In the ESNII+ EU FP7 project, a reactor physics benchmark aiming at the whole core calculation with the reflectors and detailed description of the structural elements was specified. This benchmark is based on the 2009 CEA concept of the ALLEGRO core. Fixed nominal technological data at nominal reactor state (geometry, composition) were prescribed which had to be modified in specified calculation branches according to different types of the thermal expansion and control rod positions. The parameters of the point kinetic model to be applied in a system thermal hydraulic code had to be determined this way. Static mechanical models of the expansion processes were specified by the benchmark. The goal of the calculation exercise was to verify the reactor physics codes, namely to get information about the modelling uncertainties and — after — their influence on the calculated results of the safety analyses. The obtained deviations between the participants are characterizing the user effects, the modelling uncertainties and the influence of the nuclear data differences all, without the possibility of their separation because of the complexity of the benchmark problem. A conclusion could be drawn that a step by step procedure starting from simple problems (homogenous material, Wigner-Seitz cell or subassembly in asymptotic approach) is necessary if we wish to identify the reasons of the deviations. For the Doppler effect, a decision was made in this direction already in the ESNII+ project where an infinite regular lattice problem without any leakage had to be solved. This approach of the simplicity is followed by the present benchmarks (one rod and one assembly), but extending the simple benchmarks with burnup calculations and taking into account leakage in asymptotic approximation by neglecting the complicated processes necessary in the reflector regions.

The purpose of this study is the calibration in normalization technique for the real-time measurement of void-fraction inside fluid by using VC system at each temperature condition. There is an existing method of measuring electrical response called a VC system. A VC system is a voltage applied current measured system. The VC method measures electrically between electrodes attached around the measurement subject and detects internal distribution from impedance difference of the component to be measured. We deal with water and air in this research. In this method, we measure the difference between the electrical characteristics of air and water. And the measurement technique is requirement of pre-measurement. In homogeneous case, we need to measure current for normalization with real-time measured current. This method has problem that the impedance of water depends on the temperature. So the impedance of water at pre-measurement is different as compared real-time impedance. That’s why, the technique for measuring void-fraction is not high accuracy at present. The measurement technology should be improved to obtain more accurate area of air inside fluid. This requirement is achieved by real-time measurement and real-time separation based on the different electrical properties of air and water while calibrating with temperature. In order to improve accuracy of the proposed method, we conducted a system comparison experiment before calibration with temperature and after calibration. In conclusion, we could measure the void fraction in real time more accurately.

The function of Nuclear Power Plants (NPPs) is to produce electricity. However, in order to do that electrical power is also required for the house load for operation and to keep the safety of the plant intact. The Electrical Power Systems (EPSs) are specially designed both for normal plant operation and for particular conditions other than normal operation. In this way, NPP safety is maintained and continuousness of electrical power supplies is available despite temporary interruptions or flows through regular operation and post-shutdown. The power for a plant comes from different and reliable power sources that are physically and electrically isolated. The overall EPSs design intends to assure that any single failure or malfunction will affect only a single source of power supply and the other alternative sources should not be affected. Despite the separation between electrical power sources, malicious cyberattacks [IAEA NSS 8] might disrupt the power flow and interrupt the normal operation of a NPP [IAEA NSS 17]. Therefore, to intact the nuclear safety and security of a nuclear power plant, it is very important to incorporate cybersecurity considerations into the EPS design and implementation. Particularly, threat scenarios for EPS should be developed and evaluated before a cyberattack occurs. This will help to identify possible vulnerabilities, their impact and consequences that we want to prevent. The classification of attacks and the modelling of attacks on critical infrastructure are already described in [1] as part of our SMARTEST R&D project. Here, in this paper, we are describing examples of cyber threat scenarios for the EPSs and EPS interfaces. These are necessary with regard to specifying the anticipated security impact to a physical process controlled by the Instrumentation and Control (I&C). This serves as an input for the security analysis and the closed loop virtual validation via simulation and Fault Tree Analysis (FTA).

