The expert group provided advice to the Working Party on Scientific Issues of Reactor Systems (WPRS) and the nuclear community on the scientific development needs (data and methods, validation experiments, scenario studies) of sensitivity and uncertainty methodology for modelling of different reactor systems and scenarios.
The main activity was focused on uncertainties in modelling light-water reactor (LWR) transients. In this context the objectives were to:
The goal of the Benchmark for Uncertainty Analysis in Best-Estimate Modelling for Design, Operation and Safety Analysis of Light Water Reactors (LWR-UAM) was to determine the uncertainty in light water reactor (LWR) systems and processes in all stages of calculations. It was estimated through a simulation process of nine exercises in three phases provided by the benchmarking framework.
Helium-cooled very high-temperature gas reactors are highlighted as a key technology with the potential to improve the competitiveness of nuclear energy within the Generation IV International Forum (GIF). Developing tools and methods to support this technology is seen as a priority by NEA member countries.
The Boiling Water Reactor Turbine Trip (BWRTT) Benchmark was established to challenge the coupled system thermal-hydraulic/neutron kinetics codes against a Peach-Bottom-2 (a GE-designed BWR/4) turbine trip transient with a sudden closure of the turbine stop valve.
Under the guidance of the Working Party on Scientific Issues and Uncertainty Analysis of Reactor Systems (WPRS), the Expert Group on Physics of Reactor Systems (EGPRS) will perform specific tasks associated with reactor physics aspects of present and future nuclear power systems.
This benchmark was a continuation of the V1000CT activities and defined a coupled code problem for further validation of thermal-hydraulics system codes for application with Russian-designed VVER-1000 reactors based on actual plant data from the Russian nuclear power plant Kalinin Unit 3 (Kalinin-3)
This benchmark incorporated full 3-D modelling of the reactor core into system transient codes for best-estimate simulations of the interactions between reactor core behaviour and plant dynamics and their testing on a number of transients of importance for plant behaviour and safety analysis.
The Subgroup on Uncertainty Analysis in Modelling (UAM) for Design, Operation and Safety Analysis of Sodium-cooled Fast Reactors (SFR-UAM) was formed to check the use of best-estimate codes and data.
The overall objective of the VVER-1000 Coolant Transient (V1000CT) Benchmark was to assess computer codes used in the safety analysis of water-water energetic reactor (VVER) power plants, specifically for their use in reactivity transients in VVER-1000.