Characteristics Of Natural Circulation Research Articles

Simulation study on the dynamic characteristics of natural circulation coupled phase change material for equal height difference passive containment cooling system

The passive containment cooling system (PCCS) plays a significant role in ensuring the safety of nuclear power plants. To further reduce peak pressure in the containment during a loss of coolant accident (LOCA), phase change materials (PCMs) are utilized to improve the system’s cooling effectiveness. The dynamic simulation model of PCCS has been developed using Apros software. This article focuses on examining the system’s dynamic characteristics related to various tube startup methods and comparatively analyzing the impact of different PCM thicknesses on heat transfer capacity. The results indicate that post-accident, PCM significantly suppresses peak pressure in the containment. Under the full tube startup method with 4 mm PCM, the peak pressure decreases to 1.75 MPa, whereas under the empty tube startup method, it is reduced to 1.69 MPa. With natural circulation empty tube startup method, PCM can reduce pressure change intensity in the containment but negatively affects natural circulation heat transfer. Although the full tube startup method underperforms the empty tube startup method on suppressing peak pressure, PCM can positively affect natural circulation heat transfer in the early stages of the accident. Empty tube startup offers the advantages of higher heat transfer capacity and lighter weight, making it the optimal choice for ocean nuclear platforms. Additionally, the higher peak power of PCM heat transfer under the empty tube startup method indicates that system can withstand larger thermal shocks.

Dynamic scaling characteristics of single-phase natural circulation based on different strain transformations

Dynamic scaling characteristics of single-phase natural circulation based on different strain transformations

Numerical analysis of passive safety injection driven by natural circulation in floating nuclear power plant

The traditional energy supply paradigm has been difficult to meet the demands of ocean resources development. As a combination of small nuclear reactors and offshore platforms, the floating nuclear power plant(FNPP) can be well used in offshore energy supply. Considering passive safety injection of FNPP will be deeply affected by the ocean environment, the present study brings a CFD customization model to assess the injection process. The additional inertial force is implanted into the momentum equation to simulate motion conditions. The combination of coordinate transformation and grid traversal is used to output critical parameters. Therefore, it can be seen the overall evolution process and the local flow characteristics of natural circulation intuitively apart from quantitative analysis, making up for the deficiency of the previous analysis code. Results suggested the injection process can be divided into several stages. Affected by ocean conditions, the flow rate in FNPP will decrease with periodic fluctuation after a linear increase, so the safety injection can be completed in 180 s under static conditions while needing more than 240 s under rolling. However, although additional acceleration weakens the driving force of natural circulation, liquid sloshing and disturbed flow will make the reactor core cool more effectively.

Numerical Study of Natural Circulation Flow in Reactor Coolant System during a Severe Accident

The rupturing of steam generator tubes leads to serious accidents in nuclear power plants. It causes radioactive materials to leak into the secondary system and release outside the reactor containment region. Therefore, it is important to model a technique to determine whether the natural circulation within a reactor coolant system (RCS) can cause rupture. In this study, a computational fluid dynamics (CFD) analysis methodology was incorporated as a first step to establish an RCS natural circulation evaluation technique to generate RCS natural circulation input parameters for the MELCOR analysis of thermally induced steam generator tube rupture (TI-SGTR) in nuclear power plants. Benchmarking tests were conducted against existing experimental studies; the results demonstrated a difference of 9.4% or less between the experimental and CFD analysis results with respect to the main evaluation factors. Subsequently, a steam generator tube simplification modeling technique was established for application to nuclear power plants, and CFD analysis was conducted to determine its applicability. The CFD analysis results revealed that when numerous tubes are simplified into one equivalent tube, the thermal flow characteristics generated in the RCS could be distorted. The findings of this research are expected to be helpful in understanding the thermal flow characteristics of natural circulation in the RCS. Further, the findings may potentially serve as a foundation for future CFD analysis research related to the natural circulation in the RCS of nuclear power plants.