Using supercritical carbon dioxide (SCO 2 ) as the working fluid of a closed Brayton cycle gas turbine is widely recognized nowadays, because of its compact layout and high efficiency for modest turbine inlet temperature. It is an attractive option for geothermal, nuclear and solar energy conversion. Compressor is one of the key components for the supercritical carbon dioxide Brayton cycle. With established or developing small power supercritical carbon dioxide test loop, centrifugal compressor with small mass flow rate is mainly investigated and manufactured in the literature; however, nuclear energy conversion contains more power, and axial compressor is preferred to provide SCO 2 compression with larger mass flow rate which is less studied in the literature. The performance of the axial supercritical carbon dioxide compressor is investigated in the current work. An axial supercritical carbon dioxide compressor with mass flow rate of 1000kg/s is designed. The thermodynamic region of the carbon dioxide is slightly above the vapor-liquid critical point with inlet total temperature 310K and total pressure 9MPa. Numerical simulation is then conducted to assess this axial compressor with look-up table adopted to handle the nonlinear variation property of supercritical carbon dioxide near the critical point. The results show that the performance of the design point of the designed axial compressor matches the primary target. Small corner separation occurs near the hub, and the flow motion of the tip leakage fluid is similar with the well-studied air compressor. Violent property variation near the critical point creates troubles for convergence near the stall condition, and the stall mechanism predictions are more difficult for the axial supercritical carbon dioxide compressor.

The steam-water separator is vitally important equipment to remove the droplets entrained by the vapor stream to provide dry saturated steam for the steam turbine in the nuclear power station. With the development of the large nuclear power station and the vessel nuclear power plant, the steam-water separation performance should be more efficient under the condition of higher pressure, power load and circulating ratio. The droplet motion model, which is solved by typical four steps Runge-Kutta method and validated against the experimental results, is developed according to the physical phenomenon description and the mechanism comprehension of the vapor entrained droplets moving in the wave-type vanes separator. The Euler-Lagrange methodology is adopted to simulate the moving droplet entrained by the vapor stream in the wave-type vanes separator and the separation performance is investigated. The separation efficiency of the separator and motion trajectories of droplets with various sizes are presented. Stokes Number ( St ) of diverse droplets is obtained to analyze the influence of Stokes Number on the moving droplets trajectories and the separation efficiency. The results reveal that the values of Stokes Number for most of the moving droplets in the wave-type vanes separator are beyond 1, which indicates that most of droplets are likely to collide with the solid wall of the separator. Only when the droplet velocity is smaller than 1 m/s or the droplet radius is less than 2 μm, the Stokes Number may be below 1 and the moving droplets can be entrained by the stream flow until escaping from the separator. The analysis can forecast the maximum critical separation size of the droplets that cannot be removed, and the minimum critical separation size of the droplets that can be removed throughly by the separator and guide the design of the separator.

The corrugated plate dryer is a very important equipment for the steam-water separation in the steam generator. Its separation efficiency determines the economic indicators of nuclear power plants. The study of the complicated steam-water separation mechanism of the corrugated plate dryer is very helpful to improve its separation efficiency. Flow and breakdown characteristics under horizontal shear of water film falling down vertical corrugated plate dryer is investigated. Air and water are used as the two phases in the research. This paper also derives a model calculating the critical airflow velocity of water film breakdown and completes the comparison of the experimental result and the result calculated by the model. The relationship between the water film thickness and critical airflow velocity of water film breakdown is investigated through the experiment. In addition, the water film thickness is measured according to CCD high-speed camera acquisition system and the Planar Laser Induced Fluorescence (PLIF) method is also used for the measurement. The experimental result reveals that the critical airflow velocity of water film breakdown is related to the corrugated plate structural parameters, the properties of the two phases and the water film thickness. The critical airflow velocity of water film breakdown and the water film thickness are negatively correlated. The result calculated by the model is in good agreement with the experimental result as the water film thickness is large in a certain range.

