Search

Boiling of sea water may occur in a pressure vessel of light water (nuclear) reactors to flood the nuclear fuel as an accident management procedure. Another salt water is borated water, which will be injected into the reactor core as a neutron absorber to avoid recriticality. Boiling behavior of such salt water including these mixtures is a key to gain the confidence of accident strategy to cool down the reactor core during accidental conditions. Pool boiling experiments were conducted with a simulated 5 × 5 fuel-rod bundle with three different fluids: water, condensed (two and half times higher concentration) sea water and its mixture solution of sea water and borated water. Three-dimensional void-fraction distributions in the rod bundles were quantified by the high-energy X-ray CT system with a linear accelerator. There are no significant differences in void fraction distributions between condensed sea water and mixture solution. The void-fraction has a peak at the center on horizontal plane for all the fluids. The two salt-waters shift boiling incipience toward downstream and decrease void swell level so that vertical void-fraction profiles of the salt waters are steeper than that of water. This is because created bubbles in the salt waters were smaller than those in water. The spacer has a mixing effect to increase void fraction at upstream of the spacer and homogenize the void fraction on the horizontal plane.

Numerical analyses were conducted to replicate several tests for simulating a double-ended cold-leg large break loss-of-coolant accident (LBLOCA) in the Loss-of-Fluid Test (LOFT) using the TRACE (version 5/patch level 4) code. Analytical results by the original TRACE code were so conservative that especially a first peak of cladding temperature was estimated higher than the experimental data at the blowdown phase and subsequent temperature drop corresponding to the temporal quench was not seen. We were interested in minimum film boiling temperature (Tmin) as a heat transfer model factor estimating the quench at the moment, investigated correlation equations for Tmin in previous studies and especially focused on ones given as a function of coolant mass flow because the complicated flow transient and decompression in the core region at the blowdown phase was interpreted as having an influence on the cladding temperature behavior. There are several correlations meeting the above condition but it was revealed that they are insufficient to apply for high pressure especially. Therefore, a new term including an effect of mass flow flux and time derivative of pressure was defined and added with a proportional coefficient hypothetically to the current correlation in the TRACE code for modification. The LOFT analyses were conducted again using the modified TRACE code, and it was shown by applying roughly the same proportional coefficient to all the cases of LOFT analyses that estimation of the cladding temperature behavior was improved more precisely at the blowdown phase. Also, the transition during the phase was explained phenomenologically with the wall heat transfer mode and boiling curve.

In a severe accident of a light water reactor, the reactor pressure vessel (RPV) lower head may fail due to ablation at the vessel wall boundary involving natural convection of molten core materials. Accurate prediction of RPV lower head failure is essential for assessing severe accident progression and improving accident management because it greatly influences the subsequent ex-vessel accident progressions. However, there have been still large uncertainties about RPV lower head failure mode in the Fukushima Daiichi Nuclear Accident in 2011. The Lagrangian based MPS (moving particle semi-implicit) method has advantage of analyzing such phenomena involving complex interfaces and liquid-solid phase changes over other Eulerian mesh-based method. In the preceding study, small-scale Pb–Bi hemisphere vessel ablation experiment, with silicone oil as simulated molten core, was reproduced qualitatively by original MPS method. However, ablation mechanism associated with natural convection of the high temperature liquid could not be discussed because of significant influence of numerical discretizing error. In this study, the improved MPS method coupling corrective matrix in the particle interaction model which largely suppress the numerical fluctuation was adopted to analyze the experiment. The results show that the ablated metal relocation may enhance convective heat transfer in the downstream. As a result, ablation of the vessel wall extends from the level, close to the silicone oil surface down to the bottom of the vessel rather than previously simulated localized ablation near the silicone oil surface.

