Scheme of the PSB-VVER facility, EREC

The NEA PSB-VVER project provided unique and useful experimental data for code validation using the PSB-VVER test facility. The PSB-VVER facility is a large-scale thermal-hydraulics testing facility operated by the Electrogorsk Research and Engineering Centre (EREC) in Russia which represents the scaled-down layout of the Russian-designed pressurised water reactor, namely, VVER-1000. The PSB-VVER Project was intended to provide the experimental data needed to allow full validation of the computer codes used in the thermal-hydraulic analysis of VVER-1000 reactors*. 

VVER-1000 reactors have design characteristics that are similar to western-designed pressurised water reactors (PWRs). There are however important design differences which necessitate specific experimental facilities and data to ensure that the computer codes are validated and that the results of the thermal-hydraulic analyses are of acceptable quality for this type of reactor. 

Earlier NEA work established a code validation matrix for both LWRs and VVERs. This matrix is essentially a set of phenomena and experimental data that is used to validate thermal-hydraulic computer codes. This led to the general conclusion that the VVER-1000 matrix was not complete. Consequently, it was decided to develop the NEA PSB-VVER project to obtain the required experimental data not covered by the matrix.

The project provided unique experimental data needed for the validation of thermal-hydraulic codes and to support refinements to the safety assessment tools for VVER-1000 reactors. The PSB-VVER experiments represent a relevant extension of the existing code validation database. The generated data is of relevance to the VVER-1000 systems, but could potentially also be of interest to other types of pressurised water reactors.

Five experiments were executed, dealing with loss of coolant scenarios (small, intermediate, and large break loss of coolant accidents), a primary-to-secondary leak, and a parametric study (natural circulation test) aimed at characterising the VVER system at reduced mass inventory conditions.The scope of the project addressed the following areas:

• scaling effects
• natural circulation
• small cold leg break loss-of-coolant accidents
• primary to secondary leak
• 100% double-ended cold leg break.

Extensive pre- and post-test analyses accompanied the programme throughout the entire experimental series. 

A comparative analysis regarding the large break loss of coolant accident experiment can be found at Validation of Advanced Computer Codes for VVER Technology: LB-LOCA Transient in PSB-VVER Facility.

Related research on thermalhydraulics projects can be found at Integral Test Facilities and Thermal-Hydraulic System Codes in Nuclear Safety Analysis.

Project data

The project data package can be requested via the NEA Data Bank.

* Note: In Russian, vodo-vodyanoi energetichesky reaktor meaning water-water power reactor is transliterated as VVER.