Understanding and Prediction of Thermohydraulic Phenomena Relevant to Supercritical Water Cooled Reactors (SCWRs) - Final Report of a Coordinated Research Project

FOREWORD The supercritical water cooled reactor (SCWR) is an innovative concept that operates at a pressure higher than the thermodynamic critical point of water, allowing the core outlet coolant temperature to be much higher than that of the current generation of WCRs. The key technological advantages of the SCWR include its high thermal efficiency and simplified system configuration compared with conventional WCRs. There has been a high level of interest in research and development of SCWRs in several Member States. In 2007, the IAEA started the coordinated research project (CRP) entitled Heat Transfer Behaviour and Thermo-hydraulics Code Testing for Super-critical Water Cooled Reactors (SCWRs), which promoted international collaboration among 16 institutes from 9 Member States and 2 international organizations. The CRP was successfully completed in September 2012. Information generated from that CRP was documented in numerous IAEA publications and reports. A database of thermohydraulic parameters of interest to SCWR development was compiled and is housed in the Nuclear Energy Agency’s central server. After the completion of the CRP, collaboration continued between several participating institutes. Most of these institutes expressed their strong interest in and support for a new CRP on thermohydraulics of SCWRs to continue the momentum of international collaboration. The overall objective of this second CRP, which started in 2014, was to improve the understanding of thermohydraulic phenomena and the prediction accuracy of thermohydraulic parameters related to SCWRs and to benchmark numerical toolsets for SCWR thermohydraulic analyses. Scientific investigators from participating institutes identified specific research objectives to improve the predictive capability of key technology areas (e.g. heat transfer and pressure drop for SCWR fuel related geometries, parallel channel stability boundary, natural circulation flow, critical heat flux at near critical pressures, critical flow, subchannel and plenum mixing). The predictive capability of subchannel codes and computational fluid dynamic tools was assessed through benchmarking exercises for heat transfer in tubes, annuli and bundles as well as pressure drops in annuli and bundles. In total, 12 institutes from 10 Member States and 2 international organizations were involved in this second CRP, which was completed with the planned outputs in 2019. The present publication provides the background and objectives of the CRP; descriptions of a revised Canadian SCWR design concept and a new SCWR design concept being developed at the Nuclear Power Institute of China; updated information on key technology areas (e.g. heat transfer in simple geometries, stability and critical flow) obtained since the completion of the previous CRP; new experiments and data on supercritical heat transfer in bundles and on critical heat flux; and application of the direct numerical simulation approach for supercritical heat transfer. Results of three benchmarking exercises with subchannel codes and computational fluid dynamic tools are also presented, to demonstrate successes and show areas for further improvement. Experimental information and data were contributed by participating institutes through close collaboration. This publication illustrates the state of the art of SCWR research and development. It is expected to be a key supporting publication for researchers and engineers pursuing the development of SCWRs or equipment/components operating at supercritical pressures. The IAEA is grateful for the contributions of the chief scientific investigators of all participating institutes and of the CRP chairpersons, L. Leung (Canada) and W. Ambrosini (Italy). In particular, the efforts of L. Leung to collect and organize the contributions and to continuously encourage progress in the work are gratefully recognized. The IAEA officers responsible for this publication were T. Jevremovic and K. Yamada of the Division of Nuclear Power.