Category Archives: Natural circulation data and methods for advanced water cooled nuclear power plant designs

Modelling of multi-dimensional thermalhydraulics

The increasing need for more accuracy of the codes and more detailed descriptions of new cooling concepts leads to a tendency to use at least for local detailed analyses more and more 3D and time-dependent computational fluid dynamic (CFD) codes. There are several motivations for using CFD codes in the nuclear field: Some problems can only be treated in 2D or 3D, like the flow-regime determination, thermal stratification, phase separation, and phase distributions in large pools. CFD codes are required if dynamical 2D or 3D phenomena have to be investigated, like the oscillations of velocities, pressures, temperatures, or phase distributions. The relevance of such investigations may come e. g. from ensuring safe heat transfer from all sub-assemblies under natural convection, from the thermal striping induced thermal fatigue, or flow instabilities due to geysering or sloshing, and from the phase and boron distribution dynamics coupled to neutronics.

Finally CFD codes have to be used for scaling up detailed results from model experiments to full-scale reactor conditions. They are also more and more used even for designing experiments and their instrumentation arrangement. They can also reliably be used to study in a more qualitative manner the relevance of certain phenomena in flow and heat transfer problems, so far as the governing physics is included in the equations or in the numerical modelling. Thus, they are also used to determine the dominant physical phenomena of nuclear flow problems, so that finally a reliable efficient numerical model configuration (input) can be formulated for a system code.

Most CFD codes for two-phase or multi-phase multi-component flows are based on the multi­field approach, assuming that all fluids and phases are defined everywhere in the flow field; this means, the fluids are interpenetrating. The spatial distributions of the different fluids or phases are determined by their relative local volume fractions. As a consequence of this method, one has not only a tremendous increase of the number of equations to be solved, but there are also no explicit phase boundaries existing. This kind of modelling is the basis for the typical working tools available. As such flows are physically and topologically complex, such codes give a valuable support to understand in more detail the macroscopic physical phenomena occurring in multi-phase flows.

The physical models in such codes, e. g. to model interfacial phenomena, are rather simple. Therefore, a quantitative use of such CFD codes for two-phase flows is currently limited to mainly homogeneous flows. Modelling improvements are related to bubble or particle diameters which have to be specified. New developments are going to include multi-group concepts of bubble diameter classes or to develop interfacial area concentration equation models which allow up to now for simple flow regimes to calculate the dynamics of the surface of the interface, which is an important parameter in determining momentum, heat, and mass transfer between the phases. This could also help to improve the modelling of bubble coalescence and fragmentation. For other flow regimes than bubbly or droplet flows, the modelling of interfacial phenomena needs serious improvements.

In going from 1D discretisation to 2D or 3D, one has to use turbulence models when turbulent flows have to be investigated. Some of the two — or multi-phase codes have turbulence models for two-phase flows, but our current knowledge and experimental data base on turbulence is very weak: we have only a few turbulence data for the liquid phase in bubbly flow. This basis is not sufficient to adapt, calibrate, and validate these models. Therefore, none of the existing models can reproduce all those data. For other flow regimes, data as well as models are missing; problematic flow regimes will be bubbly, churn, slug, and droplet flows. The turbulence in non-adiabatic flows is characterized by a rapid change of bubble diameters which will influence the turbulence characteristics. Another related problem occurs in the missing wall functions and in the boundary conditions for the interface phenomena at free surfaces. Thus, the turbulence modelling in two-phase flows can only be accepted as a first step. The models can be used strictly only in an interpolative manner, this means in a parameter range, in which extensive validation was done before. R&D is necessary to improve the models and to extend them for all flow regimes, so that more reliable prediction capabilities can be achieved.

Despite these model deficiencies such codes were rather successful, because many multi­phase problems are governed by large-scale interfacial phenomena. Therefore, most development in the past was based on developing adequate numerical methods to solve the set of equations of multi-field models. The challenge here comes from the weak compressibility of the fluids or phases engaged and from the strong density variations from liquids to gases across interfaces. Therefore, the development of more efficient and more robust solution schemes and solvers is always a target for further improvements.

