Rig-of-safety Assessment (ROSA) Project
Joint project

The Rig-of-safety Assessment (ROSA) Project aimed to resolve issues in thermal-hydraulics analyses relevant to light water reactor (LWR) safety using the Japanese rig-of-safety assessment/large-scale test facility (ROSA/LSTF). It intended to focus on the validation of simulation models and methods for complex phenomena that could occur during design basis events (DBE) and beyond-DBE transients.

ROSA phases

First phase (2005-2009)

The key objectives of phase one of the ROSA Project were to:

  • 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 incondensable gas flows were studied.
  • Clarify the predictability of codes used for thermal-hydraulic safety analyses as well as other advanced codes that were under development, thus creating a group among the OECD member countries who shared the need to maintain or improve the technical competence in thermal-hydraulics for nuclear reactor safety evaluations. The experimental programme was 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 incondensable 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 replaced traditional conservative approaches, 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 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 sought to address those issues.

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

  • temperature stratification and coolant mixing during emergency coolant injection
  • unstable and disruptive phenomena such as water hammer
  • natural circulation under high core power conditions
  • natural circulation with superheated steam
  • primary cooling through steam generator secondary depressurisation
  • two open tests defined by participants (one on pressure vessel upper-head break loss-of-coolant accident (LOCA) and another on pressure vessel bottom break LOCA, combined with accident management measures with symptom-oriented operator actions).

The programme contemplated a total of 12 tests, of which 8 had been carried out. 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 initial and boundary conditions for the test. They were 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, scheduled for completion in March 2009.

External publications

Second phase (2009-2013)

The ROSA-2 project aimed to resolve key light water reactor (LWR) thermal-hydraulics safety issues highlighted from the first phase of the project by using the ROSA/large-scale testing facility (LSTF) at the Japan Atomic Energy Agency.

In particular, the phase two focused on the validation of simulation models and methods for the following complex phenomena of high-safety relevance for thermal-hydraulic transients in design basis events (DBE) and beyond-DBE:

  • Generated a system-integral and separate-effect experimental database to validate predictive capability and accuracy of computer codes and models. Thermal-hydraulic phenomena coupled with multi-dimensional flows that might have included mixing, stratification, counter-current flows, parallel-channel flows and oscillatory flows were the main focus of the investigations.
  • Facilitated assessment of codes in use for thermal-hydraulic safety analyses as well as advanced codes under development including three-dimensional computational fluid dynamics (CFD) codes through active involvement of the project partners who maintained and improved the technical competence in thermal-hydraulics for nuclear reactor safety (NRS) evaluations.

The experimental program was intended to provide a valuable and broadly usable database to achieve the above cited objectives. Phase two consisted of six LSTF experiments that mainly included two groups of tests which were intermediate break loss-of-coolant accidents (LOCAs) and steam generator tube rupture (SGTR) accidents. One of the six experiments were directed to a different target following a discussion with the project partners. ROSA-2 ran from 01 April 2009 to 31 March 2013.

ROSA-2 members' area (password protected | reminder)


Belgium, Czech Republic, Finland, France, Germany, Hungary, Japan, Korea, Spain, Sweden, Switzerland, United Kingdom and United States.

Project period

ROSA: April 2005 to March 2009
ROSA-2: April 2009 to March 2013


ROSA: USD 1 million/year
ROSA-2: Not applicable