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

Design basis accidents (DBAs)

Transients without a loss-of-coolant are part of the DBAs. With the exception of containments with a pressure suppression system no major interaction with the containment atmosphere exist. Steam injected into the water pool of the suppression system is condensed in the pool and creates thus some circulation in the pool. There is experimental and operational evidence that the mixing in the pool is good; therefore, no major analytical efforts are undertaken to model this mixing.

Depending on the break size steam is injected into the containment atmosphere during loss of coolant accidents (LOCA) creating, in the short-term, injection driven flows and a pressure increase. Later condensation of steam on steel and concrete structures occurs. In the long term energy is removed through heat exchangers or to the outside of the containment pressure boundary if this boundary is designed as a steel shell. Some small amounts of hydrogen may be produced during a LOCA, due to the fuel cladding/steam reaction; recombiners in the containment will reduce a local accumulation as well as the total amount of hydrogen. During all these processes convection flows exist; usually the velocities are low.

Related computer codes have been developed since about 20 years. They are usually of lumped parameter type and can be assessed as being fully developed, in general. Well-known codes are CONTAIN (SANDIA) and COCOSYS (GRS).


The thermal hydraulic behavior of the CAREM reactor was study using generic numerical codes. Several transients and accidental situations were analyzed.

The dynamic analysis of the plant during operational transients and accidents has been performed mainly with RETRAN-02 code [1].

As an example, Figures 2 and 3 show the power and pressures evolution for a 5% decrease in the reference value during 150 seconds. In this case a classical controller was used.


FIG. 1. Primary system cool ant natural circulation.



time (s)

FIG. 2. Power evolution during a temporary decrease of the power reference value.




time (s)


FIG. 3. Pressures evolution during a temporary decrease of the power reference value.




A High Pressure Natural Convection Loop (CAPCN) 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.

Activities of passive cooling applications and simulation of innovative nuclear power plant design

F. Aglar, A. Tanrikut

Turkish Atomic Energy Authority, Turkey

Abstract. This paper gives a general insight on activities of the Turkish Atomic Energy Authority (TAEA) concerning passive cooling applications and simulation of innovative nuclear power plant design. The condensation mode of heat transfer plays an important role for the passive heat removal application in advanced water-cooled reactor systems. But it is well understood that the presence of noncondesable (NC) gases can greatly inhibit the condensation process due to the build up of NC gas concentration at the liquid/gas interface. The isolation condenser of passive containment cooling system of the simplified boiling water reactors is a typical application area of in-tube condensation in the presence of NC. The test matrix of the experimental investigation undertaken at the METU-CTF test facility (Middle East Technical University, Ankara) covers the range of parameters; Pn (system pressure) : 2-6 bar, Rev (vapor Reynolds number): 45000-94000, and Xi (air mass fraction): 0-52%. This experimental study is supplemented by a theoretical investigation concerning the effect of mixture flow rate on film turbulence and air mass diffusion concepts. Recently, TAEA participated to an international standard problem (OECD ISP-42) which covers a set of simulation of PANDA test facility (Paul Scherrer Institut-Switzerland) for six different phases including different natural circulation modes. The concept of condensation in the presence of air plays an important role for performance of heat exchangers, designed for passive containment cooling, which in turn affect the natural circulation behaviour in PANDA systems.


Nuclear energy is one of the options presently available to cope with energy needs along the forthcoming century. This challenge is requiring a tremendous effort to assure nuclear energy competence in terms of economics and safety with respect to the other potential sources of energy. In the case of water cooled power reactors, new advanced designs have been proposed of either an evolutionary or a passive type, the latter being particularly appealing for using natural forces to carry out safety functions under the most adverse conditions posed by hypothetical accidents. In this regard containment of passive reactors is to be equipped with what has been called Passive Containment Cooling Systems (PCCS).

PCCS’s features depend on specific designs. However, most of them share their reliance on steam condensation to mitigate long term pressure rise in containment. New boundary conditions and device geometries prompted renewed to investigate steam condensation to eventually demonstrate PCCS’s capability to meet their goals. As a result, experimental and analytical programs were launched worldwide, often on the basis of a fruitful international co­operation [1].

Concepts of passive safety systems with no active components have been investigated for new generation light water reactors [2]. The primary objectives of the passive design features are to simplify the design, which assures the minimised demand on operator, and to improve plant safety. To accomplish these features the operating principles of passive safety systems should be well understood by an experimental validation program. Such validation programs are also important for the assessment of advanced computer codes, which are currently used for design and licensing.