As an accident management procedure of light water (nuclear) reactors which are situated along sea shore, sea water will be injected into the reactor pressure vessel to flood the nuclear fuel which is heated by residual heat. Another salt water is borated water, which will be injected into the reactor core as a neutron absorber to avoid recriticality. Precipitation behavior of such salt water including these mixtures is a key to gain the confidence of accident strategy to cool down the reactor core during accidental conditions. Pool boiling experiments were conducted with a simulated 5 × 5 fuel-rod bundle with condensed (two and half times denser) sea water and a mixture solution of sea water and borated water. Three-dimensional salt-precipitation distributions in the rod bundles were quantified with X-ray CT system. For both solutions, salt precipitated downstream and close to the top of active fuel (TAF) height where the void fraction is the highest. The condensed sea water yields wider precipitation region in height direction than mixture solution does. Mixture solution may give localized precipitates at the same height, which is just below TAF and uniformly spread on the horizontal plane. For both solutions, dryout concentration is larger as collapsed solution level is higher. This is because that lower collapsed solution level gives longer boiling-length and higher void-fraction, which results in larger amount of salt precipitations. The proposed salt concentration is useful to evaluate dryout concentration, which is the almost constant salt concentration for heat flux levels within the experimental ranges.

Rapid thermal elevation in boiling water reactor (BWR) is an important factor for nuclear safety and there is a need to develop an analysis code for the transient phenomenon and its validation process. To evaluate the thermal property of transient boiling and its uncertainty, corroborative experimental information is crucial. In particular, the lateral propagation behavior of a vapor bubble (void) in the cross-sectional direction of fuel assembly has yet to be determined. This study evaluates the void propagation behavior in a 5 × 5 rod bundle with cross-sectional heat distribution that causes only the 3 × 3 rod bundle to generate heat; assuming rapid heating under atmospheric pressure. In this paper, using the maximum heat output applied to the nine heated rods as a parameter, from the visualization of the void behavior and the measurement of the local void fraction, the heat output conditions under circumstances where lateral propagation of voids occurs and where voids are only localized in the heated region are summarized. We quantified the time difference initially detected and the time-averaged void fraction according to the lateral propagation level.

The bubble pairing scheme was devised to quantify three-dimensional velocity of each bubble. We used two sets of Wire-Mesh Sensors to identify locations of each bubble according to bubble identification algorithm, which was developed by HZDR. The devised scheme was applied to the vertical upward air-water flow at 0.64 m/s for both air and water superficial velocities in a large diameter pipe (i.d. 224 mm). The bubble pairing scheme visualized the developing process of two-phase flow: large bubbles coalesced with each other to move toward the center, while the rest of bubbles broke up into smaller bubbles and decelerated.

We fabricated a microfluidic device with a three-dimensional (3D) structure and verified its droplet-generation performance for the stable production of droplets of around 10 μm in size. We compared the performance of the 3D device with that of conventional simple T-junction and cross-junction structures. The continuous phase sheared the dispersed phase into droplets from eight directions in the 3D device, compared with only one direction in the T-junction device and two in the cross-junction device. Droplets were produced efficiently over a wide range of fluid properties and flow conditions with the 3D device, unlike with the two conventional planar devices. Fluidic experiments were conducted using mineral oil with a surfactant as the continuous phase, deionized (DI) water as the dispersed phase, and DI water with glycerin to change the viscosity of the dispersed phase. The minimum droplet length was 47.2 μm in the T-junction device, 39.0 μm in the cross-junction device, and 22.4 μm in the 3D device when using a water and glycerin mixture with a viscosity of 9.0 mPa·s. Compared with the conventional devices, smaller droplets were produced using our 3D device, indicating that it has excellent droplet-generation performance.

In order to obtain high-resolution data for modelling of boiling two-phase flow and its validation, we designed and constructed a test loop with a vertical annulus flow path and conducted subcooled boiling experiments to investigate subcooled bubble incipience and its development process under atmospheric condition. Three kinds of the state-of-the art measurement techniques were applied to quantify key parameters such as radial and vertical distributions of void fraction, bubble velocity, interfacial area concentration (IAC), Sauter mean diameters, high-resolution temperature distribution on rod surface, bubble transport behavior, and turbulent velocity components as well as onset of nucleate boiling (ONB), and onset of significant void (OSV).

Natural circulation is a key technology for developing the molten core cooling system without an external power source from the lessons of the severe accident at Fukushima-Daiichi Nuclear Power Station. This study is devoted to quantify the void fraction which is an important parameter for the driving force of natural circulation flow, and to evaluate the effect of the void fraction correlation on the prediction accuracy of the natural circulation flow rate. Test was conducted at atmospheric pressure and room temperature, using the upward air–water two phase flow. Vertical tubes with an inner diameter of 36 and 25 mm were used as the test section. The void fraction was measured by three different methods: quick-closing valve method, pressure drop method, and conductive void-probe method. The following conclusions are obtained from this study: (1) The data of the natural circulation flow rate, void fraction and pressure drop for the upward air–water two phase flow at atmospheric pressure and room temperature were obtained to develop and verify the new model. (2) By improving the void correlation, it was found that the prediction accuracy of the natural circulation flow rate could be improved by about 10% to 5%, that is, the prediction error can be halved in the range of this study. (3) The natural circulation flow rate for 25 mm test section was saturated with increasing the air flow rate at higher air flow condition. The model cannot predict this tendency. From the point of design of the actual molten core cooling system, the model improvements in this region are necessary in the future.

