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Nuclear Engineering and Design 241 (2011) 4504–4514

Contents lists available at ScienceDirect

Nuclear Engineering and Design

journa l homepage: www.e lsev ier .com/ locate /nucengdes

mprovement in understanding of natural circulation phenomena in waterooled nuclear power plants

ong-Ho Choia,∗, John Clevelanda, Nusret Aksanb,1

International Atomic Energy Agency, Wagramer Strasse 5, P.O. Box 100, 1400 Vienna, AustriaHaldenstr. 35, 5415 Nussbaumen, Switzerland

r t i c l e i n f o

rticle history:ccepted 10 March 2011

a b s t r a c t

The IAEA has organized a coordinated research project (CRP) on “Natural Circulation Phenomena, Mod-elling, and Reliability of Passive Systems That Utilize Natural Circulation.” Specific objectives of CRP wereto (i) establish the status of knowledge: reactor start-up and operation, passive system initiation andoperation, flow stability, 3-D effects, and scaling laws, (ii) investigate phenomena influencing reliabil-ity of passive natural circulation systems, (iii) review experimental databases for the phenomena, (iv)examine the ability of computer codes to predict natural circulation and related phenomena, and (v)apply methodologies for examining the reliability of passive systems. Sixteen institutes from 13 IAEAMember States have participated in this CRP. Twenty reference advanced water cooled reactor designsincluding evolutionary and innovative designs were selected to examine the use of natural circulation andpassive systems in their designs. Twelve phenomena influencing natural circulation were identified andcharacterized: (1) behaviour in large pools of liquid, (2) effect of non-condensable gases on condensation

heat transfer, (3) condensation on the containment structures, (4) behaviour of containment emergencysystems, (5) thermo-fluid dynamics and pressure drops in various geometrical configurations, (6) nat-ural circulation in closed loop, (7) steam liquid interaction, (8) gravity driven cooling and accumulatorbehaviour, (9) liquid temperature stratification, (10) behaviour of emergency heat exchangers and iso-lation condensers, (11) stratification and mixing of boron, and (12) core make-up tank behaviour. Thispaper summarizes the achievements within the CRP for the first five phenomena (1–5).

. Introduction

Since the mid-1980s it has been recognized that the applica-ion of passive safety systems (i.e. those whose operation takesdvantage of natural forces such as convection and gravity), canontribute to simplification and potentially improve economics ofew nuclear power plant designs. The IAEA Conference on “Theafety of Nuclear Power: Strategy for the Future,” which was con-ened in 1991 (IAEA, 1991), noted that for new plants “the use ofassive safety features is a desirable method of achieving simplifi-ation and increasing the reliability of the performance of essentialafety functions, and should be used wherever appropriate”. Someew designs also utilize natural circulation as a means to remove

ore power during normal operation. The use of passive systems canliminate the costs associated with the installation, maintenance,nd operation of active systems that require multiple pumps with

∗ Corresponding author.E-mail addresses: J.H.Choi@iaea.org (J.-H. Choi), J.Cleveland@iaea.org

J. Cleveland), nusr.aksan@gmail.com (N. Aksan).1 Formerly Paul Scherrer Institute, Switzerland.

029-5493/$ – see front matter © 2011 Elsevier B.V. All rights reserved.oi:10.1016/j.nucengdes.2011.03.053

© 2011 Elsevier B.V. All rights reserved.

independent and redundant electric power supplies. However, con-sidering the weak driving forces of passive systems based on naturalcirculation, careful design and analysis methods must be employedto assure that the systems perform their intended functions.

To support the development of advanced water cooled reactordesigns with passive systems, investigations of natural circulationare conducted in several IAEA Member States with advanced reac-tor development programmes. To foster international collaborationon the enabling technology of passive systems that utilize naturalcirculation, the IAEA conducted a CRP on “Natural Circulation Phe-nomena, Modelling and Reliability of Passive Systems that UtilizeNatural Circulation” from 2004 through 2008.

2. Identification of phenomena influencing naturalcirculation

Thermal-hydraulic phenomena and related parameter rangesthat characterize the performance of passive systems do not differ,

in general, from phenomena that characterize the performance ofsystems equipped with active components. This is specifically truefor transient conditions occurring during safety relevant scenarios.For example friction pressure drops or heat transfer coefficients are

J.-H. Choi et al. / Nuclear Engineering an

Nomenclature

A flow area, m2

b constant in friction factor correlation, f = p/Reb

ˇT single-phase thermal expansion coefficient, kg/Jˇtp two-phase thermal expansion coefficient, kg/JCp specific heat, J/kg KD hydraulic diameter, mf Darcy–Weisbach friction factorg gravitational acceleration, m/s2

H loop height, mK local pressure drop coefficientL length, mp constant in friction factor correlation, f = p/Reb

Q total heat input rate, W� dynamic viscosity, Ns/m2

afgl

esascSpcppe

3

indN

3

hnEEo1

finttfimtnrT

� density, kg/m3

ffected by local velocity and void fraction and not by the drivingorce that establishes those conditions, e.g. gravity head or centrifu-al pump. The same can be repeated for more complex phenomenaike two phase critical flow or counter-current flow limiting.

