The Passive Residual Heat Removal System (PRHRS) is very important for the safety of the heating reactor after shutdown. PRHRS is a natural circulation system driven by density difference, therefore, the heat transfer performance of the Passive Residual Heat Removal Heat Exchanger (PRHR HX) has a great impact to the heat transfer efficiency of PRHRS. However, the most research object of PRHR HX is the C-shape heat exchanger at present, which located in In-containment Refueling Water Storage Tank (IRWST). This heat exchanger is mainly used for the PRHRS of nuclear power plants. In the swimming pool-type low-temperature heating reactor (SPLTHR), the PRHR HX is placed in the reactor pool, which the pressure and temperature of the reactor pool are relatively low, and the outside heat transfer mode of tube bundle is mainly natural convection heat transfer. In this study, a miniaturized single-phase pool water cooling system was built to investigate the natural convective heat transfer coefficient of the heat exchanger under the large space and low temperature conditions. The experimental data had been compared with several correlations. The results show that the predicted value of Yang correlation is the closest to the experimental data, which the maximum deviation is about 11%.

Blowdown process is an important process in LOCA. Studying the internal thermal hydraulic characteristics of the pressurizer during LOCA blowdown process can help to accurately analyze the process of the whole LOCA, which is of great significance to reduce the hazard of accidents and improve the safety of nuclear power plants. An experiment system of blowdown process under low pressure conditions is built, and it uses a pressure gauge and a differential pressure level gauge to measure the pressure and liquid level in the pressurizer-type heating container, and uses a thermocouple to measure the temperature in the pressurizer-type heating container at different heights and distances from the centerline. The changes of pressure, liquid level and temperature characteristics during the blowdown process were obtained. The research results show that the higher the heating power, the greater the slope of pressure change in the heating container, and the pressure changes from negative to positive growth; The greater the heating power is, the faster the liquid level decreases, and the liquid level decreases sharply after the blowdown inflection point. The greater the heating power is, the greater the rate of temperature change, from negative growth to positive growth. The temperature change rate after the blowdown inflection point is reduced. As the height increases, the temperature inside the container first increases and then decreases; the further away from the container’s centerline, the higher the measured temperature.

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.

Nuclear power plant (NPP) safety has become a public issue since the Fukushima daiichi NPP accident. In order to evaluate the risks caused by severe accidents (SAs), it is very important to understand the on-site source term events. One of the important unsolved source term events is the decontamination efficiency of fission products (FPs) in the suppression chamber by pool scrubbing. Therefore, a mechanistic model to analyze the particle decontamination efficiency by pool scrubbing is highly regarded. Despite the demand, particle decontamination mechanism by pool scrubbing has never been understood due to the complexity of phenomena. In our experiment, we aim to develop a reliable mechanistic model to evaluate particle decontamination efficiency of pool scrubbing by conducting separate effect tests. As to obtain the fundamental process of particle decontamination from gas to liquid-phase, we focused on decontamination factor (DF) of particle from a single bubble. However, it is very difficult to calculate the initial particle concentration inside the bubble. Therefore, in our experiment, we developed a method to measure the internal particle concentration inside the bubble by combining image processing and particle measurement. By using the experimental results, we succeeded to obtain reasonable DF for glycerin particles and CsI particles as a simulant particle for FPs. From the experimental results, detailed particle decontamination efficiency for various submergence were measured. The results tend show that DF increase linearly as submergence increases which suggests that DF is constant on bubble rise region. Moreover, the fact that glycerin particle with larger particle diameter takes a higher value shows that particle diameter significantly affects DF.

