From the point of view of technical and nuclear safety and the possible consequences of an emergency, it is necessary to have a knowledge of the state of the entire primary circuit, i.e., about all equipment and pipeline systems during the designed life and beyond - during Long-Term Operation (LTO). 

          The reactor including the reactor cover is the most critical component in Nuclear Power Plant (NPP) from the point of view of nuclear safety. This component which is irreplaceable must withstand all degradation mechanisms such as irradiation, thermal and mechanical stress and corrosion strain. The critical parts here are weld joints, cladding, nozzles, and sealing surfaces. Further, the steam generator appears as critical equipment from of view of the effects of the degradation mechanism and operation ageing. The degradation mechanisms mostly in heat exchange tubes have an influence on functionality, which is the transport of heat from the primary coolant water to the secondary circuit. Further critical parts are also the Main circulation piping - mostly their weld joints and nozzles.

          The steam generator, welds and pipping are under certain circumstances replaceable. However, any repair or replacement of equipment on the primary circuit represents a huge financial investment by the operator, in the form of downtime or even the repair costs themselves. In the case of LTO operation, this may lead to the decision that further operation of the NPP is not economically advantageous and may lead to the shutdown of the NPP. Therefore, it is necessary to know well the ageing processes of materials during operation on the entire primary circuit.

   Factors, that have an influence on the safe operation of NPPs, are the level of operation management, personnel qualifications, compliance with operating regulations and especially the state of knowledge about the operated equipment and the effect of degradation mechanisms on it. Knowledge of the operated equipment state is determined by several factors:
              a) Knowledge of degradation mechanisms acting on the material and components.
              b) Knowledge of the effects on the ageing of materials and components under operating conditions.
              c) Level and scope of maintenance activities.
              d) Level and scope of in-service inspections.
Knowledge of the effects of degradation mechanism and ageing effects during normal operating conditions is the most important factor for correct lifetime assessment of structure, construction and component. For this, it is necessary to know the operating conditions and properly monitor them for any changes.

          During the identification of critical components of the VVER, we are focused on two views: i) the increase of the safety operation at extended lifetime due to the in-time prediction of the potential failure and recommended construction, material, and maintenance optimisation, and ii) to the investigation of available materials (steam generator tubes, primary pipes) from decommissioned nuclear power plants or from material archives of the DELISA-LTO project partners.

Generally, the limiting factors for VVERs from the LTO point of view are these degradative processes:
          a)   Radiation embrittlement of the reactor pressure vessel,
          b)   Hydrogen embrittlement,
          c)   Void Swelling,
          d)   Thermal ageing,
          e)   Low-cycle fatigue,
          f)   Stress Corrosion Cracking (SCC),
          g) General corrosion,
          h)   Wear-out,                   
           i)   Loss of pre-load of bolt connections.

Radiation embrittlement of the reactor pressure vessel is one of the critical issues of the long-term operation of light water reactors, which is caused mainly by neutron irradiation. Due to neutron irradiation, the transition of reactor pressure vessel steel from a ductile state to a brittle one occurs because of internal changes in the phase composition and redistribution of defects in the crystal structure or impurity atoms. The main impurity elements supporting the formation of precipitates and leading to the brittleness are C, Cu, Cr, Si, Ni and Mn, which block the movement of dislocations and thus recombination of newly formed radiation-induced defects. By monitoring chemical composition and maintaining parameters of heat treatment during the manufacture of reactor pressure vessels, these phenomena do not present a significant problem.

The prediction of radiation embrittlement is performed usually by relevant codes and standards that are based on a large amount of information from surveillance and test irradiation programs.

In reactor pressure vessels, hydrogen arises in the primary circuit of NPP due to radiolysis of the coolant, corrosive effects of the coolant and high-temperature dissociation of water. This hydrogen can be absorbed by the material, affects its properties and can cause a degradation of material strength and stress state, which is called hydrogen embrittlement. Because of its small size, hydrogen is capable to diffuse into the material very fast even at ambient temperature. Penetration of atomic hydrogen into the steels could cause an enormous decrease in material ductility. Materials could be susceptible to cracks and formation of the brittle fractures during the loading under the yield strength - occurrence of delayed fractures. During diffusion, atomic hydrogen uses crystal lattice defects such as grain boundaries to penetrate the steel.

Void swelling is an effect of irradiation typical in austenitic steels where vacancy defects and gas bubbles are formed in the structure. This is significant in the core baffle. Its initial shape changes due to radiation, which has considerable influence on gaps between the core baffle and core barrel and can significantly affect contact stresses, as well as redistribution of the primary coolant flow rate in the reactor and change in the reactor temperature regime.

Degradation mechanism swelling almost does not occur by the reactors of VVER-440 type. Here, swelling can occur very limited by baffle bolts in specific areas of the core baffle.

The ageing effect of the radiation swelling is monitored due to the calculation of swelling, measurement of the geometric dimensions of the inner baffle diameter, and periodic non-destructive testing of the metal state through visual inspection.

Low-cycle fatigue is structural damage resulting from repeated cycles of stress/strain caused by fluctuations in loads and temperatures. As a result of repeated load cycles, microstructural damage can accumulate, leading to the appearance of micro-cracks, one of which can develop into a main crack. The controlled effects of ageing due to low-cycle fatigue of the metal components are cracking, micro-cracks initiation and change in the mechanical characteristics. There are three sources of fatigue significant to the pressure water reactors. These are system cycling, thermal cycling and flow-induced vibration. Low-cycle fatigue is a dominant degradation mechanism which decreases the fracture toughness of primary circuit components, equipment, pipelines and nozzles of both VVER-440 and VVER-1000.