In the framework of the ALFRED design (Advanced Lead Fast Reactor European Demonstrator) for DEMO-LFR, a new concept of steam generator (SG) has been proposed consisting in a double wall bayonet tube bundle which improves the plant safety reducing the possibility of water-lead interaction thanks to a double physical separation between them, and allowing an easier control of eventual leakages from the coolant by pressurizing the separation region with inert gas. In order to support the development of this innovative SG configuration, the ENEA has designed and realized the HERO (Heavy liquid mEtal pRessurized water cOoled tubes) test section, a mock-up (1:1 in length) which represents the ALFRED SG. This test section, implemented in the CIRCE pool facility, aims to investigate on the thermal-hydraulic features of the system, providing a database for STH codes validation. The experimental campaign consists of high pressure tests at about 180 bar carried out in the framework of the HORIZON2020 SESAME project (Simulations and Experiments for the Safety Assessment of MEtal cooled reactors). The secondary loop has been realized for the HERO SG feeding, consisting in an open loop circuit fed by demineralized water. The system is equipped with a volumetric pump and a heater in order to reach the water nominal working conditions of 335°C at the SG inlet and about 180 bar at the outlet. A preliminary test analysis is carried out by RELAP5-3D © thermal-hydraulic system code. A numerical 1-D model of the HERO SG and the secondary loop has been realized in order to test the loop layout and to characterize the main components from a thermal-hydraulic point of view, defining the start-up procedures for the achievement of the working conditions of the water for the high pressure tests. Furthermore, several simulations are carried out to investigate on the secondary system behavior both for steady states and transients.

A new Method Of Characteristics (MOC) based computer code, USLAM, has been used to analyze the Fire Protection System (FPS) of a Nuclear Power Plant. Some unique features of this system are that it has a complex geometry, has many operating modes, is connected to an adjacent nuclear power plant FPS, and has a significant number of high elevation standpipes or risers. In a typical FPS, firefighting water is pumped from a low elevation reservoir at atmospheric pressure to the whole system including higher elevations regions where water hammer due to water column separation & rejoining may occur. A low capacity keep-full system is employed to keep the system pressurized during standby conditions. A loss of system pressure due to the opening of the pre-action or deluge valves can lead to void formation at high elevations whose collapse can result in severe water hammer. A catastrophic valve failure due to a water hammer event at a nuclear power plant (Arastu, et al, 1999) disabled the entire FPS. The analysis presented in this paper is to evaluate the possibility of such a water hammer event and mitigate it in an effective manner. The mitigation method employed is the use of multiple vacuum breakers. As many as 11 vacuum breakers are used for this purpose. The paper discusses the analysis results without and with vacuum breakers for all operating modes. The basic methodology of the USLAM code is also discussed.

The melt jet breakup phenomenon in a pre-flooded reactor cavity during a severe accident is related to the debris bed coolability. It is important to predict the jet breakup length for the evaluation of the debris bed coolability. A large volume of works on the jet breakup length have been performed. However, the consistency between experiments and correlations was difficult to achieve. Some data follow the Saito correlation (include Froude number in the correlation) and others follow the Epstein correlation (doesn’t include Froude number). The separation of the jet breakup length correlation along the water subcooling was reported based on the experimental data using the low melting temperature materials in our previous works. Since the previous experiments show an unclear jet shape before entering the water pool which could be an uncertainty factor, a slide gate system for a clear jet shape was additionally installed. Experiments were conducted with the similar condition of previous work and different initial condition of melt jet. With a clear jet shape, the jet breakup length in the subcooled water show different tendency following the Saito correlation. To figure out the effect of the entry condition of the melt jet, the jet diameter and the method of estimating the jet breakup length are revisited. Our previo0us experiments show large uncertainties on the jet diameter, leading to the large discrepancy of the dimensionless jet breakup length. Also, early broken jet core is reported in subcooled water cases. Thus, the uncertain characteristics of the jet breakup length analysis is discussed in this paper including the jet diameter and the method to estimate the jet breakup length.

