Mechanism analysis on flow excursion of a natural circulation with low steam quality

Flow excursion was observed on the natural circulation test loop HRTL-5, which simulates the geometry and system design of the 5 MW nuclear heating reactor. By means of self-developed computational codes, a set of tools used for analyzing the flow characteristics of the natural circulation have been presented, including the characteristic curves, operational curve and the bifurcation curve. The two-dimensional disturbance analysis is adopted to explain the mechanism of the flow excursion. Analytical result shows: (1) flow excursion can occur in a natural circulation system at a suitable geometry and thermal–hydraulic conditions. (2) Characteristic curves, operational curve, bifurcation curve, and the two-dimensional disturbance analysis are the available method to analyze the flow excursion of the natural circulation. (3) The flow excursion is prior to the low steam quality density wave oscillation. (4) The onset of the flow excursion is the tangency point of the drive force curve and the flow resistance curve. (5) To operate at low heat flux to increase the inlet temperature is the effectual approach to transfer from the pressurized water state to the boiling water state, in which the flow excursion and the low steam quality density wave oscillation can be avoided.     

Flow characteristics of natural circulation in a lead-bismuth eutectic loop

Lead and lead-alloys are proposed in future advanced nuclear system as coolant and spallation target.To test the natural circulation and gas-lift and obtain thermal-hydraulics data for computational fluid dynamics (CFD) and system code validation,a lead-bismuth eutectic rectangular loop,the KYLIN-Ⅱ Thermal Hydraulic natural circulation test loop,has been designed and constructed by the FDS team.In this paper,theoretical analysis on natural circulation thermal-hydraulic performance is described and the steady-state natural circulation experiment is performed.The results indicated that the natural circulation capability depends on the loop resistance and the temperature and center height differences between the hot and cold legs.The theoretical analysis results agree well with,while the CFD deviate from,the experimental results.     

Flow characteristics of natural circulation in a lead–bismuth eutectic loop

Lead and lead-alloys are proposed in future advanced nuclear system as coolant and spallation target.To test the natural circulation and gas-lift and obtain thermal-hydraulics data for computational fluid dynamics(CFD) and system code validation, a lead–bismuth eutectic rectangular loop, the KYLIN-Ⅱ Thermal Hydraulic natural circulation test loop, has been designed and constructed by the FDS team. In this paper, theoretical analysis on natural circulation thermal-hydraulic performance is described and the steady-state natural circulation experiment is performed. The results indicated that the natural circulation capability depends on the loop resistance and the temperature and center height differences between the hot and cold legs. The theoretical analysis results agree well with,while the CFD deviate from, the experimental results.     

Flow characteristic of single-phase natural circulation under ocean motions

Natural circulation is widely used in nuclear reactor systems as the passive safety system. With the development of the floating nuclear power plant (FNPP), researchers should pay more attention to flow and heat transfer characteristics for the natural circulation under ocean conditions for the safety of FNPP. In this paper, the flow characteristics in a single-phase natural circulation system were investigated and the effects of heaving, rolling and coupled motions were analyzed. The oscillation amplitude of flow rate increases with the increase of period in a certain range and maximum acceleration under heaving motions. With the increase of oscillation intensity (higher frequency and larger maximum rolling angle), the oscillation amplitude increases and the average flow rate decreases under rolling motions. Moreover, the lateral displacement of rolling center changes the oscillation period and induces larger amplitude oscillations. The flow characteristic becomes more complex when the system is subjected to coupled motions. The oscillation period is the least common multiple of two motions’ periods. The oscillation induced by coupled motions makes the system more unstable than that induced by an individual motion. The potential superposition effect exists under coupled motions and needs to be addressed for the operation safety.     

Flow stability in a natural circulation loop: influences of wall thermal conductivity

In the study of single-phase natural circulation loops, flow stability has been an important subject because of the engineering needs for obtaining stable operating state. As an attempt of proposing method for stability control, this article studied the change of the flow stability due to the variation of the tube wall thermal conductivity. A nonlinear dynamical model was proposed to formulate the wall effects. It is suggested that in a loop made of better thermal conducting material, the flow is also more stable. Experiments were performed in three natural circulation loops; they are in the same shape and are at the same thermal boundary conditions. An unstable Lorenz flow was observed in a glass loop. In a copper loop and a copper–glass loop, both the flows were stable for any experimental cases. The results verified qualitatively the conclusions drawn by the model.     