In an application, the proposed advanced passive boiling water reactor design, simplified boiling water reactor (SBWR), utilises as a main component of the passive containment cooling systems (PCCS) the isolation condenser (IC). The function of the IC is to provide a passive heat exchanger for the removal of the reactor coolant system sensible heat, and core decay heat to a reservoir of water within the containment. In performing this function, the IC must have the capability to remove sufficient energy from the reactor containment in order to prevent the containment from exceeding design pressure shortly following design basis event and to significantly reduce containment pressure in the longer run [3]. However, it is well established that the presence of noncondensable (NC) gases in vapors can greatly inhibit the condensation process. The mass transfer resistance to condensation results from a build-up of NC gas concentration at the liquid/gas interface leading to a decrease in the corresponding vapor partial pressure and thus the interface temperature at which condensation occurs [3]. As a result, reduction in heat transfer rate is unavoidable with respect to the pure condensation case.

A part of the long-term research and development efforts of the Turkish Atomic Energy Authority (TAEA) is planned to concentrate on passive cooling systems. In this paper, a general insight on activities of the TAEA concerning condensation in the presence of air is given. Moreover, TAEA participated to an international standard problem (OECD/NEA ISP — 42) which covers a set of simulations of PANDA test facility, which is the scaled model of SBWR for different phases of natural circulation modes. The concept of condensation in the presence of air plays an important role for the performance of heat exchangers, designed for passive containment cooling, which in turn affect the natural circulation behaviour in such innovative systems.


Studying the removal of the residual heat of the reactor in a final stage of an accident with loss of coolant is very important for proving that a nuclear power plant is safe. In this stage of the accident with WWER-640 reactor, the tanks of the emergency core cooling system are empty and the emergency pool is filled with water. At this time of emergency cooling of the reactor, a rather complex heat transfer mechanism based on natural circulation of the coolant is realized. Key elements and systems of the reactor installation are involved in this process. The natural circulation circuit is a system of three vessels with a free liquid surface. These vessels are the reactor, the fuel pool, and the emergency pool. Several pipes connect them to each other, in accordance with a plan, by which the residual heat of the reactor is to be removed by using the emergency pool. The steam generated in the core during natural circulation of the coolant through the reactor enters the pressurized containment space. In this way, the steam bubbles through the liquid layer in the pools. The heat from the pressurized volume is removed by cooling the outer surface of the metallic protective shell of the containment. This is done by using a system of passive heat removal from the pressurized

shell (SPHR PS), which also utilizes the principle of natural circulation of the corresponding coolant. The condensate formed on the cooled internal surface of the protective shell is returned to the emergency pool. Thus, the necessary amount of water is maintained in the emergency pool. Therefore, during the stage of pool cooling of the reactor, the thermal- hydraulic processes that take place in the reactor, the containment, and the emergency system, are interrelated. So, to substantiate measures taken in the design for ensuring safety of the nuclear power unit, we will have to comprehensively simulate the processes mentioned above.

In order to carry out the proposed experiments connected with the WWER-640 reactor, the PACTEL was reconstructed (Fig. 2). The main parts of the PACTEL have been kept in the experiments; such as the instrumented pressure vessel, the downcomer, the lower plenum, the core, and upper plenum. The remaining components and systems were excluded.


FIG. 1. Safety systems of the reactor WWER-640 1: Lower plenum, 2: Reactor core, 3: Upper plenum 4: Reactor cavity, 5,6: Fuel storage pools, 7: Steam and non-condensable gases, 8: Emergency pool, 9: Containment shroud, 10: Heat removal system.

Two water tanks installed into the PACTEL, simulating the emergency cooling and fuel storage pools. These vessels are open to atmosphere at the top. The larger diameter of the two water tanks models the emergency cooling pool, the smaller one models the fuel storage pool. Geometrical characteristics are presented in the Table I. Each tank has a hydraulic link to the downcomer of the PACTEL with a single horizontal pipe. Similarly, each tank is linked to the upper plenum of the PACTEL with hot-leg connections. The level balancing line (LBL) interconnects the two water tanks (Fig. 3). All of the added pipelines provided with isolation valves to form a circulation loop for each scenario of the performed test series.


FIG. 2. Modified PACTEL.

Main modifications

The flow diagram of the TOPFLOW facility is given in Fig. 14; a front and a side view is shown in Fig. 15.


FIG: 14. Flow diagram of the TOPFLOW test facility.