Operation and decommissioning of nuclear facilities will produce radioactive waste, and different radionuclides in the waste will bring different hazards to the public and the environment. The waste would be sorted more reasonably by distinguishing different radionuclides. Yet it is still very difficult to measure directly the pure Beta radioactive waste in situ, though in situ Gamma-analytical and Alpha-waste-barrel measurement techniques have become more sophisticated. The aim is to propose a scientific technique to sort the radioactive waste in situ. This study focused on the 90Sr-contaminated material in China Institute of Atomic Energy and optimized the design of the existing solid waste disposal facilities. A novel technique to measure the radioactive waste 90Sr-90Y online was proposed, trying to sort the radioactive waste as optimally as possible to realize further separation of exemption waste. Theoretically, the exemption waste can be further sorted, and it can guide the design of radioactive waste disposal system.

In order to solve the problem that the current unloading operation will destroy the sealing performance of Miniature Neutron Source Reactor (MNSR) reactor vessel and the tightness can’t be restored, and to meet the application requirements that the original reactor vessel will be reloaded and operated after MNSR LEU conversion, the new unloading device is designed, which can be used without separation of reactor vessel. There has only one fuel assembly in MNSR. When the fuel assembly are unload for MNSR LEU conversion, the cover plate of the pool is removed, the cadmium string is put in, and the neutron detector is placed at first. After removing the drive mechanism and the control rod, and opening the small cover plate at the top of reactor vessel, the fuel assembly can be grabbed and unloaded by unloading tool only through the opening of the small top cover plate. The MNSR spent fuel has very high radioactivity. The auxiliary mechanical device can be used with unloading tools to realize operation in a long distance by lifting and level motion, which is convenient to shield and can reduce the works’ irradiation dose level effectively. Through calculation and analysis, the results show that the structure strength of unloading device is much larger than the actual load to ensure operation safety and reliability. The unloading device is easy to process and operate, and can be used in the practical operation of MNSR LEU conversion or decommissioning at home and abroad to simplify the operation steps and improve the working efficiency.

The Site Radwaste Treatment Facility (SRTF) was to treat the liquid radioactive waste with high concentration of electrolytes in CAP1400 plant. Unlike the forced-circulation or natural-circulation evaporator, the waste was purified by evaporation process using heat pump principle in SRTF. This process combined the benefits of heat pump technology and the separation function of evaporation. This paper gave a brief description of the complete evaporation process. The main equipments of the process included pre-heater, evaporator, demister, steam compressor, and distillate tank. The whole system and each unit operation in the process were elaborately simulated based on energy balance. The results can be further used for equipments selection and design. Through the whole discussion of the heat pump evaporation process, it showed the heat pump evaporation process has remarkable energy efficiency and great advantages in equipment layout compared to traditional evaporation treatment for liquid radioactive waste.

Delayed hydride cracking (DHC) is one of the major factors that affect the safety of zircaloy tubes. In order to effectively simulate DHC, the coupling process including hydrogen diffusion, hydride precipitation, energy flow and mechanical fields should be correctly simulated. In this study, a cohesive model considering irradiation embrittlement and damage accumulation is used to simulate the cracking behavior of zircaloy, and in the finite element model its traction-separation relation is converted to the boundary condition between the traction and normal displacement component in the surfaces to be cracked. The coupling calculation procedures to simulate the DHC behavior using the automatic finite element program generator (FEPG) are obtained and validated, in which the three-dimensional constitutive relation involves plastic strain hardening, irradiation hardening and hydride-induced anisotropic expansion. We find it effective to study the DHC behavior. And we draw the conclusion that irradiation embrittlement and irradiation hardening could visibly accelerate the cracking process. This study improves the numerical simulation method for DHC, provides a reference basis for theoretical study and lays a foundation for further researches on DHC of zircaloy tubes.

