Lu Guangyao, Ren Junsheng, Huang Wenyou, Xiang Wenyuan,

Zhang Chengang and Lv Yonghong

China Guangdong Nuclear Power Holding Co. Ltd.

China

1. Introduction

Two-phase flow exits in many nuclear power equipments, such as containment sump strainer, steam generator, steam turbine, control rod drive mechanism and so on. Experimental investigations are carried out to study the two-phase flow patterns and their transitions in these nuclear power equipments. And the results show that the two-phase flow patterns and their transitions are quite different from those in normal circular tubes.

For tube-bundle channel heat transfer enhancement technique, it has great advantages in high heat transfer efficiency and compact configuration without complex machining or additional surface processing, which has been successfully used in steam generator in nuclear power station and other industrial equipments. The characteristics of boiling flow and heat transfer have an important impact on these industrial equipments. It was found that heat transfer characteristics of fluid flowing in tube-bundle channels were different from those in circular tubes on account of the special geometric frame and different flow patterns in tube-bundle channels (Petigrew & Taylor, 1994). Boiling flow and heat transfer is a complex issue of two — phase flow, and many studies have been conducted in this research area.

Experiments of boiling flow in tube-bundle channels were carried out in order to simulate boiling flow and heat transfer in fuel module of nuclear reactor (Bergles, 1981). The experimental results showed that there were many different aspects of flow patterns and their transitions in tube-bundle channels. Grant & Chisolm (1979) and Ma (1992) made studies of air-water two-phase flow in tube-bundle heat exchangers, in which the vertical flow and horizontal flow experiments were conducted respectively. Grant & Chisolm (1979) found that there were mist flow, bubbly flow, intermittent flow, stratified-mist flow and stratified flow in tub-bundle channels. But Ma (1992) detected stratified flow and wave flow in horizontal channels, and bubbly flow and intermittent flow in vertical channels, which was different from the results gained by Grant & Chisolm (1979). Chan & Shoukri (1987) conducted visualization experiments of refrigerant-113 flowing in tube-bundle channels, of which the tubes were arranged 1×1, 3×1, 3×3, and 9×3 respectively. Sadatomi & Kawahara (2004) carried out experiments in tube-bundle channels, of which the tubes were arranged 2×3. On the basis of experiments, Sadatomi protracted flow pattern maps.

On the basis of the results obtained before, studies are carried out to investigate the characteristics of refrigerant 113 flowing in a tube-bundle channel, of which the tubes are

arranged 2×2. Furthermore, flow visualization experiments are carried out with high-speed camera. Through the comparison between the experimental results and other works, it is found flow regimes and their transitions in the tube-bundle channel are different from other normal circular tubes.

In the event of LOCA (Loss-of-Coolant Accident) or HELB (High Energy Line Break) within the containment of a light-water reactor (LWR), the primary safety concern regarding the long-term recirculation is that accident-generated debris and resident debris may be transported to the recirculation sump screens, which would result in adverse blockage effects and loss of the pumps net positive suction head (NPSH) of the emergency core cooling system (ECCS) and the containment spray system (CSS). The debris may starve the sump and the head loss of water flow through the containment sump strainers may be so large that it would exceed the available NPSH margin of pumps of ECCS and CSS systems. The pressure loss due to the accident-generated debris accumulated on the sump screens would be computed through the debris types and contents which are determined to be destroyed and transported. So the computational researches were carried out to investigate the characteristics of containment sump strainers by Lu et al. (2011a, 2011b, 2011c).

For control rod drive mechanism (CRDM) in nuclear power station, it works in high pressure and high temperature condition and single-phase water is adopted as the working liquid. But two-phase flow would come into being in the cold-state test and hot-state test. The high speed camera is also used in the CRDM visualization tests in cold-state to investigate the characteristics of two-phase flow.

2. Experiments

It was taken an aliquot of 3 mL from the 10 mL solution 3 mol L-1 HNO3 eluted of the column. Measurements of 59Ni were performed with Ultra-LEGe Detector (GUL) with a
cryostat window of Beryllium low energy y-detector containing an active area of 100 mm2, and resolution less than 150 eV (FWHM) at 5.9 keV, from Canberra Industries (USA).

2.3.1 Determination of 63Ni by LSC

It was taken an aliquot of 3 mL collected in a scintillation vial from the 10 mL solution 3 mol L-1 HNO3 eluted of the column,. It was added 17 mL of the scintillation cocktail and the vial was shaken vigorously. Before counting, in order of minimizing luminescence interferences, the vial was stored in the dark for 24 hours.

In order to calibrate the counter and to determine the counting conditions, it was prepared a 63Ni standard solution and a blank solution, in the same conditions of the sample. The counting conditions set up were a time of counting of 60 minutes and a channel interval of 50-400.

Shahab Khushnood et al.[6]

University of Engineering & Technology, Taxila

Pakistan

1. Introduction

Over the past few decades, the utility industry has suffered enormous financial losses because of vibration related problems in steam generators and heat exchangers. Cross-flow induced vibration due to shell side fluid flow around the tubes bundle of shell and tube heat exchanger results in tube vibration. This is a major concern of designers, process engineers and operators, leading to large amplitude motion or large eccentricities of the tubes in their loose supports, resulting in mechanical damage in the form of tube fretting wear, baffle damage, tube collision damage, tube joint leakage or fatigue and creep etc.

Most of the heat exchangers used in nuclear, petrochemical and power generation industries are shell and tube type. In these heat exchangers, tubes in a bundle are usually the most flexible components of the assembly. Because of cross-flow, tubes in a bundle vibrate. The general trend in heat exchanger design is towards larger exchangers with increased shell side velocities, to cater for the required heat transfer capacity, improve heat transfer and reduce fouling effects. Tube vibrations have resulted in failure due to mechanical wear, fretting and fatigue cracking. Costly plant shutdowns have lead to research efforts and analysis for flow — induced vibrations in cross-flow of shell side fluid. The risk of radiation exposure in steam generators used in pressurized water reactor (PWR) plants demand ultimate safety in designing and operating these exchangers.

