CN110543705A - Boiling simulation solving acceleration method in typical channel of nuclear reactor - Google Patents
Boiling simulation solving acceleration method in typical channel of nuclear reactor Download PDFInfo
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Abstract
The invention discloses a boiling simulation solving acceleration method in a typical channel of a nuclear reactor, which comprises the following steps: channel modeling and space dispersion, control equation dispersion, initial field assignment, algebraic discrete equation solution, cavitation bubble share overrun judgment, pressure correction and calculation convergence judgment. On the basis of solving the boiling model of the two fluid models in the prior art, the cavitation bubble share overrun judgment and pressure correction process is added, in each iteration step, the average cavitation bubble share on each section in the channel is solved, if the average cavitation bubble share exceeds the limit value, the section pressure is calculated by using an empirical relational expression to obtain a new pressure difference, and then the solution of the pressure-velocity coupling equation is carried out again. Compared with the traditional Euler two-fluid equation solution, the method can correct the pressure distribution in the channel under the high bubble share in time, avoid the backflow phenomenon caused by a mismatched pressure-speed field, quickly achieve the physical real pressure distribution, accelerate the convergence, provide fine two-phase parameters for the reactor design, and further achieve the aims of reducing the safety margin and improving the reactor design economy.
Description
Technical Field
The invention relates to a computational fluid dynamics simulation technology in a typical channel of a nuclear reactor, in particular to a boiling simulation solution accelerated convergence method.
background
Reactor safety design and operation requires prediction of the heat transfer of flow in typical channels within the reactor system. The conventional design tool is mainly a reactor system program, is a one-dimensional simplified simulation of a reactor system channel, depends on a large number of experiments for prediction, has high uncertainty, and must introduce a large safety margin in the reactor design and sacrifice economy to ensure safety. In order to improve the economy of the nuclear reactor, three-dimensional fine simulation needs to be carried out on the flow in each component channel of the reactor system, so that the design safety margin is reduced. Among the parameters related to the safety margin, boiling phenomena such as void fraction, two-phase flow pressure drop, critical heat flux density, etc. are particularly important, and therefore, boiling simulation in typical channels of a reactor is of great importance for reactor design.
in a plurality of boiling gas-liquid two-phase flow simulation methods, an Euler two-fluid model is the best choice for simulating an industrial system, and because the two phases are respectively subjected to ensemble averaging, the calculated amount required by solving is smaller than that of a model directly simulating an interface; under the condition of using a proper interphase closed model, the calculation precision can reach the industrial requirement. However, in the process of solving the boiling phenomenon using the euler two-fluid model, because the density ratio between the gas phase and the liquid phase is large, for example, the density ratio between the liquid water and the gaseous water can reach 1600 times under normal pressure, the generation of a small amount of gas phase mass will cause a large proportion of cavitation bubbles, thereby affecting the stability of equation solution.
in the process of solving the two-fluid equation, due to the coupling relation of pressure and speed, the matching between the pressure field and the speed field is the key of the convergence of the solution. The commonly used pressure-velocity coupling method, such as the pressure-velocity semi-implicit coupling method, is solved iteratively by constructing a pressure-velocity correlation. However, when the euler two-fluid model is solved by using the method, especially under the physical working condition of high bubble fraction with boiling phenomenon, the liquid phase velocity is suddenly increased due to the sudden increase of the gas phase fraction, so that the pressure-velocity coupling relation is unbalanced, and at the moment, if the pressure-velocity relation is used for solving a velocity equation, an unreasonable backflow phenomenon can occur. Eliminating the influence of the reflux requires several times of iteration steps, thus greatly reducing the convergence rate of calculation; when the backflow condition is not properly set, the calculation is even diverged, and the calculation amount and the calculation time of the simulation are directly influenced.
Disclosure of Invention
the invention aims to use the known relation between boiling flow pressure drop and void fraction in a typical channel of a nuclear reactor to correct the pressure of solution of Euler two-fluid models in the boiling simulation process, avoid pressure and speed imbalance and generate backflow, so as to accelerate calculation convergence, provide a finer simulation result for reactor design, reduce safety margin and improve the economy of the reactor.