The function of steam separator is to remove the small droplets carried by the vapor stream and to provide qualified saturated vapor for the steam turbine in the nuclear power station. The separating characteristics of the steam-water separation plant are of vital importance to the safe operation, economy as well as reliability of the power station. In order to satisfy the requirement of power increase of large nuclear power station as well as space compaction of the vessel power plant, the steam vapor quality must be improved, which requires that the steam-water separator has better separating function to make sure that it can provide the qualified steam on the condition of higher steam pressure, power load as well as circulating ratio. There are many complex phenomena when the droplet moves in the steam-water separating plant, including the droplet emergence, the droplet moving with steam vapor, the collision between droplets and with solid wall, evaporation. It is a good way to study the steam-water separating characteristics for the microcosmic behavior of the droplet. Thus, in order to know the droplet evaporation characteristics in the steam-water separator, the static droplets phase transformation model under the pressure variation condition is built according to the physical phenomenon description and mechanism comprehension when the droplet moves with the steam vapor in the steam-water separation plants. This model is solved by the typical four steps Runge-Kutta method and validated by comparing with the experimental results. Then, the influence of working pressure as well as pressure difference between the droplet surface and the environment on the static droplet evaporation characteristics is conducted. The simulation results show that the working pressure and pressure difference have great impact on the static droplet evaporation characteristics. With the increase of the working pressure, the droplet evaporation rate becomes slower, that is because the physical property parameters of the water vapor and water become closer to each other and the self-diffusion coefficient of the water vapor as well as the evaporation condensation coefficient become smaller, which results in the droplet evaporation rate becomes slower. When the pressure difference between the droplet surface and the environment rises, the droplet evaporates faster, that is because the vapor velocity around the droplet becomes larger and the droplet evaporates faster. These results of the simulation can lay the foundation for subsequent study of the droplet evaporation characteristics when the droplet moving in the separating plants and for the droplet fast evaporation characteristics when the environment pressure changes fast.

A wireless tomography system has been developed to measure the real-time behavior of solid-liquid two phase flow in the centrifuge for controlling a rotational velocity and particle supply rate. The purpose of this study is development of real-time behavior of solid-liquid two phase flow measurement technique and image conductivity distribution in rotating body because a technique for measuring the behavior of solid-liquid two phase flow in the centrifuges has not existed yet. In this study, a device to measure wirelessly behavior of solid-liquid two phase flow in stationary body has developed as a preliminary to measure in rotating body. A centrifugation technology for industry process should be improved to obtain more effective separation, shorten process time and save energy. These requirements are achieved by optimizing rotational velocity and particle supply rate. The real-time measurement of behavior of solid-liquid two phase flow for controlling the rotational velocity and the particle supply rate is essential. In other words, the real-time behavior of solid-liquid two phase flow measurement and the rotational velocity control become innovative technologies. Typical behavior of solid-liquid two phase measurement technics with easy handling in particles liquid two-phase flow uses process tomography because the process tomography is suitable for real-time measurement in a rotating body. Process tomography has high temporal resolution. This detector was used with the wireless because electrical cables are not available for centrifuges under high speed rotational condition. This wireless tomography system was used for a lab scale rotating machinery measurement experiment. Consequently, we could wirelessly measure the behavior of solid-liquid two phase flow in stationary body in real time and get data to image the behavior of solid-liquid two phase flow. This result images the behavior of solid-liquid two phase flow. Observing it enable to control a rotational velocity and particle supply rate. So this wireless tomography system can separate solid-liquid two phase flow efficiently.