The main additional test sections are the steam drum (L = 5m, D=1.5m, V = 8 m3) and two vertical pipes (D = 50 and 200 mm, L = 10 m) to be used for the instrumentation development, evaluation of flow profiles, WWER steam generator behaviour during small break LOCA and the investigation of heat up, local steam production and temperature stratification in large water pools and turbulence. Additionally several junctions are foreseen for the connection of further test sections like a PWR hot leg model [4]. The pressure vessel has been removed but is still available if needed.

The data acquisition systems will be replaced and extended, e. g. the number of the measurements as well as the acquisition rate are increased. Additionally the balance of plant system will be replaced and is now computer controlled.


FIG. 15. Front and side view of the TOPFLOW test facility.


The complex set of physical phenomena that occur in a gravity environment when a geometrically distinct heat sink and heat source are connected by a fluid flow path can be identified as natural circulation (NC). No external sources of mechanical energy for the fluid motion are involved when NC is established. In a number of publications, including textbooks, the term natural convection is used as a synonym of NC. Within the present context, natural convection is used to identify the phenomena that occur when a heat source is put in contact with a fluid. Therefore, natural convection characterizes a heat transfer regime that constitutes a subset of NC phenomena.

This report provides the presented papers and summarizes the discussions at an IAEA Technical Committee Meeting (TCM) on Natural Circulation Data and Methods for Innovative Nuclear Power Plant Design. While the planned scope of the TCM involved all types of reactor designs (light water reactors, heavy water reactors, gas-cooled reactors and liquid metal-cooled reactors), the meeting participants and papers addressed only light water reactors (LWRs) and heavy water reactors (HWRs). Furthermore, the papers and discussion addressed both evolutionary and innovative water cooled reactors, as defined by the IAEA[1].

NC principles are of fundamental interest to the designers of nuclear power plant systems and components. Making reference to the existing water cooled reactors, the consideration of NC is brought to the design of the layout of the primary circuit. The core is located at a lower elevation with respect to the steam generators and the feed-water inlet location, in the cases of pressurized and boiling water reactors, respectively. In all of the adopted geometrical configurations, NC allows the removal of the decay heat produced by the core, should the forced circulation driven by centrifugal pumps become unavailable. Furthermore, NC is the working mode for the secondary side of most steam generators in existing pressurized heavy and light water reactors. It is essential as well for the core cooling in the unlikely event of loss of primary coolant.

Reactors based on natural circulation during normal operation (e. g. the Dodewaard Reactor in the Netherlands and the VK-50 in Russia) operated for an extended period of time. Most boiling water reactors can operate in the natural circulation mode for power levels below about 40 per cent of full power. Some newly developed designs are based on natural circulation core cooling for normal operation and on the use of the natural convection heat transfer for some safety systems. Reliance on natural circulation can result in simplified systems, reduced costs and — most importantly — a very high safety level.

The accomplishment of the objectives of achieving a high safety level and reducing the cost through the reliance on NC mechanisms, requires a thorough understanding of those mechanisms. Natural circulation systems are usually characterized by smaller driving forces with respect to the systems that use an external source of energy for the fluid motion. For instance, pressure drops caused by vertical bends and siphons in a given piping system, or heat losses to environment are a secondary design consideration when a pump is installed and drives the flow. On the contrary, a significant influence upon the overall system performance may be expected due to the same pressure drops and thermal power release to the environment when natural circulation produces the coolant flow. Therefore, the level of knowledge for the thermal-hydraulic phenomena for the specific geometric conditions and governing heat transfer conditions should be deeper when NC is involved. In addition, the lower driving forces for natural circulation systems might lead to quite large equipment for which the role of 3D phenomena is essentially increased.

Within nuclear technology the renewed interest in NC is a consequence of the above, in combination with the potential for cost savings from increased use of NC mechanisms in plant designs. Relevant experiments directed to the characterization of NC have been carried out in the past because of the importance of the related mechanisms for the safety of existing reactors. Similarly, thermal-hydraulic system codes have been qualified through the comparison of predicted results and experimental data. The quality of recorded experimental data and the precision level of the available system codes, or the expected uncertainty in these predictions, are generally evaluated as satisfactory for the needs of the current reactors. However, the exigencies posed by the more extensive use of the NC in the design of evolutionary and innovative water cooled reactors require a re-evaluation of the experimental data and of the code capabilities considering the new phenomena and conditions involved.

Recent activities completed under the IAEA umbrella, e. g. Refs [1-8], and by other institutions, such as the U. S. Nuclear Regulatory Commission (U. S. NRC) [9], the OECD/NEA/CSNI (Organization for Economic Cooperation and Development/Nuclear Energy Agency/Committee on the Safety of Nuclear Installations), Refs [10-13], and the EC (European Commission), Refs [14-16], show the importance of the subject and constitute the basis for future activities in the area of NC. Potential future international activities could be directed toward:

(a) identification of still unresolved issues from evolutionary and innovative water reactor designs, and

(b) enhancing the quality levels of the available computational tools and experimental databases in relation to design needs.

