Radioactive wastes are residues with different radionuclide compositions, placing, therefore considerable demands by measurement techniques used in their characterization. All radioisotopes, at some stage, require quantitation of the isotope, which is done by measuring the intensity of radiation emitted for the three main types of ionizing radiation. Radioactive isotopes of elements are normally determined by their characteristic radiation, i. e., by radiometric methods. Radiometric determination is performed by instrumental analysis using sophisticated methods such as liquid scintillation counters that allow beta spectrometry, alpha spectrometry with semiconductor detectors and high resolution gamma spectrometry for high and low energy gamma emitting nuclides. Besides, mass spectrometric methods can be also used for the determination of radionuclides once they are normally used for determination of isotopes of elements.

There are several types of detectors that can be used for the measurement of ionizing radiation. In the specific case of 59Ni and 63Ni, the more common radiation detection systems are ultra low energy gamma detection and liquid scintillation detection on the basis of charge carriers (holes and electrons) and liquid scintillation phenomena, respectively. Furthermore, for sequential analysis of these radionuclides alpha spectrometry can be applied to alpha emitters associated and presents in ILW and LLW samples.

2.1 The semiconductors detectors

The most recent class of detector developed is the solid-state semiconductor detector. In these detectors, radiation is measured by means of the number of charge carriers set free in the detector, which is arranged between two electrodes. Ionizing radiation produces free electrons and holes. The number of electron-hole pairs is proportional to the energy transmitted by the radiation to the semiconductor. As a result, a number of electrons are transferred from the valence band to the conduction band, and an equal number of holes are created in the valence band. Under the influence of an electric field, electrons and holes travel to the electrodes, where they result in a pulse that can be measured in an outer circuit. Solid state detectors are fabricated from a variety of materials including: germanium, silicon, cadmium telluride, mercuric iodide, and cadmium zinc telluride.

Germanium detectors are mostly used for spectrometry in nuclear physics and chemistry. The Ultra Low Energy Germanium (Ultra-LEGe) detectors extends the performance range of germanium detectors down to a few hundred electron volts, providing resolution, peak shape, and peak-to-background ratios once thought to be unattainable with semiconductor detectors. According to it specification this detector offers excellent performance over a wide range of detector sizes. The resolution, for example, of a 100 mm2 Ultra-LEGe is less than 150 eV in terms of full-width-half-maximum (FWHM) at 5.9 keV.

Radionuclides commonly emit gamma rays in the energy range from a few keV to ~10 MeV, corresponding to the typical energy levels in nuclei with reasonably long lifetimes. The boundary between gamma rays and X rays is somewhat blurred, as X rays typically refer to the high energy electromagnetic emission of atoms, which may extend to over 100 keV, whereas the lowest energy emissions of nuclei are typically termed gamma rays, even though their energies may be below 20 keV. Therefore, 59Ni that decays by electron capture with emission of 6.9 keV X-rays is suitable to be detected by low energy gamma spectroscopy using Ultra Low Energy Germanium detectors.

Accidents in the second group (1.2, see Figure 5) are analysed in order to develop the measures to preserve the structural integrity of the reactor or to determine the required means and time available for subsequent cooling of the reactor core. The accidents of this group are conventional severe accidents with core meltdown as a result of misbalance between energy source and heat sink. The development of such accidents in RBMK has much in common with overheating processes of vessel-type reactors, but it differs by the RBMK features mentioned above.

The heating and melting of a RBMK core can potentially occur as a result of misbalance between heat generated in the core and removed by reactor cooling system and emergency core cooling systems. A typical example of such accident is the damage of the boundaries of the circulation loop (LOCA type accident), accompanied by the failure of the ECCS or additional loss of feedwater.