A series of nanofiltration membranes were prepared by interfacial polymerization of piperazine and terephthaloyl chloride on the surface of polyacrylonitrile (PAN) ultrafiltration membranes. ZnO nanoparticles were incorporated in the active separation layer to modify the performances of the membranes. The preparation conditions as the monomer concentration, dosage of nano-ZnO particles and the reaction time on removal of a simulated radioactive nuclide Co (II) were investigated. Fourier transform infrared in attenuated total reflection mode verified the formation of polyamide on the PAN ultrafiltration membrane. The scanning electron microscope images showed that the nano-ZnO particles can homogeneously fixed on the membrane surface. The retention of Co (II) increased with increasing the dosage of nano-ZnO in the range of 0∼0.03 g. Further adding more nano-ZnO, the rejection rate of Co (II) first decreased and then increased. The concentration of piperazine and terephthaloyl chloride showed similar effect on removal of Co (II) ion. 5 minutes polymerization time was sufficient to form an active separation layer on the substrate membrane which changed the separation mechanism from ultrafiltration to nanofiltration. The separation performance of NF3 prepared by the following conditions was optimum: 0.03g nano-ZnO, 0.6 wt% piperazine, 0.5 wt% terephthaloyl chloride, and the reaction time was 15 min. The rejection rates of 1000 mg/L Na 2 SO 4 and Co 2+ in CoCl 2 solution were 90% and 75% respectively. The Co (II) removal rate can be increased to nearly 90% by using ethylenediaminetetraacetic acid disodium salt. Increasing the operation pressure or the feeding concentration of Co (II) can also improve the performances of the membranes in this experiment.

Development of selective adsorbents is very important subject for the effective multi-nuclide decontamination related to the severe accident of Fukushima Daiichi Nuclear power Station (Fukushima NPS). In this study, the adsorption properties for nine kinds of zeolites (Zeolite A, Zeolite X, Zeolite Y, Zeolite L, Modified Chabazite, Phillipsite, Erionite, Synthetic Mordenite, Natural Mordenite and Clinoptilolite) are evaluated in the presence of sodium salts, boric acid and seawater. The present study deals with (1) selective adsorption properties for single nuclide ions (Cs + , Sr 2+ , Eu 3+ , I − , UO 2 2+ , Am 3+ and NpO 2 + ), and (2) multi-nuclide adsorption properties of 26 elements (typical elements in Advanced Liquid Processing System (ALPS) in Fukushima NPP-1) for the above zeolites. The distribution coefficient ( K d , ml/g) and uptake ( R , %) were estimated by batch method using NaI (Tl) scintillation counter, ICP-AES and AAS. Zeolites with different crystal structures have the diversity of the adsorption selectivity for various radioactive nuclides. Chabazite, mordenite and clinoptilolite with lantern or tunnel structure were very effective for the adsorption of monovalent Cs + ions even in real seawater. Zeolite A and X with three-dimensional cage structures were effective for the adsorption of divalent Sr 2+ and Co 2+ ions under the practical condition (30% diluted seawater). Zeolite L was effective for the adsorption of Eu 3+ ions under the practical condition. As for I − adsorption, Ag-zeolites are found to be effective, and the uptake (%) of I − (NaI in pure water) for Ag-zeolites was estimated to be above 98% in pure water. As for actinoid adsorption, the distribution profile, K d vs pH, had a maximum depending on the hydrolysis pH. Zeolite A, Zeolite L and Zeolite X showed an excellent adsorption property for UO 2 2+ , Am 3+ and NpO 2 + , respectively. Selective adsorption tendencies of different zeolites were evaluated for 26 elements referred to ALPS. Comparing the uptake results for different zeolites, the following tendency of adsorbability was observed. Mordenite had adsorption selectivity for monovalent alkali metal ions of Rb + and Cs + . Zeolite A and X exhibited relatively high adsorption selectivity for divalent ions of Sr 2+ and Co 2+ . Zeolite L had adsorption selectivity for trivalent lanthanide ions such as Ce 3+ and Eu 3+ . These tendencies were the same as those without boric acid. Thus, the zeolites with diverse adsorption selectivity are effective for the multi-nuclide decontamination of radioactive contaminated water.

A silica-based ammonium molybdophosphate (AMP/SiO 2 ) was successfully applied to remove Cs(I) from radioactive waste water in our previous study. The structure characterization of irradiated AMP/SiO 2 was carried out by X-ray diffraction (XRD). The effect of the initial concentration of metal ions and acidity concentration for removal Cs(I) by irradiated AMP/SiO 2 were investigated. No obvious difference was observed for AMP/SiO 2 with 300 kGy absorbed dose from the XRD patterns, except the slightly shift of (110) peak toward a small angle. Relatively large distribution coefficient (8×10 3 cm 3 /g) of Cs(I) was obtained in 3M HNO 3 by irradiated AMP/SiO 2 . The maximum adsorption capacity of AMP/SiO 2 irradiated at 300 kGy was estimated as 22.4 mg/g. The loss of molybdenum from AMP/SiO 2 in 3 M HNO 3 with 200 kGy absorbed dose is less than 5%.