This report provides an overview of the current state of the art of natural circulation data and methods, and discusses potential benefits of an integrated future effort directed toward the achievement of the two aforementioned objectives. The main attention in this report is paid to the design basis accident phenomena; severe accident issues are considered briefly in their relation to the protection of the containment as the last safety barrier.



Although a large amount of thermal hydraulic data exist from the beginning of nuclear reactor development there is still some need for additional experimental data. This is on one hand due to the focus on an area not well studied in the past — in this case natural circulation in nuclear reactor application — and on the other hand due to additional and more detailed requirements from modern CFD code developments. Because of the high costs for good experimental data it should be evident that new experiments should only be performed if existing data do not meet the requirements with respect to quality and degree of detail. If new data are required the very best instrumentation should be used, and, if necessary, new instrumentation should be developed. International cooperation is encouraged.


Natural circulation phenomena have been investigatedto a large extent. There are a large number of experiments and facilities. As an example the experiments performed with APEX (US), SPES (Italy) and ROSA/LSTF facilities have given data of the same design (AP-600) with different scaling. Instead of studying the operational performance of the whole concept, the core make-up tank (CMT) behaviour was investigated with PACTEL facility. PANTHERS, PANDA and NOKO represent facilities which simulate different systems of the same design.

As an example of coordination between different facilities with different scaling and code work is the NACUSP research project. One of the main research items that have been identified for natural-circulation flow is the stability of two-phase natural-circulation flow. The former studies have been usually financed by national sources. Within the Euratom 5th framework programme, the NACUSP project is dealing with this research item. The goal of NACUSP is to improve the economics of operating and future plants through improved operational flexibility, enhanced availability, and increased confidence level on the safety margins regarding the stability issues in boiling water reactors (BWRs). The experimental part of this project focuses on natural-circulation loops being of specific value for the ESBWR.



Experiments will be performed in 4 experimental facilities, which complement each other, ranging from small scale to large scale, and from low-pressure, low-power operating conditions, to nominal and high-power operating conditions. Figure 1 gives an overview of the facilities used.

The DESIRE (at Delft University of Technology) and CLOTAIRE (at Commissariat l’Energie Atomique) facilities will be used to study the natural circulation and stability characteristics at nominal pressures. This is important in view of the study of coupled neutronic/thermohydraulic stability, which is an issue for large BWRs. A good understanding and validation effort of the thermal hydraulic stability is essential as a basis for other work performed within NACUSP on reactor stability and in preparation for future work on regional reactor stability. Because of its flexibility the DESIRE facility will mainly be used to perform parametric studies [1], whereas the CLOTAIRE facility will be used to study the effects of upscaling. The CLOTAIRE facility was originally built to study PWR steam generators [2].

At low pressures the CIRCUS (at Delft University of Technology) and the PANDA (at Paul Scherrer Institute) facilities will be used to study the natural circulation and stability characteristics. It is known that at low-pressure, low-power operating conditions (e. g. start-up) natural circulation flow is susceptible to gravity-driven instabilities. Because of the high sensitivity of the flow to perturbations, these conditions are very challenging for computer codes to predict. CIRCUS will be used for parametric studies [1], whereas specific tests will be performed in the large-scale PANDA test facility. PANDA has already been used to study passive decay heat removal in the ESBWR [3].

The experimental work described above will be performed in close collaboration with computational work using a selection of computer codes. Table I gives an overview of the NACUSP partners and the computer codes used for analyzing the experimental data of the facilities.



Computer codes for facility data

NRG Netherlands


CEA France


DUT Netherlands


ETH Switzerland

Forsmark Sweden

FZR Germany



PSI Switzerland


UPV Spain


This section gives some examples of the NC problems being considered in new reactor concepts and relevant experimental works both performed earlier and to be performed in future.


When the water inventory of the RPV decreases from about 400 Mg to 160 Mg, the drywell around the RPV is flooded from the core flooding pool when two valves in the flooding line are opened either actively or passively (see Fig. 7). The main aim of flooding the RPV from outside is to achieve long-term retention of even a totally molten core inside an intact RPV. Experiments are in preparation which will show that cooling of the RPV is possible with thermal flux densities far below film boiling despite the numerous nozzles at the RPV bottom.

