CN-121981013-A - Pressurized water reactor fuel bundle transient nuclear thermal coupling calculation method under two-phase flow condition

CN121981013ACN 121981013 ACN121981013 ACN 121981013ACN-121981013-A

Abstract

The invention discloses a pressurized water reactor fuel bundle transient nuclear thermal coupling calculation method under two-phase flow conditions, which comprises the steps of establishing a solid domain and fluid domain model and dividing grids; generating a few-group section library based on a Monte Carlo program, parameterizing the sections into a function of temperature and density by using Python interpolation, introducing grids into a CFD solving platform, and selecting a two-phase flow thermal hydraulic model. And a neutron diffusion equation is discretely solved by adopting a user-defined scalar equation interface, a source term, a diffusion term, a transient term coefficient and a neutron physical boundary condition are defined by utilizing a user-defined function interface, the section parameters are updated in real time according to the local temperature and the density, and meanwhile, a power heat source is calculated according to neutron flux and fed back to thermal hydraulic calculation, so that bidirectional coupling is realized. And (3) calculating a steady-state characteristic value, applying disturbance to perform transient calculation after convergence, and outputting power, temperature, density, cavitation share and neutron flux space-time distribution after the termination criterion is met, so that the method is used for safety analysis of the nuclear reactor.

Inventors

SU GUANGHUI
ZHANG SHEN
LIU RUITONG
WU YINGWEI
ZHANG JING
HE YANAN

Assignees

西安交通大学

Dates

Publication Date: 20260505
Application Date: 20260205

Claims (1)

1. A method for calculating transient nuclear thermal coupling of a pressurized water reactor fuel rod bundle under two-phase flow conditions is characterized in that a unified grid is adopted in a CFD solving platform to synchronously solve a neutron diffusion equation and a thermal hydraulic control equation so as to reduce errors caused by data transmission and grid mapping and facilitate transient calculation; The method comprises the following steps: Establishing a three-dimensional geometric model containing a fuel rod bundle and surrounding coolant flow channels, and carrying out grid division on the three-dimensional geometric model to obtain a calculation grid for subsequent coupling solution; The method comprises the steps of establishing a three-dimensional geometric model of a fuel rod bundle and surrounding coolant flow channels consistent with the first step in a Monte Carlo program, carrying out Monte Carlo calculation under a plurality of groups of temperature and density conditions covering a target working condition range to obtain few group section data under the corresponding working condition and form a few group section database, constructing the few group section into a continuous function form of temperature and density by adopting a Python interpolation method based on the few group section database, and acquiring few group section parameters in real time according to local temperature and density in a nuclear thermal coupling calculation process to realize the feedback effect of thermal hydraulic parameters on neutron physical characteristics; The method comprises the steps of establishing a CFD solution platform, embedding a neutron diffusion equation in the CFD solution platform to realize nuclear thermal coupling analysis of a fuel rod bundle, adopting a user-defined scalar equation interface (UDS) framework of the CFD solution platform to solve the neutron diffusion equation, mapping the neutron diffusion equation to the UDS through equation equivalent transformation, namely, treating transient terms corresponding to the UDS transient terms, diffusion terms corresponding to the UDS diffusion terms and other terms as UDS source terms, realizing transient term coefficients, diffusion coefficients and source term coefficients of the UDS by a user-defined function interface (UDF), and combining continuous functions of a few-group section, temperature and density constructed in the step two to update neutron physical parameters in real time according to local temperature and density in the calculation process, defining transient term coefficients, diffusion coefficients, source term coefficients and boundary conditions corresponding to the UDS through the UDF interface, and loading the UDF to the CFD solution platform to participate in the solution so as to realize synchronous solution of the thermal hydraulic control equation and the neutron diffusion equation on the same calculation grid; Selecting a thermodynamic hydraulic calculation model matched with a calculation working condition in the CFD solution platform, and starting UDS (universal description space service) to provide a calculation frame for solving a neutron diffusion equation in consideration of the fact that the fuel rod bundle has high power density and the coolant is supercooled and boiled, so that a two-phase flow model is required to be started in the thermodynamic hydraulic calculation model to represent a boiling-related phase change and heat transfer process; Setting material physical parameters of a solid domain and a fluid domain in a CFD solving platform respectively, setting thermal physical parameters of density, specific heat capacity and heat conductivity coefficient of the solid domain, setting parameters of density, specific heat capacity, heat conductivity coefficient, viscosity, molar mass and standard state enthalpy of the fluid domain, and setting diffusivity parameters defined in UDF and related to UDS solving; Setting a calculation domain boundary condition in a CFD solving platform, initializing global standard, starting steady-state nuclear thermal coupling calculation, realizing bidirectional feedback coupling of neutron physics and thermal hydraulic power in a steady-state solving process, and iteratively solving until a preset convergence criterion is met; And step seven, after the steady-state nuclear thermal coupling calculation reaches convergence, storing the calculation result and carrying out post-processing analysis, starting transient nuclear thermal coupling calculation after disturbance is introduced, maintaining the two-way feedback coupling of neutron physics and thermal hydraulic power in the transient solving process, and stopping calculation and outputting the result after the set physical time or the preset termination criterion is reached.

