CN105653869B - A kind of supercritical water reactor reactor core Three dimensional transient method for analyzing performance - Google Patents
A kind of supercritical water reactor reactor core Three dimensional transient method for analyzing performance Download PDFInfo
- Publication number
- CN105653869B CN105653869B CN201610003796.5A CN201610003796A CN105653869B CN 105653869 B CN105653869 B CN 105653869B CN 201610003796 A CN201610003796 A CN 201610003796A CN 105653869 B CN105653869 B CN 105653869B
- Authority
- CN
- China
- Prior art keywords
- thermal
- transient
- reactor core
- dimensional
- calculation
- Prior art date
- Legal status (The legal status is an assumption and is not a legal conclusion. Google has not performed a legal analysis and makes no representation as to the accuracy of the status listed.)
- Active
Links
- 230000001052 transient effect Effects 0.000 title claims abstract description 150
- 238000000034 method Methods 0.000 title claims abstract description 101
- XLYOFNOQVPJJNP-UHFFFAOYSA-N water Substances O XLYOFNOQVPJJNP-UHFFFAOYSA-N 0.000 title claims abstract description 89
- 230000008878 coupling Effects 0.000 claims abstract description 88
- 238000010168 coupling process Methods 0.000 claims abstract description 88
- 238000005859 coupling reaction Methods 0.000 claims abstract description 88
- 230000008569 process Effects 0.000 claims abstract description 59
- 238000004458 analytical method Methods 0.000 claims abstract description 46
- 238000011156 evaluation Methods 0.000 claims abstract description 21
- 238000004364 calculation method Methods 0.000 claims description 125
- 230000004907 flux Effects 0.000 claims description 35
- 239000000446 fuel Substances 0.000 claims description 19
- 238000005253 cladding Methods 0.000 claims description 13
- 230000006870 function Effects 0.000 claims description 7
- 230000008859 change Effects 0.000 claims description 5
- 239000008188 pellet Substances 0.000 claims description 5
- 230000000903 blocking effect Effects 0.000 claims description 3
- 238000011438 discrete method Methods 0.000 claims 1
- 230000000694 effects Effects 0.000 abstract description 11
- 238000004088 simulation Methods 0.000 abstract 1
- 230000002159 abnormal effect Effects 0.000 description 14
- 230000009257 reactivity Effects 0.000 description 11
- 239000002243 precursor Substances 0.000 description 7
- 239000002826 coolant Substances 0.000 description 6
- 239000011159 matrix material Substances 0.000 description 6
- 238000012512 characterization method Methods 0.000 description 5
- 230000003111 delayed effect Effects 0.000 description 5
- 238000013461 design Methods 0.000 description 5
- 238000012986 modification Methods 0.000 description 5
- 230000004048 modification Effects 0.000 description 5
- 238000004613 tight binding model Methods 0.000 description 5
- 238000010586 diagram Methods 0.000 description 4
- 239000006185 dispersion Substances 0.000 description 4
- 230000004992 fission Effects 0.000 description 4
- 230000004044 response Effects 0.000 description 4
- 230000008901 benefit Effects 0.000 description 3
- 206010003671 Atrioventricular Block Diseases 0.000 description 2
- 230000009471 action Effects 0.000 description 2
- 230000000712 assembly Effects 0.000 description 2
- 238000000429 assembly Methods 0.000 description 2
- 230000005540 biological transmission Effects 0.000 description 2
- 238000011161 development Methods 0.000 description 2
- 238000009792 diffusion process Methods 0.000 description 2
- 238000011160 research Methods 0.000 description 2
- 238000001228 spectrum Methods 0.000 description 2
- NAWXUBYGYWOOIX-SFHVURJKSA-N (2s)-2-[[4-[2-(2,4-diaminoquinazolin-6-yl)ethyl]benzoyl]amino]-4-methylidenepentanedioic acid Chemical compound C1=CC2=NC(N)=NC(N)=C2C=C1CCC1=CC=C(C(=O)N[C@@H](CC(=C)C(O)=O)C(O)=O)C=C1 NAWXUBYGYWOOIX-SFHVURJKSA-N 0.000 description 1
- 230000004075 alteration Effects 0.000 description 1
- 230000001808 coupling effect Effects 0.000 description 1
- 238000000354 decomposition reaction Methods 0.000 description 1
- 238000005516 engineering process Methods 0.000 description 1
- 239000003758 nuclear fuel Substances 0.000 description 1
- 230000009897 systematic effect Effects 0.000 description 1
Classifications
-
- G—PHYSICS
- G16—INFORMATION AND COMMUNICATION TECHNOLOGY [ICT] SPECIALLY ADAPTED FOR SPECIFIC APPLICATION FIELDS
- G16Z—INFORMATION AND COMMUNICATION TECHNOLOGY [ICT] SPECIALLY ADAPTED FOR SPECIFIC APPLICATION FIELDS, NOT OTHERWISE PROVIDED FOR
- G16Z99/00—Subject matter not provided for in other main groups of this subclass
Landscapes
- Monitoring And Testing Of Nuclear Reactors (AREA)
Abstract
The invention discloses a kind of supercritical water reactor reactor core Three dimensional transient method for analyzing performance to include：Step 1：It executes reactor core stable state to calculate, power parameter, thermal parameter is provided, execute iteration coupling and calculate until reactor core steady state power parameter and thermal parameter are restrained；Step 2：Reactor core original state and component cross-section library are provided；Step 3：It executes transient state to calculate, power parameter, thermal parameter is provided, execute iteration coupling and calculate until reactor core transient power parameter and thermal parameter are restrained；Step 4：The fine power distribution of hot component lattice cell scale is obtained, hot component subchannel analysis is carried out, finally provides safety evaluation key parameter；Step 5：Judge whether the safety evaluation key parameter is more than corresponding transient state or fail- safe limit value, realize feedback that can be between accurate description physics and thermal-hydraulic, more accurate reactor core three-dimensional power distribution, the technique effect of the transient process and accident process of true simulation supercritical water reactor are provided.
Description
Technical Field
The invention relates to the field of nuclear reactor design research, in particular to a method for analyzing three-dimensional transient performance of a supercritical water reactor core.
Background
The operating condition of a supercritical water reactor (SCWR) is above the thermodynamic critical point (374 ℃, 22.1MPa) of water, the system thermal efficiency is high, the economy is good, but simultaneously, compared with a conventional water-cooled reactor, the operating parameters of the supercritical water reactor are greatly improved, the reactor core structure and the coolant flow path are relatively complex, and a large number of technical problems of reactor core physics, thermal engineering-hydraulic power, safety analysis and the like are brought, wherein the strong nuclear thermal coupling characteristic is one of the key technical problems of the design and performance analysis of the reactor core of the supercritical water reactor.
