The series of various studies were carried out in order to answer the question what energy sources should replace the lost nuclear electricity capacities (IAEA-TECDOC-1408, 2004; IAEA-TECDOC-1541, 2007). Currently all Baltic region countries cooperating and seeking to solve energy supply and energy security problems and planning to construct new nuclear reactors in Lithuania at existing NPP site. The last Ignalina Unit 2 RBMK-1500 was closed in the end of 2009 and Lithuania is considering both nuclear and fossil options for its replacement. In order to expand the research of the Lithuanian energy future another option to the analysis is also added: small and medium type nuclear reactor in the new site close to the cities with large heat demand. In general, it could be considered for small countries as alternative for the big nuclear units due to limitation imposed by the grid size and available financial resources.

The results of various studies concerning the future structure of power plants in Lithuanian energy system have showed that looking from the economical point of view the best options to replace Ignalina NPP are new nuclear unit or new combined cycle condensing units together with the existing and new units of Combined Heat and Power plants (CHP) (IAEA — TECDOC-1408, 2004; Norvaisa, 2005). Due to climate conditions in Lithuania, district heating presents a notable fraction of energy consumption in winter months, and infrastructure for its use is already in place in population centers in Lithuania. District heating is widely used in Lithuania (46% of total heat consumption), and the cities of Vilnius and Kaunas comprise the two largest consumers of district heat supply (see Table 1). So, in the future new cogeneration units are likely to be the best alternative for electricity and heat generation in Lithuania.

Among the nuclear alternatives the 330 MW(e) IRIS-like reactor was used for the conceptual investigation, as it could be operated in either the electricity only or the co-generation mode. Thus, a case study was conducted to determine the best way to provide the electricity and district heat in Lithuania up to year 2025 and to assess the tactical implications that a reduced-radius emergency planning zone might have on least cost planning with the IRIS­like reactor operating in the electricity only versus the electricity/district heat (co­generation) mode (Norvaisa & Alzbutas, 2009).

The length of any newly required hot water/steam pipelines into the cities of Vilnius and Kaunas will depend on the radius of the emergency planning zone emplaced around the IRIS-like reactor site; these pipelines represent a cost due to construction and a cost due to heat losses. Both costs increase with the pipeline length and, thereby, affect the viability of the co-generation mode. The case study was conducted parametrically for pipeline lengths of 0.5, 5, 15, and 30 km.

Extremely turbulent flow of the shell-side fluid contains a wide spectrum of frequencies distributed around a central dominant frequency, which increases as the cross-flow velocity increases. This turbulence buffets the tubes, which extract energy from the turbulence at their natural frequency from the spectrum of frequencies present. When the dominant frequency for the turbulent buffeting matches the natural frequency, a considerable transfer of energy is possible leading to significant vibration amplitudes (Chenoweth, 1993). Turbulent flow is characterized by random fluctuation in the fluid velocity and by intense mixing of the fluid. Nuclear fuel bundles and pressurized water reactor (PWR) steam generators are existing examples (Hassan & Ibrahim, 1997).

Turbulence is by nature three-dimensional (Au-Yang, 2000). Large-Eddy Simulation, (LES) incorporated in three-dimensional computer codes has become one of the promising techniques to estimate flow turbulence. (Hassan & Ibrahim, 1997) & (Davis & Hassan, 1993) have carried out Large Eddy Simulation for turbulence prediction in two-and three­dimensional flows. The primary concern in turbulence measurements is how the energy spectrum or the power spectral density (PSD) of the eddies are distributed. The PSD of the

velocity profile E(n) is numerically equal to the square of the Fourier Transform of U"(t), and is defined to be (Hassan & Ibrahim, 1997).

_____ +да

U2 = J E(n)dn

-да

where E(n) is the sum of power at positive and negative frequency n.

where T is the time period over which integration is performed, and a(n) is the Fourier Transform coefficient.

An important parameter of flow turbulence is the correlation function. The Lagrangian (temporal) auto-correlation over a time T gives the length of time (past history) that is related to a given event (Hassan & Ibrahim, 1997).

(Non-dimensional) R(r) = U (t )U ((+ r)

U'(()U'(()

1 (=t

(Dimensional) R(r) = Um r___________ — J U’ (t)U’ (t + z)dt (12b)

T t=0

Physically R(t) represents the average of the product of fluctuating velocity U" values at a given time and at a time r later. R(t) gives information about whether and for how long the instantaneous value of U depends on its previous values. Cross-correlation curves can also be obtained as a function of the time delay to give the correlation between the velocities at consecutive separated location points (Owen, 1965).

1 t=t

Ru(t) = — J U/(t)U2(t + r)dt (13)

T t=0

where R12 gives the cross-correlation of the U-velocity component at 1- and 2- point locations.

Recently (Au-Yang, 2000) has reviewed the acceptance integral method to estimate the random vibration, Root Mean Square (RMS) of structures subjected to turbulent flow (random forcing function). The acceptance integral is given by:

Jap(m) = ~ J0 J0 ^a( x ) [Sp (X > x",m)l Sp (x’ ,®)фр( x )dx’dx” (14)

When a = P, J aa is known as joint acceptance where

Yang obtained closed form solutions for the joint acceptances for two special cases of spring — supported and simply-supported beams. A review of turbulence in two-phase is presented by (Khushnood et al., 2003).

(Endres & Moller, 2009) present the experimental analysis of disturbance propagation with a fixed frequency against cross flow and its effect on velocity fluctuations inside the bank. It is concluded that continuous wavelet transforms of the signals. Figure 8 indicates the disturbance frequency to be showing steady behavior. Generally designing for enhanced heat exchange ratios in thermal equipments ignores the structural effects caused by turbulent flow.

(Pascal-Ribot and Blanchet, 2007) proposed a formulation to collapse the dimensionless spectra of buffeting forces in a single characteristic curve and gives edge to the formulation over previously normalized models in terms of collapse of data.

