Today there is a consensus that the current system thermal-hydraulic computer codes are not sufficiently validated for some conditions and for some phenomena relevant to natural circulation (low pressure, low driving heads, effect of non-condensables, boron transport at low velocities, etc. ). It is important to have assurance that the natural circulation systems are effective during all sequences in which they are required to function, and to define the needs for computer codes capable of taking into account all the important phenomena. Extensive work is needed to develop such codes and high quality experimental data are necessary to validate them. Therefore, separate effect and integral tests may be required both for additional validation of existing and/or development of new computer codes. However the targets for the computer codes accuracy should be clearly defined before asking for or designing new experiments to be sure that the financial resources required for the experiments are spent effectively.

In many cases, sufficient operating experience for the natural circulation systems/components under real plant conditions does not exist. Therefore, additional research and development works including large-scale tests for the direct substantiation of the operability of a new system may be needed individually for each advanced water cooled reactor concept. In particular, even the evolutionary plant designs may require confirmatory testing of some components and systems prior to commercial deployment.

Many experiments have been performed up to now with natural circulation systems/components used in new reactor concepts, and many of them have displayed a number of phenomena that are not adequately predicted by the existing computer codes. For example, it was found that gas stratification in the containment can significantly affect the efficiency of building condenser when the cooled H2-layer at the top displaces the steam. In advanced BWRs (e. g. the SWR-1000), vent pipes are therefore used connecting a position above the building condenser with the wet well. If heat transfer in the building condenser deteriorates because of steam shortage, the pressure in the containment increases and forces self-controlled venting of H2/steam mixture to the wet well [15].

The functional reliability of a natural circulation system depends essentially upon the way the natural physical phenomena operate in a particular system and the long-term effect of the environment on the system performance. It requires the identification and quantification of the uncertainties in the interaction between the phenomena, the immediate surroundings and the natural circulation system itself. In this respect, collection of existing and generation of new experimental data would provide information on influences on the functional reliability of the natural circulation systems.

Keeping in mind the above considerations, extensive analytical and experimental works are still needed as applied to new passive systems/components. The contents of these works and the relevant capabilities are discussed in Sections 3 and 4 of this report. The sharing of the relevant results obtained earlier and coordination of the further works on the international level would increase the value of their results and optimize the required manpower and financial resources.

(a) Increasing needs for accuracy and detail of codes result in corresponding needs regarding the details to be provided by experiments. For the development and validation of CFD codes one needs primarily single effect experiments with very detailed instrumentation. Instrumentation should be developed to provide improved local and field data for CFD codes. For design purposes and for system code improvement and validation, relatively realistic large-scale experiments are of interest. Thus some large scale more complex experiments with detailed instrumentation are likely to be needed to validate the correct interrelation between the different models used in the CFD codes;

(b) In some cases, uncertainties exist because of a lack of data for friction loss modelling and condensation (more general: phase interaction) particularly at low pressure. Perhaps dynamic flow charts can be useful here;

(c) Further experiments are likely to be needed for code validation related to the effects of non-condensable gases;

(d) As natural circulation systems used in different advanced designs are widely different in operating conditions and geometry, work on the scaling of natural circulation phenomena demands attention. Effective scaling laws will enable inter-comparison of data generated by various facilities.

The study of NC is of primary interest in the nuclear technology. In the case of PWR, primary circuit layout is designed to optimize the NC performance. Following accidents originated by recirculation pumps trip or even small break Loss of Coolant Accidents, NC may constitute the main mechanism to transfer energy from the core to the steam generators, therefore keeping the NPP in a safe condition. In addition, the ‘quasi-steady’ thermalhydraulic configuration of the reactor loops makes easier the assessment of system code capabilities used for simulating the evolution of generic transient. Owing to the above reasons, all Integral Test Facilities (ITF) built so far that simulate PWR have been used to characterize the NC. A wide experimental database has been gathered and is available.

In the present context, only data measured in ITF designed following the criteria ‘time­preserving’, ‘power-to-volume-scaling’ and ‘elevati on-scaling-factor-equal-to-unity’ are taken into account. The main features of these facilities are summarized in Table I. Additional details can be found in Refs. [8] and [9]. The actual Kv, last row of Table I, gives the ratio between the volume of the reference reactor for the largest ITF (i. e. Lstf) and the volume of the considered facility. An idea of the relative dimensions of the loops and, therefore, of the electrical core power can be derived. NC experiments have been performed in all these facilities with core power close to the decay power value typical of NPP, as documented in Ref. [1].

