Venturi scrubber is installed in nuclear power plant as a component of filtered venting system and used to remove small aerosols with fission products. There is no method for desk analysis to estimate its decontamination performance in the assumed operating pressure range. In this study, we propose a method to estimate the decontamination performance based on mechanistic thermal hydrodynamic simulation code and report simulated results of the thermal hydrodynamics and decontamination performance in the pressure range in actual environments. With decrease in inlet pressure of the Venturi scrubber as containment vessel side from venting start 650 kPaA, gas flow velocity at the throat of the Venturi scrubber was suppressed under 250 m/s by effect of compressibility of the gas, liquid flow velocity in a hole of it by self-priming changed with self-priming phenomena and the decontamination factor decreased except on some of the gas velocity condition. The tendency of the decontamination factor have high value conformed with a report of Luangdilok et al.(2009). The capability of the method to predict and evaluate the decontamination performance in actual environments was confirmed qualitatively because of the conformation.
After the Fukushima Dai-ichi nuclear accident, Filtered Containment Venting Systems (FCVSs) were developed and installed at nuclear power plants in Japan. Using silver zeolite, a design of an additional filter had been developed for FCVSs to capture radioactive organic iodine, which cannot be captured by conventional aerosol filters. For expanding the design flexibility of the FCVS with the organic iodine filter, additional laboratory tests were conducted in this study. To secure the absorption performance, silver zeolite must be kept dry during the operation of the FCVS. The additional tests were conducted to evaluate the absorption performance of silver zeolite under the conditions of relatively low degrees of superheat. And the test results demonstrated that the decontamination factor of the silver zeolite was over 50 even under the condition that the difference between gas temperature and dew point was 2.1 K and contact time was 0.129 s. In addition, absorption performance tests simulating FCVS start-up conditions were conducted. Although similar tests had been conducted in previous studies, the tests in this study demonstrated the performance under relatively short contact time conditions.
A filtered containment venting system (FCVS) reduces containment pressure and amount of the radioactive release to the environment during a severe accident of nuclear power plant. CRIEPI has developed a FCVS analysis tool which evaluate removal performance of aerosol, iodine and organic iodine in each process of FCVS as decontamination factor as a function of the FCVS system parameters. The paper addresses the integrated analysis tool of FCVS simulator based on the experimental database.
By using the FCVS technology, we had started to develop a high decontamination air cleaning system to remove multi-nuclides for radiation protection to conduct decommissioning the Fukushima NPP. High efficiency multi-nuclide aerosol filters for radiation protection during a process of cutting core debris has been developing at Hokkaido University. A plasma cutter, laser cutter, wire cutter, drilling machine, etc., will be used and will generate aerosols. Therefore, the air cleaning system should be needed for removing core debris. In order to develop an air clean up system, a metal fiver filter test was conducted. Measured DF were analyzed using FE-SEM and particle diameter analyzer was used to breakdown DF for each diameter range, It is possible to develop the high efficiency filters by mulch layer filters. Final filter system will be consisted, such as a wet-type aerosol filter, multi-stage metal fiber filters and a silver zeolite to remove organic iodine..
In this study, accelerated thermal ageing tests of electrical cables for the nuclear power plants were carried out to evaluate effect of degradation of electrical cables on flammability. To evaluate flammability of electrical cables, the GC-MS (Gas Chromatography-Mass Spectrometry) method was applied. The Eco-material, polyethylene based electrical cables and the halogen free fire resistant cables were applied to this study. Temperature and period of accelerated thermal ageing tests of polyethylene based electrical cables were 90°C, 80°C, 70°C and 10, 20, 40 days, respectively. Temperature and period of accelerated thermal ageing tests of halogen free fire resistant cables were 100°C, 90°C, 80°C and 10, 20, 40 days, respectively. Evaluation results of GC-MS method of degraded sheath material of electrical cables were performed.
In this study, a cable fire in nuclear power plant room was analyzed by FDS (Fire Dynamics Simulator). To evaluate combustion character of electrical cable was important in terms of safety assessment of nuclear power plant. However, experiment at evaluation was difficult due to scale and costs. FDS analysis software apply two fire model (one model was based on heat released rate: HRR model, the other was based on chemical reaction: CR model). HRR model and CR model were prepared by combustion tests by cone-calorie heater machine. FDS analysis of cable fire at switchgear room was conducted. The numerical result of FDS analysis showed reasonable fire spreading at CR model. It is because CR model could take oxygen concentration into consideration.
The Atomic Energy Society of Japan(AESJ) established an Investigation Committee on Development of Activity and Risk Evaluation Method for Faults by Engineering Approach. The Investigation Committee utilizes the most advanced scientific and rational judgement, and continuous discussions and efforts in the global field, in order to collect and organize these knowledges and reflect the global standards and nuclear regulations, such as risk evaluation method for the faults displacement and prevention of severe accidents, based on the accumulated database in the world. The Final Report of the Committee has published in March, 2017. In this paper, we describes the outline on the Report of the Committee.