A new type of Steam separator for CAP1400 Steam Generator is introduced. The result of characteristic experiment, the variables of which contain steam load, saturated water flow and water level, shows its separation characteristics are suitable for CAP1400 Steam Generator. Steam generator is a major equipment of pressurized water reactor power plant, one function of which is to transfer heat from reactor to feedwater and to generate qualified steam. Steam separator, located in upper shell of Steam generator, is used to separate steam from saturated water/steam mixture which exits from the tube bundle region. Steam separating process usually consists of three stages. Centrifugal separation first happens in Primary separator, where most of water is separated. Gravity separation is achieved in the space formed between Primary separator and the dryer. Finally, as the relatively low moisture content mixture flows through the chevron-type dryer, the qualified steam is obtained. The new developed primary separator, named SP3 type, consist of raise cylinder, swirl impeller, downcomer and vane. As the steam/water mixture enters the primary separator, a centrifugal force is induced by the swirl. Thus the greater dense water in the steam-water mixture is forced toward the outside diameter of the separator, while the less dense steam stays in the central portion of the separator. The separated water leaved the separator through the downcomer. Relatively high quality steam exits the separator, passing upward through the center of raise cylinder. The paper introduced the SP3 primary separator developed by hot test, which includes test device, device design, test method, and result analysis. Separation characteristics of SP3 primary separator are obtained by various tests. The result of varying steam load test shows the outlet moisture is 13.3% under the condition of 100% steam load and 4.3 recirculation ration. The result of varying saturated water flow test shows the outlet moisture increases as the saturated mixture flow rate increases. The outlet moisture increased slowly as saturated water flow rate increases when steam load is at 50% and 100%, while the outlet moisture increases fast when steam load reaches at 120%. The result of varying water level test shows the water level is not a major factor for the SP3 primary separator performance. By analyzing experimental data, a fitting formula of the primary separator performance is obtained. the steam flow rate is the most sensitive parameter affecting primary separator performance in that formula. The water flow is the subordinate parameter for the primary separator performance. the resistance characteristics of separator is also obtained, which supports the research and development of the CAP1400 steam generator..

Corrugated plate dryer is one of the key components of natural circulation steam generator in the pressurized water reactor nuclear power plant. It play a decisive role in the water separation system to making steam become a good quality after steam flow out of a vertical steam generator. With the increase of the reactor power in the future design, it requires that the performance of corrugated plate dryer must be increased in the characteristics of steam load, separating efficiency, volume and resistance to make sure that steam generators still produce qualified steam at higher steam loads and higher cycle magnitudes. In this paper, we have done some experimental studies on the corrugated plate dryer. And we concentrated on airflow parameters on the separation performance and resistance performance of dryer. (1) First of all, we have determined the experimental research program about this study, after this, we finished the design works of the test bench, and then we have completed the construction work based on the design works. (2) Afterwards, the flow and separation mechanism of the wave plate dryer was experimentally studied by means of parameter measurement and analysis combined with observation and recording. The effects of different inlet humidities and inlet velocity on separation efficiency, pressure drop and re-entrainment were studied. The above study shows that: (1) The separation efficiency of corrugated plate dryer is related to inlet humidity and inlet airflow velocity. The separation efficiency of corrugated plate dryer is increase with inlet humidity goes higher, and increased to reach a highest point then goes down when inlet airflow velocity increased. (2) The re-entrainment is related to inlet humidity. When inlet humidity increased, the inlet airflow critical velocity which lead to re-entrain would decrease. (3) The pressure drop of dryer is related to the inlet airflow velocity. When the inlet wind velocity increases, the pressure drop increases.