Two-phase Natural Circulation Research Articles

Numerical simulation on in-vessel molten corium behavior with external vessel cooling using smoothed particle hydrodynamics

The in-vessel retention through external reactor vessel cooling (IVR-ERVC) strategy is a key management strategy for early termination of a nuclear severe accident that can threaten the integrity of the reactor vessel. To simulate the physical phenomena of the molten corium, the smoothed particle hydrodynamic (SPH) method is utilized in this study. The SPH method is a Lagrangian computational fluid dynamic (CFD) method that can simulate multi-fluid stratification, turbulence, natural circulation, radiative heat transfer, thermal ablation, and crust formation. To address the external vessel cooling, it is coupled with a conventional 1-D nuclear system analysis method. The 1-D system analysis code can calculate the two-phase natural circulation of cooling water and the convective heat transfer on the external reactor vessel wall. These two simulation codes exchange the temperature and heat flux of the reactor vessel outer wall. This study numerically simulated the IVR-ERVC strategy for a Korean high-power reactor and compared it with the traditional lumped parameter method (LPM). Unlike LPM, this study provides localized detailed data about the thermal hydraulic behavior of molten corium and visualization of phenomena in the IVR-ERVC strategy. This enhances our understanding of the phenomena in IVR-ERVC strategy and introduces new perspectives.

Thermal-hydraulic phenomena and heat removal performance of a passive containment cooling system according to exit loss coefficient

The natural circulation system has been widely studied for use in various applications because of its inherent advantage. However, it has a key weakness called flow instability that makes the system unstable. Through massive previous research, the mechanisms of flow instability were analyzed, but there was an ambiguous aspect related to the effect of experimental parameters on the phenomenon. Particularly, there has been no report on the heat transfer performance of the system when flow instability phenomena were present. In this study, thermal-hydraulic phenomena of a two-phase natural circulation system that functions as a passive containment cooling system (PCCS) was investigated according to experimental parameters, namely, the temperature boundary (120–158 °C) and exit loss coefficient (0–34.5) under atmospheric pressure conditions. The experimental results showed five different flow types in the loop. The flow modes that occurred by the interaction between flashing and boiling were classified by referring to the mass flow rate, void fraction, and visualization data. The system was more unstable when the temperature boundary conditions increased, but it was more stable when the exit loss coefficient increased. These results have only been confirmed in our research. The reason for the results is that the flow conditions are located on the boundary between Density Wave Oscillation I and the stable flow region, and that boundary does not have clear criteria. In addition, comparing the heat transfer performance of a system by heat rate can confirm the effect of flow instability on the thermal performance of the passive cooling system. As a result, the high exit loss coefficient stabilizes the system better than the low case and has similar heat removal performance.

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.

Validation of coupled ATHLET-OpenFOAM simulation on a large-scale single- and two-phase flow experiment

Abstract System thermal-hydraulics (STH) codes are successfully used in the last decades for design and analysis of the physical behavior of nuclear power plants (NPPs). These programs provide reliable results at comparatively low computational cost, but have limited capabilities to predict complex 3D flow phenomena, which is possible with Computational Fluid Dynamics (CFD) tools. In order to combine the advantages of STH and CFD codes, algorithms to couple these are developed. At GRS, a new OpenFOAM solver, based on the volume of fluid (VOF) method, was developed together with new coupling boundary conditions to couple the CFD program OpenFOAM with the system code ATHLET. In a next step, verification and validation activities were carried out. These included the analysis of a Pressurized Thermal Shock (PTS) scenario. For this purpose, data from the OECD/NEA ROSA V Test 1.1, carried out in the Japanese Large-Scale Test Facility (LSTF) was used. The experiment investigated the temperature stratification after emergency core cooling (ECC) injection under single- and two-phase natural circulation conditions in the primary cooling circuit of a pressurized water reactor (PWR). This paper discusses the results of the coupled simulations, compares the numerical results with experimental data, and identifies areas for future improvements.