The structural integrity of equipment and piping systems to a fault displacement depends on the damage state of a reactor building. When there is no damage in a reactor building, the supporting function of equipment and piping systems is maintained. In this case, FEM analysis can estimate the allowable state of the entire system of the important safety equipment. In this part, an example of the analysis evaluation for the equipment and piping systems, which are important for safety, in the case of loading a fault displacement and the influence on reactor cooling system are discussed. An evaluation by superposition to a fault displacement and an earthquake is not discussed in this report.
Studies have been carried out about fault displacements under the facilities as external hazard, and risk evaluation method for influence on facilities by fault displacements using engineering approach, “accident management” for domestic nuclear facilities. In this paper it shows the scenario of accident results in fault displacements directly under the nuclear power station causes loss of safety of the plant, and shows risk evaluation method measures for the image of quantitative effect of risk reduction by engineering approach, “accident management” for the scenario, and shows examples of application of that.
Nuclear Power Plant has very huge amount of radioactive materials. Therefore, the risk on the human health , environmental impact and sociological problems are considerably huge. Fukushima-Daiichi NPP Accident revealed the environmental and sociological problems were much bigger than that of human health impact. The release of radioactive materials should be small enough even in the accidental conditions. Comprehensive Risk should be controlled to be small enough. The Risk includes not only the heal impact for workers and resident people, but also the environmental release amount and sociological impacts. However, the release were controlled to be small enough, the environmental and sociological impacts would be also very small. Therefore, the primal goal for the risk control should be radioactive material release to be small enough. The risk on the core damage was taken as one of the indicator of the comprehensive risk. The operators at the NPP have to be always considered the risk of core damage. All activities should be quantitatively and qualitatively evaluated to the risk. Online maintenance is the good tool to consider the risk with activities. The condition will be changed due to the maintenance. It may have a small relation to the core damage. Therefore, the operators also should carefully evaluate the activities. Just following the manual has no meanings. Keeping the manual, the operator should consider the comprehensive risks. These activities will improve the safety of NPP. Japanese NPP should also considered to work with reducing the risks.
Three nuclear reactors at Fukushima Daini Nuclear Power Station lost all their Ultimate Heat Sinks due to damage from the tsunami caused by the Great East Japan Earthquake on March 11, 2011. Water was injected into the reactors by alternate measures, damaged cooling systems were restored with promptly supplied substitute materials, and all the reactors were brought into a cold shutdown state within four days. Lessons learned from this experience were identified to improve emergency management, especially in the areas of strategic response planning and logistics. It was found that continuous planning activities reflecting information from plant parameters and response action results were important. Logistics were handled successfully but many difficulties were experienced. As a platform for improvement, the concept of Incident Command System was applied for the first time to a nuclear emergency management system, with specific improvement ideas such as a phased approach in response planning and common operation pictures.
The SAFER facility is to be established as a third step to support FLEX as a measure to obtain sufficient resources from outside the site in order to maintain the mitigation function indefinitely. That is, the first stage corresponds to permanent equipment for 8 hours from the event, the second phase corresponds to the support by portable equipment in the site (8 to 24 hours), the third stage corresponds to support of SAFER (after 24 hours ). For this reason, it is a requirement to carry the necessary equipment within 24 hours upon request from the power plant. We visited the facility this time and confirmed the equipment specifications and operation situation. In addition, NRC Region III exchanged views on US FLEX strategy and ROP
One of the primary lessons learned from the accident at Fukushima Dai-ichi was the significance of the challenge presented by a loss of safety related systems following the occurrence of a beyond-design-basis external event. In the case of Fukushima Dai-ichi, the extended loss of alternating current power (ELAP) condition caused by the tsunami led to loss of core cooling and a significant challenge to containment. NRC and NEI made an approach for adding diverse and flexible mitigation strategies—or FLEX— that will increase defense-in-depth for beyond design basisbeyond-design-basis scenarios to address an ELAP and loss of normal access to the ultimate heat sink (LUHS) occurring simultaneously at all units on a site. We visit to Palo Verde Nuclear Power Plant and Diablo Canyon Nuclear Power Plant so that we learn FLEX strategy in US.
Hamaoka Nuclear Power Station has always embraced the most up-to-date knowledge in its effort to enhance safety. Following the accident at TEPCO's Fukushima Daiichi Nuclear Power Station, we have been implementing countermeasures for tsunami and severe accidents, and introducing additional measures in light of the New Regulatory Requirements so as to build up our safety. While reinforcing facility measures, Hamaoka NPS is also making all-out efforts to ‘strengthen the frontline response capabilities' as it is humans who can handle those facilities to make the measures function effectively. Strengthening cooperation with local communities to enable a coordinated response in case of nuclear emergency is also essential. “No More Fukushima Daiichi Accidents”, with this firm resolve, Hamaoka NPS is making all-out efforts for safety.