In a severe accident of a light water reactor, the corium spreading behavior on a containment floor is important as it may threaten the containment vessel integrity. The Moving Particle Semi-implicit (MPS) method is one of the Lagrangian particle methods for simulation of incompressible flow. In this study, the MPS method is further developed to simulate corium spreading involving not only flow, but also heat transfer, phase change and thermo-physical property change of corium. A new crust formation model was developed, in which, immobilization of crust was modeled by stopping the particle movement when its solid fraction is above the threshold and is in contact with the substrate or any other immobilized particles. The VULCANO VE-U7 corium spreading experiment was analyzed by the developed MPS spreading analysis code to investigate influences of different particle sizes, the corium viscosity changes, and the “immobilization solid fraction” of the crust formation model on the spreading and its termination. Viscosity change of the corium was influential to the overall progression of the spreading leading edge, whereas termination of the spreading was primarily determined by the immobilization of the leading edge (i.e., crust formation). The progression of the leading edge and termination of the spreading were well predicted, but the simulation overestimated the substrate temperature. Further investigations may be necessary for the future study to see if thermal resistance at the corium-substrate boundary has significant influence on the overall spreading behavior and its termination.

The fuel support piece in a boiling water reactor (BWR) is used to brace fuel assemblies. The channel within the fuel support piece is determined to be a potential corium relocation path from the core region to the lower head during the severe accident of BWR. In the present study, the improved ∗∗∗Moving Particle Semi-implicit (MPS) method was adopted to simulate the flow and solidification behavior of the melt in a fuel support piece. The MPS method was first validated against the Pb-Bi plate ablation test that was performed by CRIEPI. The predicted ablation mass of the plate agreed well with the experimental results. Then the flowing and freezing behaviors of molten stainless steel (SS) and zircaloy in the fuel support piece were simulated by MPS method with a three dimensional particle configuration, respectively. In this study, the flow and solidification behavior of SS was simulated first. After all the SS passed through the channel, the flowing behavior of Zr in the fuel support piece was simulated. The simulation results indicated that the crust layer formed on the inner surface of the fuel support piece during the melt discharging process. The fuel support piece was plugged by the solidified zircaloy particles in the lower initial temperature case. The fuel support piece kept intact in all the calculation that were performed under the assumed order of melt injection. The present results could help to reveal the progression of a BWR severe accident.

We have established an experimental system to visualize a droplet flow in a simulated BWR fuel sub-channel optically and measure the diameter and velocity of droplet after passing through a spacer. For representative spacers of ferrule and grid type, an effect of them on downstream droplets was evaluated with the experimental system. When a ferrule type spacer was simulated and implemented in both the center and side sub-channel, the vertical velocity of droplets got faster especially in the range of small diameter compared to the case of no spacer. When a grid type spacer was simulated and implemented in the center sub-channel especially, a large dispersion of vertical velocity of droplets occurred especially in the range of small diameter compared to the case of no spacer. By a computational fluid dynamics analysis for gas phase flow to drive the droplets in the sub-channel, it was confirmed qualitatively that the characteristics of droplet behavior observed in this experiment were dependent on the structure and geometry of spacer and sub-channel. Furthermore, it was revealed that a relation between a droplet diameter and velocity can be organized with a non-dimensional function derived from a momentum equation of particle in driving fluid and its drag coefficient has linear correlation with a gas Froude number.