The LOCAs, according the PSA terminology for RBMK-1500, were categorized according to the rupture size and location in the RCS. A large LOCA means a rupture of the biggest diameter pipes in RCS, i. e. pipes with diameter of 300 — 1000 mm. Medium LOCA is a rupture of pipes with diameter of 100 — 300 mm, whereas small and very small LOCA signifies a rupture of pipes with diameters of 50 — 100 mm and 30 — 50 mm respectively. The location of LOCAs in the RCS is groped into 4 zones (see Figure 11).

Pipelines located in Zone 1 have large diameters: ~1000 mm — the main circulation pumps pressure and suction headers, ~600 mm — MCP connecting pipelines and ~300 mm — group distribution header piping. Therefore, large and medium LOCAs are possible in this zone. Pipelines located in Zone 2 have smaller diameter than the ones in Zone 1: ~300 mm — GDH piping and ~50 mm — lower water piping. Piping of Zone 4 (outside ALS compartments) consists of a part of downcomers (~600 mm diameter), feedwater lines of 500 mm diameter and steamlines of 600 mm diameter. Fuel channel break in the reactor cavity (LOCA in Zone 3) is a separate type of accident. The analysis of LOCAs in Zones 1, 2 and 4 was performed using RELAP5 model, presented in Figure 6.

In general, the surveyed research indicates two types of void fraction measurements (Feentra et al., 2000). The HEM void fraction and RAD void fraction. HEM refers to Homogeneous Equilibrium Model and RAD refers to Radiation Attenuation Method. The determination of fluid parameters (fluid density and flow velocity) are quite different when these two methods are used (Feentra et al., 2000). In RAD method (Feenstra et al., 2000, Wright & Bannister, 1970) gamma flux from radiation source which penetrates the test section will be attenuated by different amounts depending upon the average density of the two-phase flow. Void fraction a can be determined by interpolating the average density of the fluid between the benchmark measurements for one hundred percent liquid and gas according to the following equation.

a = ln(N / Nl ) / ln(Ng / Nl ) (53)

where N represents the gamma counts obtained during an experimental trial, NL and NG are the reference counts obtained prior to the experiment for 100% liquid and 100% gas respectively. Gas phase velocity, UG, and liquid phase velocity UL can be calculated by Equations below:

(1 — x)GP (1 — a)pL

where Gp is the pitch mass flux.

A logical measure of an equivalent two-phase velocity, Veq is determined from averaging the dynamic head of the gas and liquid phases as given by equation below:

Veq =V -aPGUl + (1 ~«)PlUIV P (56)

4. Conclusions

Loss of Millions of Dollars through Cross-Flow-Induced-Vibrations related problems in steam generators and heat exchangers excitations has been a cause of major concern in process, power generation and nuclear industries. Flow-Induced Vibration pose a potential problem to designers, process engineers and plant operating and maintenance personnel. Such vibrations lead to motion of tubes in loose supports of baffles of tube bundles, resulting in mechanical damage, fretting wear, leaking and fatigue etc. Heat exchanger tubes are the most flexible components of the assembly. The risk of radiation exposure is always present in case of leakage in steam generator of PWR plants due to vibration related tube failures.

A number of design consideration have been reviewed in this chapter in order to achieve design improvements to support large scale heat exchangers with increased shell-side cross­flow-velocities. The prime consideration is the natural frequency of tubes in a bundle against cross-flow-induced-vibrations. Various analytical, experimental and computational techniques for straight & curved tubes have been discussed with reference to single and multiple spans and varying end and intermediate support conditions. Earlier, Flow — Induced-Vibration analysis was based upon the concept of two types of damage numbers (Collision damage and baffle damage). Discussion on these damage numbers and on the parameters that influence damping has been included.

Next consideration is the generally accepted following four tube bundle vibration excitation mechanisms (various models have been discussed & reviewed) including steady, unsteady, analytical, FEM based, CFD based, experimental, empirical correlation based, large eddy simulation (LES) based, linear and non-linear etc.

It can not be avoided in Heat Exchangers and is caused due to turbulence.

Vortex shedding or periodic wake Shedding Self excited vibration resulting from interaction of tube motion and flow is the most dangerous excitation mechanism.