Flow excursion time scales in the advanced neutron source reactor

Flow excursion transients give rise to a key thermal limit for the proposed advanced neutron source (ANS) reactor because its core involves many parallel flow channels with a common pressure drop. Since one can envision certain accident scenarios in which the thermal limits set by flow excursion correlations might be exceeded for brief intervals, a key objective is to determine how long a flow excursion would take to bring about a system failure that could lead to fuel damage. The anticipated time scale for flow excursions has been examined by subdividing the process into its component phenomena: bubble formation, flow deceleration, and fuel plate heat-up. Models were developed to estimate the time required for each individual stage. Accident scenarios involving sudden reduction in core flow or core exit pressure have been examined, and the models compared with RELAP5 output for the ANS geometry. For a high-performance reactor such as the ANS, flow excursion time scales were predicted to be in the millisecond range, so that even very brief transients might lead to fuel damage. These results have been useful for determining the significance of momentary flow excursion events calculated for accident situations in the ANS reactor. In addition, the methods presented are applicable for evaluating the timing of flow excursion transients in other facilities as well.     

Experimental study on natural circulation and its stability in a heavy liquid metal loop

Motivated by the increasing interest in heavy liquid metal (HLM) cooled fast reactors and accelerator driven system (ADS), the TALL test facility was designed and constructed at KTH to investigate the thermal-hydraulic characteristics of HLM. In this paper, the HLM natural circulation characteristics in a HLM loop were investigated with experiments in the TALL test facility. The study includes measurements on (1) start-up of natural circulation from different initial conditions; (2) stability of natural circulation; (3) effects of influencing parameters and (4) capability of natural circulation. The experimental data are compared to predictions with a relevant code (RELAP5). Significant natural convection flow was observed in the experiments. It was found that the natural circulation was easily established and stabilized. It took only a few minutes to have a stable natural circulation prevailing from cold conditions. The natural circulation flowrate depends on the loop resistance, and the temperature difference between the hot leg and the cold leg, as determined by the power level and the heat sink capacity. The experiments show that the maximum flowrate for the natural circulation is 0.5 kg/s (corresponding to 0.5 m/s in the heat exchanger), resulting in heat removal of 15 kW from the core tank, which is comparable to the capacity of 100 W/cm of the electric heater elements. The preliminary analysis performed with the RELAP5 code is in reasonable agreement with the experimental data.     

Flashing-induced density wave oscillations in a natural circulation BWR—mechanism of instability and stability map

The SIRIUS-N facility was designed and constructed for highly accurate simulation of core-wide and regional instabilities of a natural circulation BWR. A real-time simulation was performed in the digital controller for modal point kinetics of reactor neutronics and fuel-rod conduction on the basis of measured void fractions in reactor core sections of the thermal-hydraulic loop. Stability experiments were conducted for a wide range of thermal-hydraulic conditions, power distributions, and fuel rod time constants, including the nominal operating conditions of a typical natural circulation BWR. The results show that there is a sufficiently wide stability margin under nominal operating conditions, even when void-reactivity feedback is taken into account. The stability experiments were extended to include a hypothetical parameter range (double-void reactivity coefficient and inlet core subcooling increased by a factor of 3.6) in order to identify instability phenomena. The regional instability was clearly demonstrated with the SIRIUS-N facility, when the fuel rod time constant matches the oscillation period of density wave oscillations.     

Flow excursion instability in downward flow systems part II: two-phase instability

A mechanistic model is presented in this paper to predict the onset of flow excursion in downward flows at low-pressure conditions. The model is represented in a graphical form on the subcooling number versus the Zuber (phase-change) number plane. The subcooling number and the Zuber number are measures of integral system parameter such as fluid properties, inlet subcooling, heat flux, channel geometry and coolant flow rate (all measurable quantities) resulting in an easy extrapolation of the predictive correlation to the physical system. The model addresses the distinction between the point of significant void and the onset of flow excursion and provides an analytical method for its evaluation. The model is compared with flow excursion data for downward flows and the agreement is satisfactory.     

Flow excursion instability in downward flow systems: Part I. Single-phase instability

A procedure for predicting the onset of flow excursion instability in downward flows at low-pressure and low-flow conditions without boiling is presented. It is generally accepted that the onset of significant void in subcooled boiling precedes, and is a precondition to, the occurrence of static flow instability. A detailed analysis of the pressure drop components for a downward flow in a heated channel reveals the possibility of unstable transition from single-phase flow to high-quality two-phase flow, i.e. flow excursion. Low flow rate and high subcooling are the two important conditions for the occurrence of this type of instability. The unstable transition occurs when the resistance to the downward flow caused by local (orifice), frictional, and thermal expansion pressure drops equalizes the driving force of the gravitational pressure drop. The inclusion of the thermal expansion pressure drop is essential to account for this type of transition. Experimental data have still to be produced to verify the prediction of the present analysis.     

核动力装置强迫循环与自然循环过渡过程特性研究

针对某型压水堆核动力装置,建立反应堆及一回路系统强迫循环与自然循环的计算分析模型,并与试验值进行比较,验证了建立的模型计算精度高,满足工程分析的要求.利用建立的数学模型,对自然循环与强迫循环过渡过程进行分析计算,结果表明:强迫循环向自然循环转换过程中冷却剂流量、蒸汽发生器压力、反应堆出口温度是几个约束参数;自然循环向强迫循环的转换过程中反应堆功率变化与周期变化幅度较大.     