In order to achieve the purpose, the invention adopts the following technical scheme to realize the purpose:
A boiling simulation solving acceleration method in a typical channel of a nuclear reactor is characterized in that a high bubble share judgment process is added in the solving process of a normal Euler two-fluid boiling model. If a high-void fraction area appears in the channel in the iteration process, the pressure drop in the channel is predicted by using an empirical relation between the pressure drop of a boiling flow pipe in a typical macroscopic channel and the void fraction, and meanwhile, the pressure drop in the channel is compared with the pressure drop in the channel obtained by solving the current model to obtain a pressure difference value, the pressure difference value is used as a corrected pressure value and is introduced into a pressure correction equation to obtain a more reasonable pressure field, so that the adverse phenomena of velocity swell, pressure swell, outlet backflow and the like which influence iteration convergence after the velocity equation is solved are avoided.
the invention specifically comprises the following steps:
A boiling simulation solving acceleration method in a typical channel of a nuclear reactor comprises six steps of channel modeling and space dispersion, control equation dispersion, initial field assignment, algebraic discrete equation solution, cavitation bubble share overrun judgment and pressure correction, calculation convergence judgment and the like, and is characterized by comprising the following steps:
step 1: modeling and spatial discretization of typical channels in a nuclear reactor
Establishing a typical channel geometry of the nuclear reactor by using three-dimensional modeling software; then establishing a grid, and dividing the typical channel of the nuclear reactor into a plurality of control bodies, namely, the space is discrete;
Step 2: control equation dispersion
Writing an algebraic discrete equation which is satisfied by each control body in a typical channel of the nuclear reactor, wherein the fluid in each control body obtained in the step 1 satisfies two sets of continuity equations, momentum equations and energy equations, and respectively describes the conservation of mass, momentum and energy of gas-phase fluid and liquid-phase fluid; the respective forms thereof are as follows,
equation of momentum
wherein, the velocity of the liquid phase fluid in the J-th control fluid is the velocity of the gas phase fluid in the J-th control fluid; a1, J being the sum of the pressure-independent diffusion and convection terms to which the liquid phase fluid is subjected in the control body, a2, J being the sum of the pressure-independent diffusion and convection terms to which the gas phase fluid is subjected in the control body; b1, J is the pressure action coefficient of the liquid phase fluid in the control body, B2, J is the pressure action coefficient of the gas phase fluid in the control body; nconn is the number of adjacent control volumes to the jth control volume; DPn is the pressure difference between the fluid in the jth control volume and the fluid in the nth control volume adjacent thereto;
Unified writing of continuity equations and energy equations
In formula (III), rij is a coefficient of the equation, where i ═ 1.., 4, j ═ 1.., nconn; ai is a constant term of the equation, where i ═ 1.., 4; δ α v is a correction value for the void fraction in the jth control body; δ hv is the corrected value of the vapor phase enthalpy value in the jth control gas; δ hl is a corrected value of the enthalpy of the liquid phase in the jth control body; δ PJ is a correction value of the fluid pressure in the jth control body; δ P1, δ P2, …, δ Pnconn are correction values of the fluid pressure in the 1 st, 2 nd, …, nconn th control body adjacent to the J th control body; the fourth line constitutes a pressure correction equation,
Where r44 is a coefficient of the correction pressure of the J-th control body, r4(4+ n) is a coefficient of the correction pressure of the n-th control body adjacent to the J-th control body, and these coefficients mean: a relationship between the corrected pressures obtained from the continuity equation; δ PJ is a correction value of the fluid pressure in the jth control body, δ Pn is a correction value of the fluid pressure in the nth control body adjacent to the jth control body; a4 is an equation constant term;
and step 3: initialization of flow in a typical channel of a nuclear reactor
Setting a convergence threshold value for all control bodies;
and 4, step 4: solution of algebraic discrete equations
solving equations (I), (II) and (IV) in sequence, and then judging whether the pressure-speed meets the convergence requirement; if the corrected value of the fluid pressure in any control body is larger than the pressure convergence threshold epsilon p, correcting the pressure in all the control bodies, returning to the momentum equations (I) and (II) to solve the gas phase and liquid phase fluid velocity in the control bodies again; otherwise, considering that the pressure meets the convergence requirement, and at the moment, sequentially substituting the correction value delta PJ of the fluid pressure in the J-th control body into the third line, the second line and the first line of the formula (III) to sequentially obtain the liquid phase enthalpy value hl, the gas phase enthalpy value hv and the void fraction alpha v;