Caused by some flow excitation having frequency which coincides with natural frequency.

Dynamic parameters like added mass and damping which are function of geometry, density of fluid and tube size have been targeted by a number of researches in single-phase and two-phase flow. These researches have identified seven separate sources of damping which have been highlighted.

Tube wear due to non-linear tube-to-tube support plate interactions caused by gap clearances between interacting components resulting in thickness loss and normal wear work-rates have been reviewed. Chaotic dynamics of tubes impacting generally on loose baffle plates with consideration of stability and bifurcation have been discussed.

Two-phase Cross-Flow-Induced-Vibrations in tube bundles of process heat exchanger and U-bend region of Nuclear Steam generators can cause serious tube factures by fatigue and fretting wear. Solution to such problems require understanding of vibration excitation and damping mechanism in two-phase flow. This further requires consideration of different flow regimes which characterize two-phase flow. The discussion includes the most important parameter which is void fraction, various thermal-hydraulic models, dynamic parameters, wear work-rates, void fraction measurement and application of TEMA/ ASME and other codes have been reviewed. In conclusion the objective of this chapter is to suggest improvements in the design guidelines from the available researches to use the related equipment at optimal performance level.

5. Acknowledgements

We are deeply indebted to University of Engineering & Technology, Taxila — Pakistan, PASTIC, Islamabad — Pakistan and College of EME NUST, Rawalpindi — Pakistan for providing financial, administrative and technical support. We sincerely appreciate the support provided by Mr. Zahid Iqbal, Mr. Riffat Iqbal and Mr. Muhammad Shafique in finalizing the manuscript.

2.1 Test of tube-bundle channel

High-speed camera is adopted to carry out experiments of the two-phase boiling flow in the tube-bundle channel for different heat flux and different flow rate. Several representative pictures obtained are displayed in Fig.12.

Fig. 12. Flow patterns in a vertical tube-bundle channel

Fig. 12 shows that there are four main flow patterns, bubbly flow, bubbly-churn flow, churn flow and annular flow, which is different from flow patterns of two-phase boiling flow in a circular tube. Through the analyses, it is shown that there may be two reasons for these differences. Firstly, the geometric dimensions cause the different flow patterns. The tube — bundle channel is divided into several sub-channels by the tubes, as shown in Fig. 12. And the inner tubes divide the large bubble and make disturbance on two-phase flow. Furthermore, flows in the sub-channels interact and enhance the complexity of two-phase flow in the tube-bundle channel. Secondly, the heating mode of boiling flow in the tube — bundle channel and that in the circular tubes are different. When flowing in a circular tube, the fluid is surrounded and heated by the wall of the tube. On the other hand, in a tube — bundle channel, the fluid surrounds the tube bundle, which acts as the heating source. Then, all of these might cause differences between flow patterns in vertical tube-bundle channels and that in vertical circular tubes (Hewitt & Roberts, 1969).

Two-phase boiling flow in tube-bundle channels exhibits several main flow patterns, bubbly flow, bubbly-churn flow, churn flow and annular flow, as shown in Fig. 13. There are differences from the results gained by Grant & Chisolm (1979) and Ma (1992). These differences might be caused by the different test conditions. The characteristics of flow patterns and transitions in the present experiments are analyzed as follows.

(1) Bubbly flow: In the experiments, bubbles begin to come into being in the liquid when the heat flux is small. The mainstream is liquid and discrete bubbles are dispersed in the mainstream, which indicates that bubbly flow in circular tubes is similar to that in the tube — bundle channel, as shown in Fig. 13(a). With the sustaining heating, the dimension and the quantity of bubbles increase gradually along the flow direction, and discrete bubbles begin to aggregate to combine to be large bubbles. Confined by the narrow space in the tube — bundle channel, large bubbles transfigure to be oval or crescent.