Figure 9 shows the dimensionless spectra calculated with equations 15 & 16 respectively.

Where a is the void fraction.

(Wang et al., 2006) concludes the physically realistic solutions for turbulent flow in a staggered tube banks can be realized by FLUENT (with 2-D Reynolds stress model).

Figure 10 shows the consistency of turbulence intensity contours obtained through standard wall function approach and non-equilibrium wall function approach whereas near-wall treatment model and near — wall turbulence model predicts much higher results (Wang, et al., 2006).

1. The CSB assembly was separated from the RV and set on a storage stand. The CSB assembly was lifted at 45° turns after vertical lifting of approximately 50 cm in order to prevent the DAKs from being damaged. The CSB assembly was checked continuously via a load measured by a hydra-set.

2. After the checked positions of the gauge blocks were marked using digital probes, the widths of the marked positions were measured. After removing the gauge blocks, the gaps between the CSB snubber lugs and the RV core-stabilizing lugs were used to calculated the processing dimensions of the snubber shims; the gaps were calculated at

0. 381-0.508 mm as a permissible range.

4.1.4 Installation and measurement of the shim on the RV core-stabilizing lug

1. The dimensions of the refined shims were measured and a penetration test was conducted. Before installation of the shims, the RV core stabilizing lugs and the cap screws were checked to confirm the integrity of each screw. Neolube, a dry film lubricant, was applied twice to the threaded surfaces of the cap screws and bearing surface. Fig. 13 shows the assembly of the snubber shims on the RV core-stabilizing lug.

2. When the shims were installed, the cap screws were assembled by hand and tightened according to a three-step tightness method: 160, 213 and 266 ft-lbs. It was important that after the shims were installed, the upper gaps and lower gaps were maintained as constant (0.1016 mm).

3. After the shims were installed, their full widths were measured in six positions to measure the equal intervals. The DAKs were adjusted to vertical degrees at 0.0254 mm/ft for the RV head seating surface. At this point, the vertical degrees of the DAKs and the status of the installed positions were recorded.

4. According to section 2.6 "Combination of the CSB assembly and the RV," the CSB assembly was installed in the RV. When the RV core-stablizing lugs and the RV snubber shims were connected, they were lowered using the hydra-set. The RV centerline and the CSB centerline were adjusted within 0.0254 mm using the DAKs, and the CSB assembly was completely lowered. If the positions of the DAKs changed, the measurements had to be done again.

5. The gaps between the RV and the CSB outlet nozzles were recorded and the offsets of the CSB keyways regarding the DAKs were recorded as well. The offsets of the CSB centerline in relation to the RV centerline were also calculated and recorded.

6. The air hoses, electric power cords and signal cables of the remote measurement system were reconnected to the compressor, electric power, USB hub and RS-485 converter, and electric power was supplied. The length of the snubber shims was measured five times. Once the measurement was complete, the measured data were stored and the measured lengths were recorded. The measured lengths of the shims were confirmed to be within a permissible range (0.381 — 0.508 mm). If the measured lengths of the shims exceeded the permissible range, they would be used after reprocessing.

7. All electric power was turned off, and the air hoses, electric power cords, and signal cables were respectively separated from the compressor, electric power, USB hub and RS-485 converter. Separated air hoses, electric power cables and signal cables were temporarily fixed in the CSB assembly when the RV and CSB assembly were detached in order to avoid interference with cables and pieces of equipment. Finally, the gaps between the CSB and the RV core-stop lugs were measured.

The small doses of radiation can increase the likelihood of developing cancer [Brenner

D. J., Doll R. & Goodhead D. T. 2003, NCR 2006] and is possibility appearing morbidity non-cancer origin [Hildebrandt G. 2010]. The main outcome of the 25-year study of morbidity in different categories of exposed persons in connection with the Chernobyl catastrophe is a significant increase in primary morbidity is not associated with tumor pathology. Not found increased risk of leukemia even among those engaged in reconstruction work. Only recently become apparent relationship of radiation exposure with increasing number of non-oncological pathology such as cardiovascular diseases [Preston D. L, Shimuzu Y. & Pierce D. A. 2003]

Statistically significant increase in the spread of non-neoplastic diseases is shown in [Bouzounov O. V., Tereshchenko V. M. 2010]. At the same time structure of morbidity leading place is occupied by diseases of digestive, circulatory and nervous system The maximum level and the largest number with the non-neoplastic diseases statistically confirmed by link determined in the dose > 0,25 Gy for subcohort study group. Long-term effects of chronic low-intensity exposure to low doses for exposed persons in this category until the end is unclear.

The rapid growth of non-tumor morbidity in exposed populations reflects certain changes in the systems that control the growth, development and aging, namely stereotypes matching sequence and intensity of reading the genetic information in different cells. It is shown that this task cannot perform any neural mechanism, or hormonal agents with their ability to alter the rate of metabolic processes. Neurotransmitters, hormones and their receptors do not possess sufficient ontogenetic variability and dispersal [Poletayev A. B. 2008].

According to modern ideas exactly a physiological autoimmunity throughout life provides a readout of genetic information in different cells of the whole organism [Churilov L. P. 2008, Maltsev V. N. 1983, Zaichik A. Sh. & Churilov L. P. 2008]. Singularity of the function of autoantibodies (AuAB) compared with other regulatory substances is considerably longer their half-lives ranged from 10 to 50 days. Therefore the system of autoimmunity has greater inertia. Autoantibodies regulate slow the physiological processes that continue some days and weeks [Ashmarin I. P., Freidlin V. P. 2005]. It is postulated that a mild autoimmune response to their own antigens is a necessary condition for the normal functioning of the immune system and is a prerequisite for the normal regulation and synchronization of cellular functions and morphogenesis [Churilov L. P. 2008]. Additional conditions for unmasking antigens of tissues and organs and represent them immune cells with subsequent increased production of specific AuAB appear as a result of exposure to ionizing radiation.