and 5: determination of excess of void fraction and pressure correction
Step 5-1: for a control body positioned on the same cross section in a typical channel of the nuclear reactor, calculating the average value of the void fraction on each cross section in the typical channel of the nuclear reactor by the volume average of a formula (V)
where α i is the fraction of vacuoles in the ith control volume and Vi is the volume of the ith control volume; calculating a void fraction average value for a control body on each cross section in a typical channel of the nuclear reactor;
Step 5-2: judging whether the average value of the void fraction on each cross section in the step 5-1 is higher than a threshold value 0.3 for stabilizing the calculation: if the average value of the void fraction is higher than the threshold value, performing pressure correction on the control body on the section, wherein the section is called an overrun section, and entering the step 5-3; otherwise, the pressure is not required to be additionally corrected, and the step 6 is carried out;
Step 5-3: calculating the pressure Psup on the overrun section by using an empirical relation of the pressure drop and the void fraction in a typical channel of the nuclear reactor;
Step 5-4: calculating the pressure difference of the two adjacent layers after correction, namely the pressure difference delta P of the adjacent control bodies in the main flow direction in the formula (I)
ΔP＝P-P (VI)
Wherein Psup, I and Psup, I +1 are respectively the pressures on the overrun cross sections of the layer I and the layer I +1 obtained in the step 5-3;
Step 5-5: in the same step 4, solving equations (I), (II), (III) and (IV) by using the pressure difference delta P of adjacent control bodies in the main flow direction, and recalculating the gas phase and liquid phase speeds, the void fraction and the gas phase and liquid phase enthalpy values in the typical channel of the nuclear reactor;
Step 6: computing convergence judgment
If the corrected value delta alpha v of the void fraction in the J-th control body, the corrected value delta hv of the gas phase enthalpy value in the J-th control body and the corrected value delta hl of the liquid phase enthalpy value in the J-th control body, which are obtained by the formula (III), are all smaller than the set threshold values epsilon alpha, epsilon hv and epsilon hl, the calculation is considered to be converged;
at the moment, the physical quantities of two-phase enthalpy value distribution, the distribution of bubble shares in the channel and the two-phase speed distribution under the current working condition are obtained, and reference is provided for the design of the nuclear reactor.
Compared with the traditional two-fluid model solving method, the method has the advantages that:
1) In the method, the judgment of the cavitation share overrun is added after each discrete equation solution, and when the cavitation share overrun exceeds the limit value, the pressure is corrected, so that further backflow is avoided, the calculation condition is further deteriorated, and convergence is influenced;
2) The invention changes the original two-fluid model solving method only by increasing pressure correction, and can be conveniently added into the existing solving code to realize the function of accelerating convergence;
3) The corrected value adopts an empirical pressure relation, and the empirical relations of pressure, vacuole share and apparent flow velocity of each phase obtained by a large number of experiments are utilized, so that the pressure value quickly reaches the physical and actual range, and the convergence is accelerated;
4) The method is convenient and practical, can quickly and accurately obtain the result of boiling simulation, realizes the simulation of the boiling phenomenon in the reactor component, obtains fine two-phase parameter distribution, and reduces the uncertainty of calculation, thereby reducing the safety margin of design and improving the economy of the reactor;
5) Because the method accelerates convergence, the method has great advantages when being applied to the reactor system multi-physical field coupling simulation, provides a thermal calculation basis for mechanical thermal physical coupling, further improves the calculation precision of the whole reactor design process, plays a role in replacing part of tests, and reduces the design time consumption.
Drawings
FIG. 1 is a flow chart of the method of the present invention.
FIG. 2 is a schematic diagram of a typical coolant channel within a fuel assembly of a nuclear reactor core in an axial ratio of 1: 10.
FIG. 3 is a schematic view of a control body after spatial discretization of typical coolant channels.
FIG. 4 is a schematic view of a control body at a cross-section within an exemplary coolant channel.