A large number of thermal-hydraulic phenomena that arexpected to occur in passive systems during accident are clas-ified in the OECD/NEA documents (Aksan et al., 1994; Aksannd D’Auria, 1996). The OECD/NEA list of phenomena for passiveystems was up-graded and modified in IAEA CRP on “Natural Cir-ulation Phenomena, Modelling and Reliability of Passive Safetyystems that Utilize Natural Circulation,” considering the recentlyroposed passive systems by the industry. The identification andharacterization of additional phenomena for passive systems isresented in Table 1, which includes phenomena identification andhenomena characterization based upon the individual phenom-na considering the key layout of systems.

. Characterization of phenomena

As shown in Table 1, twelve phenomena are identified as defin-ng thermal hydraulic characteristics for passive safety systems andatural circulations in advanced water cooled NPPs. This sectionescribes major improvements in understanding for phenomenaos. 1–5.

.1. Behaviour in large pools of liquid

Large pools of water at near atmospheric pressure provide aeat sink for heat removal from the reactor or the containment byatural circulation, as well as a source of water for core cooling.xamples include the pressure suppression pool (wet-well) of theSBWR, the in-containment refuelling water storage tank (IRWST)f the AP1000, the pool of the emergency condenser of the SWR-000 and the gravity driven water pool of the AHWR.

Large pools may have a very wide spectrum of geometric con-gurations. Heat transfer in a limited zone in terms of volume doesot imply homogeneous or nearly homogeneous temperature inhe pool. Three-dimensional convection flows develop affectinghe heat transfer process, which results in a temperature strati-cation. Steam generated by heat transfer or following injectionay be released from the pool into the containment and influences

he increase of the containment pressure. Compared to a homoge-eous temperature distribution, the fluid at the top of the pool mayeach the saturation temperature while the bulk fluid is sub-cooled.he three-dimensional nature of the temperature stratification may

d Design 241 (2011) 4504–4514 4505

require a Computational Fluids Dynamics (CFD) approach depend-ing on the needs.

Key experiments, which were designed to investigate the phe-nomena in large pools namely the temperature stratificationprocesses and their influence on the heat transfer capability, are:

- PANDA (Paul Scherrer Institute (PSI), Switzerland) (Aksan andLuebbesmeyer, 2003):

Several types of computer codes such as containment codes(lumped parameter and 3D), primary system codes and CFD codeswere invited to perform assessment with the ISP-42 (PANDA)experimental data. The ISP-42 PANDA test consisted of six phases:• Phase A: passive containment cooling system start-up.• Phase B: gravity-driven cooling system discharge.• Phase C: long-term passive decay heat removal.• Phase D: overload at pure-steam conditions.• Phase E: release of hidden air.• Phase F: release of light gas in reactor pressure vessel.

There were no results submitted using CFD codes for any ofthe six phases of the ISP-42, though last two phases were moreoriented to the assessment of the CFD codes.

- PUMA (Purdue University Multi-Dimensional Integral TestAssembly, USA) (Cheng et al., 2006):

Several tests were performed to investigate the thermal strat-ification, direct contact condensation, and the pool circulationby plume and jet in the suppression pool. It was found that thedegree of thermal stratification in the suppression pool is stronglyaffected by the noncondensable gas injection flow rate. Furtherinfluences by the pool water initial temperature were observed.

- NOKO (FZ Jülich, Germany) (Krepper et al., 2002):To investigate the heat removal capacity of an emergency con-

denser, comprehensive experiments were performed in the testfacility NOKO. Various tests with different primary and secondarypressures and different start-up temperatures were performed.

- TOPFLOW (FZ Dresden-Rossendorf, Germany) (Prasser et al.,2006):

The equipment with modern two phase measurement tech-niques enables the generation of data which are suitable forCFD code validation. The main purpose of the test was theinvestigation of the heat transfer capability by condensation inslightly inclined tubes, which can be found in an emergencycondenser. The result is the bundle characteristic for differentprimary pressures. Three-dimensional heating up characteristicof the secondary side in the condenser was also obtained.

- Side Wall Heated Tank (FZ Dresden-Rossendorf, Germany)(Krepper et al., 2001):

The test comprises transient heating up of water in a cylin-drical tank from the side walls. Fig. 1 shows the locations fortemperature and void fraction measurements. Temperature dif-ferences up to 50 degrees between the level 1 and the level 6are observed. The CFD simulations by CFX-4 showed that dur-ing this test period the fluid in a very thin layer near the heatedside wall moves upward very fast and almost whole fluid vol-ume within the tank moves very slowly downward. This is thekey phenomenon for establishing the strong temperature strati-fication. During the heating process, the upper region of the tankbecomes well mixed by steam, whereas in the lower region thetemperature stratification remains quite stable. The horizontalboundary between these regions moves gradually downwards.When the boundary meets a thermocouple, the measured tem-perature jumps upward to saturation temperature (see Fig. 2).

