Passive Containment Cooling System Research Articles

Modified Evaporation Model for ATHLET System Code in a Passive Containment Cooling System for Nuclear Safety

The presented work deals with the improvement of the evaporation model of the ATHLET (Analysis of Thermal and Hydraulics of Leaks and Transients) system code to be applied to a passive containment cooling system of a nuclear power plant. For the model validation, INTRAVIT (Investigation of Passive Heat Transfer in a Variably Inclined Tube) test facility setup at the University of Luxembourg was used. The first part of the paper presents a review of the existing literature on evaporation models that revealed that those models significantly simplify the physical processes that occur. Next, a modified evaporation model is proposed that offers a realistic description of various evaporation processes and the start of bubble formation using a nucleation model, and a surface density calculation model is introduced that is necessary for evaporation simulation. The final part of this work explored five different system configurations to test the evaporation model: three condenser tube inclinations (5 deg, 60 deg, and 90 deg), two riser lengths (1 m and 2.5 m), and different thermal loads. They made it possible to simulate several experiments for stable and unstable natural circulation and to verify the proposed model.

CFD Analysis of Natural Convection on the Outer Surface of the Containment of APWR Model

To improve the safety of nuclear power plants, a modern NPP such as Advance Pressurize Water Reactor (APWR) is equipped with a safety feature called PCCS (Passive Containment Cooling System), where naturally circulating air cools the outer surface of the containment which is used to remove decay heat released inside the containment vessel. The decay heat from the reactor core will be transferred out through the conduction mechanism and the natural convection mechanism on the outer surface of the containment vessel. Then, the hot air that forms in the gap between the containment and its baffles will rose through the chimney located at the top of the concrete building and cold air will enter through the inlet to create a natural circulation cycle. To obtain the effectiveness of the use of baffles, as a comparison, it is necessary to analyze the natural convection heat transfer on the outer surface of a containment vessel that is not equipped with a baffle. In this research, CFD analysis of natural convection on the outer surface of the containment of APWR reactor model has been done. Based on the model developed, the analysis was done to get temperature and convection heat transfer coefficient in the air flow on the containment surface. Heat flux in the containment surface varied from 500 W/m2 to 2.000 W/m2 with an increase of 500 W/m2 intervals. Based on the analysis results, the correlation equations are also proposed in this paper, namely the correlation of natural convection heat transfer for laminar regime on a vertical cylindrical with the heat flux constant in the form Nuq = a(Raq)b.

Machine learning-driven approach for predicting the condensation heat transfer coefficient (HTC) in the presence of non-condensable gases

This research focuses on developing a predictive model that can effectively estimate the condensation heat transfer coefficient (HTC) on vertical tube surfaces during free-fall. The model takes into consideration the presence of non-condensable gases. The main objective is to assemble an extensive and varied database that encompasses different geometric parameters and operational conditions. The motivation here stems from the urgent need to develop practical models and/or correlations that can accurately predict the condensation heat transfer efficiency for industrial heat exchangers affected by non-condensable gases, with a particular focus on passive containment cooling systems (PCCS) employed in nuclear power plants (NPPs). This innovative cooling system utilizes the force of gravity to condense water steam, effectively removing heat from the containment vessel in the event of an accident. For the development of the model, we employed the neural network toolbox in MATLAB to construct an Artificial Neural Network (ANN) known as a Multi-Layer Perceptron (MLP) network. This model was designed to predict the HTC during the condensation process involving two different types of Non-Condensable Gases (NCG): Air and Nitrogen, along with Helium as a representative light gas. To build the model, we utilized a dataset comprising 1,888 data points sourced from 21 experimental references. The input layer of our model receives a range of parameters, encompassing total pressure (Ptot), subcooling temperature (ΔTsub), mass fraction of non-condensable gases (wnc), wall length (L), hydraulic diameter (Dh), the ratio of the Helium to non-condensable gases (RHe/nc), liquid density (ρl), gas density (ρg), liquid viscosity (μl), gas viscosity (μg), saturation temperature (Tsat), latent heat (hfg), liquid thermal conductivity (kl), and gas specific heat (cp,g). The output data of our model corresponds to the condensation HTC (h). To ensure consistent scaling, the input data was normalized by scaling each feature within the range of 0 to 1. On the other hand, the output data underwent a transformation using the natural logarithm. By utilizing Pearson correlation coefficient analysis and importance analysis for parameter selection, a new correlation for heat transfer was developed and evaluated. The developed machine learning model exhibited outstanding effectiveness in precisely forecasting the condensation HTC with an accuracy range of ± 10 % and ± 20 %. The outcome of this investigation represents the notable progress in analyzing extensive datasets gathered through experiments. Specifically, it addresses the previously unexplored influence of Helium gas by harnessing the capabilities of machine learning.