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 Research for CHF Sensitivity of Heat Flux Distribution under IVR Conditions

In-vessel retention (IVR) through external reactor vessel cooling (ERVC) is one of the most effective severe accident mitigation measures in the nuclear power plants. The most influential issues on the IVR strategy are in-vessel core melt evolution, the heat fluxes imposed on the lower head, and the external cooling of reactor pressurized vessel (RPV). In the molten pool research, there are mainly two different molten pool configurations: two layers and three layers. Based on the different distributions of heat flux in molten pool configurations, a new problem was raised: whether the in-vessel heat flux distribution will affect the CHF on the outer wall of RPV and further affect the effectiveness of IVR measures? A full-height external reactor vessel cooling and natural circulating facility was conducted to study the CHF sensitivity of different heat flux distributions. The experimental results show that the characteristics of natural circulation are similar and the CHF of the RPV lower head external surface is not obviously affected under the different heat flux distributions. The varying heat flux distribution during severe accident process will not threaten significantly the success of IVR strategy.

Experimental study on fine particle deposition in supercritical water natural circulation

An experimental platform for the movement and deposition of fine particles in critical water was developed. The platform is equipped with a viewport to visually observe the movement of fluid and particles, which successfully solves the difficult problems such as easy leakage of viewport and inability of viewport materials to withstand ultra-high temperature and pressure. Through experiments and calculations, it is found that the physical properties of supercritical water and the characteristics of natural circulation are more likely to cause the deposition of finer particles. On the wall of the vertical pipe and the wall of the visible section, there are a few particles deposited due to the strong diffusion coefficient and high dielectric coefficient of supercritical water. At the bend, owing to the inertial centrifugal force, together with the temperature gradient and turbulence, a large number of particles are deposited at the bottom of the bend. In the horizontal pipeline, due to the direct action of gravity on the wall surface, together with a higher temperature gradient and turbulence force, more particulate matter will be deposited.

Experimental study on heat transfer characteristics of supercritical carbon dioxide natural circulation

An experimental study has been conducted to investigate the heat transfer characteristics of supercritical carbon dioxide (sCO2) uniformly heated in the horizontal circular smooth tube. The results illustrated that there was a significant difference in heat transfer between the top wall and bottom wall due to the buoyancy. Bulk flow acceleration cannot be negligible in the high heat flux region, which leads to heat transfer deterioration. A new heat transfer correlation is proposed, in which the buoyancy parameter and bulk flow acceleration have been taken into account. The new correlation and six classic correlations for sCO2 are examined in horizontal tubes. The comparison indicates that the new correlation has a better performance for sCO2 flowing through a horizontal heating tube under natural circulation conditions. For example, 94.9% of the calculated results using the new heat transfer correlation were within ±30% of the experimental results while only 87.9% of that using the Jackson correlation (the best of the six) were within the same error bands.

Experimental study on transient thermal–hydraulic characteristics of an open natural circulation for the passive containment cooling system

Nuclear safety has attracted increasing global attention. Passive Containment Cooling System (PCCS) is one of the several passive safety systems designed to ensure the safety of nuclear power plants (NPPs). A large-scale test facility was built to simulate the PCCS for understanding the thermal–hydraulic characteristics with operating conditions in the start-up process. Parameter analysis is carried out to reveal the influences of various parameters on the thermal–hydraulic characteristics. Results show that the transient-state performance of the PCCS strongly depends on the initial containment pressure and air mass fraction. The air mass fraction has a greater effect on the transient heat transfer process. Meanwhile, the effect of cooling water temperature on the transient characteristics can be ignored. The response time, which is defined to characteristic the heat transfer ability of PCCS, is proposed. In addition, a non-dimensional empirical correlation for response time is developed to reveal the relationship between the heat transfer capacity of the natural circulation and forced circulation, while the deviation mostly within ±20%. Studies on the relevant physical processes or phenomena are helpful for the safety analysis or accident study related with PCCS.