Wastage phenomena on adjacent tubes (target-wastage) arise from water/steam leak in steam generators of sodiumcooled fast reactors. Target-wastage is likely to be caused by liquid droplet impingement erosion (LDI) and Na-Fe composite oxidation type corrosion with flow (COCF) in an environment marked by high temperature and high-alkali (reaction jet) due to sodium-water reaction. The authors derived new wastage correlations from COCF and LDI data based on influencing factors which were formed on the periphery of an adjacent tube. In the previous report, the applicability of new wastage correlations was confirmed for the penetrated wastage tube in sodium-water reaction test with straight vertical tube bundle under practical steam generator operation condition. In this report, the authors established that the new wastage correlations were applicable to each tube of tube bundle in the above test, and the time progress of wastage was qualitatively investigated for the two penetrated tubes in the period including the blowdown.
A modification of multi-dimensional sodium fire analysis code AQUA-SF has been carried out so as to investigate a sodium fire under a severe accident condition inside a reactor containment of sodium cooled fast reactor. The GGDH (Generalized Gradient Diffusion Hypothesis) turbulence model and a radiation model for sodium droplets was newly implemented to the AQUA-SF code. A validation study has been conducted through a benchmark analysis of an upward spray combustion experiment. This paper describes detailed influencing factors in the validation. The temperature distributions are almost the same between the results of the previous and GGDH turbulence models since there is no significant difference in turbulence energy. However, the turbulence analyses shows more reasonable temperature distribution than a laminar analysis dose. The radiation model for sodium droplet increases heat transfer to structure and hence gas temperature and pressure decrease. Consequently, it has been confirmed that the experimental results is simulated reasonably by the newly introduced models in the AQUA-SF code.
For safety assessment of a steam generator in sodium-cooled fast reactors, a computational fluid dynamics code SERAPHIM, in which a compressible multicomponent multiphase flow with sodium-water chemical reaction is computed, has been developed. The original SERAPHIM code is based on the finite difference method. In this study, unstructured mesh-based numerical method was developed and introduced into the SERAPHIM code to advance a numerical accuracy for a complex-shaped domain including multiple heat transfer tubes. The multiphase flow under the tube failure accident is calculated by the multi-fluid model considering compressibility. Validity of the unstructured mesh-based SERAPHIM code was investigated through the analysis of an under-expanded jet experiment, which is a key phenomenon in the tube failure accident. The calculated pressure profile showed good agreement with the experimental data. Numerical analysis of water vapor discharging into liquid sodium was also performed. The calculated temperature field agreed with the existing experimental knowledge. It was demonstrated that the proposed numerical method could be applicable to evaluation of the sodium-water reaction phenomenon.
Accurate evaluation of three-dimensional temperature distribution is essentially required for the structural integrity analysis of large-sized straight tube steam generator (SG) of an advanced sodium-cooled fast reactor (SFR). Numerical simulation system TSG to analyze three-dimensional thermal-hydraulics in the straight tube SG has been developed. The three-dimensional simulation on sodium side by the CFD code FLUENT with porous body model was coupled with the multi-channel simulation on water/steam side. Fundamental validation of TSG code with the 1MWt straight tube SG test was performed. Temperature distribution of large-sized SG under the condition with plugged tubes was analyzed and the temperature deviation of heat transfer tubes was evaluated. Through the numerical simulation, local temperature increase near the plugged tubes was quantitatively evaluated, and the applicability of the simulation results to the structural integrity analysis of straight tube SG was indicated.
Thermal fatigue caused by thermal mixing phenomena called thermal striping is one of the most important issues in design of sodium-cooled fast reactors (SFRs). In JAEA, a numerical simulation code named MUGTHES which can deal with conjugate heat transfer between the fluid and the structure parts has been developed in order to predict the thermal response in the structure for estimation of the thermal fatigue issue. In fundamental validation of the MUGTHES, the benchmark analysis was considered using the experiment of planar triple parallel jet sodium test (PLAJEST). Since it was known by literatures that five representative flow patterns according to the velocity rate (Vr=Vs/Vc) of the side jets (Vs) to the center jet (Vc) were shown in the triple jet, three specific experimental conditions at Vr=1, 1.56, and 5.56 in PLAJEST were employed as boundary conditions for the benchmark analyses. Through the benchmarks, applicability of the large eddy simulation (LES) approach with the standard Smagorinsky model in MUGTHES to simulate thermal striping phenomena was potentially confirmed.
