Natural Circulation Phenomena Research Articles

Transient Characteristics and Performance of a Dual Compensation Chamber Loop Heat Pipe under Various Heat Loads

The dual compensation chamber loop heat pipe (DCCLHP) technology shows great promise for efficient thermal management in a variety of applications. This study focuses on refining the startup performance of the DCCLHP, particularly for low heat load conditions in terrestrial settings, by utilizing dual bayonet tubes. Empirical investigations examined the DCCLHP’s operational characteristics and improved thermal transfer capabilities across different orientations of the evaporator and compensation chamber (CC), including vertical, 45‐degree tilt angle, and horizontal configurations. Key results demonstrate the DCCLHP’s ability to manage low heat loads across diverse orientations, achieving over 400 W of heat transfer across a distance of 2.0 m while maintaining stable operation. Notably, the DCCLHP operation did not exhibit any significant instability issues. This work enhances the startup performance of the DCCLHP in various orientations and provides valuable insights into the confined natural circulation phenomenon, crucial for improving the overall thermal management capabilities of the system. In summary, the study highlights advancements in DCCLHP design and its potential to address thermal management challenges in diverse applications, particularly in the aerospace industry, through improved startup performance and enhanced heat transfer capabilities.

Comparative study of CFD and 3D thermal-hydraulic system codes in predicting natural convection and thermal stratification phenomena in an experimental facility

Natural circulation phenomena have been nowadays largely revisited aiming to investigate the performances of passive safety systems in carrying-out heat removal under accidental conditions. For this purpose, assessment studies using CFD (Computational Fluid Dynamics) and also 3D thermal-hydraulic system codes are considered at different levels of the design and safety demonstration issues. However, these tools have not being extensively validated for specific natural circulation flow regimes involving flow mixing, temperature stratification, flow recirculation and instabilities.In the present study, an experimental test case based on a small-scale pool test rig experiment performed by Korea Atomic Energy Research Institute, is considered for code-to-code and code-to-experimental data comparison. The test simulation is carried out using the FLUENT and the 3D thermal-hydraulic system CATHARE-2 codes. The objective is to evaluate and compare their prediction capabilities with respect to the test conditions of the experiment. It was observed that, notwithstanding their numerical and modelling differences, similar agreement results are obtained. Nevertheless, additional investigations efforts are still needed for a better representation of the considered phenomena.

Preliminary validation of the 1D modeling of the DYNASTY natural circulation loop against results from water experimental campaign

In the continuous improvements of the safety of nuclear power plants, the study of natural circulation is of primary importance for developing passive safety systems and Generation-IV reactors, such as the Molten Salt Fast Reactor (MSFR). In particular, removing the decay heat during the shutdown of the MSFR involves dealing with a circulating fuel driven by natural circulation; as such, the development and validation of natural circulation models are of primary importance. To validate models and provide experimental data on natural circulation with a distributed heat source, the DYNASTY (DYnamics of NAtural circulation for molten SalT internallY heated) natural circulation loop has been built on the premises of Politecnico di Milano. This paper focuses on improving the 1D modeling of natural circulation phenomena in the DYNASTY facility and validating them against the results (also reported in this work) found in the first experimental campaign, carried out with water as the working fluid. The modeling of DYNASTY is based on a 1D approach using DYMOLA®, a modeling environment based on the MODELICA simulation language. The models used are improved versions of the ones available in the literature, which better model the experimental facility as they now include the mass flow rate meter, the heat loss between the facility and the environment, and a more realistic cooler. The improved model was then used to study the outcomes of the model simulations and verify the influence of the different numerical integration algorithms on the simulation results. The first DYNASTY experimental campaign provides in-depth testing of its behavior, with different cooling fan speeds and considering all possible heating configurations. The collected results for the heating and cooling transients are the mass flow rate and four fluid temperatures measured by thermocouples placed at the beginning and the end of the four DYNASTY legs. The experimental results were compared with the model simulations, showing the good predicting capabilities of the latter one following the tuning of the heat exchange model. In particular, the comparison showed an excellent match between the model and the experimental results for the heating transient while showing improving margins for the cooling transient.