Experimental investigation of thermal characteristics on a new loop pipe model for passive cooling system

The heat pipes have the potential to be used as a passive cooling system (PCS) in thermal installations including nuclear facilities. In that context, this experimental study is correlated with the research on the new type of Loop Heat Pipe (LHP) as the PCS. This work aimed to study the characteristics of LHP with capillary pipe wick. The experiments involved varying the pool water temperature as evaporator heat load at 35, 45, 55, and 65 °C, filling ratio charged at 20, 40, 60, 80, and 100 % of LHP evaporator volume, and the flow air velocity to the fins in the condenser at 0, 1, 1.5, 2, and 2.5 m/s. The LHP was operated in sub-atmospheric conditions with an initial pressure of 2666.4 Pa. The results show that LHP has a stable two-phase natural circulation under all variations of heat loads, filling ratios, and air velocities tested. The results also show that display excellent thermal performance with the lowest thermal resistance of 0.0524 ± 0.0004 °C/W operating at the highest water temperature source, the highest filling ratio, and the highest air velocity. It is concluded that the LHP model with capillary pipe wick has similar thermal characteristics to other types of heat pipes. This wick function effectively in preventing vapor flow from the evaporator toward the condenser, and otherwise it can facilitate as condensate path from the condenser to evaporator. The stable two-phase natural circulation stability analysis showed that the LHP with capillary wick has very good thermal performance as heat absorber and heat releaser.

Design and construction of fluid-to-fluid scaled-down small modular reactor platform: As a testbed for the nuclear-based hydrogen production

This paper presents the construction results and design of the UNIST Reactor Innovation platform for small modular reactors as a versatile testbed for exploring innovative technologies. The platform uses simulant fluids to simulate the thermal-hydraulic behavior of a reference small modular reactor design, allowing for cost-effective design modifications. Scaling analysis results for single and two-phase natural circulation flows are outlined based on the three-level scaling methodology. The platform's capability to simulate natural circulation behavior was validated through performance calculations using the 1-D system thermal-hydraulic code-based calculation. The strategies for evaluating cutting-edge technologies, such as the integration of a solid oxide electrolysis cell for hydrogen production into a small modular reactor, are presented. To overcome experimental limitations, the hardware-in-the-loop technique is proposed as an alternative, enabling real-time simulation of physical phenomena that cannot be implemented within the experimental facility's hardware. Overall, the proposed versatile innovation platform is expected to provide valuable insights for advancing research in the field of small modular reactors and nuclear-based hydrogen production.

Research on characteristic of flow instability in a two-phase natural circulation loop with parallel once-through steam generators

When there are two sets of parallel once through-steam generators(OTSG) in the natural circulation loop, the thermal-hydraulic characteristics are complicated. Flow instability often occurs in such loops. In order to alleviate the harm caused by flow instability and enhance the natural circulation capacity, it is necessary to study the characteristics of flow instability in depth. In this paper, a thermal-hydraulic computational model is developed for a two-phase natural circulation loop with two parallel OTSGs based on the RELAP5 program. And the model is validated. The thermal-hydraulic parameters of the loop were calculated for different decay heat conditions. The flow instability of the loop is found to be characterized by a clear division into three zones in the decay heat interval. The concept of “load-flow ratio” is proposed to analyze the critical point of flow instability under different operating conditions. The reason for the flow instability zones of the loop is explained. The conclusions of the relevant research can provide a basis for an in-depth study of the characteristics of flow instability and the development of operational schemes to mitigate flow instability.

