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Flow instability may threaten the nuclear safety and make the nuclear power plant more difficult to control. The objective of this research is to analyze the nature of flow instability in Natural Circulation (NC) with the rod bundles by the method of experiment and RELAP5 simulation. The Pressure Drop Oscillation (PDO) occurs when the pressurizer is located in the downstream of the test section. During the PDO, boiling, bubble expansion, flashing and condensation coexist, and there is fluid exchange between the pressurizer and the NC loop. No reverse flow occurs with the restriction of the experimental conditions. The system with the pressurizer in the downstream of the test section is more stable than that in the upstream of the test section. Increasing the inlet subcooling makes the system more stable. Based on the experimental data, RELAP5 was adapted to the analyses and was used to expand the research scope. With the increase of the heat flux, the single peak will turn to double peak and triple peak flow instability. The flow instability boundary is obtained by two methods. RELAP5 code applicability in the NC with low pressure condition has been validated.

In order to study the frictional resistance characteristics of a rod bundle channel under ocean conditions, a natural circulation experiment was performed in a 3 × 3 rod bundle channel under rolling conditions. The characteristics of the instantaneous frictional resistance coefficient in the rod bundle channel under rolling conditions are obtained by analyzing newly obtained experimental data. The experimental results show that the frictional resistance coefficient varies with the rolling conditions. The larger the rolling amplitude and frequency, the larger the average value of the frictional resistance coefficient, and the greater the fluctuation of the coefficient. It is also affected by the fluid property parameters. The hotter the fluid in the channel, and the faster the fluid velocity, which means the larger heating power and the steady state Reynolds number, cause the coefficient to decrease, and the fluctuation of the coefficient will also decrease. A new correlation function of the instantaneous frictional resistance coefficient under differing rolling conditions is developed and compared with these newly presented data. The correlation function is a good predictor of the frictional resistance coefficient and is in good agreement with the experimental results over a broader range of conditions than previous functions and is also applicable for both laminar and transitions fluid flow. This new correlation function introduces the viscosity correction factor based upon the previous research, which makes it more widely applicable in predicting the instantaneous frictional resistance coefficient under rolling conditions. This newly developed function should be useful for design considerations for passive cooling for sea-based nuclear reactor cores.

A considerable amount of research has gone into understanding the thermal-hydraulics characteristics of rod bundle channels which is of great importance for the nuclear industry. This paper will show that the development of the correlation of the frictional resistance coefficient in rod bundle channels under natural circulation conditions is well understood. This paper presents experimental studies of the single-phase resistance and obtains an improved correlation of the frictional resistance coefficient under natural circulation conditions based upon thermal-hydraulics characteristics which are well understood. It is also shown that this newly developed correlation closely predicts data. The experimental results show that the frictional resistance coefficient of a rod bundle channel under natural circulation is related to the physical properties of fluid and the size of the rod bundle channel. According to different flow regions, the relationship between the frictional resistance coefficient and Reynolds number is discussed separately, and the expressions of frictional resistance coefficient with Reynolds number in laminar flow region and transition region are obtained. After considering the influence of fluid properties on the frictional resistance coefficient, this paper introduces the kinematic viscosity correction factor to correct the frictional resistance coefficient. Finally, the complete expression of the frictional resistance coefficient was obtained. This newly developed predictive expression is in good agreement with experimental data in both the laminar and transition flow regions, with a relative error within ±5%. Thus we now have a predictor of the frictional resistance coefficient under natural circulation conditions based upon more complete thermal-hydraulic parameters in a static state.

Experimental investigation has been carried out in a closed loop rectangular natural circulation facility having a single heated macro channel dimensioned and built based on concepts of similarity and scale down circuit similar to the primary loop of a Nuclear reactor. The study aims to observe and critically analyze the thermal hydraulic parameters (Temperature Variation, Heating Power rises) as it affects transfer of heat by the NC flow along its hydraulic circuit at predetermined inlet subcooled set and sustained at 40°C, 50°C, 60°C and 80°C by regulating the coolant flow rate in the secondary loop. Heating channel power varies between 7kW – 13kW to portray several conditions of the reactor power in response to temperature changes base on the inlet temperature as a way of sustaining single-Phase flow through the flow loop while suppressing instabilities in the fluid circulation. Real-time signals of thermal hydraulic responses are observed via a graphical interface for data acquisition when stable flow or self-sustained flow oscillations are achieved. Results were analyzed at the (1) inlet and outlet of heater, (2) along the heated section and (3) inlet and outlet of condenser (4) the entire Loop. Pressure drops and Flow rate with respect to Temperature Difference was also analyzed to contribute for the acquisition of important data of natural circulation phenomena to be used for the design of large scale (prototypical) NC facility.

