and further from the third coolant flow channel 013 to the second coolant flow channel 012 (also referred to as flow channel two) to the first coolant flow channel 011 (also referred to as flow channel one), and flows out from the core outlet 022 (as shown in the left side view of fig. 3), in practical applications, the coolant flow can be simplified to flow from the core inlet 021 into the second coolant flow channel 012 to the first coolant flow channel 011 (each coolant flow channel in the fuel assembly can be used as flow channel 1), and flow out from the core outlet 022 (as shown in the right side view of fig. 3), that is, the coolant flow direction is from the second coolant flow channel 012 to the first coolant flow channel 011, and the first coolant flow channel 011 is each coolant flow channel in the fuel assembly structure.
In order to obtain design parameters meeting thermal constraints, the embodiment provides an implementation manner of performing heat transfer calculation on a fluid-solid coupling heat transfer model based on target dynamic parameters to obtain multiple sets of design parameters meeting thermal constraints, and the implementation manner can be specifically executed by referring to the following steps (1) to (2):
step (1): and performing axial heat transfer calculation and radial heat transfer calculation on the fluid-solid coupling heat transfer model based on the target power parameters to obtain the wall surface temperature of the coolant flow channel, the fuel center temperature, and geometrical parameters corresponding to the main flow temperature, the reactor core outlet temperature and the reactor core outlet temperature at any axial position.
As shown in fig. 3, T in the left diagram M Wall temperature, T, of the side of the second coolant flow passage 012 remote from the fuel assembly w The wall surface temperature of the second coolant flow passage 012 on the side close to the fuel assembly. T in the right diagram Mn Is a discrete value in the axial direction of the wall surface temperature on the side of the second coolant flow passage 012 remote from the fuel assembly (the highest moderator temperature is T) Mn Maximum value of (d), T Fn The wall surface temperature of the second coolant flow passage 012 on the side close to the fuel assembly is in the axial directionDiscrete value (fuel center maximum temperature is T) Fn Maximum value of).
Axial (Z direction in fig. 3) heat transfer calculations were performed on the fluid-solid coupled heat transfer model, with a small portion of coolant flowing from third coolant flow passage 013 of the moderator assembly, through second coolant flow passage 012 and finally into first coolant flow passage 011 of fuel assembly FE, and a large portion of coolant once passing up and down within first coolant flow passage 011 of fuel assembly FE.
Total mass flow in the core
The relation between the thrust and the specific impulse is as follows
Wherein, I sp Is the dynamic parameter specific impulse of the aircraft, F is the dynamic parameter thrust of the aircraft, g 0 Is the acceleration of gravity, specific impulse I sp And core outlet temperature T e The thermal parameters satisfy the following relations:
gamma in the above formula is the ratio of constant pressure heat capacity to constant volume heat capacity, and is called specific heat capacity ratio. R is the Boltzmann constant and M is the relative molecular mass of the coolant.
Coolant mass flow through individual moderator assemblies
Comprises the following steps:
α is the fraction of the coolant mass flow through all of the moderating assemblies to the total core mass flow, and is expressed by the following equation:
N ME setting N for the number of moderating assemblies in the reactor core FE For the internal fuel assembly number of reactor core, have:
N FE +N ME ＝N 0
N 0 is the total number of components in the reactor core, N 0 The calculation formula of (a) is as follows:
N 0 ≈πr 2 /S 0
where r is the radius of the core active region, S 0 Is the cross-sectional area of the hexagonal assembly.
Introducing a number ratio beta of moderator assemblies to fuel assemblies 0 The method comprises the following steps:
β 0 ＝N ME /N FE
β 0 can be adjusted according to different arrangement rules of the in-stack components.
