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In recent years, the operational events that the water intake of heat trap was blocked by marine life in nuclear power plants have occurred repeatedly. Since 2000, there have been more than 200 shutdown events of biological (foreign body) blockage resulting in failure of cooling water system in the United States, the United Kingdom, Sweden, Arabia, India, Israel, Japan, South Korea and China. It has caused major safety risks and brought huge economic losses and has attracted wide attention from various governments, related industries and research institutions. It can be seen that water intake blockage event is a safety problem for all kinds of power plants to face. This paper introduces the typical design of water intake and analyses the potential effects of a large number of marine life into the heat trap water intake from the perspective of circulating water system (CRF) jump pump and essential service water system (SEC) loss of function. Based on the characteristics of the high frequency of such operation events and they may cause major safety risks, the reactor shutdown response time in different conditions is simulated. When a large number of marine organisms flood into the water intake of heat trap, the response of the operator on the potential impact of the unit condition and the operational response plan are analyzed. The impact of the unplanned changes in operating status caused by the blockage of the water intake is explained. Ultimately, based on the analysis as described above, suggestions on the design improvement of the heat trap water intake and seawater filtration system in the nuclear power plant are presented. At the same time, suggestions are put forward on the supervision analysis, dynamic monitoring of the surrounding environment of the water intake, establishing a rapid decision-making mechanism, improving the cold source emergency response plan and increasing the operation control strategy.

When the flow of the main feed water pump of CPR1000 unit is low, that is, when the unit is operating at low load, in order to protect the main feed water pump from damage, two low-flow protection recirculation pipelines are installed at the pressure stage pump outlet, and the two lines are installed with two minimum flow regulating valves. These two valves generally have two designs: proportional type and on-off type. The paper simulates the response of the minimum flow valve of main feed pump to water level regulation of steam generator at low load based on the full-range simulation platform of CPR1000 nuclear power unit. By analyzing the response curve of the water level of the steam generator, the advantages and disadvantages of the different valve designs are verified separately. In the proportional design, the valve position of the minimum flow valve is easy to oscillate with the fluctuation of the flow signal, which will affect the stability of the steam generator feed water control system and then affect the operation of the water level regulation system of the steam generator. At the same time, the proportional minimum flow control valve has the control dead zone. Under the transient conditions with frequent disturbance, the response speed of the minimum flow valve does not match the response of the water level adjustment of the steam generator, resulting in the unstable water level of the steam generator. In the on-off design, the stability of water level regulation will have certain advantages. After comparing the influence of the two valve designs on the water level regulation of the steam generator and analyzing the limitation of the proportional design and the importance of minimum flow valve, two improvements schemes are given for the design of the minimum continuous regulating flow valve.

When operating at high speed, aeroengines mounted on the aircrafts also perform complicated maneuvering flights along with the aircrafts. The engine rotor systems will undergo forced motions excited by base movements and mass unbalance, which may lead to severe vibrations. Squeeze film damper (SFD) is generally used for vibration damping in aeroengine, but it also introduces structural nonlinearity like bistable jump. Moreover, the reliability of SFD is also affected by base excitations. Therefore, it is necessary to investigate the nonlinear dynamic behavior of rotor-SFD-support system excited by base motions. Taking a muti-disk rotor shaft supported by two SFDs with squirrel cages as the research object to simulate a gas generator in a turboshaft engine, the steady-state responses of the rotor system were calculated by multi-dimensional harmonic balance combined with alternating frequency/time domain method (MHB-AFT). The effects of base harmonic rotations and mass unbalance on steady-state responses of rotor system were investigated. The results indicate that the time-varying parametric excitations of base motions have strong effects on amplitude-frequency response of transverse displacement of rotor system. The critical speeds and resonant amplitudes of responses change with the magnitude and frequency of several base harmonic rotations. The variation of the frequency of base harmonic motion has the most significant impact on the amplitude-frequency response of transverse displacement of rotor system. The increase of base harmonic frequency will lead to multiple local peaks in the response curves, especially in the case of base harmonic rolling motion. In addition, the combination of base harmonic rotation around pitching and yawing axes results in significantly different response characteristics. Therefore, the influence of base harmonic motions should be considered during the structural design and damping optimization of SFDs.