The NC experiments of interest, around ten, have been conducted with primary loop in single­phase and two-phase conditions:

— at constant pressure of the primary system, close to the saturation pressure of the hot leg in nominal conditions;

— with core power in the range 1%-5% of nominal value as already mentioned;

— with the SGs at nominal conditions of level and pressure;

— having available feedwater flowrate and temperature suitable for removing core power;

— by stepwise draining of primary coolant achieving a quasi steady-state at the end of each draining step.

The NC flow patterns or regimes a) to d) defined in Section 1. have been identified. Based on the results of computer codes calculations and of experiments performed in the PWR simulators (see Table I), the NC regimes are characterized in Fig. 1, taken from Ref. [9]. The mass flowrate at core inlet is given as a function of the primary system mass inventory. Other than the flow regimes, the transition zones and the occurrence of dryout situation at mass inventory roughly below the 40% of the nominal value, can be noted. The dryout occurs owing to the sharp decrease in the heat transfer coefficient in the core when void fraction and mass velocities reach a lower boundary. The wideness of the transition zones comes from uncertainties of the database, generally originated by lack of quality assurance, as well as from differences in some boundary and initial conditions. More details related to each flow regime are given below.

TABLE I. RELEVANT HARDWARE CHARACTERISTICS OF THE PWR SIMULATORS CONSIDERED FOR NATURAL CIRCULATION

Several experiments on single-phase natural circulation loops are reported in the literature. The loops studied can be categorised as Uniform Diameter Loops (UDLs) and Nonuniform Diameter Loops (NDLs).

5.1.2.1. Uniform diameter loops (UDLs)

The UDLs experimentally studied include both closed-loop and open-loop thermosyphons. Considering the shape of the loop, the closed loops can be further classified into rectangular, toroidal and figure-of-eight loops. Holman-Boggs (1960), Huang-Zelaya (1988), Misale et al. (1991), Bernier-Baliga (1992), Vijayan et al. (1992), Ho et al. (1997) and Nishihara (1997) obtained experimental natural circulation data in UDLs of rectangular shape. Uniform diameter open-loops were investigated by Bau-Torrance (1981, 1981a) and Haware et al. (1983).

Fig. 8 shows a comparison of the data with the theoretical correlations for laminar and turbulent flow for uniform diameter loops. The experimental data reported by Misale et al. (1991), Bernier-Baliga (1992) and Ho et al. (1997) are well predicted by the theoretical correlation. Most of these data points are from laminar flow region where the total local loss coefficient is negligible compared to Lt/D due to the large value of the friction factor. For the
range 2 x 107<GrmD/Lt<3 x 108 significant deviation is observed between the data and the

о

correlation. Beyond GrmD/Lt of 3 x 10 , the agreement is found to be better.

For all the steady state data reported in Fig. 8, it is assumed that Sleff=1. But for all practical configurations of loops, local losses are present so that Sleff>1. Hence, the observed deviations with the theoretical correlation may be attributed partly to the unaccounted local pressure losses in these loops. To study its effect, the local pressure loss coefficients due to elbows, bends, Tees, orifices, etc. given in Streeter and Wylie (1983) were used. The results are shown in Fig. 9. The turbulent flow data is now closer to the theoretical correlation indicating the significance of the local pressure losses at high Reynolds numbers where the friction factor is very low. The laminar region data is practically unaffected. The data for the transition region 2 x 10 <Grm/NG<3 x 10 is now closer to the theoretical correlation, but the deviation is still significant. This is the region where the correlation is not applicable as the loop is neither fully laminar nor fully turbulent. Instead the loop is partly in the laminar region and partly in either transition or turbulent regions.

FIG. 8. Steady state natural circulation flow in FIG. 9. Steady state natural circulation flow UDLs without considering local losses. in UDLs with local losses.