It is well known that oxygen absorbed by the mitochondria is converted into adenosine triphosphoric acid (ATP). About 5% of oxygen consumed by tissues is converted into free radicals such as superoxide, hydrogen peroxide, hydroxyl radical, singlet oxygen, peroxynitrite (reactive oxygen species, ROC) with unpaired valence electrons. Most of the ROS are produced continuously in cells as byproducts of normal cellular metabolism (mainly due to a small leakage of electrons to the mitochondrial respiratory chain, as well as other reactions in the cytoplasm), and do not cause damage to cells. An excess of ROS under intense ionizing radiation exceeds the protective capabilities cells and can cause serious cell disorders (eg, depletion of ATP). The increase of free radical molecules and their products have a place in the development of the state of homeostasis, has been called oxidative stress. Slow development of oxidative stress triggers apoptosis, and its intensive development leads to necrosis. Postradiation apoptosis is characterized by maintaining the integrity of the cell plasma membrane and the lack of exposure of intracellular contents from cells of the immune system. In the end, remnants of apoptotic cells are removed by exfoliation in intraorganic space and subsequently excreted from the body. But the shortage of ATP, in particular after irradiation, the energy dependence of the mechanisms of apoptosis are disabled, and the cell dies with loss of cell membrane integrity and release of macromolecular components (eg, ALT, AST, etc.) into the intercellular space. Necrosis caused an immune response in the form of inflammation — leukocyte infiltration of the affected tissue, interstitial fluid accumulation and subsequent induction of specific immune responses (specifically sensibilized T-lymphocytes and autoantibodies) to the unmasked and recognized by lymphocytes of intracellular components. According to data of many researchers the AuAB are primarily the attribute of the norm. They can be identified in healthy individuals. [Churilov L. P. 2008, Zaichik A. Sh. & Churilov L. P. 2008b, Poletayev A. B. 2005, Cohen I. R. 2005, Harel M. & Shoenfeld Y. 2006, Shoenfeld Y. 2008]. The AuAB involved in the process of apoptosis, cleaning the body from catabolic products, modulation of the activity of many enzymes and hormones, as well as perform the transport function [Poletayev A. B. 2008a]. It was shown that antibodies against nuclear antigens can penetrate into the cell nucleus in vivo and stimulate the synthesis of RNA and DNA in target cells [Zaichik A. Sh. & Churilov L. P. 1988].

Notkins in 2007 hypothesized that natural AuAB can be very informative not only precursors of autoimmune diseases, but also a variety of somatic diseases and syndromes [Notkins A. L. 2007]. It is important that changes in the content of organ natural AuAT in most cases, ahead of the clinical manifestation of appropriate forms of pathology. If for example the content of "cardiotropic", hepatotropic", "neurotropic" AuAB a concrete person within the borders of the norm, this suggests that the intensity of apoptosis, respectively, cells of the heart, liver or nerve tissue does not go beyond the norm. Persistent changes, for example, from "hepatotropic" AuAB should be regarded as a sign of the possible formation of a pathological process in liver tissue, even if at the time of examination are no clear clinical symptoms or specific biochemical changes [Poletayev A. B. 2008a, Churilov L. P. 2008, Notkins A. L. 2007, Zedman, A. J.W. & Vossenaar E. R. 2004].

The important role of the abolition of immune tolerance in the occurrence of non-viral hepatitis research shows serum levels of antibodies to liver-specific lipoprotein (LSP) for various categories of people. [Kovalev V. A. & Senyuk O. F., 2008].

The LSP which was first isolated by Meyer zum Burschenfelde and Miescher in 1971 is considered to be specific poly antigen for the liver. Native LSP is mixture of antigenic determinants of the substrate from the membranes of hepatocytes and contains soluble and membrane components were isolated by gel filtration (chromatography) supernatant after ultra — centrifugation of liver homogenate [Manns M, Gerken G. & Kyriatsoulis A, 1987, Ballot

E. , Homberg J. C. & Johanet C. 2000]. It is known that LSP is found not only in the liver, but also in some other organs [Garcia-Buey., Garcia-Monzon C. & Rodriguez S. 1995]. Therefore, the total increase in antibody levels to the PSL can be seen as a sign of abolition of immune tolerance to many organs and tissues of the human body [Kovalev V. A. & Seniuk O. F., 2008].

There are three outcomes for the cell, if the cellular radiation damage is not adequately repaired. The cell may die, or will delay it or keep playing with the viability of new qualities, or mutations as the basis for the development of remote descendants (See figure 17). The consequences of the first approach in the development of cells after irradiation described below.

The linear non-threshold concept is used as the primary standard for radiation protection and risk assessment for many years. It suggests that damage induced by low doses of radiation do not contribute significantly to increased risk of disease because a significant amount of endogenous genome damage occurs during life and they are restored in cells with high probably. In fact endogenous damages (ED) constantly appear in the cells. Some of them are due to thermodynamic processes, the hydrolysis reaction, while others arise from the effects of free radicals generated by cell during its life and still others are a necessary component of metabolism (DNA breaks accompany the process of differentiation, recombination, etc.) [Lindahl T. 1993, Bont R. D. & Van Larebeke N. 2004]. According to [Lindahl T., 1993] per one day in the DNA of one cell may have more than 50,000 endogenous damages as single-strand breaks (SSB) and 10 ones as double-strand breaks (DSB).

Many studies suggest that most of the radiation damage (RD) occurring in the DNA of cells differ significantly in their chemical nature from the ED. The main difference between DNA damage induced by ionizing radiation, from the ED was its the complexity of their chemical nature, and clustering. The proportion of complex, critical for the fate of the cell damage is much higher when exposed to IR.

