Lead-alloys are very attractive nuclear coolant because of their low melting temperature, high boiling temperature, chemical stability and neutron transparency. In addition, lead-bismuth Eutectic (LBE) itself is a very efficient spallation target for neutron generation via a high-energy proton accelerator. Thus, lead and lead alloy coolants continue to be the subject of considerable research in the USA, Europe, and Asia as well as Russian Federation, focused on accelerator-driven transmutation systems and lead and lead-alloy cooled fast reactors (LFR) that are hereafter collectively designated as lead alloy-cooled advanced nuclear energy systems (LACANES).
Accurate characterisation of the thermal-hydraulic behaviours of those LACANES under natural circulation as well as steady-state forced convection is of critical importance for the system design development effort. While benchmarking of thermal-hydraulic loop models has been extensively carried out for sodium coolants, such a systematic effort had not been carried out in parallel for lead or lead-bismuth coolant. By utilising large-scale lead-alloy coolant loop test facilities, experimental data can be examined and qualified for use in benchmarking of these models. An expert group that addressed the major issues associated with the thermal-hydraulic benchmarking for LACANES had proven beneficial to all the interested parties.
The objectives of the benchmark study were to:
Thermal-hydraulic data sets for isothermal steady-state forced convection tests and non-isothermal natural circulation tests had been produced using the HELIOS (heavy eutectic liquid metal loop for an integral test of operability and safety) facility at the Seoul National University, Seoul, Rep. of Korea. Participants modelled the loop tests and compared the results with the produced data sets. Detailed information on the geometric and thermal-hydraulic configuration of HELIOS was first disseminated to participants so that modelling input parameters could be evaluated. An isothermal convection test run was predicted by each modelling participant. Then model results were compared with HELIOS isothermal test data. The same procedure was repeated for the case of natural circulation. Unresolved important issues, if encountered, were summarised.
The time when approval was obtained to start work was defined as t0. The subgroup worked for two years to achieve the results described above. The schedule was organised into four phases:
Meetings under the WPFC Expert Group on Heavy Liquid Metal (EGHLM) Technology
The Working Party on Scientific Issues of Advanced Fuel Cycles (WPFC) studies advanced nuclear fuel cycles, including fuel cycle scenarios, innovative fuels and materials, separation chemistry, waste disposal and coolant technologies.
The expert group monitored the feedback from version 0 of the Lead-bismuth eutectic (LBE) handbook, collected, analysed and checked the consistency of expected new results from ongoing heavy liquid metal-related programmes, and updated the handbook and released version 1.
Under the guidance of the Working Party on Scientific Issues of Advanced Fuel Cycles (WPFC), the Expert Group on Reactor Coolants and Components Technology (EGCoCoT) undertakes activities to collect, evaluate and preserve relevant scientific data and "translate" fundamental scientific understanding into application.