1 Introduction

This special section is dedicated to the research performed in the Research Center Řež (CVŘ Centrum Výzkumu Řež) in commemoration of successfully completing the Sustainable Energy Project (SUSEN) and many others. Research Center Řež is a research organization founded in 2002, which is a member of the UJV (former Nuclear Research Institute — Ústav jaderného výzkumu) Group (Fig. 1). CVŘ is focused on the precommercial research in power-generation technologies, predominantly, but not exclusively, in nuclear.

The CVŘ ambition is to extend the role of national nuclear-power generation research center, to contribute to a wide European and international efforts in sustainable nuclear fission and fusion technologies. Currently CVŘ is operating two nuclear research reactors dedicated to material research, and neutron physics and industrial applications. Its research capacity has been largely increased under the SUSEN project by building cutting edge research technologies and research teams. The company is also contributing to another European research infrastructure in France, the Jules Horowitz Reactor by delivering hot cells.

Fig. 1
Aerial view of CVŘ
Fig. 1
Aerial view of CVŘ
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With more than 350 professionals—scientists, engineers, technicians, and support staff, the organization is contributing to research and technology development in Generation II&III, Generation IV reactors and fusion, development in radiopharmaceuticals and achievements in regulatory support. As one of major knowledge focused centers in the Czech Republic in nuclear technologies, it is cooperating with a number of national and international research and technology organizations, universities, and professional networks to achieve safe, secured, and efficient power sources for the future.

The first nuclear research reactor in Czechoslovakia, and one of the first reactors in Europe, was built in 1957 in Řež. It speeded up the evolution of the Czechoslovak nuclear-development program which helped the country to achieve an outstanding position. Not only the national research center in Řež with a wide range of professionals was fully equipped for the technological research, but the industry, universities, and research institutions were capable to develop nuclear technologies, and the country became fully competent in taking an active part in building and operation of six nuclear power reactors under operation in todaýs Czech Republic and four in Slovakia. The Řež “nuclear valley” became a world-wide known place for research, technological innovation, and education in nuclear technology (Figs. 2 and 3).

Fig. 2
LR-0 reactor core, which is suitable for cross section measurements
Fig. 2
LR-0 reactor core, which is suitable for cross section measurements
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Fig. 3
LVR-15 reactor core during refueling
Fig. 3
LVR-15 reactor core during refueling
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Since 1950s and 1960s, the world has significantly changed and the nuclear technologies have changed as well. Highly focused on safety and reliability, dependent on smart designs and efficient operation, it creates new research challenges. The worldwide development of nuclear energy, particularly, nuclear-power generation has to respond to important challenges in next years and decades. As a robust low-carbon-energy source, it should play an important role in the energy mix, particularly, after the market finds a new equilibrium between the variety of technologies including energy storage and electricity-based transportation.

Severe accidents can be analyzed in detail thanks to the center infrastructure, which allows to take into account specific phenomena involved. Surveillance programs are guiding safe and reliable long-term operation of materials and components subjected to neutron irradiation. Generation IV technologies have been developing in last two decades to increase efficiency of electricity production by increasing the temperature of coolants, while decreasing substantially both the long-term radioactive-waste components and its volume.

Another example, small modular reactors, are likely to utilize results of many years of research of Generation IV technologies, enabling new reactor technologies to be affordable for investors and efficient and safe for operators and other stakeholders, including general public.

With the ITER project, the technology research is getting closer to the utilization of the fusion energy. To transfer decades of experience in nuclear technology to this new area, the Czech Republic needs to utilize adequate research infrastructure and teams. For all technologies, Generation II–IV and fusion, new experimental technology is required to enable adequate testing of structures, components, and materials under conditions as close as possible to real operation ones, to model and optimize thermal and thermal-hydraulic conditions of cooling media and components or to develop new methods of radioactive-waste treatment (Fig. 4).