Detailed Description
the present invention will be described in further detail below with reference to the flowchart of fig. 1, taking an example of solving a typical single channel of a pressurized water reactor fuel assembly by a typical pressure-velocity coupling solution.
A boiling simulation solving acceleration method in a typical channel of a nuclear reactor comprises six steps of channel modeling and space dispersion, control equation dispersion, initial field assignment, algebraic discrete equation solving, cavitation bubble share overrun judgment and pressure correction, calculation convergence judgment and the like.
Step 1: modeling and spatial discretization of typical channels in a nuclear reactor
Taking a typical single channel of a pressurized water reactor fuel assembly as an example, a quadrilateral-like channel geometry is established by using three-dimensional modeling software, as shown in FIG. 2; a structured hexahedral mesh is then established, dividing the typical channels of the nuclear reactor into several control volumes, i.e. spatially discrete, as shown in fig. 3.
Step 2: control equation dispersion
Writing an algebraic discrete equation satisfied by each control body in a typical channel of the nuclear reactor; the fluid in each control body obtained in the step 1 meets two sets of continuity equations, momentum equations and energy equations, and the mass, momentum and energy conservation of the gas-phase fluid and the liquid-phase fluid are respectively described; the respective forms thereof are as follows,
Equation of momentum
Wherein, the velocity of the liquid phase fluid in the J-th control fluid is the velocity of the gas phase fluid in the J-th control fluid; a1, J being the sum of the pressure-independent diffusion and convection terms to which the liquid phase fluid is subjected in the control body, a2, J being the sum of the pressure-independent diffusion and convection terms to which the gas phase fluid is subjected in the control body; b1, J is the pressure action coefficient of the liquid phase fluid in the control body, B2, J is the pressure action coefficient of the gas phase fluid in the control body; nconn is the number of adjacent control volumes to the jth control volume; Δ Pn is the pressure difference between the fluid in the jth control volume and the fluid in the nth control volume adjacent thereto.
Equation of continuity
C[α,h,h,P,P,P,...,P]＝0 (3)
C[α,h,h,P,P,P,...,P]＝0 (4)
wherein Cl represents a continuity equation of a liquid phase fluid in the control fluid, and Cv represents a continuity equation of a gas phase fluid in the control fluid; α v represents the void fraction in the control body and represents the volume fraction of the gas phase fluid in the total volume of the control body; hv is the specific enthalpy of the gas phase fluid, hl is the specific enthalpy of the liquid phase fluid; PJ is the pressure of the fluid in the jth control body, P1, P2.., Pnconn denotes the pressure of the fluid in nconn control bodies adjacent to the jth control body, respectively.
Equation of energy
E[α,h,h,P,P,P,...,P]＝0 (5)
E[α,h,h,P,P,P,...,P]＝0 (6)
Where El represents the energy equation for controlling the liquid phase fluid in the fluid and Ev represents the energy equation for controlling the gas phase fluid in the fluid.
the expressions (3) to (6) can be written as vector forms
The continuity equation and the energy equation in the form of column vectors are fluid parameters in the solved control body, including a void fraction, a gas phase enthalpy value, a liquid phase enthalpy value, the pressure of the fluid in the J-th control body and the pressure of the fluid in the nconn control bodies adjacent to the J-th control body. The remaining symbols in the above formula have the same meanings as those in the formulae (3) to (6). Using Newton's iteration method, the iteration format of equation (7) can be constructed
wherein the corrected value of the fluid parameter in the control body to be solved is the derivative of the field variable and has the following form
By Gaussian elimination, equation (8) can be transformed into an upper triangular matrix form as follows
In equation (10), rij is a coefficient of the equation, where i ═ 1.., 4, j ═ 1.. once, nconn; ai is a constant term of the equation, where i ═ 1.., 4; the remaining variables have the same meaning as in formula (8). Equation (10) each line constitutes a separate equation, wherein the fourth line constitutes a pressure correction equation,
where r44 is a coefficient of the correction pressure of the J-th control body, r4(4+ n) is a coefficient of the correction pressure of the n-th control body adjacent to the J-th control body, and these coefficients mean: a relationship between the corrected pressures obtained from the continuity equation; δ PJ is a correction value of the fluid pressure in the jth control body, δ Pn is a correction value of the fluid pressure in the nth control body adjacent to the jth control body; a4 is the equation constant term.