(Erskine & Waddington, 1973) have carried out a parametric form of investigation on a total of nineteen exchanger failures, in addition to other exchangers containing no failures. They realized that these failures represent only a small sample of the many exchangers currently in service. The heat exchanger tube vibration workshop (Chenoweth, 1976) pointed out a critical problem i. e., the information on flow-induced vibration had mostly been withheld because of its proprietary nature.

Failure of heat exchanger tubes in a bundle due to flow-induced vibrations is a deep concern, particularly in geometrically large and highly rated units. Excessive tube vibration may cause failure by fatigue or by fretting wear. Each tube in a bundle is loosely supported at baffles, forming multiple supports often with unequal support spacing. Reactor components like heat exchanger tubes, fuel rods and piping sections may be modeled as beams on multiple supports. It is important to determine whether any of the natural frequencies be within the operating range of frequencies. Considerable research efforts have been carried out, which highlight the importance of the problem.

Tube natural frequency is an important and primary consideration in flow-induced vibration design. A considerable research has been carried out to calculate the natural frequencies of straight and curved (U-tubes) by various models for single and multiple, continuous spans, in air and in liquids for varying end and intermediate support conditions. (Chenoweth, 1976), (Chen & Wambsganss, 1974), (Shin & Wambsganss,1975), (Wambsganss, et al., 1974), (Weaver, 1993), (Brothman, et al., 1974), (Lowery & Moretti, 1975), (Elliott & Pick, 1973), (Jones, 1970), (Ojalvo & Newman, 1964) and (Khushnood et al., 2002), to name some who have carried out research and highlighted the importance of the calculation of natural frequencies of heat exchanger tubes in a bundle.

The dimensionless parameters required for modeling a system may be determined as follows (Weaver, 1993):

• Through non-dimensionalizing the differential equations governing the system behavior.

• From application of Buckingham Pi-theorem.

• This theorem only gives the number of ks, and not a calculation procedure. So we rely on (i) essentially.

(Shin & Wambsganss, 1975), and (Khushnood et al., 2000) gave the basics of model testing via dimensional analysis. (Blevins, 1977) has described non-dimensional variables such as geometry, reduced velocity, dimensionless amplitude, mass ratio, Reynolds number and damping factor as being useful in describing the vibrations of an elastic structure in a subsonic steady flow. However, other non-dimensional variables such as Mach number, capillary number, Richardson number, Strouhal number and Euler number are also useful in case effects such as surface tension, gravity, supersonic flow or vortex shedding are also considered.

It is generally accepted that the tube bundle excitation mechanisms are (Weaver, 1993, Pettigrew et al., 1991) • cylinder width. Fluid-elastic instability is by far the most dangerous excitation mechanism and the most common cause of tube failure. This instability is typical of self-excited vibration in that it results from the interaction of tube motion and flow. Acoustic resonance is caused by some flow excitation (possibly vortex shedding) having a frequency, which coincides with the natural frequency of the heat exchanger cavity.

With regard to dynamic parameters, including added mass and damping, the concept of added mass was first introduced by DuBuat in 1776 (Weaver, 1993). The fluid oscillating with the tube may have an appreciable affect on both natural frequency and mode shape. Added mass is a function of geometry, density of fluid and the size of the tube (Moretti & Lowry, 1976). Several studies including (Weaver, 1993, Lowery, 1995, Jones, 1970, Chen et al., 1994, Taylor et al., 1998, Rogers et al., 1984, Noghrehkar et al., 1995, Carlucci, 1980, Pettigrew et al., 1994, Pettigrew et al., 1986, Zhou et al., 1997) have targeted damping in heat exchanger tube bundles in single-phase and two-phase cross-flow. (Rogers et al., 1984) have given identification of seven separate sources of damping.

(Ojalvo & Newman, 1964) have presented design for out-of-plane and in-plane frequency factors for various modes. (Jones, 1970) carried out experimental and analytical analysis of a vibrating beam immersed in a fluid and carrying concentrated mass and rotary inertia. (Erskine & Waddington, 1973) conducted parametric form of investigation on a total of 19 exchanger failures along with other exchangers containing no failures, for comparative purpose, indicated the incompleteness of methods available till then and emphasized the need for a fully comprehensive design method. Finite element technique applied by (Elliott & Pick, 1973), concluded that the prediction of natural frequencies was possible with this method and that catastrophic vibrations might be prevented by avoiding matching of material and excitation frequencies. Lack of sufficient data to support comprehensive analytical description for several fundamentally different vibration excitation mechanisms for tube vibration have been indicated in Ontario Hydro Research Division Report (Simpson & Hartlen, 1974). The report also gives response in terms of mid-span amplitude to a uniformly distributed lift for a simply supported tube. A simple graphical method for predicting the in-plane and out-of-plane frequencies of continuous beams and curved beams on periodic, multiple supports with spans of equal length have been presented by (Chen & Wambsganss, 1974). They have given design guidelines for calculating natural frequencies of straight and curved beams. (Wambsganss, et al., 1974) have carried out an analytical and experimental study of cylindrical rod vibrating in a viscous fluid, enclosed by a rigid, concentric cylindrical shell, obtaining closed-form solution for added mass and damping coefficient. (Shin & Wambsganss, 1975) have given information for making the best possible evaluation of potential flow-induced vibration in LMFBR steam generator focusing on tube vibration. A simple computer program developed by (Lowery & Moretti, 1975), calculates frequencies of idealized support with multiple spans. (Chenoweth, 1976), in his final report on heat exchanger tube vibration, pointed out the slow progress and inadequacy of existing methods and a need for field data to test suitability of design procedures. It stressed the need for testing specially built and instrumented industrial — sized heat exchangers and wind tunnel based theories to demonstrate interaction of many parameters that contribute to flow-induced vibrations. (Rogers et al., 1984) have modeled mass and damping effects of surrounding fluid and also the effects of squeeze film damping. (Pettigrew et. al., 1986) have treated damping of multi-span heat exchanger tubes in air and gases in terms of different energy dissipation mechanisms, showing a strong relation of damping to tube support thickness.