This small reactor has one fuel assembly of 8 concentric cylinders of metal fuel in the core diameter 0.62m .Regularly (every 6 months) the bottom part of the fuel assembly is cut off, and the fuel assembly is reset, and so the breed-burn wave is concentrated to upper direction. Regularly (every 5 years) the supply of blanket fuel including 1~3% fissile fuel is added to the top of the fuel assembly with welding. That maintenance-work of the fuel assembly takes 6 days with robot technology, and so the fuel assembly works almost always at full load. The naked cylinders are thick of 15mm so that the breeding ratio and the strength of breed-burn wave are improved to high-level. Those features make this reactor to work as TWR within max. burnup 14%. The thick cylinders grooved deeply are cooled by the boiling sodium. (boiling point 700°C 16kPa) The recycled U and TRU from used fuels of LWR are useful for this reactor.
In the in-service inspection of sodium cooled fast reactors, an examination by continuous leak monitoring is considered for components constituting a sodium boundary. The continuous leak monitoring examinations are premises in a LBB (leak before break) being established. In previous LBB evaluation, the axial crack of the elbow flank has been evaluated, because in the piping system of the fast reactor with high thermal expansion, high stress is generated in the elbows flank. However, the LBB evaluations for circumferential cracks are important in the point of continuous leak monitoring in the in-service inspection. In this study, the load conditions in the LBB assessments for circumferential cracks were examined and LBB characteristics of 1 class 1 pipes of the prototype fast breeder reactor "Monju" were evaluated as an example.
In the Fukushima Daiichi Nuclear Power Plant unit 1, isolation condenser (IC) should play an important role in core cooling during station blackout. To evaluate cooling capability of the IC, two phase flow analyses with the TRAC code and experiments for high pressure conditions were conducted. The analyses for the actual BWR indicated that decay heat could be removed if the IC worked properly at 18:18. However, our results showed that core uncovery occurred at 17:30, which will lead to hydrogen production from the heated core. Further efforts should be made to properly consider the effect of hydrogen accumulation for degradation of core cooling. The experiments simulating IC operation indicated that the natural convection occurred and decay heat could be removed by the IC at high pressure conditions.
The purpose of this study is improving the air-cooled heat exchanger of spent fuel pool cooling system. It is necessary to improve the performance of this heat exchanger to compact this system. We decided the most suitable heat exchange pipe shape by experiment. And we could increase natural circulation flow rate of the air by improving the chimney shape of the heat exchanger. Then, we succeeded in the reduction of the heat exchanger. Finally, we confirmed that the improved cooling-system without power supply could radiate heat by heat of 2MW.
A new dose information management system is proposed in this paper. This system consists of wearable devices and measures dose information simultaneously with position information. The dose information is visualized in being plotted on a map so high exposure dose area is identified clearly. This paper shows the result of implementing the dose information collecting devices and exposure dose visualization function.
The filtered venting system has the functions of preventing the overpressure breakage of a primary containment vessel of a reactor and reducing the release amount of the radioactive materials to the environment. One of the filtered venting systems is Venturi scrubbers to remove small particles of the radioactive materials. However, the physics of the two-phase flow in the Venturi scrubber has not been clarified. In this study, we report the experiment of the water-vapor two-phase flow in a Venturi tube. As the gas flow velocity at the throat of the Venturi tube increased, the supply amount of scrubbing water decreased and eventually stops, because the pressure at the throat increased with increasing the gas density. This phenomenon was a phenomenon caused by the gas compressibility. By visualizing the two-phase flow in the Venturi tube, we cannot measure the spraying from the water supply port, which is considered to affect decontamination of the Venturi scrubber. However, we confirmed that the water film on the wall at the diverging part of the Venturi tube, and the dispersed water droplets was formed intermittently from the water film.
A concentric-tube two-phase thermosyphon to use sodium for as a refrigerant for heat transportation use of the small nuclear reactor for Mars is promising. For the design, a model to apply a bubble pump theory, and to evaluate a rise of liquid level of heating section to greatly influence the limit heat transport rate was built. And after producing the thermosyphon using water and the low boiling point refrigerant as a refrigerant, and measuring the heat input that was necessary for a rise of liquid level of heating section, a calculated value and experimental value agreed on order is confirmed. And, in the case of sodium for a refrigerant the heat transportation performance of the concentric-tube two-phase thermosyphon of outside pipe 17mm in diameter was predicted theoretically. Then it was estimated that heat transport rate 6800W could be realized under Mars gravity by optimization of the inner tube diameter if difference of liquid level 500mm is secured. In addition, it was estimated that the theory limit of this thermosyphon is largely bigger than a conventional single tube thermosyphon.
