Steam injector (SI) are attracting attention as countermeasures against severe-accident in nuclear reactors. It is a static jet pump which operates using driving force to draw steam and water by internal pressure being reduced by direct contact condensation of these two fluids. In addition, capability of SI as a heat exchanger with high heat-transfer is expected. The absence of a drive unit such as an external power supply and rotating machine is significant characteristic of SI, and it can be expected to suppress the cost of installation and maintenance. It is also possible to produce a discharge pressure higher than the inlet pressure. From these facts, SI is expected to be applied as a static safety system that can cool the reactor core even if power lose at the nuclear power plant. Although SI has been used for steam engines since long ago, the mechanism of its operation has not yet been clarified. Thus, elucidation of the mechanism of operation of SI is indispensable for introduction to a nuclear power plant. A one-dimensional analytical model which predicts the operating characteristics assuming full condensation and evaluated discharge pressure is constructed (Narabayashi et al., 1996). In addition, from detailed observation, it was confirmed by that there is a boundary of luminance in the diffuser section (Abe et al., 2012). This is considered as the boundary where the two-phase flow condenses. However, this phenomenon is not considered in the current analysis model. The aim of this research is to clarify the flow structure in the diffuser section of SI. For that purpose, the state of the diffuser section of the transparent SI test part was observed with a highspeed camera, and the pressure at each point in it was measured simultaneously. The boundary of the luminance is confirmed to approach the throat as closing the back-pressure valve. In addition to this boundary, it was confirmed that the bright region intermittently propagated downstream. This phenomenon is supposed to be caused by pressure increasing, and the propagations assumed as a pressure wave moving at the sound speed. Thus, void fraction is estimated by calculating this propagation speed with image processing. Furthermore, experiments were carried out using three types of large, medium and small test parts, respectively. From the above results, the internal flow structure in the SI diffuser section was discussed.

In order to maintain the reactor core integrity in case of severe accidents, multi-layered installation of backup safety systems with the passive feature is crucial, as was evident from the Fukushima Daiichi accident. The supersonic steam injector (SI), also known as a passive jet pump driven by high-pressure steam and water, is capable of operating without external power supplies nor rotating machinery. Along with the Reactor Core Isolation Cooling (RCIC) system, the SI can possibly be considered as one of the candidates as a backup safety system for current and next-generation reactors. Although the fundamental experiments on SI have been carried out for past few decades, its operation characteristics are still not well understood. Especially, its stable operation condition as a passive jet pump hasn’t been investigated thoroughly. Our study shows that the SI’s operating range is highly dependent on inlet water flow rate and steam pressure. In addition, due to the direct contact condensation between steam and water jet within SI unit, the water jet stability may influence SI’s operation. In order to deepen understanding of SI’s operation characteristics, this study aims to investigate on the inlet water temperature effect, which has a direct influence on steam-water condensation as well as the water jet stability within the mixing nozzle section of SI unit.

The Experimental Advanced Superconducting Tokamak (EAST) is one of the fully superconducting tokamak, its aim at the long-pulse operation (1000s) to study the physics of steady-state operation for nuclear fusion sciences. In order to support the steady-state operation and physical research, the high power neutral beam injection (NBI) system need to be employed on the EAST for the plasma heating and current driving. According to the scientific study schedule of the EAST, the designed NBI system includes two beam lines which will be constructed in two phases. Each beam line will deliver a deuterium neutral beam with beam energy of 50–80 keV with beam duration of 10–100 s. Each beam line has the maximum beam power of 4MW. The high current ion source is one of the most important parts in the high power NBI. A hot cathode positive ion based source was developed for EAST-NBI, which shown in Fig. 1. The ion source contains a bucket hot cathode arc chamber with 650 mm long, 260 mm width and 300 mm depth. There are 32 hairpin filaments with diamond of 1.5 mm and 160 mm long to supply primary electrons. A tetrode type accelerator with slit type used to extract the ions from the plasma and accelerated to the desired energy. The beam extraction area is 120 mm × 480 mm (can be changed) with beam transmittance of 60 %. The designed beam species is deuterium with beam power of 2–4 MW and beam energy of 50–80 keV and beam pulse length of 10–100 s. The ion source needs to be conditioning before operation on the EAST-NBI. An ion source test bed was designed and developed for the ion source performance tests and the optimization verification. The characteristics of ion source were tested with hydrogen beam, and each ion source should achieve 4MW ion beam with beam energy of 80 keV. The optimum beam perveance and arc efficiency were analyzed too. The optimum beam perveance was 2.8 μp with beam energy of 50 keV and the arc efficiency was 0.55A/kw. Long pulse operation was one of the requests of EAST ion source. The real-time feedback control method was employed and can got stable plasma and ion beam. The beam extraction was tested to achieve 100 s on the test bed. Consider the high power deposited on the calorimeter, the beam was modulated with suitable frequency and duty ratio. When the conditioning finished, the ion sources were moved to the EAST-NBI. The deuterium beam was extracted and injected into the EAST plasma. Details of the performance of positive ion source on the test bed and EAST-NBI will be presented.