(Price, 1995) has reviewed all known theoretical models of fluid-elastic instability for cylinder arrays subject to cross-flow with particular emphasis on the physics of instability mechanisms. Despite considerable difference in the theoretical models, there has been a general agreement in conclusions. (Masatoshi et al., 1997) have carried tests on an intermediate heat exchanger with helically coiled tube bundle using a partial model to investigate the complicated vibrational behavior induced by interaction through seismic stop between center pipe and tube bundle. They have indicated the effect of the size of gap between seismic stop and tube support of the bundle.(Botros & Price, 2000) have carried out a study of a large heat exchanger tube bundle of styrene monomer plant, which experienced severe fretting and leaking of tubes and considerable costs associated with operational shutdowns. Analysis through Computational Fluid Dynamics (CFD) and fluid-elastic instability study resulted in the replacement of a bundle with shorter span between baffles, and showed no signature of vibration over a wide range of frequencies. (Yang, 2000) has postulated that crossing — frequency can be used as a measure of heat exchanger support plate effectiveness. Crossing — frequency is the number of times per second the vibrational amplitude crosses the zero displacement line from negative displacement to positive displacement.

The wear of tube due to non-linear tube-to-tube support plate (TSP) interactions is caused by the gap clearances between the two interacting components. Tube wall thickness loss and normal work-rates for different TSP combination studies have been the target. Electric Power Research Institute (EPRI), launched an extensive program in early 1980’s for analyses of fluid forcing functions, software development and studying linear and non­linear tube bundle dynamics. Other studies include (Rao et al., 1988), (Axisa & Izquierdo, 1992), (Payen et al., 1995), (Peterka, 1995), (Hassan et al., 2000), (Charpentier and Payen, 2000) and (Au-Yang, 1998).

Generally, there are three geometric configurations in which tubes are arranged in a bundle. These are triangular, normal square and rotated square. Relatively little information exists on two-phase cross-flow induced vibration. Not surprisingly as single-phase flow-induced vibration is not yet fully understood. Vibration in two-phase is much more complex because it depends upon two-phase flow regime and involves an important consideration, the void fraction, which is the ratio of volume of gas to the volume of the liquid gas mixture. Two-phase flow experimentation is much more expensive and difficult to carry out usually requiring pressurized loops with the ability to produce two-phase mixtures of desired void fractions.

Two-phase flow research includes the models, such as, Smith Correlation (Smith, 1968), drift- flux model developed by (Zuber and Findlay, 1965), Schrage correlation (Schrage, 1988), and Feenstra model (Feentra et al., 2000). (Frick et al., 1984) has given an overview of tube wear — rate in two-phase flow. (Pettigrew et al., 2000), (Mirza & Gorman, 1973), (Taylor et al.,1989), (Papp, 1988), (Wambsganss et al., 1992) and others have carried out potential research for vibration response. Earlier reviews on two-phase cross flow are provided by (Paidoussis, 1982), (Weaver & Fitzpatrik, 1988), (Price, 1995), and (Pettigrew & Taylor, 1994).

Two-phase cross-flow induced vibration in tube bundles of process heat exchangers and U — bend region of nuclear steam generators can cause serious tube failures by fatigue and fretting wear. Tube failures could force entire plant to shut down for costly repairs and suffering loss of production. Vibration problems may be avoided by thorough vibration analysis. However, this requires an understanding of vibration excitation and damping mechanism in two-phase flow. A number of flow regimes (Table 1) can occur for a given boundary configuration, depending upon the concentration and size of the gas bubbles and on the mass flow rates of the two-phases. Two-phase (khushnood, et al., 2004) flow characteristics greatly depend upon the type of flow occurring.

Vibration of tube in two-phase flow displays different flow regimes i. e., gas and liquid phase distributions, depending upon the void fraction and mass flux. It is known that four mechanisms are responsible for the excitation of tube arrays in cross-flow (Pettigrew, et al., 1991). These mechanisms are: turbulence buffeting, vortex shedding or Strouhal periodicity, fluid-elastic instability and acoustic resonance. Table 2 presents a summary of these vibration mechanisms for single cylinder and tube bundles for liquid, gas and liquid-gas two-phase flow respectively. Of these four mechanisms, fluid-elastic instability is the most damaging in the short term, because it causes the tubes to vibrate excessively, leading to rapid wear at the tube supports. This mechanism occurs once the flow rate exceeds a threshold velocity at which tubes become self-excited and the vibration amplitude rises rapidly with an increase in flow velocity.

Typically, researchers have relied on the Homogeneous Equilibrium Model (HEM) (Feentra et al., 2000) to define important fluid parameters in two-phase flow, such as density, void fraction and velocity. This model treats the two-phase flow as a finely mixed and homogeneous in density and temperature, with no difference in velocity between the gas and liquid phases. This model has been used a great deal because it is easy to implement and is widely recognized which facilitated earlier data comparison. Other models include Smith correlation (Smith, 1968), drift-flux model developed by (Zuber and Findlay, 1965), Schrage Correlation (Schrage, 1988), which is based on empirical data, and Feenstra model (Feentra et al., 2000), which is given in terms of dimensionless numbers.