According to conservation of energy, mass flow is
Outlet temperature T 2 Flows out through the second coolant flow passage 012 at a mass flow rate of
Temperature of T 1 The coolant is mixed into the second coolant flow passage 011, the inlet temperature T of the mixed coolant 0 The following relation is satisfied:
in order to accurately calculate the maximum fuel temperature and the maximum moderator temperature in the axial direction, n nodes are divided in the axial direction Z, as shown in fig. 3. The temperature of the inner wall surface of each node of each flow channel and the temperature of the main flow meet the following conditions:
-q＝h(T w -T b )
in the above formula, q is the heat flux density, T w Is the wall surface temperature, h is the convective heat transfer coefficient, and the mainstream temperature T of the adjacent node b The following relation is satisfied:
q n the heat generated for the flow at node n (i.e., the heat resulting from the product of the heat flow density and the heat transfer area at node n), c p For specific heat capacity, assume that the linear power density of the coolant channel is q l Length of flow channel is L, mass flow rate is
At the same time, the power distribution presents a cosine distribution, and the temperature of the main flow at any axial position z in the flow direction can be calculated according to the following formula:
radial (X direction in fig. 3) heat transfer calculations were performed on the fluid-solid coupled heat transfer model, with radial primary focus on FE (fuel assembly) to ME (moderator assembly) heat transfer. T is a unit of Fn (fuel core temperature) vs. T Mn The heat transfer (moderator surface temperature) can be regarded as the heat conduction of a multilayer flat plate without an internal heat source, and then the FE conducts heat flow phi to ME 2 ：
p＝Φ 1 +β 1 Φ 2
β 1 Represents the average number of fuel assemblies adjacent to each moderator assembly in the core, and p is the power delivered by a single fuel assembly. The linear power density q in the axial heat transfer calculation can be further determined from the power p emitted by the fuel assembly l And fuel core temperature.
At a given initial aircraft performance parameter, such as thrust (25 k)N-75 kN) and thermal initial parameters such as the reactor core inlet temperature (about 120K), and the thermal parameters (the reactor core outlet temperature) and the reactor core geometric parameters (the reactor core height, the radius and the coolant flow channel diameter) of the reactor can be obtained by programming and simultaneously and iteratively solving the equations. The convergence criterion in the iterative calculation includes: maximum fuel center temperature max (T) Fn )<3000K, maximum moderator surface temperature max (T) Mn )<773K. In each iteration, a set of design parameters (i.e., core outlet temperature, and corresponding core geometry parameters) satisfying thermal constraints can be obtained.
Step (2): and repeatedly executing the axial heat transfer calculation and the radial heat transfer calculation until a first preset condition is reached to obtain a plurality of groups of design parameters.
The first preset condition includes: the fuel core temperature reaches the highest fuel temperature, and/or the temperature of the moderator in the structure of the moderator reaches the highest moderator temperature. And (2) repeatedly executing the axial heat transfer calculation and the radial heat transfer calculation in the step (1), when the fuel center temperature and/or the temperature of the moderator in the moderator assembly meet the first preset condition, reaching a convergence condition, ending the heat transfer calculation, namely when the fuel center temperature in the fluid-solid coupling heat transfer model reaches the maximum fuel temperature or when the temperature of the moderator in the moderator assembly structure reaches the maximum moderator temperature, determining that the heat transfer calculation reaches a thermodynamic convergence condition, and stopping the heat transfer calculation so that the fuel center temperature (including the maximum fuel center temperature) reaches the maximum fuel temperature and/or the temperature of the moderator in the moderator assembly (including the maximum moderator temperature) reaches the maximum moderator temperature. Because a group of design parameters can be obtained in each heat transfer calculation, a plurality of groups of design parameters can be obtained after a plurality of rounds of heat transfer calculation. In practical application, the maximum fuel temperature is less than 3000K, and the maximum moderator temperature is less than 773K. The moderator temperature may be a wall temperature T of the second coolant flow passage 012 on the side away from the fuel assembly M When discrete value T in axial direction Mn To the maximum value ofWhen the temperature of the moderator is the highest, the heat transfer calculation is determined to reach the thermal convergence condition. The above-mentioned fuel core temperature is the wall surface temperature of the second coolant flow passage 012 on the side close to the fuel assembly when the discrete value T is in the axial direction Fn When the maximum value of the heat transfer coefficient reaches the maximum fuel temperature, the heat transfer calculation is determined to reach the thermal convergence condition.