Evaluating direct vessel injection accident-event progression of AP1000 and key figures of merit to support the design and development of water-cooled small modular reactors

The passive safety systems (PSSs) within water-cooled reactors are meticulously engineered to function autonomously, requiring no external power source or manual intervention. They depend exclusively on inherent natural forces and the fundamental principles of reactor physics, such as gravity, natural convection, and phase changes, to manage, alleviate, and avert the release of radioactive materials into the environment during accident scenarios like a loss-of-coolant accident (LOCA). PSSs are already integrated into such operating commercial reactors as the Advanced Pressurized Reactor-1000 MWe (AP1000) and the Water-Water Energetic Reactor-1200 MWe (WWER-1200) are adopted in most of the upcoming small modular reactor (SMR) designs. Examples of water-cooled SMR PSSs are the passive emergency core-cooling system (ECCS), passive containment cooling system (PCCS), and passive decay-heat removal system, the designs of which vary based on reactor system-design requirements. However, understanding the accident-event progression and phases of a LOCA is pivotal for adopting a specific PSS for a new SMR design. This study covers the accident-event progression for direct vessel injection (DVI) small-break loss-of-coolant accident (SB-LOCA), associated physics phenomena, knowledge gaps, and important figures of merit (FOMs) that may need to be evaluated and assessed to validate thermal-hydraulics models with an available experimental dataset to support new SMR design and development.

Overview of separate effect and integral system tests on the passive containment cooling system of SMART100

SMART100 has a containment pressure and radioactivity suppression system (CPRSS) for passive containment cooling system (PCCS). This prevents overheating and over-pressurization of a containment through direct contact condensation in an in-containment refueling water storage tank (IRWST) and wall condensation in a CPRSS heat exchanger (CHX) in an emergency cool-down tank (ECT). The Korea Atomic Energy Research Institute (KAERI) constructed scaled-down test facilities, SISTA1 and SISTA2, for the thermal-hydraulic validation of the SMART100 CPRSS. Three separate effect tests were performed using SISTA1 to confirm the heat removal characteristics of SMART100 CPRSS. When the low mass flux steam with or without non-condensable gas is released into an IRWST, the conditions for mitigation of the chugging phenomenon were identified, and the physical variables were quantified by the 3D reconstruction method. The local behavior of the non-condensable gas was measured after condensation inside heat exchanger using a traverse system. Stratification of non-condensable gas occurred in large tank of the natural circulation loop. SISTA2 was used to simulate a small break loss-of-coolant accident (SBLCOA) transient. Since the test apparatus was a metal tank, compensations of initial heat transfer to the material and effect of heat loss during long-term operation were important for simulating cooling performance of SMART100 CPRSS. The pressure of SMART100 CPRSS was maintained below the design limit for 3 days even under sufficiently conservative conditions of an SBLOCA transient.

Heat transfer characteristics of passive containment cooling system with equal height difference under steady-state conditions

In the nuclear energy utilization, the passive containment cooling system (PCCS) holds paramount importance, with its heat transfer performance standing as a critical aspect of nuclear safety. This study simulated the flow and heat transfer behaviors of PCCS with equal height difference by APROS software under a broad range of steady state operating conditions. At first, the flow pattern of natural circulation is categorized by phase state of flow through operational parameters. A criterion model of boundary of phase change is proposed with prediction deviation of less than ±25 %. Furthermore, the influence of various operation parameters such as temperature, non-condensable gas content, and pressure on the heat transfer behaviors of natural circulation was also revealed. A new heat transfer coefficient correlation for both single-phase and two-phase conditions is proposed with prediction deviation of ±15 %. Finally, the thermal stratification phenomenon in the containment vessel with non-condensable gas was investigated, indicating that local heat transfer characteristics in the upper region surpass those in the lower section on account of deposition of the non-condensable gas at bottom of the vessel.