Dynamic parameters such as added or hydrodynamic mass and damping are very important considerations in two-phase cross-flow induced vibrations. Hydrodynamic mass depends upon pitch-to-diameter ratio and decrease with increase in void fraction. Damping is very complicated in two-phase flow and is highly void fraction dependent. Tube-to-restraint interaction at the baffles (loose supports) can lead to fretting wear because of out of plane impact force and in-plane rubbing force. (Frick et al., 1984) has given an overview of the development of relationship between work-rate and wear-rate. Another important consideration in two-phase flow is the random turbulence excitation. Vibration response below fluid-elastic instability is attributed to random turbulence excitation.

(Pettigrew et al., 2000), (Mirza & Gorman, 1973), (Taylor et al.,1989), (Papp, 1988), and (Wambsganss et al., 1992) to name some, have carried out research for Root Mean Square (RMS) vibration response, encompassing spatially correlated forces, Normalized Power Spectral Density (NPSD), two-phase flow pressure drop, two-phase friction multiplier, mass flux, and coefficient of interaction between fluid mixture and tubes. More recently researchers have expanded the study to two-phase flow which occur in nuclear steam generators and many other tubular heat exchangers, a review of which was last given by (Pettigrew & Taylor, 1994). A current review on this topic is given by (Khushnood et al., 2004)

The use of Finite Element Method (FEM), Computational Fluid Dynamics (CFD) and Large Eddy Simulation (LES) have proved quite useful in analyzing flow-induced vibrations in tube bundles in recent years. Earlier on, only pressure drop and heat transfer calculations were considered as the basis of heat exchanger design. Recently, flow-induced tube vibrations have also been included in the design criteria for process heat exchangers and steam generators.

A remote measurement system was developed to measures the gaps between the RV core- stabilizing lug and the CSB snubber lug using a DT/20/P digital probe sensor. The major characteristics of the remote measurement system are as follows:

— The remote measurement system consists of a measurement sensor section, a pneumatic supply and control section, a power supply section, and a remote control computer and software program.

— The measurement sensor section is intended to measure gaps between the RV core — stabilizing lug and the CSB snubber lug. Those sensors, placed at 0° and 60°, are measured by 24 sensors with a signal cable connected to the channel box #1. Those sensors placed at 120° and 180° are measured by 24 sensors with a signal cable connected to the channel box #2, and those sensors placed at 240° and 300° are measured by 24 sensors with a signal cable to connected the channel box #3. A measurement sensor section is composed of 72 digital probes. This system is able to measure 72 points at once and operate by pneumatic actuation.

— The pneumatic supply and control section is intended to supply air to actuate the sensors by a remote control computer and a software program. The pneumatic supply section consists of an air compressor, an air filter, an air pressure regulator and an air tube. The pneumatic control section is composed of a flow control valve, a solenoid valve, a solenoid valve manifold, a USB orbit module and a T-CON. The solenoid valve and the solenoid valve manifold control operation of measurement sensors and the USB orbit module and the T-CON receive signal data of the digital probe, DT/20/P.

— The power supply section supplies electric power to the electric equipment, including the T-CON.

— The remote control computer and the software program consist of a laptop computer and the software to control digital probes and to process and store the measurement results.

— The channel box contains 24 digital probes, the T-CON, 24 solenoid valves, and four solenoid valve manifolds. The channel is designed for three channel boxes. These boxes were very easy to handle due to their suitable weight and size.

— All of the cables and air tubes are easily connected to the channel boxes.

— The network of the remote measurement system is very stable and causes no disturbance to the EMI environment.

USB HUB{4 Port)

Fig. 14. Block diagram of developed remote measurement system

Reactor

Vessel

CSB

Cablelsignal. power) & Air tube

Remote Measuring system

Solenoid Valve, T-CON, Air Distributor

LSS

Sectional View

Fig. 15. Setup position of the remote measurement system

Table 2 presents a list of the parts of the remote measurement system, and Fig. 14 shows a block diagram of the developed remote measurement system.

As shown in Fig. 15, the channel boxes are located above the LSS in the CSB. The digital probes should be set on the CSB snubber lugs before assembly with the RV. The channel box should be connected to the air compressor, to the remote control computer, and to the electric power source after assembly of the RV and the CSB.

The HPC, figure 5, consists of a lower cylindrical section, an eccentric cone section, an upper cylindrical section and an hemispherical upper end head. For scaling reasons, in order to

have an adiabatic boundary condition in all the wall of the HPC except through the heat transfer plate wall where the condensation has to take place, four groups of containment heaters have been installed permitting the heat transfer takes place only between the CPV and HPC containment. These heaters are located in the exterior surface of the HPC, under the insulation, and above the containment water level. The temperatures of heaters located on the walls of the HPC and the temperatures within the walls of the HPC, between the heaters and the water, are measured. The entire HPC is covered by Thermo-12 hydrous calcium silicate insulation. The HPC level and pressure are measured.

Fig. 5. OSU-MASLWR containment structures (Reyes et al., 2007; Mascari et al., 2011d; Mascari et al., 2011e).

The CPV consists of a tall right cylindrical tank covered by Thermo-12 hydrous calcium silicate insulation. The CPV level and water temperature are measured. One disk rupture is connected between the HPC and the CPV.

The heat transfer plate, having the same height of the HPC without the hemispherical head, provides the heat conduction between the HPC and CPV. The heat transfer plate is scaled in order to model the heat transfer area between the MASLWR design high pressure containment vessel and the cooling pool in which it sits.

Five thermocouples are located at six different elevations to measure the temperature distribution from the HPC to the CPV. In particular one group of thermocouples measures the water temperatures located inside the HPC near the heat transfer plate, one measures the water temperatures located inside the CPV near the heat transfer plate, one measures the wall temperatures at the midpoint of the heat transfer plate between the CPV and the HPC, one measures wall temperatures within the heat transfer plate between the CPV and the HPC nearest to the HPC and one measures wall temperatures within the heat transfer plate between the CPV and the HPC nearest to the CPV.

