[en] The Indian Pressurized Heavy Water Reactor (IPHWR) is a heavy water moderated and heavy water cooled nuclear reactor which employs short fuel bundles stacked one after another inside each of the horizontal fuel channels. The 540 MWe reactors, under operation in India, employ 37- rod bundles. Upcoming IPHWR designs of 700 MWe rating allow boiling of the coolant at the exit of channels. Therefore, there is a two-phase flow near at the exit of fuel channel, which causes the pressure drop to increase in the channel due to its two-phase conditions. In the absence of two-phase pressure drop data for 37-rod bundle, it is of prime interest and requirement to estimate the pressure drop across 37- rod fuel bundle under two-phase conditions. The experiments have been conducted in existing Boiling Water Loop (BWL) for different flows and qualities at a pressure of 70 bar across the simulated fuel channel of 700 MWe IPHWR. The data generated for pressure drop across fuel bundle have been analyzed to evaluate the two-phase friction multipliers. This paper deals with the experimental set-up, the experiments carried out and the analysis of the results in details. (author)