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 correlations 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. (C) 2011 Elsevier B.V. All rights reserved.