In the reactivity abnormal events of the supercritical water reactor, most of the reactivity abnormal events are related to control rod actions, the control rod action abnormal events cause the power distribution distortion of the supercritical water reactor core, along with strong physical and thermal-hydraulic coupling effects, the power distribution change directly influences the maximum cladding temperature of the supercritical water reactor fuel, the cladding temperature is not over limit, which is a very important transient safety analysis criterion of the supercritical water reactor, for the power distribution abnormal events, a point reactor or a one-dimensional neutron dynamics model cannot correctly describe the spatial distribution of power, the change of the power along with the time cannot be accurately calculated, in order to ensure the calculated envelopment, a large number of conservative assumptions must be introduced, in addition, the supercritical water reactor has strong nuclear thermal coupling characteristics, the transient process of the reactivity and power distribution abnormal events is more complex, and the conservative method cannot provide the real response process, therefore, designers are difficult to deeply research and analyze the reactivity abnormal event, and can accurately describe the feedback between physics and thermal engineering-hydraulic engineering only by coupling the three-dimensional neutron space-time dynamics with the thermal engineering-hydraulic engineering to analyze the transient process, provide more accurate three-dimensional power distribution and truly simulate the transient process and the accident process of the supercritical water reactor.
In the prior art, for the power distribution abnormal event of the supercritical water reactor, a transient analysis method is internationally adopted, wherein a point reactor or a one-dimensional model is adopted in a sub-dynamic part, the three-dimensional power calculation capability of the nuclear thermal coupling of the reactor core is not provided, and the transient process of the power distribution abnormal event such as reactivity is difficult to truly or accurately simulate.
In summary, in the process of implementing the technical solution of the embodiments of the present application, the inventors of the present application find that the above-mentioned technology has at least the following technical problems:
in the prior art, the existing supercritical water reactor power distribution transient analysis method has the technical problems that the existing supercritical water reactor power distribution transient analysis method does not have the core-core thermal coupling three-dimensional power calculation capability and is difficult to truly or accurately simulate the transient process such as reactivity and power distribution abnormal events.
Disclosure of Invention
The invention provides a supercritical water reactor core three-dimensional transient performance analysis method, which solves the technical problems that the existing supercritical water reactor power distribution transient analysis method does not have the core thermal coupling three-dimensional power calculation capability and is difficult to truly or accurately simulate the transient process of an abnormal event such as reactivity and power distribution, realizes the technical effects of accurately describing the feedback between physics and thermal-hydraulic power, providing more accurate reactor core three-dimensional power distribution and truly simulating the transient process and the accident process of a supercritical water reactor.
In order to solve the technical problem, an embodiment of the present application provides a method for analyzing three-dimensional transient performance of a supercritical water reactor core, where the method includes:
step 1: the supercritical water reactor core three-dimensional stable physical-thermal hydraulic coupling module executes core stable calculation, the core fuel management module provides power parameters for the sub-channel thermal-hydraulic module, the sub-channel thermal-hydraulic module provides thermal parameters for the core fuel management module, and iterative coupling calculation is executed until the core stable power parameters and the thermal parameters are converged;
step 2: the reactor core three-dimensional steady-state physical-thermal hydraulic coupling module provides a reactor core initial state and a component section library for the supercritical water reactor core three-dimensional transient physical-thermal hydraulic coupling module;
and step 3: the three-dimensional transient physical-thermal hydraulic coupling module of the supercritical water reactor core executes transient calculation, the three-dimensional neutron space-time dynamics module provides power parameters for the sub-channel thermal-hydraulic module, the sub-channel thermal-hydraulic module provides thermal parameters for the three-dimensional neutron space-time dynamics module, and iterative coupling calculation is executed until the transient power parameters and the thermal parameters of the reactor core converge;
and 4, step 4: performing power reconstruction on the thermal assembly in reactor core calculation to obtain fine power distribution of the thermal assembly grid cell scale, performing thermal assembly subchannel analysis by using the subchannel thermal-hydraulic module again, and finally giving out safety evaluation key parameters;
and 5: and judging whether the safety evaluation key parameters exceed corresponding transient or accident safety limit values or not, and evaluating the safety performance of the supercritical water reactor core in the transient process.
Further, the safety evaluation key parameters include: maximum cladding wall temperature, pellet enthalpy.
Further, fewer groups of cross sections are used in the reactor core three-dimensional neutron science calculation, the few groups of cross sections are preprocessed into a cross section library for the reactor core three-dimensional neutron science module in a segmented interpolation mode according to the change of thermal parameters and preset parameters, and the preset parameters comprise: the fuel assembly is burnable with or without control rods.
Further, the performing iterative coupling calculation until the core steady-state power parameter and the thermal parameter converge specifically includes:
starting;
reading an input file, and initializing the section of the assembly few groups and the neutron flux in the reactor core;
performing steady state calculation on the sub-channel thermal module;
updating the section;
performing three-dimensional neutron science module steady-state calculation;
performing steady state calculation on the sub-channel thermal module;
judging whether the power distribution of the reactor core is converged, if not, returning to the step to renew the section again; if yes, the steady state calculation is finished, and the transient state calculation is started.
Further, the three-dimensional transient physical-thermal hydraulic coupling module of the supercritical water reactor core executes transient calculation, which specifically comprises:
starting transient state calculation;
updating the time step, and preparing an initial value at the moment n;
performing transient calculation on a time step by a sub-channel thermotechnical module;
updating the section;
performing transient calculation on a first time step by a three-dimensional neutron space-time dynamics module;
judging whether the reactor core power distribution is converged, if not, returning to execute time step updating, and preparing an initial value at n moments; if n is n + 1; judging whether tn is smaller than T, wherein tn is the time corresponding to the nth time step, T is the time from the preassigned transient calculation to T, if yes, returning to execute time step updating, and preparing the initial value of the n time; if not, the calculation is finished.
Further, the supercritical water reactor core three-dimensional steady-state physical-thermal hydraulic coupling module executes core steady-state calculation and the supercritical water reactor core three-dimensional transient physical-thermal hydraulic coupling module executes transient calculation, and the same space solving method is adopted: the second class of boundary condition blocking green's function methods. Therefore, the self-consistency of the neutron calculation in the three-dimensional steady-state-transient calculation analysis process is ensured.
Furthermore, the time dispersion method adopted by the three-dimensional neutron space-time dynamics module for executing the reactor core three-dimensional transient physical computation is a backward Euler method. Among them, the backward euler method has absolute-stability and facilitates serial coupling development of programs.