When treatment of mammalian cells to H2O2 ratio of DSB to SSB is 1: 10 000 [Bradley M. O. & Kohn K. W. 1979] while under the influence of IR is much higher -1:20 [Shikazono N., Noguchi M., & Fujii K. 2009]. Many of the DNA RD are not accidental, are located in close proximity and have the cluster grouping. They are formed as a result of coincidence of two or more single damages within 1-2 rotation the DNA helix [Hada M., Georgakilas A. C. 2008, Sutherland B. M., Bennet P. V. & Sidorkina O. 2000]. Especially the massive clustering of DNA damage occurs when the ionization tracks pass along chromatin fiber. In this case, they may cover DNA regions with an average size of about 2000 base pairs [Radulescu I., Elmroth K. & Stenerlow B. 2004]. At the same time, we know that the probability of occurrence of endogenous clustered DNA damage is extremely low [Bennett P. V., Cintron N. S. & Gros L. 2004]. Accumulation of ED does not occur because in the cells are constantly functioning mechanisms of reparation, specifically targeted at removing various types damages [Friedburg E. C., Walker G. C. & Siede W. 2006].

The complex nature of RD of DNA and the presence of cluster groups can be regarded as the first cause, which creates difficulties for repairing systems cells. Damages repair processes within the cluster can break down in various stages of excision repair and lead to the formation of additional or inaccurate DSB repair, important for cell survival, mutagenesis, and the risk of malignancy [Ide H., Shoulkamy M. I. & Nakano T. 2010]. The second reason for the low efficiency of repair of RD may be a relatively low amount of DNA damage. Therefore, radiation effects in low dose range and low power are certain features associated not only with destructive modifications but with deducing the cellular genome at a different level of activity. High doses of ionizing radiation via activation of cell cycle control points — checkpoints, blocking the synthetic phase of the cell cycle (S) and the transition from G1 to S phase and G2 to M (mitotic) phase and support the repair of DNA. In this case, small doses can not activate the G2/M chekpoynt-arrest and DNA repair are not activated when the number of DSB DNA damage and the MNF up to 10 — 20 per cell.

In this case, heterochromatin little relaxes, and access of repair enzymes to sites of DNA damage worsens [Fernet M, Megnin-Chanet F. & Hall J. 2010, Grudzenskia S., Rathsa A., Conrada S., 2010, Marples B., Wouters B. G. & Collis S. J. 2004, Gaziyev A. I. 2011]. There is the third of the possible causes of low efficiency of repair of critical DNA damages.

The major part of radiation induced DNA damages are represented by DSB and crosslink between strands, which slowly and inefficiently repaired and are responsible for various end effects — from the radiation death of cells to the appearance of chromosome aberration, gene mutations and neoplastic transformation. [Pfeiffer P., Gottlich B. & Reichenberger S. 1996]. Therefore radiation effects in low dose range and low rate dose have the certain features associated not only with destructive changes than with deducing the cellular genome at another level of activity. Analysis of many studies suggests that DNA damage caused by the IR, increase linearly with dose, but the reaction of cells on these lesions, the efficiency of repair of the most complicated critical damages can be nonlinear. Because irradiation is decreased expression of proteins, enzymes of different systems providing the stability of DNA as a result of the accumulation of DNA damage is not recovered or recovered with errors and fix mutations. Dysfunction of many genes, regulatory systems and cellular processes that are ultimately linked to the development of various pathologies, including carcinogenesis. A significant part of replication errors — spontaneous mutations — can be harmful to an organism. There are the basis for the occurrence of hereditary diseases, carcinogenesis, etc. were revealed.

The advanced evaluation of phase analyses of corrosion products from different parts of VVER-440 steam generators via Mossbauer spectroscopy is our active and unique contribution in this area. The scientific works go over 25 years. The first period (mostly 80- ties) was important for improving Mossbauer technique. The benefit from this period is mostly in experience collection, optimization of measurement condition and evaluation programs improvement [5]. Unfortunately, not all specimens were well defined. Having in mind also different level of technique and evaluation procedures, it would be not serious to compare results from that period with the results obtained from measurement after 1998.

Nuclear fuel used in the RBMK-1500 (the reactor of Ignalina NPP in Lithuania) is slightly enriched with uranium in the form of uranium dioxide. According to RBMK-1500 design, low-enriched (2%) uranium fuel was used since the begging of Ignalina NPP exploitation. Later this fuel was mostly fully replaced by a little higher-enrichment (2.4% and 2.6%; 2.8%) uranium fuel with a burnable erbium absorber. The change of fuel allows improving safety and economic parameters of the plant.

Fuel pellets have a 11.5 mm outer diameter and are 15 mm long. The fuel pellets have hemispherical indentations in order to reduce the fuel column thermal expansion and thermo-mechanical interaction with the cladding. The 2 mm diameter hole through the axis of the pellet reduces the temperature at the center of the pellet, and helps to release the gases formed during the operation. The pellets placed into a tube with an outside diameter of 13 mm compose a fuel rod. The active length of RBMK-1500 fuel rod is approximately 3.4 m.

The tube (fuel cladding) material of the fuel rod is an alloy of zirconium with one percent niobium. The fuel rods are pressurized with helium and sealed. The fuel pellets are held in place by a spring. 18 fuel rods, arranged within two concentric rings in a central carrier rod, contain the fuel bundle with an inside diameter of 8 cm [1].

Active core height is 7 m in RBMK type reactors. Thus, the complete fuel assembly is made up of two bundles, which are joined by means of a sleeve at the central plane. The lower bundle of the fuel assembly is provided with an end grid and ten spacing grids. The central tube and the end spacer are also made from the zirconium-niobium alloy. The remaining spacers are made from stainless steel and are rigidly fixed (welded) to the central tube. Apart from the spacers, the top bundle also has intensifying grids, which act as turbulence enhancers to improve the heat transfer characteristics. The fuel tubes are mounted so that axial expansion of the upper or lower bundles takes place in the direction towards the center of the core. The total mass of uranium in one fuel assembly is approximately 110 kg [1].