In order to obtain design parameters satisfying neutron physical constraints, this embodiment provides an implementation manner of performing neutron modeling and physical analysis based on multiple sets of design parameters to obtain target geometric parameters and target thermal parameters satisfying neutron physical constraints, which can be specifically executed with reference to the following steps 1 to 3:
step 1: and selecting any one group of design parameters from the multiple groups of design parameters to obtain a first design parameter, and performing neutron modeling based on the geometric parameters in the first design parameter to obtain a target model.
Because each group of design parameters comprises geometric parameters and thermal parameters, the geometric parameters comprise the height of a reactor core, the radius of the reactor core and the diameter of a coolant flow channel, and a target model corresponding to the geometric parameters can be obtained by performing neutron modeling based on the geometric parameters in any group of design parameters.
Step 2: and carrying out reactor physical analysis on the target model to obtain physical parameters of the target model.
The physical parameters of the target model, such as the effective multiplication coefficient of the reactor core, can be obtained through analysis by performing physical analysis of the reactor on the target model obtained through the neutron modeling.
And step 3: and judging whether the physical parameters meet a second preset condition, and if so, respectively taking the geometric parameters and the thermal parameters in the first design parameters as target geometric parameters and target thermal parameters.
And judging whether the physical parameter meets a second preset condition according to the size of the effective multiplication coefficient of the reactor core, specifically, judging whether the effective multiplication coefficient is greater than 1, and if the effective multiplication coefficient is greater than 1, determining that the physical parameter meets the second preset condition. When the physical parameters meet a second preset condition, the target model obtained by geometric parameter modeling in the first design parameters meets neutron physical constraint conditions, namely the first design parameters meet the neutron physical constraint conditions, the geometric parameters in the first design parameters are used as target geometric parameters, and the thermal parameters in the same group as the target geometric parameters are used as target thermal parameters.
In consideration of the case that the physical parameter does not satisfy the second preset condition, the method provided by this embodiment further includes: if the physical parameters do not meet the second preset condition, selecting any one group of design parameters except the first design parameters from the plurality of groups of design parameters to obtain new first design parameters; and performing neutron modeling and physical analysis based on the geometric parameters in the new first design parameters until target geometric parameters and target thermal parameters are obtained. And when the physical parameters do not meet a second preset condition, selecting any one group of design parameters except the first design parameters from the multiple groups of design parameters to serve as new first design parameters, or sequentially using the multiple groups of design parameters as new first design parameters, and repeatedly executing the steps 1-3 based on the new first design parameters until target geometric parameters and target thermal parameters meeting neutron physical constraint conditions are obtained.
The reactor core parameter determination method of the nuclear thermal propulsion reactor provided by the embodiment provides a complete calculation thought for parameter design of a low-enrichment enriched uranium reactor core of the nuclear thermal propulsion reactor, utilizes a reasonably established simplified heat transfer model, is suitable for different number distribution rules and arrangement rules of fuel assemblies and slowing assemblies in the reactor core, is also suitable for different distribution conditions of coolant mass flow in the two assemblies, can conveniently and flexibly carry out optimization design aiming at different assembly arrangement modes and flow distribution, and ensures the effectiveness of design parameters through constraint conditions of three aspects of aircraft performance constraint, thermodynamic hydraulic constraint and neutron physical constraint.
On the basis of the foregoing embodiments, the present embodiment provides an example of performing parameter design on a low-enrichment enriched uranium core by using the core parameter determination method for a nuclear thermal propulsion reactor, and the method may be specifically executed with reference to the following steps a to d:
a, step a: determining target power parameters meeting aircraft performance constraints in the nuclear thermal propulsion reactor, and determining value ranges of geometric parameters and thermal parameters in the nuclear thermal propulsion reactor.
The power parameters comprise parameters such as specific impulse, thrust and the like, the geometric parameters comprise height, radius, diameter of a coolant runner and the like of the reactor core, and the thermal parameters comprise outlet temperature of the reactor core and the like. Referring to a design flow chart of the low enrichment enriched uranium core reactor parameters shown in fig. 4, firstly, a target power parameter meeting aircraft constraint is determined according to the initial performance of an aircraft, and then, the value ranges of geometric parameters and thermal parameters in a nuclear thermal propulsion reactor are set.