The water flows downward, driven by gravitational force alone, and fills the space around the RPV within about 30 minutes. At first the water is subcooled, but it heats up due to the hot RPV until evaporation begins. The generated steam flows through openings in the insulation of the RPV to the containment cooling condensers. The condensate again flows downward into the core flooding pool and back to the space round the RPV.

In this case, natural circulation is very effective due to both steam generation at a low eleva­tion and condensation at a high elevation in the same circuit. Because no water is lost in this natural circulation circuit, the passive transport of decay heat takes place as long as the CCCs are filled with water on the secondary side.




Steam outlet



flooding device

Core melt (metal fraction)

Core melt (oxyd fraction)


Gaps between
CRD housings
and insulation


FIG. 7. Conceptual drawing of flooding of RPV from outside during a core melt accident.




. Wind channel test for passive containment cooling system

The wind channel test will research air flow resistance characteristic and effects of air channel shape and air inlet location on natural circulation flow.

The influence of environment wind and surrounding buildings on the natural convection flow had been especially considered. Moreover, the velocity field at the lower turning of air baffle

and the surface wind pressure distribution of containment were needed for design. The main purposes of the test were: a) To verify the feasibility of passive containment cooling system design. Main emphasis was put on the influence of the environmental wind; b) To supply a database for preliminary design and design improvement, especially the experimental data of velocity field in annuli and pressure drop of each section. The velocity field in the low baffle end zone was calculated and tested.

The model tests (1/10 in scale) with different flow deflectors were done to study the way of improvement. Both test and calculated results indicated that there were an enclosed vortex in the stagnant bottom and an obvious separate bubble formed at the rear of the baffle. The flow deflectors could reduce the separate bubble. Another model test was run in 0.2—0.5 m/s water velocity. A larger vortex at the upper stream of flying object protect shielding and a smaller bubble at the down stream of it were proved by both test and calculation.

A pressure distribution test in the surface of containment was done in a low velocity wind tunnel with various wind speeds, air entrances, yaw and pitch angles. The test shows that the pressure is positive in the area of -35° < 0 < 35°. The influence of wind on natural convection of containment was also done with a 1/50 integral model. The test results revealed that natural convection flow rate was enhanced in general by outside wind and horizontal wind (a = 0) had the better effect than a < 0° or a > 0°. The position of chimney might influence the air flowing around the containment. The distance between containment and chimney should be larger than 4 times of chimney diameter. However, smoke wind tunnel test showed no exhausted hot air was re-circulated under any d outside wind conditions.


FIG. 9. Diagram of containment structure. FIG. 10. Flow field near flying protect



1.2. Summary

1) The studies of passive ERHR system, CMT injection system and passive containment cooling system prove that the design of all these passive systems are feasible and reliable in principle and can meet basically the required safety functions.

2) Some undesired thermal hydraulic phenomena were found and identified in these studies. For example, the flow vortex in the containment air duct, and “water hammer” of ERHR test may have bad impacts on its safety functions and should be avoided in the next step tests and AC600/1000 design.

3) All data obtained have been already used for design improvement and next R&D program planning.


The calculations were performed with APROS version 4.06

The first pressure peak occurred too early in the APROS calculation and one additional pressure peak was observed after the pressure had already started to drop in the experiment. The code simulated the general trend of the downcomer flow behaviour well until the PSIS stagnated. During the PSIS stagnation period, between 10000 and 12000 seconds, the downcomer flow decreased and stopped completely in the experiment. This did not happen in the APROS simulation. The CMT started to empty too early in the APROS calculations. This happened since too much water flowed into the pressuriser and the steam begun to flow to the cold legs too early. The recirculation flow through the PSIS decreased as the hot water filled the CMT in the APROS calculation, but the flow did not stop completely. The calculated recirculation flow at the end of the recirculation phase was about about 75% of the initial value in the APROS calculation.

Passive Safety Injection Test GDE-43 (1 mm cold leg break)


FIG. 3. Water temperature in the CMT. Experiment GDE-43 vs. APROS calculations.

The effects of numerical diffusion were clear in the APROS calculation of CMT water temperature (Fig. 3). The calculated temperature rise was smoother than in experiment. Too much water accumulated in the pressuriser in the all calculations during the primary flow stagnation periods, indicating problems in modelling pressuriser heat losses. At the end of the transient, too much water accumulated in the pressuriser in the APROS simulation. The accumulation of water in the pressuriser partly explains the fact that the PSIS flow stagnation did not occur in the APROS calculation. The core water level dropped too fast in the APROS simulation in the early phase of the transient and the codes did not predict the core water level lowering during the PSIS flow stagnation