CN110457802B - Precision optimization implementation method for SFCOMPO fuel consumption experiment benchmark question check simulation - Google Patents
Precision optimization implementation method for SFCOMPO fuel consumption experiment benchmark question check simulation Download PDFInfo
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Abstract
A precision optimization implementation method for calibration simulation of SFCOMPO fuel consumption experiment benchmark questions is characterized by establishing a single grid element model and defining total reflection boundary conditions around the single grid element model; establishing a super-grid cell calculation model, completing the initialization of the model, carrying out geometric and burnup modeling on the experimental sample through a burnup component calculation program, and carrying out simulation calculation on the burnup process of the experimental sample; comparing the simulated and calculated spent fuel component with the error of the actually measured spent fuel component provided by the database to adjust the calculation precision of the fuel consumption library used by the fuel consumption component calculation program; and finally, simulating and calculating the burnup depth through nuclear species check of a spent fuel component Nd-148, so that the calculated burnup depth is consistent with the actual burnup depth. The invention utilizes the stock of Nd-148 nuclide to adjust the burnup depth in the numerical simulation calculation, eliminates the deviation of numerical simulation and experimental measurement means, and leads the simulation calculation to be closer to the real burnup process.
Description
Technical Field
The invention relates to a technology in the field of nuclear engineering, in particular to a precision optimization realization method for benchmark question check simulation of a spent fuel database (SFCOMPO) fuel consumption experiment.
Background
The burnup calculation accuracy of the component program is of great significance to the aspects of power distribution, refueling life and reactivity control design of the nuclear reactor core. In order to develop a MOX fuel and burnable poison multi-group constant library, a burnup benchmark experimental calculation example of the multi-group constant library needs to be established for burnup benchmark inspection of the library.
Disclosure of Invention
Aiming at the defects in the prior art, the invention provides a precision optimization realization method for the calibration simulation of the SFCOMPO fuel consumption experiment benchmark questions, the calculated fuel consumption depth is basically consistent with the actual fuel consumption depth, and the influence of different fuel consumption depths on the calculation result is reduced, so that powerful support is provided for the verification of a multi-group constant library.
The invention is realized by the following technical scheme:
the invention comprises the following steps:
1.1 query sfcomp database for corresponding material and geometry information, obtained by but not limited to www.oecd-nea.
1.2 analyzing and calculating the physical environment of the sample point, wherein the change rate of the neutron flux density at the sample point specifically comprises the following steps: and analyzing and calculating the types of fuel rods around the burnup sample point, whether a neutron absorber, a control rod guide tube and the like exist. This causes distortion of the neutron flux density at the point of burnup of the sample.
1.3, establishing a single grid cell model and a super grid cell calculation model according to the analysis condition.
2.1 obtaining the initial uranium density in g/cm, from the initial fuel composition3。
2.2 stock of nuclides in g/gU according to SFCOMPO libraryiConverted into corresponding nuclide density with the unit of g/cm3。
And 2.3, obtaining the nuclear density of the nuclide corresponding to the stock according to the relation of the amount of the substance.
3.1 obtaining the final Nd-148 nuclide stock by utilizing a fuel consumption component calculation program according to the fuel consumption history queried by the SFCOMPO database.
3.2 comparing the Nd-148 nuclide stock in the SFCOMPO database with the Nd-148 nuclide stock obtained by the fuel-up component calculation program.
3.3 when the difference between the Nd-148 inventory in the database and the Nd-148 inventory obtained in the step 3.2 is more than 0.03 percent, adjusting the fuel consumption depth of the fuel consumption component calculation program.
The burnup component calculation program adopts, but is not limited to, reactor simulation software DRAGON developed by the nuclear engineering system of Montreal university, Canada.
Technical effects
Compared with the prior art, the method utilizes the storage of the Nd-148 nuclide to adjust the burnup depth in the numerical simulation calculation, eliminates the deviation of numerical simulation and experimental measurement means, and enables the simulation calculation to be closer to the real burnup process.