A Very High Temperature Reactor (VHTR) is a next generation nuclear reactor system. From the view point of the safety characteristic, the passive cooling system should be designed for the VHTR as the best way of reactor vessel cooling system (VCS). Therefore, the gas cooling system with natural circulation is considered as a candidate for the VCS of the VHTR. Japan Atomic Energy Agency (JAEA) is advancing the technology development of the VHTR using the HTTR and is now pursuing design and development of commercial systems such as the 300MWe gas turbine high temperature reactor GTHTR300C (Gas Turbine High Temperature Reactor 300 for Cogeneration). The objective of this study is to investigate the heat transfer characteristics by natural convection in the vertical rectangular channel inserting the porous materials with high porosity. It is also to examine heat transfer characteristics of one side heated vertical rectangular channel with natural circulation in order to construct the passive cooling system of the VHTR. This paper describes the heat transfer coefficient and the amount of the removed heat in the proposed channel. In order to enhance the heat transfer, a porous material was inserted into the channel with high porosity. In order to apply to the VHTR such as GTHTR300C, it is necessary to perform a design study of the passive cooling system. This study is to refer to make design of the passive cooling system for the VCS.
The optimal condition to generate micro-bubbles and its effect of flow was investigated by using a modified pressurized dissolution method. Two nozzles were connected to modify the pressurized dissolution method in the present study. The effect of the amount of carbon dioxide (CO2) as the dissolved gas on the generation of micro-bubbles and its effect on the flow were studied. In the present study, the outlet pressure condition in converging-diverging nozzles plays the important role to generate micro-bubbles. Therefore, the optimal conditions to generate micro-bubbles depend on the percentage of dissolve gas and outlet pressure conditions.
Estimation of two-phase flow regime was studied by using sound data which were taken from the flow through vertical pipe made of acrylic in the anechoic box. The sound data was classified into four flow regime based on the video data. The effective sound pressure of two-phase flow could show the features in each flow, almost these features existed in low water flow rate. By using frequency spectrum analysis, two-phase flow could be classified into other three kinds of flow regime. These are gas phase and liquid phase are continuous, gas phase and liquid phase are discontinuous, and gas phase is discontinuous and liquid phase is discontinuous. The flow regime that gas phase and the liquid phase is discontinuous contains a slug flow and churn flow.
An air cavity is formed behind a spherical particle immersing into a liquid bath. The residual bubble attached to the sphere has an important effect on the performance of CaO particle used for desulfurization of molted iron. In this study, the bubble volume behavior of air cavity was observed for the spheres coated by two kinds of water repellent preparations. In the preceding report, the volume of the residual bubble was estimated theoretically when the sphere is immersed into water quasi-statically. The system energy (i.e., work of wetting, potential energy and surface energy of gas-liquid interface) was calculated at the breakdown of the axi-symmetric meniscus and the bubble volume was determined from the energy minimum condition. Here in this report, the bubble volume was theoretically obtained for the sphere immersion with finite velocities (0.05～25mm/s). First the critical depth of the sphere at the breakdown of liquid meniscus was considered from a dynamic equation in which the movement of triple-phase contact line is analyzed from the energy gradient on the sphere surface. Then the bubble volume was estimated at the critical depth from the same energy minimum principle as that treated in the preceding report. The experimental results of bubble volume can be roughly approximated by the model proposed in this study.
The quenching of high temperature stainless steel cylinder was investigated experimentally under the attachment of a honeycomb porous plate (HPP) on TiO2 nanoparticle deposited surface (NPDS). The experiments were performed under saturated conditions at atmospheric pressure. The results demonstrate that quenching was enhanced by HPP and/or NPDS in distilled water. Especially, for the case of HPP installed on the nanoparticle deposited surface, the time period from film boiling to nucleate boiling became extremely fast in distilled water. The enhanced quenching speed by the honeycomb porous plate was interpreted as liquid were supplied to the heated surface due to capillary action of the honeycomb porous plate. On the other hand, for the case of NPDS, the film boiling region disappears, and quenching speed was shortened. It was considered that these two effects combined are responsible for a shorter cooling time under the condition of HPP on NPDS.
Microbubble emission boiling (MEB) is an interesting phenomenon because of extremely high heat flux beyond the CHF. However, the mechanism of occurrence and the heat transfer characteristics are still not known. We carried out the experiment observation of bubble behaviors on the heating surface by using a platinum wire and a planar copper surface. The MEB was not observed in the subcooled boiling on the wire, but we found the similar behavior to the MEB on the planar surface. In the experiment, the huge coalesced bubbles were observed intermittently on the planar heat transfer surface, showing higher heat flux than the CHF. Because such large bubble was not formed on the fine wire, we are considering that the formation of the large coalesced bubbles has an important role on the occurrence of MEB. It is necessary to understand the effects of the size of heat transfer surface and the heat capacity. We also apply the prediction by the Microlayer Model.