PWR natural circulation at a small break LOCA

In this paper, a one-dimensional quasi-steady model has been developed to analyse the potential impacts of reduced primary coolant inventory on natural circulation. This model focuses on the development of expressions for the coolant inventory, temperature distributions, and mass flow rate. The model encompasses a range of fluid flow scenarios, including single-phase, combined single-and-two-phase, and two-phase natural circulation, as well as the phenomenon of reflux condensation. The model is also utilized to examine the impact of reactor core power on natural circulation parameters. Comparisons with experimental findings and earlier research are made, highlighting a satisfactory level of agreement with the analytical findings. Based on the findings of this study, it has been observed that within the range of primary coolant inventory of 100–60% and core decay power of 1.5–5% of full power, natural circulation in a PWR type system has the potential to serve as an effective mechanism for transferring core decay heat to the secondary loop. The flow rates of natural circulation in pressurized water reactors (PWRs) are observed to have a significant dependence onthe amount ofliquid inventory, whereas their relationship with power level is comparatively less pronounced. The flow rate of a pressurized water reactor (PWR) system tends to exhibit its highest value within a specific range of inventory, typically falling between 60% and 80%. The transition between the different modes of natural circulation is observed to exhibit a smooth transition, with varying system inventories.

Analysis of coupled two-phase natural circulation loops by developing HEM and DFM based numerical models

The coupled two-phase natural circulation loop based passive systems find numerous applications, particularly in nuclear reactors where safety is paramount. However, the loops coupling, system conditions (specific to the reactor type), and inherent coupled loops feedback make their dynamics more complex. Hence, the models used should be capable of predicting such intricate coupled dynamics for precise forecasting of passive systems performance. With this objective, the present analyses deliberate the transients of such coupled two-phase loops linked to coupling configurations and operating conditions. Generalized two-phase models based on Homogeneous Equilibrium (HEM) and Drift Flux (DFM) with flashing, wall and fluid conduction, water-steam property formulation, etc., are developed and employed. Validation studies substantiate the prediction capability of these models for various cooler-heater orientations. Subsequently, predicted the start-up transients of vertically and horizontally coupled two-phase loops. The analyses revealed that coupling configurations, heater-cooler orientations and system conditions significantly influence the coupled loops. Further, the influences of various geometrical and operating parameters on the dynamics of coupled loops are investigated. It elucidated the differences in the parameter's effect on single-phase and two-phase coupled loops. The study also identified the vital parameters strongly influencing the coupled loops based on sensitivity analysis. The models and findings will be applied to evaluate the performance of coupled loop-based passive safety systems in nuclear reactors.

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.

Analysis of the heat-carrying performance of natural circulation loop using modified RELAP5/MOD3.4

Since the successful manufacture of the separate heat pipe (also known as Two-phase Loop Thermosyphon or Natural Circulation Loop), it has been used in many applications, especially for waste heat recovery/remove and cooling systems. We use the modified system analysis program RELAP5 to study the medium-sized separated heat pipe setup MSL2 built by Seok-Ho Rhi. To better calculate heat pipe performance under low pressure, we have improved a series of models, including the bubble nucleation model, the heat flux partition model, the interfacial heat transfer, the drift flux model, the wall heat transfer model, etc. Then we use the low-pressure subcooled boiling experiment to verify the modified code. The validation results show that the modified code accurately captures the subcooling boiling phenomenon under low pressure and low flow conditions. We use the modified code to investigate the factors influencing heat-carrying performance and analyze their action mechanisms. The results show that heating power, volumetric filling rate, and cooling water temperature influence heat-carrying performance. Additionally, we analyze the heat pipe operating parameters over time to obtain the characteristics of the heat pipe at various stages.

Predicting reverse flow characteristics in equal-height-difference passive heat removal system for ocean nuclear power

The equal-height-difference passive heat removal system (EHDPHRS) via natural circulation driving and utilizes ocean as infinite heat sink for heat dissipation, showing broad application prospect in ocean nuclear power field; but the heat removal pipe normally encounters the reverse flow of cool seawater into the pipe, leading to operation instability and even condensation-induced water hammer. In this study, the integration of volume of fluid model and condensation model based on surface renewal theory is used to simulate the reverse flow characteristics in the heat removal pipe under single-and two-phase natural circulation conditions. The interfacial heat and mass transfer are calculated by user-defined functions (UDF). The numerical results reveal that the buoyancy force (Fb) and condensation driving force (FΔP) both accelerate the reverse flow, while the inertial force (Fi) restrains the reverse flow; the reverse flow length presents monotonous decrease with increasing the flow velocity under single-phase conditions and parabolic-like trend for two-phase flow. A prediction correlation for the reverse flow length is developed using dimensionless Reynolds number, Grashof number and vapor volume fraction, covering single- and two-phase conditions. The predicted reverse flow length by the correlation agrees with 94% of the results within deviations of ±15%.