Step b: and determining a fluid-solid coupling heat transfer model based on the flowing rule of the coolant in the nuclear thermal propulsion reactor.
The fluid-solid coupling heat transfer model meets the following conditions: 1) There are two different components within the core: the structure of the moderator assembly is relatively complex, and the materials and the structures of the two assemblies are shown in FIG. 2; 2) The heat transfer between the highest temperature point S1 of the fuel and the temperature point S2 of the inner wall surface of the second coolant flow channel 012 of the slowing component is flat-wall heat conduction without an internal heat source; 3) The coolant flow heat exchange from the third coolant flow passage 013 to the second coolant flow passage 012 and then to the first coolant flow passage 011 is simplified to from the second coolant flow passage 012 to the first coolant flow passage 011; 4) The core nuclear heat follows a cosine distribution in the axial direction.
Step c: and carrying out heat transfer calculation on the fluid-solid coupling heat transfer model to obtain a plurality of groups of design parameters meeting thermal engineering constraints.
The plurality of sets of design parameters (including geometric parameters and thermal parameters) can form a value range of the design parameters. The mainstream temperature, the wall surface temperature and the fuel center temperature of the corresponding coolant flow channel can be obtained through the heat transfer calculation, the thermotechnical convergence condition is set by utilizing the fuel limit temperature and the tolerance temperature of the moderator, and when the design parameters do not meet the thermotechnical convergence condition, the values of the geometric parameters and the thermotechnical parameters in the heat transfer calculation are updated according to the ranges of the geometric parameters and the thermotechnical parameters, so that multiple groups of design parameters meeting thermotechnical constraints are obtained.
Step d: and sequentially carrying out neutron modeling on each group of design parameters in the plurality of groups of design parameters, and carrying out physical analysis on the obtained model so as to judge whether each group of design parameters meets neutron physical constraints or not.
When the design parameters meeting the neutron physical constraints are obtained, the target thermal parameters and the target geometric parameters meeting the neutron physical constraints are used as reactor core design parameters, whether the reactor core design parameters (including the target power parameters, the target thermal parameters and the target geometric parameters) obtained by calculating the target power parameters meet requirements (or are satisfied with the reactor core design parameters) or not can be further judged, if the reactor core design parameters meet the requirements, the low-enrichment enriched uranium enriched reactor core design parameters are obtained, and if the reactor core design parameters do not meet the requirements, the initial target power parameters (thrust and specific impulse) can be modified in the step a to recalculate the design parameters until the satisfied reactor design parameters meeting the three-aspect constraint conditions are obtained.
Corresponding to the method for determining the core parameter of the nuclear heat propulsion reactor provided in the foregoing embodiment, an embodiment of the present invention provides a core parameter determining apparatus of a nuclear heat propulsion reactor, and referring to a schematic structural diagram of a core parameter determining apparatus of a nuclear heat propulsion reactor shown in fig. 5, the apparatus includes the following modules:
the obtaining module 51 is configured to obtain a fluid-solid coupling heat transfer model of the nuclear thermal propulsion reactor.
The calculation module 52 is configured to obtain a target power parameter that satisfies the aircraft performance constraint, and perform heat transfer calculation on the fluid-solid coupling heat transfer model based on the target power parameter to obtain multiple sets of design parameters that satisfy the thermal engineering constraint; wherein each set of design parameters includes geometric parameters and thermal parameters.
And the analysis module 53 is configured to perform neutron modeling and physical analysis based on multiple sets of design parameters to obtain target geometric parameters and target thermal parameters satisfying neutron physical constraints.
And the parameter determination module 54 is used for taking the target power parameter, the target thermal parameter and the target geometric parameter as the core design parameters of the nuclear thermal reactor.