Critical heat flux is one of the most important keys for various systems involving two-phase flow and heat transfer, such as nuclear reactors and boilers. But under downward flow condition, it is very difficult to predict critical heat flux because of complex flow caused by buoyancy. In series studies, obtained critical heat flux for downward flow were classified into four modes, however there are still not enough information about critical heat flux on stagnant region. In this investigation, critical heat flux experiment was carried out on a forced convective boiling system with a stainless steel tube of 15 mm in inner diameter and 200 mm in heating length. As the results, in the stagnant region, critical heat flux occurred in intermittent flow region, and the critical heat flux value was evaluated by a critical heat flux model taking into account of Kelvin-Helmholtz instability. In the high mass flux region, critical heat flux was observed in subcooled region. The obtained critical heat flux could be estimated by a departure from nucleate boiling model with modified departing bubble diameter correlation.
Experimental investigations have been carried out on fluid flow and heat transfer of opposing combined forced and natural convection of water adjacent to a heated vertical plate. The length of the plate, L, are L = 50, 100 and 150 mm. The plate is heated with a constant heat flux. The experiments cover the range of Reynolds and Rayleigh numbers as; 4×102 < ReL < 5×103, 4×106 < RaL* < 3×1011．The flow fields around the plate were visualized with dye. The result shows that the separation of the laminar boundary layer of forced convection appears first at the bottom edge of the plate when the non-dimensional parameter (GrL*/ReL2.5) = 0.4 and the separation point reaches to the reading edge when (GrL*/ReL2.5) = 3. The local heat transfer coefficients from the plates were subsequently measured with thermocouples. The result shows that the coefficients deviate from those of the pure-forced convection with the onset of flow separation at the bottom. It is also found that the combined convection region was determined as; 0.35 < (GrL*/ReL2.5) < 4.0.
This study deals with the heat transportation by oscillatory flow using Microencapsulated Phase Change Material Suspension (MPCMS) as the working fluid. The MPCMS consists of maicroencupsulated paraffin (diameter 10 μm, melting point 32 oC) and water. Experiments were performed to measure the thermal resistance of the heat transportation tube with inner diameter of 3 mm and length of 500 mm in the MPCMS mass concentration range of 0, 1, 3, and 5 wt%. The oscillation frequency and amplitude were 1 Hz and 90 mm. As the results, the thermal resistance indicates a tendency to decrease with increasing the mass concentration of MPCMS.
Jet noise reduction is one of the major issues in an aircraft engine. This paper describes a research on a jet noise reduction device, a chevron nozzle. Although chevron nozzles reduce the jet noise by an effect of promoting the mixing of high velocity jet and external flow, they tend to provide a thrust loss at the same time. In this research, experiments have been conducted using a small turbo-jet engine to evaluate the effect of chevron nozzles on the engine performance from both aspects of noise reduction and thrust loss. Five kinds of chevron shapes and two kinds of nozzle exit geometries have been used in the experiment and results have been compared with the reference nozzle.
Thermal power plants using steam turbines are required to be high efficient. However, various parts including nozzle guide vanes and rotor blades degrade through long time operations. To keep the high efficiency of the power plant, appropriate maintenances have to be done as necessary. It is important to understand aerodynamic characteristics in deteriorated steam turbines in order to realize the appropriate maintenance. In the present study, the aerodynamics of the deteriorated steam turbine is examined numerically. The nozzle and the rotor blade profiles of the 1st stage were obtained from a real steam turbine during the maintenance. The numerical boundary conditions were similar to the operation condition of the real thermal power plant. The numerical results show that the turbine power with the deteriorated blades is larger than that with the new blades because the mass flow of the main steam is larger. However, the turbine efficiency with the new blades is larger than that with the used blades.
A model has been developed to investigate the evaporation behavior of water droplets in a uniform duct flow under gas turbine environments. It is assumed that the change of water droplet temperature is caused by convective heat transfer through droplet surface. The temperature of a water droplet has been calculated by treating the droplet as a lumped mass with a representative temperature uniformly distributed in the droplet. The proposed model consists of two steps. The 1st step is the heat transfer calculation between a water droplet and surrounding air flow and the 2nd step is the calculation to take the latent heat of vaporization into consideration. The evaporation rate is calculated based upon mass transfer between water droplet and surrounding air flow.
Since non-condensable gas is contained in the steam at geothermal power plant, it is required for the purpose of simplifying a gas discharge system for plant efficiency improvement that emission gas should be made to fully cool and a volume flow should be made small. It is common to adopt how to cool the gas by direct contact for geothermal energy. We researched the cooling capability by the direct contact with the falling liquid film using structured packing. Experimental apparatus (Figure1) mainly consists of an evaporator, a cooling unit, and a vacuum pump. In the cooling section at the test equipment, Mixed gas(steam and air) and cooling water were floated by opposed flow : Mixed gas flows upward, and steam flows downward. Our conclusion in the experiment is that the cooling capability of falling liquid film has sufficient capability to cool mixed gas with high non-condensable gas density to cooling water temperture.