One or more technical solutions provided in the embodiments of the present application have at least the following technical effects or advantages:
the method for analyzing the three-dimensional transient performance of the supercritical water reactor core is designed to comprise the following steps: step 1: the supercritical water reactor core three-dimensional stable physical-thermal hydraulic coupling module executes core stable calculation, the core fuel management module provides power parameters for the sub-channel thermal-hydraulic module, the sub-channel thermal-hydraulic module provides thermal parameters for the core fuel management module, and iterative coupling calculation is executed until the core stable power parameters and the thermal parameters are converged; step 2: the reactor core three-dimensional steady-state physical-thermal hydraulic coupling module provides a reactor core initial state and a component section library for the supercritical water reactor core three-dimensional transient physical-thermal hydraulic coupling module; and step 3: the three-dimensional transient physical-thermal hydraulic coupling module of the supercritical water reactor core executes transient calculation, the three-dimensional neutron space-time dynamics module provides power parameters for the sub-channel thermal-hydraulic module, the sub-channel thermal-hydraulic module provides thermal parameters for the three-dimensional neutron space-time dynamics module, and iterative coupling calculation is executed until the transient power parameters and the thermal parameters of the reactor core converge; and 4, step 4: performing power reconstruction on the thermal assembly in reactor core calculation to obtain fine power distribution of the grid cell scale of the thermal assembly, performing thermal assembly subchannel analysis by using the thermal module again, and finally giving out safety evaluation key parameters; and 5: judging whether the safety evaluation key parameters exceed corresponding transient or accident safety limit values or not, and evaluating the safety performance of the supercritical water reactor core in the transient process, namely performing transient process analysis by coupling three-dimensional neutron space-time dynamics and thermal engineering-hydraulics, wherein the neutron dynamics part adopts a three-dimensional model, so that the reactor core nuclear thermal coupling three-dimensional power calculation capacity is realized, and the technical effect of three-dimensional transient analysis of the supercritical water reactor core is realized; therefore, the technical problems that the existing supercritical water reactor power distribution transient analysis method does not have the core-core thermal coupling three-dimensional power calculation capability and is difficult to truly or accurately simulate the transient process such as reactivity and power distribution abnormal events are effectively solved, and the technical effects that the feedback between physics and thermal-hydraulic power can be accurately described, more accurate core three-dimensional power distribution is provided and the transient process and the accident process of the supercritical water reactor are truly simulated by establishing the core three-dimensional stable state-transient physics-thermal hydraulic coupling calculation process are realized.
Furthermore, the method can truly simulate the supercritical water reactor core transient process, improve the knowledge of the supercritical water reactor core transient working condition, estimate the safety margin more accurately, improve the reactor core design and improve the reactor core performance.
Drawings
In order to make the objects, technical solutions and advantages of the present invention more apparent, the present invention is further described in detail below with reference to examples and accompanying drawings, and the exemplary embodiments and descriptions thereof are only used for explaining the present invention and are not meant to limit the present invention.
FIG. 1 is a schematic diagram of a three-dimensional steady-state-transient-physical-thermal-hydraulic coupling calculation process of a supercritical water reactor core;
FIG. 2 is a schematic diagram of a steady state calculation process of a supercritical water reactor core three-dimensional transient performance analysis method;
FIG. 3 is a schematic diagram of a transient calculation process of a supercritical water reactor core three-dimensional transient performance analysis method;
FIG. 4 is a schematic diagram of a backward Euler method transient neutronics calculation process.
Detailed Description
The invention provides a supercritical water reactor core three-dimensional transient performance analysis method, which solves the technical problems that the existing supercritical water reactor power distribution transient analysis method does not have the core thermal coupling three-dimensional power calculation capability and is difficult to truly or accurately simulate the transient process of an abnormal event such as reactivity and power distribution, realizes the technical effects of accurately describing the feedback between physics and thermal-hydraulic power, providing more accurate reactor core three-dimensional power distribution and truly simulating the transient process and the accident process of a supercritical water reactor.
The technical scheme in the implementation of the application is to solve the technical problem. The general idea is as follows:
the method for analyzing the three-dimensional transient performance of the supercritical water reactor core is designed to comprise the following steps: step 1: the supercritical water reactor core three-dimensional stable physical-thermal hydraulic coupling module executes core stable calculation, the core fuel management module provides power parameters for the sub-channel thermal-hydraulic module, the sub-channel thermal-hydraulic module provides thermal parameters for the core fuel management module, and iterative coupling calculation is executed until the core stable power parameters and the thermal parameters are converged; step 2: the reactor core three-dimensional steady-state physical-thermal hydraulic coupling module provides a reactor core initial state and a component section library for the supercritical water reactor core three-dimensional transient physical-thermal hydraulic coupling module; and step 3: the three-dimensional transient physical-thermal hydraulic coupling module of the supercritical water reactor core executes transient calculation, the three-dimensional neutron space-time dynamics module provides power parameters for the sub-channel thermal-hydraulic module, the sub-channel thermal-hydraulic module provides thermal parameters for the three-dimensional neutron space-time dynamics module, and iterative coupling calculation is executed until the transient power parameters and the thermal parameters of the reactor core converge; and 4, step 4: performing power reconstruction on the thermal assembly in reactor core calculation to obtain fine power distribution of the grid cell scale of the thermal assembly, performing thermal assembly subchannel analysis by using the thermal module again, and finally giving out safety evaluation key parameters; and 5: judging whether the safety evaluation key parameters exceed corresponding transient or accident safety limit values or not, and evaluating the safety performance of the supercritical water reactor core in the transient process, namely performing transient process analysis by coupling three-dimensional neutron space-time dynamics and thermal engineering-hydraulics, wherein the neutron dynamics part adopts a three-dimensional model, so that the reactor core nuclear thermal coupling three-dimensional power calculation capacity is realized, and the technical effect of three-dimensional transient analysis of the supercritical water reactor core is realized; therefore, the technical problems that the existing supercritical water reactor power distribution transient analysis method does not have the core-core thermal coupling three-dimensional power calculation capability and is difficult to truly or accurately simulate the transient process such as reactivity and power distribution abnormal events are effectively solved, and the technical effects that the feedback between physics and thermal-hydraulic power can be accurately described, more accurate core three-dimensional power distribution is provided and the transient process and the accident process of the supercritical water reactor are truly simulated by establishing the core three-dimensional stable state-transient physics-thermal hydraulic coupling calculation process are realized.
In order to better understand the technical solution, the technical solution will be described in detail with reference to the drawings and the specific embodiments.