Research of aerosol removal by passive containment cooling system

Passive containment cooling system (PCCS) can effectively remove the heat from the containment and maintain the pressure and temperature in case of accident, thus ensuring the containment integrity. PCCS introduces thermophoresis and diffusiophoresis aerosol deposition and changes the distribution of fission products in the containment, thus affecting containment dose evaluation and off-site source term assessment. In this paper, a Generation III PWR nuclear power plant model with PCCS is established by using ASTEC (Accident Source Term Evaluation Code), and the removal effect of PCCS on fission product aerosol is studied under the large break loss of coolant accident. The results show that one train of PCCS can increase the aerosol removal factor for approximately 0.05/h. Taking five trains of PCCS in this study as an example, aerosol deposition effect introduced by PCCS reduces suspended aerosol in containment by 10 %–20 %, and the effect is more significant in the PCCS-locating compartment. Meanwhile, the gas flow among compartments changes the aerosol mass in the adjacency of the PCCS-locating compartment. This phenomenon weakens the removal effect in the PCCS-locating compartment and make the aerosol concentration in adjacent compartments tend to be consistent.

Thermal-hydraulic characteristics analysis of steam–air condensation in vertical porous-wall corrugated tubes

In the passive containment cooling systems (PCCS) of Generation III nuclear power plants, the condensation performance of steam containing non-condensable gas (NCG) in condensate tubes is critical for removing heat from the nuclear power plant containment. By applying a novel porous-wall corrugated structure in the condenser tubes of this system, it was found that under certain conditions, a double vortex is formed inside and outside the porous layer, which is conductive to the transportation and condensation of steam to the condensing wall. The VOF multiphase relation and Brinkman–Forchheimerextended Darcy model in combination of local heat balance, condensation and species transport equations are employed to describe the heat and mass transfer in the porous-wall corrugated tube during condensation. The numerical analysis are conducted to investigate the effects of waveform channel, porous layer and its thicknesses, subcooling of cooling wall and inlet air mass fractions on flow and condensation under different inlet velocities, in which simulations agree with the experimental data. The results show that the condensation heat transfer is greatly influenced by the vortex distribution and air accumulation. At inflowing velocity of 1 m/s, the waveform with amplitude of 1 mm and wavelength of 18 mm can achieve the highest performance evaluation factor, φ (about 67.5 % enhancement than the smooth tubes). The ratio of waveform amplitude to wavelength needs to be chosen properly in consideration of convection between solid matrix and steam–air mixture in porous layer, and non-condensation gas distribution near cooling wall of condensation tube.

Experimental and numerical study on the falling film flow process on the outer wall of dome cylinder

The passive containment cooling system (PCCS) is crucial for ensuring the safety of nuclear power plants, and the thickness and temperature of the falling film on its outer wall are important parameters that determine the heat and mass transfer performances. In this paper, a compact liquid film thickness and temperature simultaneous measurement system was developed to provide a measurement method, which is based on the absorption spectroscopy technology, and then the measurement accuracy was verified by using a calibration tool. The results showed that the mean relative error of the liquid film thickness and temperature measurements from the known values were about 3.4% and 0.4%, respectively. On this basis, this system was used to study the falling film process on the outer wall of the dome cylinder analogous to the containment under different working conditions and combined with numerical simulation for comparison. The results showed that the two methods have good consistency. The variation of the liquid film heat transfer coefficient for various operating conditions was further analyzed. It was found that the average thickness of the falling film is positively correlated with the liquid flow rate, but negatively correlated with the temperature and heat transfer coefficient. In summary, this system can accurately study the falling film process on the outer wall of the dome cylinder, which is expected to accurately analyze the heat and mass transfer process in the PCCS and provide data support for its design and optimization.

The modeling of passive containment cooling system and sensitivity analysis using the SPRUCE code

Passive Containment Cooling Systems (PCCSs) have attracted wide attention to improve the inherent safety of nuclear power plants. This study summarizes a system analysis method of a closed-loop natural circulation of the PCCS via separately evaporation and condensation. A lumped-parameter model is built using SPRUCE, a code of system simulation of severe accidents, to investigate the natural circulation characteristics and heat-removal efficiency of the PCCS. The analysis of heat-removal capacity of the PCCS is carried out under typically thermal conditions of the containment, such as pressure from 0.28 to 0.4 MPa, and the results are validated by comparison with the experimental data obtained from a single-loop testing facility of the PCCS in full scale. The SPRUCE model agrees with the data, despite slightly underestimating the rate of condensation heat transfer. It conservatively satisfies the requirement of design capacity. To further verify the effectiveness of mitigation on containment overpressure, a representative scenario of LBLOCA (Large Break Loss of Coolant Accident) is implemented using the code. Meanwhile, the effects of non-condensable gas are also presented, as a reference for optimization of PCCS design.