Variable cycle turbofan-ramjet (VCTR) engine is a promising propulsion system candidate for future Mach 4+ vehicles. Compared with conventional turbine engines, the VCTR engine has a higher thrust-to-weight ratio, a smaller diameter and a more compact structure. A component-based simulation model is developed to explore the operating characteristics of a high-speed VCTR concept. Modeling methods of variable-geometry components are characterized in this paper, and then the way to build the VCTR model solver is described. The VCTR engine simulation model is validated by the numerical and experimental results of double bypass variable cycle engines from NASA technical reports. Performance and control law of the VCTR engine along the flight trajectory with flight Mach number varying from 0.5 to 4.0 are obtained to maximize net thrust and compared with that of a conventional two-spool afterburning mixed-flow turbofan (MFTF). Results show that the VCTR’s performance is highly dependent on the fan characteristics and limited by various operating constraints, such as fan surge margin, turbine inlet temperature, and compressor discharge temperature. In addition, the thrust level of the MFTF with the variable-area low-pressure turbine is close to the VCTR engine.

When operating in a high-temperature and high-pressure environment, the high-speed rotor systems of aero-engines are inevitably subjected to thermal and aerodynamic loads. However, conventional simulations and experiments of rotor dynamics did not consider temperature distribution and aerodynamic load. Therefore, it’s necessary to investigate the influence mechanism of temperature distribution and aerodynamic loads on rotor dynamics. Using polynomial function fitting for the continuously various parameters in elements, combined with the temperature distribution, the axial distribution of physical parameters in shaft elements was established. The stiffness coefficient matrix considering temperature load was derived based on Timoshenko beam theory. By introducing the tension (compression) potential energy, the additional stiffness coefficient matrix was derived from Lagrange equation. Using state-space vector method, the differential equations of motion for rotor-support system were solved, and dynamic characteristics of rotor system subjected to temperature distribution and aerodynamic force were analyzed. The simulation method was verified with a typical fan rotor under the maximum operating condition. The results indicate that the rotor dynamic behavior is more sensitive to temperature distribution than aerodynamic load when operating in the temperature range of 40∼190°C. And the decrease of first two critical speeds is mainly caused by thermal load. Due to the relatively low temperature, the first-order critical speed of fan rotor considering the additional load is reduced slightly, while the second-order one is decreased by 3.42%. When a mass unbalance of 500 g·mm is located at the first disk, the resonant amplitude of unbalance response rises slightly (0.25%) with additional load. If such amount of unbalance is located at the third disk, the peak amplitude is increased by 1.28%. In addition, such a simulation method could be extended to core engines or even dual- or triple-rotor engines with higher temperature and aerodynamic load, where the dynamic characteristics and unbalance responses are expected to be significantly affected.

In this paper, upstream unsteady passing wake effects on the rotor blade tip with film cooling have been numerically examined. The geometry and flow conditions of the first stage of GE-E3 high-pressure turbine have been used to obtain the unsteady three-dimensional blade tip flow and heat transfer characteristics. The first stage of GE-E3 high-pressure turbine has 46 guide vanes and 76 rotor blades. In the process of calculation, compromise the computational resources and accuracy to simplify the number of the guide vane and rotor blade to 38:76. Namely, each computational domain comprises of one guide vane and two rotor blades. The computational boundary conditions are consistent with the GE-E3 annular cascade test conditions. In the case of 1% blade span tip clearance, through comparing the steady results, time-averaged and time-resolved unsteady results, the investigation of the unsteady passing wake effects on flat tip aerodynamic and aerothermal performance without film cooling holes (NC) and three cases with film cooling holes are conducted. To be specific, near the leading edge (C123), at the central area (C456) and near the trailing edge (C789). This paper emphasizes the variation of leakage flow and heat transfer coefficient at different unsteady instants. The results show that the time-averaged leakage flow is pretty similar to the steady results, but the increment of the leakage flow can rise to more than 8% at the maximum envelope. Moreover, the heat transfer coefficient discrepancy between steady results and time-averaged results is almost below 4%, but the dramatic growth of the instantaneous heat transfer coefficient along the pressure side is in excess of 20% due to the shift of the pressure spot.

