pieprofvaharneo.cf/liquid-cooling-guidelines-for-datacom-equipment-centers.php Because replacing worn or malfunctioning safety-related components, especially those that were designed and manufactured decades ago by now-defunct manufacturers, is a challenge, qualification programs for new components have been developed. As regulations have changed over the years, so have the requirements for component qualification. This paper explains newly established qualification tasks, the corresponding testing infrastructure, and the state of the art of testing technology. By way of example, the paper describes the program and possible sequence of qualifying NPP safety-related valves.
Depending on the design and operational conditions, such a qualification program may include testing containment environmental effects, sealing capability, end loading, seismic performance, and proper functioning under various conditions, such as high-energy pipe rupture, thermal shock, particle-loaded fluids, to name a few. Qualification of nuclear safety electrical equipment is an important part of ensuring the safety of nuclear power. Combining the status of Chinese nuclear power under EPC mode, the paper puts forward the basic idea of how to construct equipment qualification system.
The eigenfrequencies of anchored cylindrical liquid storage tanks are investigated with different methods in this paper. Three methods are applied to analysis the natural frequencies of these liquid storage tanks: formula method, added-mass method, and fluid-solid coupling method. The results show great consistencies. Formula method provides the most conservative result in the three; added-mass method is the most widely used and easy to setup; and fluid-solid coupling method provides better accuracy.
Mode shapes of the tank-liquid coupling system can be divided in three types: Liquid sloshing, beam vibration, and circle wave. The influence of liquid volume to frequencies is also analyzed. Results from this investigation could be used for evaluation of anti-seismic performance of anchored cylindrical liquid storage tanks.
The aim of this work is to enhance the efficiency and improve the quality control in supervision. And this paper can be used as a beneficial reference for the improvement of supervision efficiency and quality management of NPS. The characteristics of sealing material for equipment on containment pressure boundary may be changed under the high temperature and high pressure and high irradiation conditions caused by accident. This will result in loss of containment tightness. So we need to perform the identification test to verify that the equipment under accident conditions can keep the seal function.
In view of the identification, standard for design basis accidents and serious accidents is different, cannot cover each other, and therefore should be identified respectively. Influence factors for identification and mainly consider the normal condition of thermal aging, normal condition of irradiation aging, mechanical aging, accident condition of thermal environment, the accident condition of irradiation aging, accident conditions of chemical environment,etc.
And assume that seal material in the life of the final suffered a design basis accident or serious accident. When evaluating the test results, the acceptance value should be between the leakage rate limit and intact seal leakage rate, ensure enough design margin, and guarantee the evaluation test pass.
This topic starts from one of the most common non-destructive testing methods: ultrasonic testing. Connecting tightly with the reactor vessel internal manufacturing process, some typical ultrasonic tests have been listed in the essay; the reasons why these tests should be executed at that certain period of time and actual problems confronted during these processes have been issued as well. The problems mentioned include austenitic stainless steel plate, forging piece, and those happened during RVI machining processes, which can help NDT engineers comprehend criterion better and can be used as a feedback for technical preparations and quality surveillance during RVI manufacturing processes.
As one of the main equipments in nuclear island, Reactor Vessel Internal abb: RVI provides space for nuclear fuel assemblies installment and location, steers CRDM Controlling Rod Drive Mechanism, one of the main equipments in nuclear island to control the start-up, power adjustment, and shutdown of the reactor, leads a channel for coolant to the reactor core and dominates the rate of flow to assure the normal running of the first circle in the reactor, shields RPV Reactor Pressure Vessel, another main equipment in nuclear island from radiation, and supports the reactor both during normal and emergency conditions as well.
There are series of screw holes and pinholes in the surface of RVI, ensuring the accurate positioning of baffles and formers which points out the foundation of the functions RVI operates in the core mentioned above. These holes are with different dimensions and diverse types, which making machining them a tough task to accomplish. This essay focuses on the machining and dimension inspection processes of these holes, establishing a risk-prevention system to distinguish the technical risks, and formulating corresponding actions to control the final quality.
Besides, the machining processes, dimensional testing technology and modularization management mentioned in this essay can be used as a beneficial feedback for technicians and quality surveillances involved in the nuclear power field as well. The disbonding mechanism of submerged arc stainless steel strip cladding has been studied.
