Nuclear Engineering and Design 241 (2011) 3898–3909 Contents lists available at ScienceDirect Nuclear Engineering and Design jo u r n al hom epage : www.elsevier.com/locate/nucengdes Pressure drop modeling and comparisons with experiments for single- and two-phase sodium flow A. Chenu a,b,∗ , K. Mikityuk a , R. Chawla a,b a Paul Scherrer Institute (PSI), 5232 Villigen PSI, Switzerland b Ecole Polytechnique Fédérale de Lausanne (EPFL), 1015 Lausanne, Switzerland a r t i c l e i n f o Article history: Received 21 May 2010 Received in revised form 16 June 2011 Accepted 5 July 2011 a b s t r a c t The thermal-hydraulic code TRACE is currently being extended at the Paul Scherrer Institute (PSI) for enabling the study of sodium-cooled fast reactor (SFR) core behavior during transients in which boiling is anticipated. An accurate prediction of pressure losses across fuel bundles – under both single- and two- phase sodium flow conditions – is necessary in this context. The present paper addresses the assessment, and implementation in TRACE, of appropriate friction factor models for round tubes and wire-wrapped fuel bundles, as well as local pressure drop models for grid spacers. Validity of the implemented cor- relations has been confirmed via the analysis of a range of experiments conducted earlier at the Joint Research Centre, Ispra. The measurements utilized are those of single- and two-phase pressure loss for sodium flow in tubes and 12-pin bundles, as a function of the inlet velocity under quasi steady-state conditions. The reported study thus represents an important further development step for the reliable simulation of two-phase sodium flow in TRACE. © 2011 Elsevier B.V. All rights reserved. 1. Introduction The sodium-cooled fast reactor (SFR) is a fast reactor concept proposed by the Generation IV International Forum (GIF) and is considered, by several countries, as the prime candidate for the large-scale implementation of breeder reactor technology in the medium-term future. Since the void distribution in an SFR core is one of the principal parameters controlling the reactivity during hypothetical transients, and also since boiling may lead to dryout and the melting of material, the accurate modeling of sodium two- phase flow is an important requirement in safety studies for the SFR. For this purpose, appropriate extension of the thermal-hydraulic code TRACE is being carried out at the Paul Scherrer Institute (PSI) in Switzerland (Chenu et al., 2009). The work is being performed within the frame of the FAST project, which aims at the devel- opment of a unique computational code system for the transient analysis of different advanced fast reactor concepts (Mikityuk et al., 2005). Most fast breeder concepts use multi-pin fuel bundles with axial coolant flow, and an accurate prediction of the hydraulic losses in these bundle geometries is important for design calculation and safety analysis. The present paper reports on the modeling of the pressure drop in single- and two-phase sodium flow. A review of available correlations for modeling of the pressure drop in different ∗ Corresponding author at: Paul Scherrer Institute, 5232 Villigen PSI, Switzerland. E-mail address: aurelia.chenu@psi.ch (A. Chenu). geometries is first presented. Thereby, the two basically different types of spacers applicable to cylindrical fuel elements, i.e. wire and grid spacers, have both been considered. For two-phase flow, the classical approach of multiplying the friction factor for single-phase flow (at the same mass flux) by an empirical multiplier has been applied. In this context, a number of existing models available in the open literature have been selected and implemented in TRACE, and their predictive capability for sodium has been evaluated on the basis of a appropriate experiments. The sodium boiling experiments selected for assessing the valid- ity of the physical models were performed at the Joint Research Centre at Ispra, Italy, in the 1980s (Kottowski and Savatteri, 1984; Savatteri et al., 1986). Sodium flow characteristics were measured in tubes and pin bundles with identical geometries but differ- ent spacers (wire-wrapper or grid-spacers). A large experimental database was generated by the research program, consisting of single- and two-phase pressure loss measurements as a function of the velocity at the test-section inlet under quasi steady-state boiling conditions. The comparison of the experimental data with numerical predictions has enabled a quantitative estimation of the quality of the correlations, enabling recommendations to be made for future modeling. These recommendations are in terms of the single-phase friction factor for simple geometries, for wire-spaced bundles and across grid spacers, as well as the two-phase flow multipliers needed for both friction and grid-spacer pressure losses. The following section gives a summary of the tested models, while Section 3 describes the experiments currently used for ver- ification purposes. Results of the TRACE analysis are presented in 0029-5493/$ – see front matter © 2011 Elsevier B.V. All rights reserved. doi:10.1016/j.nucengdes.2011.07.009