For such cases, in which separated phases have to be considered, special numerical tools were developed to keep the interface sharp, like interface capturing, interface tracking, or some approximate numerical methods like surface sharpening. In case of dispersed flows special highly accurate numerical schemes would be required to avoid or at least minimize numerical diffusion, but most codes use standard lower order schemes; thus one still has to live with more or less strong numerical diffusion.

In the past, most codes for nuclear applications with multi-phase flows were developed mainly in the large national research centres or in the industry, like AFDM (LANL/FZK), ATHLET (GRS), CATHARE (CEA), COBRA-TF (Pacific NW Lab), ESTET-ASTRID (EDF), IVA (Siemens), MATTINA (FZK), MC-3D (CEA), RELAP5-3D (INEEL, DOE), ERHRAC (NPIC), TRAC (SNL, USNRC)… They have the advantage that the required nuclear specific modelling is included, and the models and numerics were selected according to the special requirements of nuclear applications. Now, powerful, and comfortable commercial CFD codes are becoming available, which are increasingly suitable for nuclear applications, like CFX-4,

COMET, FIDAP, FLOW-3D, FLUENT, PHOENICS, and others. There are major differences between the principal modelling capabilities available in the commercial codes. None of those has all the modelling required for certain flow phenomena in innovative reactor systems; thus it has to be decided from case to case which code should be used. This brings some concerns regarding the certification effort when one is progressing towards a licensing procedure of a reactor.

With increasing requirements on accuracy and details of codes, also requirements regarding the accuracy and details to be provided by experiments are growing. Whereas for design purposes and for system code improvement and validation mainly realistic large scale experiments are of interest, for the development and validation of CFD codes for two-phase flows one needs primarily single effect experiments with very detailed instrumentation, preferably providing not only local data, but also field data, like PIV and tomography. Topics should be the interfacial phenomena and turbulence near bubbles with rapid diameter change, near free surfaces, and near walls with boiling, the bubble formation, coalescence and fragmentation, data on the interfacial area, and the turbulence in the continuous phases; similar investigations are also needed for all other flow regimes, and the flow regime transitions. Nevertheless, some large scale more complex experiments with a detailed instrumentation are also required to validate the correct interrelation between the different models engaged in the CFD codes.

COMPARISONS TO STABILITY AND CHF DATA

To compare to the prediction of the first and second roots (instability lines) at low and high quality there are many data in the literature. We must know, or be able to estimate or calculate the loss coefficients and use only data where the channel pressure drop is constant, relevant to the reactor situation.

Definitions for the experimental onset of instability are not universal, and were often reported as the onset of density wave oscillations, not static instability, obtained by usually raising the power until some unstable or oscillating flow was observed. Data review yielded over 300 data points, extracted from 30 years of the literature, covering a range of pressures from 0.1 to 19 MPa,(i. e. nearly 1-190 bars), for tubes, rod bundles and parallel channels. In Figure 3, the data and theory are shown, with the density wave data cluster at the lower subcoolings and powers.

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Подпись: Ns

FIG. 2. Typical stability maps for different flow directions.

 

World Data and Theory Limits

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FIG. 3. Comparison of data and stability theory showing stability and CHF boundaries.

 

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In non-dimensional Np versus Ns space, the data clearly map out a central unstable (uninhabited) region. The best fit through the lower limit of the saturated data is given by a line of slope (Np /Ns) = 3), which is the asymptotic limit of Equation (8).

From this form of plot we see that stability is possible when operating in the high Np and lower Ns region, and the first and second unstable lines are given by,

Np /Ns = N* (13)

For the second characteristic line at high quality, we have N* is O(3), from both theory and experiment: for the first characteristic line in the low quality subcooled region, stability is possible when N* is 0(1).