- BARC Experiments (BARC, India) (Saha and Sinha, 2005):

The experiments were directed on aspects of the passive decay

heat removal system of AHWR. To understand the basic flowbehaviour near an immersed heater, phenomenological experi-ments were carried out with visualization to obtain flow pattern

4506 J.-H. Choi et al. / Nuclear Engineering and Design 241 (2011) 4504–4514

Table 1Identification and characterization of phenomena for passive safety systems for advanced water cooled NPPs.

Phenomena identification Characterizing thermal-hydraulic aspect

1 Behaviour in large pools of liquid Thermal stratificationNatural/forced convection and circulationSteam condensation (e.g. chugging, etc.)Heat and mass transfer at the upper interface (e.g. vaporization)Liquid draining from small openings (steam and gas transport)

2 Effects of non-condensable gases on condensation heat transfer Effect on mixture to wall heat transfer coefficientMixing with liquid phaseMixing with steam phaseStratification in large volumes at very low velocities

3 Condensation on containment structures Coupling with conduction in larger structures4 Behaviour of containment emergency systems Interaction with primary cooling loops5 Thermo-fluid dynamics and pressure drops in various geometrical

configurations3-D large flow paths e.g. around open doors and stair wells, connection of bigpipes with pools, etc.Gas liquid phase separation at low Re and in laminar flowLocal pressure drops

6 Natural circulation in closed loop Interaction among parallel circulation loops inside and outside the vesselInfluence of non-condensable gasesStabilityReflux condensation

7 Steam liquid interaction Direct condensationPressure waves due to condensation

8 Gravity driven cooling and accumulator behaviour Core cooling and core flooding9 Liquid temperature stratification Lower plenum of vessel

Down-comer of vesselHorizontal/vertical piping

10 Behaviour of emergency heat exchangers and isolation condensers Low pressure phenomena11 Stratification and mixing of boron Interaction between chemical and thermo-hydraulic problems

Time delay for the boron to become effective in the core12 Core make-up tank behaviour Thermal stratification

Natural circulation

Modified from the table provided in Aksan and D’Auria (1996).

Water level at the beginning(210 mm)

T1

T3

T5

T7

T9

T11

T0

T2

T4

T6

T8

T10

T13

T14

T15

T16

T17

T12 Level 1

Level 2

Level 3

Level 4

Level 5

Level 6

V7

V8

V9

V10

V11

V12

V1

V2

V3

V4

V5

V6

ThermocoupleVoid sonde

fo

•

Level 4

Level 5

Level 6

Level 1

Level 2

Level 3

first kind of temperature jumps

second kind oftemperature jumps

160012008004000t[s]

280

300

320

340

360

380

T [K

]

Fig. 1. Locations for temperature and void fraction measurements.

around a single strip heater immersed in a rectangular water tank.When heater is supplied with a power, the particles clinging to theheater move up along the surface of the heater to the top free sur-face. At the free surface, the detaching boundary layer is reflecteddownwards. However, due to its higher temperature (lower den-sity), it is observed to rise back to the free surface and flows alongthe free surface horizontally towards the wall of the container.

For CFD simulation of natural circulation in large pools, theollowing challenges are issued and further study on model devel-pment is required:

The range of validity of the Boussinesq approximation might beexceeded for flow condition with high Rayleigh Numbers.

Fig. 2. Temperature transients in the centre of tank.

• The calculation of the Reynolds stress and the Reynolds flux isbased on isotropic approaches. For natural circulation, however,the gravity plays an important role which requires an anisotropicsimulation.

The competitiveness of CFD is continuously growing due to therapid developments in computer technology. However, computercapacity is still a limiting factor for CFD calculations to produceaccurate results. The different types of errors in CFD simulationsare divided into the two main categories:

• Numerical errors, caused by the discretisation of the flow geom-

etry and the model equations, and by their numerical solution.

• Model errors, which arise from the approximation of physicalprocesses by empirical mathematical models.

J.-H. Choi et al. / Nuclear Engineering and Design 241 (2011) 4504–4514 4507

r stea

amriat

3t

aioeIefagte

acst

y

Fig. 3. Schematic of test set-up fo

Numerical errors should be quantified and reduced to ancceptable level, before comparison with experimental data. Thateans the CFD solution has to be shown as grid-independent, i.e.

esults do not change when the grid is refined further. A grid-ndependent solution can be defined as a solution that can beccepted by the end-user, in view of the purpose of the calcula-ions.

