As a result of the melt-down accident, it was assumed that the Units 1 to 3 reactor cores at the Fukushima Daiichi Nuclear Power Plant (1F) fell down from the bottom of the reactor pressure vessels (RPVs) to those of the primary containment vessels (PCVs). However, the conditions inside the PCVs were unknown despite the fact that it is totally essential to confirm the existence of fuel debris to examine the methodologies and procedures for the retrieval in the next stage. Under this circumstance, International Research Institute for Nuclear decommissioning (IRID) and Toshiba ESS jointly developed devices that can access to the lower part of a PCV and conducted surveys to confirm the conditions inside the PCVs. As well as the specifications for the remotely-operated devices developed and used for the surveys, this paper presents the findings, including the identification of sand, pebbles and clay-like deposits sitting in the bottom of pedestals; damaged components remaining in the pedestals；and solid substances adhered on the components, which considered to be melted fuel debris.

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Workable and operable robots remotely, in high radiation environment have developed for decommissioning Fukushima Daiichi Nuclear Power Station (1F). This paper has described the development of shape changing robot (PMORPH) and flexible structure arm (Muscle robot). PMORPH has developed to conduct investigations inside primary containment vessel (PCV) for confirmation of fuel debris diffusion conditions out of a pedestal. PMORPH can change its shape suitably to pass through a small diameter pipe and travel on rough planes with few electronic devices on it in order to improve its radiation tolerance. PMORPH investigated inside PCV, taking images with distribution of dose rate. A Muscle robot has developed to apply for fuel debris retrieval; the Muscle robot consists of hydraulic cylinders and springs without any electronic devices in order to improve its radiation tolerance, applying to various work of fuel debris retrieval with replacing the attachment tools and mechanism elements. Five (5) prototype robots have verified the Muscle robot performance.

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When the nuclear power plant suffered a severe accident, it is assumed that the debris reach to the bottom of the reactor pressure vessel and falls down to the bottom of the primary containment vessel. In that case, there is a possibility that the concrete of the primary containment vessel is eroded by Molten Core Concrete Interaction(MCCI) and the boundary of the primary containment vessel is fault, and the radioactive materials is released to the inside of the reactor building. In order to alleviate the influence of MCCI and to maintain the boundary of the primary containment vessel, we developed and installed Corium shield at Units 6 and 7 of Kashiwazaki Kariwa Nuclear Power Station. This paper describes the development policy and outline of Corium shield.

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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..

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Performance of the Filtered Containment Venting System (FCVS), developed by Tokyo Electric Power Company (TEPCO), has been verified by test programs include scaled filter venting test facility. The results of the tests showed TEPCO’s FCVS achieved high decontamination performance against aerosol particles under various conditions (various particle diameters, various carrier gas flow rates) and the FCVS of the tested design has been installed at Kashiwazaki-Kariwa Nuclear Power Station unit 6 and 7. For more advancement of nucler plant safety, we, now, try to improve the scrubber nozzle for attaing higher decomtamination performance. In this paper, we show some test results of decontamination factors (DF) for aerosol particles and pressure drops with various scrubber nozzles. The test results showed that, with increase of flow velocity from the scrubber nozzles, pressure drops on the nozzles increse, and the DF on water scrubber increase.

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When a severe accident occurs in nuclear power plant, irradiated iodine would be formed. The irradiated iodine is basically trapped into the cooling water by existing in the ionized condition. However, it is assumed that the gaseous iodine might be re-volatilized from the cooling water in the suppression pool chamber by the irradiation effect. The iodine volatilize behavior is said to be affected by pH control of the pool water. In this study, volatilize behavior of inorganic/organic iodine with pH control under gamma-ray irradiation is evaluated, based on simulated actual plant environments. The pH control to alkaline environment is effective to the inhibition of re-volatilization of iodine. The emission behavior of organic iodine gas is mainly due to the reaction between the epoxy paint of gaseous phase in PCV inner wall and volatilized inorganic iodine gas.

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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. According to previous studies, to secure the absorption performance, silver zeolite must be kept dry during the operation of the FCVS. For evaluating whether silver zeolite keep dry condition or not, steam blow test was conducted with gas flow rate and dew point distance which simulated the operating condition. Furthermore in previous studies, when vent stops for some reasons, and restart venting once again, silver zeolite will be wet and organic iodine absorption performance will be dropped. But if silver zeolite can be reconditioned during vent stop, this performance drop can be prevented. So we tested reconditioning method, and evaluate effectiveness.