Fig. 4
Cold crucible (open left, in operation right)
Fig. 4
Cold crucible (open left, in operation right)
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The project SUSEN has been built within the Research Center Řež in the close partnership with the University of West Bohemia under the grant of the EU Structural funds and the Czech Ministry of Education, Youth, and Sport financial contributions. It is opening new opportunities for the research at universities, research institutions, and for the industry. These opportunities have been highlighted by unique testing facilities built in Řež and Plzeň, by increasing attraction for professionals of all kinds of their seniority and nationalities and by opportunities for broad and fruitful international cooperation of Czech research institutions. The valuable synergies offered by connections between the SUSEN technologies and another large research infrastructure, the research reactors LVR-15 and LR-0, make the Řež valley even more a symbol of the current and future nuclear knowledge, technologies, and scientific and engineering cooperation in the Czech Republic and abroad.

The new infrastructure and teams will support all three important aspects of a robust nuclear power generation country. It serves as a fundamental basis for knowledge creation and retention, in close cooperation with the industry. It will support further development of methods and technologies for safe and reliable operation. And it will help to educate new generation of nuclear engineers and specialists to continue already 60 years long and excellent history of the Czech nuclear technologies.

SUSEN project comprises of four research programs:

  1. Technological Experimental Circuits Research Programme;

  2. Structural and System Diagnostics Research Programme;

  3. Nuclear Fuel Cycle Research Programme;

  4. Material Research Programme.

2 Technological Experimental Circuits Research Program

The program is focused on research and development of Generation IV nuclear reactors and nuclear fusion. The main research topics are: (1) thermodynamic and hydraulic properties of the coolants and (2) their interactions with structural materials and radiation fields.

To cover these research topics some large-scale experimental facilities have been commissioned. The gathered experimental data will extend the knowledge on materials behavior under specific conditions and will allow a deeper understanding of effects of different coolants on structural materials, as well as their thermodynamic and thermal-hydraulic properties.

Fig. 5
Hot-cell facility

Such data will improve calculation codes and enlarge the material-property databases as a necessary condition for further development of the considered types of advanced reactors. This program presents a strong cross-cut with the materials program.

The program includes:

  • Supercritical water—medium for the primary circuit of a supercritical water-cooled reactor.

  • Helium—medium for the primary circuit of a (very) high-temperature reactor ((V)HTR) and gas fast reactors (GFR).

  • Helium—coolant for the first wall and breeding blanket of a fusion reactor.

  • Supercritical carbon dioxide—potential medium of the secondary circuit for heat transfer from the primary circuit of Generation IV reactors.

  • Eutectic lead–lithium (Pb–Li) alloy—coolant and medium for generation of tritium as fuel in fusion reactor systems.

Under the program, important facilities have been built, e.g.,

  • supercritical-water loop with in-pile capabilities at the LVR–15 research reactor;

  • high-temperature helium loop with in-pile capabilities at the LVR–15 research reactor;

  • experimental facility for hydrogen generation through high-temperature water electrolysis;

  • test-blanket module to test maintenance and remote handling procedures of ITER;

  • testing facility for cyclic thermal stresses of ITER first-wall panels; and

  • high-temperature helium loop for thermal-hydraulic tests of a GFR.

The program has been aimed toward:

  • Coolant chemistry control, materials and fuel cladding research and development for gas-cooled high-temperature reactors (GFRs, VHTRs), supercritical-water-cooled reactor, and heavy-liquid-metal cooled reactors.

  • Component development and testing for gas-cooled high-temperature reactors and heavy-liquid-metal reactors, including the advance balance of plant options such as supercritical-CO2 power cycle.

  • Thermal hydraulic and physics experiments and simulations for fusion and Generation IV reactors.

  • Evaluation of materials and components for fusion technologies and development of manipulation techniques for fusion applications.