This paper proposes a meters-class anti-gravity loop heat pipe with Shirasu Porous Glass(SPG) as a wick for the smart thermal management of house. SPG shows excellent capillary force against anti-gravity and high permeability based on submicron pore radius and porosity over 50 %. Furthermore, five layers PTFE filters which have low thermal conductivity were applied on the SPG wick to reduce the heat leak from the evaporator to the compensation chamber. The proposed anti-gravity LHP were designed and fabricated based on the one-dimensional steady-state numerical model. Experiments of heat load were conducted under 2 m and 4 m anti-gravity conditions and those results showed the stable operation. Finally, the LHP performance was evaluated by calculating the thermal resistance between the evaporator to the condenser.
The aim of this study is to understand the spray properties and to clarify an influence of spray formation on the cooling efficiency of air. A phase Doppler anemometry (PDA) provides spray properties. Air temperatures before cooling and after cooling are measured. A impaction pin type nozzle and hole type nozzle are tested. For both nozzles, evaporation of droplets becomes inactive when humidity is high. For pin type nozzles, spray droplets exist in large area and are smaller than those of hole type nozzle. For hole type nozzle, droplets size is larger and spray area is smaller compared with pin type nozzle. The cooling efficiency for pin type nozzle is larger than that of hole type nozzle. The cooling efficiency becomes large when droplets are small and spray area is large.
For energy efficient operation of steam facilities, it is often required to measure a wide range of vapor flow rate. Clamp-on flowmeters are advantageous when it is necessary to measure vapor flow rates in various locations within one facility. On the other hand, ultrasonic flowmeters are well-known for their wide range of measurable gas flow rate. In the past, a clamp-on ultrasonic gas flow meter was developed, however there was a need for a damping material able to propagate ultrasonic wave through high temperature vapor pipe. The development of a clamp-on ultrasonic steam flowmeter with a high temperature endurance silicone rubber damping material is presented here. Silicone with high loss modulus has demonstrated good damping characteristic of ultrasonic wave. Using this newly developed flowmeter, 25A SGP 0.3 MPa steam flow can be measured up to 28m/s velocity.
We've improved our clamp-on steam flowmeter so as to measure as low as 0.3MPa by developing the high performance dumping material to reduce the ultrasonic noise that propagate the pipe wall. This newly developed clamp-on ultrasonic steam flowmeter was put into the actual factory site and held a field test. From the measurement results, we found that the flow rate and the valve opening setting signal were changing synchronously, so we believe correct measurement was being held. Although the change in the flow rate with respect to the valve opening setting signal was not smooth, this is considered to be due to a malfunction of the valve as we've already proven that the clamp-on flowmeter output change smoothly to the real flow rate. We will continue to make field tests to find issues in the real site and improve the quality of the flowmeter.
This paper presents the influence of the waveform on velocity profile measurement using ultrasonic time-domain correlation (UTDC) method. UTDC is a type of flow metering method that is applicable to the on-site calibration. The advantages of this method are the possibility of no-limitation of velocity range and applicability to flow fields without enough reflectors. Previous studies reported, however, that UTDC has also a limitation of velocity range because of the false detection. To investigate the influence of the waveform of pulsed ultrasound on the false detection, experimental measurements were performed at the national standard calibration facility of water flowrate in Japan. The frequency of ultrasound is 2 MHz and the flow rate is set from 200 to 400 m3/h to cover the velocity range of the conventional pulsed Doppler method. The results indicate that the false detection ratio increases with increasing the number of cycles in a pulsed ultrasound.
Effect of transient variation in the fraction of components in the multi-component fuels on the characteristics of a Bunsen flame has been investigated experimentally. Methane, ethane and propane were used as fuel, and air was used as oxidizer. The flame motion under the linear transition from one fuel component to another was examined. The experiment was conducted under two conditions; fuel flow rate constant and equivalence ratio constant. The transition time was varied from 1 s to 10 s. The flow velocity was 0.8 m/s and the equivalence ratio before the transition was 0.8. The variation in the flame height was measured, using high speed video camera at 125 fps. Experimental results showed that the flame height variation before and after the transition was larger in the case of fuel flow rate constant. On the other hand, in the case of equivalence ratio constant, the overshoot of the flame height during the transition was observed. The magnitude of the overshoot was larger for shorter transition time and for larger difference of volume fraction of fuel components. It is considered that a control method between an air supplier and a fuel supplier is important in the case of an actual combustor.