Design of large-scale sodium thermal-hydraulic integral effect test facility, STELLA-2

The STELLA program was launched to support the PGSFR development in 2012 and for the 2nd stage, the STELLA-2 facility was designed to investigate the integral effect of safety systems including the comprehensive interaction among PHTS, IHTS and DHRS. In STELLA-2, the long-term transient behavior after accidents can be observed and the overall safety aspect can also be evaluated. In this paper, the basic design concept from engineering basis to specific design is described. The design was aimed to meet similarity criteria and requirements based on various non-dimensional numbers and the result satisfied the key features to explain the reasoning of safety evaluation. The result of this study was used to construct the facility and the experiment is on-going. In general, the final design meets the similarity criteria of the multi-dimensional physics inside the reactor pool. And also, for the conservation of natural circulation phenomena, the design meets the similarity requirements of geometry and thermo-dynamic behavior.

Investigation into different 1D/3D co-simulation methodologies applied to a natural circulation loop

1D/3D co-simulation is a promising approach for the modelling and simulation of complex systems. These systems are typically computationally intensive. By combining the desired features of a one-dimensional (1D) systems code and a three-dimensional (3D) Computational Fluid Dynamics (CFD) code, the required resources are optimized whilst still retaining an acceptable level of accuracy. Co-simulation has been applied in various domains and scenarios, often without a comprehensive study on the impact to simulation- stability, -time, and results. The purpose of this study is to explore the behaviour of some of the strategies by applying them to the simulation of the natural circulation occurring in a simplified Reactor Cavity Cooling System (RCCS). Four different co-simulation approaches were explored to investigate the radiative- and convective- heat transfer as well as the natural circulation phenomena occurring in a RCCS facility. Typically, in these four modelling approaches a 3D methodology was employed to model the radiation and convection heat transfer in the cavity between the heater wall and the outer wall of the riser. The 1D methodology was employed to model the remainder of the RCCS system. The 1D systems code, Flownex, and the 3D CFD code Ansys Fluent were used to perform the 1D/3D co-simulations. Each of the codes modelled a relevant part of the RCCS facility and the models were coupled by explicitly exchanging boundary conditions or relevant data between the 1D and 3D regions. The validity of the coupled 1D/3D approaches was investigated by comparing them with a full 1D simulation.

Analisa karakteristik aliran pada fenomena sirkulasi alam fasa-tunggal di Untai FASSIP-01 Mod.2

FASSIP-01 rectangular loop is a facility which owned by PTKRN-BATAN that used for observe the phenomenon of natural circulation in large-scale. This natural circulation phenomenon will be applied as a passive cooling system in nuclear power plant that will be built up in Indonesia. Some FASSIP-01 experiments have done before, but those experiments show that the fluid flow in FASSIP-01 is not in optimal conditions because there are some errors happened and causes some losses. The second modification of FASSIP-01 rectangular loop has purpose to minimize losses happened in previous modification of FASSIP-01. The purpose of this research is to know the characteristics of flow and the trend of the data from the flow in FASSIP-01 mod.2 rectangular loop before do experiments. Simulations is done by using 2-dimensional images and varying the heating temperature at 353,15 K and 363,15 K and the cooler at temperatures of 278.15 K, 283.15 K, 288.15 K and 293.15 K. Based on the simulation results it is known that the velocity flow rate will increase if the temperature in the heating tank is increased, but will decrease if the temperature in the cooling tank is increased. The results of the calculation of the Reynold Number and Nusselt Number based on the velocity flow rate are 2973,668 the lowest and 5775,356 the highest and for the Nusselt number obtained 17,36 and the highest is 28,091.

Appliance of scaling methodologies for designing counterpart experiments dominated by natural circulation phenomena

The utility of the Integral Test Facilities (ITF) for the study of thermohydraulic phenomena that can occur in nuclear power plants has been demonstrated. However, direct extrapolation of data between different scales is challenged due to the uncertainty and distortion that implies the application of scaling laws. In this sense, the analyses of counterpart experiments help to clarify the capabilities and limitations of scaling methodologies.This work focuses on the design and study of a counterpart experiment in the LSTF facility based on the ATLAS A1.1 test. The transient at issue reproduces a SBO-type scenario with the asymmetric and delayed supply of auxiliary feedwater. The scenario in ATLAS and LSTF facilities is simulated with the TRACE5 thermal-hydraulic code and the results are compared. Then, a global system scaling analysis is performed and the calculation of dimensionless groups reveals, through an analytical approach, the relevance of the transfer processes (mass, enthalpy and so on) in the evolution of the transient and the scaling distortion between both facilities.This study assesses a great similarity in the evolution of the main thermal-hydraulic phenomena throughout the experiment, despite a significant discrepancy is evidenced in the chronology. Likewise, the results support the novel procedure to design the counterpart test.