3 Structural and System Diagnostics Research Program

The program aims to R&D supporting long-term operation of the current generation of nuclear power plants (NPPs). It examines degradation of nuclear-reactor structural materials and component properties after long-term operational exposure as a critical input for an assessment of residual service life, reliability, and safety of nuclear reactors (Fig. 5).

The program covers:

  • Irradiated-material studies to characterize a degradation of mechanical properties of structural materials within nuclear reactors after long-term operation in complex, i.e., mechanical, corrosion testing, and postirradiation evaluation of irradiated materials.

  • Development of a new nondestructive evaluation (NDE) procedure for ferritic, austenitic, dissimilar weld joints, concrete structures and components of a complex configurations. NDE qualification using test pieces with artificial defects realistically simulating cracks. Development of universal hardware and software manipulator control systems for inspections in harsh environment (high temperature, steam, dust, or lower space).

  • Development of new procedures for verification of thermal and radiation resistance and properties of NPP's components under extreme accidental conditions. The LOCA (loss of coolant accident) Simulation Laboratory will develop new procedures for verifying thermal- and radiation-aged materials and components resistance and behavior under extreme conditions.

Under the program, important facilities have been built, e.g.,

  • Complex of ten hot cells for characterization of a degradation of structural and mechanical properties of nuclear reactors after long-term operation in complex, i.e., mechanical and corrosion mechanical testing and postirradiation evaluation.

  • Center of highly sensitive microstructural analyses to provide studies of degradation processes of irradiated material on the microstructural level.

  • Modern nondestructive testing laboratory, which provides complex solutions of metallic-material technical issues, concrete structures, and nuclear fuel including development of manipulator control systems for testing welds of critical nodes of the primary and secondary NPP circuits.

  • A laboratory for simulating LOCA in NPPs, including pre-exposure of components by modern gamma-irradiation laboratory, for achievement of high dose rates under temperatures between –180 °C and +400 °C simulating radiation conditions upon LOCA accidents, as well as studying materials for space applications (Fig. 6).

Fig. 6
LOCA test facility
Fig. 6
LOCA test facility
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The program has been aimed toward:

  • Obtaining a complete description of degradation characteristics and durability of structural materials of nuclear reactors after a long-term operational exposition that will serve for evaluation of durability of components, evaluation of safety and reliability.

  • Design, optimization, and production of irradiated test specimens for material characteristics research and a surveillance program (tensile testing, fracture toughness, crack growth rate in cyclic loading and at elevated temperature, fatigue and creep from samples of un-irradiated as well as highly irradiated material), including prolonging the service life.

  • Development of methodologies to determine the velocity of crack spreading during cyclic loading at temperatures ranging from ambient to 800 °C, developing and mastering fatigue tests at elevated temperatures (up to 800 °C) and its interaction with creep.

  • Development of new methods for NDE of concrete structures, ferrite, austenitic, and heterogeneous welded joints and components and their certification (NDE qualification) using test bodies with artificial imperfection and realistic operational crack simulations and mathematical modeling of measurement and discontinuities indication responses.

  • Development of new procedures for thermal and radiation resistance verification and behavior of structural materials and systems under the extreme conditions of severe accidents.

4 Nuclear Fuel Cycle Research Program

The program is focused on the R&D of the fuel cycle back-end. It consists of multiple areas, which may support nuclear-fuel cycle of Generation IV technologies, radioactive-waste management including its final disposal, severe accidents experimental studies, activities related to nonproliferation programs and the environmental impact. Nuclear-waste processing, including chemical aspects, is embedded to the program as well.

The program deals with:

  • Research and development of deep-repository support.

  • Development of analytical methods related to the need of deep-repository development and design.

  • Research and development of pyrochemical methods of spent-fuel recycling/reprocessing aimed to close the nuclear-fuel cycle.

  • Technology improvements for radioactive-material fixation for permanent depository acceptance.

  • Material innovation technologies (geo-polymer materials).

  • Modeling and investigation of nonstandard situations at nuclear-power reactors (in- and ex-vessel severe accidents).