The hardened layer with higher hardness and lower toughness is lath martensite. Type-II grain boundary and porosity appeared in transition zone and martensite layer, respectively, is the internal cause for disbonding. Numerical simulation results indicate that the residual stress improving with the number of the cladding layers increasing, and the residual stress reach to MPa when cladding six layers. Type-II grain boundary as the crack source expands inducing the disbanding, greater residual stress is the external cause for disbanding. The structural features of generation III nuclear island main components are introduced.
Described the welding processes, welding filler metal classification and welding filler metal special requirement for products in detail, and discuss about the Storage requirement and apply requirement of welding filler metal in briefly. These introductions will be help to lay a solid foundation for the subsequent III generation nuclear equipment. Nuclear fuel is the first barrier for safety operation of the nuclear power plant. High quality of the fuel is the foundation for their good performance in the reactor.
Generally, the process of fuel manufacturing includes several steps, chemical conversion, pellets preparation, fuel elements and fuel manufacturing. The process control of the nuclear fuel manufacturing is implemented in the means of person, machine, material, methods, environment and metering, thus to ensure the quality of fuel and to ensure the nuclear security.
The target of the implementation of the process control is to set the whole processes under control and no issues occurred that can lead to the process failure. The article discusses the Key points to realize the controlled fuel manufacturing process, based on the theory of man-machine system analysis and fuel manufacturing. Meanwhile, it also gives the actions taken according to the key points to prevent human error, equipment malfunction and the process failure.
Different practices of the process control are compared to demonstrate that there are different choices to realize the process control due to different approaches of fuel manufacturing. All the means taken is useful and effective to prevent the malfunction of the processes. The paper also gives the good experiences about the manufacturing process control, such as establish of management procedures to prevent the mis-operation, implement of prevent repair and maintenance to ensure the reliability of equipments etc.
DANESS provides a simulation environment enriching the decision-making process on short- to long-term intra-nuclear options. Short-term ones being essentially nuclear fuel cycle optimisation with longer term options the increasing synergies one may envisage in keeping nuclear energy as competitive as possible while responding to a multitude of constraints and criteria within the economic, environmental and socio-political sustainability space.
Thorium-based nuclear reactors use thorium as fuel through breeding to uranium The analysis of boron level in thorium fuels is essential due to the stringent specifications for boron. A method has been developed for the determination of trace boron in high-purity ThO 2 powders. About 0. The solution was then heated to evaporate surplus HNO 3 and HF, and then an amount of 3 M HNO 3 was added and the solution was subsequently used for the chemical separation.
The number of extractions was optimized to reduce the thorium concentration in the aqueous phase to avoid spectral interference of thorium in the determination of boron by inductively coupled plasma-atomic emission spectroscopy ICP-AES. The results of blank experiments showed that the limit of detection of the developed method was about 0. By spiking 0. These results indicated that the proposed method could meet the requirements for the determination of trace boron in high-purity ThO 2 powders.
Fluoride salt cooled High temperature Reactor FHR attracts particular attention in the world because of its excellence in high power density and low operation pressure. However, it is difficult to simulate the discharge burnup of FHR at equilibrium state because of the changes of fuel element and fuel pebble position.
At present, there are three methods used to estimate the attainable burnup of FHR: full core method, equilibrium bed method and infinite uniform bed method. These methods including refueling or searching equilibrium state consumes more computation time. Based on the infinite uniform bed method, this work introduces a new method to estimate the discharge burnup without needing to search equilibrium state.
It shows that the developed method can be used to estimate the discharge burnup in pebble-bed reactor. The spherical fuel element pebble fuel for high temperature gas-cooled reactor HTGR is the fourth-generation nuclear fuel element with completely independent intellectual property rights. The most notable feature of pebble fuel is its inherent safety, which is guaranteed by the four coating layers serving as its major security barrier and the external matrix graphite.
The pebble fuel is composed of inner core and nonfuel zone, the former is composed of coated uranium dioxide particles and matrix graphite powder and molded by quasi-cold isostatic pressing CIP technique, while the nonfuel zone is matrix graphite shell wrapping on the core pebble by CIP again. After subsequent processes of low temperature carbonization, lathing and high temperature purification, spherical fuel elements complying specific performance requirements can be obtained. Fuel particles and matrix graphite powder will be mixed up uniformly and premolded into core pebble using CIP, then by adding matrix graphite powder wrapping the core pebble to use CIP again for nonfuel zone creation.