In the framework of the European Commission LEADER project, an experimental campaign of seven tests was performed in the LIFUS5/Mod2 facility, at ENEA CR Brasimone, for investigating the postulated Steam Generator Tube Rupture (SGTR) event in a relevant configuration for the Spiral-Tube Steam Generator (STSG) of the European Lead Fast Reactor (ELFR). The LIFUS5/Mod2 facility is composed by a water tank of 15 L injecting subcooled water up to 200 bar into the reaction tank of 100 L (420 mm of diameter), which is connected by a 3 inch pipe to the dump tank of 2 m 3 . A dedicated test section was designed, assembled and implemented in the reaction tank. It is composed by 188 tubes, vertically disposed with triangular pitch inside a cylindrical support. This tube bundle is representative of a portion of the STSG of ELFR. The cylindrical support is closed at the lower and upper end by two tube plates and has a perforated lateral shell (300 mm of diameter and 400 mm high). The reaction tank is filled by Lead-Bismuth Eutectic alloy (LBE) at 400°C up to the top tube plate, with an argon cover gas at about 2 bar. The water is injected at about 180 bar and 270°C through the central tube, at middle height of the bundle. The water-LBE interaction is characterised by high quality data acquisition system: 6 fast Pressure Transducers (PTs) working at 10 kHz for precisely characterize the first narrow injection peaks, 70 low constant time Thermocouples (TCs) to understand the vapour evolution path and 13 strain gages (SGGs) for measuring the strain of the bundle and main vessel. The overall LEADER experimental campaign is constituted by seven tests, divided in three series (B1, B2 and B3), characterized by different injection orifice diameters of 4, 8.9 and 12.6 mm, respectively. This paper presents the experimental results of the first two tests of series B2 (B2.1 and B2.2) having 8.9 mm of injection orifice. The first test analysed showed a first narrow pressure peak of about 32 bar, some milliseconds after the cap rupture instant. The following pressurization due to the evaporation of water entered into the reaction vessel was of an analogues magnitude for both the tests (about 50 bar) and lasted some tenths of second. The water/LBE interaction lower temperature was reached on the inner ranks of tubes, about 150°C. The outer rank was cooled down to about 300°C. The strain gage measurements showed a decreasing deformation on the tubes toward the outer positions. No ruptures were observed on tubes surrounding the injector. The amount of LBE transported into the dump tank was strongly dependent on the LBE level in the reaction tank at the start of the tests and about 200 kg.