Investigation of two-phase natural circulation with the SMART-ITL facility for an integral type reactor

A two-phase natural circulation test using SMART integral test loop (SMART–ITL) was conducted to explore thermo-hydraulic phenomena of two-phase natural circulation in the SMART reactor. Specifically, the test examined the natural circulation in the primary loop under a stepwise coolant inventory loss while keeping the core power constant at 5% of the scaled full power. Based on the test results, three flow regimes were observed: single-phase natural circulation (SPNC), two-phase natural circulation (TPNC), and boiler–condenser natural circulation (BCNC). The flow rate remained steady in the SPNC, slightly increased in the TPNC, and dropped abruptly and maintained in the BCNC. Using a natural circulation flow map, the natural circulation characteristic in the SMART–ITL was compared with those in pressurized water reactor simulators. In the SMART–ITL, a BCNC regime appeared instead of siphon condensation and reflux condensation regimes because of the use of once-through steam generators.

Analysis of steam generator tube rupture accident for OPR 1000 nuclear power plant

Evaluation of radionuclide releases to the environment from operating nuclear power plants has increased in importance in the Republic of Korea following the 2015 amendment to its Nuclear Safety Act that includes not only design basis accidents but also severe accident management. Containment bypass accidents are among the crucial scenarios that should be considered under the new circumstances due to their high consequences, even though the probability of such accidents is low. In this study, analysis of a thermally induced steam generator tube rupture (TI-SGTR) accident caused by a severe accident has been conducted with MELCOR. To increase the reliability of the analysis results, qualification processes were performed in the process of producing the input. In addition, computational fluid dynamics analysis was performed to simulate the natural circulation phenomena in the reactor coolant system (RCS) during the accident and major flow parameters were obtained for the target plant. Thermal-hydraulic and fission product behavior was found during the accident for the base case scenario, and uncertainty band for CsOH release fraction to the environment were presented through the uncertainty analysis on the natural circulation flow parameters. Sensitivity calculations were also performed on uncertain parameters such as the number of ruptured tubes and the depressurization time of the secondary system, and the CsOH release fraction to the environment without passing through the containment were evaluated.

CATHARE simulation results of the natural circulation characterisation test of the PKL test facility

In the past, several experimental investigations aiming at characterizing the natural circulation (NC) behavior in test facilities were carried out. They showed a variety of flow patterns characterized by an inverted U-shape of the NC flow curve versus primary mass inventory. On the other hand, attempts to reproduce such curves using thermal-hydraulic system codes, showed 10–30% differences between the measured and calculated NC mass flow rate. Actually, the used computer codes are generally based upon nodalization using single U-tube representation. Such model may not allow getting accurate simulation of most of the NC phenomena occurring during such tests (like flow redistribution and flow reversal in some SG U-tubes). Simulations based on multi-U-tubes model, showed better agreement with the overall behavior, but remain unable to predict NC phenomena taking place in the steam generator (SG) during the experiment.In the current study, the CATHARE code is considered in order to assess a NC characterization test performed in the four loops PKL facility. For this purpose, four different SG nodalizations including, single and multi-U-tubes, 1D and 3D SG inlet/outlet zones are considered. In general, it is shown that the 1D and 3D models exhibit similar prediction results up to a certain point of the rising part of the inverted U-shape of the NC flow curve. After that, the results bifurcate with, on the one hand, a tendency of the 1D models to over-predict the measured NC mass flow rate and on the other hand, a tendency of the 3D models to under-predict the NC flow rate.