And step 3: initialization of flow in a typical channel of a nuclear reactor
each control body in a typical channel of a nuclear reactor is assigned a calculation initial value. In this example, the fluid pressure in the control body is set to a hydrostatic pressure value calculated from the height in the channel, and the initial values of the other physical quantities are set to values at the inlet of the channel.
A convergence threshold is set for all the control volumes. In this example, the criterion for calculation convergence is set as: and the convergence thresholds Epsilon, Epsilon alpha, Epsilon hv and Epsilon hl are respectively set by controlling the pressure, the void fraction, the gas phase ratio enthalpy value and the liquid phase ratio enthalpy value in the body.
and 4, step 4: solution of algebraic discrete equations
In this example, a pressure velocity semi-implicit coupling method well known in the art is selected for solving.
Step 4-1: for all the control bodies obtained in step 1, respective momentum equations are calculated. Specifically, for the jth control body, using the pressure value of the fluid in the jth control body, the pressure difference (Δ p) n between the jth control body and the fluid in the nth control body adjacent thereto can be obtained, thereby calculating equations (1) (2) to obtain the velocity sum of the fluids of the respective phases in the jth control body
Step 4-2: the coefficients and constant terms in the pressure correction equation (11) are updated by equations (7) to (10) with the flow velocity of each phase and the equations (3) to (6) in the jth control body obtained in step 4-1. Then, the pressure correction equations (11) for all the control bodies are solved simultaneously, and the pressure correction values δ PJ of the fluids in all the control bodies can be obtained.
Step 4-3: if the corrected value of the fluid pressure in any control body is larger than the pressure convergence threshold epsilon p, correcting the pressure in all the control bodies, returning to the momentum equations (1) and (2) to solve the gas phase and liquid phase fluid velocity in the control bodies again; and otherwise, considering that the pressure meets the convergence requirement, and at the moment, sequentially substituting the correction value delta PJ of the fluid pressure in the J-th control body into the third line, the second line and the first line of the formula (10) to sequentially obtain the liquid phase enthalpy value hl, the gas phase enthalpy value hv and the void fraction alpha v.
And 5: determination of excess of void fraction and pressure correction
Step 5-1: for a control body which is positioned on the same cross section in a typical channel of the nuclear reactor and is shown in figure 4, calculating the average value of the void fraction on each cross section in the typical channel of the nuclear reactor by the volume average of the formula (12)
Where α i is the fraction of vacuoles in the ith control volume and Vi is the volume of the ith control volume; the average value of the void fraction is calculated for the control body on each cross section in the typical channel of the nuclear reactor.
Step 5-2: and (4) judging whether the average value of the void fraction on each cross section in the step 5-1 is higher than a threshold value 0.3 for stabilizing the calculation, if so, performing pressure correction on the control body on the cross section, wherein the cross section is called an overrun cross section, and entering the step 5-3. Otherwise, the pressure is not modified additionally and step 6 is entered.
Step 5-3: the pressure on the overrun cross section is calculated using an empirical relationship between the pressure drop in the typical channel of the nuclear reactor and the void fraction. In this example, the pressure drop is calculated using the Chisolm relationship
Wherein Δ PChisolm is the pressure difference between the channel inlet and the overrun cross-section; Δ z is the distance between the channel inlet and the overrun cross-section in the primary flow direction; the pressure drop coefficient is obtained by using a Chisolm relational expression according to the average flow velocity in the channel and the section void fraction alpha; u is the channel cross-sectional perimeter; a is the channel cross-sectional area; is the average density of the channel inlet fluid; is the average velocity of the fluid at the entrance of the channel.
The pressure Psup on the overrun cross-section can then be obtained
P＝P+ΔP (14)
Where Pin is the pressure at the channel inlet and Δ PChisolm is the pressure difference between the channel inlet and the overrun cross-section as determined by equation (3).
