Two-phase Natural Circulation Flow Research Articles

Assessment of MARS-KS prediction capability for natural circulation flow in passive heat removal system

Considering that system analysis codes are used for the evaluation of the performance of Passive Safety Systems (PSSs), it is important to investigate the capability of the system analysis code to reliably predict the heat transfer and natural circulation flow, which are the main phenomena governing the performance of a PSS. Since MARS-KS has been widely validated for heat transfer models, this study focuses on evaluating its capability to predict the single and two-phase pressure drops and natural circulation flow. The straight pipe simulation results indicate that the pressure drop predictions are reliable within ±5 % error margin for the single-phase flow and the errors of pressure drop up to −30 % for the two-phase flow. Through single-phase natural circulation flow analysis, it is concluded that the use of the appropriate K-factor modeling based on the flow regimes is important since the natural circulation flow rate in MARS-KS is mainly affected by the form loss factor modeling. With two-phase natural circulation flow analysis, this study emphasizes the behavior of the system could change significantly depending on the two-phase wall friction and pressure loss modeling. With the analysis results, modeling considerations for the PSS performance evaluation with the system analysis codes are proposed.

Mechanisms and inhibition method investigation on condensation induced water hammer in an equal-height-difference passive heat removal system

The equal-height-difference passive heat removal system utilizes the ocean as an ultimate heat sink for residual heat removal, demonstrating obvious advantages for ocean nuclear power plant applications. However, the structure with a zero-height difference can trigger the reverse flow of subcooled water. This could cause serious condensation induced water hammer events. In this study, a fully visual experimental setup was established to investigate the occurrence features. First, two kinds of formation mechanisms were defined by the visual experimental results. In this regard, the subcooling and the isolated steam slug are the necessary conditions for the formation and development. In addition, the occurrence positions were found to be concentrated in the test pipe section near the water tank. Then, a butterfly check valve was designed to avoid the condensation induced water hammer events based on the formation mechanisms. The experimental results showed that the check valve usage can effectively inhibit the reverse flow of the subcooled water and thus achieve its purpose of eliminating condensation induced water hammer events. Moreover, the check valve transforms the two-phase natural circulation flow from the condensation oscillation mode to the stable boiling two-phase flow mode. As a result, it enhances the two-phase natural circulation capacity.

Experimental study on transient flow characteristics in an equal-height-difference passive heat removal system for ocean nuclear power plants

The equal-height-difference passive heat removal system (EHDPHRS) utilizes the ocean as the heat sink for infinite heat dissipation and has broad application prospects in ocean nuclear power plants. However, the equal height difference between the heat source and heat sink is susceptible to severe flow instability. In this study, a fully visualized experimental facility was constructed to analyze the transient flow characteristics in the EHDPHRS. The experimental results revealed three typical two-phase natural circulation flow modes in the EHDPHRS. The formation mechanism of each flow mode was analyzed using thermal–hydraulic parameters. The results revealed that the flow mode is determined by the steam–water direct contact condensation and reverse flow of subcooled water, and both factors promote each other. Moreover, condensation-induced water hammer (CIWH) was found to be the most severe flow instability in the EHDPHRS. CIWH events can lead to violent pressure peaks and reduce the natural circulation capacity. In addition, when the occurrence position of CIWH events is concentrated at the pipe outlet, the influence of CIWH events on the two-phase natural circulation flow capacity gradually decreases. These research results aid in the optimal design of the EHDPHRS and prevention and control of CIWH events.

Numerical Evaluation of Coolability Limits of External Reactor Vessel Cooling Using an Improved Thermal-Hydraulic System Analysis Code

The external reactor vessel cooling provides a means to stabilize the molten core inside the reactor vessel during the severe accident in nuclear power plants. For a credible evaluation of the chances of success of in-vessel melt retention, one requires high-fidelity predictions on the natural circulation flow around the reactor vessel and, in particular, the corresponding critical heat flux (CHF) on a hemispherical lower head. In this study, the coolability limit of external reactor vessel cooling in the Korean APR1400 reactor was investigated numerically by improving a thermal-hydraulic system analysis code, MARS-KS1.5. Since such system analysis codes did not incorporate a CHF model for a downward-facing hemisphere, three CHF correlations were newly implemented in the MARS-KS code, two of them taking into account the effect of the mass flux. A transient analysis on the CHF and surface heat flux on the lower head was carried out while the surrounding cooling water was heated to create the two-phase natural circulation flow. The heat transfer mode on the heated surface was used as a measure of evaluating the success of the strategy, and the simulation results were compared with those obtained via the default CHF model in the code. The sensitivity analysis on the heat load from the core melt was also performed.