Modeling and validation of RELAP5 for natural circulation flow in a single PWR fuel assembly

The natural circulation loops have been widely used in the design of next generation nuclear reactors and spent fuel storage facilities. However, the friction correlations for fuel bundle and grid spacers in RELAP5 code are not applicable for natural circulation flow. This paper summarizes the modeling methods in previous researches, and proposed a flow resistance model for rod bundle and grid spacers based on experimental data. A RELAP modeling method of radiation heat transfer in fuel assembly is proposed for natural circulation flow of air in the fuel assembly. The new model is verified using experimental data from the Spent Fuel Project by NRC. The results of the new model are in good agreement with the experimental data. The current model will be useful for analyzing the flow and heat transfer characteristics of natural circulation in the reactor core of next generation nuclear reactors and spent fuel storage facilities.

Analysis on CHF deviation caused by geometry of the heater blocks in a validation test of IVR-ERVC technology

The innovative technology of In-Vessel Retention – External Reactor Vessel Cooling (IVR-ERVC) is a mitigation countermeasure for severe accidents, which has been applied in the design of advanced pressurized water reactors. The validity of IVR-ERVC is restricted to boiling crisis and critical heat flux (CHF) of the reactor vessel lower head. A cylindrical slice heater block was usually employed to simulate the reactor vessel lower head in the majority of existing experiments, which may cause some uncertainties in CHF under natural circulation conditions. In the present paper, an experiment was constructed to study the characteristics of natural circulation and critical heat flux. Apart from the conventional cylindrical slice, a spherical slice was also employed as the heater block. The experimental data indicated that the natural circulation flow rate increases with increasing inlet water temperature, and shows large oscillations with the lift of circulation height. Besides, CHF decreases with the increase of inlet temperature. It was found that the CHF in the spherical slice is lower than that in the cylindrical slice from 10.3% to 28.4%, depending on the inlet water temperature. For the first time, the CHF uncertainty caused by the geometrical difference of the heater block was clarified.

Numerical study on transient response characteristics of natural circulation in coupled loops under ocean condition

The application of passive safety systems (PSSs) can greatly improve the safety level of the Offshore Floating Nuclear Power Plants. As one of the important component of PSSs, the secondary passive residual heat removal system can be seen as the multiple natural circulation (NC) loops coupled system. The transient response characteristics between coupled natural circulation loops may affect the core decay heat removal progress under accident conditions. Besides, the ship motions caused by ocean waves can induce oscillations of the system parameters and variations of the reactor thermal hydraulic characteristics. The superposition effects of the transient response characteristics of the multiple NC loops coupled system and the system parameters fluctuation caused by ocean conditions can generate more complex system behaviors. In this paper, effects of the transient response of key parameters during process of increasing power in coupled natural circulation loops under ocean condition was analyzed by the modified RELAP5 code. The simulation results show that the response time of each parameter of the system changes greatly under inclination and rolling condition, but the effect of pitching and heaving motion on the response time of multi-loop natural circulation system is basically negligible.

Thermal hydraulic characteristics of two-phase natural circulation for secondary side passive residual heat removal system

Abstract The Secondary Circuit Passive Residual Heat Removal System (PRHRS) is designed for advanced nuclear power system. A research on thermal hydraulic characteristics of two-phase natural circulation process for secondary circuit PRHRS has been carried out using the RELAP5 code. For this, the natural circulation steam generator (SG) and the PRHRS are modeled. Five thermal boundary conditions are set to verify the function of PRHRS. The phenomenon of flow instability in the two-phase natural circulation process is of great significance for system operation, as well as the performance of SG and condenser. Therefore, the focus is on the natural circulation flowrate oscillation of steam and condensed water. The mixed water level difference between descending pipe of PRHRS and secondary side of natural circulation SG is found to be the meaningful parameter to characterize the flow instability of natural circulation. The comparative analysis is also performed by considering the effects of height between the SG and condenser, as well as system resistance, which can provide support for PRHRS design and optimization.