The first embodiment is as follows:
in one embodiment, a method for analyzing three-dimensional transient performance of a supercritical water reactor core is provided, referring to fig. 1 to 4, the method includes:
step 1: the supercritical water reactor core three-dimensional stable physical-thermal hydraulic coupling module executes core stable calculation, the core fuel management module provides power parameters for the sub-channel thermal-hydraulic module, the sub-channel thermal-hydraulic module provides thermal parameters for the core fuel management module, and iterative coupling calculation is executed until the core stable power parameters and the thermal parameters are converged;
step 2: the reactor core three-dimensional steady-state physical-thermal hydraulic coupling module provides a reactor core initial state and a component section library for the supercritical water reactor core three-dimensional transient physical-thermal hydraulic coupling module;
and step 3: the three-dimensional transient physical-thermal hydraulic coupling module of the supercritical water reactor core executes transient calculation, the three-dimensional neutron space-time dynamics module provides power parameters for the sub-channel thermal-hydraulic module, the sub-channel thermal-hydraulic module provides thermal parameters for the three-dimensional neutron space-time dynamics module, and iterative coupling calculation is executed until the transient power parameters and the thermal parameters of the reactor core converge;
and 4, step 4: performing power reconstruction on the thermal assembly in reactor core calculation to obtain fine power distribution of the grid cell scale of the thermal assembly, performing thermal assembly subchannel analysis by using the thermal module again, and finally giving out safety evaluation key parameters;
and 5: and judging whether the safety evaluation key parameters exceed corresponding transient or accident safety limit values or not, and evaluating the safety performance of the supercritical water reactor core in the transient process.
In the embodiment of the present application, the security evaluation key parameters include, but are not limited to: maximum cladding wall temperature, pellet enthalpy.
In practical application, the safety limit values corresponding to the safety evaluation key parameters are respectively: the safety limit value under the transient working condition of the maximum cladding wall surface temperature is 850 ℃, and the safety limit value under the accident working condition is 1260 ℃; the safety limit value under the transient working condition of the enthalpy of the pellets is 170cal/g, and the safety limit value under the accident working condition is 230 cal/g. (safety guidelines require no excessive core damage under accident conditions; and safety guidelines require no systematic fuel rod damage, fuel pellet damage under transient conditions.)
In the embodiment of the present application, a few groups of cross sections are used in the core calculation, and the few groups of cross sections are pre-processed into a cross section library for the core module in a segmented interpolation manner with respect to the change of thermal parameters and preset parameters, including but not limited to: burnup, with or without control rods.
In this embodiment of the present application, the performing iterative coupling calculation until the core steady-state power parameter and the thermal parameter converge specifically includes:
starting;
reading an input file, and initializing the section of the assembly few groups and the neutron flux in the reactor core;
performing steady state calculation on the sub-channel thermal module, and calculating the density of a moderator and the density distribution of a coolant in a reactor core according to the initial assembly small group cross section and the reactor core neutron flux;
updating the cross section, and performing interpolation calculation on the cross section of the less group of the components according to the density distribution of the moderator and the density distribution of the coolant;
performing three-dimensional neutron module steady state calculation, calculating to obtain the average power distribution of the assemblies according to the sections of the new assemblies with few groups, and updating the neutron flux of the reactor core;
performing steady state calculation on the sub-channel thermal module, and calculating the density of a moderator and the density distribution of a coolant in a reactor core according to the section of the new assembly with few groups and the neutron flux of the reactor core;
judging whether the power distribution is converged, if not, returning to the step to renew the section; if yes, the steady state calculation is finished, and the transient state calculation is started.
In an embodiment of the present application, the performing transient calculation by the supercritical water reactor core three-dimensional transient physical-thermal hydraulic coupling module specifically includes:
starting transient state calculation;
updating the time step, and preparing an initial value at the moment n;
performing transient calculation on the sub-channel thermal module for a time step, and calculating the density distribution of a moderator and a coolant in a reactor core according to the section of the initial assembly with few groups and the power distribution of the reactor core;
updating the cross section, and performing interpolation calculation on the cross section of the less group of the components according to the density distribution of the moderator and the density distribution of the coolant;
performing transient calculation on a first time step by a three-dimensional neutron space-time dynamics module, and calculating to obtain reactor core power distribution according to the section of the new assembly small group;
judging whether the power distribution is converged, if not, returning to execute time step updating, and preparing an initial value at n moments; if so, judging whether the time tn corresponding to the nth time step is less than the pre-specified transient calculation time T or not, if so, returning to execute time step updating and preparing an initial value of the n moment; if not, the calculation is finished.
Further, the supercritical water reactor core three-dimensional steady-state physical-thermal hydraulic coupling module executes core steady-state calculation and the supercritical water reactor core three-dimensional transient physical-thermal hydraulic coupling module executes transient calculation, and the same space solving method is adopted: the second class of boundary condition blocking green's function methods. Therefore, the self-consistency of the neutron calculation in the three-dimensional steady-state-transient calculation analysis process is ensured.
The solving idea of the second class of boundary condition block Green function method is as follows:
the form of a three-dimensional neutron space-time kinetic equation to be solved is as follows:
wherein G represents an energy group, and the total number of the energy groups is G; v. ofgCharacterization of group g neutron velocity in cm s-1；Characterization of group g neutron flux in cm-2·s-1；DgCharacterizing the diffusion coefficient of the group g in cm; sigmar,gSum-sigmaf,gCharacterizing the g-th group shifted-out cross-section and the fission cross-section, Σ, respectivelys,g′→gCharacterization of the scattering Cross-section from the g 'th group to the g' th group in cm-1；χgCharacterizing an instantaneous neutron fission spectrum; v characterisation of number of neutrons released per fission(ii) a i, representing a precursor core group of delayed neutrons, wherein the total number of the precursor core group is ND; chi shapeg,iRepresenting the spectrum quota of the ith group of delayed neutrons; lambda [ alpha ]iThe decay constant of the ith group of delayed neutron precursor nucleus is represented in the unit of s-1；The concentration of the delayed neutron precursor nucleus of the ith group is represented in unit of cm-3；βiAnd characterizing the fraction of delayed neutrons in the ith group.