The fuel channels, where the fuel assemblies are placed, consist of three segments: top, center and bottom. The center segment is an 8 cm inside diameter (4 mm thick wall) tube, made from zirconium-niobium alloy. The top and bottom segments are made from stainless steel tube. The center segment of fuel channel, set in the active core region, and zirconium — niobium alloy warrant the low thermal neutron absorption cross-section.

The fuel channel tubes are set into the circular passages which consist of aligned central openings of the graphite blocks and stainless steel guide tubes of the top and bottom core plate structures to maintain the core region hermetically sealed. The reactor core is constructed of closely packed graphite blocks stacked into approximately 2500 columns with an axial opening. Most of the openings contain fuel channels. A number of them also serve other purposes (e. g. instrumentation, reactivity regulation). The total mass of graphite is about 1700 tons. The fuel channels together with graphite stack are placed inside the leaktight reactor cavity.

The fuel channel tubes also provide cooling for the energy deposited in the graphite moderator of the core region. In order to improve heat transfer from the graphite stack, the graphite rings surround the central segment of the fuel channel. These rings are arranged next to one another in such a manner that one is in contact with the channel, and the other with the graphite stack block. The minimum clearance between the fuel channel and the graphite ring is 1.15 mm, and between ring and graphite stack — 1.38 mm. These clearances prevent compression of the fuel channel tube due to the radiation and/or thermal expansion of the graphite stack [1].

(Smith, 1968) assumes that kinetic energy of the liquid is equivalent to that of the two-phase mixture and a constant fraction k of liquid phase is entrained with the gas phase. The value к = 0.4 was chosen to correspond with the best agreement to experimental data for flow in a vertical tube. Using the Smith correlation, the slip is defined as follows.

where x is the mass quality, pg is the density of the gas phase and p1 is the density of the liquid phase.

3.3.2 Drift-flux model

The main formulation of drift-flux model was developed by (Zuber and Findlay, 1965). This model takes into account both the two-phase flow non-uniformity and local differences of velocity between the two phases. The slip is defined as follows.

where U gJ is averaged gas phase drift velocity.

Where m is the mass flux

The remaining two unknowns are empirical and (Lellouche et al., 1982) is used to estimate these.

3.3.3 Schrage correlation

The correlation by (Schrage, 1988) is based on empirical data from an experimental test section, which measures void fraction directly. This test section has two valves capable of isolating a part of the flow almost instantaneously.

The correlation is based on physical considerations and assumes two different hypotheses:

The Schrage correlation is as follows:

Eg / Egh = 1+0.123 Fr-°-191lnx

with

This correlation was established with an air-water mixture, but it remains valid for any other phase flow.

Regulators all over the world formally base their decisions about the acceptability of a particular radioactive waste disposal system upon the related performance (safety) assessment. The key element of this assessment is dose evaluation. However, the requirements for the certainty/accuracy/validity of such evaluation are not clearly defined in advance and are a subject of development in the dose evaluation process itself. Therefore, dose evaluation, as well as the associated licensing procedure that builds on compliance assessment, seems to be a less appropriate approach due to the uncertainty involved. An alternative method for assessing human exposure in the framework of long­term safety assessment should be developed. Such a method, integrated with the concept of reasonable assurance (IAEA, 1997), should build on indicators of future exploitation of the environment — therefore a clear link to spatial planning in site selection process where human activities remain the basis for future exposure assessment (Kontic et al., 1999).

Since this approach is more fundamental, direct and transparent than dose or risk assessment, it is expected that it will be more powerful in confidence building among different social groups, i. e. scientists, regulators, the public and politicians. Eventually, it is also expected to be effectively applied in comparative evaluations of various energy options. In addition, certain ethical dilemmas in the licensing process connected to regulatory decision making in the presence of uncertainty and in the context of the disposal of long lived radioactive wastes, could also be reduced if dose or risk are avoided as individual numerical safety indicators.

3.1.2 Scientific treatment of specific uncertainty and predictions related to spent fuel

In this sub-section an analysis of the impact of uncertainty (associated with the quantity of radioactive waste produced by Krsko NPP in its anticipated operational time) on the waste-disposal strategy, particularly the selection of the disposal option, is presented. The dilemma is whether to build a shallow land repository for the LILW and to treat all high — level and long-lived waste separately; or to adopt deep geological disposal as the option for all waste types produced in the country. Tightly connected to these questions is the credibility of the evaluation of health consequences due to radioactive waste disposal. Indicators can, for example, be the dose and risk in the presence of uncertainty associated with the waste characteristics on the one hand, and societal characteristics and human habits in the distant future on the other. The approach and methods applied in the analyses were as follows: [7]

324. Using 1,876 MW as the nominal power of the plant, and 48.7 t of uranium per cycle, one obtains 11,857 MWd/ tU. When this is rounded off, 12,000 MWd/ tU for burn-up and 320 effective days of operation at full power are obtained.

• A twelve-month cycle was assumed (i. e. the cycle lasts 365 days); the operational period is 320 days and the cooling (decay) period between cycles is 45 days (actually used for refuelling and maintenance).

• One batch of fuel consists of 40 elements, containing 16.24 tonnes of uranium, and on average represents one-third of the total amount of fuel in the cycle (there are three different batches in the reactor during operation). Each batch remains in the reactor for the three following cycles — except the first, second, penultimate and last batches. The real situation is more complicated but corresponds roughly to these assumptions.

• Being aware of the differences between these assumptions and the real operational data for Krsko NPP, a screening calculation of the activity of spent fuel for the first 13 cycles was made, for the purpose of further calibrating the model. The results for the model and those based on operational data differ very little — see Table 3 for details.

• The content (mass) of uranium isotopes per fuel batch is given in Table 4.

• The mass of zircaloy (Zr-40) per batch is 4012.5 kg; the mass of oxygen (O-16) is 2183.5 kg.