Effect of geometrical and operating parameters on performance of a rectangular two-phase natural circulation loop

Natural Circulation Loops (NCLs) find application in a wide range of engineering devices and it is often necessary to predict and ensure the performance of such loops under various operating conditions. In this paper analytical expressions are presented in non-dimensional form for steady-state mass flow rate in typical rectangular, two-phase natural circulation loops having Vertical-Heater Vertical-Cooler (VHVC) and Vertical-Heater Horizontal-Cooler (VHHC) configurations. The normalized loop mass flow rates are expressed in terms of several geometrical and operating variables such as, normalized heat input, density ratio, volume/diameter of heater and cooler sections, aspect ratio of loop, inlet subcooling, fractional length of heater and cooler sections, two-phase friction multiplier and the loss coefficient at heater inlet. The analytical mass flow rate expressions are first validated using published data for various loops and shown to reasonably capture the gravity and friction dominated regimes in two-phase pipe flow. Parametric studies are then conducted to elucidate the effect of key variables on steady-state loop mass flow rates. For the chosen geometrical configuration of the loops, the possibility of existence of multiple steady-state solutions is checked for the entire range of values used in this study for each variable. The discussions in this paper are limited to explaining the influence of individual variables by comparing with a base case analysis.

Behaviors of type-I instabilities in natural circulation with a heating channel under ocean conditions

Characteristics of two-phase natural circulation and density wave instability with a single-heating channel under ocean conditions were studied by using PNCMC PROGRAM (Program for Natural Circulation under Motion Condition). The superposition laws of oscillations of both mass flow rate and density wave caused by ocean conditions are found, and the corresponding stability boundaries of the natural circulation system are also proposed here. The use of a simplified mathematical description of fluid flow to describe the additional motions with waves reduces the difficulty of analysis. Specifically speaking, based on the study of stability boundary in vertical condition, first of all, it was found that inclined ocean condition can weaken the ability of natural circulation and can move down stability boundary of natural circulation, and the greater the inclination angle, the greater the amount of going-down. Secondly, the stability boundary of the system moves up in heaving circumstances, and the smaller the period, the greater the amount of going-up. Thirdly, the amplitude of rolling plays an important role in the rolling situation. In the case of rolling with large amplitude, the period of flow complex oscillation is clear, which is half the rolling period. However, in the case of rolling with a short period, the period of flow complex oscillation is ambiguous, the oscillation frequency of density wave under vertical condition is preserved in the flow complex oscillation. The stability boundary of natural circulation is moved down by rolling with a short period, where the frequency of flow complex oscillation was equal to that of density wave oscillation under vertical conditions. In contrast, the stability boundary of natural circulation is moved up by rolling with a long period, where the short-period local oscillation appears in the process of flow complex oscillation. Finally, the oscillation superpositions of rolling and heaving motion will result in a complex long-period oscillation with motion amplitudes being important.