Now, in a natural circulation loop with n channels, we define the region of instability as given by the intersection of the unstable region from the instability with the natural circulation flow. We find that we can maintain stable natural-circulation flow below the critical subcooling number. However, for larger subcooling, the natural-circulation line is in an unstable region and therefore, flow will be unstable. Basically, the downcomer level should be maintained high enough to have imposed pressure higher than the high-void-region extreme of the pressure drop-flow curve.

The intersection of the natural circulation flow with the unstable region for larger downcomer heights (L* ~ 0.3) is at a value of Np/Ns of about 2, which implies the maximum stable power level for the system is close to the theoretical and experimental stability limit. For lower downcomer heights, L* ~ 0.5, the intersection is closer to the minimum value of the unstable region. Thus, we also have the analytical result which determines the instability boundary and the onset of CHF in a natural circulation system.

THE NATURAL CIRCULATION FLOW MAP

The similitude of the geometry and of the operational parameters of the PWR simulators (or Integral Test Facilities, ITF) allowed a direct comparison between results of NC experiments. The database that was gathered from ten experiments performed in the six ITF listed in Table I has been used to establish a Natural Circulation Flow Map, Refs. [2] and [3]. In all the considered ITF, the NC regimes discussed in the previous chapter are experienced. The linear power of fuel rod simulators, the fraction of nominal core power and the primary system pressure mainly differentiate the considered experiments. In relation to pressure, PKL experiments have been performed at a pressure value roughly one half the value adopted in the other facilities. The identified differences and the ranges of design and of other operational parameters (e. g. pipe diameter, system volume, number of steam generators, heat losses to the environment, not explicitly discussed here), produce an envelope for any expected NC situation in a typical PWR. The envelope for NC system behaviour is assumed applicable when decay heat removal conditions are established.

Measured values of core inlet flowrate (G, Kg/s), core power (P, MW), primary system fluid mass inventory (RM, Kg) and net volume of the primary system (V = const., m3), have been used for setting up the NCFM. The diagram G/P versus RM/V has been preferred for the NCFM over other possible choices including non-dimensional quantities. The experimental database from ITF (six ITF, ten experiments) and the envelope of curves are given in Figs 2 and 3, respectively. The envelope in Fig. 3 is assumed to constitute the NCFM of PWR at decay core power.

Finnish facilities for studies on innovative nuclear power plant designs

J. Vihavainen, J. Banati,

Lappeenranta University of Technology, Finland

H. Purhonen,

VTT Energy, Finland

Abstract. Several series of experiments has been carried out at Lappeenranta University of Technology (LTKK) in co-operation with VTT Energy in Finland to investigate systems to be used in advanced and innovative reactors. The integral and separate effects test facilities and the flexibility to use and modify existing instruments enables the variety of the possible applications. The main goal of the experimental work is to provide test data for validation of computer codes used for nuclear safety analysis. VTT Energy started the studies on Innovative Nuclear Power Plant designs by investigating the Core Make-up Tank (CMT) behaviour on the Parallel Channel Test Loop (PACTEL) facility. The main objective was to provide new and independent information about Passive Safety Injection System (PSIS) performance. The latest project on the PACTEL test facility in European Commission 4th Framework Programme “Assessment of Passive Safety Injection Systems of Advanced Light Water Reactors” involved 15 experiments in three series. The purpose of these experiments was to increase information about condensation and heat transfer processes in the CMT, thermal stratification of water in the CMT, and natural circulation flow through the PSIS lines. The PACTEL facility was modified for investigations of the WWER-640 emergency cooling system for reducing primary pressure in Loss-of-Coolant Accident by equalising the primary-system and containment pressures and removing the residual heat from the core by natural circulation to the emergency cooling pool. The results of the experiments have been used for computer code validation in low pressure and low power conditions. Another innovative experimental project is a hydraulic scram system for SWR-1000 power plant concept by Siemens-KWU. The system was studied in the series of separate effect experiments. The system uses pressurised water and steam as a driving force of the hydraulic scram system. Electric heaters located below the water level generate and maintain the steam volume and a layer of saturated water in the tank. After the scram signal the energy of the steam volume is used to move the control rods into the core. In the experiments orifices simulated the response of the control rods.