.2. Effect of non-condensable gases on condensation heatransfer

Presence of even a small amount of non-condensable gas (e.g.ir, N2, H2, He, etc.) in the condensing vapor leads to a signif-cant reduction in heat transfer during condensation. The effectf non-condensable gases on the condensation of steam has beenxtensively studied for both natural and forced convection flows.n each of them, geometries of interest (e.g. tubes, plates, annulus,tc.) and the flow orientation (horizontal, vertical) can be dif-erent for various applications. The condensation heat transfer isffected by parameters such as mass fraction of non-condensableas, system pressure, gas/vapor mixture Reynolds number, orien-ations of surface, interfacial shear, Prandtl number of condensate,tc.

Literature review on experimental and theoretical studiesnd mechanistic models on condensation in presence of non-ondensable gases was done for condensation of steam in the

tagnant environment and for condensation of steam inside verticalube.

The researchers generally employ two approaches to the anal-sis. The first approach solves the governing equations in the

m condensation outside the tube.

boundary layer. The second approach considers the mass andenergy balance at the interface and application of heat and masstransfer analogy. The thermal resistances in condensate film andgas/vapor boundary layer are estimated by correlations along withheat and mass transfer analogy. The boundary layer solutions,although insightful from theory point of view, are not useful forcontainment analysis due to their complexity. The analogy basedmodels can be easily implemented into the nodal system codes usedin the containment accident analysis and they provide an insightof the phenomena easily.

New theoretical models are developed within the CRP for thecondensation of steam in presence of non-condensable gas in thestagnant atmosphere and for the condensation heat transfer whensteam/non-condensable gas mixture is flowing in a vertical tube.The results obtained from the specific studies conducted withinthe CRP are summarized here.

3.2.1. BARC experiments on condensation of steam in presence ofair in stagnant atmosphere (Maheshwari et al., 2006)

In order to model the tubes of passive external condenser ofAdvanced Heavy Water Reactor (AHWR), a stainless steel tube ofsize 21.3 mm OD, 15.2 mm ID and 750 mm length is taken as con-densing section for the experiment. Fig. 3 shows the schematic ofthe test set-up in which the condenser model is kept in horizontalorientation.

The variation of heat transfer coefficient was shown as a func-

tion of air mass fraction. The pressure and average wall sub-coolingwere kept almost constant at 2.8 atm and 55 ◦C, respectively. A com-parison of theoretical heat transfer coefficient with experimentaldata is also given in Fig. 4.

4508 J.-H. Choi et al. / Nuclear Engineering and Design 241 (2011) 4504–4514

Fig. 4. Comparison between experimental and theoretical heat transfer coefficient.

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Hea

t tra

nsfe

r coe

ffici

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2 .o C)

Distance from the tube inlet (m)

Air mass fraction =12.8% (Expt. # E20701) Air mass fraction =24.8% (Expt. # E20703)

ig. 5. Test set-up for condensation of steam in the presence of non-condensableas.

.2.2. BARC experiments on condensation of steam in presence ofir inside vertical tubes immersed in a water pool (Maheshwarit al., 2003)

Experiments on condensation of steam inside the vertical tubesf condenser have been carried out in presence of air. Fig. 5 showsschematic diagram of the test set-up.

For a fixed steam inlet flow rate, Fig. 6 shows the axial variationf heat transfer coefficient at two different values of the inlet airass fraction. The condensing heat transfer coefficient decreases

long the length of the condensing tube because the condensationf steam decreases the total mixture flow rate and increases the airass fraction. The experiments were also performed for two differ-

nt inlet pressures of 3.26 × 105 and 1.7 × 105 Pa and at about 25%nlet air mass fraction and 0.0043 kg/s inlet steam flow in one tubesteam mass flux 3.0 kg/m2s). The local heat transfer coefficientor higher pressure (inlet saturation steam temperature) is always

ower as compared to the case with lower saturation temperatureecause of higher accumulation of air in the mixture boundary layerthe vapor/liquid interface temperature is lower due to higher wallub-cooling).

Fig. 6. Heat transfer coefficient as inlet air mass fraction.

3.2.3. Theoretical studies for the condensation in the presence ofnon-condensable gases inside the vertical tube performed by UPV(De La Rosa et al., 2008)

The Polytechnic University of Valencia (UPV) has developeda non-iterative model for the condensation in the presence ofnon-condensable gases inside vertical tubes. The new proposedmechanistic models solve explicitly the real interface temperatureby means of a cubic or a fourth order equation. The models havebeen validated with the Vierow experimental data (Vierow, 1990).