(2) Bubbly-churn flow: When the little discrete bubbles aggregate and combine to be large bubbles, the tubes agitate and divide these large bubbles. This makes there be no slug flow in the tube-bundle channel. And the flow pattern begins to be transformed from bubbly flow to churn flow.

(3) Churn flow: With the augmentation of heat flux, discrete bubbles continue to aggregate and combine and the aggregation bubbles become bigger in size. These aggregation bubbles present unstable state, on account of the agitation and division of the tubes. And the aggregation bubbles begin to burst into many little discrete bubbles with unequal geometric dimensions. Then, the flow pattern is transformed to churn flow.

When churn flow occurs in the tube-bundle channel, there might be many bubbles with unequal geometric dimensions. The liquid moves up and down in the channel and the two — phase flow exhibits surge state. [8]

In the experiments, there is no mist flow due to the limit of the experiment condition, which is detected by Grant & Chisolm (1979).

There may be several differences between flow patterns and their transition in circular tubes and that in tube-bundle channels. Firstly, slug flow is one of the main flow patterns for two-phase flow in a circular tube. But the tube-bundle channels, the inner tubes divide the large bubbles and flows in the sub-channels interact, which makes bubbles can not aggregate and combine to be a slug. Thus there is a lack of slug flow in the tube-bundle channel and the flow pattern is transform directly from bubbly flow to churn flow. Secondly, churn flow, as the transition from slug flow to annular flow, exists transitory in circular tubes. Under some circumstances, there might be a lack of churn flow in circular tubes. But in the tube-bundle channel, churn flow presents itself as one of the main flow patterns. And churn flow exists in many experimental conditions and present a long time in the tube-bundle channel, as shown in Fig. 12.

With the same flow rate, there may be different flow patterns in the tube-bundle channel due to the different heat flux. For example, when the flow rate is 281.8 (kg/m2s), bubbly flow occurs in the tube-bundle channel where the heat flux is comparatively small. But the amount of bubbles begin to increase with the enhanced heat flux. And the flow pattern changes from bubbly flow to churn flow when the heat flux is equal to 5.1X104 (w/m2). This difference in flow patterns will be more obvious along with the decrease of the flow rate, which is shown in Fig. 12. When the flow rate is 133.9 (kg/m2s), there will appear three flow patterns, which are bubbly flow, bubbly-churn flow and churn flow, under different heat flux in the tube-bundle channel. And when the flow rate is getting smaller, the flow patterns with differnet heat flux will be differentiated more distinctly. As shown in Fig. 12, there appear four different flow patterns, which are bubbly flow, churn flow, churn-annular flow and annular flow, when the flow rate is equal to 63.3 (kg/m2s). From above all, it is shown that heat flux will affect the flow pattern more remarkably under the small flow rate condition.

Furthermore, it is found that there may be two different flow patterns in the same cross­section when the experiments run. The flow pattern transitions exhibit unsynchronized in different sub-channels. This unsynchronized phenomenon is caused by the different geometric dimensions, different heat flux and different quantity of discrete bubbles generated in different sub-channels. It was found the same phenomenon in a tube-bundle channel by adopting microprobe to detect the flow pattern transitions (Bergles 1981).

The flow pattern map obtained in the experiments is shown in Fig. 14. Comparisons are made with the results gained by Hewitt & Roberts (1969), which is figured by dashed line. Two-phase flow patterns and their transition in vertical circular tubes were experimentally studied by Hewitt & Roberts. The mass flow rate in the present experiments is less than that of Hewitt & Roberts owing to the experimental condition limits. In Fig. 6, the abscissa p’ j ^ and the ordinate p" jg are calculated by (Hewitt & Roberts, 1969),

G2 x2

P" Jg =—IT

s p"

It is shown in Fig. 14 that there are great differences between flow pattern transitions in a tube-bundle channel and that in a circular tube. The generation regions of bubbly flow and churn flow in a tube-bundle channel move left in Fig. 14, compared with the region in a circular tube. And it is shown that the regions of bubbly flow and churn flow are larger in a tube-bundle channel, which is caused by that the inner tubes have effects of disturbance and division on the bubbles and these effects make it impossible for discrete bubbles to converge and unite to be a slug. In addition, on account of being heated by the inner tubes, the fluid generates bubbles in the core of the channel, which makes the probability increase for bubbles to converge and unit to be a continuous axle center in the core of the tube-bundle channel, and then the flow pattern transform to annular flow in a tube-bundle channel is earlier than that in a circular tube.

