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CHF relational expression DNBR limit value statistical determination method based on grouping method

A deterministic and relational technique used in the field of thermal-hydraulic design and safety analysis of nuclear reactors

Active Publication Date: 2020-01-24
NUCLEAR POWER INSTITUTE OF CHINA
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  • Claims
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Problems solved by technology

[0005] The object of the present invention is a method for statistically determining the DNBR limit value of the CHF relational formula based on the grouping method, which solves the problem of strict, accurate and relatively conservative acquisition of the DNBR limit value in the CHF relational formula

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  • CHF relational expression DNBR limit value statistical determination method based on grouping method
  • CHF relational expression DNBR limit value statistical determination method based on grouping method
  • CHF relational expression DNBR limit value statistical determination method based on grouping method

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Embodiment Construction

[0079] The present invention will be described in further detail below in conjunction with the accompanying drawings and specific embodiments.

[0080] like figure 1 Shown, a kind of CHF relational formula DNBR limit statistical determination method based on grouping method, this method specifically comprises the following steps:

[0081] Step 1, collecting and obtaining the CHF experimental data of the fuel assembly;

[0082] Step 2. Obtain the M / P data of the position of the experimental burn point on the basis of the determined fuel assembly CHF relational expression;

[0083] Step 3, carry out Bartlett test to the M / P data of experiment burn point position;

[0084] Utilize the Bartlett test to check the homogeneity of variance of the M / P data (M / P: the CHF value predicted by the experimentally measured CHF value / CHF relational formula) of the burnt point position of the collected fuel assembly;

[0085]

[0086]

[0087]

[0088] Among them, ν t for sample ...

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Abstract

The invention relates to the technical field of nuclear reactor thermal hydraulic design and safety analysis, and particularly discloses a CHF relational expression DNBR limit value statistical determination method based on a grouping method. The method comprises the following steps: 1, acquiring CHF experimental data of a fuel assembly; 2, obtaining M / P data of an experiment burning point position; 3, carrying out a Bartlett test; 4, after passing the Bartlett test, carrying out homogeneity test of a data mean value; 5, carrying out normal distribution inspection by adopting an Epps-Pulley inspection method; 6, determining a DNBR limit value by using an Own criterion; 7, performing W-M-W inspection and K-W unilateral variance analysis respectively; 8, determining a unilateral 95 / 95 limitvalue of free distribution, and obtaining a DNBR limit value; 9, after the Bartlett test is not passed, carrying out a homogeneity test of a data mean value; 10, performing data grouping inspection; 11, determining a DNBR limit value. According to the method, a rigorous, accurate and relatively conservative CHF relational expression DNBR limit value can be obtained, key parameters can be calculated for CHF relational expression development and CHF experimental data evaluation, and the most concerned design limit value is provided for a nuclear safety department.

Description

technical field [0001] The invention belongs to the technical field of thermal-hydraulic design and safety analysis of nuclear reactors, and in particular relates to a method for statistically determining the DNBR limit value of a CHF relational formula based on a grouping method. Background technique [0002] The critical heat flux (Critical Heat Flux, CHF) relational expression is used to predict the critical heat flux value of the core, and the safety and economy of the core are closely related to the CHF relational expression. DNBR (Deviation from Nucleate Boiling Ratio) is defined as the ratio of the heat flux value calculated by the CHF relation to the local actual heat flux value. Accurately predicting the DNBR value of the reactor core is the core content of thermal-hydraulic design and safety analysis of the reactor, and is an important criterion for the safety of the reactor core in the steady-state thermal-hydraulic analysis and accident analysis of Type I, II and...

Claims

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Application Information

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Patent Type & Authority Applications(China)
IPC IPC(8): G06F17/18G21D3/00G21D3/06
CPCG06F17/18G21D3/005G21D3/06Y02E30/30
Inventor 刘伟杜思佳张渝张虹黄慧剑徐良剑刘余陆祺冷贵君李庆
Owner NUCLEAR POWER INSTITUTE OF CHINA
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