Sea water shall be injected into water-cooled nuclear reactors during severe accidents, which are located along coastal side to flood the nuclear fuel, which is heated by residual heat. Precipitation growth to narrower flow path area is a key to gain the confidence of accident mitigation procedure to cool down the reactor core during accidental conditions. A pool boiling experiment was conducted with a simulated 5 × 5 full-height BWR fuel-rod bundle with condensed (two and half times higher concentration) sea water. The temperature on the center rod surface in the top spacer rose rapidly, since the flow area inside the top spacer was filled with the precipitated salt. Dryout below the top spacer escalated temperatures of the heater surface. On the other hand, the heater above the top spacer was cooled stably by pool boiling. An example calculation estimates that the dryout due to salt precipitation may occur 19 h after sea water injection for an ABWR, which had operated at 3.926 GWt for 13 months on the basis of critical dryout concentration of 50 wt%.

During a severe nuclear power plant accident, the release of fission products into containment and an increase in containment pressure are assumed to be possible. When the containment is damaged by excess pressure or temperature, radioactive materials are released. Pressure suppression pools, containment spray systems and a filtered containment venting system (FCVS) reduce containment pressure and reduce the radioactive release into the environment. These devices remove radioactive materials via various mechanisms. Pressure suppression pools remove radioactive materials by pool scrubbing. Spray systems remove radioactive materials by droplet−aerosol interaction. FCVS, which is installed in the exhaust system, comprises multi-scrubbers (venturi-scrubber, pool scrubbing, static mixer, metal−fiber filter and molecular sieve). For the particulate radioactive materials, its size affects the removal performance and a number of studies have been performed on the removal effect of radioactive materials. This study has developed a new means of evaluating aerosol removal efficiency. The aerosol number density of each effective diameter (light scattering equivalent diameter) is measured using an optical method, while the decontamination factor (DF) of each effective diameter is evaluated by the inlet outlet number density ratio. While the applicable scope is limited to several conditions (geometry of test section: inner diameter 500 mm × height 8.0 m, nozzle shape and air-water ambient pressure conditions), this study has developed a numerical model which defines aerosol DF as a function of aerosol diameter ( d ) and submergences ( x ).

Steam explosion experiments with a melt layer spreading at the bottom of a shallow water pool, namely the PULiMS-E6 and SES-S1 by KTH, Sweden, were simulated by the steam explosion simulation code, JASMINE. The observed impulses in the experiments were successfully reproduced by simulations with assumed premixing conditions. With those simulation results, the adequacy of the kinetic energy evaluation method used for the experiments were examined by comparison of the kinetic energy directly obtained in the simulation, E k , and the one evaluated based on the impulse and the water mass limited to the center area above the premixing zone, E kic . It showed that the impulse based kinetic energy evaluation gives about five times overestimation. The impact of the water pool geometry on the validity of the impulse based kinetic energy evaluation method was further examined by a parametric study with variations of the pool geometry in the simulations of PULiMS-E6 and SES-S1 as well as high pressure bubble expansion simulations. The results for the relation of E kic / E k and the geometric factors were consistent between the cases for the experiments and the bubble expansion. The results showed that: (1) for the shallow water pool regime, E kic / E k shows a trend of convergence to 4–5, (2) for deep water pool regime, the impulse based kinetic energy evaluation with the whole water mass, E ki , rather than E kic , gives a good estimation. A set of empirical formulas was obtained for E kic / E k .

In order to evaluate residual the stress distribution in oxides formed on zirconium alloys, synchrotron X-ray diffraction (XRD) was performed on the oxides formed on Zircaloy-2 after autoclave treatment at a temperature of 360° C in pure water. The use of a micro-beam XRD and a micro-sized cross-sectional sample achieved the detailed local characterization of the oxides. The oxide microstructure was observed by TEM following the micro-beam XRD measurements. The residual compressive stress increased in the vicinity of the oxide/metal interface of the pre-transition oxide. Highly oriented columnar grains of a monoclinic phase were observed in that region. Furthermore, at the interface of the post-first transition oxide, there was only a small increase in the residual compressive stress and the columnar grains had a more random orientation. The volume fraction of the tetragonal phase increased with the residual compressive stress. The results are discussed in terms of the formation and transition of the protective oxide.