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Silver zeolite AgX plays a catalyst role in the reaction of hydrogen and oxygen. When a severe accident happens, it is necessary to remove the hydrogen timely in order to prevent containment vessel from being broken. For this reason, we have investigated the suitability of AgX. As the results, AgX shows a weak catalytic activity under the conditions of steam 50%, hydrogen 3% and 75 ºC. On the other hand, when the temperature is higher than 140 ºC and residence time is as short as 0.05 second, 90% hydrogen can be removed.

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In Fukushima accident on 2011, molten core supposed to be leaked on the concreate floor in containment vessel. Development of small type of core catcher which can be installed in operating nuclear power plant is needed. In this investigation, the thermal deterioration of concreate and Magnesia-Carbon brick observed trough surface observation by microscope and measurement of compressive strength, weight reduction, and difference of void ratio. In addition, permeability measurement for burned concreate was operated. All factor was measured on room temperature situation with enough burned and cooled specimens. As a result, both of concreate and Magnesia-Carbon brick lost its weight, volume, and strength mostly relate with heating temperature. Permeability of concreate also rose as heating temperature rise. Magnesiacarbon brick which heated over 800℃ showed more severe deterioration than concreate heated same situation. Thermal deterioration of both material was observed.

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At Kansai Electric Power Co., we acquired licenses for restarting at seven plants including the operation extension plant and decided to decommission them at four plants. Currently, even in BWR, licensing and approval progresses, and in PWR, permission and approval of special facilities are progressing. In the re-operating plant, safety assessment evaluations have been conducted by business operators, and activities for maintaining and improving safety are constantly being carried out in NRA through revision of the novelty standards and the like. In this section, we will introduce the state of compliance with licenses and approvals by business operators, and the correspondence situation for maintaining and improving safety after re-operation, referring to the outline of the national nuclear policy.

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In severe accident at a nuclear power plant, we developed a new "alternative coolant circulation system" in order to continue the heat removal while maintaining the primary containment vessel boundary. In considering the alternative coolant circulation system, we decided to achieve the objective by combining existing facilities rather than creating new equipment. A system of alternative coolant circulation system was constructed using existing heat exchangers of residual heat removal system, emergency core cooling system piping and condensate makeup water pump as existing facilities. Alternative coolant circulation system satisfied Japanese regulatory provisions.

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In Fukushima accident on 2011, molten core supposed to be leaked on the concreate floor in containment vessel. After the loss of both the emergency core cooling system and IC core cooling, primary containment vessel (PCV) pressure increased. Water level measurement drifted because of water evaporation in the reference leg. Radiation level increased at a turbine building (T/B). There was a hydrogen explosion the after suppression chamber (S/C) wet venting. Modular Accident Analysis Program (MAAP) code analysis results and actual data suggest that depressurization of the reactor pressure vessel (RPV) began before its bottom failed. The data shows that the measured water level measurement drifted by more than 4 m owing to water loss in the reference leg. This is likely to have been caused by the high-temperature superheated core. Water should have been supplied to the water level reference leg through instrumentation piping. The system is named the back fill system. The system test was conducted and succeeded the recovery of water level gage.

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Safety enhancement of nuclear power plant is pushed forward by addition and remodeling of the facility based on a new regulation. The performance of the person operating it makes effort improvement by periodical training. We have an experience of repair work not having advanced as we think by an unexpected condition in the accident of Fukushima nuclear plant. In addition, it became clear from the conversation with the local government that there were some common issues by doing accident prevention training. Some results of analysis and measures plan for performance improvement in the accidents are introduced in this presentation.

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This study proposes to introduce a fusion reactor with an MA transmutation system into the nuclear fuel cycle. In this study, the effects of the coolants (molten salts, liquid metal, water, He gas) on the MA transmutation were evaluated. Neutron transport and burn-up simulations were performed using simple rectangular model. The results indicated that liquid metal PbBi most contributed to reduce the heavy metal nuclides because it less moderates neutrons and has low cross section of neutron absorption. In contrast, the amounts of the heavy metal nuclides reduction were small in the case of water coolant. Water effectively decelerates neutrons, which results in that MA turned to other heavy metal nuclides. Plutonium, mainly ²³⁸Pu and ²³⁹Pu, gradually accumulate in any case of coolants. The produced Pu is applicable to fast reactor fuels from the viewpoint of neutronics. The results, therefore, imply that the feasibility of the fuel cycle between light water reactor, fast reactor, and fusion reactor.