Step 5-4: calculating the pressure difference of the two adjacent layers after correction, namely the pressure difference delta P of the adjacent control bodies in the main flow direction in the formula (1)
ΔP＝P-P (15)
wherein Psup, I and Psup, I +1 are the pressures on the overrun cross-sections of the layer I and the layer I-1 obtained in the step 5-3 respectively.
step 5-5: in the same step 4, the pressure difference delta P of the adjacent control bodies in the main flow direction is used for solving equations (1) (2) (10) (11), and the gas phase and liquid phase speeds, the void fraction and the gas phase and liquid phase enthalpy values in the typical channel of the nuclear reactor are recalculated.
Step 6: computing convergence judgment
If the correction value δ α v of the void fraction in the J-th control body, the correction value δ hv of the vapor phase enthalpy value in the J-th control body, and the correction value δ hl of the liquid phase enthalpy value in the J-th control body, which are obtained by equation (10), are all smaller than the set threshold values ε a, ε hv, and ε hl, the calculation is considered to be converged.
at the moment, physical quantities such as two-phase enthalpy value distribution, air bubble share distribution in the channel, two-phase speed distribution and the like under the current working condition are obtained, and reference is provided for the design of the fuel assembly.
The above description is a specific example of the present invention, but they are not intended to limit the present invention, and those skilled in the art can make various modifications and equivalent changes without departing from the spirit and scope of the present invention, and still fall into the protection scope of the technical solution of the present invention.
Claims (1)
1. a boiling simulation solution acceleration method in a typical channel of a nuclear reactor is characterized by comprising the following steps:
step 1: modeling and spatial discretization of typical channels in a nuclear reactor
Establishing a typical channel geometry of the nuclear reactor by using three-dimensional modeling software; then establishing a grid, and dividing the typical channel of the nuclear reactor into a plurality of control bodies, namely, the space is discrete;
Step 2: control equation dispersion
writing an algebraic discrete equation which is satisfied by each control body in a typical channel of the nuclear reactor, wherein the fluid in each control body obtained in the step 1 satisfies two sets of continuity equations, momentum equations and energy equations, and respectively describes the conservation of mass, momentum and energy of gas-phase fluid and liquid-phase fluid; the respective forms thereof are as follows,
Equation of momentum
Wherein, the velocity of the liquid phase fluid in the J-th control fluid is the velocity of the gas phase fluid in the J-th control fluid; a1, J being the sum of the pressure-independent diffusion and convection terms to which the liquid phase fluid is subjected in the control body, a2, J being the sum of the pressure-independent diffusion and convection terms to which the gas phase fluid is subjected in the control body; b1, J is the pressure action coefficient of the liquid phase fluid in the control body, B2, J is the pressure action coefficient of the gas phase fluid in the control body; nconn is the number of adjacent control volumes to the jth control volume; Δ Pn is the pressure difference between the fluid in the jth control volume and the fluid in the nth control volume adjacent thereto;
unified writing of continuity equations and energy equations
In formula (III), rij is a coefficient of the equation, where i ═ 1.., 4, j ═ 1.., nconn; ai is a constant term of the equation, where i ═ 1.., 4; δ α v is a correction value for the void fraction in the jth control body; δ hv is the corrected value of the vapor phase enthalpy value in the jth control gas; δ hl is a corrected value of the enthalpy of the liquid phase in the jth control body; δ PJ is a correction value of the fluid pressure in the jth control body; δ P1, δ P2, …, δ Pnconn are correction values of the fluid pressure in the 1 st, 2 nd, …, nconn th control body adjacent to the J th control body; the fourth line constitutes a pressure correction equation,
Where r44 is a coefficient of the correction pressure of the J-th control body, r4(4+ n) is a coefficient of the correction pressure of the n-th control body adjacent to the J-th control body, and these coefficients mean: a relationship between the corrected pressures obtained from the continuity equation; δ PJ is a correction value of the fluid pressure in the jth control body, δ Pn is a correction value of the fluid pressure in the nth control body adjacent to the jth control body; a4 is an equation constant term;
and step 3: initialization of flow in a typical channel of a nuclear reactor
Setting a convergence threshold value for all control bodies;