3.3. Condensation on the containment structures

This phenomenon involves heat and mass transfer from the con-tainment atmosphere towards the surrounding structures. Steamcondensation is largely affected by conditions which can be splitinto two groups depending on the relevance of the physical dimen-sions of the system. The “scale-independent factors” are variableslike the fraction of non-condensables, the pressure, the gas com-position and so on, the effect of which could be well investigatedthrough Separate Effect Tests. The “scale-dependent factors” arethose phenomena that are required to be investigated in actual orscaled geometries (i.e., Integral Effect Tests) since physical dimen-sions largely influence their quantitative effect. Examples of thiskind are the natural convection process at both sides of the metal-lic structures and the potential gas stratification. This phenomenonfocuses on the condensation on plate wall in the presence of non-condensable gases.

3.3.1. Models based on theoretical approach (De La Rosa et al.,2009)

The theoretical models can be classified into two main groups.

- Models based on the numerical solution of the conservation equa-tions: Their structure consists in the field equations (continuity,momentum, energy and conservation of vapor species) in com-bination with some thermodynamic relationships and empiricalcorrelations (if required) related with the heat transfer. Thesemodels have usually one or two-dimensional nature and includea specific approach for the interface and, therefore, they are iter-

ative with respect to the interfacial temperature. These models,which will be named “integral models”, are able to simulate thephenomenon of condensation calculating the liquid and gas mass

J.-H. Choi et al. / Nuclear Engineering an

Tab

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d Design 241 (2011) 4504–4514 4509

flows, updating the thermodynamic conditions for each phase,etc.

- Models that directly (and only) solve the equations related to thefundamental parameters of condensation (condensation flow orHTC): These models will be termed as “separate models”. Thesemodels need to be initialized from the boundary conditions for thephenomenon of condensation, whose calculation is conducted bymeans of the field equations or taken from experimental data,such as the local steam velocity or the noncondensables massfraction. Since these models are not restricted in its use to a spe-cific form of the field equations, their implementation is muchmore flexible and, therefore, can be easily employed in thermal-hydraulic codes.

Literature survey on integral and separate models was donewithin CRP. Both the integral and separate models present someproblems when they are implemented in a thermalhydraulic con-tainment code.

- Implementation in a lumped-parameter (LP) code: The integralmodel is practically impossible, due to the necessity of dealingwith large volumes, while the field equations require local valuesfor the variables. Since the separate models deal with the vari-ables of the bulk, their implementation will be also complicatedin this type of codes, because the three-dimensional distributionof the gaseous flow (temperature, mass fraction, etc.) will preventthe values for the bulk from being adequately averaged by meansof the typical volumes of a LP code.

- Implementation in a Computational Fluid Dynamics (CFD) code:The integral models are also difficult to be implemented, since thespecific equations of condensation depend on the way the fieldequations are formulated. With respect to the separate models,their inclusion must take into account that the concept of “bulk”does no exist in CFD codes and therefore different nodalizationsmust be tested in order to obtain adequate values for the variablesin the bulk.

3.3.2. Models based on experiment (De La Rosa et al., 2009)Unlike other thermalhydraulic phenomena, the modelling of

wall condensation in the presence of noncondensable gases hasbeen developed more theoretically than experimentally. Never-theless, the use of empirical correlations in the modelling of thephenomenon has become predominant, especially for two reasons:

- Empirical correlations are easy to apply compared to a modelbased on a system of differential equations.

- The lack of knowledge of the phenomenon required the use ofthe empirical correlations in the evaluation models in order to beconservative.

Due to these reasons, the major part of the licensing activitiesaimed to the calculation of the conservative condition for pressureand temperature in the containment, and to the determination ofthe conservative scenario for the effectiveness of the EmergencyCore Cooling System have been employed very conservative empir-ical correlations such as Uchida and Tagami correlations (Tagami,1965; Uchida et al., 1965).

Experimental facilities that can be used whether for the vali-dation of models or for the characterization of the condensation

phenomenon within the containment are surveyed from the liter-ature and sorted according to the classification. Subsequently, theranges of the fundamental parameters of condensation for eachfacility are summarized in Table 2.

4510 J.-H. Choi et al. / Nuclear Engineering and Design 241 (2011) 4504–4514

Table 3Condensation models implemented in thermal-hydraulic codes (De La Rosa et al., 2009).

Code Type of code Nature of the code Condensation model

GOTHIC Nuclear LP/field code Uchida correlation, and Gido and Koestel correlationCOMMIX-1D Nuclear Field code Collier’s model (HMTAa)PCCSAC Nuclear Field code Collier’s model (HMTA)CFX4 Commercial Field code Non-implementedFLUENT Commercial Field code Non-implementedMAAP-DBA Nuclear LP Uchida and Tagami correlationsCONTAIN Nuclear LP Collier’s model (HMTA)CONTEMPT-LT Nuclear LP Collier’s model (HMTA)CONTEMPT-4/MOD5 Nuclear LP Tagami correlation and othersMELCOR Nuclear LP Collier’s mass flux and Chilton–Colburn correlation (HMTA)COMPACT Nuclear LP Uchida and Tagami correlationsTONUS Nuclear LP/field code Chilton/Bird condensate rate formulation with McAdams correlation (HMTA)

a ‘HMTA’ means heat and mass transfer analogy.