The complex fluid flow events occurring during the operation of NPPs are influenced by the automatic operational control processes. Nevertheless, as a consequence of the manifold manual intervention opportunities equal technological processes may induce different local loading sequences for the components. In other words, an assessment of components exclusively based on operational measuring instrumentation is insufficient. Local data acquisition and monitoring of local loads at the fatigue relevant components is the better solution. Local effects such as the swapping flow after feeding interruption can only be recorded in the load data set this way. It is to be pointed out that the safety check against cyclic loads of the components has to be a permanent operation accompanying procedure. The German KTA rules regulate this issue as part of the rule for operational monitoring (KTA 3201.4) [3].

1.2 Modified codes

On one hand, the checks have to be harmonized with the valid design code. On the other hand, the state of the art in science and technology has to be considered. Recently, the detrimental influence of the medium (high temperature reactor water) on the fatigue process — which has been examined since the 1980ies — is the subject of code modifications tending towards tightening code rules. The term environmentally assisted fatigue (EAF) is synonymous to the corrosive influence of the medium on the fatigue behavior. The usual way of considering EAF in fatigue analysis is the application of penalty factors Fen. The modified code rules mostly based on [2] have to be considered and applied both within the lifetime extension projects and the new built projects of NPPs.

The Safety-by-Design™ approach, used by the designers of IRIS to eliminate the possibility of occurrence of certain severe accidents caused by internal events, had been extended to the external events.

The normally operating IRIS systems and their non-safety, active backup systems were typically located within substantial structures that can withstand some degree of external event challenges. This equipment included the backup diesel generators. IRIS had non­safety related backup diesels for normally available active equipment that could bring the plant to cold shutdown conditions.

IRIS plant safety features, once actuated, relied on natural driving forces such as gravity and natural circulation flow for their continued function. These safety systems did not need diesel generators as they are designed to function without safety-grade support systems (e. g. AC power, component cooling water, or service water) for a period of 7 days.

All the IRIS safety related equipment, including the batteries that provide emergency power, and the passive habitability system, were also located within concrete structures. The reactor, containment, passive safety systems, fuel storage, power source, control room and backup control were all located within the reinforced concrete auxiliary building and were protected from on-site explosions.

Actually, IRIS had a very low profile, which was very important when considering aircraft crash, especially by terrorists. The IRIS containment was completely within the reinforced concrete auxiliary building and one-half of it (13 m) was actually underground. The external, surrounding building was only about 25 m high, thus offering a minimal target. The integral vessel configuration eliminated loop piping and external components, thus enabling compact containment (see Figure 1) and plant size.

Fig. 1. IRIS Containment

The Refuelling Water Storage Tank (RWST) which is the plant’s ultimate heat sink would be also protected from some external events by locating it inside the reinforced concrete auxiliary building structure. In addition, the IRIS RWST was designed to be replenished by alternative water sources such as fire trucks, therefore being completely independent by the plant power resources.

Because of these and other reasons, it was expected that the impact of external events at the site would be lower than that for current plants. In addition, typical design approaches, that could contribute to achieve such robustness in advanced NPPs design are:

• Capability to limit reactor power through inherent neutronic characteristics in the event of any failure of normal shut-down systems, and/or provision of a passive shut-down system not requiring any trip signal, power source, or operator action.

• Availability of a sufficiently large heat sink within the containment to indefinitely (or for a long grace period) remove core heat corresponding to above-mentioned events.