The marine nuclear power plant operating in the marine environment has complicated motion under the influence of wind and waves. The movement of marine nuclear power plant will affect the thermal-hydraulic characteristics of its nuclear reactor system. Compared with other typical motion conditions, the effects of rolling conditions on the thermal-hydraulic characteristics of the nuclear reactor system are the most complex. In order to study the effects of rolling conditions on the thermal-hydraulic characteristics of the nuclear reactor system, a thermal-hydraulic system code for motion conditions named STAC was developed. The STAC code was verified by the experiments conducted in Japan. The effects of rolling conditions on the thermal-hydraulic characteristics of the nuclear reactor system under forced circulation and natural circulation are studied with the STAC code. The simulation results show that the thermal parameters of the reactor system under rolling condition fluctuate periodically. The fluctuation period of the thermal parameters of the core is half of the rolling period, and the fluctuation periods of other thermal parameters are the same as the rolling period. The effect of rolling condition on thermal-hydraulic parameters under forced circulation is smaller than that under natural circulation. The fluctuation amplitudes of the thermal parameters increase with the angle amplitude of the rolling condition. There is a rolling period with the smallest fluctuation amplitude. Under the rolling condition with short period, the fluctuation amplitudes of the thermal parameters increase and their average values change rapidly as the rolling period decreases. Under the rolling condition with large period, the fluctuation amplitudes of the thermal parameters increase with the rolling period, and they tend to fixed values.

Corrugated plate dryer is a extremely vital equipment for steam-water separation in the fields of heat transfer and nuclear engineering. The corrugated plate is also a commonly used steam-water separator in steam generators in nuclear power plants. It is meaningful to study the breakdown characteristics and mechanism of the water film on corrugated plate wall. Water film thickness of steady flow is measured based on plane laser induced fluorescence (PLIF) technique and time series and its fitted equation of water film thickness are obtained, respectively. Besides, fluctuation characteristics of water film are analyzed by probability density function (PDF). Based on the dimensionless approach, the water film breakdown model at the corner of the corrugated plate is established. And the calculation equation of the relative position of the water film breakdown at the corner is deprived. The specific conclusions are as follows. The theoretical equation agrees well with the relative position of the water film breakdown at the corrugated plate corner. The evolution of the surface wave of water film is carried out in time and space. The PDF curve have no significant peak characteristics. Therefore, the spectrum has no characteristic frequency, that is, the water film has multi-frequency characteristics. Gravity of water film can be ignored in the water film model. The thickness sequences for falling film is measured and fitted. The two-dimensional model of water film breakdown at the corner is set up. The equation for the film thickness when the water film is just ruptured is obtained. Relative position of the water film rupture at the corner of the corrugated plate is theoretically related only to the structural parameters of the corrugated plate, the parameters of the gas phase and the liquid phase, and the Reynolds number of the liquid film. However, in the low Reynolds number region, the airflow velocity is extremely large, which causes certain fluctuations and nonlinear characteristics of the water film boundary position. Therefore, the theoretical formula is not particularly good at predicting the relative position of the breakdown in this region. I think that this nonlinear feature has obvious chaotic characteristics. The study of the chaotic characteristics generated by shearing the liquid film by high velocity flow airflow at the corner of the corrugated plate may become a prospect for future research.

Departure from nucleate boiling (DNB) may occur in the process of reactor accident. The departure from nucleate boiling (DNB) design basis is one of the most important bases for reactor core thermo-hydraulic design. In order to evaluate whether the DNB design basis meets the demand of thermal hydraulic design, the departure from nucleate boiling ratio (DNBR) should be calculated. The traditional DNBR calculation method is the direct substitution method (DSM). However, the research has pointed out that the DSM was inaccurate and non-conservative to represent the real safety margin. There is also a modified DNBR method by combining with the heat balance equation, which was called heat balance condition method (HBM). The essence of both DNBR methods is to evaluate the reactor safety margin by the difference between the heat fluxes of the actual condition and the critical condition. So the critical heat flux (CHF) value needs to be measured by CHF tests in actual condition parameters, which results in high heating power. To avoid this deficiency of DNBR methods, a new safety margin evaluation method has been developed in the research, which was called quality margin method. In this method, the safety margin of the reactor was evaluated by the difference between the critical quality and the actual quality. The method required lower test power, so greatly decreased the difficulty and cost of CHF experiments.

We investigated a single droplet impinging on an inclined heated wall with different inclination angles. A high-speed camera was used to observe this impinging process at 10000 frames / second. The phenomena of the droplet spreading, shrinking, rebounding, boiling, break-up and splashing were observed. The effects of the wall temperature (T w , 40–262 °C), the Weber number of the droplet (We d , 0.66–589) and the wall inclination angle (α, 0–45.6°) on the spreading behavior of the droplet after impinging on the wall were analyzed. Energy conservation equation was used to analyze experimental results. The results show that increasing the inclination angle of the wall is beneficial to the forward spreading of the droplet along the wall, but not to the downward spreading. When α > 45.6°, the droplet will break through the flow resistance of the wall and slide down all the time. The increase of the wall temperature and the Weber number of the droplet is beneficial to the backward and forward spreading of the droplet. Increasing α, increasing T w and decreasing We d will promote the shrinking of the droplet after spreading. In addition, the experimental phenomenon also shows that the larger the wall inclination and the higher the wall temperature, the easier the droplet will break away from the wall.