PCCS

3

tPmft

tiplitih

ads

Fig. 7. Schematic of PUMA–

.3.3. Validation of computer codesSince the validation of the condensation phenomenon belongs

o the more general objective of validating the behaviour of theassive Containment Cooling System (PCCS) in the reactor contain-ent, the validation of a condensation model must be achieved

rom the validation of the thermalhydraulic code simulation wherehe model is implemented.

Since the phenomenon of condensation is strongly affected byhe boundary conditions, especially for the flow thermodynam-cs (regime, flow patterns, velocity, composition, temperature andressure), but also for the sub-cooling grade of the wall and the

ength of the condensing surface, the code capability in simulat-ng the condensation phenomenon will be strongly affected byhe capability of simulating the flow conditions. Validation heres restricted only to the capability of simulating the condensationeat and mass transfer.

Table 3 shows the major computer codes used for containmentnalysis and condensation models implemented in the codes. Vali-ation of computer codes was summarized based on the literatureurvey. Due to the commercial nature of many CFD codes, a con-

separate effect test facility.

densation model is usually not included and must be implementedby the user.

3.4. Behaviour of containment emergency systems

Nuclear power reactor containments are equipped with safetysystems which protect the containment integrity under variousaccident conditions. The focus of this phenomenon is the natu-ral circulation cooling and heat transfer in various containmentpassive cooling systems under accident conditions to remove theenergy out of the containment by natural circulation and conden-sation heat transfer. Typical systems adopted in advanced designsare:

- Vertical PCCS condenser: The Simplified Boiling Water Reactor

(SBWR) and Economic SBWR (ESBWR) developed by GeneralElectric (GE) were designed to employ a vertical PCCS. The PCCSis a system that passively removes heat from the containment tothe liquid pool surrounding the PCCS heat exchangers.

ing and Design 241 (2011) 4504–4514 4511

-

-

-

3

ostivwc(

ve

scicoAdwhdtoelsg

3

aPrspea

3P

an

Fig. 8. Effect of NC gas mass fraction on PCCS condensation heat transfer coefficientfor bypass and cyclic venting modes.

J.-H. Choi et al. / Nuclear Engineer

Horizontal PCCS condenser: The other type of the PCCS was pro-posed by the Japanese industries for the ABWR-II. For this designconcept, the PCCS has only a function of the heat removal from thecontainment during severe accidents with failures of the activesafety systems.External air cooling system: The AP600 and AP1000, advancedpressurized water reactors developed by Westinghouse, utilize aPCCS to remove heat released inside the containment vessel fol-lowing a postulated Design Basis Accident (DBA). During a DBA,heat released to the interior of the steel containment vessel isremoved by evaporation of a continuously flowing thin liquid filmon the outside surface of the vessel, which lowers the tempera-ture of the steel vessel wall so that steam condenses on its insidesurface.Containment cooling condenser: The SWR-1000, an advanced BWRdeveloped by AREVA, utilizes the containment cooling con-densers to remove residual heat passively from the containmentatmosphere to the shielding/storage pool located above the con-tainment.

.4.1. PUMA–PCCS separate effect tests (Choi, 2009)A series of experiments have been conducted in this study to

btain experimental data for the downward co-current flow of ateam-air mixture through the condenser tube bundles during thehree operational modes of the PCCS: the bypass mode during thenitial blowdown phase, the cyclic venting mode with the periodicenting of non-condensable (NC) gas through the PCCS vent linehich is submerged into the wet well water, and the long-term

ooling mode with a very low NC gas concentration in the dry wellFig. 7).

In Table 4, test conditions are summarized for the bypass, cyclicenting, and long-term cooling modes for the PUMA–PCCS separateffect tests.

The experimental results showed that as the PCCS inlet pres-ure increases, the heat removal rate increases while the averageondensation heat transfer coefficient decreases. The PCCS poolnventory is one of the key parameters that determine the PCCSooling capability. The condensation heat transfer characteristicsf the PCCS in the presence of NC gas have also been investigated.s the PCCS inlet NC gas concentration increases, the average con-ensation heat transfer coefficient decreases (Fig. 8). Comparedith Kuhn’s (1995) and Vierow’s (1990) models among others, theeat transfer coefficients and heat removal rates from the PUMAata have the same trends as those from the two models, howeverhe average heat transfer coefficients from the PUMA data werever-predicted by the two models. The over-prediction could bexplained by the difference of PCCS boundary conditions; singleoop experiment did not contain the prototypic boundary conditionuch as PCCS pool boiling and cyclic venting of non-condensableas.