STH/CFD coupled calculations of postulated transients from mixed to natural circulation conditions in the NACIE-UP facility

In the present paper, coupled STH/CFD calculations are performed addressing two operating conditions in the frame of the NACIE-UP experimental campaign. In order to simulate mixed and natural circulation phenomena related to liquid metals thermal hydraulics, the RELAP5/Mod3.3 STH code and the CFD code ANSYS Fluent are adopted in an integrated manner. The combined use of STH and CFD codes aims to help the user overcoming the limits of the individual techniques and thus achieving a better predicting capability with respect to each stand-alone application.Measured values and calculated results are compared. In particular, the coupled application provides the best results in terms of predicted mass flow rate and comprehensiveness of information giving the possibility to analyse the local and cross-sectional distributions. Furthermore, as an additional beneficial aspect, CFD results can be used as a tuning tool for some of the STH code parameters. The aim of the present work is highlighting the advantages and drawbacks of the adopted techniques, also reporting the margins of error that may presently be expected from these applications. Though perfectly matching results were not obtained, coupled STH/CFD application report promising modelling capabilities and may be consequently considered as a sound candidate for addressing liquid metal thermal hydraulics in complex geometries and large systems.

Closed-loop experimental investigation of single-phase natural circulation flow phenomena based on temperature and heating power variations

Experimental investigation of single-phase, low pressure natural circulation flow phenomena has been carried out. The study aims to critically analyze the thermal hydraulic parameters (temperature and heating power) variations as it affects transfer of heat in a closed-loop rectangular natura l circulation facility having a single heated channel at predetermined inlet sub cooled temperature varied from 40 to 80 °C. Heating power varies between 7 kW−13 kW to portray several conditions of the reactor power in response to temperature changes while sustaining single-phase flow circulation through the hydraulic loop. Real-time signals of thermal hydraulic responses are observed via a graphical interface for data acquisition when stable flow or self-sustained flow oscillations are achieved. Results obtained were analyzed at several sections of the hydraulic circuit which are basically pressure drops and flow rate due to temperature differences and heating power increments. These results and observations made through them thus contribute to the database which is important for characterizing natural circulation phenomena and acquisition of dominants parameters for the scaling analysis of future designs.

Single-phase natural circulation in a rectangular loop with a finite capacity heat sink positioned below the heater section

One of the passive safety systems employed in nuclear reactors is a large water pool positioned at an elevation, which acts as a heat sink to remove decay heat from the reactor core in the event of an accident using the natural circulation phenomena. A large mass of water at an elevated position has concerns regarding the safety of support structure, and therefore an alternative methodology where the heat sink is positioned just below the heater section is investigated in this study. Results of an experimental and numerical investigation for the thermal-hydraulic behavior of single-phase natural circulation in a rectangular loop of height 4300 mm and width 1900 mm with the heat sink located at an elevation just below the heater section is reported here. The heat sink was a finite one, a constant volume insulated tank filled with water, whereas the heat source was a constant heat flux type. The loop was constructed with circular pipe elements with a nominal inner diameter equal to 40 mm. The heater section was 1m in height and three input power values equal to 1.15 kW, 1.55 kW, and 2 kW were used in the study. The flow behavior in a single and twin parallel heater system with a single riser was studied. The loop fluid is initially in a single-phase and becomes two-phase as the heat sink temperature rises, but only the results for the loop fluid in the single-phase regime are presented in the current study. The measured mass flow rate, after an initial transient, settles down to values which slowly decreases with time. The average temperature of the heat sink is shown to rise continuously with increasing time, demonstrating the effectiveness of the heat sink in removing the energy input to the working fluid in the heater section. However, the heat sink removes only 60–70% of the input heat while the rest is either lost to the ambient or utilized to raise the enthalpy of the working fluid.The Nusselt numbers in the heater section are also presented which steadily rise after an initial transient. In addition, the values are much higher than those for the corresponding Reynolds number values in forced convection due to the influence of buoyancy-induced convection. The Nusselt number is well correlated by the expression Nu‾=1.35(Ra*‾)0.3 for 2×105≤Ra*‾≤6.5×105and 600≤Re≤1000. Numerical predictions for loop flow rate, temperature variation, and sink temperature using the RELAP5/MOD3.2 code for the experimental configuration of the current study are presented which compare well with the measurements the maximum deviation being less than 10%. Numerical results for pressurizing the working fluid and permitting boiling in the heat sink to drive the system towards steady-state operation are also presented.

