The equipment cooling water heat exchanger is an important device in the component cooling water system. In this paper, the Bell-Delaware method was applied to achieve a preliminary design of equipment cooling water heat exchanger and a very mature design software HTRI6.0 was used to check the results. On the basis of HTRI6.0 results, the relative error of the total heat transfer coefficient and the total heat transfer area were 1.35% and 0.2%, respectively. The effective length of the heat exchanger tube and the number of the baffle plates are consistent. The design result of the Bell-Delaware method meets the actual heat transfer requirements and the heat exchanger design satisfies the requirements of vibration.
Three dimensional PWR-core analysis code CORAL developed by Wuhan Second Ship Design and Research Institute, which provides all functions required by PWR-core analysis calculation. These functions are neutron diffusion within the core and reflector, macroscopic depletion or microscopic depletion calculation analysis, multi-channel or sub-channel thermal-hydraulic analysis, one-dimensional heat transfer from nuclear fuel to the coolant, critical search by boron concentration or control rod position, integral and differential worth of neutron absorbers, neutron kinetics parameters for transient analysis, in-core neutron detector response simulation etc. CORAL is convenient to update and maintain in consider of modular, object-oriented programming technology. In order to verify the computational capabilities of the reactor core analysis code, the BEAVRS benchmark is adopted as the research object. The BEAVRS problem is a benchmark problem based on real commercial PWRs with detailed material description, geometric information, core operating history, and detector measurement data. In this paper, the CORAL code is used to carry out physical analysis and calculation for the BEAVRS benchmark, and the response rate distribution of radial and axial detectors can be obtained. By comparing with the measurement results provided by the benchmark, it can be found that the calculation results of the CORAL code are in good agreement. This shows that the CORAL code can well simulate the detector response distribution in the core.
It is well know that the two-fluid single pressure model is currently widely used to analyze reactor transient accidents such as LOCA. Current mainstream reactor safety analysis codes such as RALAP5 and CATHARE are based on this single pressure model. However this model has been proved to be ill posed in the sense that the equation system is non-hyperbolic which will lead to numerical unphysical oscillation. Currently reactor safety analysis codes use the interfacial pressure and virtual mass force to solve the ill-posed non-hyperbolicity issue. The development of China’s self-owned reactor high fidelity system analysis code has attracted much attention in recent years, and studying on the ill-posedness of the two-fluid model and the improvement of the ill-posedness is an important basis to analyze these reactor accidents. In this paper, the ill-posedness regions of the two-fluid single pressure model is first investigated using the eigenvalue analysis method based on the Cauchy problem with initial conditions. Then the effect of the virtual mass force and the interfacial pressure is studied by this eigenvalue analysis method, and three types of virtual mass term in the momentum equation are discussed. The results show that the appropriate virtual mass force and interfacial pressure can well improve the ill-posedness of the single pressure model, and the appropriate combination of them can significantly improve the ill-posedness.
Natural circulation refers to the phenomenon of fluid circulation driven by the driving force formed by the density difference and height difference. Natural circulation is very common in nuclear power system, which has an important impact on reactor safety. Therefore, accurate simulation of natural circulation phenomenon is one of the most important research contents in current reactor thermal hydraulic analysis. The low order difference method is widely employed to discretize basic conservation equations for current main thermal-hydraulics analysis code such as RELAP5, CATHARE, TRACE, etc. However, the low order difference scheme has the characteristics of large numerical truncation error and low calculation accuracy when simulating natural circulation problem. In this paper, the high-order accuracy numerical algorithm for two-fluid two-pressure model is used to simulate Welander natural circulation problem. Numerical results imply that the higher order difference scheme can accurately predict the unstable boundary and chaotic behavior with fewer grids, while the lower order difference scheme may get wrong results when the mesh is not fine enough, that is, the unstable mode of natural circulation can be predicted as stable mode. Therefore, high-order accuracy numerical algorithm could prevent excessive numerical diffusion effectively and improve the prediction accuracy, which demonstrates the advantage of using high-order schemes.
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