.4.2. ISP-42 PANDA tests data baseThe main interest for this ISP was code validation in relation to

range of LWR and ALWR (mainly) containment issues. The ISP-42ANDA test consists of six phases, A through F. These phases rep-esent a sequence of operating modes or processes as used for theimulation and study of the behaviour of ALWR containments withassive safety systems. Each of these phases is in fact a separatexperiment, with its own initial and boundary conditions (Aksannd Luebbesmeyer, 2003).

.4.3. JAEA’s research project on horizontal heat exchanger for

CCS (Kondo et al., 2005)

The research project was conducted at JAEA to clarify thepplicability of the PCCS with horizontal heat exchangers to aext-generation BWR. The design requirement of the PCCS of this

Fig. 9. Comparison of the experimental data and predictions with existing and pro-posed models.

concept is to prevent the containment damage due to pressuriza-tion for at least one day during severe accidents with failures of theactive safety systems. The research program consisted of the fun-damental thermal-hydraulic experiments, the bundle experiments,the code validation and modification, and the reactor analyses withthe modified code.

The fundamental thermal-hydraulic experiments were per-formed first to investigate the performance of a horizontalcondenser tube under wide ranges of experimental conditions.Experimental results have shown that the condensation heat trans-fer coefficients for the annular flow regime are under-predicted byexisting correlations (Fig. 9). A new model was developed, whichwas based on the measured relationship between roll wave passingfrequencies and the heat transfer rates.

The large-scale experiments were performed secondly to con-firm the total performance of the horizontal heat exchanger andto validate analysis codes to predict overall thermal-hydraulic

behaviour of the horizontal heat exchanger. Various flow regimeswere observed in the secondary side of the tube bundle includingsingle-phase liquid, bubbly, and churn flows with the increase inthe elevation in the tube bundle.

4512 J.-H. Choi et al. / Nuclear Engineering and Design 241 (2011) 4504–4514

Table 4Test matrix for PUMA–PCCS separate effect tests.

Test mode Steam flow rate (kg/s) Pressure (kPa) PCCS pool water level (m) NC gas mass fraction (%)

Long-term cooling (LT) 0.018–0.028 200 0.92 <1230 0.60260

Cyclic venting (CY) 0.031–0.042 220 0.92 0.3, 2, 4240 0.60260

0.92 0, 10, 15

3g

spv(viggsp

diotdosg

fhft

R

wGtlra

W

w∑

W

Ap

-

105 106 107 108 109 1010 1011 1012 1013 1014 1015

102

103

104

105

Turbulent flow equation( C=1.96, r=0.364)

-25%

+25%

Laminar flow equation( C=0.1768, r=0.5)

Re s

s

Grm/ NG

Generalized correlation Ress=C [Grm/NG]r

Loop with ID:1/2" Loop with ID:3/4" Loop with ID:1"

Bypass (BY) 0.060–0.075 220260300

.5. Thermo-fluid dynamics and pressure drops in variouseometrical configurations

It is customary to express the total pressure drop in a flowingystem as the sum of its individual components such as distributedressure loss due to friction, local pressure losses due to suddenariations of shape, flow area, direction, etc. and pressure lossesthe reversible ones) due to acceleration (induced by flow areaariation or by density change in the fluid) and elevation (grav-ty effect). An important factor affecting the pressure loss is theeometry. In a nuclear reactor, we have to deal with several basiceometrical shapes (circular pipes, annuli, etc.) and a number ofpecial devices like rod bundles, heat exchangers, valves, headers,lenums, pumps, large pools, etc.

Another very important issue is concerned with the driving forceepending on whether the flow is sustained by a density difference

n the fluid (natural circulation) or by a pump (forced convection),r whether there will be feedback between the pressure loss andhe extracted power or not. Normally the pressure loss inside aevice depends on the nature of flow through the device and notn the nature of driving head causing the flow. However, underome circumstances, because of local effects, the pressure loss mayet influenced by the nature of driving force.

Pressure drop relationships for single and two-phase availablerom literature survey are reviewed and some new correlationsave been developed within the CRP. BARC developed a generalized

orm of flow correlation for single and two-phase natural circula-ion loops (Vijayan and Austregesilo, 1994; Gartia et al., 2006):

ess = C(

Grm

NG

)r

(1)

here Ress is steady state Reynolds number, Grm is modifiedrashof number, NG is dimensionless parameter, and C and r are

he constants determined by the constants of friction factor corre-ation. Based on this generalized correlation, the steady state flowate for single and two-phase natural circulation loop was obtaineds

ss =[

2 g �20 ˇT H Qh

R Cp

] 13

for single-phase (2)

here the total hydraulic resistance of the loop is given by, R =Nti=1((fiLi/Di) + Ki)(1/A2

i).

ss =[

2p

g �r ˇtp H Q Dbr A2−b

r �l

�br NG

] 13−b

for two-phase (3)

ssessment on single and two-phase natural circulation loop waserformed for:

Effect of friction factor on steady state flow rate in a single-phasenatural circulation loop (uniform and non-uniform diameterloop).