Machining feature recognition is a key technology in design for manufacturing, inter-operation of neutral format solid models, and reuse of design models in down-streaming applications. Although many approaches have been proposed, the detection and processing of intersection features are still challenging. In this paper, we studied the local property of B-rep model and usage in intersection feature detection. To support hint based reasoning and processing, we proposed a symmetry property definition with relation to edges of solid models. This symmetry property is based on conventional edge convexity, which is an extension of the feature hint theory. The Transformation of global or partial graph matching to local property matching would reduce the complexity in feature matching stages. We implemented a prototype module to evaluate the feasibility of the proposed approach.

ATHAS is the subchannel analysis code, independently developed by Nuclear Safety and Operation Laboratory (NUSOL) of Xi'an Jiao Tong University (XJTU), the original version of the ATHAS is only suitable for pressurized water reactor. This paper is aimed at the characteristics of sodium cooled fast reactor core mentioned above, developing the analysis program AYHAS-LMR, which is applicable for LMFBR. In the design of sodium cold fast reactor, flow blockage must be considered to meet the design criteria. In view of this, a new model is proposed in this paper and added in ATHAS-LMR, considering the flow blockage which changes the flow area and the wetted perimeter. This model is based on the original grid model, we use the grid to simulate the presence of block, increase the resistance of the blockage area artificially. Then, ATHAS-LMR is able to simulate the flow blockage in LMFBR. In this paper, simulations of FFM5B blockage test were done for the high flow case with ATHAS-LMR, and we compare the simulation results with COMMIX-1. The result indicates that ATHAS-LMR is suitable for the analysis of the flow blockage.

Exposed to the high radiation environment, the PWR fuel rod will be swelling. The deformation is quite different azimuthally which caused by the different subchannels coolant temperature and some manufacture issues. And the asymmetry deformation will have a great effect on fuel rod safety for the increasing pellet center temperature and FCMI due to the change of gas gap. An accurate prediction on fuel strain which is associated with the temperature distribution closely should be done to assure the safety of the fuel rod. The conventional finite difference method has deficiency on treating with the deformation and the finite element method (FEM) could satisfy the requirement. Based on the principle of FEM, the model of thermal conduction, stress-strain analysis is established. The two dimensional model is chosen considering the characteristic of fuel. The analysis code SSTFEM is developed and verified using the experimental data from the opening literature and results of authorized code. After that the code is used to analyze the effect of the asymmetry deformation which is caused by the asymmetry coolant temperature distribution in adjacent subchannels and the manufacture eccentricity between the fuel rod and the cladding. The effect law of deformation and the stress is obtained after a systematic analysis.

Normally, nuclear power plants operate at 100% full power level. However, in some special cases, nuclear power plants need to decrease power and operate at lower power level. For example, because of the typhoon, test and equipment maintenance requirements etc., nuclear power plants decrease power to specific power level and stay for several hours at this power level. Because there is inserted time limit of the power compensation rod (G rod) in the technical specification, inserting control rod to decrease power cannot always be chosen, and the strategy that decreases power by boron injection should be taken. In this paper, powers down strategies are introduced. And the similarities and differences between decreasing power by inserting control rod and decreasing power by injecting boron are also introduced. Before reducing power, reactivity equilibrium calculation should be performed and the predicted amount of boron injection can be calculated. Then according to the power down rate, Boron injection rate can be estimated. Finally, Boron injection operation can be performed. During the power reducing, R rod and some important equipment associated with boron injection should be monitored closely, especially the drain tank water level of condensate extraction system (CEX) and the response of feedwater flow control system (ARE). A key control of power down is to control the axial power deviation (delta I) that is one of the most difficult technical problems. Delta I must be kept in the functional area and it is inhibited that the operating point goes outside limits (left limit and right limit) of region II in any situations. It is performed on CPR1000 full scope simulator (FSS) to decrease power by inserting control rod and boron injection in this paper. Several curves obtained during power reducing are analyzed in detail. According to analysis results, the key procedure, operation risks and operation control during power decreasing are proposed as reference. The purpose of them is to meet the requirements of nuclear safety during power down operation.