To solve the system of transient equations (1), it can be notedΔt＝t-t0，y1＝y(t)，y0＝y(t0) Where t represents the current moment to be solved, t0The last moment is characterized. f (t, y) characterizes the sum of the non-time partial derivative terms in equation set (1), then equation set (1) can be abbreviated as:
the time dispersion of the equation is obtained using the a-stable backward euler format:
y1＝y0+Δt×f(t,y1) (3)
applying the backward Euler format, the precursor nucleus concentration equation is discretized as:
similarly, the neutron flux equation is discretized as:
substituting the neutron diffusion equation into the equations (4) and (5) to obtain the equation (6), and converting the problem into a fixed source problem:
the fixed source equation (6) is solved by adopting a second class boundary condition-based block Green function method, and the solving idea is as follows:
1. introducing a second type of boundary condition Green's function to obtain product decomposition of a transverse integral partial neutron flux equation;
2. obtaining a corresponding matrix of an interface average net neutron flow by utilizing the non-uniform flux of the joint interface and the continuity condition of the flow;
3. the partial neutron flux, the precursor nucleus concentration and the transverse leakage item are all expanded by adopting a 2-order Legendre polynomial, and a partial neutron flux expansion coefficient is solved by adopting a residual weight method;
4. and solving a nodal neutron balance equation to obtain the average neutron flux, and further correcting the expansion coefficient of the bias neutron flux.
In the solving process of the block Green function, the reactor core is divided into a plurality of basic computing units with the same nuclear section, which are called as blocks. Taking the block as a minimum unit of space dispersion, k represents the block number, and the definition domain of x, y and z direction coordinates is as follows:
the x direction is taken as an example for explanation. Integrating equation (6) in a section k along two directions of y and z to obtain an x-direction partial neutron flux equation:
in the formula,respectively representing transverse integral flux, a source term, a fixed source term and a transverse leakage term.
And introducing a Green function to obtain an integral solution of the partial neutron flux equation:
in the formula,the green's function based on the second class of boundary conditions is characterized.
The non-uniform flux and net neutron flux are continuous across the interface of adjacent segments. The expression for non-uniform flux continuity is:
wherein,and g groups of discontinuity factors representing positive end points of the k blocks in the x coordinate direction.The negative endpoint discontinuity factor of the k +1 segment in the x coordinate direction is characterized.
By substituting formula (9) for formula (10) while using the net neutron flux continuum condition:
equation (11) gives the net neutron flux coupling relationship at the k-block x-coordinate direction internal interface. At the external interface, the neutron flow conditions can be obtained from the core boundary conditions, and the net neutron flow response matrix of the column of blocks is obtained. Taking the x-direction left boundary of the core as an example:
1. reflecting the symmetric condition from the nodal boundaries yields:
2. from the incident flow of 0, we can obtain:
by substituting formula (9) for formula (13), it is possible to obtain:
3. from the condition that the boundary flux is 0, it can be obtained:
by substituting formula (9) for formula (15), it is possible to obtain:
4. from the 90 ° rotational symmetry condition, we can obtain:
the boundary conditions are suitable for the radial direction of the reactor core, and the x-direction and y-direction segments of the reactor core can be uniformly solved.
The neutron balance equation can be obtained by dividing the volume of equation (6) in the block k:
wherein u is x, y, z,and respectively representing the block volume average flux, the source term and the fixed source term.
A partial neutron flux equation integral solution formula (9) based on a Green function, a net neutron flux coupling relation formula (11) and a neutron balance equation (18) based on a nodal block method form an equation set of partial integral flux distribution, interface net flow and nodal block average flux. The x, y, z coordinates are represented by coordinates u, and the one-dimensional distribution quantities of the bias flux, the source, the lateral leakage and the like are expanded by a Legendre polynomial of order 2, so that:
in the formula:
the net flow response matrix can be obtained by substituting (19) for equations (11) to (17). And (3) solving the partial neutron flux expansion coefficient by adopting a residual weight method, wherein the weight function is shown as a formula (20). The matrix form of partial neutron flux solution is:
in the formula, underlining represents a 3-dimensional vector.The expansion coefficient vector is characterized. Andthe same can be obtained.Andcharacterization of eachAndlegendre expansion coefficient of; [ A ]u]Andis a 3 × 3 square matrix, and the coefficients thereof satisfy:
and solving the expansion coefficient of the transverse leakage term by adopting a 2-order approximation method. Assuming that the lateral leakage term relationships are all true in 3 adjacent segments, 3 equations can be obtained from the average lateral leakage of the 3 segments to determine the expansion coefficients. The coefficient expression is:
the average leakage of a segment can be determined from the net neutron flux:
the basic equation of the nodal Green's function transient solution based on the second type boundary condition is composed of a net flow response matrix composed of equations (11) to (17), a neutron balance equation (18) and a bias flux expansion coefficient equation (21), and can be solved by adopting a source iteration method.
Furthermore, the time dispersion method adopted by the three-dimensional neutron space-time dynamics module for executing the reactor core three-dimensional transient physical computation is a backward Euler method. Among them, the backward euler method has absolute-stability and facilitates serial coupling development of programs.
The transient neutron science calculation process adopting the backward euler method, please refer to fig. 4, which specifically includes: firstly, converting the correlation quantity of the last moment; then preparing a cross section; then updating the source item; then, the average net flow of the interfaces in all directions is solved in sequence; then solving the block average flux and partial neutron flux coefficients; then judging whether all the groups can meet the requirements, if not, returning to execute updating source items, if so, judging whether the fission sources are converged, if not, returning to execute updating source items, and if so, updating the precursor nucleus concentration; then the next moment calculation is performed.
The present invention will be described in further detail with reference to the accompanying drawings and examples.
The invention provides a supercritical water reactor core three-dimensional transient performance analysis method, wherein a core three-dimensional steady-state-transient physical-thermal hydraulic coupling calculation process applying the analysis method of the invention is shown in figure 1, a steady-state calculation process applying the analysis method of the invention is shown in figure 2, a transient calculation process applying the analysis method of the invention is shown in figure 3, figures 1, 2 and 3 only show one embodiment of the concept proposed by the invention, and the invention is further explained by combining figures 1, 2 and 3 with specific embodiments:
as shown in figure 1, the three-dimensional steady-state-transient-physical-thermal-hydraulic coupling calculation process of the supercritical water reactor core comprises five steps of (1) SCWR (design language reactor) core three-dimensional steady-state physical-thermal-hydraulic coupling calculation, (2) data transmission from a steady-state module to a transient module, (3) SCWR core three-dimensional transient-physical-thermal-hydraulic coupling calculation, (4) thermal component sub-channel thermal-hydraulic calculation, and (5) key parameter safety evaluation, wherein the SCWR core three-dimensional steady-state physical-thermal-hydraulic coupling calculation (1) comprises two parts of core fuel management module calculation of the steady-state core and sub-channel thermal-hydraulic module calculation of the steady-state core, the data transmission from the steady-state module to the transient module (2) comprises an initial core state and a component cross-section library, and the SCWR core three-dimensional transient-physical-thermal-hydraulic coupling calculation (3) comprises a three-dimensional neutron The method comprises the steps that a hydraulic module calculates two parts of a transient reactor core, a thermal component sub-channel thermal-hydraulic calculation (4) utilizes a thermal module to perform sub-channel calculation analysis on a thermal component after power reconstruction, the maximum cladding wall surface temperature (MCST) is given, and a key parameter safety evaluation (5) evaluates the safety performance of the supercritical water reactor core in the transient process by judging whether the MCST exceeds a corresponding transient or accident safety limit value.