Under the program, important facilities have been built, e.g.,

  • Cold crucibles enabling severe accident conditions modeling experiments.

  • Anaerobic laboratory to study conditions in the deep depository.

  • Experimental technological line and molten-salt-oxidation line for radioactive waste management research.

  • Molten salt loop to study conditions for the molten salt-based reactors.

The program has been aimed toward:

  • Research of the processes, which may occur in a deep repository of high-level radioactive wastes.

  • Modeling and evaluation of nonstandard events at nuclear-power reactors (in- and ex-vessel severe accidents).

  • Technology improvements for radioactive-material fixation for permanent depository acceptance.

  • Development of micro- and nano-analytical techniques for monitoring of nuclear installations and advanced material characterization.

  • Development of pyrochemical reprocessing of simulated spent nuclear fuel with fluoride volatility reprocessing method.

  • Material innovation technologies (geo-polymer materials) for nuclear-waste-repository applications.

5 Material Research Program

The program pursues evaluation of mechanical properties and microstructure of the new advanced materials under extreme conditions, namely, the behavior at high temperatures at static and dynamic stresses and in environments simulating operating conditions of power generation.

The program deals with:

  • Research and development of kinetics of corrosion mechanism.

  • Evaluating of mechanical properties in environments simulating operating conditions.

  • Mechanical behavior of power plants components during long-term operation.

  • Research and development of materials improved by advanced surface-modification techniques.

Under the program, new experimental technologies have been developed:

  • Laboratory for static- and cyclic-mechanical testing in corrosion environment (water, liquid metals, and molten salts).

  • Laboratory for chemical analysis.

  • Laboratory of surface and substructure analysis.

  • Laboratory for fusion welding.

The Program has been aimed toward:

  • Material research for the development of ferritic/martensitic steels.

  • Deployment of austenitic stainless steels and Ni alloys for components of Generation IV nuclear reactors and supercritical fossil power plants.

  • Mechanical behavior (fatigue, creep, and their interaction, stress corrosion cracking and corrosion fatigue) of structural material used or currently being developed for the application primarily in power engineering applications.

The new infrastructure built under the SUSEN projects significantly extended the research infrastructure of CVŘ. However, other fields are also pursued. Many of the papers presented in this special section are directly linked toward the SUSEN projects, but not exclusively. Other important CVŘ's research topics are covered as well.

The scientific content of the Řež Special Section of the ASME Journal of Nuclear Engineering and Radiation Science covers all major research areas of Research Center Řež. The important topic connected with the improvement of cross sections is presented in this issue by Schulc et al. and Burianova et al. Connected to this topic, Czakoj et al. introduced work devoted to the sensitivity and uncertainty analysis of the LR-0 reactor used for the averaged cross section measurements. The neutron spectrometry issues at the LVR-15 research reactor are described in paper of Kostal et al. Departments dealing with severe accidents form the modeling point of view contributed by (i) findings on VVER 1000 and (ii) suggesting new methodology of source term estimation in severe accidents in fusion installation, both presented by Mazzini et al. Experimental possibilities of LOCA-type accident analysis by cold crucible method are summarized by Rot et al., which deals with issues of aerodynamics as well in his other article. The aerodynamics issues are dealt also in the work of Duda et al. Groups representing material sciences are reporting about their specimen program (Hojna et al., Gavelova et al., Samek et al., corrosion (Galek et al. and Hadrava et al.) and defects investigation (Hodac et al.). The fuel-cycle back end and fuel management is dealt by Kopec and Kotowski papers about their deep-geological-repository research.


Václav Dostál, Guest Editor Scientific Director Research Center Rez


Michal Košťál, AE, Guest Editor Expert scientist in LR-0 Reactor Lab, Research Center Rez


Martin Schulc, Guest Editor Senior scientist in LR-0 Reactor Lab, Research Center Rez