Conceptual design of small modular reactor driven by natural circulation and study of design characteristics using CFD & RELAP5 code

A detailed computational fluid dynamics (CFD) simulation analysis model was developed using ANSYS CFX 16.1 and analyzed to simulate the basic design and internal flow characteristics of a 180 MW small modular reactor (SMR) with a natural circulation flow system. To analyze the natural circulation phenomena without a pump for the initial flow generation inside the reactor, the flow characteristics were evaluated for each output assuming various initial powers relative to the critical condition. The eddy phenomenon and the flow imbalance phenomenon at each output were confirmed, and a flow leveling structure under the core was proposed for an optimization of the internal natural circulation flow. In the steady-state analysis, the temperature distribution and heat transfer speed at each position considering an increase in the output power of the core were calculated, and the conceptual design of the SMR had a sufficient thermal margin (31.4 K). A transient model with the output ranging from 0% to 100% was analyzed, and the obtained values were close to the Thot and Tcold temperature difference value estimated in the conceptual design of the SMR. The K-factor was calculated from the flow analysis data of the CFX model and applied to an analysis model in RELAP5/MOD3.3, the optimal analysis system code for nuclear power plants. The CFX analysis results and RELAP analysis results were evaluated in terms of the internal flow characteristics per core output. The two codes, which model the same nuclear power plant, have different flow analysis schemes but can be used complementarily. In particular, it will be useful to carry out detailed studies of the timing of the steam generator intervention when an SMR is activated. The thermal and hydraulic characteristics of the models that applied porous media to the core & steam generators and the models that embodied the entire detail shape were compared and analyzed. Although there were differences in the ability to analyze detailed flow characteristics at some low powers, it was confirmed that there was no significant difference in the thermal hydraulic characteristics' analysis of the SMR system's conceptual design.

Natural circulation systems in nuclear reactors: advantages and challenges

Following the Fukushima disaster in Japan in March 2011, a number of questions and doubts about the safety of nuclear reactors have been raised globally and many nations have begun revising and improving the nuclear installations. Natural circulation plays a major role in nuclear accident. This significance of the natural circulation-based systems led us to study the natural circulation phenomena for better understanding and in order to ensure the safety and efficiency of natural circulation systems. This review focuses on the advantages and challenges of natural circulation systems.

A multi-dimensional dataset for two-phase instability in low pressure natural circulation based on direct transient local measurement

Thermal hydraulic instability can challenge the control and the safety of nuclear reactors. This experimental research captures and investigates the instability in a low-pressure (1–10 atmospheres) natural circulation loop with an adiabatic chimney installed above the heated section. Direct transient local measurement of two-phase flow is performed at multiple axial locations along the flow channel with thermocouples and conductivity probes traversed by customized high-precision mechanisms. Based on the periodicity of oscillations, a post-processing method is proposed to align the data captured non-simultaneously to reconstruct the instability behavior in one representative period. The averaging operation over multiple oscillation cycles is performed to reduce the random errors. Cyclic phenomena are clearly demonstrated by the post-processed data, and rapid void fraction change induced by the flashing is observed to be significant in the driving mechanism for low pressure oscillations. Parametric effects of inlet subcooling, heat flux, inlet flow restriction and system pressure are studied within the experiment range, and qualitative results are explained and compared with existing literature. The ratio between the oscillation period and the travelling time of the fluid particle through the test section is compared with the theoretical value for the density wave oscillation. The new dataset contains directly measured transient multi-dimensional two-phase data along the flow channel for natural circulation under density wave oscillations. The generated dataset is valuable for the validation of system analysis codes and CFD codes in the simulation of natural circulation phenomena and dynamic two-phase flow.

Flow instability analysis of natural circulation under ERVC condition for SMR

In this work, the internal natural circulation flow of ERVC system in SMR is calculated by the thermal-hydraulic code RELAP5. Both the transient characteristics and the flow instability are analyzed in detail to reveal the flow mechanism. Firstly, to verify the applicability of RELAP5 models, multiple cases of steady-state natural circulation with low heating power are simulated by comparing with the REPEC experiment. Next, a sliced model of ERVC loop is built to obtain the transient phenomena of natural circulation. As the circulation loop has no cold source, the inlet temperature of the heating section will keep rising to transform the flow patterns into stable-oscillating-stable in turns. During this transition, the flow is dominated by the subcooled boiling of heating section, and the mass flux is about 60–160 kg/m2·s. Based on the features of flow oscillation, the unstable flow patterns can be further divided into low-frequency and high-frequency oscillations. Both of them belong to the density wave oscillation, and their trajectory diagrams can present the attributes of the limit cycle. Finally, in the plane formed with the inlet subcooling (horizontal axis) and heating power (vertical axis), the boundaries of flow instability are divided. The instability domain change with the effects of heating power, back pressure, inlet resistance coefficient, and rising section height is discussed. The results indicate that increasing the back pressure will compress the instability domain. Reducing the inlet resistance coefficient will shift the instability boundaries towards the direction of power increase. When lengthening the height of the rising section, the flow can be gradually dominated by the flashing with the decrease of inlet subcooling, which may also introduce another instability.