ONE-DIMENSIONAL MIXTURE MODEL CODE FOR THE MODELING OF TWO-PHASE FLOW THERMOSIPHON

The numerous coupled physical phenomena involved in natural circulation two-phase flow loops make their analytical or numerical modelization difficult. A 1D mixture model code has been developed for the sizing of semi-closed thermosiphon loops. The results of the code have been compared to existing experimental data of several two-phase flow thermosiphons. A reasonable agreement was found between the experimental and computed mass fluxes. Additionally, a sensitivity analysis was conducted on the pipe diameter, the imposed pressure, and the position of the condenser. The model was finally used to size an experimental thermosiphon loop.

A mathematic model of two-phase flow taking account of sub-cooled boiling, condensation, and void flashing

Abstract Based on a physical model of the test loop HRTL-5 simulating the 5 MW nuclear heating reactor, a drift-flux model of two-phase flow has been derived, in which the sub-cooled boiling and saturated boiling in the heated section, condensation and void flashing in the adiabatic riser etc, are taken into account. It is suitable to investigate the natural circulation two-phase flow, especially at heating reactor conditions. By introducing the concept of condensation boundary layer, a condensation formula has been derived and complemented into the drift-flux model, which was used to investigate the behavior of the HRTL-5 in our previous research works. Based on the mathematical model, a computer program has been developed. At first, through comparing the calculations with the experimental results, the parameters in the two-phase drift-flux model have been determined. Then calculations on the flow characteristics of HRTL-5 have been performed, including the simulation of the mass flow rate and the distribution of various physical variables. At last, the characteristic curves of natural circulation have been simulated; thereby, also useful experiences have been acquired.

Operator Action–Induced Two-Phase Flow Condition Resulting in Performance Degradation of Interfacing Passive System

This study investigates the degradation of the heat transfer performance of a closed-circuit intermediate natural circulation heat transport loop used as a passive safety system in a nuclear power plant (NPP). The degradation arises from the strong thermal-hydraulic (TH) coupling of the loop operating characteristics, saturation temperature and pressure, and natural circulation flow rate, which determine the heat rejection rate to the TH boundary conditions imposed on the hot side of the loop by the transitory state of the primary reactor coolant system (RCS) of the NPP. Several operator actions related to a feed-and-bleed emergency operating procedure (F&B) are postulated, and system TH code simulations are performed to demonstrate how the F&B can induce two-phase flow conditions in the RCS. Natural circulation two-phase flow regimes in the RCS hot leg can significantly reduce the heat transfer to the circulating working fluid of the interfacing heat transport loop over long periods, sometimes lasting over 24 h, of passive system mission time. A transient performance indicator for the passive system mission is introduced for use in the passive reliability assessment and quantitative comparison of transient simulations. The need to consider human factors in the design and operation of NPP passive safety systems is stressed.

Validation and Uncertainty Quantification for Two-Phase Natural Circulation Flows Using TRACE Code

The work presents validation of the TRAC/RELAP Advanced Computational Engine (TRACE) code for natural circulation two-phase flow in a vertical annulus. Natural circulation experiments were recently conducted for a vertical internally heated annulus at the Multiphase Thermo-Fluid Dynamics Laboratory at the University of Illinois. The experimental matrix consists of 107 experiments with system pressure in the range of 145 to 950 kPa and heat flux up to 275 kW/m2. Void fraction, gas velocity, and interfacial area concentration were measured in five axial locations along the test section with six measurements of bulk liquid temperature and pressure. To validate the capability of the TRACE code under natural circulation flow conditions, a complete model of the experimental facility was created and validated using forced convection and single-phase natural circulation data. Sensitivity and uncertainty quantification were performed. The sensitivity to important simulation parameters was studied using Sobol’s variance decomposition and the Morris screening method. The sensitivity of boundary conditions on void fraction measurement was investigated. The sensitivity study has shown significant differences in model sensitivity between different experimental conditions. With heat flux being the most influential parameter for high-pressure cases without flashing and pressure, temperature and heat flux have a combined strong effect in the case of low-pressure experiments when flashing occurs. Additionally, higher uncertainty in void fraction prediction was observed for experimental conditions at low pressure with flashing.

