The Benchmark for Uncertainty Analysis in Best-Estimate Modelling (UAM) for Design, Operation and Safety Analysis of Light Water Reactors (LWRs) is an international high-visibility benchmark for unce...

International Fuel Performance Experiments (IFPE) is the public domain database on nuclear fuel performance experiments for the purpose of code development and validation.

Understanding fuel performance is a key condition for successful reactor operation. Scientific knowledge in this domain was conceived by experimental tests encoded into software routines to ensure ac...

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-100...

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....

NEA Working Party on Scientific Issues of Reactor Systems (WPRS) annual meeting, 19-21 February 2020.

DATIF, the DATabase for IFpe, was first developed in 2018 in order to enhance accessibility to the IFPE legacy collection of integral fuel performance experiments data, jointly put together by the NE...

Questions relating to the space, time end energy distribution of neutrons and radiation in different media are the domain of reactor physics.

Average fuel burn-up in light water reactors (LWRs) has steadily increased with time as technological advances have been made. The practical limit is currently in the region of 50 GWd/t. The main dri...

This publication provides an introduction to minor actinide nuclear properties and discusses some of the arguments in favour of minor actinide recycling, as well as the potential role of thermal reac...

Sodium fast reactors offer the most promising type of reactors to achieve such Generation IV goals at a reasonable time scale given the experience accumulated over the years. However, it is recognise...

High-temperature gas-cooled reactors (HTGRs), also known as very-high-temperature reactors (VHTR) are Generation IV reactors that can operate at very high temperatures and use a graphite-moderated ga...

The high-temperature gas-cooled reactor (HTGR) is a Generation IV reactor concept that uses a graphitemoderated gas-cooled nuclear reactor with a once-through uranium fuel cycle. This design permits ...

The Expert Group on Reactor-based Plutonium Disposition (TFRPD) dealt with the status and trends of reactor physics, fuel performance and fuel cycle issues related to the disposition of weapons-grade...

Thermal hydraulics and mechanics deals with the physics and mechanics of the flow and energetic transfer of liquids, and its interactions with the structures around them in large complex systems, suc...

In the field of coupled neutronics/thermal-hydraulics computation there is a need to enhance scientific knowledge in order to develop advanced modelling techniques for new nuclear technologies and co...

In the field of coupled neutronics/thermal-hydraulics computation there is a need to enhance scientific knowledge in order to develop advanced modelling techniques for new nuclear technologies and co...

Based on the success of the OECD/NRC BFBT Benchmark, Japan released data based on the NUPEC PWR Sub-channel and Bundle Tests (PSBT) Benchmark and asked Pennsylvania State University (PSU) in the Unit...

The need to refine models for best-estimate calculations based on good-quality experimental data has arisen for various nuclear applications. One of the most extensive and valuable databases availabl...

The objective of the benchmark was twofold. First, the benchmark aimed to evaluate currently available computational approaches in an effort to understand the strengths and weaknesses of current ther...