1. INTRODUCTION

PACTEL (Parallel Channel Test Loop) [1] is a full height, medium-scale integral test facility (volumetrically scaled 1:305) designed to simulate the thermal-hydraulic phenomena characteristic of the Finnish Loviisa PWR. VTT Energy together with the Lappeenranta University of Technology run the facility. PACTEL has three primary coolant loops with pressurizer, primary coolant pumps and horizontal steam generators, high-pressure emergency core cooling system (ECCS), and low pressure ECCS with two accumulators. The peak operating pressures in the primary and secondary sides are 8 MPa and 4.6 MPa, respectively. The reactor vessel is simulated with a U-tube construction consisting of separate downcomer and core sections. The core comprises of 144 full length, electrically heated fuel rod simulators with a heated length of 2.42 meters. The maximum total core output is 1 MW, or 22% of scaled full power. The three coolant loops with double capacity steam generators model the six loops of the reference power plant. Each steam generator has 118 U-tubes with an average length of 2.8 m.

The passive safety injection experiments in Lappeenranta started already in 1992 with a series of five experiments. These experiments simulated hot leg SBLOCA’s. The PSIS included a CMT and two PBL’s. The second series included four cold leg SBLOCA experiments in 1993. These experiments used similar PSIS as in the first series. In 1996, a new project of passive safety injection experiments with a CMT and one PBL started. The new experiments are a part of the European Commission 4th Framework Programme Nuclear Fission Safety program. Within this new project, three series of five experiments were completed, project continued until September 1998.

The PACTEL facility was modified for investigations of the WWER-640 emergency cooling system for reducing primary pressure in Loss-of-Coolant Accident by equalising the primary — system and containment pressures and removing the residual heat from the core by natural circulation to the emergency cooling pool. The results of the experiments are not yet finally analysed but they are being used for computer code validation in low pressure and low power conditions.

Another innovative experimental project has been a hydraulic scram system for SWR-1000 power plant concept by Siemens-KWU. The system was studied in the series of seven separate effect experiments during December 1999. The project was a part of VTT’s Advanced Light Water Reactor (ALWR) technology programme. It was funded by Siemens, Teollisuuden Voima Oy (TVO), TEKES and VTT. The system uses pressurised water and steam as a driving force of the hydraulic scram system.

Emergency Condenser 2. Bundle

It could be calculated that the tube wall is responsible for more than 50 percent of the total heat transfer resistance. Therefore, a second bundle was used with the geometrical data.

total length: about 10 m

inner diameter: 44.3 mm

wall thickness: 2.0 mm.

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FIG. 4. The design of the Emergency Condenser (EC).

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FIG. 5. Power transfer to the water pool with the EC with four tubes.

In addition, the bundle was oriented in a vertical position, under an angle of 40.9° and in a horizontal position.

The experimental results and a comparison with ATHLET calculations are shown in Fig. 6. From sensitivity studies it has been evaluated that the thermalhydraulic codes like ATHLET or CATHARE are calculating well the integral power transferred, but deviate substantially in the in-tube heat transfer by condensation processes. Therefore, special tests with a single tube only and an improved instrumentation were performed. In addition, special tests and measurements for 3D-temperature fields in the water pool have been performed, see the isothermes in Fig. 7 for one of the tests.

SYSTEM CODES DESCRIPTION

It has been shown before that a number of plant situations can only be simulated with computer codes.

The computer code development started together with the development of nuclear reactor systems. With increasing computer capabilities, a sound experimental basis and the set-up of validation matrices the computer code development for system codes has now practically reached its final stage. As examples, short descriptions for ATHLET, RELAP 5 and CATHARE are given below.