Natural circulation two-phase flow characteristics in a rod bundle channel under stable and rolling motion conditions

Natural circulation two-phase flow characteristics are studied under stable and rolling motion conditions in a rod bundle channel. The mass quality, flow rate and frictional pressure drop variation results with increasing heat flux are given and the effects of system pressure and inlet subcooling are analyzed. The flow rate decline is observed under high mass quality conditions since the frictional pressure drop is large enough. The flow instability characteristics for different stages with increasing heat flux are given. The main reason for the flow instability characteristics variation is that the dominant factor changes from subcooled boiling to the pressure drop oscillation. The experimental results indicate that increasing the system pressure and the inlet resistance of the test section, including moving the compressible volume to a downstream position, can delay the onset of the flow instability. The time-average flow parameters under rolling motion conditions are mainly influenced by the stable mass quality and the rolling parameters. For high stable mass quality conditions, the time-average two-phase flow parameters decrease a little compared to stable results. With the stable mass quality decreasing, the time-average mass quality gradually increases with the rolling frequency. For further lower stable mass quality conditions, the time-average values are even larger than stable values. For the frictional resistance calculation, a few correlations are evaluated and recommended. Additionally, new correlations are fitted for stable and time-average two-phase frictional pressure drop prediction in high mass quality conditions.

Impact of dissolved salts on two-phase flow and boiling heat transfer in a natural circulation loop

Two-phase natural circulation and boiling experiments are conducted in this study with artificial seawater as a heat transfer fluid or process coolant. This experimental work is focused on understanding the role of dissolved salts in water on the steady state two-phase natural circulation flow rates. Mass flow rates were measured at different heat flux levels with artificial seawater, deionized water and tap water in the experimental loop. It was found that the average quasi-steady state mass flow rates were the same at any particular heat flux level regardless of the fluid being used. Artificial seawater coolant exhibited more stable natural circulation at higher heat flux as compared to the other two coolants. Mass flow measurements recorded at 2 s intervals show that magnitude of flow oscillations are much higher in case of tap water and deionized water. Also, high speed camera images revealed that the bubble departure diameters are smaller in seawater compared to tap or deionized water. This can be attributed to the increased local wettability due to the presence of dissolved salts in the fluid. Visual observations of the riser section showed that bubble coalescence and formation of slugs is inhibited in the seawater. These observations also explain the reduced pressure drop across test section when seawater was used as a coolant.

An experimental study on natural circulation flow characteristics under passive IVR-ERVC conditions

In vessel retention (IVR) of molten core debris via water cooling at the external surface of the reactor vessel is an important severe accident management feature of advanced passive plants. During postulated severe accidents, the heat generated by the molten debris relocated to the lower head of reactor pressure vessel needs to be removed continuously to prevent vessel failure.Besides the local critical heat flux (CHF) of outer wall surface which is of the first importance, a stable feature of natural circulation flow and an effective natural circulation capability within the external reactor vessel cooling (ERVC) channel tend to be rather crucial for the success of IVR. Under this circumstance, a full-height ERVC test infrastructure, reactor pressure vessel external cooling II test facility (REPEC-II) was designed and constructed in Shanghai Jiao Tong University (SJTU). During test campaign on the REPEC II facility, the one-dimensional natural circulation boiling flow characteristics with the experimental observation and measurement are investigated under IVR-ERVC conditions.Based on the abundant results acquired in the test campaign, it is attempt to evaluate the ERVC performances and trends under various practical engineered conditions in this paper. From this study, the following conclusions can be made: The influence of different power shapes on the two-phase natural circulation flow is not obvious, while the impact on the resistance characteristics of the outlet section is very significant. Among all the test conditions (power shapes, subcooling, flow path configuration and water level), the resistance characteristics of the outlet section change the most compared with other sections. It is expected that these conclusions may help designers to have a reliable estimate of the impact of some related engineered factors on real IVR-ERVC performance.