As shown in fig. 2, in the steady state calculation process of the supercritical water reactor core three-dimensional transient performance analysis method, the three-dimensional neutronics module and the sub-channel thermal module respectively perform core steady state calculation, and perform iterative coupling calculation until power convergence.
As shown in fig. 3, in the transient calculation process of the supercritical water reactor core three-dimensional transient performance analysis method, the sub-channel thermal module and the three-dimensional neutron space-time dynamics module respectively perform core transient calculation, and perform iterative coupling calculation until power convergence.
The supercritical water reactor core three-dimensional transient performance analysis method establishes a core three-dimensional steady state-transient physical-thermal hydraulic coupling calculation process, can accurately describe feedback between physics and thermal-hydraulic, provides accurate core three-dimensional power distribution, and achieves the purpose of truly simulating the transient process and the accident process of the supercritical water reactor.
The following is presented as a specific example:
by applying the three-dimensional transient analysis method for the supercritical water reactor core, the transient problem of the supercritical water reactor in the states of thermal state zero power (HZP) and thermal state full power (HFP) is analyzed, so that the applicability of the three-dimensional transient analysis method is checked. The problem core is a supercritical water reactor CSR1000 core, and the initial power of the core under the HZP and HFP states is 2300W and 2300MW respectively. The transient problem is the pop-up problem, and the pop-up position is the II-flow component position E11. In the HZP slug problem, the initial position of the slug was 7.7cm from the bottom of the active zone and the slug reached the end position within 0.1s, 420 cm. In the HFP pop-up problem, the pop-up rod initial position was 210cm from the bottom of the active zone and popped up to the end position within 0.1s, 420 cm. The pop-up bar values for the HZP and HFP problems were 683pcm and 293pcm, respectively. By utilizing the three-dimensional transient analysis method, firstly, a CSR1000 reactor core three-dimensional steady-state physical-thermal hydraulic coupling module is executed to execute reactor core steady-state calculation, a reactor core fuel management module provides power parameters for a sub-channel thermal-hydraulic module, the sub-channel thermal-hydraulic module provides thermal parameters for the reactor core fuel management module, and iterative coupling calculation is executed until the reactor core steady-state power parameters and the thermal parameters are converged; then, providing a reactor core initial state and a component section library for the reactor core three-dimensional transient physical-thermal hydraulic coupling module by using the reactor core three-dimensional steady-state physical-thermal hydraulic coupling module; then executing a CSR1000 reactor core three-dimensional transient physical-thermal hydraulic coupling module to execute transient calculation, wherein the three-dimensional neutron space-time dynamics module provides power parameters for the sub-channel thermal-hydraulic module, the sub-channel thermal-hydraulic module provides thermal parameters for the three-dimensional neutron space-time dynamics module, and iterative coupling calculation is executed until the reactor core transient power parameters and the thermal parameters are converged; then, power reconstruction is carried out on the thermal assembly in reactor core calculation to obtain fine power distribution of the grid cell scale of the thermal assembly, thermal assembly sub-channel analysis is carried out by utilizing a thermal module again, and finally a safety evaluation key parameter, namely the maximum cladding wall surface temperature, is given out; and finally, judging whether the maximum cladding wall surface temperature exceeds the accident safety limit value or not, and evaluating the safety performance of the CSR1000 reactor core in the rod ejection transient process. The three-dimensional transient analysis result shows that the core peak power of the HZP rod ejection problem is 0.68NP (rated power 2300MW), the maximum cladding wall surface temperature is 440 ℃, and the temperature is lower than the safety limit value 1260 ℃ under the accident condition; the peak power of the reactor core of the HFP rod ejection problem is 1.74NP, the maximum cladding wall temperature is 1075 ℃, and the maximum cladding wall temperature is 1260 ℃ lower than the safety limit under the accident condition. The three-dimensional transient analysis of the CSR1000 core rod ejection problem shows that the maximum cladding wall surface temperature peak value in the accident process is lower than the accident safety limit value 1260 ℃ no matter whether the HFP rod ejection or the HZP rod ejection is carried out, and the safety requirement is met.
The technical scheme in the embodiment of the application at least has the following technical effects or advantages:
the method for analyzing the three-dimensional transient performance of the supercritical water reactor core is designed to comprise the following steps: step 1: the supercritical water reactor core three-dimensional stable physical-thermal hydraulic coupling module executes core stable calculation, the core fuel management module provides power parameters for the sub-channel thermal-hydraulic module, the sub-channel thermal-hydraulic module provides thermal parameters for the core fuel management module, and iterative coupling calculation is executed until the core stable power parameters and the thermal parameters are converged; step 2: the reactor core three-dimensional steady-state physical-thermal hydraulic coupling module provides a reactor core initial state and a component section library for the supercritical water reactor core three-dimensional transient physical-thermal hydraulic coupling module; and step 3: the three-dimensional transient physical-thermal hydraulic coupling module of the supercritical water reactor core executes transient calculation, the three-dimensional neutron space-time dynamics module provides power parameters for the sub-channel thermal-hydraulic module, the sub-channel thermal-hydraulic module provides thermal parameters for the three-dimensional neutron space-time dynamics module, and iterative coupling calculation is executed until the transient power parameters and the thermal parameters of the reactor core converge; and 4, step 4: performing power reconstruction on the thermal assembly in reactor core calculation to obtain fine power distribution of the grid cell scale of the thermal assembly, performing thermal assembly subchannel analysis by using the thermal module again, and finally giving out safety evaluation key parameters; and 5: judging whether the safety evaluation key parameters exceed corresponding transient or accident safety limit values or not, and evaluating the safety performance of the supercritical water reactor core in the transient process, namely performing transient process analysis by coupling three-dimensional neutron space-time dynamics and thermal engineering-hydraulics, wherein the neutron dynamics part adopts a three-dimensional model, so that the reactor core nuclear thermal coupling three-dimensional power calculation capacity is realized, and the technical effect of three-dimensional transient analysis of the supercritical water reactor core is realized; therefore, the technical problems that the existing supercritical water reactor power distribution transient analysis method does not have the core-core thermal coupling three-dimensional power calculation capability and is difficult to truly or accurately simulate the transient process such as reactivity and power distribution abnormal events are effectively solved, and the technical effects that the feedback between physics and thermal-hydraulic power can be accurately described, more accurate core three-dimensional power distribution is provided and the transient process and the accident process of the supercritical water reactor are truly simulated by establishing the core three-dimensional stable state-transient physics-thermal hydraulic coupling calculation process are realized.
