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3 Journal of Power Technologies

3 2014

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Modeling of liquid film flow in annuli

Anglart H.

Journal of Power Technologies

2014

Vol. 94, nr 5

8--15

One of the challenges in thermal-hydraulic analyses of BWRs is correct prediction of dryout occurrence in fuel assemblies. In practical applications the critical powers in fuel assemblies are found from correlations that are based on experimental data. The drawback of this approach is that correlations are valid only for these fuel assemblies on which the experiments have been conducted. Other restrictive factors are the limited ranges of experimental working conditions including pressure, mass flux and axial power distributions. To overcome the above-mentioned limitations, several different approaches have been proposed to predict the dryout occurrence. One of them is to employ a phenomenological model of annular flow, in which the mass transfer between the liquid film and the gas core is based on entrainment and deposition correlations. Most of these correlations are derived from water-air flows in vertical tubes and their applicability to other geometries in general, and rod-bundles in particular, should be analysed. This paper presents an analysis of the entrainment rate in vertical annuli. Using the standard approach to calculate the entrainment rate, one can demonstrate that the results deviate from measurements. It has been shown that modifying the entrainment correlation based on data obtained in the annulus geometry leads to an essential improvement in the predictive capability of the phenomenological model of annular two-phase flow.

Efficiency of selected methods of hydrogen removal from a nuclear reactor's containment building

Bury T.

Journal of Power Technologies

2014

Vol. 94, nr 5

96--104

Large amounts of gaseous hydrogen may be released into the containment building during a severe accident in a water cooled nuclear reactor. The main methods of hydrogen removal from the containment are described in brief in this paper. HEPCAL - an in-house lumped parameter computer code - was used for simulation purposes and the results were used to evaluate the efficiency of various hydrogen removal systems.

Computer codes in the safety analysis for nuclear power plants. Computational capabilities of thermal-hydraulic tools, using the example of the RELAP5 code

Skrzypek E., Skrzypek M.

Journal of Power Technologies

2014

Vol. 94, nr 5

41--50

Safety is a paramount concern of the Nuclear Power Program in Poland. To this end there is a need to investigate the design of the proposed reactor and its operation principles and perform multiple analyses both before the reactor start-up (The Pre-Construction Safety Report (PCSR) and during its operational life. In the worldwide nuclear community hundreds of people are involved in this complicated and complex process. Due to the sophistication of the phenomena occurring during operation and accidents, the number of analyses is increasing rapidly. Currently, much interest in this field is focused on the use of computer codes and high computational power.