Steam condensation on a circular tube in the presence of non-condensables (air) in passive containment cooling system

Steam inside the containment should be condensed to reduce the system pressure after severe accidents such as loss of coolant accident (LOCA) and main steam line break (MSLB). Due to the absence of an active coolant system, the passive containment cooling system (PCCS) can be used to improve safety by condensing the steam on a test section (coolant) tube surface. However, the steam condensation is degraded by the presence of non-condensable gases like air. The condensation heat transfer coefficient is dependent on operating parameters such as system pressure, air-mass fraction, and wall subcooling and geometric parameters such as tube diameter, tube length, and angle of orientation.In the present study, experiments are conducted on a vertical tube to estimate the condensation heat transfer coefficient in the presence of air. The system pressure ranges from 2 - 5.5 bar, air-mass fraction of 0.3 - 0.8 and wall subcooling of 27 - 80°C. The influence of test section tube diameter (48.2 mm, 32 mm, 22 mm, and 16 mm) and inclination angles (90° and 68° from the horizontal) are investigated. Based on the experimental data, a correlation for condensation heat transfer coefficient in terms of system pressure, air-mass fraction, wall subcooling, tube diameter and angle of inclination is proposed. The condensation heat transfer coefficient increases with the increase in the system pressure and decreases with the increase in the air mass fraction, wall subcooling and condenser tube diameter. With the decrease in the diameter of the vertical condenser tube from 48.2 mm to 16 mm, an increase in the condensation heat transfer coefficient of 66% is observed. A non-dimensional correlation for condensate Nusselt number as a function of Grashof number, air-mass fraction and Jakob number is also proposed. This correlation predicts 80% and 90% experimental data within 20% and 30% uncertainty range, respectively.

Film condensation with high heat fluxes and scaled experiments using pure steam for reactor containment cooling

Condensation tests were performed using a newly developed test facility for scaling the passive containment cooling system (PCCS) to a small modular reactor (SMR). The PCCS of the SMR plays a pivotal role in ensuring greater safety, reliability, and compactness than what is afforded by traditional reactors. Therefore, a well-designed PCCS is essential to SMRs. However, previous studies and test data were unsuitable for scaling, due to high variation in the test geometry and operating conditions. This study intends to close this research gap by using a novel designed scaled test facility consisting of vertical condensing test sections featuring 1-, 2-, and 4-inch-diameter condensing tubes with annular water cooling, and by applying superheated and saturated steam with different steam mass flow ranges of 5–25 g/s. The primary test data, including axial temperatures, mass flow rates, and pressures, were used in conjunction with a standard data reduction method to estimate critical parameters such as heat fluxes, heat transfer coefficients, and condensation rates. These scaled test data would support improving empirical correlations and validating condensation models to identify scaling distortion for SMR PCCSs.

PANDA experimental database and further needs for containment analyses

PANDA is a large-scale, multi-compartment, thermal-hydraulics facility located at the Paul Scherrer Institute (PSI) in Switzerland and is operated for nuclear safety projects. PANDA is continuously upgraded with new components and instruments to meet the requirements of each new project. The modularity of the facility and its instrumentation provide unique opportunities to perform experiments ranging from basic containment phenomena studied mainly in a single vessel to integral behavior of Passive Containment Cooling Systems (PCCS) requiring up to six PANDA vessels and four passive condensers (PCCs, IC) submerged in four water pools. Experimental data analyses are performed using system codes, lumped parameter (LP) severe accident codes and Computational Fluid Dynamics (CFD) codes. In this paper, we provide an overview of the PANDA experimental database and identify further data needs for containment analyses, considering the main results of recently completed PANDA projects.