Furthermore, the method can truly simulate the supercritical water reactor core transient process, improve the knowledge of the supercritical water reactor core transient working condition, estimate the safety margin more accurately, improve the reactor core design and improve the reactor core performance.
While preferred embodiments of the present invention have been described, additional variations and modifications in those embodiments may occur to those skilled in the art once they learn of the basic inventive concepts. Therefore, it is intended that the appended claims be interpreted as including preferred embodiments and all such alterations and modifications as fall within the scope of the invention.
It will be apparent to those skilled in the art that various changes and modifications may be made in the present invention without departing from the spirit and scope of the invention. Thus, if such modifications and variations of the present invention fall within the scope of the claims of the present invention and their equivalents, the present invention is also intended to include such modifications and variations.
Claims (6)
1. A three-dimensional transient performance analysis method for a supercritical water reactor core is characterized by comprising the following steps:
step 1: the supercritical water reactor core three-dimensional stable physical-thermal hydraulic coupling module executes core stable calculation, the core fuel management module provides power parameters for the sub-channel thermal-hydraulic module, the sub-channel thermal-hydraulic module provides thermal parameters for the core fuel management module, and iterative coupling calculation is executed until the core stable power parameters and the thermal parameters are converged;
step 2: the reactor core three-dimensional steady-state physical-thermal hydraulic coupling module provides a reactor core initial state and a component section library for the supercritical water reactor core three-dimensional transient physical-thermal hydraulic coupling module;
and step 3: the three-dimensional transient physical-thermal hydraulic coupling module of the supercritical water reactor core executes transient calculation, the three-dimensional neutron space-time dynamics module provides power parameters for the sub-channel thermal-hydraulic module, the sub-channel thermal-hydraulic module provides thermal parameters for the three-dimensional neutron space-time dynamics module, and iterative coupling calculation is executed until the transient power parameters and the thermal parameters of the reactor core converge;
and 4, step 4: performing power reconstruction on the thermal assembly in reactor core calculation to obtain fine power distribution of the thermal assembly grid cell scale, performing thermal assembly subchannel analysis by using the subchannel thermal-hydraulic module again, and finally giving out safety evaluation key parameters;
and 5: judging whether the safety evaluation key parameters exceed corresponding transient or accident safety limit values or not, and evaluating the safety performance of the supercritical water reactor core in the transient process; the safety evaluation key parameters comprise: maximum cladding wall temperature, pellet enthalpy.
2. The method of claim 1, wherein the few groups of cross sections are used in the calculation of the reactor core three-dimensional neutron space-time dynamics module, and the few groups of cross sections are preprocessed into a cross section library for the reactor core three-dimensional neutron space-time dynamics module by means of segmented interpolation according to the change of thermal parameters and preset parameters, wherein the preset parameters comprise: the fuel assembly is burnable with or without control rods.
3. The method of claim 1, wherein the performing iterative coupled calculations until the core steady-state power parameters and the thermal parameters converge comprises:
starting;
reading an input file, and initializing the section of the assembly few groups and the neutron flux in the reactor core;
performing steady state calculation of the sub-channel thermal-hydraulic module;
updating the section;
performing steady-state calculation of a three-dimensional neutron space-time dynamics module;
performing steady state calculation on the sub-channel thermal module;
judging whether the power distribution of the reactor core is converged, if not, returning to the step to renew the section again; if yes, the steady state calculation is finished, and the transient state calculation is started.
4. The method of claim 1, wherein the supercritical water reactor core three-dimensional transient physical-thermal hydraulic coupling module performing transient calculations specifically comprises:
starting transient state calculation;
updating the time step, and preparing an initial value at the moment n;
performing transient calculation on a time step by a sub-channel thermotechnical module;
updating the section;
performing transient calculation on a first time step by a three-dimensional neutron space-time dynamics module;
judging whether the reactor core power distribution is converged, if not, returning to execute time step updating, and preparing an initial value at n moments; if n is n + 1; judging whether tn is smaller than T, wherein tn is the time corresponding to the nth time step, T is the time from the preassigned transient calculation to the T moment, if yes, returning to execute the updating of the time step, and preparing the initial value of the n moment; if not, the calculation is finished.
5. The method of claim 1, wherein the supercritical water reactor core three-dimensional steady-state physical-thermal hydraulic coupling module performs core steady-state calculations and the supercritical water reactor core three-dimensional transient physical-thermal hydraulic coupling module performs transient calculations, both using the same spatial solution method: the second class of boundary condition blocking green's function methods.
6. The method of claim 1, wherein the three-dimensional neutron spatiotemporal dynamics module employs a time discrete method of: backward euler method.
