Commercial Services > Nuclear Regulation Support (TSO)

Nuclear Regulation Support (TSO)

Mission of the Section

The mission of the Nuclear Regulation Support Section (TSO) of the State Office for Nuclear Safety (SUJB) is to provide state authorities with technical and professional support, which is independent of organizations providing safety reports for nuclear power plants (NPPs). Services and research activities are aimed at supporting the state administration, particularly SUJB, in the field of nuclear safety and nuclear power. This support is particularly realized by expert review of safety analyses reports presented to SUJB by Dukovany and Temelín NPPs.

Services offered

Expert technical support in decision-making and licensing of nuclear power plants, provided on the basis of safety documentation submitted by the licensee in the areas of:

  • thermal hydraulic characteristics
  • safety analyses
  • in the related parts of the document entitled Limits and Conditions

We also provide the following independent safety analysis in accordance with the recommendations of the International Atomic Energy Agency along with the following related development activities:

  • development of qualified thermal hydraulic models of nuclear power plant units for system computational codes on the basis of the results of commissioning and experiments
  • development and testing of a best-estimate method and uncertainty evaluation for the implementation of deterministic safety analysis of nuclear power plant accidents

Realized projects and their description

National R&D projects:

2016 – 2015
Project of Ministry of Interiors CR: "Prevention, preparedness and mitigation of severe accidents of  Czech nuclear power plants in line with new findings of the stress tests after the accident in Fukushima"

CVR was  in the period April 2013 – December 2015 part of the research consortium of the project "Prevention, preparedness and mitigation of severe accidents Czech nuclear power plants in line with new findings of the stress tests after the accident in Fukushima"
CVR focus of the work consisted in the development of software for the analysis of severe accidents, namely the adaptation of computing platforms MELCOR for VVER 1000 reactors.

MELCOR is the second-generation computational tool for the analysis of nuclear reactor severe accidents, which replaced the original code STCP (Source Term Code Package) developed in the US since 1982. It was developed at Sandia National Laboratory for the needs of US NRC.
MELCOR is a fully integrated system computational code designed to modelling the course of severe accidents at nuclear power plants with light-water reactors. MELCOR is a general platform that must be adapted to the specific nuclear power plant design.

Developed software structure for VVER 1000 allows computational simulation of:
– thermal-hydraulic response of the cooling system and reactor containment
– gradual overheating of the reactor core and its degradation
– subsequently moving melt the fuel to the bottom of the pressure vessel
– the release of fission products and their spread

License to obtain and use the code can be obtained free of charge through the State Office for Nuclear safety (SUJB) and US NRC, subject of meeting the license conditions.
CVR and its research team is ready to provide expert support in adapting the platform and use the model VVER within their services offered.

Internal identification: CVR-TSO-02/15

Contact person: Miroslav Hrehor (+420 266 173 432)

The software is based on developed model of VVER 1000 nuclear power plant in the programming platform MELCOR for analysis of accidents with core melt. Development work has included creating a set of input data from operational and design characteristics of Temelin NPP, the setting up of initial inventory of the active core, its isotopic composition for medium burnout and especially nodalization technological systems and the reactor containment. Computational model allows to get detailed information about the timing of selected accident scenarios. Calculations of accident scenarios, whose common feature was long complete loss of electric power plant (station blackout), provided the key parameters characterizing the accident, such as temperature and pressure inside the reactor pressure vessel and in the containment, time evolution of gases (e.g. hydrogen) and in particular the mass and dynamics of fission products released into the containment (the source terms).


"Development and validation of thermal-hydraulic models of NPP units with VVER reactors for the purpose of conducting independent safety analyses".

"Development and validation of the method of best-estimate thermal-hydraulic NPP model for deterministic safety analyses".

International projects:

"The project Bemuse Programme (BEMUSE: Best Estimate Methods, Uncertainty and Sensitivity Evaluation)".

The Project, carried out in the framework of the OECD/NEA, was focused on the application of computational methods for uncertainty evaluation in loss of coolant accidents in a pressurized water reactor. Its aim was to evaluate the usability, quality and reliability of best-estimate methods, including uncertainty and sensitivity evaluation, in applications addressing the safety of nuclear reactors. In addition, its aim was to promote and facilitate the application of uncertainty methodologies in the licensing process for the regulatory authorities as well as for the industry

"PSB-VVER project on experimental modelling of loss of coolant accidents for VVER-1000 reactors"