Bifurcation and hysteresis phenomena in the two－phase natural circulation system

The static bifurcation of the two-phase natural circulation (TPNC) system was studied theoretically and numerically.By the DERPAR algorthm the solution diagram was calculated,which shows that the static bifurcation occurs under some conditions in the TPNC systems.Also,it shows that the static bifurcation occurs under some conditions in the TPNC systems.Also,it shows that,in a region of multiple solutions.the static instability may occur.It is defined as a region of thermal-siphon instability induced flow rate jumping.By means of the solution diagram,the stability margin can be determined in this region.Furthermore.the heat input at the peak of the solution diagram is defined as the maximum capacity of heating load that can be used to judge the capacity of the TPNC of a given geometry topological structure,Meanwhile,it is interesting that the TPNC systems have the hystersis phenomenon defined as thermal-siphon hysteresis.Some parametric effects related were also studied.     

Study on Pb-Bi natural circulation phenomena

For the development of 45w%Pb-55w%Bi cooled direct contact boiling water small fast reactor (PBWFR), experimental study on Pb-Bi-water direct contact boiling two-phase flow has been performed using Pb-Bi-water direct contact boiling two-phase flow loop. For stable start-up of the boiling flow operation, Pb-Bi single-phase natural circulation must be realized in a Pb-Bi flow system of the loop before water injection into Pb-Bi. The Pb-Bi flow system consists of a four-heater-pin bundle, a chimney, an upper plenum, a level meter tank, a cooler, and an electromagnetic flow meter. A stable Pb-Bi single-phase natural circulation was realized in the range of flow rate from 1.5 l/min to 4.8 l/min by heating Pb-Bi in the heater-pin bundle with a power up to 7.7 kW. The inlet and outlet temperatures of the heater bundle were in the ranges from 243°C to 278°C, and from 251°C to 278°C, respectively. The natural circulation flow was simulated analytically using one-dimensional flow model including frictional, form and drag forces. Total hydraulic head through the loop were calculated from Pb-Bi densities at measured Pb-Bi temperatures in the loop. It was found that the calculated flow rate agreed well with the measured ones, which indicated the validity of the analytical models.     

Startup instability in natural circulation driven nuclear reactors

Natural circulation driven nuclear reactors are prone to flow instability during the startup transients. This paper intends to provide the state-of-the-art reviews on the theoretical analysis and experimental studies on flow instability in three types of natural circulation driven reactors, ranging from conventional nuclear reactors to small modular reactors. Brief overviews of three categories of startup flow instability, i.e., density wave oscillations, flashing instability, and Geysering instability, are provided. A critical review is conducted for the scaling analysis and design of small scaled test facility. The review of obtaining quasi-steady state stability maps in the dimensionless stability plane through frequency domain analysis and experimental tests provides the state-of-the-art methodology of analyzing the flow instability. Experimental startup instability during different initial startup procedures is reviewed. Although extensive efforts have been made to study the flow instability, further work is required to improve the scaling ability of experimental investigation and the accuracy of code analysis. Some discussions for future research directions are given.     

Analysis of natural circulation phenomena in VVER-1000

In all light water reactors (LWR), natural circulation is an important passive heat removal mechanism. In the present paper, the natural circulation phenomena are studied with reference to step-wise coolant inventory reduction and a small break loss-of-coolant-accident (SBLOCA) in the cold leg of VVER-1000. The natural circulation flow map (NCFM) approach is considered to evaluate the natural circulation performance of the VVER-1000 NPP also comparing VVER-1000 and PWR systems. Three different elevations between heat source (core) and heat sink (steam generators) zones have been considered in order to characterize the buoyancy force in a VVER-1000. The influence of power and the cold legs loop seal upon the natural circulation performance is also evaluated. In the second part, a series of SBLOCA simulations with break area ranging from 0.5 to 11.7% of the cold leg cross sectional area are performed starting with the VVER-1000 system in nominal conditions. The effect of Emergency Core Cooling System (ECCS) including passive and active parts of ECCS are evaluated. The simulations were performed by the help of the system code RELAP5. Within the framework of the qualification of the adopted computational tools, the results are compared with experimental data from Kozloduy NPP unit 6 test and PSB-VVER integral test facility available from the literature. Namely, the qualification of the adopted nodalisation in steady state conditions is achieved by using experimental data. The accuracy of selected results have been estimated in quantitative terms by applying the fast Fourier transform based method (FFTBM). Finally, the relevance and the potential for the occurrence of the reflux condensation mode, i.e., one of the Natural Circulation regimes, for cooling of reactor core in VVER-1000 are discussed.