NEA Rig-of-Safety Assessment (ROSA) Project

The NEA ROSA project aims to resolve issues in thermal-hydraulics analyses relevant to light water reactor (LWR) safety using the Japanese ROSA/LSTF facility. In particular, it intends to focus on the validation of simulation models and methods for complex phenomena that may occur during design basis events (DBE) and beyond-DBE transients.

The key objectives of the NEA ROSA project are to:

  1. Provide an integral and separate-effect experimental database to validate code predictive capability and accuracy of experimental thermal-hydraulic models. In particular, phenomena coupled with multi-dimensional mixing, stratification, parallel flows, oscillatory flows and non-condensable gas flows are to be studied.
  2. Clarify the predictability of codes currently used for thermal-hydraulic safety analyses as well as other advanced codes presently under development, thus creating a group among the OECD member countries who share the need to maintain or improve the technical competence in thermal-hydraulics for nuclear reactor safety evaluations. The experimental programme is defined to provide a valuable and broadly usable database to achieve these objectives.

When evaluating the safety of light water reactors, computer codes are used to simulate their behaviour during design basis events and beyond-DBE transients. This involves complex multi-dimensional single-phase and two-phase flow conditions, which can include non-condensable gas in many cases. Although current thermal-hydraulic safety analysis codes have a very high predictive capability (especially for one-dimensional phenomena such as flows in piping at high flow rates) there is a need for experimental work and code development and validation for these complex flow conditions. Given the increased use of best-estimate analysis methods in licensing, which is replacing traditional conservative approaches, the validation and quantification of uncertainties in the simulation models and methods is required.

Many experimental facilities have contributed to the thermal-hydraulic databases available today. However, most of current data are insufficient for future codes that incorporate multi-dimensional simulation capabilities, mainly because the spatial resolution of measurement is not enough to assess the simulation models and methods. The ROSA project will seek to address these issues.

The project consists of the following six types of ROSA large-scale experiments:

The programme contemplates a total of 12 tests, of which eight have been carried out so far. Four tests were performed in 2007, one on temperature stratification, one on water hammer and two on primary cooling through depressurisation. The remaining four tests were discussed by the project steering bodies, which defined the test initial and boundary conditions. They will be conducted in 2008 and in the first part of 2009. Project members also discussed the issues to be addressed in a possible follow up of the project, which is scheduled for completion in March 2009.

Project participants

The project is supported by safety organisations, research laboratories and industry from the following countries: Belgium, the Czech Republic, Finland, France, Germany, Hungary, Japan, Republic of Korea, Spain, Sweden, Switzerland, the United Kingdom and the United States.


A. Jasiulevicius, O. Zerkak, R.l Macian-Juan (2007), Simulation of NEA ROSA Test 6.2 Using TRACE, ICONE15-10192, Proceedings of the 15th International Conference on Nuclear Engineering (ICONE-15), Nagoya, Japan.

S. Marshall (2007), NEA ROSA Program, 19th Annual Regulatory Information Conference (RIC 2007), Washington, USA.

T. Takeda, Hideaki Asaka, M. Suzuki, H. Nakamura (2007), RELAP5 Analysis of ROSA/LSTF Vessel Upper Head Break Loca Experiment, 12th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-12), Pittsburgh, USA.

T. Watanabe (2007), Simulation of Temperature Stratification During Eccs Water Injection Using Fluent: Preparatory Analysis for OECD/NEA ROSA Project, 12th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-12), Pittsburgh, USA.

M. Suzuki, T. Takeda, H. Asaka and H. Nakamura (2007), Effects of three-dimensional steam flow on exit temperatures during core boil-off in PWR vessel top break LOCA simulation experiments by using ROSA/LSTF, 2007 Fall Meeting of the Atomic Energy Society of Japan, Kitakyushu, H03 (in Japanese).

Project period

April 2005 to March 2009


USD 1 million per year

Related links

Last updated: 22 September 2014