Radiation Material Study Department

(Head – Ph.D. L. I. Chyrko)

The department was established in 1977. Up to 1993 the department was headed by Dr. Sci., Professor V. S. Karasev.

There are two laboratories for the works with radioactive materials at the department, including unique in Ukraine “hot” cells with the allowable activity up to 25000 Ci. The laboratories are fitted out with the equipment for the study of physical-mechanical characteristics of materials irradiated with the high neutron doses, surveillance specimens of the reactor vessel metals of the Ukrainian NPPs. The works on exchange of industrial and medical devices completed with the sources of ionizing radiation are also performed in the laboratories. For the transportation of the radioactive materials there are special protective containers at the department. The department has appropriate licenses and permissions from the regulatory bodies of Ukraine for carrying out the above mentioned works.

Main directions of scientific activity:

• Determination of mechanisms of radiation damage of the solids, selection of the most advanced structural materials for nuclear reactor industry.
• Radiation material science aspects of operating reactors safety, investigation of the radiation embrittlement of the reactor vessel steels and determination of the WWER-1000 type reactor vessels safe operation lifetime in particular.
• Study of the effect of radiation damage on the dislocation mobility.
• Study of the point defect kinetics under irradiation and the role of radiation and thermal vacancies in the grain-boundary relaxation.
• Radiation helium effect on the metal and steel stability.
• Hydrogen embrittlement of zirconium alloy of fuel channels.

The most important scientific results: 

• The effect of the reverse decrease of the material shift modulus is discovered experimentally (immediately under irradiation). The analytical calculations showed the effect to be induced by the increase of the dislocation diffusion mobility at the expense of the overflow of the interstitial atoms. The discovered correlation allows assessing the material radiation resistance without performing long-term reactor experiments.
• Grain-boundary stress relaxation is stated to be defined by the total rate of flows of radiation and thermal vacancies at the expense of the absorption of interstitial atoms by matrix dislocations. The value of the vacancies overflow to the grain boundaries at various rates of point defect generation is assessed. The obtained results allow predicting the tendency of a material to the radiation swelling.
• Binding energy of helium with the grain boundaries and its concentration in the specimens is calculated. The boundary neutron fluence is assessed.
• Neutron flux effect mechanism on the microstructure of austenitic steel for the fuel cladding is elucidated.
• For the WWER-1000 reactor vessel steel such an effect is revealed: in a definite flux interval the lower the neutron flux is, the higher the degree of embrittlement is, the neutron fluence being the same. Nickel is shown to be a deleterious alloying element of the irradiated vessel steel and the synergetic effect of Ni, P and Cu intensifies the negative effect of these elements on the steel resistance.
• Hydrogen embrittlement is proved to be the main factor of zirconium alloys physical-mechanical properties degradation. It is intensified by the availability of the impurity particles which swelling under irradiation leads to the internal stress leading in its turn to the further increase of the level of the surface oxidation and the alloy inundation. The obtained results demonstrate the fuel channels of all the units of ChNPP not to terminate their lifetime of the safe operation in accordance with the standard of the brittle fracture. Data on the effect of the neutron irradiation on the microstructure of steel-zirconium weld fuel channels allowed formulating the criterion of their safe operation.