According to the reactor core parameter determination device for the nuclear heat propulsion reactor, the heat transfer calculation is performed on the fluid-solid coupling heat transfer model of the nuclear heat propulsion reactor based on the target power parameters meeting the aircraft performance constraint, the neutron modeling and the physical analysis are performed, the reactor core design parameters meeting the aircraft performance constraint, the thermal constraint and the neutron physical constraint at the same time can be obtained, the effectiveness of the reactor core design parameters is guaranteed, the method can be applied to the determination of the low enrichment enriched uranium reactor core design parameters, and the reliability of the reactor core parameter design in the low enrichment enriched uranium nuclear heat propulsion reactor is improved.
In one embodiment, the geometric parameters include core height, core radius, and coolant flowpath diameter; the obtaining module 51 is further configured to determine a fuel assembly structure and a moderating assembly structure in the nuclear thermal reactor respectively based on the height of the core, the radius of the core, and the diameter of the coolant flow channel; and obtaining a fluid-solid coupling heat transfer model of the nuclear heat reactor based on the fuel assembly structure, the moderating assembly structure and the flow direction of the coolant.
In one embodiment, the fuel assembly structure and the moderator structure are each a prismatic structure, the height and cross-sectional shape of the fuel assembly structure and the moderator structure are the same, the fuel assembly structure comprises a plurality of coolant flow channels, the moderator structure comprises a second coolant flow channel, the coolant flow direction is from the second coolant flow channel to the first coolant flow channel, and the first coolant flow channel is each coolant flow channel in the fuel assembly structure.
In one embodiment, the thermal parameter comprises core outlet temperature; the calculating module 52 is further configured to perform axial heat transfer calculation and radial heat transfer calculation on the fluid-solid coupling heat transfer model based on the target power parameter, so as to obtain a wall temperature of the coolant flow channel, a fuel center temperature, a mainstream temperature at any axial position, a core outlet temperature, and a geometric parameter corresponding to the core outlet temperature; and repeatedly executing the axial heat transfer calculation and the radial heat transfer calculation until a first preset condition is reached to obtain a plurality of groups of design parameters.
In one embodiment, the first preset condition includes: the fuel core temperature reaches the highest fuel temperature, and/or the temperature of the moderator in the structure of the moderator reaches the highest moderator temperature.
In one embodiment, the geometric parameters include core height, core radius, and coolant flowpath diameter; the analysis module 53 is further configured to select any one set of design parameters from the plurality of sets of design parameters to obtain a first design parameter, and perform a neutron modeling based on a geometric parameter in the first design parameter to obtain a target model; carrying out reactor physical analysis on the target model to obtain physical parameters of the target model; and judging whether the physical parameters meet a second preset condition, and if so, respectively taking the geometric parameters and the thermal parameters in the first design parameters as target geometric parameters and target thermal parameters.
In one embodiment, the physical parameter comprises the effective core multiplication factor; the analysis module 53 is further configured to determine whether the effective multiplication coefficient is greater than 1, and if so, determine that the physical parameter satisfies a second preset condition.
In one embodiment, the apparatus further comprises:
the repeated analysis module is used for selecting any one group of design parameters except the first design parameter from the plurality of groups of design parameters to obtain a new first design parameter when the physical parameters do not meet a second preset condition; and performing neutron modeling and physical analysis based on the geometric parameters in the new first design parameters until target geometric parameters and target thermal parameters are obtained.
The reactor core parameter determining device of the nuclear thermal propulsion reactor provided by the embodiment provides a complete calculation thought for parameter design of the nuclear thermal propulsion reactor low-enrichment enriched uranium reactor core, utilizes a reasonably established simplified heat transfer model, is suitable for different number distribution rules and arrangement rules of fuel assemblies and moderating assemblies in the reactor core, is also suitable for different distribution conditions of coolant mass flow in two assemblies, can conveniently and flexibly carry out optimization design aiming at different assembly arrangement modes and flow distribution, and guarantees validity of design parameters through constraint conditions of three aspects of aircraft performance constraint, thermal hydraulic constraint and neutron physical constraint.