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In order to investigate heat transfer deterioration due to salt precipitation, a pool boiling experiment was conducted with a full-height 5×5 rod-bundle of BWR simulated fuel in sea water. The temperature on the center rod surface at the top spacer rose rapidly, since the flow area inside the top spacer was filled with the precipitated salt. Dryout occurred below the spacer, which results in the temperature escalation of the heater surface. On the other hand, the heater above the top spacer was cooled continuously by pool boiling.

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Internal flooding can be a threat to safety of nuclear power plants (NPPs) since it could cause both of an initiating event and loss of mitigation functions. In this paper, time transient of the flooding volume and velocity are considered to evaluate the water level in each area and the time of submersion of equipment. And also, to simplify the problem and solve the flow transient quickly, each area is considered as a node with volume and floor area, and the propagation path of the flooding water is considered as a junction with correlation equation for flooding rate evaluation.

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The prevention of excessive deformation by thermal ratcheting is important in the elevated temperature design of fast breeder reactors (FBR). In an experimental study that simulated a fast breeder reactor vessel near the coolant surface, it was reported that the long distance travel of temperature distribution causes a new type of thermal ratcheting, even in the absence of primary stress. Without excessive stress and travel distance, the inclement of the plastic strain derived from the above mechanism reduced each cycle of thermal transient, and finally reaches to approximately 0. In this paper, we investigate path independency on “saturated” distribution of plastic strain. For this purpose, we carried out several cases of finite element analyses with initial strain distribution, which simulated plastic strain distribution in the middle of accumulating. As the result, we confirmed that the shape of saturated distribution has small dependency on the shape of distribution in the middle of accumulating. Additionally, in the case using conservative initial strain distribution, further accumulation of plastic strain was not observed. This result suggests the possibility of the method that evaluate whether the accumulation of the plastic strain go beyond design limits.

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This paper describes the calculation method of debris bed cooling model for a Sodium-Cooled Fast Reactor (SFR). The existing method requires much calculation time in transient by a high calculated load due to small mesh and time step in boiling region. Thus, we study the method for calculated load reduction. The consideration of the cooling model equation suggests that the boiling region be steady behavior in transient. In addition, the behavior is also confirmed in the transient calculation for the typical SFR debris bed conditions. On the basis of the results above, we propose the new calculation method which is calculated in two steps by dividing the cooling model into two. In the first step, the transient calculation is performed with the cooling model in which the equivalent thermal conductivity of boiling region is changed to the simplified large value. By the simplification, we can use large mesh size and time step in boiling region, which reduces the calculated load. On the other hand, as a result of the simplification, the temperature and saturation distributions in the boiling region are not properly calculated in transient calculation. Thus, in second step, we calculate them with the steady calculation. The validation of this model is confirmed for the typical SFR debris bed conditions. The calculation speed of the proposed method is significantly faster than the existing method.

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Sodium has large electronegativity as compared with other metals and is therefore high chemical activity. Due to this property, it is an obstacle in sodium fire and stabilization treatment of sodium. Therefore, the authors devised method for extinguishing / treating that ionically combines Na with Cl of calcium chloride to produce neutral and stable sodium chloride. Therefore, the authors devised a method for extinguishing / treating that Na⁺ form an ionic bond with Cl^- of calcium chloride (CaCl₂) to produce sodium chloride (NaCl) possessing the strong ionic bond and high thermal stable. From the understanding of basic characteristics (thermal analysis, elemental analysis etc.) and small scale demonstration tests, we obtained results that can be expected to be applied.

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In a postulated severe accidental condition of sodium-cooled fast reactor (SFR), eutectic reaction between boron carbide (B₄C) and stainless steel (SS) may probably occur. Elucidation on the behavior of cited eutectic reaction is very important in terms of evaluation of core disruptive accidents in SFRs. For the first step to clarify the kinetic feature of B4C-SS eutectic reaction, the preliminary thermogavimetry-differential thermal analysis (TG-DTA) measurements using individual reagent were performed to obtain the fundamental information and to confirm the applicability of sample crucibles. It was found that alumina crucible was applicable in terms of eutectic behavior. When accomplishing the TG-DTA measurement at a high temperature, a temperature correction using standard materials is indispensable to minimize the significant measurement uncertainty. Nickel (m.p.:1728 K) as a standard material was used for maximum temperature limit of the temperature correction. After its temeprature correction, the TG-DTA meareuement of B4C-SS eutectic reaction were carried out. Based on the DTA curves at different heating rates, the kinetic parameters were roughly estimated by using Kissinger method.