Fig. 10. Effect of friction factor on steady state flow rate in a two-phase naturalcirculation loop.

- Effect of friction factor on steady state flow rate in a two-phasenatural circulation loop (uniform and non-uniform diameterloop).

- Effect of two-phase friction factor multiplier on steady state flowrate in a two-phase natural circulation loop.

- Effect of pressure on steady state flow rate in a two-phase naturalcirculation loop.

- Effect of friction factor on stability of a single-phase natural cir-culation loop.

- Effect of two-phase friction factor multiplier on stability of a two-phase natural circulation loop.

In Fig. 10, the experimental results obtained from three differ-ent natural circulation loops are compared with theoretical resultsbased on the above relationships. Further, data on both uniformdiameter loop (UDL) and non-uniform diameter loop (NDL) werecompared with the two-phase generalized correlation (Fig. 11). Ingeneral, a reasonably good agreement is obtained with all reporteddata. However, in experiments where complex geometries areinvolved (e.g. in NDL), the friction factor correlation used may beinsufficient to obtain reasonable agreement. Hence, large devia-tions are found in case of non-uniform diameter loop data as shownin Fig. 11. In such cases, it may be required to determine the pres-sure drop experimentally and then the flow rate can be calculatedas suggested in Eq. (3).

Fig. 12 shows the stability map for a single-phase naturalcirculation loop with HHHC (Horizontal-Heater and Horizontal-Cooler) orientation (Vijayan, 2002). Where Stm is modified Stanton

number. This figure shows that the stability boundary changeswith the choice of friction factor correlation even in single-phaseloops. Fig. 13 (Nayak et al., 2007) shows the stability map fora two-phase natural circulation loop along with the instability

J.-H. Choi et al. / Nuclear Engineering an

107 109 1011 1013 1015102

103

104

105

106

- 40%+40%

Re ss

Grm / NG

Generalized correlation UDL data NDL data

Fig. 11. Steady state performance of uniform and non-uniform diameter two-phasenatural circulation loops.

10987654321010-7

100

107

1014

1021

1028

1035

Lt=7.23m, Lt/D=267.29, HHHC Orientation

Stable

Stable

Uns

tabl

e

Gr m

Stm

Turbulent flow (p=0.184, b=0.2) Turbulent flow (p=0.316, b=0.25)

Fig. 12. Effect of friction factor on stability in a single-phase natural circulation loop.

353025201510500

20

40

60

80

100

120ΔTsub= 4K Homogenous model

Lockhart-Martinelli model Martinelli-Nelson model Chisholm-Laird model Sekoguchi model Threshold of Instability

[Furutera (1986)]

Stable

Unstable

Pow

er (k

W)

Pressure (bar)

Fig. 13. Effect of two-phase friction factor multiplier on the stability of a two-phasenatural circulation loop.

d Design 241 (2011) 4504–4514 4513

threshold obtained experimentally. It is clear that the threshold ofinstability predicted by the code may vary with the choice of two-phase friction factor multiplier in a two-phase natural circulationloop.

Within the range of parameter studied so far, relationships forforced circulation as given in TECDOC-1203 (IAEA, 2001) werefound to be adequate for studying natural circulation and stabil-ity of natural circulation. More accurate prediction capability isrequired at low mass flux and for large area flow paths. However,this issue is not unique to only natural circulation systems.

4. Conclusions

A great improvement in understanding of natural circulationphenomena and modelling in water cooled nuclear power plantshas been achieved through the IAEA CRP on “Natural CirculationPhenomena, Modelling, and Reliability of Passive Systems that Uti-lize Natural Circulation”. The major results of this CRP will bepublished in detail as two complementary TECDOC reports. Thefirst one (IAEA, 2009) was published in 2009 and the last one isscheduled to be issued in 2011.

Acknowledgments

The IAEA expresses its appreciation to all experts participatedin the CRP: N. Aksan (CRP Chairperson, Formerly PSI, Switzerland),M. Giménez (CNEA, Argentina), M. Marques (CEA, France), E. Krep-per (FZ Rossendorf, Germany), F. D’Auria (Univ. of Pisa, Italy), D.Saha (BARC, India), Y. Sibamoto (JAEA, Japan), Y.-J. Chung (KAERI,Rep. of Korea), K. Korotaev (Gidropress, Russia), W. Krotiuk (USNRC,USA), J. Reyes & B. Woods (Oregon State Univ., USA), L. Burgazzi(ENEA, Italy), P. Matejovic (IVS, Slovakia), R. Bolado (JRC Institutefor Energy, EC), M. Ishii (Purdue Univ. USA), B. Williams (Idaho StateUniv., USA), J. De La Rosa (Univ. of Valencia, Spain), and H. Khartabil(AECL, Canada).

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