The device provided by the embodiment has the same implementation principle and technical effect as the foregoing embodiment, and for the sake of brief description, reference may be made to the corresponding contents in the foregoing method embodiment for the portion of the embodiment of the device that is not mentioned.
An embodiment of the present invention provides an electronic device, as shown in a schematic structural diagram of the electronic device shown in fig. 6, where the electronic device includes a processor 61 and a memory 62, where a computer program operable on the processor is stored in the memory, and when the processor executes the computer program, the steps of the method provided in the foregoing embodiment are implemented.
Referring to fig. 6, the electronic device further includes: a bus 64 and a communication interface 63, and the processor 61, the communication interface 63 and the memory 62 are connected by the bus 64. The processor 61 is for executing executable modules, such as computer programs, stored in the memory 62.
The Memory 62 may include a high-speed Random Access Memory (RAM) and may also include a non-volatile Memory (non-volatile Memory), such as at least one disk Memory. The communication connection between the network element of the system and at least one other network element is realized through at least one communication interface 63 (which may be wired or wireless), and the internet, a wide area network, a local network, a metropolitan area network, and the like can be used.
The bus 64 may be an ISA (Industry Standard Architecture) bus, a PCI (Peripheral Component Interconnect) bus, an EISA (Extended Industry Standard Architecture) bus, or the like. The bus may be divided into an address bus, a data bus, a control bus, etc. For ease of illustration, only one double-headed arrow is shown in FIG. 6, but that does not indicate only one bus or one type of bus.
The memory 62 is configured to store a program, and the processor 61 executes the program after receiving an execution instruction, and the method executed by the apparatus defined by the flow process disclosed in any of the foregoing embodiments of the present invention may be applied to the processor 61, or implemented by the processor 61.
The processor 61 may be an integrated circuit chip having signal processing capabilities. In implementation, the steps of the above method may be performed by integrated logic circuits of hardware or instructions in the form of software in the processor 61. The Processor 61 may be a general-purpose Processor, and includes a Central Processing Unit (CPU), a Network Processor (NP), and the like. The device can also be a Digital Signal Processor (DSP), an Application Specific Integrated Circuit (ASIC), a Field-Programmable Gate Array (FPGA), or other Programmable logic devices, discrete Gate or transistor logic devices, discrete hardware components. The various methods, steps, and logic blocks disclosed in the embodiments of the present invention may be implemented or performed. A general purpose processor may be a microprocessor or the processor may be any conventional processor or the like. The steps of the method disclosed in connection with the embodiments of the present invention may be directly implemented by a hardware decoding processor, or implemented by a combination of hardware and software modules in the decoding processor. The software module may be located in ram, flash memory, rom, prom, or eprom, registers, etc. storage media as is well known in the art. The storage medium is located in the memory 62, and the processor 61 reads the information in the memory 62, and combines the hardware thereof to complete the steps of the method.
Embodiments of the present invention provide a computer-readable medium, wherein the computer-readable medium stores computer-executable instructions, which, when invoked and executed by a processor, cause the processor to implement the method of the above-mentioned embodiments.
It can be clearly understood by those skilled in the art that, for convenience and brevity of description, the specific working process of the system described above may refer to the corresponding process in the foregoing embodiments, and is not described herein again.
The computer program product of the method and the apparatus for determining the core parameter of the nuclear thermal propulsion reactor provided by the embodiment of the present invention includes a computer readable storage medium storing program codes, where instructions included in the program codes may be used to execute the method described in the foregoing method embodiment, and specific implementation may refer to the method embodiment, and will not be described herein again.
In addition, in the description of the embodiments of the present invention, unless otherwise explicitly specified or limited, the terms "mounted," "connected," and "connected" are to be construed broadly, e.g., as meaning either a fixed connection, a removable connection, or an integral connection; can be mechanically or electrically connected; they may be connected directly or indirectly through intervening media, or they may be interconnected between two elements. The specific meanings of the above terms in the present invention can be understood in specific cases to those skilled in the art.