The interfacial behavior of upward liquid film flow is an important phenomenon to evaluate interfacial transfer accompanying the entrainment and deposition of droplets. This research focuses on a vertical annular channel, and an air–water liquid film flow experiment was conducted under atmospheric pressure conditions. The diameters of inner and outer pipes in the annular channel were 12 and 18 mm respectively. The experiment featured multi-point electrode sensors installed in both the inner and outer pipe surface at the same height, and the ability to measure the liquid film distribution on both surfaces in the annular channel simultaneously. As for the sensor structure, 10 × 32 measuring points were arranged in a lattice pattern on the sensor surface and the spatial resolution was 2 × 2 mm, hence the liquid film thickness distribution could be measured rapidly, at over 1250 slices per second. Since the sensor was manufactured by a flexible multilayer substrate, it was applicable to a cylindrical channel surface. In the experiment, water was supplied from the inner pipe surface and uniformly distributed in the circumferential direction, whereupon liquid film distributions were measured 300 mm downstream from the water supply position. The time series data of the liquid film distribution demonstrated circumferential distributions of liquid film thickness and interfacial wave velocity. When the superficial gas velocity was smaller than 20 m/s, a liquid film formed on both inner and outside pipe surfaces, regardless of the superficial liquid velocity. With increasing superficial gas velocity, the film thickness of the outer pipe surface became thinner than that of the inner pipe surface. Measurement of the liquid film thickness on both surfaces of the annular channel also showed that a liquid slug with wavelength of several millimeters passed concurrently through both surfaces in the annular channel.

Given the severe accident of a light water reactor (LWR), stratification and solidification/melting are important phenomena in melt corium behavior within the reactor lower head, influencing the decay heat distribution and ablation of penetration tube and vessel wall. Numerical calculation is a necessary and effective approach for mechanistic study of local melt corium behavior. In this study, the improved moving particle semi-implicit (MPS) method was applied for investigating the stratification and solidification/melting phenomena. The implicit viscous term calculation technique and stability improvement technique were adopted to enable MPS to simulate the stratification process of materials with high viscosity in phase transition stage. The solid–liquid phase transition model was also coupled with MPS method. The validation experiment was carried out with low-melting-point metal tin and NeoSK-SALT. The layer configurations and temperature profiles obtained from MPS calculation showed good agreement with the experimental results. Meanwhile, the calculation results indicated that the material freezing behavior could affect the layer formation, and the layer configurations also significantly influenced the temperature profiles and heat flux distributions. The present results demonstrated that MPS method has the capacity to understand the local melt behavior in detail that is relevant to stratification and phase transition.

An analytical model was developed to describe thermal stratification in a primary containment vessel (PCV) and transient thermal-hydraulics coupled with a reactor pressure vessel (RPV) using TRACE code version 5.0 patch level 4. Geometries of a dry well (D/W) and a suppression chamber (S/C) were represented by a nodalization of TRACE code to simulate multi-dimensional flow in the PCV. An additive loss coefficient (so called ‘K-factor’) was focused as a sensitivity parameter to limit flow rate in a pool. For the first step, a validation analysis was conducted against a steam discharge experiment of S/C. The TRACE result was in good agreement with the measurement and showed a thermally-stratified temperature distribution in the S/C pool. For the second step, an analysis to simulate the accident at Fukushima Daiichi Unit 3 power plant (1F3) was conducted. It was proved to be able to explain the pressure increase in the PCV at the beginning of accident by demonstrating thermal stratification in the S/C pool. Sensitivity study revealed an optimal K-factor value for a macroscopic viscous drag in a liquid phase fluid to demonstrate thermal stratification in a pool.

Void fraction (i.e., the volume fraction occupied by gas) is a key parameter for determining the coolability and neutron-moderating performance of a water-cooled nuclear reactor. To develop computational multi-fluid dynamics models for determining the void-fraction distribution, experimental data of comparable quality are required. We have developed a high-energy X-ray computed tomography (CT) system to acquire three-dimensional void-fraction distributions. The CT system comprises a linear-accelerator-driven high-energy X-ray source and a linear detector array. We quantified a boiling two-phase flow in a 5 × 5 heated rod bundle at high pressure, simulating a fuel-rod bundle in a boiling water reactor (BWR). Because the axial travel of the CT system is 4 m and includes the entire BWR fuel-rod bundle, we optimized the CT imaging conditions and reconstruction method for rod-bundle visualization to reduce uncertainties due to density fluctuations in the boiling flow and imaging artifacts. We conducted a boiling experiment at a low flow rate and low thermal power and acquired three-dimensional distributions of the void fraction over a wide pressure range of 0.1–7.2 MPa. The experiment provided three-dimensional void-fraction distributions with high spatial resolution, especially in subchannel regions surrounded by rods, and the results are suitable for validating three-dimensional thermal-hydraulic analysis codes.