The fast running SFE method draws up an overview of the temperature changes and a qualitative stress estimation for every monitored component. The fatigue handbook and the knowledge of these parameters determine the pertinence of a fatigue analysis for the component. In that way a more detailed and automatic method, FFE, can be used to calculate the CUFs.

The determination of the time-history of loads and the resulting local stresses are the basis of the method. On a measuring section, close to the fatigue relevant location, FAMOS measures the temperature at the outside surface of the pipe. First of all the interpretation of this temperature has to be explained.

In a homogeneous isotropic solid, where the temperature is a function of time and space T = f (x, y, z, t), the equation of heat conduction is as follows:

d2T 82T 82T pc 8T n

—т +—v + —v ———- = 0

8×2 8y2 8z2 X 8t

К… thermal conductivity of the solid [W/(m-K)] p… density of the solid [kg/m3] c… specific heat of the solid [J/(kg-K)]

This equation can be written in a cylindrical coordinate system:

82T 18T 1 82T 82T pc 8T n

82 + + —882 + ~8———— X~8f ~ ° (2)

8r2 r 8r r2 80z 8z2 X 8t

The stratification effects will not be considered here and the application of the method will be restricted to plug flow events. In this case, the temperature evolution is independent of the circumferential direction T = f (r, z, t). Moreover the measuring section on the pipe is located relatively far away from geometrical discontinuities and T (z — Sz) = T (z + dz) holds true. Following this presumption, the temperature in the assessed section can be written as: T = f (r, t). This involves:

82T 1 cT = pc cT 8r2 r 8r X 8t

This equation handles the thermal evolution inside the thickness of the pipe. The solution of the equation depends on the applied boundary conditions. As the varying load is the medium temperature flowing throughout the component, the heat transfer between the fluid and the inner surface is governed by a Newton’s law of cooling:

= h(T — TM )r =ri (4)

r =ri

h… heat transfer coefficient [W/(m2 K)]

For further explanations of the mathematical background of the method see e. g. [12].

All the difficulty to apply this equation during unsteady fluid temperature states is due to the determination of the heat transfer coefficient h. Indeed, this parameter depends on the velocity, the thermo-hydraulics conditions and the geometry of the surface. The time dependent knowledge of all these parameters with sufficient accuracy is hardly compatible with a fast determination of the real loads. To solve this problem, the inverse philosophy was developed to calculate the stresses in the structure. Indeed, to perform a structural analysis, the knowledge of the temperature distribution throughout the wall is sufficient. The FFE method is based on the time history determination of the inner wall temperature by solving the inverse problem of conduction of heat. The according flowchart is shown in Figure 6.

Newton‘s law

Solving the inverse conduction equation of heat is done by application of potential functions (unit transients). A unit transient is applied at the inner surface of the pipe (boundary condition), the equation of conduction of heat is solved, the resulting time-history of temperature at the outer wall is observed. The resolution of the equation of heat can be done by means of an analytical method or with the help of a finite element program (ANSYS®). In that case a two-dimensional model of the section of the pipe is generated. The benefit of this last choice is the opportunity to integrate the thermal influences of the thermocouple installation at the outer surface of the pipe in the solution (see Figure 7).

The determined temperature response calculated in the thermocouple (outer wall) will be considered as a reference. Its evolution is characteristic by the applied unit transient at the inside surface of the pipe (characterized by a temperature rate of changes and a thermal amplitude ATref). Thus the FAMOS measured outside temperature will be scanned step by step (typically every second). The temperature difference at the outer wall between two time steps is compared with the simulated outer wall temperature (reference). The factor resulting from this comparison is through linearity properties also available at the inner side of the structure. Thus, step by step the inside temperature of the pipe can be restituted. A computation algorithm of this process was developed. The acquired measured data of FAMOS are read into the FFE program. A preparatory work consists of calculating, for the different observed piping sections, the thermal references. These last ones depend on the material, pipe thickness and measurement thermocouple. After this pre-processing work, the computation of the transient inner wall temperatures is completely automated.

Fig. 7. FE calculation of the temperature response at the outside of the pipe

The determined inner wall temperature will be used to calculate the thermal stress at the fatigue relevant locations. An appropriate temperature transfer function can readily be used for correction of the axial dependency of the temperature if the FAMOS section is far away from the stress calculation locations T(z -8z) Ф T(z + 8z). The procedure is shown schematically in Figure 8.

The thermal stress determination is done according to a similar process as previously explained. A two — or three — dimensional finite element model of the monitored component is generated (nozzle, heat exchanger,…). A unit (elementary) transient is used as a reference load of a thermal calculation. Thus, the thermal field in the structure is calculated. Subsequently, the thermal stresses are calculated by a linearly elastic structural analysis.

The resulting thermal stresses are determined for typical fatigue relevant locations. The calculated stress components are the response to a reference load characterized by a temperature rate of change and a thermal amplitude ATref. The exemplary procedure is shown in Figure 9.

Fig. 9. FE calculation of the stress responses at a fatigue relevant location

The inside temperature calculated in the previous step by means of FFE, is scanned step by step. Between two time steps, the temperature difference is interpreted as a unit transient the amplitude of which is compared to the reference unit transient of amplitude ATref. Because of linearity in the thermal stress calculation, the comparison between the measured amplitude and the reference gives a coefficient to be applied to the reference stress matrix in order to obtain the stress contribution resulting from the thermal load at the calculated time. The time-dependent stress components are then obtained by the summation of all these single contributions. The process is also completely automated within the FFE program. The stress matrix references have to be calculated previously in an FE program. The results are then added to the database of FFE: it is the pre-processing work. Subsequently, the calculation at the selected locations can be processed. Within a few minutes, thermal loads and stress components of the entire operating cycles are calculated (see e. g. Figure 10).