Description

Pressurized water reactor fuel bundle transient nuclear thermal coupling calculation method under two-phase flow condition Technical Field The invention belongs to the technical field of pressurized water reactor fuel bundles and nuclear thermal coupling, and particularly relates to a pressurized water reactor fuel bundle transient nuclear thermal coupling calculation method under a two-phase flow condition based on a CFD solving platform. Background Pressurized water reactor is the commercial nuclear reactor type with the most widely applied world currently, and has the characteristics of mature technology, rich operation experience, strong engineering adaptability and the like. With the development of core design to high burnup, high power density and compactness, the thermal margin of the fuel assembly is reduced under normal operation and certain transient working conditions, and two-phase flow phenomena such as supercooling boiling and the like of the coolant can occur in a local high heat flow density region, so that the flow heat exchange characteristic, cavitation share and density distribution are changed locally. In a pressurized water reactor core, a remarkable bi-directional feedback coupling relation exists between neutron physics and thermal hydraulic power, namely neutron flux influences reactor power density distribution, so that a heat source item of a fuel rod is determined, and further fuel temperature, cladding temperature, a coolant temperature field and a coolant density field are influenced, and on the other hand, the change of thermal hydraulic power parameters such as temperature, density, void share and the like can cause neutron section change, so that neutron flux distribution is influenced in turn. If the decoupling mode is adopted to respectively perform core physical and thermal hydraulic calculation, the real response under the condition of strong feedback is difficult to accurately reflect. Particularly, when local boiling, rapid temperature density change or transient disturbance is strong, a large error may be introduced in decoupling analysis, so that judgment on core safety margin and key parameters (such as power peak value, temperature peak value and cavitation share distribution) is affected, and therefore nuclear thermal coupling characteristic research is necessary for the pressurized water reactor fuel rod bundle. The nuclear thermal coupling analysis method can be divided into two general types, namely a cross-software/cross-program coupling calculation method, namely a core physical program and a thermal hydraulic program are adopted to solve respectively, parameter exchange is realized through an external interface, a script or an intermediate data file and iteration is carried out, and a nuclear thermal coupling method is realized in a unified calculation platform or the same solver framework, namely the solution of a neutron equation and a thermal hydraulic equation is completed in the unified platform. Compared with a cross-software coupling mode, the unified platform coupling can generally reduce errors and resource consumption caused by links such as grid mapping and data transmission among software, has shorter data transmission link and stronger coupling consistency, is more beneficial to developing transient calculation and improves calculation stability and robustness. From the existing research and engineering application conditions, the thermal model under the assumption of single-phase flow or equivalent single phase is still used as the main part in the pressurized water reactor fuel rod bundle nuclear thermal coupling analysis, and the research and the method for simultaneously considering two-phase thermal hydraulic power and reactor core physical feedback in a unified platform can be relatively less for the nuclear thermal coupling problem including two-phase flow conditions such as supercooling boiling and the like. Therefore, a method for realizing stable and reliable nuclear thermal coupling calculation under the condition of considering two-phase flow is needed to improve the analysis accuracy of steady-state and transient state working conditions of a high-power density pressurized water reactor fuel rod bundle. Disclosure of Invention In order to overcome the problems in the prior art, the invention aims to provide a pressurized water reactor fuel bundle transient nuclear thermal coupling calculation method under a two-phase flow condition based on a CFD solving platform. Aiming at the problems that under higher power density, two-phase flow phenomena such as supercooling boiling and the like can occur to a pressurized water reactor fuel rod bundle, decoupling analysis is difficult to accurately reflect actual physical effects due to strong feedback coupling of temperature/density/cavitation share distribution and neutron physical characteristics, grid mapping and data transmission errors are easy to generate in a cross-program coupling pro