The NEA Working Group on the Analysis and Management of Accidents convened for its annual meeting to discuss ongoing work related to accident strategy and management.

NEA Working Group on Analysis and Management of Accidents (WGAMA) initiated this activity focusing on computational fluid dynamics (CFD) in nuclear reactor safety (NRS).

The eighth instalment of the workshop on computational fluid dynamics for nuclear reactor safety (CFD4NRS-8) was organised by the Nuclear Energy Agency (NEA) Committee on the Safety of Nuclear Instal...

The 9^th edition of the Computational Fluid Dynamics for Nuclear Reactor Safety workshop was hosted by Texas A&M University on 20-22 February 2023. 

Prevention, mitigation and management of potential accidents in nuclear power plants

When looking for a validation experiment for CFD application to reactor simulation for either design or safety studies, it appears that available data often suffer from a lack of local measurements, ...

The Workshop on Experimental Validation and Application of CFD and CMFD Codes to Nuclear Reactor Safety Issues, CFD4NRS-7 was the seventh in a series of workshops focusing on Computational Fluid Dyna...

Single-phase computational fluid dynamics (CFD) is used more and more for design and safety issues related to light-water reactor (LWR) thermal hydraulics. Over the past ten years, the Working Group ...

In May 2002, an exploratory meeting of experts to define an action plan on the application of Computational Fluid Dynamics (CFD) Codes to nuclear reactor safety problems was held in Aix-en-Provence, ...

Following recommendations made at an exploratory meeting of experts to define an action plan on the application of Computational Fluid Dynamics (CFD) codes to nuclear reactor safety (NRS) problems, h...

This document is an extension of the NEA/SEN/SIN/AMA(2006)2 report produced by the Writing Group 3 on the extension of CFD to two-phase flow safety problems, which was formed following the recommen...

This report initially provides a brief overview of the main features of a PWR, BWR, CANDU and VVER reactors. It also provides an overview of the ex-vessel corium retention (core catcher) and a genera...

Following the CFD4NRS workshops held in Garching, Germany (September 2006), Grenoble, France (September 2008) and Washington D.C., USA (September 2010), this workshop, held in Daejeon, Korea on 10-12...

The OECD/NEA-KAERI MATiS-H benchmark was initiated to test the ability of state-of-the-art computational fluid dynamics (CFD) codes to predict important turbulence parameters downstream of a generic ...

The OECD/SETH-2 project was carried out during the period 2007-2010 under the auspices of the Nuclear Energy Agency (NEA) and with support provided by the Czech Republic, Finland, France, Germany, Ja...

The OECD Nuclear Energy Agency developed an initiative to form writing groups of experts with the specific task of assessing the maturity of computational fluid dynamics (CFD) codes for nuclear reaco...

The OECD/NEA-Vattenfall T-Junction Benchmark was initiated to test the ability of state-of-the-art computational fluid dynamics (CFD) codes to predict the important parameters affecting high-cycle th...

This document is an extension of the NEA/SEN/SIN/AMA(2006)2 report produced by Writing Group 3 on the extension of computational fluid dynamics (CFD) to two-phase flow safety problems. Writing Group ...

Following recommendations made at an “Exploratory Meeting of Experts to Define an Action Plan on the Application of Computational Fluid Dynamics (CFD) Codes to Nuclear Reactor Safety (NRS) Problems”,...

The main objective of ISP-47 was to assess the capabilities of lumped parameter (LP) and computational fluid dynamics (CFD) codes in the area of containment thermal-hydraulics. Following the recommen...

In May 2002, an “Exploratory Meeting of Experts to Define an Action Plan on the Application of Computational Fluid Dynamics (CFD) Codes to Nuclear Reactor Safety Problems” was held in Aix-en-Provence...

Computational fluid dynamics (CFD) is to an increasing extent being adopted in nuclear reactor safety analyses as a tool that enables specific safety relevant phenomena occurring in the reactor coola...