In order to improve the technical level of nuclear and radiation safety personnel, strengthen the nuclear safety culture, safeguard nuclear and radiation safety, China established Registered Nuclear Safety Engineer Qualification License System (hereinafter referred to as the registered qualification system). The registered qualification system has been implemented for more than ten years. The paper briefly discusses the background and situation of the construction of the registered qualification system, introduces the achievements and problems, and focuses on the analysis of the key factors restricting the development of the registered qualification system. Finally, some pertinent suggestions are given.

At present, there is still lack of detailed software V&V guidance standards in China, while a number of US nuclear power units and I&C platform are introduced and applied. So the software verification and validation work in our country usually cited the methods in IEEE 1012. With reference to the requirements of IEEE 1012, the V&V process of the software can be mainly divided into three forms: audit evaluation, special analysis and testing. This paper focuses on these parts and gives a detailed description and annotations of the technical methods and their life cycle stages in IEEE 1012, which cover multiple V&V phases. At the same time, the author puts forward his own understanding of the special analysis approach and procedure, such as criticality analysis, interface analysis, traceability analysis, hazard analysis, risk analysis and security analysis, and gives his own experience and related recommendations.

For the optimal design of the selective non-catalytic reduction (SNCR) for a 600MW W-flame boiler, the SNCR process was simulated through method of chemical kinetics analysis and fluid dynamics analysis. The design temperature, de-nitrification efficiency in theory, position of spray gun and other parameters were determined and 46% de-nitrification rate was finally obtained. Chemical kinetics analysis, without considering the effect of reducing agent mixed with NOx, the theoretical efficiency is higher. Fluid dynamics analysis, taking into account the effect of mass transfer, the de-nitrification efficiency is lower than the theoretical value. In practical engineering, the mixed mass transfer is an important factor affecting the efficiency of SNCR. (CSPE)

The dynamic transport of gas-solids in a riser leads to highly non-uniform and complex flow distributions in both axial and radial directions. This study presents a continuous modeling approach that simultaneously computes the axial and radial non-uniform distribution of gas and solid phase transport properties in the risers. The radial non-uniform distributions of transport properties of gas and solids are approximated by the 3rd order polynomials, which have been validated by available experimental data from literatures. The radial heterogeneity is due to a combined effect of riser wall boundary, the radial transport by the collision-induced diffusion, and the turbulent convection of solids. Some important transport properties, such as core-wall boundary and back-mixing ratio, are flow-coupled and solved by the proposed model. The model predictions have been validated against some published experiment data, including the distributions of solid concentration, velocity and pressure gradient along the risers.Copyright © 2014 by ASME

The fluid catalytic cracking (FCC) riser reactor consists of a bottom section of liquid feed injection and vaporization and an upward straight riser of vapor-catalysts transport and reaction. The product yield, obtained at the top of riser, is an accumulative result of liquid feed injection, vaporization by liquid contacting with hot catalysts, and subsequent catalytic cracking of feed vapor while being transported concurrently with catalysts through the riser. The FCC process involves not only these sequential sub-processes but also complicated coupling among multiphase fluid hydrodynamics, heat and mass transfer between phases, and catalytic kinetic reactions of vapor components in each sub-process. It is essential to build up a model covering all sub-processes/mechanisms mentioned above through riser reactor and giving prompt results, especially for real-time online optimization of industrial operation. This paper aims to develop a parametric model, integrated from bottom feed nozzle to top exit of riser, that can quickly predict both hydrodynamic and kinetic characteristics throughout the riser as well as various parametric effects on production yield and selectivity. Highlights of modeling contributions in this integrated model include a mechanistic and spatial-structural model of multiple-nozzle feeding with strong interactions not only among sprays themselves but also with cross-flowing steam and catalysts, a heat transfer model between gaseous and catalyst phases, and a more-rigorously derived model of reactant conservation in the multiphase flow transport. The convective nature dominating the nozzle feeding, riser transport and kinetic reactions allows us to simplify the governing equations in this integrated model to a set of coupled first-order ordinary differential equations whose solutions can be obtained quickly via Runge-Kutta algorithm. Compared to the published plant data, the predicted VGO conversion and gasoline yield from the proposed model shows a much better agreement to those from previous parametric models, which suggests the newly-added sub-models of previously overlooked mechanisms can be quite important. Some parametric effects, such as the effect of catalyst-to-oil ratio and catalyst inlet temperature, on production yield and selectivity are further predicted. The results show that a higher CTO or catalyst temperature normally leads to higher cracking conversion, higher gasoline production and lower coke content. However, a very high inlet temperature of catalysts does cause over-cracking and lower the gasoline selectivity.Copyright © 2014 by ASME