Time-Average Heat Transfer Coefficient for Steam-Air Condensation during the Dropping of Containment Pressure

The passive containment cooling system (PCCS) has been applied in the new generation nuclear power plant. Previous studies mainly focused on steady-state heat transfer, but the actual heat transfer of PCCS is the transient process. Thus, investigating the transient heat transfer during the containment pressure dropping is necessary. The present study proposes a time-average condensation heat transfer coefficient (time-average condensation HTC) to characterize the intensity of transient condensation heat transfer. The time-average condensation HTC is defined as the heat absorbed per unit heat transfer area and per unit wall subcooling in a unit time. Experiments were performed to research the effect of initial gas-mixture pressure and air mass fraction on transient HTC. The result showed that the more significant gas-mixture pressure and lower air mass fraction could promote the transient-state heat transfer, especially the low air mass fraction. Based on the experimental result, a detailed empirical correlation for the time-average heat transfer coefficient is developed with an error of ±20%. Another simplified empirical correlation that only includes air mass fraction is also proposed to predict the transient heat transfer roughly. Besides, research on self-sustaining stability is also conducted to evaluate the stable operation characteristic of PCCS. The system can respond quickly and then run to a new steady state after interference during the long-term operation of PCCS. The phenomenon above implies that PCCS has good self-sustaining stability.

MARS-CUPID 연계코드를 이용한 피동원자로건물냉각계통 격납건물 내부 열수력거동 분석

A passive containment cooling system (PCCS) has been designed to effectively remove the energy released to a containment by using condensation heat during a loss of coolant accident (LOCA) in a nuclear reactor. During the past few years, thermal-hydraulic performance of the PCCS has been analyzed using system scale codes, MARS and GOTHIC codes. In this work, thermal-hydraulic performance of the PCCS was analyzed again using the MARS-CUPID coupled code, which is a combinded code of the system scale and component scale, respectively. For a more realistic simulation, we modified Uchida correlation to take into account the effect of pressure in the MARS-CUPID coupled code. The simulation result showed the peak pressure of 4.31 bar, which is similar to the peak pressure obtained using the GOTHIC code. In this study, thermal-hydraulic performance of the PCCS considering uncertainty parameters, such as mass&energy release, heat transfer coefficient, and heat sink area, is also discussed.

Development of a PCCS module in the ATHROC for simulation of passive containment cooling system

Advanced reactors, such as HPR1000 and VVER1200, have adopted passive containment cooling systems (PCCSs) to maintain containment integrity under accident conditions. In order to meet the requirements of behavior and effect analysis of PCCS in containment, a multi-compartment containment code ATHROC was further developed by adding PCCS module. Then the code benchmarking based on RELAP5 MOD3.4 and the experimental validations based on PANGU experiment were conducted. Finally, the thermo-hydraulic analysis of the containment of CPR1000 (a gigawatt pressurized water reactor) with PCCSs during LOCA was carried out using ATHROC. The results show that although two sets of PCCS are able to inhibit the containment pressure rise under LBLOCA with the failure of CSP (Containment Spray system), the coolant in water tanks evaporates too fast to meet the requirement of a 72-h passive heat removal from the containment. This study newly proposed a validated and benchmarked multi-compartment containment code called ATHROC for thermo-hydraulic analysis of a containment with PCCS in single-phase and two-phase states. Both containment and PCCS transients can be captured by ATHROC. It is demonstrated that the ATHROC code could offer good practicability and adaptability for capturing both containment and PCCS transients.

Thermal‐hydraulic investigation of a proposed concept for a passive containment cooling system and evaluation of heat exchanger performance

Reduction in ambient pressure within the containment of a light water reactor in the event of a hypothetical Loss of Coolant Accident (LOCA) is crucial for containment integrity. This article introduces the distinct concept of a Passive Containment Cooling System (PCCS). In the proposed PCCS concept, the containment is divided into two compartments with appropriate volumes. An external heat exchanger is connected between the two compartments of the containment for heat transfer and pressure reduction in the case of LOCA. This concept has unique characteristics and provides many advantages over previous designs. A RELAP5 model was developed to perform thermal-hydraulic analysis of the proposed PCCS concept. The performance of the PCCS after its deployment on a small reactor (998.6 MWth) has been assessed under various conditions following LOCA. The results indicate that partitioning of the containment creates a differential pressure that acts as a driving force for flow through the PCCS heat exchanger. Consequently, a decrease in long-term containment pressure is observed. In the case of equal volume fractions in the two containment compartments, the PCCS reduces long-term pressure in the containment by 16.5%. Sensitivity analysis was conducted in which the flow area of the PCCS inlet line, the heat transfer area, and the size of the heat exchanger tank were varied. The PCCS heat transfer capacity increases as these parameters are increased. However, the former is less sensitive than the latter two. The performance of the PCCS heat exchanger under steady-state conditions was validated using RELAP5 simulations, empirical correlations, and analytical models. The evaluation of heat exchanger parameters by various methods (presented in Appendix 1 ) was found in close agreement.