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It is necessary to simulate a eutectic melting reaction and relocation behavior of boron carbide (B₄C) as a control rod material and stainless steel (SS) during a core disruptive accident in an advanced sodium-cooled fast reactor designed in Japan. On that account, a new project has been started to conduct eutectic melting experiments, thermophysical property measurement of the eutectic melt, and physical model development for the eutectic melting reaction. The eutectic experiments involve the visualization experiments, eutectic reaction rate experiments and material analyses. The thermophysical properties are measured in the range of liquid and solid states. The physical model is developed for a severe accident computer code based on the measured data of the eutectic reaction rate and the physical properties. This paper describes the project overview and progress by JFY2017.

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For the thermal-hydraulic design regarding a fuel assembly of sodium-cooled fast reactors, a subchannel analysis code ASFRE has been developed by JAEA. ASFRE was applied to numerical simulations of several kinds of water and sodium experiments as its validation studies and it was confirmed that pressure drops and temperature distributions measured in the experiments can be well reproduced. To enhance safety of sodium-cooled fast reactor, it is required to evaluate thermal-hydraulics in a core during decay heat removal by natural circulation. It is necessary to estimate radial heat transfer phenomena between fuel assemblies. In this study, a numerical simulation of a 37-pin bundle sodium experiment with radial heat flux was carried out and it was confirmed that ASFRE can be qualitatively reproduced temperature distributions in a fuel assembly affected by radial heat transfer.

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Overheating tube rupture of adjacent tubes arises from water/steam leak in steam generators of sodium-cooled fast reactors. It is very important to predict the tube wall stress (tube wall temperature) with a high degree of accuracy on evaluation of overheating tube rupture, and is crucial to estimate quantitatively material strength standard which is one of the major influencing factor. Therefore, in previous study, the authors carried out tube rupture experiments with rapidly-heating which were simulated the tube thermally-affected by sodium-water reaction jet using thick single-walled tube, and evaluated quantitatively failure hoop stress and failure time. Then, the authors confirmed that existing stress strength standard was applicable to thin diameter and thick-walled single tube in case of sodium-water reaction exceeding 1300 °C under practical steam generator operation conditions. In this report, rapid heating tube rupture experiments were conducted on the double-walled tube under the same experimental conditions as those in the single-walled tube experiments, and evaluated quantitatively failure hoop stress and failure time. The authors confirmed the validity of the existing stress strength standard under the conditions that the double-walled tubes were uniformly heated.

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A depressurization accident is the one of the design-basis accidents of a Very High Temperature Reactor (VHTR). It is an accident unique to a VHTR where air enters a large amount in the rector and oxidizes the graphite core structure, which may damage radioactive material emission and core shape. In order to predict or analyze the air ingress phenomena during the depressurization accident, it is important to develop the method for prevention of air ingress during the accident. We have conducted an experiment to obtain the mixing process of two component gases and natural circulation flow characteristics. The experimental apparatus consists of a cylindrical double coaxial vessel and a horizontal double coaxial pipe. The end of the double coaxial pipe are closed by a valve. The experimental method is as follows. This apparatus is filled with helium gas. Then, the inner cylinder is heated, and at the same time the outer cylinder is cooled. Finally, after confirming the steady state of the temperature of gas and wall in the apparatus, the valve of the horizontal double coaxial pipe are opened. From the experimental results, the gas mixing process and air ingress process (counter current flow, localized natural convection, molecular diffusion) occurring during the depressurization accident can be reproduced roughly. However, since the inner cylinder doesn't reproduce the thin flow path of the actual core, natural circulation flow occurred immediately after opening the valve.

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This reactor has many Blocks of Metal-Fuel (U-Pu-10%Zr) piled as Fuel-core. So the Fuel-core involves no metallic structure, the Fuel-core has high density of heavy-metal (7336kg/m3 ) in spite of having enough space of cooling channel, and is enclosed by the reflectors so that this reactor has high Plutonium-breeding capacity. The breed-burn wave goes down from upper burning blocks to lower blanket-blocks like Traveling Wave Reactor. When the top burning block goes over burn-up limit (7%), this block is removed from the reactor with no shutdown times, and the blanket-blocks are supplied every 2 years so that the breed-burn wave keeps stable. The blanket-blocks are made of recycled LWR-used fuel so that the Nuclear Fuel Cycle works effectively.