A series of experimental study on natural circulation pressure drop oscillations (PDO) under rolling motion are carried out. PDO under rolling motion is a chaotic motion, the thermal–hydraulic mechanism and evolution route are analyzed by maximal Lyapunov exponent and spectrum analysis. When rolling motion has different periods with PDO, composite oscillations are formed. The evolution route can be divided into 3 parts: thermal–hydraulic stable region, frequency-locking oscillation region and rolling effect weakening region. When heat flux is low, PDO is forced to have the same frequency as rolling motion. Frequency-locking oscillation has increasing maximal Lyapunov exponents and a broadband spectrum. But while under high heat flux, rolling effect is weakened. Evolution route of co-frequency oscillations has 3 parts: thermal–hydraulic stable region, bimodal oscillation region and resonance oscillation region. Bimodal oscillation region shows strong chaotic characteristic and cannot form regular waveforms. Rolling motion reduces chaos characteristic when resonance oscillation occurs.

Flow characteristics during loss of coolant to the heat exchanger (HX) via the secondary loop cold leg of a natural circulation (NC) test facility referred here as loss of heat sink transient (LOHS) in a closed rectangular NC system having a single heated channel at two different heating powers and system pressures has been investigated experimentally. Two LOHS experiments (LOHS-1 and LOHS-2) were conducted in a well dimensioned closed loop NC test facility designed to represent the primary system of a nuclear reactor. Operating conditions were set and maintained at 6.0 kW and 9.0 kW with system pressures of 0.2 MPa and 0.5 MPa respectively for both single-phase and single-to-two-phase flow transition to determine their thermal hydraulic responses at these predetermined parameters. The transients were initiated by closing the cold leg of the secondary loop supplying coolant to the tube-in-shell heat exchanger while NC flow phenomena were monitored at each vital section along the flow map based on the pre-determined inlet operating conditions. LOHS experiment stops when critical heat flux (CHF) occurs or heat exchanger outlet (HXoutlet) temperature rises too high to approximately equal its inlet temperature. Results were acquired using a system-design platform and development environment LabVIEW, programmed in a dedicated computer that was interfaced with the coupled sensors in the measuring devices. Common computational tools which include MATLAB, Ms Excel and Origin software were used for the analysis of the acquired data. Results at the inlet and outlet of heated sections as well as the inlet and outlet of condensers of both tests were analyzed and discussed in details at various flow regimes with the associated phenomena. It is concluded that while stable flow was maintained before transient with slug flow occurring in the heating channel, the system experienced transient stability resulting in steady temperature increase and subsequently drops into an instability region when the transient scenario was activated. In addition, the NC system transited into stable two-phase flow and further loses stability when HXoutlet temperature surges to its inlet value with CHF occurrence. These results are presented to contribute to a more extensive utilization of NC systems and understanding of flow behavior during LOHS in advanced reactors.

Loss of heat sink (LOHS) accident in a natural circulation system has a complex event sequence scenario. Flow instability phenomena in LOHS accident involves multiple flow patterns and variety instability mechanisms. Through experimental investigation, a methodology analyzing instability under LOHS is established. It is concluded while slug flow occurring in heating channel, the system falls into geyser instability region which is proven as pressure drop oscillations (PDO) accompanied by wave mode changing. While annular flow occurs, natural circulation system forms stable two-phase flow transitorily and losses stability again. The post-annular-flow instability cannot form a regular vibration wave mode, and annular flow absorbs flowrate oscillation instead of pressurizer. Finally, average flowrate approaches to 0 and critical heat flux (CHF) happens.

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

A set of simplified iterative model coupled with experimental procedures has been derived to aid in computing the flow rate of a closed rectangular NC Circuit at low pressure/low power steady state condition. The approach involves partitioning the NC loop into different pressure loses segments depending on the associated factor, thereafter the various pressure drops in each segment is determined with applicable mathematical correlations. The iterative model allows for varying of the outlet temperature, at a predetermined Heating Power (Decay Heat) and inlet temperature such that the flow rate is determined when the pressure drops equals to the driving force ( Δ P drive = Σ Δ P loss ) . The model was applied in computing the flow rate of the NC loop considered based on the assigned parameters with results validated with experimental data showing good agreement. Microsoft excel was employed in the computation procedures to allow for simplicity and easy comprehension.

In order to enhance the safety of nuclear energy, natural circulation served as a passive technology is widely used. Accordingly, an investigation associated with thermal-hydraulic characteristics for natural circulation system is made in this work. Numerical simulation based on RELAP5 code is performed especially for flow instability phenomenon. The results indicate that flashing and condensation are two important factors for low pressure natural circulation two-phase system. For all this, the tolerance of natural circulation to flashing oscillation could be enhanced effectively with the increase of system pressure. Under high pressure condition, the flowrate might shrink to a new equilibrium state from flashing-induced oscillation. The simulation results and analysis could contribute to a more extensive utility of natural circulation mechanism for nuclear energy.