A scalability study of one-dimensional two-phase flow drift-flux closure relations for use in RELAP5 rod bundles and new, well-scaled low liquid flow rod bundle data
Abstract
The present research is a detailed study of the scalability of one-dimensional two-phase flow drift-flux closure relations for use in RELAP5 rod bundle geometries. The scalability of constitutive relations from test to prototypic conditions is a fundamental problem of utmost importance in nuclear reactor thermal-hydraulics due to the popularity of computation fluid dynamic computer codes, such as RELAP5, to validate nuclear reactor safety systems. Of particular importance is the calculation of the water level in the reactor core, which is a function of the void fraction, during accident scenarios. The core must remain covered with coolant to remove the decay heat which continues to be produced, even after the reactor is shut down. To perform this study, an intensive literature survey is completed to collect all of the available drift-flux models and correlations developed for use in rod bundle geometries, as well as available data. Each drift-flux relation is reviewed in detail. A detailed review of the calculation scheme for the interfacial friction in the reactor core in RELAP5 is presented. Based on the importance of the length scale in two-phase rod bundle systems, the geometric scalability of the gathered data is determined. Using data determined to be geometrically scalable, the predictive capability of each of the drift-flux correlations is assessed. The correlation by Chexal and Lellouche (1986), which is currently implemented into the water level determination scheme in RELAP5, is evaluated using the code scaling, applicability, and uncertainty methodology in the context of a small break loss of coolant accident in a prototypic four loop pressurized water reactor. The physical and geometric dependencies of this correlation are reviewed. The correlation by Chexal and Lellouche (1986) is found to have acceptable results predicting the void fraction in a rod bundle system when used in the context of the drift-flux general expression, however, individually the distribution parameter and drift velocity calculated from this correlation have difficulties accurately representing the physics of the two-phase system. Due to the deficiencies identified in the correlation by Chexal and Lellouche (1986) and the poor predictive capabilities or limitations of the other correlations studied, new adiabatic area-averaged void fraction data in low liquid flow conditions at atmospheric pressure is collected in a well-scaled 8 × 8 rod bundle test section. This data is utilized along with the drift velocity from Hibiki and Ishii (2003) to back calculate the distribution parameter from the drift-flux general expression. The distribution parameter obtained shows that for increasing liquid superficial velocities, the distribution parameter behaves more like that in pipe flow as described by Ishii (1977). As an extension of the present research, this new data can be employed to extend the correlation by Chen et al. (2012b) and replace the correlation by Chexal and Lellouche (1986) in RELAP5 rod bundle geometries. The new correlation should allow for more accurate and fundamentally sound results in calculating the interfacial friction, and hence the water level, in the reactor core.
Degree
M.S.E.
Advisors
Hibiki, Purdue University.
Subject Area
Nuclear engineering
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