DESIRE is a simulated fuel bundle of a natural-circulation BWR with freon-12 as scaling fluid. Figure 1 shows an overview of the primary loop of the current facility, which is a scaled model of the Dodewaard natural circulation reactor[6]. The fuel bundle consists of 35 fuel rods in a 6 x 6 array. The fuel rods are 958 mm long with an heated length of 880 mm. The diameter is 6.35 mm. The fuel rods have either a chopped cosine or a flat uniform axial profile. Six independent power supplies can be connected arbitrarily with individual fuel rods, facilitating a wide range of radial power distributions. The nominal power is 22.3 kW, but the maximum power is about 50 kW. The pressure ranges from 8 to 13 bar (nominal pressure is 11.6 bar). In this manner, using freon-12 as a coolant, a full 1.8 m BWR bundle is simulated operating at 75 bar and 1116 kW nominal conditions. The inlet friction of the fuel bundle can be varied by means of a valve. The riser section can be varied in length (1.1 m-1.9 m) by means of a telescopic riser section. At the top of the riser a free surface exists at which the vapour is separated from the liquid which returns through the downcomer. Part of the vapour is also drawn into the downcomer, the so-called carry-under. The temperature distribution in the loop is measured with chromel-alumel thermocouples. The absolute pressure is measured at different positions in the loop, and the pressure difference is measured over the upper part of the downcomer tube and part of the steam dome, which gives an indication of the collapsed liquid level above the riser exit. The total recirculation flow as well as the steam flow and the feedwater flow is measured with vortex flow meters. Gamma transmission is used to determine the average void fraction at a given height in the core region. This void fraction is also being used to implement an artificial feedback on the power of the fuel rods. The fuel rod thermal time constant as well as the void reactivity coefficient can be varied artificially. Gamma transmission can also be used to reconstruct the void-fraction on a subchannel basis, using tomographic techniques. In order to study the stability characteristics of the system, noise-analysis techniques are being used.

DESIRE has been extensively used to study the natural circulation and stability characteristics of the Dodewaard natural circulation reactor[7,8]. Recently a riser exit restriction has been implemented to be able to destabilize the system in order to study limit cycles and non-linear effects with the facility[9]. With this riser exit restriction the effect of steam separator friction can be studied. In the near future the intention is to use DESIRE in the framework of the European NACUSP project. A large set of thermohydraulic experiments without artificial feedback will be performed covering a wide operational range of the facility (power, feedwater temperature, pressure, riser liquid level, power distribution) for a large range of geometrical settings (inlet friction, outlet friction, riser length). For this purpose the instrumentation of the loop is currently being updated and extended. A dedicated set of experiments will be performed to study the capability of 3D-codes to calculate the void — distribution in the fuel bundle. The focus of these experiments is at low void fractions. Finally, part of the experiments will be devoted to optimize the artificial nuclear feedback currently being used in the facility. This optimization is needed to be able to extend the facility to multiple “parallel bundles” with which coupled neutronic/thermohydraulic regional oscillations can be simulated.

2. CIRCUS

The CIRCUS-facility is a full-height scaled steam/water loop of the Dodewaard reactor. Figure 2 shows a schematic view of the current facility. The reactor core is simulated by 4 electrically heated fuel rods in coolant channels and by 4 separate bypass channels. The channels are made of glass in order to be able to visualize the flow. This is one of the important features of the CIRCUS facility, which enables the researcher to combine the measurements with observations of the flow. The friction of each individual coolant channel or bypass channel can be varied independently. On the top of the core section a long adiabatic glass tube is used to simulate the riser section. The two-phase mixture is condensed and cooled by means of a heat exchanger. The length and the secondary flow rate of the heat exchanger can be varied. This controls in combination with the heater power in the buffer

vessel the subcooling of the liquid at the core inlet. The system pressure is regulated with a steam vessel, representing the steam dome in a BWR. A pressure vessel is used to pressurize the system. When the level in the steam dome drops below a certain level, the pressure vessel can be disconnected. The pressure vessel can also be used to perform alternative measurements with pressure feedback of the pressure vessel itself instead of pressure feedback of the steam volume in the steam dome. The position of the steam dome will be changed to a place before the heat exchanger to avoid problems with subcooled liquid entering the steam dome. Table 1 shows the main characteristics of the facility.

