NCBJ is currently engaged in the development of the High-Temperature Gas-cooled Reactor (HTGR) project. The technical description of the facility is being prepared as part of a grant from the Polish Ministry of Education and Science. The initial outcome of this project is the conceptual design, which was was finished in December 2022 and publically announced  in June 2023. According to the team from the Division of Nuclear Energy and Environmental Analyses at the National Center for Nuclear Research, the HTGR-POLA reactor, utilizing helium as a coolant, will provide a thermal power output of 30 MWth (thermal).

The basic technical parameters of the HTGR-POLA are as follows:

  • The reactor core is of the prismatic type, composed of hexagonal blocks. Graphite will serve as the moderator.
  • The reactor will utilize TRISO fuel with preliminary range of 8-12% enrichment (HALEU – High-Assay Low-Enriched Uranium). Detailed specifications will be determined during the development of the basic design, after conducting final simulations.
  • The maximum thermal power output of the reactor is 30 MWth.
  • The reactor will operate in an open fuel cycle. The primary cooling circuit will employ forced circulation helium at a pressure of 6 MPa. The helium temperature at the outlet of the reactor will be 750 ⁰C, while at the inlet, it will be 325 ⁰C. The reactor will feature a secondary water-steam cooling circuit operating at a pressure of 13.8 MPa.
  • The external dimensions of the reactor pressure vessel are 4.1 m in diameter and 12.3 m in height.
  • Both passive and active safety systems will be implemented. Reactivity control will be achieved through a specific system of control elements (rods), the use of reactivity reducing materials (poisons) permanently housed in the fuel blocks, and reserve reactivity reducing capsules.
  • Cogeneration operation is possible, with a maximum gross electric power output of 10 MWe and the capability to produce high-temperature steam at 540 ⁰C.
  • For industrial processes, the reactor can deliver 17 MWth with a maximum capacity of 25 t/h, and it can generate low-temperature steam for municipal purposes at maximum 16.5 MWth.
  • The planned lifetime of the reactor is 60 years.

The project is being developed in collaboration with the Japan Atomic Energy Agency (JAEA), which possesses its own HTTR gas-cooled high-temperature reactor. The conceptual design serves as the starting point for the development of the basic design which is now processing, and will include safety analyses and tests of the reactor's construction materials.

Research HTGR project

Operational and Safety Analyses of Nuclear Reactors, and Environmental Impact Studies

Scientists at NCBJ are developing computational tools for safety analyses and optimization of the operation of light-water power and research reactors, as well as high-temperature reactors. They conduct operational and safety analyses and prepare safety assessments for nuclear power plants. Their research also encompasses thermal-hydraulic issues, fuel cycles, fuel recycling and transmutation, and severe accidents. The dedicated Hazard Analysis Center MANHAZ specializes in advanced Computational Fluid Dynamics (CFD) simulations, modeling the dispersion of contaminants in the environment, inverse problem-solving, probabilistic safety analyses, and reliability analyses especially for HTR. The MARIA reactor serves as a testing ground for some nuclear reactor analyses conducted at NCBJ.