Cooling holes are playing critical roles in the impingement-effusion cooling system. The present experimental investigation have studied the effects of the hole diameters on the cooling efficiency and discharge coefficient. The main flow was heated up via combustion, with a maximum outlet temperature 850K. Cooling air, on the other hand, is supplied by an MAM-200 compressor. Temperature distribution on the hot side of the effusion plate is measured by a thermal infrared imaging camera calibrated by 5 thermocouples embedded in the tested plate. A filter is used to minimize the radiation from the hot gas. Cooling holes, both the effusion holes and impingement holes are designed in staggered arrangement; and the number of effusion holes is almost equal to the impingement holes. The inclination angle of the effusion holes in the streamwise direction is 20 degree with respect to the liner surface. The pressure drop across the cooling system ranges from 3% to 5% of the total pressure of the inlet flow. Temperature distributions are measured with an infrared thermal imaging system. The length of the holes is kept to be constant. Three various effusion hole diameters/length ratio 0.3, 0.35, 0.43, with impingement hole diameters/length ratio 0.4, 0.47 and 0.6 correspondingly, have been tested and compared. The experimental results indicate that with the increasing of cooling hole diameters, discharge coefficient is also increased. Nevertheless, cooling efficiency reached its maximum with the effusion hole diameter/length ratio 0.35.

A low-cost chemical sensor based on shrink polymer was presented in this paper. Graphene nanoplatelets were self-assembled on the inner surface of microchannel formed by shrink polymer. The low-cost sensor system can give stable and precise results by only consuming 1 μL reagent, demonstrating great advantage over planar sensor. With assistance of self-assembly and shrink polymer, the sensor can be realized in a simple low-cost way.Copyright © 2013 by ASME

Large break LOCA (LBLOCA) is one of the limit design basic accidents in nuclear power plant. The large flow water in the advanced accumulator is injected into primary loop in early short time. When the vessel pressure drops and reactor core is re-flooded, the advanced accumulator provides a small injection flow to keep the reactor core in flooded condition. Thus, the startup grace time of the low pressure safety injection pump is extended, and the core still stays in a long-term cooling state. By deducing the original accumulator model in RELAP5 accident analysis code, a new model combining the advanced and the traditional accumulator is obtained and coupled into RELAP5/ MOD 3.3. Simulation results show that there is a large flow in the advanced accumulator at the initial stage. When the accumulator water level is lower than the stand pipe, a vortex appears in the damper, resulting in a large pressure drop and small flow. The phenomenon meets the demand of the advanced accumulator design and the simulation of the advanced accumulator is accomplished successfully. Based on this, the primary coolant loop cold leg double-ended guillotine break LBLOCA in CPR1000 is analyzed with the modified RELAP5 code. When the double ended cold leg guillotine accident with 200s delayed startup of the low pressure safety injection occurs, maximum cladding temperature in the core with traditional accumulator is 1860K which seriously exceeded the safety temperature (1477K)[1] prescribed limits while the maximum cladding temperature with advanced accumulator has the security temperature-1277K. From this point of view, adopting passive advanced accumulator can strive a longer grace time for LPSI. Thus the reliability, security and economy of reactor system were improved.Copyright © 2013 by ASME