From the viewpoint of an importance of safety, the nuclear power plant should be managed to prepare severe accidents. The performance of safety dropped by an accident is strongly to be minimized during the situation of station blackout. The installation of a steam injector (SI) into the nuclear power plant has long been expected. In the SI, the steam condenses due to the direct contact at the surface of water jet, resulting in the force attracting water. The force drives the circulation of an amount of coolant water. SI also works as a reactor condenser thanks to its high efficient performance during the condensation. Because any external forces to circulate water and steam are not required, SI can be operated without the electric powers. The structure of SI is similar to a convergent-divergent nozzle. After the flow acceleration at a throat, the discharged pressure is expected to exceed the inlet pressure. Owing to its quite simple structure, the reduced cost of installation and maintenance is also expected. The following previous studies for four cases of throat diameter clarified two-phase flow structures and heat transfer characteristics in water jet and performance of SI: (i) Narabayashi et al. (2000) examined for 5.5 and 6.5 mm in diameter; (ii) Osakabe et al. (2004) for 3.4 mm; (iii) Koizumi et al. (2006) for 4 mm; (iv) Abe et al. (2014) for 4, 6.5, and 8 mm. Although these clarified the operative state which formed a water jet, operative condition was not elucidated. Furthermore, the scale effect for various diameters of SI has not been discussed in detail. The aim of this study is to clarify scale effect of a test section on operating criteria and performance. Experiment was performed to clarify the scale effect by using three types of throat diameters: 4, 6.5, and 8 mm. As a result, three formations of a water jet were observed: (i) formation, (ii) incomplete formation, and (iii) no formation. We proposed a classification which enables us to categorize complex flow patterns into five regimes. We clarified the operating criteria of them by comparing water flow rate with steam flow rate. SI did not form a water jet on the condition with low steam flow rate. The suppling water was stopped, and only steam was supplied to the test section for the condition that steam latent heat was larger than subcooled water enthalpy.

Steam Injector (SI) is a passive pump activated by steam and water, and it does not require any external power supplies or rotating machineries. Moreover, SI has high ability as a heat exchanger by undergoing direct contact condensation mechanism. From these characteristics, SI has a capability to be applied as a passive safety system in the nuclear power plant. In the present study, experiments targeting the operating range and pump performance were carried out to obtain SI’s detailed characteristics under following supply conditions; inlet steam pressure was 0.02 ∼ 0.81MPaG, inlet water flow rate was 0.21 ∼ 0.80 kg/s. In the former experiment, operating range was investigated by changing inlet conditions, and the influence of steam inlet pressure and water inlet flow rates on SI operating range was tested. In the latter experiment, pump performance of SI was evaluated by investigating the maximum discharge pressure in each inlet condition by using back pressure valve. From the results of experiments, it was confirmed the operating range of SI was limited by supplied steam pressure and supplied water flow rate, and some clear trends were found in the operation boundary map. In addition, SI could discharge water at least 1.2 times higher than inlet steam pressure under above-mentioned conditions, which verified SI’s capability to be applied for the nuclear power plant as a core cooling system. And furthermore, existing 1D analysis model’s predictive capability was tested based on these experimental results.

In the framework of Gen IV Heavy Liquid Metal Fast Reactors (HLMFRs) development, research activities are carried out for improving the Advanced Lead Fast Reactor European Demonstrator (ALFRED) project. In particular, the functioning and safety of the Steam Generator Bayonet Tube (SGBT) of ALFRED reactor (developed by ANSALDO Nucleare) needs to be experimentally investigated and numerical tools have to be verified and validated against reliable high quality data of SGBT stationary and accidental transient conditions. At the ENEA CR Brasimone, a full scale portion of the SGBT of ALFRED, composed by seven bayonet tubes arranged as triangular pitch inside a hexagonal shroud, was designed and assembled. It is called Heavy liquid mEtal-pRessurized water cOoled tube (HERO) and it was implemented as heat exchanger in the Integral Circulation Experiment (ICE) test section. The ICE-HERO test section is presently ready to be set in the large pool CIRCE facility for carrying out integral tests in Lead Bismuth Eutectic alloy (LBE) pool at about 400°C. ICE test section is equipped with a fuel pin simulator (FPS) of 37 electrically heated pin for an overall maximum thermal source of 1 MW and an argon injector above FPS for performing enhanced circulation (gas lift). The hot LBE (about 480°C) exiting from FPS moves upwards passing through the thermally insulated riser pipe and reaches an upper small tank, isolated from the outer colder LBE pool, called separator, from which descends shell side of the HERO heat exchanger and returns to the LBE pool, closing the integral circulation. The secondary side of HERO component (7 bayonet tubes) is fed by about 0.35 kg/s of water at 180 bar and 335°C. Each bayonet tube is composed by three concentric tubes. In the first (inner) tube, liquid water flows downwards in subcooled conditions and thermally insulated from outer parts. Between the first and second tube (annular cross-sectional area) water moves upwards and it is heated up to superheated steam condition. The annular gap between the second and third tube is filled by helium at 5 bar and steel powder for monitoring the possible rupture of the third or second tube and do not reduce excessively the heat transfer coefficient. The HERO component is highly instrumented for characterising the heat exchanger capabilities of the bayonet tubes in stationary and transient conditions. An overall number of 65 thermocouples are implemented, acquiring water, LBE and helium temperatures. Two pressure transducers monitor the feed water and steam line pressure. Each bayonet tube has four differential pressure transducers for measuring the pressure drops on the descending, ascending, and total water path and computing the water mass flow rate of each tube by a calibrated orifice. A Coriolis flowmeter is adopted to precisely determine the feed water mass flow rate. Moreover the signals of thermocouples, pressure transducers and bubble tubes implemented in the ICE test section will be acquired for an exhaustive analysis of the tests. A sensitivity analysis performed by RELAP5 code was carried out for defining the geometry and material to be adopted for the bayonet tube construction. The materials foreseen to be implemented have undergone modifications, AISI powder instead of SiC, air gap in place of the thermal insulating paint and AISI pipe instead of T91. The RELAP5 thermodynamic analysis predicted that the bayonet tube in updated configuration is able to provide superheated steam at about 400°C. The expected performance of HERO-TS is retained acceptable for the application to a prototypic unit and can be scaled to ALFRED.

