Theory and implementation of nuclear safety system codes – Part I: Conservation equations, flow regimes, numerics and significant assumptions

• We compare the conservation equations for multiple nuclear system codes.
• We discuss the two-phase flow regimes used in the nuclear system codes.
• Details of the numerical methods and code assumptions are provided.
• The capabilities of the codes are evaluated for new nuclear power plant systems.
• The information provided aids in the selection of appropriate system codes.

The design and analysis of the thermal/hydraulic systems of nuclear power plants necessitates system codes that can be used in the analysis of steady-state and transient conditions. Due to the dispersed development of system codes over many laboratories and universities, there are several system codes available for use. Many of the available codes have multiple similar versions developed for specific user needs.The code comparisons provided in the two parts of this article series allow users to select the appropriate system code for their specific problems. In this comparison, the governing equations for mass, momentum and energy conservation are evaluated. It will be shown that the governing equations do not vary substantially between the codes considered. Most of them utilize a lumped approach with only two fields to represent two phase flow.Two-phase flows are divided into flow regimes based on their appearance and the flow structure. The regimes are used to select appropriate closure relationships to model heat transfer, interfacial drag, and other flow conditions.In addition, major assumptions about the governing and closure equations in these codes are compared and discussed. The most significant of the assumptions is that the governing equations can be discretized in time.The numerical approach of the codes is compared to one another since the numerical approach not only affects the speed of the system codes but also the accuracy of the results.In the second part of this article, the closure relations, their major assumptions, experimental verification and validation are discussed.The results of these articles also guide the development of these system codes, the underlying thermal/hydraulic models, and indicate areas where models must be improved to adequately address issues with new reactor design and development activities.