Numerical Study on Two-phase Natural Circulation Flow by External Reactor Vessel Cooling of iPOWER

Numerical Study on Two-phase Natural Circulation Flow by External Reactor Vessel Cooling of iPOWER

Experimental study on natural circulation two-phase flow resistance in a vertical 3 × 3 rod bundle channel

Natural circulation boiling flow resistance characteristics in a 3 × 3 rod bundle channel are experimentally studied in this paper. In order to achieve a stable two-phase natural circulation flow, the boiling flow phases are identified and measures such as decreasing the inlet fluid subcooling degree and increasing the inlet local resistance are adopted. The ONB during the experiment is determined by the wall temperature drop. The variation of pressure drop components, including the gravity pressure drop, the frictional pressure drop and the acceleration pressure drop, is given. For high mass quality conditions, the mass flow rate is observed declining with the increasing heat flux when the frictional pressure drop and the acceleration pressure drop grow faster than the gravity pressure drop reduction. A few homogeneous flow models and separated flow models for the two-phase frictional pressure drop are compared against experimental results. The homogeneous flow models can well predict the frictional pressure drops for low mass quality conditions and the Duckler model predictions are the most accurate with a mean relative error of 15.7%. In terms of the separated flow models, the Mishima and Sun models predictions are in good agreement with experimental results respectively for low and high mass quality conditions. Based on the Chisholm correlation, a new correlation is fitted by experimental results. The friction factors obtained in single-phase experiments are used to calculate the liquid phase frictional pressure drops. The comparison against experimental results and data from other literatures shows the new correlation can well predict the two-phase frictional pressure drops under given flow conditions.

The comprehensive analysis of coolability limits of passive external reactor vessel cooling under in-vessel retention

The induced two-phase natural circulation flow and the critical heat flux (CHF) on the outer surface of lower head are crucial factors for the assessment of the external reactor vessel cooling (ERVC) under in-vessel retention which is considered as an effective management strategy to mitigate the core degradation accident. In this work combined with the two-phase drift flux model, an analysis program for the induced two-phase circulation flow has been established based on one-dimensional steady-state mass, momentum and energy conservation equations. The calculation results are compared with the simulation results using RELAP5 within the same geometric parameters to verify the rationality and correctness of the analysis program. Combining the analysis program with the SULTAN CHF correlation, the characteristics of induced two-phase flow and the CHF on the outer surface of lower head are studied comprehensively to analyze the coolability limits of ERVC under in-vessel retention. The effects of concerned parameters including subcooling of water, decay power of molten corium, flooding levels, height of riser channel, loss coefficient at the inlet and gap clearance between the reactor pressure vessel and thermal insulation are studied.

Classification of natural circulation two-phase flow image patterns based on self-organizing maps of full frame DCT coefficients

Many of the recent nuclear power plant projects use natural circulation as heat removal mechanism. The accuracy of heat transfer parameters estimation has been improved through models that require precise prediction of two-phase flow pattern transitions. Image patterns of natural circulation instabilities were used to construct an automated classification system based on Self-Organizing Maps (SOMs). The system is used to investigate the more appropriate image features to obtain classification success. An efficient automated classification system based on image features can enable better and faster experimental procedures on two-phase flow phenomena studies. A comparison with a previous fuzzy inference study was foreseen to obtain classification power improvements.In the present work, frequency domain image features were used to characterize three different natural circulation two-phase flow instability stages to serve as input to a SOM clustering algorithm. Full-Frame Discrete Cosine Transform (FFDCT) coefficients were obtained for 32 image samples for each instability stage and were organized as input database for SOM training. A systematic training/test methodology was used to verify the classification method. Image database was obtained from two-phase flow experiments performed on the Natural Circulation Facility (NCF) at Instituto de Pesquisas Energéticas e Nucleares (IPEN/CNEN), Brazil.A mean right classification rate of 88.75% was obtained for SOMs trained with 50% of database. A mean right classificationrate of 93.98% was obtained for SOMs trained with 75% of data. These mean rates were obtained through 1000 different randomly sampled training data. FFDCT proved to be a very efficient and compact image feature to improve image-based classification systems. Fuzzy inference showed to be more flexible and able to adapt to simpler statistical features from only one image profile. FFDCT features resulted in more precise results when applied to a SOM neural network, though had to be applied to the full original grayscale matrix for all flow images to be classified.