According to the calculated axial and radial power distribution of reactor core, a numerical analysis of thermal-hydraulic behavior of coolant in supercritical water-cooled reactor of CGNPC (CGN-SCWR) was performed by using sub-channel analysis software ATHAS. The coolant flow in CGN-SCWR reactor core were in two path mode. In order to ensure the maximum cladding temperature (MCT for short) reached the requirements throughout the entire life cycle, the sub-channel analysis of the MCT in the BOC, MOC and EOC had been carried out. The temperature and density properties of coolant and moderator in the fuel assembly along the axial and radial direction were investigated. The MCT of the fuel assembly was obtained, and the selected typical sub-channels were analyzed in detail. The results show that the MCT, coolant outlet temperature and other parameters in CGN-SCWR reactor core meet the thermal-hydraulic design requirements through the whole life time.Copyright © 2013 by ASME

The spent fuel pool (SFP) is mainly used for cooling spent fuel assemblies (SFAs) discharged from the reactor core. Besides, it can also shield the radiation. The decay heat can be removed through normal operation cooling system, otherwise it can only rely on the natural circulation in the pool when the coolant pump loses power or the heat exchanger fails. Thus the pool water temperature will continue to rise until it begins to boil. During this period, if no active cooling measures are triggered, the water level will gradually decrease as water boiling. Once the water level drops to the top of the fuel assemblies, the fuels begin to be exposed in the environment. In this paper, the CPR1000 spent fuel pool was chosen as the analysis object and the best estimate system thermal hydraulic code RELAP5 was employed to investigate the process in spent fuel pool in case of loss of heat sink. The results of calculations show that when losing two sets of cooling line, the increase in water temperature in the pool from 55 °C up to 100 °C takes approximately 9.1 h, the evaporation of water volume above the SFAs takes approximately 75.4 additional hours; while in case of losing one set of cooling line, the water temperature of the pool surface reaches 76.6 °C, at which the pool water would not going to boil under the atmospheric environment condition. The results of performed analysis are useful for safety analysis and storage of the SFAs, and can be used to provide a reference for spent fuel pool engineering design.

The CHF in PWR fuel assemblies is usually predicted by the local flow correlation approach based on subchannel analysis while the effects of spacer grids, cold walls, non-uniform heat flux, etc are investigated. By using the subchannel code ATHAS to calculate each set of bundle CHF data, the local thermal-hydraulic parameters at DNB occurrence point were obtained. In present study, the minimum DNBR point method was applied to develop a new CHF correlation for PWR fuel assemblies. The so-called “three-step method” and “magnitude analysis method” were used to determine the shape and the expression of each item, respectively and the least square method was applied to determine the coefficients of the correlation. Based on the large database of CHF tests, the CHF correlation named ACC correlation has been developed to calculate the risk of DNB. The analysis and assessment results indicate that the ACC correlation can fit the experimental data well with high prediction accuracy and correct parametric trends. Coupled with subchannel code ATHAS, this correlation can simulate the thermal-hydraulics performances of PWR fuel assemblies exactly.

A barrier that limits the development of the nuclear power plants is the problem of the depressurization for severe accidents. The present work addresses this issue by adding the dedicated severe accident depressurization system that is placed on the top of the pressurizer. This design enables the excessive heat to be directly pumped into the atmosphere to lower the pressure in the containment. The theoretical and numerical results indicated that the addition of the dedicated severe accident depressurization system does not affect the performance of the second generation of PWRs. More importantly, this design not only significantly reduces the LERF value but also facilitates the modification of the present structure.