The functions, if implemented in the form of software functional units and sold or used as a stand-alone product, may be stored in a computer readable storage medium. Based on such understanding, the technical solution of the present invention may be embodied in the form of a software product, which is stored in a storage medium and includes instructions for causing a computer device (which may be a personal computer, a server, or a network device) to execute all or part of the steps of the method according to the embodiments of the present invention. And the aforementioned storage medium includes: a U-disk, a removable hard disk, a Read-Only Memory (ROM), a Random Access Memory (RAM), a magnetic disk, or an optical disk, and various media capable of storing program codes.
In the description of the present invention, it should be noted that the terms "center", "upper", "lower", "left", "right", "vertical", "horizontal", "inner", "outer", etc. indicate orientations or positional relationships based on the orientations or positional relationships shown in the drawings, and are only for convenience of description and simplification of description, but do not indicate or imply that the device or element referred to must have a specific orientation, be constructed and operated in a specific orientation, and thus, should not be construed as limiting the present invention. Furthermore, the terms "first," "second," and "third" are used for descriptive purposes only and are not to be construed as indicating or implying relative importance.
Finally, it should be noted that: the above-mentioned embodiments are only specific embodiments of the present invention, which are used for illustrating the technical solutions of the present invention and not for limiting the same, and the protection scope of the present invention is not limited thereto, although the present invention is described in detail with reference to the foregoing embodiments, those skilled in the art should understand that: those skilled in the art can still make modifications or changes to the embodiments described in the foregoing embodiments, or make equivalent substitutions for some features, within the scope of the disclosure; such modifications, changes or substitutions do not depart from the spirit and scope of the embodiments of the present invention, and they should be construed as being included therein. Therefore, the protection scope of the present invention shall be subject to the protection scope of the appended claims.
Claims (8)
1. The method for determining the reactor core parameters of the nuclear heat propulsion reactor is characterized by being applied to a low-enrichment-degree concentrated uranium nuclear heat propulsion reactor, and comprises the following steps:
acquiring a fluid-solid coupling heat transfer model of the nuclear thermal propulsion reactor;
acquiring target power parameters meeting aircraft performance constraints, and performing heat transfer calculation on the fluid-solid coupling heat transfer model based on the target power parameters to obtain multiple groups of design parameters meeting thermal engineering constraints; each group of design parameters comprises geometric parameters and thermal parameters;
performing neutron modeling and physical analysis based on the plurality of groups of design parameters to obtain target geometric parameters and target thermal parameters which meet neutron physical constraints;
taking the target dynamic parameter, the target thermal engineering parameter and the target geometric parameter as reactor core design parameters of the nuclear thermal reactor;
the thermal parameters comprise reactor core outlet temperature; the step of carrying out heat transfer calculation on the fluid-solid coupling heat transfer model based on the target dynamic parameter to obtain a plurality of groups of design parameters meeting thermal engineering constraints comprises the following steps:
performing axial heat transfer calculation and radial heat transfer calculation on the fluid-solid coupling heat transfer model based on the target power parameters to obtain the wall surface temperature and the fuel center temperature of a coolant flow channel, the mainstream temperature and the reactor core outlet temperature at any axial position and the geometric parameters corresponding to the reactor core outlet temperature;
repeatedly executing the axial heat transfer calculation and the radial heat transfer calculation until a first preset condition is reached to obtain a plurality of groups of design parameters;
the geometric parameters comprise the height of a reactor core, the radius of the reactor core and the diameter of a coolant flow channel;
the step of performing neutron modeling and physical analysis based on the plurality of sets of design parameters to obtain target geometric parameters and target thermal parameters meeting neutron physical constraints comprises the following steps:
selecting any one group of design parameters from the multiple groups of design parameters to obtain first design parameters, and performing neutron modeling based on the geometric parameters in the first design parameters to obtain a target model;
performing reactor physical analysis on the target model to obtain physical parameters of the target model;
and judging whether the physical parameters meet a second preset condition, and if so, respectively taking the geometric parameters and the thermal parameters in the first design parameters as target geometric parameters and target thermal parameters.