The temperature distribution in the loop is measured with chromel-alumel thermocouples and two Pt-100 temperature sensors for reference measurements. Absolute pressure is measured at the top of the riser and at the inlet of the core. The liquid level in the steam dome is measured with a differential pressure sensor. The differential pressure over the friction settings of the individual channels is a measure for the flow distribution over the coolant channels and bypass channels. The total flow in the loop is measured at 2 different positions with electromagnetic flow meters. The void fraction at a given height can be measured with gamma transmission techniques. At a fixed height at the top of the riser the radial void distribution is measured with a wire-mesh sensor, which measures the conduction of the two- phase mixture on a two-dimensional grid. Furthermore, laser doppler anemometry is used to study the local liquid velocity in the core or in the riser.

A limited set of experiments has been performed in which large flashing-induced oscillations have been observed[10]. As well as DESIRE, CIRCUS will also be used in the near future in the framework of the NACUSP project. A large set of thermohydraulic experiments will be performed covering a wide operational range of the facility (power, inlet subcooling, pressure, steam dome level, power distribution) for a range of geometrical settings (inlet friction, friction distribution). A dedicated set of experiments will be performed in order to study the flashing effect in the riser in detail. The gamma transmission technique and the wire-mesh sensor are of course important tools for this study. A third extensive set of experiments will be performed with two parallel riser sections instead of one common riser as in the current configuration.

TABLE. I. MAIN CHARACTERISTICS OF THE CIRCUS FACILITY

3. CONCLUSIONS

An overview of the facilities at the Delft University of Technology to study the natural circulation and stability characteristics of natural-circulation cooled BWRs has been given. The results of these studies are not only useful for these types of reactors but also for forced- circulation BWRs, because of the study on type-II instabilities. The results are also useful for LWR-reactors in general, because of the experimental data that will be generated at low power and low pressure. The experiments in the near future will be performed in the framework of the European NACUSP project in which the natural circulation characteristics and stability performance of natural-circulation cooled BWRs are studied.

The reactor system is enclosed in a leak-tight containment to protect the public from release of radioactive products; this implies the prevention of leakages from the containment atmosphere into the environment.

Containments of commercial reactors have a volume from several thousand to a maximum of 80000 m3; some small containments are equipped with a pressure-suppression — system (ice boxes or a water pool). Containments are somewhat compartmentalized.

D. F. Delmastro

Centro Atomico Bariloche,

San Carlos de Bariloche, Argentina

Abstract. CAREM is an innovative reactor with an integrated self-pressurized primary system, developed by Argentina. The primary system coolant circulation is of natural circulation type and several passive safety systems are included. The thermal-hydraulic behavior of CAREM reactor was study using generic numerical codes. Several transients and accidental situations were analyzed. A High Pressure Natural Convection Loop was constructed and operated to produce data in order to verify the thermal hydraulic tools used to design CAREM reactor, mainly its dynamical response. This is accomplished by the validation of the calculation procedures and codes for the rig working in states that are very close to the operating states of CAREM reactor. Several dynamical experiments were performed and new ones are planned. The data obtained is being used to test our numerical procedures and codes. In this paper an overview of the thermal-hydraulic aspects of CAREM reactor is presented. The analytical dynamical studies and experimental facility, studies and results are briefly presented.

1. INTRODUCTION

The CAREM nuclear power plant design is based on a light water integrated reactor. The whole high-energy primary system, core, steam generators, primary coolant and steam dome, is contained inside a single pressure vessel.

The flow rate in the reactor primary systems is achieved by natural circulation. Figure 1 shows a diagram of the natural circulation of the coolant in the primary system. Water enters the core from the lower plenum. After been heated the coolant exits the core and flows up through the riser to the upper dome. In the upper part, water leaves the riser through lateral windows to the external region. Then it flows down through modular steam generators, decreasing it enthalpy. Finally, the coolant exits the steam generators and flows down through the down-comer to the lower plenum, closing the circuit. The driving forces obtained by the differences in the density along the circuit are balanced by the friction and form losses, producing the adequate flow rate in the core in order to have the sufficient thermal margin to critical phenomena. Reactor coolant natural convection is produced by the location of the steam generators above the core. Coolant acts also as moderator.

Self-pressurisation of the primary system in the steam dome is the result of the liquid-vapour equilibrium. The large volume of the integral pressuriser also contributes to the damping of eventual pressure perturbations. Due to self-pressurisation, bulk temperature at core outlet corresponds to saturation temperature at primary pressure. Heaters and sprinkles typical of conventional PWR’s are thus eliminated.

