Nuclear reactor core channel analysis method based on two fluid models

By using two-fluid three-flow field model and second-order discrete method in the core channel analysis of nuclear reactor reactor, the problem of failure to effectively consider the droplet phase and difficulty in capturing transient working conditions in the prior art is solved, and high-precision calculation of the key parameter distribution of pressurized water reactor core is achieved.

CN119720671AActive Publication Date: 2025-03-28XI AN JIAOTONG UNIV
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Patent Information

Application Number
CN202411841543.6
Authority / Receiving Office
CN · China
Patent Type
Applications(China)
Current Assignee / Owner
Filing Date
2024-12-13
Publication Date
2025-03-28
Estimated Expiration
2044-12-13

AI Technical Summary

Technical Problem

The prior art fails to effectively consider the influence of the droplet phase in the sub-channel analysis of nuclear reactor cores, and adopts a first-order discrete format, making it difficult to capture the changing trend of key parameters under transient operating conditions.

Method used

The sub-channel analysis method based on the two-fluid three-flow field model is adopted to independently consider the continuous gas phase, the continuous liquid phase and the discrete droplet phase, and the control equation is discrete through the second-order discrete method composed of the central differential format and the second-order numerical flux limiting function of OSPRE.

Benefits of technology

High-precision calculation of key parameter distributions such as coolant flow, temperature, pressure and DNBR under stable/transient operating conditions of pressurized water reservoir core is realized, and the stability and accuracy of the calculation are improved.

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Abstract

The invention discloses a nuclear reactor core channel analysis method based on two fluid models, which comprises the following steps of: 1, establishing a fluid domain computational grid according to the structure of a nuclear reactor core or an assembly; 2, respectively establishing a control equation set and a heat conduction model of the fuel rod for the continuous gas-liquid phase and the discrete liquid drop phase; 3, considering physical phenomena such as flow heat exchange, resistance pressure drop and turbulence mixing in a reactor core, and implanting a corresponding constitutive model; 4, discretizing the control equation based on a second-order discrete format to obtain a discrete equation set; 5, dividing parallel tasks and parallel regions through an MPI parallel strategy to form a large coefficient matrix of a pressure equation; and 6, solving a discrete equation set and a large coefficient matrix through a pressure coupling semi-implicit algorithm and a stable biconjugate gradient method. According to the method, the key physical parameter distribution of the pressurized water reactor can be more accurately and reasonably calculated, so that technical guidance is provided for safety analysis and engineering design of the reactor.
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