There is growing interest in CFD computer codes that are capable of modelling 2-D and 3-D fields, and which have the potential to simulate complex geometries. However, more developments and validation is necessary. The FLUENT code will be described below as an example of a CFD code.

EMERGENCY CONDENSERS

The emergency condenser system consists of four separate heat-exchanger subsystems, each having a nominal heat transfer capacity of 55 MW at about 71 bar reactor pressure. The emergency condensers are connected to the RPV without isolating elements, and thus actually form part of the RPV. Each emergency condenser consists of a steam line leading from the RPV nozzle to a heat exchanger tube bundle. This tube bundle is located inside the core flooding pool at a low elevational position. The outlet on the heat exchanger primary side is the reflooding line with integrated anti-circulation loop.

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FIG. 2. Conceptual arrangement of emergency condenser (at normal and reduced RPV water levels).

The working principle of the emergency condenser design is illustrated in Figure 2. Given the normal water level inside the RPV there prevails a stratified condition inside the emergency condenser. The upper part of the steam line is filled with steam while the lower part is filled with water. The water remains cold (except for a small layer below the RPV water level), as the anti-circulation loop prevents hot water from the RPV from entering the reflooding line from below. No convection occurs and thus thermal losses are negligible as long as the water level in the RPV remains normal. The water level in the steam line of the emergency conden­ser is several meters lower because the density of the water inside the RPV is lower than that of the water in the emergency condenser.

This stratified condition changes to natural circulation if the water level inside the RPV drops by more than 0.7 m. Consequentially, when the water level in the emergency condenser then drops by more than 0.5 m, steam enters the heat exchanger bundle. The steam then condenses inside the heat exchanger tubes and the resultant condensate flows via the reflooding line back into the RPV. If the water level inside the RPV is lower than the inlet nozzle of the reflooding line, the maximum driving pressure differential will be reached at a pressure of about 0.5 bar. This pressure differential is used to overcome the flow resistances in the steam line, heat exchanger tube bundle and reflooding line. The emergency condenser continues to function as long as the water level inside the RPV remains lower than 0.7 m below the normal RPV water level. This has been experimentally tested and verified under the direction of Prof.

E. F. Hicken on the emergency condenser test facility at Germany’s Julich Research Center.

On the secondary side, natural circulation also occurs once the emergency condenser begins to work. At low heat transfer rates there is single-phase flow, while at higher rates two-phase flow occurs due to water evaporation. Normally, the water inventory of the flooding pool below the heat exchanger bundle could not be used as a heat sink due to stratification. To overcome this problem, the heat exchanger is enclosed in a chimney. Water enters the chimney at the bottom of the pool and exits at the top several meters above the heat exchanger bundle.

Experiment research and calculation method of natural circulation flow for AC600/1000

S. Zhang

Nuclear Power Institute of China,

China

Abstract. Passive safety concept is extensively used in the design for next generation advanced PWR nuclear power plant. The decay heat of reactor core can be removed through natural circulation flow of coolant following an accident. This not only increases reliability of engineered safety systems and reduces core melt frequency, but also simplifies systems and increases plant economy. Nuclear Power Institute of China (NPIC) has performed preliminary experiment research and relative theoretical analysis for passive characteristics of advanced PWR nuclear power plant AC600/1000. Three tests about natural circulation flow have finished as the following: residual heat removal through SG secondary side, core makeup tank behavior and wind flow of containment. The above mentioned three mechanism tests have verified natural circulation flow concept of AC600/1000. By the end of this year NPIC will finish other two single tests in order to research the following key technology of the passive safety systems: The natural circulation characteristics of tandem system of SG secondary side loop and air flow loop for emergency residual heat removal system (ERHRS) after station blackout accident; The water flow behavior in primary coolant system contained by core makeup tank, pressurizer, accumulator and reactor pressure vessel after small break accident; Computer code development and verification. Meanwhile, NPIC will cooperate with Karlsruhe Technology Center of Germany to research natural circulation characteristics of air in the annular channel between the steel shell and the concrete shell of containment. NPIC plans to build two large integral test facilities. One of which is used to research natural circulation flow and residual heat removal through primary loop, secondary loop and air flow loop from reactor core to ultimate sink —atmosphere after station blackout accident. It is also used to research the passive safety injection features for emergency core cooling system. The second integral test facility will be used to research the comprehensive heat removal behavior of passive containment cooling system. The paper will describe the utilization of natural circulation concept in the passive safety systems, experiment research performed by NPIC and computer codes as well for AC600/1000.