The analysis documented in the previous section has been extended. Loop seal piping and pumps available in the PWR-1 nodalisation have been modified and deleted, respectively. The loop seal, ‘U-shaped’ cold leg piping connecting SG outlet with the MCP, has been substituted by a ‘L-shaped’ piping connecting the SG outlet and the horizontal part of the cold leg. New calculations have been performed with results documented in Table VII and in Figs 9 and 10.

Table VII has the same structure as Table VI. However, the columns reporting primary system pressure, upper plenum fluid temperature and upper plenum void fraction in Table VI have been substituted by columns giving the steam line flowrate, the maximum rod surface temperature and the maximum void fraction in the core, respectively. Thermalhydraulic conditions in the core (void fraction and maximum cladding temperature) and at the outlet of the SG (dome temperature and steam line flowrate) can be seen in Figs 9 and 10, respectively.

Starting from the situation depicted by the data in Table VI, a number of strategies can be pursued to design a natural circulation PWR, taking the goal of full core power and allowing for two-phase flow in the core. For instance, system geometry (number of SG U-Tubes, mutual position between RPV and SG, core hydraulic diameter) can be substantially modified, steam generator pressure can be lowered, feedwater temperature can be lowered, primary system pressure can be increased. Strategies requiring (almost) no system geometry changes and minimizing the necessary variation in the values of operational parameters have been followed:

a) The occurrence of dryout, (stable condition with limited temperature excursion) is allowed.

b) The feedwater temperature and the operating pressure of the SG are decreased, thus allowing a loss of thermal efficiency of the plant.

TABLE VII. REMOVABLE POWER BY NATURAL CIRCULATION IN PWR-1 WITHOUT PUMPS AND LOOP SEALS.

The database supplied by the two calculations identified as ND03 and NI03 in Table VII, constitutes the result of the activity. System configurations considered as input for calculations ND03 and NI03 comply with the strategies at items a) and b) above. In the Table VII, calculations KK01 and KN08 are taken from Table VI in order to set reference values for the considered parameters.

The comparison between results of calculations KN08 and NC00 shows the advantage derived from adopting the ‘L-shaped’ cold leg at the outlet of the SG instead of the ‘U — shaped’ cold leg. Core flowrate in NC conditions at 60% core nominal power increases from about 1400 to about 1800 Kg/s showing that pumps at rest conditions introduce a noticeable pressure drop in the loop. Keeping the ‘L-shaped’ configuration and all operational parameters of run NC00 (or KN08), it has been shown that 85% core power can be removed by NC without the occurrence of dryout. This value can be compared with the 70% limit applicable to the ‘U-shaped’ cold leg configuration, (run KN10 in Table VI).

The results from calculations ND03 and NI03 (Table VII and Figs 9 and 10) bring to the following remarks:

• The removal capability of 100% core power by NC has been demonstrated starting from a steady state condition for the system fixed at 3% core power with primary loop in subcooled conditions and pressurizer and SG at nominal pressure.

• Film boiling cooling establishes on the surface of fuel rods in a region of the core (calculation ND03). This conditions is characterized by a stable temperature jump (DTsat, difference between local clad temperature and fluid saturation temperature) of about 180 K. Maximum values of void fraction that are calculated for the core are typical for the nominal operation of Boiling Water Reactors.

• Excursion from the nucleate boiling is achieved at two of the ten axial levels of the core region (upper half) that are characterized by a power peak factor close to 1.5. A proper design of core power distribution, i. e. consistent with the moderation features of the new system, could reduce the spatial extension of the film boiling region.

• The decrease of steam temperature at the SG outlet of about 20 K, consequence of the pressure decrease, prevents dryout occurrence on the fuel rod clad surface, calculation NI03. In this case, no attempt has been made to optimize the combination between feedwater temperature and SG pressure keeping the goal of maintaining the core in nucleate boiling.

• The mass inventory for the primary loop stabilizes at about 80% and 93% of the initial value in the two operational configurations calculated by code runs ND03 and NI03, respectively. This includes the fluid in the pressuriser. Stable liquid level in the pressuriser, i. e. buoyant fluid in a dynamic condition, establishes as the result of the procedure considered for power rise. The level value is such to allow the operation of heaters, though they are not called into operation in the present framework.

• The G/P and the RM/V values that characterize the working points in the NCFM for code runs ND03 and NI03 lie in the area of the square symbols of Fig. 7, though the present values are not reported in that figure.