Experimental and Numerical Simulation Investigation of Single-Phase Natural Circulation in a Large Scale Rectangular Loop

In order to anticipate station blackout, the use of safety system based on passive features is highly considered in advanced nuclear power plant designs, especially after the Fukushima Dai-ichi nuclear power station accident. An example is the application of natural circulation in the emergency cooling system. To study the reliability of such an application, a research project on natural circulation was carried out. This paper describes the investigation results on the natural circulation phenomena obtained using a large rectangular experimental loop named FASSIP-01. The experiments were conducted at two different heat source powers. The experimental results are analysed using existing correlation and numerical model simulation. The RELAP5 system code is applied to model the natural circulation. FLUENT computational fluid dynamic code is used to visualize the flow distribution. The experimental results show the establishment of stable natural circulation in all heat power input with the mass flow rate of about 0.0012 kg/s. Calculation using the existing correlation shows that the experimental Reynold numbers are lower than predicted by the correlation. The computational fluid dynamics-based tool could show the three dimensional distribution of the temperature, while the model of RELAP5 predict well the dynamic of the single-phase natural circulation established in the experimental loop. It is concluded that the stable natural circulation have been established in the large rectangular loop and the model of the RELAP5 could simulate the observed natural circulation phenomenon reasonably well. /* Style Definitions */ table.MsoNormalTable {mso-style-name:"Table Normal"; mso-tstyle-rowband-size:0; mso-tstyle-colband-size:0; mso-style-noshow:yes; mso-style-priority:99; mso-style-qformat:yes; mso-style-parent:""; mso-padding-alt:0cm 5.4pt 0cm 5.4pt; mso-para-margin:0cm; mso-para-margin-bottom:.0001pt; mso-pagination:widow-orphan; font-size:11.0pt; font-family:"Calibri","sans-serif"; mso-ascii-font-family:Calibri; mso-ascii-theme-font:minor-latin; mso-fareast-font-family:"Times New Roman"; mso-fareast-theme-font:minor-fareast; mso-hansi-font-family:Calibri; mso-hansi-theme-font:minor-latin; mso-bidi-font-family:"Times New Roman"; mso-bidi-theme-font:minor-bidi;}

SMART 원자로 자연대류 현상 모의 개별효과실험 및 해석

SMART 원자로 자연대류 현상 모의 개별효과실험 및 해석

STUDI AWAL KARAKTERISTIK TEMPERATUR DAN TEKANAN PADA ALIRAN SIRKULASI ALAM DI FASILITAS UJI UNTAI FASSIP-02 MENGGUNAKAN SIMULASI CFD

TFASSIP-02 loop is a test facility that is used for research and development of safety technologies for future nuclear power plants based on natural law. This test facility is designed to study natural circulation phenomena caused by differences in fluid density due to temperature differences in the one-phase heat dissipation system during the simulation of heat removal from the reactor core when an accident occurs. FASSIP-02 loop consists of piping components, water heating tanks, water cooling tanks and expansion tanks. The purpose of this study was to understand the conditions of temperature change and pressure of water working fluid based on temperature changes in the heater section which were simulated on the loop geometry FASSIP-02. The research method was carried out in a simulation of Computational Fluid Dynamics using FLUENT 6.3 software. The working fluid in the FASSIP-02 loop uses water with a temperature of 27°C, the flow rate is varied 0.3 m/s and 0.45 m / s, while the temperature in the heating section is 70°C. CFD simulation results show that the increase in the working fluid temperature of the water with a flow rate of 0.3 m/s after passing through the heating section is 39°C, while the temperature increase of the working fluid of the water with a flow rate of 0.45 m/s is 36.6°C. Pressure drops at flow rates of 0.3 m/s and 0.45 m/s each occur in water working fluid before entering through WHT and after passing through the heating section.