Recent advances in horizontal cask designs for commercial spent nuclear fuel have significantly increased maximum thermal loading. This is due in part to greater efficiency in internal conduction pathways. Carefully measured data sets generated from testing of full-sized casks or smaller cask analogs are widely recognized as vital for validating thermal-hydraulic models of these storage cask designs. While several testing programs have been previously conducted, these earlier validation studies did not integrate all the physics or components important in a modern, horizontal dry cask system. The purpose of this investigation is to produce data sets that can be used to benchmark the codes and best practices presently used to calculate cladding temperatures and induced cooling air flows in modern, horizontal dry storage systems. The horizontal dry cask simulator (HDCS) has been designed to generate this benchmark data and complement the existing knowledge base. Transverse and axial temperature profiles along with induced-cooling air flow are measured using various backfills of gases for a wide range of decay powers and canister pressures. The data from the HDCS tests will be used to host a blind model validation effort.

Small reactors have received more and more attention for their high safety, reliability, low power density, and short construction period. And the gas-steam pressurizer is widely used in small reactors due to its characteristics of simple structure, saves the heating and spray equipment, and prevents the coolant from boiling. The gas-steam pressurizer is a pressure control equipment for the reactor coolant system, and its characteristic of transient response is an important factor that affect operation stability of nuclear reactor systems. An experimental system was established to study the effect of pressure response for an insurge transient and influence factors were analyzed quantitatively. Experimental investigation shows that for the gas-steam pressurizer, the increase of coolant loading capacity (insurge) can cause system pressure rising. And the change of system pressure has much consistency with the change of liquid level and gas space temperature. The liquid phase exists temperature fluctuations and overall shows a downward trend during the insurge transient. And there exists a temperature gradient from bottom to top in the pressurizer liquid phase region during the insurge transient. The change of water vapor quantity curve is the oscillating curve during the transient and water vapor quantity is in a decreasing trend overall during the insurge transient. What’s more, the experiments also analyzed the pressure response and temperature response during the insurge transient.

The Transient Reactor Test (TREAT) Facility at Idaho National Laboratory (INL) started testing new fuels and reactor technologies once again in 2018 and new experiments and tests are currently being designed like for example the water loop “TREAT Water Environment Recirculating Loop” (TWERL). During the design of such experiments, the designer must assess how close the experiment reproduces the physics (and other important phenomena) happening during a transient of interest compared to the full-size reactor the experiment attempts representing. Traditionally, to assess this “representativity” of the experiment, scaling theory involving expert judgment is needed. This paper presents a step towards a systematic modeling and simulation (M&S) informed methodology for experiment design. The new methodology compares a model of the full system and a model of the mock-up facility that are subject to the same perturbations. In this way, the “overlap” of the perturbed experiment and full-size facility model outputs can be analyzed and the “representativity” of the experiment determined. The paper presents a RELAP5-3D analysis, where TWERL LOCA calculations are compared to prototypic PWR LOCA calculations with respect to representativity. To inform the design of the TWERL experiments, i.e. to find the most “representative” configuration for the TWERL loop, different design parameters for TWERL have been optimized in the study.

A theoretical model for Density Wave Oscillations (DWOs) flow instability in parallel rectangular channels under periodic heaving motion is established with a lumped mathematical model based on homogenous hypothesis. The parallel rectangular channels comprise of the entrance section, the heating section, the riser section and the upper- and lower plenums, which guarantee the isobaric pressure drop condition between channels and the model consists of boiling channel model, pressure drop model, parallel channel model, additional pressure drop model generated by heaving motions, the constitutive and numerical models. The effect of periodic perturbation is introduced through additional pressure drop in the momentum equation. The model is validated with experimental data of a twin-rectangular-channel flow instability experiment under static condition. Then the flow instability in parallel-rectangular-channel system is studied under periodic perturbation and the margin of flow instability and the threshold power of the system under static condition is calculated as basis condition for comparison. The effect of the amplitude and period of perturbation is analyzed analytically and the results show that the amplitude and period of perturbation shows little effect on flow instability. While when the additional pressure difference introduced by heaving motion is comparable with that under static condition, the effect of amplitude becomes stronger. And the period of perturbation strongly effects the threshold power when it is identical to that of natural period of the system, which can be explained by resonance between the perturbation and the system. And this effect is even stronger when the asymmetric heating condition is introduced.