Experimental study of steam-air mixture thermal stratification during the dropping of containment pressure

The passive containment cooling system (PCCS) as a passive technology has been applied in Generation Ⅲ nuclear power plants. Experiments were conducted in a scale model to analyze the thermal stratification in containment during the containment pressure dropping. Results indicate that the obvious thermal stratification occurs in the containment during the transient-state heat transfer process but not during the steady-state heat transfer process. For the transient-state process, the steam-air mixture temperature increases by 2 °C for each 100 mm increase in the elevation direction of the pressure tank. The nonuniform distribution of air in containment causes thermal stratification, which is more pronounced for lower air mass fraction and higher initial containment pressure. The thermal stratification effect on the condensation heat transfer process is assessed via diffusion layer heat transfer and film heat transfer to evaluate its effect. The diffusion layer heat transfer coefficient (HTC) is strongly affected by thermal stratification, whereas no effect is observed on the film HTC. This leads to the total condensation HTC at the top of the heat exchanger is approximately 3–5 times than that at the bottom in the first 100 s of the containment pressure dropping process. To improve the overall heat transfer performance of PCCS, approaches that enhance the heat transfer capacity at the bottom of the heat exchanger should be considered for prioritization, especially the methods of steam diffusion heat transfer capacity.

Multicomponent gas mixture parametric CFD study of condensation heat transfer in small modular reactor system safety

Safety is always the primary concern for designing and analyzing nuclear reactor systems. The requirements for the safety margin for advanced small modular reactor (SMR) systems are targeted even higher than the conventional commercial large-scale nuclear reactors incorporating the passive and inherent safety systems. The SMR systems are designed with the condensation passive containment cooling system (PCCS), which plays a critical role in removing reactor heat during a steam release accident case. However, the presence of non-condensable gas (NCG), like air, reduces the heat transfer performance. This physics phenomenon becomes multifactorial for nuclear reactor containment during a fuel failure accident case that releases hydrogen gas. Besides, the mixture component of steam-air-hydrogen varies in reactor accident cases, which needs simulation and validation keeping parameters of importance. Reviews showed that previous studies for SMR’s PCCS did not cover the condensation heat transfer (CHT) in the presence of multicomponent NCG mixture parametric computational fluid dynamics (CFD) simulation and validation, making a research gap in the SMR design safety. A comprehensive CHT parametric CFD study was performed for SMR PCCS to fill this research gap. This study used experimental data as simulation 3D physics domain inlet and outlet boundary conditions. However, the wall boundary conditions were constant temperature, curve-fit, and annular coolant for verifying the turbulence models. Parametric simulations were performed, verified, and optimized for steam-NCG mixtures. The multicomponent gases, multiphase mixtures, and fluid film condensation models were applied with associated turbulence models. The results of the parametric study were evaluated for realistic reactor conditions. Results showed that parametric study provided critical insight about the dependency of multicomponent gas mixture parameters that supports reactor safety design, analysis, and licensing.

The combination method of functional failure and device fault for passive safety system in nuclear power plant

Functional failure is one of the important causes for passive system failure, and some methods are developed for functional failure evaluation based on the thermal–hydraulic (T-H) model. However, the system equipment configuration is the prerequisite of the T-H behavior analysis and setting in the model, and the system configuration and the functional failure have one-to-one logical relationship which should be considered in the passive system reliability evaluation. In this paper, a logical combination method is put forward based on Fault Tree (FT) tool, and the simplification way is also proposed to improve the calculation efficiency. The Passive Containment Cooling System (PCCS) in AP1000 is analyzed as an example, the result shows that the functional failure under the configuration of one or more of the three parallel lines in valve gallery are available is the main contributor to the system failure, for the plant site in a subtropical area.