In this paper a new technique to handle solid radioactive materials inside a liquid matrix is presented. The conceptual design of the device profits of the experience and know-how gained in decontamination procedures. The proposed system makes use of an ejector for the suction of a water-highly radioactive swarf mixture from the purifiers pool of the Italian E. Fermi NPP and moving it in a suitable container for the subsequent conditioning. A dedicated circuit with an ejector to demonstrate the feasibility of the method was realized. A minimum inlet flow was found to have swarf suction. No swarf packaging conditions were observed in the performed tests, demonstrating the feasibility of the method.

In the framework of the EC FP7 LEADER project, an experimental campaign was performed in the LIFUS5/Mod2 facility, at ENEA CR Brasimone, for investigating the postulated Steam Generator Tube Rupture (SGTR) event in a relevant configuration for the spiral tube Steam Generator (SG) of the European Lead Fast Reactor (ELFR). Two tests are analysed. The LIFUS5/Mod2 facility implemented a test section composed by 188 tubes, vertically disposed with triangular pitch, in a shell closed by top and bottom flanges and having a perforated cylindrical wall. The central tube injected water at about 180 bar and 270°C, at middle height of the tube bundle, in the reaction tank partially filled by Lead-Bismuth Eutectic alloy (LBE) at 400°C with an argon cover gas at about 2 bar. It was connected to a 2 m 3 dump tank, due to the high injection pressure. In the reaction tank fast instrumentation was set: 6 fast Pressure Transducers (PTs) acquiring data at 10 kHz for precisely characterize the first injection peaks; 70 low constant time Thermocouples (TCs) to understand the vapour evolution path; and 13 strain gages (SGGs) to measure the strain of the bundle and main vessel. The first test analysed showed a first pressure peak of about 25 bar, due to pressure wave propagation at the cap rupture instant. It did not appear in the second test as consequence of a leakage from the cap before the complete rupture. The following pressurization caused by the entering of water into the reaction vessel was of an analogues magnitude for both the tests (about 30 bar). The water/LBE interaction lower temperature was reached on the inner ranks of tubes, about 160°C. The outer rank was cooled down to 340°C. The strain gage measurements showed a decreasing deformation on the tubes toward the outer positions. No ruptures were observed on tubes surrounding the injector. The amount of LBE transported into the dump tank was strongly dependent on the LBE level in the reaction tank at the start of the tests.