The effects of rolling condition on the flow instability characteristics of natural circulation (NC) in rod bundle channel were experimentally studied. A 3 × 3 rod bundle channel is used as the testing section. The experimental system pressure range is 0.2 to 0.6 MPa, and the range of inlet subcooling is 10 to 70 °C. The ranges of rolling motion amplitude and period are 10 ∼ 20° and 10 ∼ 30s, respectively. Two typical two-phase flow instabilities in rod bundle channel under rolling condition were found in experiments: (a) the trough-type oscillation caused by the vapor generation at the minimum point of flow fluctuation and (b) the compound oscillation formed by the superposition of the trough-type oscillation and DWO I . Experimental results show that the rolling motion can reduce the threshold heating power of trough-type oscillation and cause the occurrence of NC flow instability in advance. But the rolling motion cannot affect the dimensionless boundary of DWO I in rod bundle channel.

Based on Armenian Nuclear Power Plant (VVER-440/270) example the response of confinement to leaktightness improvement in case of design basis accident was investigated. The main interest of the investigation is to determine the influence of confinement leaktightness improvement on the reduction of available NPSH of active safety systems pumps (ECCS and spray systems). Results of analyses showed that confinement leaktightness improvement will lead to strong underpressure condition and endanger ECCS pumps operation due to cavitation. Hence, in parallel with confinement leaktightness improvement confinement vacuumization prevention means must be considered.

This paper presents the experimental results and analyses of the structure velocity of air-water two-phase flow in a 3 × 3 rod bundle channel. A total of 56 flow conditions were tested and investigated for rod-gap, sub-channel, rod-wall and global regions of rod bundle geometry. The experimental tests were carried out under bubbly and cap-bubbly flow regimes with superficial gas and liquid velocities of 0–1 m/s and 1–1.7 m/s, respectively. The conductivity probes were set at different heights to measure the global and local void fractions. The structure velocity of air-water two-phase flow is the average bubble velocity calculated by the method in this study. The structure velocity were determined by utilizing the cross-correlation technique to analyze the time lags of the bubbles passing through the conductivity probes. The results of this study indicated that the structure velocity may increase with increasing superficial gas and liquid velocities. In low superficial gas velocity region, the structure velocity may first slightly increase and follow by a sudden jump which appear in most regions. After the sudden jump, the structure velocity may keep increasing mildly. The present structure velocities have been compared with the area-averaged gas velocities predicted by the drift flux model, and it appears that most structure velocities show a good agreement with the averaged gas velocities from the drift flux model after the jump.

Heat transfer between cladding and coolant during transient scenarios remains a critical area of uncertainty in understanding nuclear reactor safety. To advance the understanding of transient and accident scenarios involving critical heat flux (CHF), an in-pile experiment for the Transient Reactor Test facility (TREAT) at Idaho National Laboratory (INL) was developed. The experiment, named CHF-Static Environment Rodlet Transient Test Apparatus (CHF-SERTTA), consists of a hollow borated stainless-steel heater rod submerged in a static water pool heated via the (n, α) reaction in boron-10. This paper presents a novel inverse heat transfer method to determine CHF by using the optimization and uncertainty software Dakota to calibrate a RELAP5-3D model of CHF-SERTTA to temperature measurements obtained from a thermocouple welded to the surface of the rod.