Numerical Investigation on the Natural Circulation Thermal Hydraulic Characteristics of a Pressure Tube Type Boiling Water Reactor

Natural Circulation Boiling Water Reactors (BWRs) are susceptible to boiling two-phase flow instabilities under certain conditions, which can lead to flow oscillations in the reactors. These oscillations could be in-phase or out-of-phase in nature depending on the geometry and operating conditions of the system.This paper reports on a study on the effect of both thermal hydraulics as well as neutron kinetics on the characteristics of boiling two-phase natural circulation flow instabilities in a pressure tube type natural circulation BWR. RELAP5/MOD3.2 code has been used to simulate the natural circulation behavior in multiple channels of the reactor. Before applying the RELAP5 model for simulation of natural circulation in this reactor, the code was benchmarked with the experiment conducted in a multichannel boiling natural circulation loop, having geometry similar to this reactor. The results showed that the RELAP5 model is able to capture the boiling induced-flow instabilities. Then the model was applied to simulate the reactor behavior. The prediction showed that the reactor could experience both Type-I and Type-II density wave oscillations depending on the channel inlet subcooling and channel power. Unlike Type-II instability wherein clear cut outs of phase oscillations among multiple channels were observed, in Type-I instability it was observed that mixed mode oscillations could be present, especially at low subcooling. The phase difference among the channels were found to change with time in Type-I instability. These are completely new findings with regard to characteristics of boiling two-phase natural circulation.The fuel in this reactor is a combination of (Th-233U)O2 and (Th-Pu)O2, which is different from conventional BWRs. Also, the coolant (light water) is present in different pressure tubes which are physically separated from moderator (heavy water). The effect of neutronic feedback due to this fuel and geometrical configuration on characteristics of Type-I and Type-II instabilities has not been investigated before. In view of this, a systematic investigation was done to study the effect of neutronic feedback on Type-I and Type-II oscillations observed in this reactor. The simulation showed that the threshold power for both Type-I and Type-II instability slightly stabilizes with introduction of neutronic feedback. Since the magnitude of void reactivity feedback is very small in the present fuel composition, the stability boundary was only slightly altered with the introduction of neutronic feedback. Regarding the oscillation characteristics, it was found that change in magnitude of void reactivity has almost no effect on Type-I oscillations whereas the Type-II oscillations get stabilized when void reactivity magnitude was increased. This kind of effect due to void reactivity feedback is in contrast to findings based on conventional BWR and is an important finding of the present study.

Performance evaluation of controllable separate heat pipes

A controllable separate heat pipe (CSHP) combined with cool storage can be an efficient and low-cost solution to release the necessary cool power and to improve the precision in the temperature control of cool-storage refrigerators powered by solar energy or electricity with time-of-use price policy. However, the precise temperature control performance of separate heat pipes under active control depends on system design and thus should be confirmed. In this study, a CSHP was developed and an active control method was achieved by interrupting its two-phase natural circulation flow. A one-dimensional steady-state model was used to calculate the critical fill ratio, heat transfer rate, and average working temperature of CSHP. A series of experiments was conducted to analyze the start–stop and heat transfer performances of CSHP at various R134a fill ratios and heat sink temperatures. Under different test conditions, the optimum working points were obtained through fitting. The most suitable control mode was determined and the start–stop performance of CSHP was examined in detail. CSHP starts up quickly in approximately 15 seconds and the quickest stopping time of CSHP approximates to 80 seconds. Therefore, the performance of CSHP under active control can be applied to cool-storage refrigerators.