Through specific process, matrix graphite powder, the main raw material for spherical fuel elements, has certain composition of artificial graphite, natural graphite, and phenolic resin. In the production process of spherical fuel element, some destructive tests on graphite pebble without fuel particle but prepared by the same process for fuel pebble are required, among which the crushing strength is one of the most important testing items.
For each batch of fuel pebbles, the crushing strength on directions both axial and perpendicular to the molding direction is tested. The results of the trial production of matrix graphite spheres indicate that compressing strength of the core pebble directly affects crushing strength of graphite pebble. Compressing strength of core ball depends both on the characteristics of matrix graphite powder including particle morphology, particle size distribution, content of phenolic resin, and the compressing pressure.
In this paper, major study efforts are concentrated on the effect of matrix graphite powder characteristics on core ball compressing strength in order to discover the relations between powder characteristics and the crushing strength of substrate graphite sphere. The molten salt reactor MSR adopts the fissile materials dissolved in fluoride salt as fuel. In the core, nuclear fission occurs in the fuel salt, while only nuclear decay happens when the fuel salt flows to the outer-loop.
It would make the core burnup to perform differently from that in the reactor adopting solid fuel. In the core burnup calculation, accurately evaluating nuclide density of Xe is important because of the large neutron absorption cross section. In this paper, the flow effect on nuclide density of I and Xe is investigated at various flow speeds, core to outer-loop volume ratios, as well as core power scales.
The result shows that the nuclide density of I decreases significantly when the fuel salt flow is considered due to nuclide decay of I in the outer-loop. At low core power scale, the flow effect on nuclide density decrease of Xe is apparent, which results in hundreds of pcm increase of the core reactivity, because the nuclide decay of Xe in the outer-loop is considerable comparing with the Xe generation in the outer-loop resulting from I nuclide decay.
It is mitigated as the core power scale is increased which increases the nuclide density of I and the nuclide generation of Xe in the outer-loop, which decreases the effect of Xe nuclide decay in the outer-loop on the Xe nuclide density. Extraction of the target isotope ions in the laser photoionized plasmas is one of the key processes in the atomic vapor laser isotope separation. The key plasma parameters, including the electron number density, the temperatures of electrons and heavy species, may have significant influences on the characteristics of the ion extraction process.
In recent years, the ion extraction simulation facility has become an important tool to study the ion extraction mechanisms. In the present paper, a low-pressure, alternating-current AC gas discharge experimental setup is developed. Based on this gas discharge plasma source, the electrical and optical measurements are conducted for studying the features of the argon discharge plasma jet under different operating conditions.
In particularly, a collisional-radiative model is established for argon plasmas and is used for the derivation of the electron temperature and number density based on the measured spectral line-intensity ratios, while the gas temperature is obtained by fitting the recorded optical emission spectra of the plasma jet ranging from The preliminary experimental results show that both the power input and driving frequency of the power supplies have significant influences on the number density and temperature of electrons in the expanded plasma jet region with constant chamber pressures of 6.
The measured key plasma parameters, including the temperature and number density of electrons and the gas temperature, of the argon plasma jet are, on the order of magnitude, close to those of the laser photoionized plasmas in the laser isotope separation process, which provides a good basis for the ion extraction simulation experiments. The determination of nuclear fuel pellets in Gd 2 O 3 and abnormal core blocks by the method of eddy current testing, the factors affecting the detection process were analyzed and processed, through the design of standard bar detection on equipment performance is verified.
From the result of the verification, the eddy flow detection method in the detection speed of 4. Preliminary neutronics analysis of thorium utilization in Fluoride salt-cooled High-temperature Reactor FHR has been carried out using a similar reactor model and fuel composition as the one researched in UC Berkeley. Compared with uranium-based fuels, thorium-based fuel i. The usage of thorium will save uranium resources considerably. In this paper, Zr-doped ThO 2 ceramic microsphere was prepared via external gelation process.
Appropriate calcining and sintering programs were set on the basis of Thermal Gravity Analysis TG curve of microsphere drying in the air. The appearance of the product was investigated by microphotography. The influence of doping concentration and sintering temperature on grain size and microstructure was discussed. Different types of systems and equipment for the nuclear fuel cycle facilities are complex and difficult to maintain, used in the special environment, which inclusive medium is radioactive and corrosive fluids or gases.