A new theoretical model of steady-state characteristics of supercritical carbon dioxide natural circulation

Due to its commendable thermal properties in the supercritical region, supercritical carbon dioxide shows great promise as the working fluid of the Brayton cycle, which could achieve remarkable higher efficiency compared with steam Rankine cycle. The steady-state characteristics are essential for the design of supercritical carbon dioxide power conversion system. In the present article, the steady-state characteristics of supercritical carbon dioxide natural circulation were theoretically and experimentally studied. An one-dimensional theoretical model of steady-state characteristics of supercritical carbon dioxide natural circulation was put forward, which was developed in an analytical form. So far as is known to the authors, no similar theoretical model has been carried out ever. Experiments on the steady-state characteristics of supercritical carbon dioxide natural circulation were performed. The influence of system pressure, inlet temperature and enthalpy difference on the steady-state characteristics was discussed in detail. Results indicated that the steady-state characteristics of supercritical carbon dioxide natural circulation were closely related to geometrical parameters as well as thermal properties at the pseudo-critical point. The new theoretical model was further validated with present experimental data and available experimental results from the open literature.

Experimental Study on Characteristics of a Three-Phase Natural Circulation with TiO<sub>2</sub>, Water and Air

Experimental Study on Characteristics of a Three-Phase Natural Circulation with TiO<sub>2</sub>, Water and Air

Process of Air Ingress during a Depressurization Accident of GTHTR300

A depressurization accident is the design-basis accidents of a gas turbine high temperature reactor, GTHTR300, which is JAEA’s design and one of the Very-High-Temperature Reactors (VHTR). When a primary pipe rupture accident occurs, air is expected to enter the reactor core from the breach and oxidize in-core graphite structures. Therefore, it is important to know a mixing process of different kinds of gases in the stable and unstable density stratified fluid layer. In order to predict or analyze the air ingress phenomena during the depressurization accident, we have conducted an experiment to obtain the mixing process of two component gases and the characteristics of natural circulation. The experimental apparatus consists of a storage tank and a reverse U-shaped vertical rectangular passage. One side wall of the high temperature side vertical passage is heated and the other side wall is cooled. The other experimental apparatus consists of a cylindrical double coaxial vessel and a horizontal double coaxial pipe. The outside of the double coaxial vessel is cooled and the inside is heated. The results obtained in this study are as follows. When the primary pipe is connected at the bottom of the reactor pressure vessel, onset time of natural circulation of air is affected by not only molecular diffusion but also localized natural convection. When the wall temperature difference is large, onset time of natural circulation of air is strongly affected by natural convection rather than molecular diffusion. When the primary pipe is connected at the side of the reactor pressure vessel, air will enter the bottom space in the reactor pressure vessel by counter-current flow at the coaxial double pipe break part immediately. Afterward, air will enter the reactor core by localized natural convection and molecular diffusion.

Experimental investigation of the oscillation characteristics of natural circulation of R134A in a two-phase loop

Natural circulation (NC) loops were widely used in many fields, but still limited in the instability of the loop. The instability characteristics of NC are not fully understood, such as oscillation boundary, amplitude and frequency. To understand the oscillation characteristics, experiments of two-phase natural circulation were performed for R134a in a single loop. Two types of the oscillation were observed and identified by the characteristic frequency in their oscillation spectra. The characteristic frequencies and amplitudes of the oscillations were affected by the system pressure, the heating power and the evaporator inlet subcooling. For the type 1 oscillation, with the increase of the heating power, the main characteristic frequency increased, while the oscillation amplitude decreased; with the increase of the system pressure, the oscillation frequency decreased; with the increase of the subcooling, the frequency decreased. For the type 2 oscillation, with the increase of the heating power, the main characteristic frequency increased, the oscillation amplitude increased; with the increase of subcooling, the frequency decreased, and the amplitude increased. An instability map was built based on the dimensionless numbers: the subcooling number N_sub versus the phase change number N_pch. The stable region of the system was wider at low subcooling condition. Two types of the oscillation may occur at the same time, and the stable region disappeared when subcooling number higher than 3. Combining dimensionless analysis and spectrum analysis, we divided the two-phase natural circulation into five regions in the instability map. Through the oscillation spectrum, this study provided a deep perspective for the instability of two-phase natural circulation. Also, the experimental results of this study can lay the foundation for future research.