Steam injector (SI) is a passive jet pump with a converging-diverging structure. A SI operates without an electrical power source by direct contact condensation of a supersonic steam flow and a subcooled water jet in a mixing section. Furthermore, there are advantages that a SI has high heat-transfer performance and discharges water at high pressure. Therefore a SI is expected to apply to the safety system that is able to condense steam efficiently and inject water into a core reactor when severe-accident occurs in a nuclear power plant. However it is not cleared about the operating range of a SI which is taken account of the discharged flow structure. The objective of the present study is to reveal the influence of two-phase flow behavior on operating limits of a SI. The test section of the SI is made by transparent material to observe flow structure in it. The pressure distributions along the flow direction and the discharge pressure were measured by changing the inlet steam pressure and a load on exit of the SI. At the same time, the discharged flow at the diffuser was observed with a high speed camera. From the observation results, it was confirmed that a boundary which the flow structure changed in the discharged flow. The area from the throat to this boundary is seemed a two-phase region that steam which has not been condensed completely in the mixing nozzle remains. It was found that this boundary moved to upstream as the load on the exit increased and significant pressure rise occurred at the position the boundary reached. Additionally white propagations toward downstream were observed. This propagation is seemed a pressure wave propagation. The velocity of the propagation was estimated by image processing. Assuming a pressure wave propagates at sonic speed, void fraction at the discharged flow was estimated by existing homogenized model. From the above, the influence of two-phase flow in discharged flow on the operating limits of the SI is discussed.

Supersonic steam injector (SI) is a passive jet pump with a venturi structure which operates by condensation of the supersonic steam flow in a subcooled water jet. There are advantages that a SI which operates without electric power has high heat-transfer performance and discharges water at high pressure. Therefore the SI is expected to be applied to the safety system which is able to condense steam efficiently and inject water into a core reactor when severe-accident occurs in a nuclear power plant. According to the previous studies about a SI, there are two conditions about the definition of a SI, one is formation a water jet in the mixing nozzle, the other is stopping drainage from the drain. The 1-dimension analysis model, which the steam condensed completely, at the mixing nozzle, is developed by Narabayashi et al. (1996). Abe et al. (2011) quantified the interfacial behavior of water jet with high resolution observation in the mixing nozzle and evaluated correlation between the interfacial behavior and heat and momentum transfer characteristics. However the operating limit of a SI which is taken account of flow structure at the diffuser is not cleared. The objective of the present study is to reveal the effects of two-phase flow behavior in the diffuser on the operating limits of the SI. In the present study, static pressure distributions along flow direction were measured in the SI with change about the back pressure. The discharge pressure was measured and compared with the value of discharge pressure predicted by 1-dimention analysis model. On the other hand, the discharged flow in the diffuser was observed with a high speed camera. According to the measurement of pressure, a significant pressure rise in the diffuser was measured as the back pressure increased. In addition, the value of the discharged pressure just before the SI becomes inoperative by the back pressure rise was defined as the maximum discharge pressure. It is suggested that this maximum discharge pressure is one of the indicators which predicts the operating limit of the SI. From visualization result of the diffuser, it was confirmed that there is propagation of white color. The void fraction in the discharge flow was estimated at about 2% from the velocity of the propagation. From the experimental results, effect of two-phase flow behavior of discharged water jet on the operating limits of a SI is discussed.

Adequate design solution and maintenance of circuits with fast reactors cooled by lead and lead-bismuth coolants require taking into account the peculiarities of hydrodynamics of these coolant flows. The design pressure of saturated vapors of lead and its alloys at temperatures of 400–550 °C is 10 −18 −10 −10 atm, which is significantly less than that of sodium or water. Processes of traditional cavitation cannot occur in a flow of heavy liquid-metal coolants because of their specific character. The main circulation pumps of reactor circuits are one of their basic elements. In fact, the flow-type parts of these pumps and other vane pumps operating in lead and its alloys cannot be calculated by traditional methods in terms of cavitation characteristics; appropriate calculation formulas are not currently available. To study cavitation processes in a heavy liquid-metal coolant flow, the authors have carried out the experiments aimed at: - determining the conditions of disconnection of liquid lead and lead-bismuth eutectic column; - determining the cavitation characteristics of a centrifugal pump transferring lead at a temperature of 500 °C; - studying the characteristics of ejector (Venturi nozzle) in a liquid metal; - studying the cavitation erosion effect of the lead coolant on impeller vanes of an axial-flow pump in a limited volume in the FT-4-A stand at a lead flow rate of up to 1200 t/h; - studying the cavitation characteristics of an axial-flow pump in the FT-4 stand at shaft speeds of 13.34–25 Hz. These studies are performed with the lead coolant at temperatures of 450°−550°C, oxygen in lead from 10 −4 −10 −5 to 10 0 , flow rates from 20 to 1800 t/h, which corresponds to velocities of the lead coolant flow from 1.0 to 26 m/s. The experiments have shown that as distinct from water, traditional cavitation processes in a heavy liquid-metal coolant (HLMC) flow are not recorded. The probable cavitation mechanism is gas cavitation. The allowance for the specific character of hydrodynamics of HLMC flows is necessary for adequate design engineering and maintenance of some elements of the reactor circuit.