• The average power per tonne of uranium is 37.5 MW; the average power of the batch is 609 MW.

Table 3. Comparison of the calculated (model) results and operational data

The calculated changes of activity of activation products (AP), actinides (ACT), fission products (FP) and total activity per fuel batch with time are presented in Figure 4. The
illustration is for model Batch 6; however, the figures are similar for other model batches. Batch 6 goes into the reactor in the fourth cycle (year). At the moment of irradiation, the total activity immediately increases by about eight orders of magnitude. Before that, the activity is constant at a level of 1.9*106 MBq (the activity of approximately 16 tonnes of non — irradiated fuel). During irradiation, this activity rises slightly from 1.23 to 1.28*1014 MBq, while during the cooling period of 45 days it drops by approximately two orders of magnitude. Each batch stays in the reactor for three successive cycles (except the first, the second, the penultimate and the last), whereupon the batch goes into the spent fuel pit for ultimate cooling and decay. It should be noted that the scale of both axes is logarithmic, so that the origins of axes are avoided in the illustrations.

Fig. 4. Activity of model batch 6 over a million years

Model results for all the fuel are presented in Figure 5, which shows time changes in total activity. With regard to activity during the first 34 cycles, an almost linear increase can be identified due to the collection of spent fuel in the spent fuel pit — one batch per cycle/year. After the 35th cycle, i. e. at the end of the assumed operation of the plant, all three batches from the reactor will be placed in the spent fuel pit at the same time, which is seen as an noncontinuous increase in activity. Activity then decreases, depending on the radionuclides contained in the spent fuel. Note again that the scale of the axes is logarithmic. The values of total activity and decay heat for all spent fuel at selected time — points are summarised in Table 5.

The model adequately represents the overall operation of the plant. This was proved in the process of calibrating the model, where data for the past thirteen cycles were used for comparison. However, fuel enrichment, as well as other key operational elements in future cycles, may not remain constant, since an upgrade of the plant’s power in parallel with the replacement of the steam generators has been achieved. Extension of the fuel cycles was also adopted/made. This was the reason for the analysis of the changes in the activity and radionuclide inventory of spent fuel, due to different fuel enrichment and the prolonged operation of the plant. The adopted variation in fuel enrichment was 1% above and below the value applied in previous calculations, i. e. 4% of U-235.

With regard to the prolonged operation of the plant, a five-year variation was applied. All the variations were simulated for the period following the replacement of the steam generators, i. e. after the 17th cycle. The differences are presented in Figures 6 and 7, respectively. It is clear that the differences are so small that they can be neglected, since they are not relevant to the overall waste management strategy. Moreover, the conclusion which can be drawn from this result is that no benefit can be expected in terms of improved safety connected with radioactive waste disposal whether Krsko NPP were closed down immediately or operated for almost a further 12, or even 40 years as the new National Energy Programme suggests.

The results of the modelling show that the main contributors to fuel activity during the period approximately 200 years after irradiation are the fission products; after that, actinides will prevail. The total expected activity of the spent fuel after one million years is 4,8*1014 Bq. The main contributors to this activity are the radionuclides of the U — and Np-chains. Residual thermal power is about 1.0*105 W approximately 200 years after irradiation, about 1.0*104 W after 10,000 years, and about 250 W after one million years.

The problem of uncertainty, which can be treated scientifically, is manageable. It was recognised that the basic characteristics of this waste can be accurately predicted, since all the sources of uncertainty are well defined, understandable and therefore controllable. Residual uncertainty does not change the overall picture of the waste, which would mean that the predictions could clearly be used as a basis for policy-making, i. e. creating a strategy for radioactive waste management, decision-making and also for communication with the public.

With regard to confidence building connected to radioactive waste disposal, the recommendation is that prompt, clear and complete informing of all interested parties and the general public should take place. It should be clearly stated that the spent fuel from Krsko NPP, and a part of the decommissioned waste, will remain radioactive above today’s prescribed levels for hundreds, thousands or even a million years from now. Consequently, a strategy built upon waiting for the activity to "disappear" cannot be effective. Doubts and uncertainties regarding safety assessments in a timeframe of a million years should also be revealed. At the same time, efforts should be made to present the concept of reasonable

assurance (IAEA, 1997) as the most reliable method, and as the basis upon which a waste

management strategy can rely.

4. Concluding remarks

Strategic environmental considerations of nuclear power should inevitably cover the

following (however, not restricted to):

• Comparative evaluation of electricity generation technologies; the evaluation should, in addition to topical consideration, thoroughly deal with the ways on how to overcome specifics and details of individual analysis of a certain technology which is usually influenced by the specific characteristics of the site compared to others in its category, the manufacturing and design characteristics, the power, lifetime and the operating conditions. Results are therefore difficult to transfer from a country to another or one generation unit to another, as most major environmental, economic and social impacts, with the exception of e. g., climate change, are heavily site-dependent. Application of proper indicators in such a comprehensive comparative evaluation may be of practical help and guidance;

• The energy system as a whole; the electricity grid and market issues are rarely taken into consideration when making comparative evaluation. Similarly, the issue of increased share of intermittent RES, its impact on the energy system, and consequent need for the adaptation of environmental impact appraoches by taking into account actual share of each generation technology as provided in the energy system;

• Uncertainties; uncertainties and limitations of various methodologies may be acknowledged by the authors of the studies but those are rarely taken into account when results (or only some of them) are used by policymakers. Strategic considerations should provide guidance/recommendations on how to deal with the uncertainties in the decision-making process associated with comparative evaluation of different electricity generation technologies. This is especially relevant when deciding about long-term impacts, e. g. nuclear waste disposal or societal and spatial consequences of climate change.