Modification of SIMPLE algorithm to handle supercritical natural circulation in a loop

The conventional SIMPLE algorithm for the pressure–velocity coupling has been adopted by many commercial CFD codes. Since it encounters convergence problem when it is used for numerical analysis of a two-dimensional unsteady natural convection flow in a rectangular cavity with zero-isothermal compressibility, the modification of SIMPLE algorithm has been proposed for such a fluid. In this paper, differences between solutions for a one dimensional natural circulation of a super-critical carbon dioxide in a loop with a horizontal heater and a horizontal cooler configuration obtained by using the conventional SIMPLE algorithm and the modified version of the algorithm are investigated. The modification of the algorithm includes updating the density at each time step based on its value at the previous time step to satisfy the mass conservation. The differences of the velocity and temperature are relatively small comparing with the two-dimensional natural convection case. As an example of utilizing the modified SIMPLE algorithm, characteristics of the unsteady natural circulation of super-critical carbon dioxide in a rectangular loop are revealed.

Neutronic/thermal-hydraulic coupling analysis of natural circulation IPWR under ocean conditions

In this paper, a thermal-hydraulic analysis code under non-inertial system was developed based on a RELAP5/MOD3 code by establishing dynamic simulation models of typical ocean conditions. The simulation results under periodic force field were in good agreement with the reference data, proving the accuracy of the program. In order to analyze the reactor core flow distribution and power distribution under different ocean conditions, the thermal-hydraulic code was coupled with a two-group three-dimensional neutron kinetics code. Based on the modified RELAP5 code, the effects of heeling, heaving and rolling conditions on neutronic/thermal-hydraulic coupling characteristics of natural circulation integrated pressurized water reactor (IPWR) were studied. The results indicate that the uneven distribution of coolant flow increases with the increasing inclination angle, but the reactor power distribution is subject to little changes. This trend leads to an uneven distribution of coolant temperature at the core outlet. The flow or power fluctuation has a 180° phase difference under rolling conditions, and the reactor power and coolant flow oscillation increases with the increasing rolling period and amplitude. In the case of heaving motion, the peaks of the oscillation amplitude of the flow and power lying in the hottest channel as the additional forces on the fluid of each channel are spatially uniform. Furthermore, the impact of the heaving amplitude is more significant than the heaving period, whether on the flow or power.

Capability of RELAP5 code to simulate the thermal-hydraulic characteristics of open natural circulation

Many advanced reactor designs incorporate the passive systems mainly to enhance their inherent safety. Some reactors even are designed to remove the accidental heat in the containment by an open natural circulation system, without using a mechanical pumps, e.g. HPR1000, APR+. The designs and safety analyses of these reactors are usually carried out by using the best estimate codes, such as the RELAP5, TRAC and CATHARE. These codes have been well validated by the pumped circulation systems and high-pressure natural circulation systems, but the applicabilities to simulate the thermal-hydraulic behaviors of open natural circulation have not been well proven by literatures.In the present study, one of the extensively used best estimate code RELAP5/MOD3.3 was used to simulate the thermal-hydraulic behaviors of an open natural circulation loop (PCCS-TF). The simulation results were validated against the corresponding experimental data performed on the PCCS-TF. The comparison results showed that RELAP5 code can successfully simulate most of the thermal-hydraulic transients, the flow mode distribution and the static characteristics of the open natural circulation system, in view of the simulation results reproduce in good agreements with the experimental data. Moreover, the influences of the structure parameters on the thermal- hydraulic behaviors were also researched, with the help of RELAP5 code to provide the technical supports for the design of the open natural circulation.