Experimental Study on Separate Heat Pipe-type Passive Residual Heat Removal System of Swimming Pool-type Low-Temperature Heating Reactor

The use of separate heat pipe in Passive Residual Heat Removal System (PRHRS) for Swimming Pool-type Low-Temperature Heating Reactor (SPLTHR) has a good application prospect. An experimental study on miniaturized separate heat pipe-type PRHRS has been carried out. The heat transfer medium in the separate heat pipe is water and is at sub-atmospheric. The boundary conditions of the experiment refer to the operating range of SPLTHR. The two-phase natural circulation characteristics of the separate heat pipe-type PRHRS and the influence of boundary conditions on the system heat transfer capacity, flow instability and heat stratification of pool water are studied. Within the experimental range, the heat transfer rate is from 72.0 kW to 459.2 kW and the steam mass flowrate is from 0.031 kg/s to 0.197 kg/s. During the operation of the system, there is long-term flow instability. The process and causes of system instability are analyzed from the perspective of heat transfer. It is believed that the natural convective heat transfer outside the tube bundle hinders the heat transfer, resulting in instability. The heat transfer capacity of the system under three filling ratios of 30%, 50% and 70% was studied, and the results showed that the filling ratio within this variation range had little effect on the heat transfer capacity of the system. Furthermore, the thermal stratification in the pool has been analyzed.

Analytical study of flow instability behaviour in a boiling two-phase natural circulation loop under low quality conditions

Abstract Analytical investigations have been carried out to study the flow instability behaviour in a boiling two-phase natural circulation loop under low quality conditions. For this purpose, the computer code TINFLO-S has been developed. The code solves the conservation equations of mass, momentum and energy and equation of state for homogeneous equilibrium two-phase flow using linear analytical technique. The results of the code have been validated with the experimental data of the loop for both the steady state and stability. The study reveals that the stability behaviour of low quality flow oscillations is different from that of the high quality flow oscillations. The instability reduces with increase in power and throttling at the inlet of the heater. The instability first increases and then reduces with increase in pressure at any subcooling. The effects of diameter of riser pipe, heater and the height of the riser on this instability are also investigated.

Experimental investigation on the heat removal capacity of secondary side passive residual heat removal system for an integrated reactor

An experimental facility named Intermediate Heat Exchange Loop (IHEL) is built to simulate the secondary side passive residual heat removal system. The IHEL mainly includes a helically coiled Once-Through Steam Generator (OTSG), a condenser, a cooling water tank, and associated piping and valves. The steam is produced by the water-to-water heat exchange method in the SG secondary side. The heat from the SG primary side is carried to the cooling water tank by the IHEL through a two-phase natural circulation method. The natural circulation characteristics of the IHEL under the low pressure (nearly atmospheric pressure) and low water inventory conditions are investigated. The result shows that the heat removal capacity of the IHEL is good at these conditions, and it is determined by the parameters of the SG primary side and IHEL together. In the present test conditions, for the parameters of the SG primary side, the heat removal rate will increase with the increase of mass flow rate or heat source temperature. For the parameters of the IHEL, the heat removal rate is not sensitive to the initial non-condensable gas content. The heat removal rate is not sensitive to the resistance level when the heat source temperature is low, while it will decrease with the increase of resistance level when the heat source temperature is high. A very meaningful conclusion is that the natural circulation has been established successfully in the very lower water inventory condition (L = 14.8% in the present study), and the heat removal rate almost remains constant with the increase of the water inventory, which indicates the IHEL is robust and is beneficial for the system design.

CFD investigation of flow reversal in inverted U-tube steam generator under two-phase natural circulation

In this paper, a critical mass flow rate (CMFR) has been obtained for three different steam generators (i.e. two marine-type and one commercial) at different inlet void fractions using the CFD method to study the sustainability of natural circulation (NC) due to depressurization in two-phase NC mode. Because of depressurization, the transients in the inlet void fraction have been considered as multiple steady-state inputs, and the range of safe operating mass flow rate has been obtained. The results show that the characteristics curves shift with inlet void fraction thus limiting the operating mass flow rate range between 0.046 kg/s to 0.050 kg/s for M_SG1 and 0.051 kg/s to 0.055 kg/s for M_SG2. The effect of pipe roughness on CMFR has also been studied. The results can be used for the optimized design of the U-tube steam generator (UTSG) that will safely perform the heat removal operation during a small break loss of coolant accident (SBLOCA).