Safe and stable operation of the equipment will be directly affected. Higher reliability program is guarantee for safety of nuclear fuel recycle. This thesis analyzes the reality and problems of nuclear fuel recycle. We also provide some solutions for those problems. Although the closed nuclear fuel cycle strategy has been adopted in China, there are still some issues existing in the developing strategy, such as ignoring the utilization of recycled uranium, just depending on the U—Pu fuel cycle, etc.
Because of the high neutron economy, simple fuel bundle design, and on-power refueling, PHWR could use recycled uranium economically and efficiently, and could be the breakthrough of nuclear utilization of thorium. According to the storage mode of the spent fuel assemblies in the spent fuel pool and shielding calculation model of some nuclear power plant spent fuel pool, appropriate improvement and optimization can be made for the relevant shielding calculation model according to the detail conditions.
Some simplified methods can be adopted based on the symmetry principle and equivalence principle. The calculation model and analysis method can provide reference for the shielding calculation and simplified analysis of other nuclear power plant spent fuel pool. Similarly, it can be applied in other cases that involve a large numbers of the same type and neatly arranged radiation sources, such as shielding calculation and analysis for solid radioactive waste storage facilities and repositories. A backlog of approximately 23, tons of used nuclear fuel will be waiting for reprocessing in in China.
France has a long history with fuel recycling with the first plant commissioned in La Hague in and later with its Melox plant that started MOX fuel production in in an industrial scale. Is the supply chain expertise used for the past plants up to date and ready to face this challenge? Parts of the process are the intellectual property of the supply chain like ROBATEL which designed and fabricated essential components of the plants of la Hague and Melox. Like for instance the decanter centrifuge systems or the centrifugal extractors, and to the capacity question, the answer is yes.
The purpose of this paper is to present a description about the first AP fuel fabrication line in Baotou, China. Manufacturing process development has been performed, manufacturing of key process equipments have been studied and localized. Since the completion of the civil work and equipment installation, the project has come into the qualification stage, and the entire line is scheduled to be qualified in by manufacturing of four fuel assemblies. Under the AP Technology Transfer Contract, Westinghouse has been involved in depth in the project and playing an important role during the plant design stage and the process qualification stage.
The first reload fuel fabrication is expected to be scheduled in Taking the importance of protection of occupational exposure in the spent nuclear fuel SNF reprocessing facility into account, based on the international and domestic key technologies of design and construction and commercial operation experiences this paper preliminarily concluded several important aspects on radiation protection design including significant system and pipeline layout, radiation shielding calculation and analysis, radiation monitoring design, and so on.
Some suggestions to the protection of occupational exposure of SNF reprocessing facility are made as well. Reverse fluidic device RFD is a kind of maintenance-free equipment which is used for radioactive liquids nuclear fuel transportation Smith and Lewis in Design of a pulsed-mode fluidic pump using a venture-like reverse flow diverter. The parameters of system are difficult to be measured. In addition, every phase of cycle operation is a nonlinear process Fallows et al. Tucson, AZ, . So the control method of RFD is hard to be ascertained. By the research of traditional method to ascertain time-control parameters, a new method based on level feedback of supply tank is suggested.
The time-control parameters to be ascertained by this method are used for RFD system operation. System can stably operate for a long time. Meanwhile, the method can eliminate the influence of different initial levels on system operation. As a kind of maintenance-free equipment, reverse fluidic device RFD for short can be used for radioactive liquid transportation at spent nuclear fuel plant Rajeev et al.
Indira Gandhi Centre for Atomic Research, . The control method of RFD is one of the most important key points at engineering application for purpose of system long-time steady operation by some measuring and control methods. Now instruments on the market cannot be used for liquid level real-time measurement of RFD and cannot be maintained or replaced. So it is very difficult to ascertain a safe, reliable, and steady control method of RFD.
In this paper, by the analysis of RFD transportation characteristics and the research of traditional method to ascertain time-control parameters, a new method based on level feedback of supply tank is suggested, and time parameters ascertained by this new method are used in RFD system control to verify its effectiveness. The welding process of cladding tube and end plug is a key process of PWR fuel assembly manufacture.