If information on the time dependent pressure or piping section forces and moments are available based on operational instrumentation, the resulting mechanical stress components can be calculated equally by means of FFE (scaling of unit loads). Thermal and mechanical
stress components are added and the equivalent stress is calculated. The use of a rain-flow algorithm will classify the stress ranges, a standard conform comparison with the fatigue curves will give the fatigue level of the selected locations.

Finally, if the calculated fatigue usage factor is lower than the allowable limit, the fatigue check will be successfully finished. If not, further analyses according to the detailed code based fatigue check will be performed.

In order to optimize the costs and user flexibility, the FFE program was based on a modular architecture. Thus, only information required by the customer/user is calculated. This architecture also permits an easy upgrade of the program to implement new modules e. g. as a consequence of changes of nuclear standards (new fatigue curves, environmental factor integration,…) or further calculation methods (automated stratification consideration).

Usually, the calculation results of such analysis depend on initial parameters and modelling techniques. In any case, having quite uncertain parameters, the influence of the main initial parameters to the main results may be investigated performing uncertainty and sensitivity analysis. In particular case, the analysis focuses on how the calculation results could change, when the initial parameters describing IRIS-like technology in MESSAGE model are changed. In addition, the most important parameters for the precision of calculation results were also indentified.

The main parameters and their possible values (describing the IRIS-like technology in the model) for different scenarios are presented in Table 3.

The reference values of some parameters presented in this table are taken from (Alzbutas & Maioli, 2008). The possible variations of these parameters are based on calculation assumptions. For instance, in the calculations it was assumed that the EPZ could change from 0 to 30 kilometers. In the model this is represented by the length of additional heat supply pipe in order to connect IRIS-like NPP with cogeneration option to the existing district heating network. In addition, the number of IRIS units in the MESSAGE model is adapting depending on specific conditions in the modelled energy system.

Following the uncertainty analysis the sensitivity measure PCC (see Fig. 9) describes how the initial conditions and model parameters (see Table 3) influence the result. From the sensitivity analysis we can see that the 3rd parameter (discount rate) has the largest (negative) influence on the total system costs (main modelling result). When this parameter increases, the considered model result decreases most significantly. Alternatively, the increase of nuclear fuel price (the increase of 6th parameter) in the considered range has the lowest influence.

In general, a high discount rate gives more weight or importance to present expenditures than to future ones, while a low discount rate reduces these differences and thus favours technologies that have high investment cost but low operation costs (for example NPPs).

I parameter

Fig. 9. Sensitivity measure and determination coefficient (R**2) for total system cost

In this case study, the sensitivity measure, which is a product of the statistical analysis, shows which sources of uncertainty are contributing most to the uncertainty in the predicted energy system performance (see Fig. 9). But it is possible that sensitivity measure, in this case a Partial Correlation Coefficient (PCC), explains too small a fraction of the variability of the model output values, for instance, if coefficient of determination is less than 0.5. However, for analyzed case the coefficient of determination is 0.99. Thus, in this case the sensitivity measure PCC in a very good way express the relation in variability and analyst can easily determine which model parameters should be controlled better in order to decrease unfavourable changes of results. Alternatively, the analyst can determine which parameters could be less precise without substantially affecting results.

2. Conclusions

1. While innovative design solutions are possible in an early design stage to cope with extreme internal events, the need for integrating external events considerations on a probabilistic basis at a relatively early design stage is going to be another challenge for effective and balanced use of PSA as a support of the design phase.

2. Further progress of PSA application and EPZ definition could be achieved via discussion with national regulatory authorities in those IAEA Member States that are considering performance-based and risk-informed licensing approaches for future NPPs.

3. Construction of SMR units is very attractive option (looking from economical point of view) for the future electricity and heat generation. The option with SMR cogeneration mode may cause the lowest total discounted cost among the scenarios analyzed.

4. In the case, when IRIS cogeneration unit should be installed away from existing district heating networks (due to EPZ), the attractiveness of this unit is decreasing gradually with distance, because of investment cost and heat losses in addition district heating pipelines.

5. The sensitivity analysis may be essential as it shows how particular parameter is important to the modelling results and where the accuracy of primary data could be increased (in order to decrease the uncertainty of the results). Alternatively, the analyst can determine which parameters could be less precise without substantially affecting results.

6. In our case, the discount rate has the highest influence on the total system costs, while the increase of nuclear fuel price in the considered range has the lowest influence to the total system costs.

3. Acknowledgment

Due to the international nature of the IRIS project, the cooperation can be treated as a trade mark of the IRIS project. This was even truer for the IRIS PRA. The authors wish to acknowledge the large support and valuable assistance of the IRIS heads M. D. Carelli and

B. Petrovic as well as of other members of the IRIS PRA team, especially D. J. Finnicum and

C. L. Kling. We also want to acknowledge the advises and useful discussions with L. E. Conway and L. Oriani from the IRIS design team. And last but not least we would like to extend thanks to J. Augutis and M. Ricotti for the great personal support provided during the initial stage of PRA related research. The publication of this chapter was funded by Westinghouse Electric Company, LLC.

In flow-induced vibration design of heat exchanger tube bundles, resonant conditions must be suppressed by ensuring separation of natural frequencies of the tubes and exciting frequencies (Shin & Wambsganss,1975). A number of techniques are available for computing natural frequencies of straight, curved, single and multiple span tubes. These tubes may be subjected to varying end conditions. Loose baffles act like "kinfe-edged rings" supports (Timoshenko, 1955) . Tubes are rigidly fastened to the tube sheet and supported at intermediate points along their lengths by baffles or support plates. Some tubes in the centre may be supported by every baffle, whereas tubes that pass thorough baffle window may be supported by every second and third baffle. Table 5 (MacDuff & Feglar, 1957, Kissel, 1977) gives some of the formulas/techniques for estimating the natural frequencies of straight-and curved-or U-tubes.