2. The method of claim 1, wherein the geometric parameters include core height, core radius, and coolant flowpath diameter;
the step of obtaining the fluid-solid coupling heat transfer model of the nuclear thermal propulsion reactor comprises the following steps:
respectively determining a fuel assembly structure and a moderating assembly structure in the nuclear heat reactor based on the core height, the core radius and the coolant runner diameter;
and obtaining a fluid-solid coupling heat transfer model of the nuclear heat reactor based on the fuel assembly structure, the moderating assembly structure and the flow direction of the coolant.
3. The method of claim 2, wherein the fuel assembly structure and the moderator structure are each prismatic structures, the fuel assembly structure and the moderator structure have the same height and cross-sectional shape, the fuel assembly structure includes a plurality of coolant flow channels, the moderator structure includes a second coolant flow channel, the coolant flows from the second coolant flow channel to a first coolant flow channel, the first coolant flow channel being each coolant flow channel in the fuel assembly structure.
4. The method according to claim 1, wherein the first preset condition comprises: the fuel core temperature reaches a maximum fuel temperature, and/or the temperature of the moderator in the structure of the moderator reaches a maximum moderator temperature.
5. The method of claim 1, wherein the physical parameters include core effective multiplication factor;
the step of judging whether the physical parameter meets a second preset condition includes:
and judging whether the effective multiplication coefficient is larger than 1, and if so, determining that the physical parameter meets the second preset condition.
6. The method of claim 1, further comprising:
if the physical parameters do not meet the second preset condition, selecting any one group of design parameters except the first design parameters from the multiple groups of design parameters to obtain new first design parameters;
and performing neutron modeling and physical analysis based on the geometric parameters in the new first design parameters until target geometric parameters and target thermal parameters are obtained.
7. The utility model provides a reactor core parameter determination device of nuclear heat propulsion reactor which characterized in that is applied to low enrichment concentration uranium nuclear heat propulsion reactor, reactor core parameter determination device of nuclear heat propulsion reactor includes:
the acquisition module is used for acquiring a fluid-solid coupling heat transfer model of the nuclear thermal propulsion reactor;
the calculation module is used for acquiring target power parameters meeting aircraft performance constraints and performing heat transfer calculation on the fluid-solid coupling heat transfer model based on the target power parameters to obtain a plurality of groups of design parameters meeting thermal engineering constraints; each group of design parameters comprises geometric parameters and thermal parameters;
the analysis module is used for performing neutron modeling and physical analysis based on the plurality of groups of design parameters to obtain target geometric parameters and target thermal parameters meeting neutron physical constraints;
the parameter determination module is used for taking the target dynamic parameter, the target thermotechnical parameter and the target geometric parameter as reactor core design parameters of the nuclear heat reactor;
the thermal parameters include core outlet temperature; the calculation module is used for performing axial heat transfer calculation and radial heat transfer calculation on the fluid-solid coupling heat transfer model based on the target power parameters to obtain the wall surface temperature of a coolant flow channel, the fuel center temperature, the main flow temperature at any axial position, the reactor core outlet temperature and the geometric parameters corresponding to the reactor core outlet temperature; repeatedly executing the axial heat transfer calculation and the radial heat transfer calculation until a first preset condition is reached to obtain a plurality of groups of design parameters;
the geometric parameters comprise the height of a reactor core, the radius of the reactor core and the diameter of a coolant flow channel; the analysis module is used for selecting any one group of design parameters from the plurality of groups of design parameters to obtain first design parameters, and performing neutron modeling based on the geometric parameters in the first design parameters to obtain a target model; performing reactor physical analysis on the target model to obtain physical parameters of the target model; and judging whether the physical parameters meet a second preset condition, and if so, respectively taking the geometric parameters and the thermal parameters in the first design parameters as target geometric parameters and target thermal parameters.
8. A computer-readable storage medium, on which a computer program is stored which, when being executed by a processor, carries out the steps of the method according to any one of the claims 1 to 6.
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| CN114694861A (en) * | 2022-03-28 | 2022-07-01 | 哈尔滨工程大学 | Nuclear reactor core parameter simulation method and device and electronic equipment |
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