Drawings
FIG. 1 is a schematic flow diagram of the present invention;
FIG. 2 is a schematic diagram of a single grid cell model;
FIG. 3 is a schematic diagram of a model for computing a super cell;
FIG. 4 is a graph of the relative error comparison of the calculation results of the embodiment;
in the figure: 1 is moderator area, 2 is cladding tube, 3 is fuel area, 4 control rod guide tube, 5 composite flammable absorbent.
Detailed Description
In this example, samples of the pressurized water reactor Takahama-3 assemblies NT3G23 and NT3G24 series were selected to carry out the above method, wherein an NT3G23 assembly was provided with SF95 and SF96 series fuel rods, an NT3G24 assembly contained SF97 series fuel rods, SF95 and SF97 fuel rods were surrounded by the same fuel rods, control rod-free guide tubes and burnable poison distorted to neutron flux density,the SF96 fuel rod is a burnable poison rod, and strong poison absorption can cause kinfThe values are low and even the calculation cannot be performed.
The embodiment specifically comprises the following steps:
step 1) establishing a single-grid cell model aiming at SF95 and SF97 series fuel rods, and defining a total reflection boundary condition around the single-grid cell model.
As shown in fig. 2, the single cell model includes cells containing moderator, and cladding tubes containing fuel and disposed in the cells, wherein: the outer radius of the cladding tube 2 is 0.475cm, the radius of the fuel area 3 is 0.411cm, the side length L of the grid element is 1.26cm, the calculated fuel temperature is defined as 900K, and the cladding temperature is 600K. Meanwhile, the coolant temperature was set to 575.5K based on the average value of the core inlet-outlet coolant temperature.
The boundary conditions of total reflection are as follows: j. the design is a square-|x0＝J+|x0Wherein: j. the design is a square-|x0Representing the density of neutron flux emitted at the boundary of the calculated region, J+|x0Representing the neutron flux density reflected from outside the calculated area back to the calculated area at the boundary.
Step 2) calculating a model for the SF96 series fuel rod super-grid cells.
As shown in fig. 3, the super cell computation model includes: the 3x3 grid cells containing the moderator are 9, each grid cell is internally provided with a cladding tube (provided with fuel), a control rod guide tube (4) internally provided with the moderator or a cladding tube internally provided with the composite combustible absorbent, wherein: the external radius of the cladding tube with the built-in moderator is 0.475cm, the radius of the fuel area is 0.411cm, and the side length of the grid cell is 1.26 cm.
The composite combustible absorbent is UO2-Gd2O3。
The material of the control rod guide tube 4 is the same as that of the fuel cladding, the inner diameter is 10.0838mm, and the outer diameter is 10.922 mm.
And 3) converting the nuclear stock unit of the SFCOMPO library by the ratio relationship of the amount of the substance to the molecular nuclear ratio because the nuclear stock unit of the component information of the spent fuel database SFCOMPO is not consistent with the nuclear stock unit of the component program calculation resultIs a unit of nuclear density, i.e., units/cm, consistent with the calculation results3。
The ratio relation between the amount of the substance and the molecular formula nuclide satisfies the following conditions:
wherein: the specific meaning of each variable is expressed as: n represents the number of material particles. m represents the mass of the substance. M represents the relative atomic or molecular mass of a substance. N is a radical ofARepresents an avogalois constant.
And 4) when a large error exists between the burnup modeling and the actual burnup history, adjusting the final burnup depth through a spent fuel component Nd-148 to ensure that the actual burnup depth is basically consistent with the calculated burnup depth.
The final burn-up depth is adjusted by: when the inventory of Nd-148 in the database is large, the burn-up depth of a Dragon calculation program is increased, and when the inventory of Nd-148 nuclides obtained by the Dragon program is large, the burn-up depth calculated by the program is reduced; and (4) adjusting until the difference between the Nd-148 inventory and the Nd-148 inventory in the database is within 0.03 percent under the condition of calculating the burnup depth, finishing the adjustment, and finally calculating the burnup depth by taking the current burnup depth as the final burnup depth.
Through specific practical experiments, experimental data obtained by operating the method with a Takahama3 reactor as an example and SF95, SF96 and SF97 samples as calculation simulation objects are shown in Table 1 and FIG. 4.
