[en] This document provides a set of engineering data for an international benchmark analysis in the area of coolant thermal hydraulics of fast breeder reactors. The target phenomenon is temperature stratification of liquid sodium in the reactor vessel upper plenum of Japanese Prototype Fast Breeder Reactor 'Monju', observed at a turbine trip test from 40% power output conditions conducted in December 1995. Structural integrity of sodium-cooled fast reactors should be secured against high temperature conditions and against sharp temporal and spatial temperature gradient due to high thermal conductivity and to small specific heat of liquid sodium. This brings special importance of single phase liquid thermal hydraulics on FR designs. Japanese prototype fast breeder reactor Monju is now a precious facility capable to provide actual information on crucial issues on FR safety to other FR developing countries. Japan Nuclear Cycle Development Institute (JNC, predecessor of Japan Atomic Energy Agency) sought for fruitful way of contribution to the IAEA/TWG-FR member states for their FR development effort since a technical meeting in Tsuruga, 1st and 2nd, Dec. 2004, to reach conclusion that provision of opportunity of benchmark analysis of thermal stratification in Monju reactor vessel upper plenum is one of the best proposals along with discussions in TWG-FR annual meetings in 2005 and 2006, considering that reactor vessel is a common component among various plant configurations of FRs, and that thermal stratification inside plenum is also a common safety issue among various FR designs. This proposal was approved at the IAEA General Conference in September 2007. The set of engineering data consists of the 'Monju' plant major features, outline of the turbine trip test at December 1995, geometry of the reactor vessel upper plenum, and flow inlet boundary conditions at the reactor core top surface. (author)