• Availability of very reliable passive heat transfer mechanisms for transfer of core heat to this heat sink.

It was observed that the implementation of innovative design measures needs to be supported (and encouraged) by a rational, technical and non-prescriptive basis to exclude any severe accident (core melt need not be presupposed to occur). The rational technical basis should be derived from realistic scenarios applicable for the plant design. Most of the innovative reactor designs aimed to eliminate the need for relocation or evacuation measures outside the plant site, through the use of enhanced safety features in design. Many of these designs also aimed to take advantage of these advanced safety characteristics to seek exemption from maintaining a large exclusion distance around the nuclear power plants.

Using a quasi-static analysis, (Connors & Parrondo, 1970) and later (Dalton & Helfinstine, 1971) developed the fluid-elastic instability prediction for cylinders (single row of cylinders) subjected to cross-flow. Connors measured the fluid forces instead of predicting these using pitch to dia. ratio of P/D=1.41. He observed many different model patterns at instability, but suggested that the most dominant was elliptical motion (whirling). Using the measured fluid stiffness, Connors obtained energy balances in the in-and cross-flow directions, which must be satisfied simultaneously giving

where K is the so-called Connors constant, fn is the frequency of oscillation. Vpc is the so — called pitch velocity given by

VP

P — D

where P being the centre-to-centre inter cylinder pitch

(Blevins, 1974) has derived Equation 3 by assuming that the fluid forces on any cylinder are due to relative displacements between itself and its neighboring cylinders. Later, (Blevins, 1979) modified his original analysis to account for flow dependent fluid damping giving

where and ^y are total damping factors in the in-and cross-flow directions.

1. Gauge blocks were installed and the cap screws were then tightened. They needed to be tightened according to a three-step tightness method, as follows: 160, 213 and 266 ft-lbs. The final torque was 256 — 276 ft-lbs and the tightness sequence was as follows:

2. The number of cap screws was 1, 2, 3, and 4 from top to bottom. The tightness sequence of step 1 was 2, 3, 1, and 4, and the torque was 160 ft-lbs.

3. The tightness of step 2 was identical to that of 2) and the torque was 213 ft-lbs.

4. The tightness of step 3 was identical to that of 2) and the torque was 266 ft-lbs.

5. The cap screws were unscrewed in the reverse sequence of the tightness sequence: 4, 1, 3, and 2.

6. One more time, the tightness was carried out by the sequence of 2), 3), and 4).

7. The remaining gauge blocks were installed on the CSB snubber lugs according to the sequence of 2), 3), 4), 5), and 6). After installation of the gauge blocks, the gaps between the upper part and the lower part were uniformly maintained at 0.1016 mm.

4.1.3 Dimensions of the CSB Snubber Lug

1. After installation of the gauge blocks, all widths of the gauge blocks at the same six intervals were measured. Subsequently, the dimensions of the CSB snubber lugs were measured. The measured parts of the CSB snubber lugs were the inside widths of both of their surfaces. In total, six holes of the CSB snubber lugs were measured.

2. The measured positions of the CSB keyway were marked and the widths measured.

3. Four dummy alignment keys (DAKs) were installed on the RV keyways and adjusted so that they could be positioned within 0.254 mm of the RV centerline. The vertical degree of an RV head seating surface of a DAK was adjusted so that it could be positioned within 0.0254 mm/ ft. The position of the DAK and the vertical degree were measured again.

An OSU-MASLWR TRACE model (Mascari et al., 2008, 2009a, 2009b, 2010b, 2011a, 2011b, 2011c, 2011d) is developed in order to evaluate the TRACE code capabilities in predicting the thermal hydraulic phenomena typical of the MASLWR design as natural circulation, heat exchange from primary to secondary side by helical SG in superheated condition and primary/ containment coupling during transient scenario.

The TRACE nodalization, developed by using SNAP, models the primary and the secondary circuit. The containment structures consisting of the HPC, CPV and heat transfer plate are modeled as well, figure 12.