The welding seam of cladding tube and end plug is the weakest part of the whole fuel rod. The life of assembly and safety of reactor is directly influenced by the quality of welding seam, so the fuel plant has very high requirements on the quality of welding seam. And choosing proper weld method is important to the manufacture. And based on large numbers of experiments, the influences of different welding parameter on the metallography, burst test, and corrosion test are analyzed and studied in details. This analysis and study provides the theoretical base for controlling of fuel rod end plug welding process to manufacture the qualified fuel rods.
The zirconium cladding tube is the main component of PWR nuclear fuel assembly, and as the first protective barrier to prevent radiation production escaping, it is significant to environmental protection. After cladding tube production, a non-destructive testing is required, ultrasonic inspection is a main method. Through the analysis of the types of defects in the process of manufacturing, design, and production, artificial defects different in length, width, depth, and angles, both vertical and horizontal orientation, simulate analysis of the ultrasonic response to different component of defect.
Aiming at the typical defects found during the inspection, develop inspect items, such as longitudinal, transverse defect, with longitudinal and transverse position defect, density defect, surface roughness, etc. Adopt metallographic anatomical analysis for some typical defects, and measure the actual shape and size to verify the inspection results. A deconvolution algorithm with response function based on energy resolution calibration was presented. A Eu gamma spectrum was detected by a LaBr 3 Ce scintillator.
The energy resolution calibration was best fitted by a square root of a quadratic function, based on which the detector response matrix was constructed connected with the energy, and the boosted Gold deconvolution algorithm was applied. The deconvolution results are better than the fixed response function with a constant standard deviation. In addition to electricity generation, CANDU6 reactors are used to produce Co, which is an important radioactive nuclide used in many industrial areas.
Unlike the conventional Co production scheme using the adjuster rods in CANDU6 reactors, the central fuel element is replaced by Co together with graphite or SiC in this study and Co is obtained after the fuel bundle is discharged. To investigate such a new and innovative Co production scheme in CANDU6 reactors, various lattice depletion studies are performed to determine the possible amount of Co loading and to derive an optimal material composition in the center pin of a fuel bundle.
Through the lattice depletion analysis, the achievable fuel burnup is evaluated and the possible Co production capacity has also been calculated with several neutronic assumptions. For the lattice analysis, the fuel discharge burnup is calculated based on the nonlinear reactivity theory. Lattice power distribution is also evaluated for different cases. All the neutronic analyses are performed with the continuous-energy Monte Carlo code Serpent in this work.
The application of Lead-Rubber Bearing base-isolation technology was studied in this paper, including base-isolation method scheme, selection of seismic isolation devices, determination of isolation structure models, etc. Taking one SMR nuclear safety-related building for example, the influence on isolation effectiveness of buildings from various spaces and various hysteretic models for the isolation layer were investigated based on the dynamic analysis of isolation structures.
Tokamak, a type of fusion experimental device, is considered as the most promising device on which net fusion power could be outputted for electricity production, i. Divertor, as one of the core components in Tokamak, has to sustain very high heat flux from high temperature plasma, up to tens of MWs, so that the design of divertor plasma-facing unit PFU is quite important. However, due to high activation of Cu element by neutron irradiation, CuCrZr is not appropriate for the material of heat sink anymore for future fusion reactor.
The issue was discussed and the solution was preliminary proposed. This can provide the necessary theoretical basis for improvement of heat flux handling capacity to divertor PFU in future application. For comparison, a non-ODS steel 9Cr was also examined. The oxide films were rather porous in 9Cr steel, while the porosity in 9Cr-ODS almost could not be observed. A three-layered oxide films formed on 9Cr-ODS steel, which was composed of an outer oxide layer, an inner oxide layer, and an internal oxide zone which was featured by gradual variation of alloying elements and oxygen from the oxide to the matrix, the Cr 2 O 3 in internal oxide zone play a key role of oxidation behavior.
The kinetic data of 9Cr and 9Cr-ODS steels showed that oxide films formation obeyed a parabolic rate law. Semiconductors used for neutron detector are always coated by a thin-film, which is used to convert free neutrons into charged particles. The detection efficiency of these detectors coated with single thin-film varied with the energy of neutrons. So the energy response performance of personnel dosimeter with these detectors is not satisfactory.