The Second Series of Passive Safety Injection Experiments (GDE-11 through GDE-14)

The second experiment series included four SBLOCA tests with break in one cold leg of PACTEL. Two break sizes (2 and 4 mm or 0.5 and 2%, respectively) were used. The tests also included studies of primary system depressurization and reproducibility of the phenomena (one test was repeated twice with the same initial and boundary conditions). The PACTEL operators depressurized the primary system by opening a pressurizer relief valve. The core power in the tests was 80 kW corresponding to about 1.8% of the scaled thermal power of the reference reactor. In all tests, only the CMT provided ECC water to the primary loop.

The primary pressure used in the tests was again lower than the nominal operation pressure of PACTEL. The maximum operation pressure of the passive accumulator limited the maximum experiment pressure to 3.8 MPa. The experiment loop geometry in the second series have been described in reference [3]. Munther [4] and Munther et al. [7, 8, 9] have published the results of these experiments. Since one loop of PACTEL was equipped with a different steam generator model the tests run with only two loops in operation. The main results of the second test series can be summarised as the following:

• only a limited period of single phase natural circulation through the cold leg pressure balancing line to the CMT was observed,

• the CMT level started to drop immediately after the opening of the break,

• the saturated water layer formed to the CMT remained thin,

• the ECC water injection was stopped totally during injection, due to rapid vapour condensation in the CMT,

• two types of mixing in the CMT during condensation was observed: complete mixing or only the mixing of the uppermost water layer in the tank.

There was still water in the CMT when the tests were terminated after about 2000 seconds from the opening of the break. During the rapid condensation period water level in the core simulator dropped close to the top of the core. However, also in these test no core heat-up occurred.

In 1996, a new project for the investigation of passive safety injection systems of ALWR’s begun. The project, entitled "Investigation of Passive Safety Injection Systems of Advanced Light Water Reactors", was a part of the INNO cluster of the European Commission Nuclear Fission Safety (NFS-2) Programme. The project received funding from the European Commission. The project had four partners: VTT Energy and Lappeenranta University of Technology from Finland, AEA Technology from the UK and the University of Pisa from Italy. VTT and LTKK were responsible for the experiments in PACTEL and for the APROS simulations of selected experiments. AEA Technology and University of Pisa were responsible for simulation of selected experiments with RELAP5 and CATHARE codes. The general objectives of the new project were:

• to provide new and independent information about passive safety injection system performance,

• to contribute to a public data base for the users and developers of thermal-hydraulic computer codes on the phenomenological behaviour of PSIS’s in LOCA conditions, and

• to identify the accuracy, uncertainties and limitations of thermal-hydraulic computer codes in the modelling of passive safety injection system behaviour.

The next sections summarise the main results of the three series (called here the third, fourth and fifth Passive Safety Injection experiment series) of the EC funded programme.

The isolation condenser

Fullscale tests with an Isolation Condenser (IC) as proposed for the SBWR from GE have been performed in the PANTHERS facility. Because the results are proprietary special tests have been performed in the PANDA facility, see Appendix B.

The design of an IC is schematically shown in Fig. 7; vertical tubes are connected to an upper (inlet-) header and a lower header.

Tests with pure steam as well as tests with a mixture of steam and helium (up to about 14 per cent) were performed.

In Fig. 8 test results with pure steam and calculated results with ATHLET are shown.