The "slice nodalization" technique is adopted in order to improve the capability of the code to reproduce natural circulation phenomena. This technique consists in realizing the mesh cells of different nodalization zones, at the same elevation, with the same cell length (Mascari et al., 2011a). In this way it is avoided the error due to the position/ elevation of the cell nodalization center that can influences the results of the calculated data when natural circulation regime is present. If the "slice nodalization" technique is not used, this error has to be taken into account and its effect increases if larger nodalization cells are used. In this case it can be reduced by using a "fine nodalization". In general, its effect on the results is less important when forced circulation regimes are simulated. However the "slice nodalization" technique could require nodes of small length increasing the numerical error and the computational time. The "code user" has to take into account these disadvantages during the nodalization development.

The primary circuit of the TRACE model, comprises the core, the HL riser, the UP, the PRZ, the SG primary side, the CL down comer and the LP. After leaving the top of the HL riser, the flow enters the UP divided in two thermal hydraulic regions connected to the PRZ. After living the UP, the flow continues downward through the SG primary section and into the CL down comer region. The core is modeled with one thermal hydraulic region thermally coupled with one equivalent active heat structure simulating the 56 electric heaters. The PRZ is modeled with two hydraulic regions, connected by different single junctions, in order to allow potential natural circulation/ convection phenomena. The three different PRZ heater elements are modeled with one equivalent active heat structure. The thick baffle plate is modeled as well. The direct heat exchange by the internal shell between the hotter fluid, in the ascending riser, and the colder fluid, in the descending annular down comer, is modeled by heat structures thermally coupled with these two different hydraulic regions. SG coils are modeled with one "equivalent" group of pipes, in order to simulate the three separate parallel helical coils. The steam line of the facility is modeled as well.

In order to simulate the OSU-MASLWR-001 test the HPC is divided in two thermal hydraulic regions, connected by single junctions, in order to allow the simulation of possible natural circulation phenomena. The ADS lines are modeled.

The RPV, HPC and CPV shell and the connected insulation are modeled.

V. Slugen, J. Lipka, J. Dekan, J. Degmova and I. Toth

Institute of Nuclear and Physical Engineerining Slovak University of Technology Bratislava, Bratislava

Slovakia

1. Introduction

Corrosion at nuclear power plants (NPP) is a problem which is expected. If it is managed properly during the whole NPP lifetime, consequences of corrosion processes are not dramatic. For adequate protection against corrosion it is important to collect all relevant parameters including exact phase composition of registered corrosion products.

Corrosion is more frequent and stronger in secondary circuit of NPP. Steam generator (SG) is generally one of the most important components from the corrosion point of view at all NPP with close impact to safe and long-term operation. Various designs were developed at different NPPs during last 50 years. Wide type of steels was used in respect of specific operational conditions and expected corrosion processes. In our study we were focused on the Russian water cooled and water moderated reactors (VVER). These reactors are unique because of horizontal position of SGs. It takes several advantages (large amount of cooling water in case of loss of coolant accident, good accessibility, large heat exchange surface, etc. …) but also some disadvantages, which are important to take into account during the operation and maintenance. Material degradation and corrosion/erosion processes are serious risks for long-term reliable operation. In the period of about 10-15 year ago, the feed water pipelines were changed at all SG in all 4 Bohunice units (V-1 and V-2, in total at 24 SGs). Also, a new design of this pipeline system was performed. Actually, there is a time to evaluate the benefit of these changes.

The variability of the properties and the composition of the corrosion products of the stainless Cr-Ni and mild steels in dependence on the NPP operating conditions (temperature, acidity, etc.) is of such range that, in practice, it is impossible to determine the properties of the corrosion products for an actual case from the theoretical data only. Since the decontamination processes for the materials of the VVER-440 secondary circuits are in the progress of development, it is necessary to draw the needed information by the measurement and analysis of the real specimens [1].