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Nuclear reactor simulated fuel assembly flow measurement method and system

A technology for simulating fuel assemblies and flow measurement, which is applied in the direction of liquid/fluid solid measurement, volume/mass flow generated by mechanical effects, and measurement devices. It can solve the problems of large changes in Reynolds number and achieve wide range flow measurement. adaptive effect

Active Publication Date: 2020-11-24
NUCLEAR POWER INSTITUTE OF CHINA
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  • Abstract
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Problems solved by technology

[0003] In the nuclear field, when the overall hydraulic simulation test of the reactor measures the flow rate of the simulated fuel assembly, since there are various working conditions in the reactor operation, the Reynolds number varies greatly under different working conditions, which makes the flow measurement method in general technology It is not applicable to the measurement of the flow rate of nuclear reactor simulated fuel assemblies, and there is no relevant literature for reference

Method used

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  • Nuclear reactor simulated fuel assembly flow measurement method and system
  • Nuclear reactor simulated fuel assembly flow measurement method and system
  • Nuclear reactor simulated fuel assembly flow measurement method and system

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Embodiment

[0063] like figure 1 As shown, a method for measuring the flow rate of a nuclear reactor simulated fuel assembly of the present invention includes the following steps:

[0064] S1: use a venturi to calibrate the relationship between the flow coefficient α and the Reynolds number Re of the fluid in the simulated fuel assembly to generate a calibration relationship function; the independent variable of the calibration relationship function is the Reynolds number Re, and the strain variable is the flow coefficient α;

[0065] S2: Fitting an ascending segment in the calibration relation function to generate an ascending segment fitting function; fitting a descending segment in the calibration relation function to generate a descending segment fitting function;

[0066] S3: generating an ascending segment model by combining the fitting function of the ascending segment and the theoretical relationship function of the venturi; generating a descending segment model by combining the f...

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Abstract

The invention discloses a nuclear reactor simulated fuel assembly flow measurement method. The invention further discloses a nuclear reactor simulation fuel assembly flow measurement system. On the premise that the geometrical structure of a classical venturi tube is not changed, the measuring range ratio of the classical venturi tube is widened, and the measurable minimum flow of the classical venturi tube is further reduced. Based on the fact that the corresponding flow coefficient alpha in the widened Re range has a special change trend, the invention provides a flow calculation method of the wide-range classical venturi tube, iterative calculation is not needed by utilizing the calculation method, and efficient and high-precision flow measurement of the wide-range venturi tube is realized through a method of directly solving an equation.

Description

technical field [0001] The invention relates to the field of nuclear technology, in particular to a method and system for measuring the flow of a simulated fuel assembly of a nuclear reactor. Background technique [0002] In "GB / T 2624-2006 Measuring Full Pipe Fluid Flow with Differential Pressure Devices Installed in Circular Section Pipes", the fourth part of the introduction to the use of classic Venturi tubes has constraints that require Reynolds number Re ≥ 2 ×10 5 , Under real use conditions, the Venturi flowmeter is often required to have a higher range ratio and can measure smaller flow rates. It is necessary to widen the flow range of the flowmeter calibration test without changing the geometric structure of the classic Venturi flowmeter. In this case, the Reynolds number Re will be less than 2 × 10 5 , in the national standard GB / T 2624-2006, Appendix B, "Classic Venturi used beyond the scope of GB / T 2624.4", there is such a description, when the Reynolds number ...

Claims

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Application Information

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IPC IPC(8): G01F1/44G01F1/36G01F25/00
CPCG01F1/44G01F1/36G01F25/10
Inventor 孟洋
Owner NUCLEAR POWER INSTITUTE OF CHINA