The purpose of the article is designing a suitable neutron detector for personnel dosimeter. In order to overcome the disadvantage of the energy response performance the neutron detectors are coated by a single thin-film. In these designs a neutron personnel dosimeter detector fabricated with three silicon diodes are coated by two different thin-films. One silicon diode is uncoated and will be used to subtract the response to gamma rays of the coated silicon diodes.
The total personnel doserate results will be obtained by summing-up each response of the coated silicon diodes multiplied with different gain factors. Liquid-fueled molten salt reactor is the only reactor system using liquid salt as fuel in the six generation IV advanced reactor systems, which can be characterized by remarkable advantages and competitiveness in nuclear safety, economical efficiency, natural resource protection, sustainable development, and proliferation resistance of nuclear energy.
Each country in the world is carrying out research on the molten salt reactor, and system analysis is an important research content. RELAP5 code is the best estimation code of pressurized water reactor. In this paper the neutronics and thermal hydraulics models in the RELAP5 are extended, so as to develop a system analysis code for liquid-fueled molten salt reactor. Then the code is verified using molten salt reactor experiment MSRE experimental benchmarks. The results indicate that the computational results coincide well with MSRE experimental benchmarks, and preliminary verifies the validity of the extended system analysis code.
Cooldown and depressurization by the mode of automatic depressurization system ADS and passive core cooling system are calculated post-accident. The variation of main parameters post-accident and the accident advancement and results have been analyzed.
Operation intervention is given and the effects with it are discussed. And the emergency strategy for development and verification of emergency operating procedures EOP is given. The objective of our research is to improve fundamental knowledge of the way to develop heat transfer correlation HTC for supercritical carbon dioxide SCCO 2. According to literature review, most of the existing SCCO 2 heat transfer correlations were extended from subcritical by three modification methods, i.
Based on different modification method, this paper also presents an analysis of 17 heat transfer correlations developed for supercritical fluids SCFs flowing in vertical bare tubes. The analysis on the structure of the correlations based on heat transfer evaluation was also carried out. Deposition of corrosion products increases radiation in primary coolant loops of reactor.
The effects of temperature and pH on the deposition of corrosion products in fusion reactor were experimentally investigated. Magnetite particles with State of loop was made stable by heating and pressurizing. Thickness of deposition in primary and test pipelines is determined by oxide skin thickness gauge model OMD , with accuracy of 0. Higher value of pH is benefit for corrosion products transport from fusion reactor core to steam generator.
In addition, higher curvature causes larger flow resistance, which makes corrosion products easier to deposit. As a prospective approach for inertial confinement fusion, fast ion ignition may achieve high gain with low driver energy and simple target fabrication. However, studies show that fast ion ignition requires an ion beam with extremely large energy fluence and high quality. In comparison with conventional accelerators, laser-driven ion accelerators have advantages of compact size, high density, and short bunch duration. Nevertheless, it is still challenging to simultaneously enhance the yield and quality of laser-driven ion beams for fast ion ignition.
In this work, we propose a scheme to address this challenge. First, the ions of a target can be uniformly accelerated in hole-boring radiation pressure acceleration by matching the laser intensity profile with the target density profile. Second, the oscillation of the electric field for ion acceleration can be effectively suppressed by using two-ion-species targets. Particle-in-cell PIC simulation demonstrates that almost all ions in a solid target of a few microns can be uniformly accelerated to relativistic speeds by using our scheme.
The resulting ion beam with large energy fluence and high quality may drive fusion ignition and more generally create matter with unprecedented high-energy density. Using MCNP code, lead cooled fast reactor physics calculation model has been set up. Detailed calculation of several kinds of critical physics calculations has been established, including fuel elements, fuel assemblies, control rods, reflector and reactor pressure vessel. Then the three-dimensional models were developed for the whole core.
The calculation results include the following core physics parameters: primary effective multiplication factor, power distribution, control system worth, reactivity coefficient.
The results indicate that its physics characteristics of the LFR can satisfy the core design requirement; the power distributions are reasonable; the control system can meet the request for shutdown. The calculation results can provide necessary parameters for thermal hydraulic and transient analysis for lead cooled fast reactor. A neutron activation study has been performed through exposing potential cladding materials for high-temperature reactors to neutrons at the low power research reactor of Saskatchewan Research Council in Saskatoon of Canada.
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