Stratification behavior is of great significance in the late in-vessel stage of core melt severe accident of a nuclear reactor. Conventional numerical methods have difficulties in analyzing stratification process accompanying with free surface without depending on empirical correlations. The Moving Particle Semi-implicit (MPS) method, which calculates free surface and multiphase flow without empirical equations, is applicable for analyzing the stratification behavior of fluids. In the present study, the original MPS method was improved to simulate the stratification behavior of two immiscible fluids. The improved MPS method was validated through simulating classical dam break problem. Then, the stratification processes of two fluid columns and injected fluid were investigated through experiments and simulations, using silicone oil and salt water as the simulant materials. The effects of fluid viscosity and density difference on stratification behavior were also sensitively investigated by simulations. Typical fluid configurations at various parametric and geometrical conditions were observed and well predicted by improved MPS method.

An accurate and detailed experimental database is crucial for modeling the multidimensional two-phase flow and for validating the numerical calculation results. In particular, a two-phase flow in the rod bundle flow channel is so complicated that it is difficult to measure a multidimensional flow structure. Based on the available reference, a point-measurement sensor for acquiring void fractions and bubble velocity distributions do not infer interactions of the subchannel flow dynamics, such as a cross flow and flow distribution, etc. In order to acquire multidimensional two-phase flow in a 10 × 10 rod bundle with an o.d. of 10 mm and length of 3110 mm, a new sensor consisting of 11 × 11 wire and 10 × 10 rod electrodes was developed. The electrical potential in the proximity region between the two wires creates a void fraction in the central subchannel, like a so-called wire-mesh sensor. A unique feature of the devised sensor is that the void fraction near the rod surface can be estimated from the electrical potential in the proximity region between one wire and one rod, meaning the additional 400 points of void fraction and phasic velocity in the 10 × 10 rod bundle can be acquired. The devised sensor demonstrates multidimensional flow structures, i.e. void fraction, phasic velocity, sauter mean diameter and interfacial area concentration distributions. Acquired data exhibit complexity of two-phase flow dynamics in a rod bundle flow channel, such as coalescence and the breakup of bubbles in transient phasic velocity distributions.

A gas–liquid two-phase flow in a large diameter pipe exhibits a three-dimensional flow structure. The wire-mesh sensor (WMS) can acquire a quasi-three-dimensional void fraction distribution. Furthermore, the WMS can acquire a phasic-velocity distribution on the basis of the time lag of void signals between both sets of WMS. Previously, the acquired phasic velocity was one-dimensional distributions. The authors propose a method to estimate the three-dimensional phasic-velocity distribution from the same WMS data. A three dimensional velocity vector was determined on the basis of cross-correlation analysis. The flow direction is determined by the WMS measuring-point combination, whereby the cross-correlation coefficient between both sets of WMS measuring points reveals the peak. In addition, the flow structure can be extracted by size on the basis of a wavelet analysis. The proposed method was applied for two sets of 64 × 64 mesh sensors in an air–water flow in a vertical pipe with inner diameter of 224 mm. The proposed method can successfully visualize a swirl flow structure where large and small bubbles tend to move respectively in inward and outward directions in turn.

Cladding materials with superior corrosion resistance and anti-hydrogen pickup have been developed for high burnup nuclear fuel. We have suggested a surface modification of the cladding materials for this purpose and invented a new surface modification method “Fresh-Green”. The Fresh-Green treatment oxidizes and carbonizes a material surface in the same process. Zircaloy-2 with the Fresh-Green treatment showed the improvement of corrosion resistance in autoclave tests. In order to investigate the effect of surface modifications on the corrosion resistance, a synchrotron radiation experiment and a TEM observation were performed on different oxide layers formed on Zircaloy-2. The oxide layers were formed by air-oxidation, an autoclave test and the Fresh-Green treatment. Crystal structures of all the samples were transformed as Zr > Zr3O > tetragonal ZrO2 > monoclinic ZrO2 from the matrix to the surface. Columnar grains of monoclinic zirconia were arranged unidirectionally in the Fresh-Green oxide layer treated at a low temperature. Diffusing capacity for oxygen influenced the crystal structure of the oxide layers.