The radial power distribution is uneven in the Pressure Tube type SCWR bundle because of the higher fuel enrichment and the separation of light water coolant and heavy water moderator. The uneven profile will usually result in high cladding temperature. In order to optimize bundle design, the analysis of coupled neutronics/thermal-hydraulics was carried out. The neutronics code WIMS-AECL and the subchannel code ATHAS are selected to conduct the coupled neutronics/thermal-hydraulics analysis for the PT-SCWR 54-elements bundle design. The analysis of the reference case showed that the maximum cladding surface temperature of the bundles at both Beginning of the Cycle (BOC) and End of the Cycle (EOC) exceed the design limits. The optimization was carried out through (1) adopting staggered radial uranium enrichment profile and (2) optimizing the pitch-circle diameters of element rings. The results showed that the optimized bundle can achieve lower maximum cladding surface temperature, which is well below the limiting criteria, and meet the criteria of neutronics demands.

This paper describes the preliminary safety analysis of a thermal-spectrum SCWR concept (CSR1000), which was proposed by Nuclear Power Institute of China (NPIC). The passive safety system and the design of the two-pass core concept characterize the safety performance of CSR1000. With code SCTRAN (a one-dimensional safety analysis code for SCWRs), loss of coolant flow accidents (LOFA) and loss of coolant accident (LOCA) as well as some other typical transients and accidents were analysed. The maximum cladding surface temperature (MCST) was regarded as an important criterion. The sensitivity analyses of some crucial parameters are helpful for the safety evaluation. Thus some parameters about the safety system and the actuation conditions, such as the delay time of the ADS actuation, the break area in LOCA analysis, were also involved in this paper. The analyses have shown that the proposed passive safety system is capable to mitigate the consequence of the selected abnormalities. The results will be a useful reference for the future development of CSR1000.

A model was established to discuss the effect of operating parameters and nozzle shapes on the performance of injectors. The comparison pointed out that heat load has worst effect on injector geometries, while little effect on area ratio. On the other hand, the area ratio is very sensitive to suction temperature, which is critical to injector performance. The least energy loss is achieved when nozzle converges gently at the same nozzle entrance and exit diameter.Copyright © 2013 by ASME

Understanding the solids wetting by collisions between flowing solids and spray droplets is essential to optimizing the design and operation of related industrial processes. This paper presents an innovative experimental method to identify the statistical distributions of liquid attachment and collision probability of the impingement of a liquid spray onto free-fall particles. To separate the liquid attachment from other contributions to the mass measurements of particles and liquids, individual impingement experiment with hydrophobic or hydrophilic particles of the same otherwise properties is respectively performed under the same conditions. In addition, a series of collection bins are distributed along the spray path to determine the statistical probability of collision properties including coating thickness. Based on these measurements, the collision-induced momentum transfer and velocity redistribution of solids and droplets can be estimated with the aid of a Lagrangian trajectory model. Consequently a general two-parameter of collision probability density function is proposed.Copyright © 2012 by ASME

Safety analysis is one of the chief difficulties during the research and design of SCWRs. Currently, the development of the SCWR safety analysis code is still in its infancy all around the world, and very few computer codes could carry out the trans-critical calculations where significant changes in water properties would take place. In this paper, a safety analysis code SCTRAN for SCWRs, has been developed based on RETRAN-02, the best estimate code used for safety analysis of Light Water Reactors. The ability of SCTRAN to simulate transients where both supercritical and subcritical regimes are encountered has been verified by comparing with the APROS and RELAP5-3D codes, respectively. The results show that the SCTRAN code developed in this study is capable of performing safety analysis for SCWRs, and the results are reliable. SCTRAN developed in this study is of great value and engineering significance, and could be utilized in the Chinese SCWR research and development.Copyright © 2012 by ASME

Safety evaluation of urea reactors containing defects was studied based on acoustic emission examination and subsequent dissection. First, acoustic emission examination was conducted in the process of pressure testing during the hydro test and the existed defects of the urea reactor were detected and comprehensively evaluated. Second, the dissection results of the multilayer urea reactor confirmed that the acoustic emission source locations were heavy weld radial expansion cracks. Third, crack stress intensity factors were calculated in accordance with the finite element analysis results and actual crack sizes which were acquired during urea reactor dissection. Based on the crack stress intensity factors, the remaining life of the multilayer urea reactors can be estimated during the acoustic emission testing process in accordance with the acoustic emission source location appearance pressures.