Steam injector (SI) is a passive jet pump which is driven by high-performance steam condensation onto water jet and it is expected to be active at severe accident of nuclear power plant with no electricity. SI is mainly consists of convergent-divergent nozzle. Supersonic steam flow condenses onto water jet in the mixing nozzle and mass, momentum, and energy of steam is transferred to water in the mixing nozzle. Condensed water jet is accelerated at the throat and kinetic energy is converted into pressure in the diffuser, which produces higher pressure than inlet steam pressure. It is easy to apply the SI to nuclear power plant since SI has quite simple and compact structures. The objectives of the present study are to clarify the mechanism of heat and momentum transfer in the mixing nozzle and to determine operating range of SI for practical use. A transparent test section is adopted to conduct visualization of the flow structure with a high-speed video camera as well as measurement of pressure distribution in mixing nozzle, throat, and diffuser with changing back pressure. Fundamental parameters change between operative and inoperative state of the injector were evaluated by measuring pressure and temperature distribution along axial direction of the test section. Discharge pressure as one of operating characteristics of the injector was also measured in changing back pressure by decreasing the opening ratio of the back pressure valve attached downstream of the test section. It was confirmed that discharge pressure increased and the injector became inoperative unsteadily with decreasing opening ratio of the back pressure valve just after it produced the maximum discharge pressure. In the present investigation, this maximum discharge pressure is evaluated as the operation limit of the injector. Furthermore, discharge pressure from diffuser, which is one of the indicators of operating performance as well as operating limit is predicted from inlet condition adopting one-dimensional analysis model proposed previously. By comparing analytical result with experimental data, as well as visualization of flow structure in throat and diffuser, physics model including two-phase flow structure with shock wave which was observed at throat and diffuser are discussed in order to predict injector’s operation with high accuracy.

Supersonic steam injector is a passive jet pump which operates without rotating power source or machinery and it has high heat-transfer performance due to the direct contact condensation between supersonic steam flow and subcooled water jet at the mixing nozzle. Since the supersonic steam injector has a quite simple and compact structure, it has been considered to apply to the safety system for the Next-generation nuclear power plant. There are various researches about the formulation and modeling of operating, flow structure and heat transfer characteristics of both vapor and liquid flow. However, there are few models which are capable of evaluating heat and momentum exchange at the boundary layer between supersonic steam flow and water jet. Since heat and momentum exchange is considered to have a major impact to operating characteristics of a supersonic steam injector, it is necessary to formulate the model which simulates such complex phenomena at the boundary layer with high accuracy. The objective of the present study is to investigate the relation between the thermal characteristics and interfacial behavior between the flows to develop a model which is able to assess the heat and momentum transfer characteristic and the flow structure of the supersonic steam injector in detail. In the present study, a visible test section of water jet-centered supersonic steam injector was adopted to conduct visualization of the water jet with high speed video camera. In addition, special measurement instrumentations of temperature and pressure were applied to obtain radial distribution of temperature and pressure in the mixing nozzle of the injector. There were large velocity and temperature gaps between the water jet and the supersonic steam flow which indicated the existence of large momentum and heat exchange at the boundary layer of the flows. It was clarified that there was pressure gradient which was considered to stabilize the water jet in the mixing nozzle from calculation of radial distribution of total pressure gradient. From the visualization measurement, it was also clarified the existence of a complex wavy behavior on the surface of the water jet. The wave velocity was estimated by the image processing technique and the cross-correlation method. It was found that there was a relation between the wave velocity and heat transfer characteristics in the supersonic steam injector. It is suggested that the enthalpy ratio between the liquid subcool enthalpy and steam condensation enthalpy as well as the Jacob number between both flows could be an indication factor for the effect of the wavy behavior on the condensation.