TABLE 1 comparison of the results
Compared with the prior art, the difference between the spent fuel nuclide stock and the database nuclide stock is smaller.
The foregoing embodiments may be modified in many different ways by those skilled in the art without departing from the spirit and scope of the invention, which is defined by the appended claims and all changes that come within the meaning and range of equivalency of the claims are therefore intended to be embraced therein.
Claims (1)
1. A precision optimization implementation method for calibration simulation of SFCOMPO fuel consumption experiment benchmark questions is characterized in that a single grid element model is established, and total reflection boundary conditions are defined around the single grid element model; establishing a super-grid cell calculation model, completing the initialization of the model, carrying out geometric and burnup modeling on the experimental sample through a burnup component calculation program, and carrying out simulation calculation on the burnup process of the experimental sample; comparing the simulated and calculated spent fuel component with the error of the actually measured spent fuel component provided by the database to adjust the calculation precision of the fuel consumption library used by the fuel consumption component calculation program; finally, simulating and calculating the burnup depth through nuclear species check of a spent fuel component Nd-148, so that the calculated burnup depth is consistent with the actual burnup depth;
the single grid cell model comprises grid cells containing a moderator and a cladding tube which is arranged in the grid cells and is internally provided with fuel, wherein: the outer radius of the cladding tube is 0.475cm, the radius of the fuel area is 0.411cm, the side length of the grid element is 1.26cm, the calculated temperature of the fuel is defined as 900K, the cladding temperature is 600K, and the coolant temperature is set to be 575.5K according to the average value of the coolant temperature at the inlet and the outlet of the reactor core;
the boundary conditions of total reflection are as follows: j. the design is a square-|x0＝J+|x0Wherein: j. the design is a square-|x0Representing the density of neutron flux emitted at the boundary of the calculated region, J+|x0Representing the neutron flux density reflected from outside the calculated area back to the calculated area at the boundary;
the super grid cell calculation model comprises: the 3x3 grid cells containing the moderator are 9, each grid cell is internally provided with a cladding tube with fuel inside, a control rod guide tube with the moderator inside or a cladding tube with the composite combustible absorbent inside, wherein: the outer radius of the cladding tube with the built-in moderator is 0.475cm, the radius of the fuel area is 0.411cm, and the side length of the grid cell is 1.26 cm;
the simulation calculation refers to the ratio of the amount of the substance to the molecular formula nuclideThe relationship converts the SFCOMPO stock unit of nuclides into the density unit of the nuclei consistent with the calculation result, namely, the number of the nuclei per cm3(ii) a Wherein: the ratio relation between the amount of the substance and the molecular formula nuclide satisfies the following conditions:
wherein: the specific meaning of each variable is expressed as: n represents the number of particles of the substance, M represents the mass of the substance, M represents the relative atomic or molecular mass of the substance, NARepresents an avogalois constant;
the adjustment is as follows: adjusting the final burnup depth by comparing Nd-148 of the spent fuel component in the database with the calculated Nd-148 stock to ensure that the actual burnup depth is basically consistent with the calculated burnup depth;
the adjustment specifically comprises:
1) obtaining the final Nd-148 nuclide stock by utilizing a burnup component calculation program according to the burnup history inquired by the SFCOMPO database;
2) comparing the Nd-148 nuclide stock in the SFCOMPO database with the Nd-148 nuclide stock obtained by the fuel consumption component calculation program;
3) when the difference between the inventory of the Nd-148 in the database and the inventory of the Nd-148 obtained in the step 3.2 is larger than 0.03 percent, adjusting the fuel consumption depth of the fuel consumption component calculation program;
the burnup component calculation program adopts DRAGON.
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| CN113470766B (en) * | 2021-06-23 | 2023-11-10 | 中国原子能科学研究院 | Automatic fission product burnup chain testing method and device |
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| CN115099049B (en) * | 2022-07-08 | 2023-05-16 | 西安交通大学 | Fission yield data optimization method based on actual measured value of nuclide stock of irradiated sample |
| CN116487079B (en) * | 2023-04-03 | 2024-01-30 | 中国核动力研究设计院 | Method for qualitatively testing nuclear fuel microcell burnup based on electronic probe |
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