And 4, step 4: solution of algebraic discrete equations
Solving equations (I), (II) and (IV) in sequence, and then judging whether the pressure-speed meets the convergence requirement; if the corrected value of the fluid pressure in any control body is larger than the pressure convergence threshold epsilon p, correcting the pressure in all the control bodies, returning to the momentum equations (I) and (II) to solve the gas phase and liquid phase fluid velocity in the control bodies again; otherwise, considering that the pressure meets the convergence requirement, and at the moment, sequentially substituting the correction value delta PJ of the fluid pressure in the J-th control body into the third line, the second line and the first line of the formula (III) to sequentially obtain the liquid phase enthalpy value hl, the gas phase enthalpy value hv and the void fraction alpha v;
and 5: determination of excess of void fraction and pressure correction
Step 5-1: for a control body positioned on the same cross section in a typical channel of the nuclear reactor, calculating the average value of the void fraction on each cross section in the typical channel of the nuclear reactor by the volume average of a formula (V)
Where α i is the fraction of vacuoles in the ith control volume and Vi is the volume of the ith control volume; calculating a void fraction average value for a control body on each cross section in a typical channel of the nuclear reactor;
step 5-2: judging whether the average value of the void fraction on each cross section in the step 5-1 is higher than a threshold value 0.3 for stabilizing the calculation: if the average value of the void fraction is higher than the threshold value, performing pressure correction on the control body on the section, wherein the section is called an overrun section, and entering the step 5-3; otherwise, the pressure is not required to be additionally corrected, and the step 6 is carried out;
step 5-3: calculating the pressure Psup on the overrun section by using an empirical relation of the pressure drop and the void fraction in a typical channel of the nuclear reactor;
step 5-4: calculating the pressure difference of the two adjacent layers after correction, namely the pressure difference delta P of the adjacent control bodies in the main flow direction in the formula (I)
ΔP＝P-P (VI)
Wherein Psup, I and Psup, I +1 are respectively the pressures on the overrun cross sections of the layer I and the layer I +1 obtained in the step 5-3;
Step 5-5: in the same step 4, solving equations (I), (II), (III) and (IV) by using the pressure difference delta P of adjacent control bodies in the main flow direction, and recalculating the gas phase and liquid phase speeds, the void fraction and the gas phase and liquid phase enthalpy values in the typical channel of the nuclear reactor;
Step 6: computing convergence judgment
if the corrected value delta alpha v of the void fraction in the J-th control body, the corrected value delta hv of the gas phase enthalpy value in the J-th control body and the corrected value delta hl of the liquid phase enthalpy value in the J-th control body, which are obtained by the formula (III), are all smaller than the set threshold values epsilon alpha, epsilon hv and epsilon hl, the calculation is considered to be converged;
at the moment, the physical quantities of two-phase enthalpy value distribution, the distribution of bubble shares in the channel and the two-phase speed distribution under the current working condition are obtained, and reference is provided for the design of the nuclear reactor.
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| CN113792449A (en) * | 2021-08-10 | 2021-12-14 | 中国船舶重工集团公司第七一九研究所 | Nuclear reactor monitoring method and system |
| CN114757122A (en) * | 2022-04-19 | 2022-07-15 | 西安交通大学 | Method for establishing fine thermal hydraulic calculation model for sodium-cooled fast reactor core disintegration accident |
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| CN112613240A (en) * | 2020-11-26 | 2021-04-06 | 中国核电工程有限公司 | Calculation method for flow analysis in containment vessel under serious accident |
| CN112613240B (en) * | 2020-11-26 | 2024-07-12 | 中国核电工程有限公司 | Calculation method for flow analysis of containment vessel under serious accident |
| CN113792449A (en) * | 2021-08-10 | 2021-12-14 | 中国船舶重工集团公司第七一九研究所 | Nuclear reactor monitoring method and system |
| CN113792449B (en) * | 2021-08-10 | 2023-12-19 | 中国船舶重工集团公司第七一九研究所 | Nuclear reactor monitoring method and system |
| CN114757122A (en) * | 2022-04-19 | 2022-07-15 | 西安交通大学 | Method for establishing fine thermal hydraulic calculation model for sodium-cooled fast reactor core disintegration accident |
| CN114757122B (en) * | 2022-04-19 | 2024-02-23 | 西安交通大学 | Method for establishing fine thermodynamic hydraulic calculation model of sodium-cooled fast reactor core disassembly accident |
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