The sequential determination is based on radiochemical procedure that consists of three steps performed by anion-exchange chromatography, extraction chromatography, using Eichrom resins, and precipitation techniques. For each aliquot was added 2 mL of Ni (0.01 mol L-1), 1 mL of Sr (0.02 mol L-1) and 2 mL of Fe (0.01 mol L-1) as carriers and yield monitor. In the first step, the separation of Pu by ion exchange chromatography, anion exchange column (Dowex 1×8, Cl-form. 100-200 mesh, Sigma Chemical Co. USA), is based on the formation of anionic complexes of the Pu(IV) with NO3- or Cl — in concentrated HNO3 or HCl. In the second one the effluent from the anion exchange column was used to separate Am and Sr by co-precipitation with oxalic acid of U, Fe and Ni that remains in the filtrate.

In the third step we use the filtrate to separate Ni from U and Fe. The filtrate was heated to dryness and the solid obtained was dissolved in 30 mL of concentrate nitric acid and heated to dryness in order to destroy the excess of oxalic acid. The solid obtained was hot dissolved in 30 mL of 3:2 nitric acid and was diluted to 200 mL with deionized water. The pH of the solution was corrected to 9.0 with ammonia hydroxide for co-precipitation of iron hydroxide and uranium while Ni forms a soluble [Ni(NH3)4]2+complex.

After filtration, the filtrate was heated to dryness and retaken with 20 mL of HCl concentrate and again heated to dryness. The solid obtained was dissolved in 25 mL of 1 mol L-1 HCl and was added 1 mL of ammonium citrate to the sample being the pH adjusted to 8-9 with ammonium hydroxide (Eichrom Technologies, 2003). A nickel resin extraction chromatography column (Eichrom Industries Inc. USA) was pre-conditioned with 5 mL of solution 0.2 mol L-1 ammonium citrate that has been adjusted to pH 8-9 with ammonium hydroxide. The column was loaded with the sample and rinsed with 20 mL of solution 0.2 mol L-1 ammonium citrate. Nickel was eluted with 10 mL of solution 3 mol L-1 HNO3. Figure 1 represents the flowchart for sequential separation of radionuclides in a sample of radioactive waste.

The second group of accidents from the second category of BDBA (total damage of the core or its components with the reactor maintaining its overall structural integrity) are the accidents when heat-up of the reactor core occurs after the reactor shutdown (2.2 — see Figure 5). The heat-up, damage and melting of the reactor core could occur in the late phase after the reactor shutdown due to loss of the long-term cooling. The results of the Level 1 probabilistic safety assessment of the Ignalina NPP showed that in the topography of the risk, transients dominate above the accidents with LOCAs and the failure of the core long-term cooling are the main factors of the frequency of core damage. The initiating events leading to the loss of long term cooling accident are such: [4] impossibility to feed RCS by water. The results of the analysis [20, 25] suggest that approximately 1.5 hours after the beginning of the accident, a dangerous heat-up of fuel rods and FC walls starts. Three ways of potential accident management for the loss of the long-term core cooling were discussed in [25]:

• decay heat removal by ventilation of DS compartments,

• decay heat removal by direct water supply into the reactor cavity,

• de-pressurisation of the reactor coolant system and water supply to the GDH from ECCS hydro-accumulators, deaerators or using non-regular means.

The results showed that the first two ways are inexpedient. The ventilation of DS compartments and the direct water supply into the RC are not sufficient to remove the decay heat from the core. However, the de-pressurisation of RCS and the following water supply from regular and non-regular means to the GDH in the case of loss of long-term cooling gives considerably better results compared with the other two measures. The performed analysis [25] demonstrated that the reactor core cooling by RCS depressurisation and water supply from deaerators give additional four hours for the operator to install water supply from the external (artesian) water source. Thus, this way of accident management is recommended to be included in the RBMK-1500 accident management programme.

In the case of loss of intermediate cooling circuit or loss of service water, the consequences of the accidents will be very similar. In these last cases there are no direct signals for the reactor shutdown; however, these initiating events lead to the loss of feedwater supply and MCP will be tripped due to the insufficient cooling. Reactor scram would appear on secondary parameters and further sequence of the accident will be the same as in case of the station blackout: due to decay heat the water in the core is evaporating and the core heat-up process starts. Core overheating can be avoided by using water stored in the ECCS hydro­accumulators, deaerators and other water sources located at NPP.

Figure 21 shows the behavior of fuel, claddings, channel walls, and graphite stack temperatures, calculated using RELAP5 and RELAP5/SCDAPSIM codes (models presented in Figure 6 and Figure 7), in case of RBMK-1500 reactor blackout, without any additional water supply. Such event, which is developing into a severe accident, serves as an example for the discussion of severe accident phenomena in RBMK type reactors. As it is presented in Figure 21, the failure of fuel channels could occur after ~3 hours. The failure of FCs is expected because the pressure in RCS is nominal, and the acceptance criterion for fuel channel (650 °C) will be reached (Figure 21). It is assumed that after ~3.8 hours the operator opens one steam relief valve to discharge steam from RCS. This action allows depressurizing RCS and prevents ruptures of FCs. If no operator actions were taken (no manual depressurization), then approximately 50 FCs with a higher power level could be ruptured. If all FCs ruptured within short time interval, the reactor cavity would be destroyed and the consequences of the accident would be similar to the Chernobyl accident.