2. CONCLUSIONS

The database gathered in natural circulation tests performed in PWR simulators and the results of system code calculation related to the same experiments constituted the starting point for the present investigation. The experimental database has been utilized to set up a reference natural circulation flow map. This allowed the judgement of the performance, in natural circulation conditions, of current PWR systems not necessarily of the same type as those used for deriving the map.

It was found that the PWR equipped with Once-Through steam generators have a poor natural circulation performance when primary mass inventory is decreased. Otherwise, reasonable natural circulation performances of Russian designed reactors WWER were characterized. This is mainly true for the WWER-1000. Passive systems in the AP-600 innovative reactor are effective in keeping the primary system under single-phase natural circulation notwithstanding removal of coolant mass.

Power removal by natural circulation strongly depends upon primary and secondary system boundary condition. Two thermohydraulic thresholds have been considered and the neutronic — thermalhydraulic coupling has been neglected. The void formation in the core that prevents a PWR from being ‘pressurised’ and the technological limit constituted by the critical heat flux are the reference thresholds. When primary and secondary system boundary conditions are kept close to their nominal values, it was found that:

• single phase natural circulation is effective for removing up to around 20% of core nominal power,

• two phase natural circulation, with the primary system in a boiling condition, is effective for removing up to about 70% of core nominal power avoiding the occurrence of the dryout.

A deeper investigation showed that 100% core power can be removed in two-phase natural circulation, provided the steam generator pressure and the feedwater temperature are lowered to values of the order of 3 MPa and 100°C, respectively. An increase of primary pressure also brings to the increase of the power removal capability by natural circulation.

The substitution of the current ‘U-shaped’ loop seal with a ‘L-shaped’ pipe and the elimination of the pump component improved the natural circulation performance of the considered system. Two reference sets of natural circulation conditions suitable for additional investigation have been established at 100% reactor core power:

a) film boiling allowed within a limited extension core region;

b) reduced pressure in steam generator.

In the former case, roughly one-tenth of the fuel rod clad surface was calculated with a temperature 180 K larger than the saturation value. Steam conditions at the outlet of the steam generator are the same as in current design PWR. In the latter case, all fuel rod clads are predicted in the nucleate boiling heat transfer regime, though core average void fraction is close to 40%. Better core cooling conditions are obtained at the expenses of lower temperature of the steam and then of lower thermal efficiency for the plant.

Continuation of the study, if interest is shown from the technical community derived from the economic benefit of the proposed solutions, requires the investigation of the following aspects:

• role of the pressurizer;

• design of suitable neutron kinetic parameters for the core region and analysis of the thermalhydraulic-neutronic feedback;

• design of suitable start-up procedure;

• accident analysis.

The last series of the EC funded project included additional five SBLOCA experiments. VTT Energy and LTKK ran the experiments in August and September 1997. Based on the experiences from the previous series, the new experiments were performed in the third and final series of the EC funded programme.

In this series the CMT position was moved to a 1 m higher elevation than in previous experiments. The purpose was to increase the driving head of the passive safety system. The result was about 10% higher flow rate during the recirculation phase of the system. Also, a very small break size was used to investigate if the driving force could disappear due to filling of the CMT with warm water. As a result, the experiment showed this phenomenon.

A Saxony collaboration consisting of the Institut fur Sicherheitsforschung of the Forschungszentrum Rossendorf (FZR) e. V., the University of Dresden (TUD) and the Hochschule fur Technik, Wirtschaft und Sozialwesen Zittau-Gorlitz (HTWS) are going to take over the NOKO facility, which will be reconstructed as TOPFLOW (Transient Two Phase Flow Test Facility) at FZR. TOPFLOW will mainly be used for the investigation of generic and applied steady state and transient two-phase flow phenomena in power and process industries. Main fields of activities are the investigation of:

— transient flow regimes in horizontal, vertical and inclined tubes,

— critical mass flows and oscillations during depressurization,

— natural convection in parallel channels and feed pipes,

— condensation phenomena in horizontal tubes,

— dynamic behaviour of interphase area in bubble columns,

— non-equilibrium effects.

The philosophy of the facility is, that working groups throughout Europe shall be invited to come to Rossendorf with their ideas and to perform their experiments here making use of the wide range of parameters (power, water and steam mass flow, pressure range, measuring instrumentation) offered by TOPFLOW.