To investigate the heat transfer characteristics of the chrome-plated tube and hope it could be used on the internal heat exchanger of passive containment cooling system (PCS), an experimental investigation has been conducted. In this experiment, a series of steam condensation experiments are performed under pure steam and steam-air mixed conditions over chrome-plated tube for a variety of chromium coating thickness (1μm and 10μm), total pressure, air mass fraction and wall subcooling. Condensation heat transfer coefficient was obtained for the total pressure ranging from 0.2 MPa to 0.4 MPa, air mass fraction ranging from 0.10 to 0.71, and wall subcooling from 10°C to 70°C. Moreover, the designed visualization experimental device makes the experimental phenomenon can be directly observed through the observation window. Under the pure steam condition, the results show that droplet condensation and filmwise condensation is co-existed on both two kinds of chrome-plated tubes, the chrome coating thickness of 10μm tube shows better heat transfer ability. Under the steam-air mixed condition, the condensation heat transfer coefficient of both two kinds of tubes increases with total pressure, and decrease with the air mass fraction and wall subcooling, while the influence of chrome coating thickness on heat transfer is no longer noticeable. The results also indicate that the thickness of the chromium coating will affect the surface microstructure of the chrome-plated tube and then affect the heat transfer ability of the chrome-plated tube.

In the third generation pressurized water reactor AP1000 plant, the Automatic Depressurization System (ADS) is one of the most important passive safety system. However, the steam Direct Contact Condensation (DCC) microscopic mechanisms are very complicated, which are not very clear yet. Moreover, the high-pressure and high-temperature experiment is very expensive to be conducted for many different test conditions. So in the present work, both the experimental and numerical methods are employed to investigate the steam DCC behavior. The steam DCC experimental bench has been built up, and the key parameters including the flow patterns and steam core temperature distributions are measured to provide validation data for the numerical results. In aspect of the numerical work, CFD simulation on the steam condensation is conducted. The heat and mass transfer process is simulated through the three-dimension commercial software FLUENT 16.0. Some of the key heat and mass transfer correlations are added by User Defined Function (UDF). The key parameters including the condensation steam fraction, temperature, and pressure, etc. are analyzed, which reflect the major heat transfer characteristics. According to the results, the expansion-compression-steam tail could be observed in both the numerical and experimental results. In essential, the steam fraction, temperature, and pressure distributions are determined by the equilibrium and transformation between the thermal dynamic energy and kinetic energy. The results provide working references for the practical ADS steam spraying condensation process in AP1000 reactor.

To better understand the flow features of pebble cluster in pebble bed, discharging of the pebble cluster were simulated by DEM. The pebble entangled cluster was composed of eight particles connected by rigid bonds and the simulated cluster models are divided into two types: axisymmetric u-particle and distorted z-particle. The simulation starts with the closed discharge outlet and the bonded clusters with different ID are randomly added from the entrance section. The pebbles fall freely and accumulate freely in the pebble bed. The discharge hole opens after all the pebbles being stationary for a period. Then the pebbles are discharged from the pebble bed under gravity. The discharging process is time-dependent bulk-movement behavior. There is not much mixing between layers on the boundary. The vertical end makes the packing loose, but also intensifies the interaction between particles due to entanglement. Consequently, the discharge features of pebble clusters of different included angles were quantified. The results show that the pebble discharging speeds depend on entanglement angle (α of u-particle and η of z-particle) and discharging outlet diameter. A large included angle may play the role of retarding or inhibiting the discharging flowrate. Therefore, the entanglement of particles component also always plays the key role of retarding the discharge.

Isolation Condenser (IC) is one of the passive core cooling systems with natural circulation flow, which is effective for safety measures against station black out. Once core uncover occurs, hydrogen generated in the core affects operating condition of ICs. To use ICs as an important safety measure not only for transient conditions but also for accident conditions, robustness of ICs against hydrogen inflow must be understood well. In this study, experiments with high pressure steam were conducted using experimental setup simulating IC, where helium was injected to simulate hydrogen effects. When the pressure in an accumulator increased high enough, natural circulation flow generated in the experimental loop. After the long-term operation, the pressure and the natural circulation flow rate achieved nearly constant. The pressure at quasi-steady state increased with increasing the helium injection amount. The pressure difference in a section including outlet side of a vertical pipe was slightly increased when helium was injected which may have indicated that the helium accumulated in the section and caused increment of the pressure loss. The startup pressure of the IC simulator also increased when helium was injected, where the driving force by the water head difference also decreased. Though long-term operations were performed after helium injection, the effect of injected helium on operating conditions of the IC remained for quasi-steady state conditions.