Branko Kontic

Jozef Stefan Institute Slovenia

1. Introduction

The key topics of this chapter are i) comparative evaluation of various energy options, and ii) radioactive waste disposal. Both are treated from the strategic planning and assessment points of view and are supported by a discussion of multi-objective decision-making. Environmental considerations are foremost. The discussion is focused on the uppermost level of societal energy planning, and attempts to answer strategic questions concerned with the comparative evaluation of various energy options and waste disposal. It is guided by a number of questions as illustrated in Table 1. The Table also indicates in which sub-chapter a certain, more specific discussion can be found.

The author is a natural scientist, experienced in research and preparation of different types of environmental impact and risk assessments. At the present time — January 2012 — after more than 30 years of practice in the field he is astonished by the increasing inefficiency of formal guidance on evaluation of environmental impacts. He wonders why is this so and is especially disappointed when seeing that even the highest administrative level EU institutions, the DG Environment and DG Regional Policy, do not succeed in implementing the guides on performing strategic environmental assessments. For example, the DG Regional Policy and Cohesion provided a guide for the ex-ante evaluation of the environmental impact of regional development programmes in 1999 (EC, 1999) as complementary to the Handbook on Environmental Assessment of Regional Development Plans and EU Structural Fund Programmes (EC, 1998). These were a kind of predecessor of the EU Directive 2001/42/EC (usually referred to as the strategic environmental assessment — SEA Directive). Despite the fact that the guides clearly stress the importance of establishing an interactive relationship between evaluation and planning — the objective of the integration is to improve and strengthen the final quality of the plan or programme under preparation — more than 10 years afterwards Member States fail to follow them and report on a number of difficulties in SEA implementation (EC, 2009). The most important deficiency in the current practice of SEA in certain EU countries is still the approval/permitting context of the use of SEA instead of the planning context and optimisation of plans, and the mixed use (misuse) of project level environmental impact assessment — EIA and SEA. SEA is very often used for the evaluation of specific projects, while EIA is used at higher, i. e. strategic, levels, sometimes even for the evaluation of sustainability of plans and programmes (Kontic & Kontic, 2011). This situation stimulated the author to prepare the present condensed overview

of research and consultancy results on strategic considerations of nuclear power. His aim is that this will contribute to the desired change of implementation of strategic evaluation in the area of energy production and elsewhere.

Comparative information about the environmental impacts of various energy systems can assist in the evaluation of energy options and consequent decision making. Over the last thirty years a number of studies have attempted to quantify such impacts for a wide range of energy sources. These estimations have taken different approaches, from impacts of fuel acquisition through to waste disposal (IAEA, 2000). Recent major studies have been completed and new studies begun in which nuclear power is either supported — justification through e. g. climate change issues or low-carbon society — or criticised — justification through e. g. accidents at Chernobyl and Fukushima, or waste related issues. The results of the studies provide useful insights and help to promote further studies of impacts for many technologies and sites. However, the strategic level of these considerations still remains less well covered and a number of questions are still unanswered. This chapter is aimed as a contribution to filling these gaps.

Related to the radioactive waste issue, the siting of a disposal facility or final repository is a task with unique traits that are clearly associated with changes in the surrounding world. A number of questions can be posed regarding how ongoing and future changes in technology, views, politics and practices in other parts of the world, concerning e. g. energy supply, nuclear power and nuclear waste, may affect national decisions regarding the approach and decisions involved in successful and safe disposal of the waste. National trends in politics, economy and opinion also influence events and views, locally and nationally (SKB, 2011). The decision-making process has to fulfil certain democratic expectations and criteria: openness, transparency, participation. So far, known and applied approaches have not been efficient or effective in solving the primary issue of participatory decision-making in this area, i. e. proper, fair and balanced consideration of specific priorities and interests. Neither weight assignment, as a representative method rooted in (expert) opinion and value judgements, nor methods based on statistics and probability theory (applicable for measurable attributes) have proved successful for this purpose. Maybe ‘approval voting’ (Laukkanen et al., 2002) is the closest to what is widely understood as participative/democratic decision making. It appears, on the other hand, that a continuous engagement process, sound and consistent, scientifically supported and respected by all involved parties, which deals adequately with uncertainties related to long-term predictions/evaluations — as applied in Finland and Sweden — can provide satisfactory results (SKB, 2011). The approach applied in Slovenia for identifying and approving a site for a low and intermediate level radioactive waste disposal facility could also be seen as being successful, and is presented in more detail in Section 3. In summary, it builds on social acceptance of predictive uncertainty based on so-called "local partnership" i. e. the community is actively involved in the siting process and has a right of veto, together with a comprehensive investigation of the perceptions of the types of consequences rather than the likelihood of their occurrence. The underlying basis of the approach is that it is more promising to investigate which consequences of a certain alternative are more likely to be accepted by society than how likely these consequences are to occur. Thus, as many feasible alternatives as possible should be evaluated, so that the parties involved can express their preferences rather than just "yes/no", or "accept/reject" responses. This is clearly in line with the basic philosophy of SEA and strategic considerations of nuclear power.