Collapse of a water jet flowing out from a nozzle to the atmosphere was examined. The diameters of nozzles used in the experiments were 3, 6 and 8 mm. The flow state of the water jet was recorded with a high speed video camera. The collapse length was derived from recorded images. When the flow velocity was quite low, the surface of the water jet was smooth and small perturbations appeared at the lower position of the water jet. As the flow velocity was increased, the position where the small perturbations appeared came close to the nozzle outlet. The perturbations grew as these went downstream and lumps of water were formed at the lower position. When the flow velocity was further increased, successive waves came around on the surface of the water jet. The collapse of the water jet occurred in such a state that the lump of water was torn off from the jet. When the water jet velocity was high, the jet turned into a dispersed flow and the collapse occurred. The agreement of the measured results and the predicted results was poor. It was considered that the instability of the surface of the water jet seemed important for the jet collapse in the present experimental range. The Kelvin-Helmholtz instability wave length was compared with the measured wave length on the water jet. When the wave length reached the Kelvin-Helmholtz instability wave length, the jet collapse occurred except the case that the transition to the dispersed flow caused the jet collapse.

This paper reports new approaches for increasing environmental protection during severe accidents at NPPs. For NPPs with two protective shells and pressure release system such as WWER-1000 we suggest a new comprehensive, passive-mode environmental protection system of decontamination of the radioactive air-steam mixture from the containment and the intercontainment area, which includes the “wet” stage (scrubbers, etc.), the “dry” stage (sorption module), and also an ejector, which in a passive mode is capable of solving the multi-purpose task of decontamination of the air-steam mixture. For Russian WWER-440/V-230 NPPs we suggest three protection levels: 1) a jet-vortex condenser; 2) the spray system; 3) a sorption module. For modern designs of new generation NPPs, which do not provide for pressure release systems, we proposed a new passive filtering system together with the passive heat-removal system, which can be used during severe accidents in case all power supply units become unavailable.

An optical fiber has interesting and useful characteristics, which are able to be applied to scientific measurement. Its phase detection characteristic based on refraction difference between two phases is applied to bubbles/droplets measurement. Recently, demands for measurement of tiny bubbles/droplets increase in research fields of steam injectors, sprays, automotive engines, fine chemical reactors, and so on. Meanwhile, laser science and engineering, particularly femtosecond pulse laser, has made remarkable advances lately. Their unique properties of the interaction between the femtosecond pulses and materials can be utilized for microfabrication. The optical fiber probe methods have been repeatedly improved in order to measure bubbles/droplets efficiently and reliably in gas-liquid two-phase flows. However, one has been taking it for granted that simultaneous measurement of their diameters and velocities needs at least two optical fiber probes. To break through this situation, we newly developed a Single-Tip Optical Fiber Probe (F-STOP) microfabricated by femtosecond pulse laser (fs-pulses) which realizes simultaneous measurement of diameters and velocities of tiny bubbles/droplets. In the F-STOP measurement, the following properties are used to realize the simultaneous measurement of diameters and velocities of bubbles/droplets: the relation between the reflected-light intensity at the wedge-shaped probe tip and the tip-surface area covered with a phase; reflected-light intensity at the groove microfabricated by fs-pulses. The first aim of the present study is to provide data for evaluation of the influences of surface tension and wettability on the bubble measurement in order to develop precise and reliable F-STOP method. The second aim is to describe the process to make F-STOP via fs-pulses. The third aim is demonstration of the newly developed probe in real measurement of bubbles/droplets. On the basis of these, the performance of the new probe is discussed.

One of the most interesting devices for Next-generation nuclear reactor systems aimed at simplified system, improvement of safety and credibility is a supersonic steam injector. Supersonic steam injector is a passive jet pump without rotating machine and high efficient heat exchanger because of the direct contact condensation between the supersonic steam and the subcooled water jet. It is considered that flow behavior in the supersonic steam injector is governed by the complicated turbulent flow with large shear stress induced by velocity difference between the steam and water and direct contact condensation. There are various studies about formulation of operating characteristic of the steam injector and jet structure in it [1][5]. According to these investigations, heat and momentum transfer of the steam injector are supposed to be related to the turbulent eddy diffusion. However, the studies about the turbulent flow under the large shear stress with direct contact condensation are not enough. Especially mechanisms of the momentum and the heat transfer are not clarified in detail. Objective of the present study is to investigate the turbulent behaviors of the water jet and the interface between the water and the supersonic steam that play an important role in the heat and the momentum transfer. Radial velocity distribution was measured by a special pitot tube. Fluctuation of the total pressure which doesn’t depnd on the density of the fluid is measured by the pitot measurement. It was found that the streamwise velocity increased as it approaches downstream of the injector. Fluctuation of the total pressure is large at the mixing region of the water and the steam. Visual measurement of the turbulent water jet was conducted by high speed camera in order to identify the position of unstable interface and its behavior. It was confirmed that wave propagated on the interface. And its velocity is obtained.