Figure 22 shows the long-term behavior of fuel, claddings, fuel channels, and graphite stack temperatures, calculated using RELAP5/SCDAPSIM code (model presented in Figure 7), in case of RBMK-1500 reactor blackout, with operator intervention (opening one of steam relief valves for RCS depressurization). As the operator starts depressurization, the accident scenario continues at low pressure. Due to the pressure decrease the rest of coolant in pipelines below the reactor core starts boiling and steam cools down the core for a short

time. As it is shown in Figure 22, the temperatures of the core components decrease for a short period of time. However, after 4.3 hours the second (repeated) heat-up of the reactor core elements begins. When the temperature of fuel cladding increases above 800 °C, the failure of fuel claddings occurs due to ballooning. The ballooning happens because at that time the pressure in RCS (outside fuel rods) is close to the atmospheric and the pressure inside fuel rods is high. The conditions for fast oxidation of claddings and fuel channels,

-10 12 3 4

Time, h

Time, h

Fig. 22. Station blackout, when operators depressurizes RCS. Main consequences in case of station blackout

made from zirconium-niobium alloy, are reached after the fuel cladding and fuel channel temperatures exceed 1000 — 1200 °C (~15 hours after the beginning of the accident). Due to steam-zirconium reaction the generation of hydrogen starts. The oxidation and hydrogen generation processes terminated after the pressure in RCS decreases down to atmospheric. This indicates that there is no steam in RCS, thus the steam-zirconium reaction is impossible. Later the processes would continue at low pressure in RCS and RC would remain intact.

When the temperatures of fuel claddings and FC walls reach 1450 °С, the melting of stainless steel grids and zirconium starts at 1760 °С (Figure 22). Probably at the same time the fuel channels will fail. At temperature of 1930-2050 °С and 2330 °С the melting of aluminum oxide (control rods claddings) and boron-carbide (control rods elements) in the separate control rods channels starts. The formation of ceramic (U, Zr, ZrO2) starts at temperature of 2600 °С. The analysis performed using RELAP5/SCDAPSIM code shows that fuel melting (melting of ZrO2 and UO2) starts at low pressure, approximately 50 hours after the beginning of the accident at temperatures of 2690 °С and 2850 °С respectively (Figure 22). Such comparably slow core heat up process is due to the high inertia of graphite stack, which provides a heat sink. Hence, the high pressure melt ejection and direct containment heating — the phenomena more related to PWR design — could not occur at RBMK-1500 reactor due to the limited space inside the reactor cavity. However, to cool down the reactor, it is necessary to start water supply into the fuel channels within the first 15 hours after the beginning of the accident. The water supply in later phases could lead to a fast steam-zirconium reaction and it could accelerate core damage processes.

6. Conclusion

In this paper the specifics of RBMK reactors design was presented. Based on the specific feature of RBMK, possible Beyond Design Basis Accidents were divided into four groups:

• accidents with no severe damage of the core;

• accidents leading to a severe core damage accompanied by containment of the core fragments in the reactor cavity and accident localization system or other reactor buildings;

• accidents when heat-up of the reactor core occurs during the reactor operation or within the first seconds after the reactor shutdown;

• accidents when heat-up of the reactor core occurs after the reactor shutdown.

The deterministic analysis of all these groups of BDBA was performed using a system of thermal hydraulic computer codes RELAP5 and RELAP/SDAPSIM. The consequences of these BDBAs and possible accident mitigation measures were discussed. [5]

when there are no possibilities to inject water in the reactor using regular means, the operators can supply the water into reactor from ECCS hydro-accumulators, later to perform the RCS de-pressurization by opening manually steam relief valve. Finally, after the pressure in RCS is reduced, the low-pressure non-regular water sources can be used (deaerators and artesian water).

• The accidents in the second group (accidents leading to a severe core damage accompanied by containment of the core fragments in the reactor cavity and accident localization system or other reactor buildings) are initiated due to misbalance between energy source and heat sink. If the emergency core cooling system is not activated, or the amount of supplied water is less as required, the core meltdown can occur. Based on the performed deterministic analysis the setpoints for ECCS activation were selected. The capacity of reactor cavity venting system was increased to prevent failure of reactor cavity in case of multiple fuel channel rupture (up to simultaneous rupture of 16 fuel channels).

• The third group contains the accidents when heat-up of the reactor core occurs during the reactor operation or within the first seconds after the reactor shutdown, when decay heat is high. Because fast process of heat-up of fuel rods in this case — there is no time for operator actions in this case. The new algorithms for reactor shutdown and fast emergency core cooling system activation were proposed for RBMK-1500 to prevent overheating of fuel in local flow stagnation or flow blockage in the group of fuel channels cases. Also the new ECCS activation algorithm was developed to cooldown the reactor in the case of loss of natural circulation due to a sharp decrease of pressure in the RCS. To prevent the catastrophic core damage in the anticipated transients without reactor scram case (when main reactor shutdown system fails to shutdown reactor) the additional emergency protection was implemented in the RBMK-1500.

• The forth group of accidents — the accidents when heat-up of the reactor core occurs after the reactor shutdown. The performed analysis shown, that even in the case of failure of all design (regular) and non-regular means to cooldown the rector in the case of loss of long term core cooling, the core heat-up process is slow in RBMK-type reactors. In the station blackout case, to prevent failure of reactor cavity at high pressure, the operators are required to open the steam relief valve manually, to start RCS depressurization. Due to the high inertia of graphite stack, which provides a heat sink, the melting of fuel stats at low pressure not earlier as 50 hours after loss core cooling.

The analysis was performed for the RBMK-1500 reactor (Ignalina NPP, Lithuania), but the main ideas of the accident mitigation are also valid for the RBMK-1000, which are still operating in Russia.

7. Abbreviations

ALS Accident Localization System

ATWS Anticipated Transients Without reactor Scram

BDBA Beyond Design Basis Accidents

BWR Boiling Water Reactor

CPS Control & Protection System

DAZ Acronym for Russian — Additional emergency protection

DS Drum Separator

ECCS Emergency Core Cooling System

FC Fuel Channel

GDH Group Distribution Header

LWR Light Water Reactor

LOCA Loss Of Coolant Accident

LWP Low Water Pipes

MCP Main Circulation Pump

NPP Nuclear Power Plant

RBMK Acronym for Russian — graphite-moderated boiling water reactor type

PWR Pressurized Water Reactor

RC Reactor Cavity

RCS Reactor Cooling System

RCVS Reactor Cavity Venting System

SRV Steam Relief Valve

TCV Turbine Control Valve