This study investigates the effects of mixing vane (MV) attached to a grid spacer on pressure drop for air-water annular flows in a vertical circular pipe of 16 mm I.D. In order to know the effects of MV, grid spacers with or without MV were installed in the test pipe. Three types of grid spacers with MV which has four MVs inclined 20, 30 degree from the vertical flow axis (4-MV20, 4-MV30) and two MVs inclined 30 degree (2-MV30) were used. In addition, two liquid injection methods (wall and center jet injections) are adopted to control liquid droplet entrainment faction in gas core. In the experiment, pressure drop along the channel upstream and downstream from the spacer was measured with a pressure transducer. In the analysis, a correlation of the pressure drop across the spacer is newly developed based on the present data.
In the previous study, we have evaluated countercurrent flow limitation (CCFL) at the sharp-edged upper end of the vertical pipe in the pressurizer surge line of a pressurized water reactor, and derived CCFL correlation by using air-water CCFL data with the diameter range of D = 19-140 mm. However, steam-water data are limited, and effects of fluid properties on CCFL are not clear. Recently we found steam-water data by Ilyukhin et al. (1988) with the diameter of 20 mm and pressures of P = 0.6- 4.1 MPa. By using these data and air-water data at 0.1 MPa, we evaluated effects of fluid properties on CCFL, and found that the viscosity ratio of gas and liquid is better than the density ration of gas and liquid (which was used by Ilyukhin et al.) to modify the Kutateladze parameters in the CCFL correlation proposed by Wallis.
In the Fukushima Daiichi Nuclear Power Plant accident, loss of cooling capability for Spent Fuel Pool (SFP) occurred. As a result, the water level of the SFP decreased due to the evaporation by the decay heat and was concerned that the spent fuels were damaged. Therefore, safety measures for SFP cooling are required in order to prevent the failure of the spent fuels. A portable spray system is considered as one of the safety measures for SFPs. When the spray system is applied to SFPs, a capability of spray cooling has to be evaluated to validate the applicability of portable spray systems as the safety measure for severe accident in SFPs. Then we started the research project to evaluate the spray cooling capability for SFPs. In this research project, the numerical simulation code for evaluating the capability of spray cooling is being developed. To develop this method, we focus on thermal-hydraulic behaviors of two-phase flow, which affect the cooling capability, such as Counter Current Flow Limiting (CCFL) phenomenon. This paper the research plan for spray cooling is described.
In recently, loop heat pipes (LHPs) have been attracting attention as a thermal control device. The LHP is a two phase loop system with phase change of a working fluid, and it is able to transport large capacity heat by using latent heat. However, previous studies of the LHP have been reported to exhibit temperature overshoot and vibration at startup. This transient phenomenon is considered to influence the behavior of the liquid-vapor interface in the evaporator. In this study, in order to understand and observe the behavior of the liquid-vapor interface in the porous material, the LHP experiment system that the maximum heat flux applied to the evaporator is 20 W/cm2 was designed and fabricated using a glass tube evaporator and far infrared heaters. The LHP could operate well. The vapor -liquid distribution in the groove and wick was observed. Effect of the initial vapor -liquid distribution in the vapor line, groove and CC on the startup characteristics was investigated. Three patterns of vapor -liquid phase displacement were observed. Temperature overshoot was observed when both the vapor line and grooves are filled with liquid.
Pool boiling is used for cooling in numerous thermal energy dissipation systems because of its high heat flux removal capacity. However, the heat removal in a pool boiling is limited by the occurrence of critical heat flux conditions (CHF). The primary concern in the thermal management requiring high heat flux removal is CHF enhancement. For such systems, we have demonstrated that pool boiling CHF using a honeycomb porous plate (HPP) was enhanced by more than twice (2.5 MW/m2) that of a plain surface. The plausible mechanisms for the CHF enhancement are liquid supply through (1) capillary action and (2) vapor escape channels from the top surface due to gravity. On the other hand, there is no significant CHF enhancement when the thickness of the HPP is about the same thickness as macrolayer formed under coalesced bubble. As a result of the dry out is occurred inside the porous medium during the hovering period of coalesced bubble, hence CHF is reached. In addition, the effects of the formation of coalesced bubble on CHF is far greater in the case of the downward-facing heated surfaces than the upward-facing heated surfaces. In this study, we report the effect of the heater surface orientation using two layers structured HPP on CHF enhancement in a saturated pool boiling. First layer, an alumina HPP is installed just on the heated surface, has very fine pores to supply water toward the heated surface by strong capillary force. Second layer, A metal HPP is placed just on the alumina HPP stated above, is structured to hold a sufficient amount of water in order to prevent the drying out completely inside the metal HPP during hovering period of a coalescent bubble.