R&D Sector for Nuclear and Structural Materials

Sector Activity Areas

  • organization and implementation of research and technological developments to create new fuel, absorber and structural materials for fuel rods and fuel assemblies of enhanced reliability, performance and economic efficiency for application in nuclear engineering and other industry sectors;
  • organization and implementation of material science research and technological developments related to structural materials (zirconium and its alloys, stainless steels, and other materials) and fabrication of products therefrom in support of NFC STE NSC design engineering developments for creation of enhanced-safety nuclear reactors;
  • substantiation of nuclear and structural materials for normal operation, transients, design-basis and beyond-design basis accidents;
  • establishing of research, technological, and test frameworks for nuclear fuel cycle related activities;
  • research and developments in the area of nuclear weapon material utilization to manufacture fuel for NPPs;
  • development of methodologies and experimental substantiation of reactor core component performance;
  • development and creation at NFC STE of the regulatory and technical documentation database by the sector's activity areas;
  • participation in the development and implementation of state conversion programs, government programs for fundamental and applied research into the utilization of nuclear materials and nuclear-and-radiation technologies for various branches of economy and other government needs;
  • carrying out R&D and applied activities in conjunction with national and foreign laboratories and companies;
  • research in the area of spent nuclear fuel management;
  • organization, on behalf of NFC STE, of relations with NSC KIPT departments, enterprises and companies in Ukraine and foreign companies and enterprises with regard to R&D on structural materials to support NFC STE activities on the nuclear fuel cycle;
  • participation in and development of work programs on structural materials for the nuclear fuel cycle;
  • participation in organization of manufacturing and testing of pilot batches of materials samples and products therefrom to substantiate their quality, reliability, and endurance;
  • expert review of the work performed by other organizations and enterprises in the areas related to the sectors activities;
  • issuing reports, designs, and other documentation based on the results of in-house developments; development of analytical reports and overviews on the sector's activity areas and as requested by the National Academy of Sciences of Ukraine, Ministry of Energy and Coal Industry and other organizations;
  • organization of the quality system for the NFC STE and NSC KIPT developments in the sector's activity areas; participation in organization of the quality system at industrial facilities for manufacturing materials and products therefrom in the sector's activity areas.

Sector's Equipment

Division for autoclave tests of fuel and absorber element materials in water environments similar to the WWER-1000 coolant at operating temperatures and pressures
Division for autoclave tests of various materials and product mockups in gas environments
Autoclaves for testing reactor materials, fuel and absorber element mockups at temperatures 300… 350оС and water environment pressure up to 16.5 MPa. Designed to determine performance of zirconium alloys, stainless steels, and product mockups at WWER-1000 coolant parameters
Specialized titanium and stainless steel autoclaves to research corrosion processes in reactor materials
Autoclave for express testing of reactor materials, fuel and absorber element mockups at a temperature of 400оС and water environment pressure of 20.0 MPa. Designed for express testing of reactor materials at parameters exceeding the operating pressures and temperatures
Autoclave control and computer monitoring panel
Facility for water steam flow testing of materials at temperatures up to 1200оС to imitate the effects of accident-induced overheats on fuel and absorber rod materials in WWER conditions Parameters: temperature 400…1200оС, water steam pressure 0.1 MPa
Division for high-precision weighing and heat treatment of reactor materials
Facility for hydrogenation of zirconium alloy and other material samples Purpose: study of hydrogen effects on the properties of materials and orientation of hydride precipitation in zirconium alloys
Facility with quartz spring balance for continuous weighing in the water steam flow at temperatures up to 1200оС. Purpose: study of zirconium alloy oxidation at accident-induced overheat temperatures in reactors
High-vacuum facility for heat treatment of reactor materials Purpose: annealing and oxidation of metals and alloys in vacuum conditions in gas environments at gas pressures from 0.1 MPa to 2·10-6 mm of mercury and temperatures of up to 1100оС
The facility is designed for high-frequency heat treatment of tubes and rods using the VChG-1-25/0,44 generator. A tube or a rod is displaced at a specified rate through a coil where it is heated to a specified temperature (in excess of 2000°С) to be further cooled by an annular jet of water in a shower (the tube or rod cooling rate is 500-1200°С/s)
A vacuum annealing facility with a vertical electrical furnace of 110 mm in diameter and 700 mm in height designed for heat treatment of materials and products in vacuum
Facility for study of various alloys' thermoelectromotive force A vacuum facility with a horizontal electrical furnace of 40 mm in diameter designed to study specific electrical resistance during heating and cooling of various materials
Facility for vacuum melting of uranium-containing materials
The “Vertikal” facility Designed for UHF treatment of full-size Zr-2.5% Nb channel tubes 40-120 mm in diameter and about 10 m in length to create a quasi-isotropic radiation-resistant structure
Facility for study of various alloys' viscosity
The “Termoshok” [Thermal Shock] facility to study performance of thin-walled cladding tubes in normal operating conditions, during transients, and LOCA-type design-basis accidents

Based on fundamental theoretical and experimental research into structural state, physical and mechanical properties, and radiation behavior of products from commercial zirconium alloys (Zr-1%Nb, Zr-2.5%Nb), NFC STE NSC KIPT developed an effective method for enhancing radiation stability of these products, which are used in power reactor cores as channel and fuel rod tubes. The method consists in comprehensive thermo-mechanical treatment of zirconium alloys during fabrication of the said products, including a number of successive operations: deformation, annealing, beta-heat treatment, and ageing. The treatment creates an isotropic strengthened small-grain structure with uniform distribution of fine secondary phases.

Dependence of Zr-2.5% Nb tube radiation growth deformation lengthwise and transversely versus fast neutron fluence 

Exterior view of a channel tube

1 – annealing 550°С, 5 h
2 – UHF heat treatment
a, c – lengthwise
b, d – transversely

Radiation tests of 1 and 2.5%Nb zirconium alloy products in the temperature range 80-3500С and fluences from 1×1024 n/m2 to 1×1027 n/m2 confirmed their high radiation stability – absence of radiation growth, 10-time higher resistance to radiation creep, and minimum irradiation hardening of these alloys compared to the same alloy products after commercial treatment.

The method developed can be recommended for fabrication of guide thimble and central tubes of WWER and PWR reactors, and shroud and channel tubes of RBMK and CANDU reactors as a means of significant improvement of their radiation resistance.


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