2. Comparative evaluation of environmental impacts of various energy systems

Contaminated of the wide territories in Ukraine not only with radionucludes 137Cs and 90Sr, and with fission products of uranium and transuranium elements is an essential consequence of the accident at the IV block of Chernobyl Nuclear Power Plant that is classified as a global ecological catastrophe. The biota behaviors and adapt in this areas captured dose from radionuclide with long half-value period decay isotopes. As dose related amount of the isotopes 137Cs and 90Sr during long time after accident were decreased. But only the amount of the radioactive isotope 241Am depend of time is increasing exactly in environmental alienation zone of Chernobyl. Radionuclide 241Am as а-emitter is a daughter product of 241Pu isotope appeared after |3-decay. The activity in environment of the isotope 241Am is increasing with during time owing to |3-decay of the 241Pu isotope. The biota behavior in this areas captured dose from radionuclide with long half-value period decay isotopes. The peculiarity of radionuclides contamination associated with the Chernobyl accident is verified of physical and chemical forms of radioactivity elements through out into the environment [Rashydov N. M. 1999, Rashydov N. M., Konoplyova A. A., Grodzinsky D. M. 2004, Rashydov N. M., Kutsokon N. K. 2005, Rashydov N. M., Grodzinsky D., Berezhna V. 2006]. A part of the radioactivity isotopes is registered in water soluble droplets-liquid state, an other part — as "hot" particles, the interrelation between there forms being unstable and change under the influence of biotic and abiotic environmental factors. As rule in this conditions accumulation of radionuclides in plants which occurs mainly at the expense of their water-soluble and exchangeable forms, reflects rather complicated transitional processes in the soil, the rate and direction of these ones is determined by biological activity of all component of the plant rhizosphere inhabited layer of the soil. After Chernobyl accident already during 25 year a lot of "hot" particles transferred into fine dispersive conditions, which easy movements in outdoors where captured by biota which could characterize by help of transfer coefficient (TC) radionuclide ongoing. The transfer coefficient is ratio specific activities (kBq/kg) of plant to specific activity of soil (kBq/kg) where its grow that characterize go over a radionuclide from soil to vegetative plant on experimental plot. Necessary mentioned that the TC not constant and it differed on depend of parts of plant were determinate. For radionuclide 241 Am observed the value TC a lot of plants and mushrooms several order less than for isotopes 137Cs, 90Sr. Especially for matured seed the value of the TC observed less than for other vegetative parts of the plant [Rashydov N., Berezhna V., Kutsokon N. 2007, Rashydov N. M., Kutsokon N. K. 2008, Rashydov N. M., Berezhna V. V., Grodzinsky D. M. 2009, Rashydov N., Berezhna V. 2010, Rashydov N. M. 2010, Rashydov N. M. 2011]. To study of the TC peculiarity modification is reason elucidation of our field research in alienation Chernobyl zone around object "Shelter".

Contamination of plant in natural experimental fields at the alienation zone of Chernobyl significant added by flying dust with very small size radioactivity particles less than "hot particles" in environment. The results received for plants soybean (content: 137Cs — 3.6 kBq/kg and 90Sr — 11.84 kBq/kg) and flax (content: 137Cs — 0.78 kBq/kg and 90Sr — 3.55 kBq/kg) which grown in Chistogalovka (specific activity of soil is 20.65 kBq/kg and 5.18 kBq/kg for radionuclide 137Cs and 90Sr, accordingly) and Chernobyl confirmed this hypothesis. The value TC for above mentioned seeds specimens collected from plant which grow on Chistogalovka was approximately 22.3 (soybean) and 6.63 (flax) times (for isotope 137Cs) and 13.97 and 4.71 (for isotope 90Sr) times higher by comparison with control variants which grown in Chernobyl where specific activity was 1.41 kBq/kg for radionuclide 137Cs and 0.55 kBq/kg for isotope 90Sr, correspondingly.

The peculiarity distribution in controlled laboratory conditions the radionuclide 241 Am in Arabidopsis thaliana plant on high level first layer leaves, in petiole and in carry out fascicles of the leaves significantly that go into this isotope from root system to top of plant very slow and membrane of cells played as discrimination barrier in this processes as mentioned in our previously investigations [Rashydov N. M., Berezhna V. V., Grodzinsky D. M. 2009].

In laboratory conditions for autoradiography investigation purpose the seedlings Arabidopsis thaliana were aseptically grown in hard agar cultured medium containing 241AmCl3 in concentration with specific activity 50 kBq/kg. After 25 days some leaves and top of stems of plants witch had not direct contact with medium were carefully cut off so that to avoid contact with medium. Selected parts of plants Arabidopsis thaliana settled down on the microscopic glass slides and dried a few days. During this process they were gluing to the slides themselves. The slides with parts of plants were coated with photo emulsion LM-1 in gel (Amersham — Biosciences UK) and exposure during time 20 days at temperature +40 C. After development the samples of slides were observed of the track of a-particles from radionuclide 241 Am with light microscope. A lot of datum confirmed that the coefficient
uptake very small for radionuclide 241Am and this element maldistribution by organs and tissues. We observed that accumulation the radionuclide of 241Am depended of carry out fascicles system of the leaves and localization of the layer leaves not far from length root collar of plant which grow in laboratory conditions. The first layer leaves were taken up high-level amount radionuclide 241 Am. As result the capture dose also may tissues of plant distribute no uniform. It is known that mineral nutrients are transported apoplastically, i. e. in the wall system outside the plasma membrane, or symplastically, i. e. in the cytoplasm from cell to cell deal with through plasmodesmata. The nutrient elements that penetrate into the cytoplast can also be shuttled into the vacuole via various mechanisms depending of biological function in cell life behaviors for mentioned isotope.

For field experiments we use plant white blow (Erophila verna (L.) Bess.) for autoradiography investigation from Chistogalovka and Yaniv contaminated soil sites the distribution radionuclide essential differs in spite of above-mentioned experiment. On the top shoot apex leaves and flower observed a lot of tracks of the particles a — and P — decays [Rashydov N. M., Berezhna V. V. 2010] (Figure 36).

Our experimental data confirms that radioactivity fallout in environment essentially differed important amendment of the TC. Thus extra-root nutrition that included micro — or nano- size "hot" particles had essential role of plant behavior in environment. But for plants that harvested from contaminated sites distribution of the radionuclide 241Am by tissue and organs essentially differed from plants which grown in laboratory conditions.

(a) (b)

Contamination with radionuclide in natural experimental fields significant added tracks elementary particles from flying in air very small dust such as nano — and micro-size with radioactivity similarly "hot" particles in environment by help foliar pathway uptake into top leaves and aboveground apical apex of plants, especially around the object "Shelter".