It has been proposed to introduce a steam injector into boiling water reactors as a feed water heat exchanger and a safety injection pump. In the present paper, the heat transfer characteristics in the steam injector were examined. The nozzle size of the water jet in the steam injector that was used in the present experiments was 5 mm. The length of the mixing section of the water jet and the steam flow in the injector was 53 mm. Subcooled water was supplied to the water nozzle. Saturated steam at approximately 0.1 MPa was also supplied to the mixing section of the steam injector. Water jet velocities tested in the present experiments were in the range of 9.7 ∼ 21 m/s. The velocities corresponded to the Reynolds number of 5.9×10 4 ∼ 1.5×10 5 . Radial and axial temperature distributions of the water jet in the steam injector were measured. Velocity distributions of the water jet were also measured. From the measured temperature and the velocity distributions, heat exchange rates from the steam flow around the water jet to the water jet were derived. The obtained results indicated that the heat exchange rates were greatly larger than those of usual turbulent flow in a pipe. A flow state of the water jet was also visually examined. The results of the visual observation revealed that the interface between the water jet and the steam flow was very wavy. It was supposed that the wavy motion on the water jet surface created the effective-large-internal circulation flow in the water jet, which resulted in the tremendously effective heat transport into the center portion of the water jet. From the pictures of the water jet surface recorded by a high speed video camera, characteristics of waves on the surface; the wave height, the wave velocity, the wave length and the wave frequency, were obtained. The heat transfer of the water jet in the steam injector was correlated with the wave characteristic properties. The heat transfer of the water jet was also analyzed by using the commercial CFD code of STAR-CD. When the wavy interface was introduced into the STAR-CD code analysis, the radial heat transport was drastically improved. This analytical result also supported that the tremendously more effective radial heat transport than that of the usual turbulent flow was caused by the wavy motion of the water jet surface.

A steam injector (SI) is a simple, compact and passive pump and also acts as a high-performance direct-contact compact heater. We are developing an innovative idea by applying SI system for core injection system in emergency core cooling systems (ECCS) to further improve the safety of nuclear power plants. The passive core injection system (PCIS) driven by high-efficiency SI is a system that, in an accident such as a LOCA (loss of coolant accident), attains discharge pressure higher than the supply steam pressure to inject water into the reactor by operating the SI, by supplying water from a pool in a containment vessel and the steam from a reactor pressure vessel (RPV). The SI, passive equipment, is used to replace large rotating machines such as pumps and motors, eliminating the failure probabilities of such active equipment. When the water and steam supply valves open, the SI-driven PCIS (SI-PCIS) will automatically start to inject water into the core to keep the core covered with water. The SI-PCIS works for the range of steam pressure conditions from atmosphere pressure through high pressures, in which the analytical simulations of SI were carried out based on the plenty amount of experimental data using reduced scale SI. We further simulated and evaluated the core cooling and water injection performance of SI-PCIS in BWR using RETRAN-3D code for the case of small LOCA. A reactor, such as ESBWR, equipped with the passive safety system by gravity-driven cooling system (GDCS) and the depressurization valves (DPVs) should be inevitable to lead to large LOCA even for the case of small LOCA by forcibly opening the DPVs to inject water from the GDCS pool due to that the GDCS water head is up to ∼0.2MPa. On the contrary, our simulation exhibited that SI-PCIS could save the reactors from leading to large LOCA by discharge of the water into a core for the cases of small LOCA or DPV unexpectedly open. In addition, we conducted the analytical simulations of SI, which grew in size for the actual nuclear power plant. A part of this report are fruits of research which is carried out by Tokyo Electric Power Company (TEPCO), Toshiba corporation, and seven universities in Japan, funded from the Ministry of Economy, Trade and Industry (METI) of Japan as the national public research-funded program.