Experiments were conducted to investigate two-phase flow instabilities due to flashing in a boiling natural circulation loop with a chimney at low pressure. The SIRIUS-N facility was designed to have non-dimensional values nearly equal to those of typical natural circulation boiling water reactor (BWR). The observed instability is suggested to be flashing-induced density wave oscillations, since the oscillation period correlated well with the passing time of single-phase liquid in the chimney section regardless of system pressure, heat flux, and inlet subcooling. Stability maps were obtained in reference to the inlet subcooling and the heat flux at the system pressures of 0.1, 0.2, 0.35, and 0.5 MPa. The flow became stable below a certain heat flux regardless of the channel inlet subcooling. The stable region enlarged with increasing system pressure. Thus, the stability margin becomes larger in a startup process of a reactor by pressurizing the reactor sufficiently before heating according to the stability map.

Improving the limit of boiling heat transfer or critical heat flux requires that the cooling liquid can contact the heating surface, or a high wettability, highly hydrophilic heating surface, even if a vapor bubble layer is generated on the surface. In our previous study, contact angle, an indicator of macroscopic wettability, of a water droplet on metal oxide at room temperature was measured by image processing of the images obtained by a CCD video camera. The results showed that the surface wettability on metal oxide pieces of titanium, zircaloy No. 4, SUS-304, and copper was improved significantly by the radiation induced surface activation (RISA) phenomenon. To delineate the effect of RISA on heat transferring phenomena, the Leidenfrost condition and quenching of metal oxides irradiated by γ-rays were investigated in this study. In the Leidenfrost experiment, when the temperature of the heating surface reached the wetting limit temperature, water–solid contact vanished because a stable vapor film existed between the droplet and the metal surface; i.e., a Leidenfrost condition obtained. The wetting limit temperature increased with integrated irradiation dose. After irradiation, the wet length and the duration of contact increased, and the contact angle decreased. In the quenching test, high surface wettability, or a highly hydrophilic condition, of a simulated fuel rod made of SUS was achieved, and the quenching velocities were increased up to 20–30% after 300 kGy 60 Co γ-ray irradiation.

Effects of mitigation and suppression of vapor explosion have been investigated by adding a surfactant, polymer, and neutral salt into the water droplet impinging onto a molten alloy pool surface. This configuration was selected to attain good reproducibility and visibility, because premixing events prior to the triggering restrained. In addition, an adequate amount of a surfactant can be adsorbed at the vapor/liquid interface compared to that in the case of conventional configurations such as a molten alloy droplet injecting into a water pool. Dilute anionic and nonionic surfactant aqueous solutions did not affect the triggering conditions or the pressure pulse generated by vapor explosion, even if the surfactant was added up to a density 25 times higher than the critical micelle concentration. Polymeric additives, which increase the viscosity of a fluid, have little suppression effect on vapor explosion. Spontaneous vapor explosion was, however, suppressed by a 200 wppm polyethylene glycol (molecular weight of 4×10 6 ) solution, since deposition of the solute due to cloudy-point phenomenon may stabilize the vapor film and prevent the solution from mixing finely. In order to exert this effect, molecular weight should be heavier so that a cloudy-point temperature is below the boiling point at the tested system pressure. Additionally, this threshold concentration became denser as the impingement velocity increased. Thus, a denser concentration and heavier molecular weight should be used to suppress vapor explosion when the vapor film may be subjected to large inertia and/or external force. When a neutral salt was added, the initial molten alloy temperature range where vapor explosion occurred shifted to a higher temperature and became wider. Vapor explosion was observed on the molten zinc surface, which does not explode spontaneously for water.

An analytical study is presented on the thermo-hydraulic stability of a boiling natural circulation loop with a chimney at low pressure start-up. The effect of flashing induced by the pressure drop in the channel and the chimney due to gravity head on the instability is considered. A method to analyze linear stability is developed, in which a drift-flux model is used. The analytical result of a stability map agrees very well with the experimental one obtained in a previous report. Instability does not occur when the heater power is too low to generate voids in the chimney and only natural circulation of single phase can be induced. Instability tends to occur when boiling occurs only near the chimney exit due to flashing. This instability phenomenon has some similarities with density wave oscillation, such as the phase difference of temperature between the boiling region and non-boiling region, and the oscillation period which is near to the time required for fluid to pass through the chimney. However, there are also some differences from density wave oscillation, such as the boiling region is very short, and pressure fluctuation can affect void fraction fluctuation.