Automation of RELAP5 input calibration and code validation using genetic algorithm

Validation of system thermal-hydraulic codes is an important step in application of the codes to reactor safety analysis. The goal of the validation process is to determine how well a code can represent physical reality. This is achieved by comparing predicted and experimental system response quantities (SRQs) taking into account experimental and modelling uncertainties. Parameters which are required for the code input but not measured directly in the experiment can become an important source of uncertainty in the code validation process. Quantification of such parameters is often called input calibration. Calibration and uncertainty quantification may become challenging tasks when the number of calibrated input parameters and SRQs is large and dependencies between them are complex. If only engineering judgment is employed in the process, the outcome can be prone to so called “user effects”.The goal of this work is to develop an automated approach to input calibration and RELAP5 code validation against data on two-phase natural circulation flow instability. Multiple SRQs are used in both calibration and validation.In the input calibration, we used genetic algorithm (GA), a heuristic global optimization method, in order to minimize the discrepancy between experimental and simulation data by identifying optimal combinations of uncertain input parameters in the calibration process. We demonstrate the importance of the proper selection of SRQs and respective normalization and weighting factors in the fitness function.In the code validation, we used maximum flow rate as the SRQ of primary interest. The ranges of the input parameter were defined based on the experimental data and results of the calibration process. Then GA was used in order to identify combinations of the uncertain input parameters that provide maximum deviation of code prediction results from the experimental data. Such approach provides a conservative estimate of the possible discrepancy between the code result and the experimental data.

Detailed evaluation of two phase natural circulation flow in the cooling channel of the ex-vessel core catcher for EU-APR1400

The ex-vessel core catcher of the PECS (Passive Ex-vessel corium retaining and Cooling System) is installed to retain and cool down the corium in the reactor cavity of the EU (European Union)-APR (Advanced Power Reactor) 1400. A verification experiment on the cooling capability of the PECS has been conducted in the CE (Cooling Experiment)-PECS. Simulations of a two-phase natural circulation flow using the RELAP5/MOD3 computer code in the CE-PECS and PECS have been conducted to predict the two-phase flow characteristics, to determine the natural circulation mass flow rate in the cooling channel, and to evaluate the scaling in the experimental design of the CE-PECS. Particularly from a comparative study of the prototype PECS and the scaled test facility of the CE-PECS, the orifice loss coefficient in the CE-PECS was found to be 6 to maintain the coolant circulation mass flux, which is approximately 273.1kg/m2s. The RELAP5 results on the coolant circulation mass flow rate are very similar to the CE-PECS experimental results. An increase in the coolant injection temperature and the heat flux lead to an increase in the coolant circulation mass flow rate. In the base case simulation, a lot of vapor was injected into the downcomer, which leads to an instability of the two-phase natural circulation flow. A super-step design at a downcomer inlet is suitable to prevent vapor injection into the downcomer piping.

Static bifurcation analysis of natural circulation inner evaporative cooling system in wind turbine

Wind power is one of the most attractive renewable clean energies under development at present. On a global scale, wind power generation development was very rapid in recent years. As the wind power generation tends to develop toward large-scale and offshore, the traditional cooling methods gradually expose their own shortcomings. As the large wind turbine installation tower is high, and the installation site is dispersed, the installation and maintenance of wind turbine generator are difficult. So, the generator is required to have a small weight and less maintenance. Self-circulation inner evaporative cooling system (SCIECS) has the following advantages: self-circulation without pump, high cooling efficiency, safe and reliable operation, and basically maintain-free, etc. The self-circulation of cooling system can be realized by the 35 between wind turbine generator and the horizontal direction, and it is very suitable for being used in a large-scale wind power generator. Owing to the intrinsic nonlinearity of two-phase self-circulation system, changes of operation condition and circuit topology in a large range may lead to an instability of the cooling system, causing the system parameters to severly change. The instability of the cooling system can cause the local overheating and even burning of the generator, which provides a huge security risk for the cooling system, thus threatening the safe and stable operation of the generator. The stability of SCIECS is very important for the safe operation of wind turbine. In this paper, static stability of the SCIECS in wind power generator is studied based on the nonlinear bifurcation analysis theory and its numerical continuation method. System static bifurcation diagram is obtained to analyze the evolution characteristics of the SCIECS. Parameter effect of the system static bifurcation is analyzed at the same time. In order to verify the theoretical prediction of the static bifurcation of the small-angle two-phase natural circulation, an experimental platform is built. Static bifurcation of the SCIECS is observed experimentally. The experimental results show that the static bifurcation phenomenon exists in the natural circulation two-phase flow of small angle, and the theoretically predicted m-Q bifurcation curves are in good agreement with the experimental curves, which verifies the correctness of the theoretical analysis.