Priority Applications (1)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| CN201610003796.5A CN105653869B (en) | 2016-01-05 | 2016-01-05 | A kind of supercritical water reactor reactor core Three dimensional transient method for analyzing performance |
Applications Claiming Priority (1)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| CN201610003796.5A CN105653869B (en) | 2016-01-05 | 2016-01-05 | A kind of supercritical water reactor reactor core Three dimensional transient method for analyzing performance |
Publications (2)
| Publication Number | Publication Date |
|---|---|
| CN105653869A CN105653869A (en) | 2016-06-08 |
| CN105653869B true CN105653869B (en) | 2018-09-11 |
Family
ID=56491620
Family Applications (1)
| Application Number | Title | Priority Date | Filing Date |
|---|---|---|---|
| CN201610003796.5A Active CN105653869B (en) | 2016-01-05 | 2016-01-05 | A kind of supercritical water reactor reactor core Three dimensional transient method for analyzing performance |
Country Status (1)
| Country | Link |
|---|---|
| CN (1) | CN105653869B (en) |
Families Citing this family (14)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| CN108256212B (en) * | 2018-01-16 | 2021-08-31 | 中南大学 | Parallel performance analysis method, system and storage medium for multi-rod fuel elements |
| CN108763670B (en) * | 2018-05-15 | 2021-04-20 | 西安交通大学 | Method for solving Brayton cycle transient process of supercritical carbon dioxide reactor |
| CN108846190B (en) * | 2018-06-05 | 2022-04-12 | 哈尔滨工程大学 | A nuclear-thermal coupled simulation method for pressurized water reactor fuel assemblies |
| CN108875213B (en) * | 2018-06-19 | 2022-04-12 | 哈尔滨工程大学 | Reactor core thermal hydraulic multi-scale analysis method |
| CN109192341B (en) * | 2018-09-13 | 2020-01-14 | 中国核动力研究设计院 | Large reactivity measurement method based on three-dimensional space-time dynamics |
| CN109215822B (en) * | 2018-09-13 | 2022-08-05 | 中国核动力研究设计院 | Rod drop reactivity measurement method |
| CN111680458B (en) * | 2020-06-03 | 2021-10-19 | 西安交通大学 | Thermodynamic hydraulic transient calculation method suitable for sodium water direct-current steam generator |
| CN112347645B (en) * | 2020-11-06 | 2022-03-22 | 中国核动力研究设计院 | Method and device for reconstructing burnup characteristics of reactor core grid cells |
| CN112613156B (en) * | 2020-11-19 | 2022-04-19 | 中国核动力研究设计院 | Fine fuel rod performance analysis method |
| CN113076682B (en) * | 2021-04-19 | 2022-08-05 | 西安交通大学 | Reactor core physical-thermal coupling simulation method based on multi-physical field frame |
| CN113536537B (en) * | 2021-06-10 | 2024-01-12 | 中国核动力研究设计院 | Method and system for analyzing large-break water loss accident |
| CN113409975B (en) * | 2021-06-17 | 2022-11-15 | 中国核动力研究设计院 | Reactor core power distribution monitoring method and system based on model order reduction and data assimilation |
| CN114239279B (en) * | 2021-12-17 | 2023-10-31 | 中国核动力研究设计院 | Reactor thermal safety design cooperative device, method, terminal and storage medium |
| CN115358125B (en) * | 2022-08-22 | 2023-06-09 | 中广核工程有限公司 | Pressurized water reactor three-dimensional reactor core nuclear thermal coupling method and system |
Citations (1)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| CN104133965A (en) * | 2014-07-30 | 2014-11-05 | 中国核动力研究设计院 | Transient analysis method applied to two-pass core |
-
2016
- 2016-01-05 CN CN201610003796.5A patent/CN105653869B/en active Active
Patent Citations (1)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| CN104133965A (en) * | 2014-07-30 | 2014-11-05 | 中国核动力研究设计院 | Transient analysis method applied to two-pass core |
Non-Patent Citations (6)
| Title |
|---|
| Improved core design of the high temperature supercritical-pressure light water reactor;A. Yamaji et al.;《Annals of Nuclear Energy》;20051231;第32卷(第7期);第651-670页 * |
| SCWR堆芯三维瞬态物理-热工水力耦合程序开发;王连杰 等;《核动力工程》;20141231;第35卷;第186-189页 * |
| SCWR堆芯三维稳态性能分析程序系统开发及验证;王连杰 等;《核动力工程》;20150831;第36卷(第4期);第41-44页 * |
| SCWR堆芯稳态物理-热工水力耦合计算程序系统CASIR的开发;马永强 等;《核动力工程》;20130228;第34卷(第1期);第87-91页 * |
| 基于格林函数节块法的物理与热工-水力耦合方法研究;赵文博 等;《原子能科学技术》;20130228;第47卷(第2期);第266-270页 * |
| 超临界水冷堆三维瞬态耦合程序研究;程万旭;《中国优秀硕士学位论文全文数据库工程科技Ⅱ辑》;20150115(第01期);第5、11、14、17-18、25-28页 * |
Also Published As
| Publication number | Publication date |
|---|---|
| CN105653869A (en) | 2016-06-08 |
Similar Documents
| Publication | Publication Date | Title |
|---|---|---|
| CN105653869B (en) | A kind of supercritical water reactor reactor core Three dimensional transient method for analyzing performance | |
| CN111414722B (en) | Simulation method for physical and thermal coupling of nuclear reactor core | |
| CN107273582A (en) | A kind of computational methods for fast neutron reactor neutron transport burnup coupling analysis | |
| Weber et al. | High-fidelity light water reactor analysis with the numerical nuclear reactor | |
| CN112906271A (en) | Reactor transient physical thermal full-coupling fine numerical simulation method and system | |
| Rohde et al. | Development and verification of the coupled 3D neutron kinetics/thermal-hydraulics code DYN3D-HTR for the simulation of transients in block-type HTGR | |
| Valtavirta et al. | Coupled neutronics–fuel behavior calculations in steady state using the Serpent 2 Monte Carlo code | |
| CN113254860B (en) | A Calculation Method for Neutron Flux of Core Cells | |
| CN112906272B (en) | Reactor steady-state physical thermal full-coupling fine numerical simulation method and system | |
| Laureau et al. | Coupled neutronics and thermal-hydraulics transient calculations based on a fission matrix approach: application to the Molten Salt Fast Reactor | |
| Magedanz et al. | High-fidelity multi-physics system TORT-TD/CTF/FRAPTRAN for light water reactor analysis | |
| Xie et al. | Development of 3D transient neutronics and thermal-hydraulics coupling procedure and its application to a fuel pin analysis | |
| Suikkanen et al. | Development of coupled neutronics and fuel performance analysis capabilities between Serpent and TRANSURANUS | |
| Dong et al. | The development of nuclear reactor three-dimensional neutronic thermal–hydraulic coupling code: CorTAF-2.0 | |
| CN106021184A (en) | A spatial-temporal dynamics solving method for hexagonal component geometric reactor cores | |
| Hursin et al. | Synergism of the method of characteristic, R-functions and diffusion solution for accurate representation of 3D neutron interactions in research reactors using the AGENT code system | |
| Liu et al. | Development and verification of the hexagonal core simulation capability of the NECP-X code | |
| Zhou | Operator split, Picard iteration and JFNK methods based on nonlinear CMFD for transient full core models in the coupling multiphysics environment | |
| Fiorina et al. | Creation of an OpenFOAM fuel performance class based on FRED and integration into the GeN-foam multi-physics code | |
| Feng et al. | Development of a multi-physics coupling system based on ICoCo interface and its validation on NEA-OECD core transient benchmark | |
| CN110705184A (en) | A Virtual Body Force Momentum Source Method for Refined Numerical Solution of Reactor Core | |
| Gu et al. | Verification of a HC-PK-CFD coupled program based a benchmark on beam trip transients for XADS reactor | |
| Girieud et al. | Science Version 2: the most recent capabilities of the Framatome 3-D nuclear code package | |
| Sartori | Reduced order methods: applications to nuclear reactor core spatial dynamics | |
| He et al. | The development of the combined fission matrix method in small advanced reactor design and optimization |
Legal Events
| Date | Code | Title | Description |
|---|---|---|---|
| C06 | Publication | ||
| PB01 | Publication | ||
| C10 | Entry into substantive examination | ||
| SE01 | Entry into force of request for substantive examination | ||
| GR01 | Patent grant | ||
| GR01 | Patent grant |