Thermal ageing leads to a change in the characteristics of hardness, ductility and strength of steels and proceeds due to a change in the solubility of carbon in α-iron with increasing temperature. Thermal ageing processes in steels occur in the temperature range from 350 to 500 °C or during a long-time operation at lower temperatures of 280 to 300 °C.

The degradation mechanism of thermal ageing is monitored together with other degradation mechanisms of risk parts of the components because these degradation mechanisms are demonstrating more significant and earlier cause the component degradation.For example, steel 15Kh2NMFA is little subject to thermal ageing at operating temperature and her shift of ductile-brittle transient temperature is not more than 10–20 ºС, while due to irradiation can be shifted more than 100°C. The low thermal ageing effect of the reactor pressure vessel could be explained by the positive effect of alloying elements Mo and V which make stable compounds with C and N.

The effect of thermal ageing manifests itself mostly at the initial operation stage (up to 15 years). Therefore, it is assumed that this degradation mechanism should not have a dominant effect on lifetime extension for reactor components.

In the Czech Republic, thermal ageing is assessed using hardness measurements of specimens from standard surveillance specimen programs. In Ukraine, thermal ageing is monitored by a periodical control of the tensile strength of miniature specimens cut from the outer surfaces of nuclear power plant components after 100 000 and 200 000 operational hours. In Slovakia, monitoring of thermal ageing is performed by two approaches: Samples from operating components are analysed and long-term thermally exposed original material is monitored.

Stress corrosion cracking is a special case of corrosion damage or corrosion cracking, which is characterised by quasi-brittle failure without detectable corrosion products. Its formation is conditioned by a simultaneous action of three factors i.e., sufficiently high tensile stress, aggressive environment, and sensitive material.

Stress corrosion cracking is a dominant degradation mechanism for such components as heat exchange tubes and flange connections of steam generators. Operation experience of steam generators of VVER-1000 confirms a high probability of detecting cracks in certain zones with large areas of damage. Effective prevention of stress corrosion cracking is a periodic volumetric ultrasonic test and replacement well before failure.

In the Czech Republic, the bolts from stainless steel showed intergranular stress corrosion cracking caused by a high level of tensile stress and a complex corrosion environment of secondary media under the secondary collector lid. The stress corrosion cracking probably formed a crack also in the main circulation pump in the timing wheels and on the sealing surface of the flanges. The cause of the cracks was determined to be the presence of Cl on the surface of the component in the first years of operation.

The basic causes of the development of cracks are bad water chemistry at the location of the pockets, high temperature and high tensile stress due to the thermal deformation of the pipeline.

Irradiation-assisted stress corrosion cracking can occur in mostly in critical parts such as weld joints, cladding, nozzles, sealing surfaces, etc.

Stress corrosion cracking in steam generator welded joins of VVER-1000 is a common case not only in Ukraine but also in Russia. In 2001, there was detected a through-wall crack in Ukraine VVER-1000 (type V-302).  Such type of defects was detected in steam generators of 9 Ukrainian VVER-1000 Units. According to the results obtained, the damage resulted from stress corrosion cracking originating in the wet and chloride-contaminated environment underneath the thermal insulation on the outer surface of the piping. The occurrence of cracks was confirmed also at VVER-440 type.

In Slovakia, there was a cracking of the hot leg primary collector of the steam generator, where the crack was initiated on the secondary side in the region of water oscillations and was propagated through the collector wall.

Corrosion is a type of material degradation that is caused by its chemical or electrochemical reaction to the surrounding environment. It is characterised by a loss of material and deterioration of its mechanical properties. This mechanism affects the outer surfaces of the reactor pressure vessel, steam generators and other structural parts that are not protected by austenitic anticorrosive cladding.

The general corrosion and pitting are the main degradation mechanism of steam generator heat exchanger tubes together with the stress corrosion cracking. In most cases, the degradation of tubes starts with pitting and after some time, due to the simultaneous increase of tension stress, also the stress corrosion cracking affects the tube with pitting. Thus, the combination of pitting and stress corrosion cracking leads to the failure of heat exchange tubes.

Material degradation and corrosion/erosion processes are serious risks for a long-term reliable operation. In the period of about 10-15 years ago, the feed water pipes originally from carbon steel GOST 22K steel were changed by stainless steel components in more units.

Wear-out is characterised by loss of material due to contact between two mutually moving surfaces of the components. It results in a continuous formation of plastically deformed exposed micro-surfaces. The wear-out affects mostly the main circulation pump and main shut-off valve. In the feed water collector of the steam generator, the erosion-corrosion wear can cause the occurrence of insufficiently heated water on the primary circuit SG collector and as a result lead to unacceptable thermal cyclic stresses.

Loss of pre-load of bolt connections – this degradation mechanism is applied to bolt joints of the main dividing plane and bolts of control assembly drive nozzles. The loss of pre-load is a result of the relaxation of tensions from increased temperature or due to vibrations when the loosening of bolt connections occurs.

The dominant mechanism of the bolt damage was intergranular stress corrosion cracking caused by a high level of tensile stress and a complex corrosion environment of secondary media under the secondary collector lid. The only way to avoid cracking of the bolts was to decrease the pre-stress loads together with changing the sealing method. Also, swelling-induced stresses may cause irradiation-assisted stress corrosion in reactor basket bolts. The prevention is a periodic volumetric ultrasonic test and replacement well before failure.