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In order to measure steam flowrate at an existing pipe, clamp-on type ultrasonic flowmeter is suitable. However, the steam wetness increases with the heat loss, and the liquid film and the droplets may disturb the ultrasonic propagation for the flowmeter. In this study, ultrasonic propagation in a pipe was measured under air-water two-phase flow. Two ultrasonic sensors located at up- and down-stream positions were used for the measuring ultrasonic time-of-flight that depends on the flowrate. The signals were evaluated using cross-correlation values. Further, the same experiments were conducted at wet-steam facility. The results showed that the accuracy of the flowrate measurement strongly depended on the flow regime. In wavy flow, the time-difference of ultrasonic propagation were good agreement with the reference values, and the cross-correlation value was high. However, the measurement became difficult in annular mist flow. Thus, it can be said that the liquid film at inner pipe wall was significantly affect the ultrasonic propagation.

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Recently, necessity of energy saving is more increasing in the industrial field with relation to energy cost increase. Measurement of steam flow rate is indispensable for an energy solution in facilities such as factories and power plants. However, a sensor installation for steam flow is generally difficult. Because steam pipes are necessary to be cut for the installation and the facilities needs to be out of operation temporarily. Accordingly, clamp-on type steam flowmeter is hoped to be developed. Therefore, we devised “heater method” for steam flow measurement. A circumferential heater is attached to the outside of a tube. Then, axial temperature distribution on the tube outside is measured. Steam velocity is analyzed on the basis of the temperature distribution, considering heat transmission in the pipe and thermal conductivity in steel pipe. We demonstrated that measurement value by the heater method fully agreed with reference value under the condition of steady flow in our previous paper ⁽¹⁾ . However, it takes a long time to analyze steam velocity because the analysis was manually executed after inputting measured temperature data. Therefore, we developed on-line measurement system by the heater method and we conducted measurement for unsteady flow using this system. Then, the system performance such as measurement response is discussed.

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Steam is mostly utilized as a means for thermal energy supply in industrial fields. It is important to be aware of the steam flow rate in the view point of energy management. However, steam becomes wet in many cases in the process to be sent through steam pipes to machinery using steam. It is well known that the wetness of steam sometimes causes measurement errors of the steam flow rate, and there has scarcely been the established method for estimating the error caused by the wetness of steam flow. Accordingly, we conducted the flow rate measurement of wet steam whose wetness was between 15 % and 40 % using a vortex flow meter, to clarify the measurement error caused by the wetness of steam, following our previous paper whose wetness was less than 15 %. The experiments were conducted with the conditions in changing the flow rate, pressure and wetness. As a result, the correlation between the measurement error and the flow condition was clarified.

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As demands for the reduction of energy consumption in steam heating system increased, it is very important to measure accuracy of mass flowrate in gas phase of wet steam. In this paper, we described the method for realizing steam flowmeter calibration by general-purpose steam facility. In addition, we conducted the calibration of clamp-on type ultrasonic flowmeter in dry and wet steam flow and discussed the uncertainty analysis of tested flowmeter.

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A steam flow meter is required in industrial fields to save and manage energy. An invasive flow meter, e.g. orifice plate flowmeter, is difficult to measure a steam flow rate accurately since it disturbs flow fields. Therefore, we focused on a clamp-on type ultrasonic flow meter. An ultrasonic flow meter based on propagation time difference method was developed and verified experimentally. A steam flow test facility was established to confirm validity of clamp-on type ultrasonic flow meter. The test was conducted under the condition of the steam flow rate of 0 – 40 kg/h and the pressure of 0.2 MPa. As a result, the propagation ultrasonic waveform with high S/N ratio was obtained using super critical angle transducers of the clamp-on type ultrasonic flow meter. In addition, the followability of measured steam flow rate value by the developed flow meter to the measured value by vortex flow meter was confirmed, therefore, the applicability of this system for steam flow was clarified.

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Counter-current flow limitation (CCFL) characteristics at the sharp-edged upper end (CCFL-U) of a vertical pipe with the diameter of D = 20 mm were measured, and effects of diameters on CCFL-U were evaluated by using CCFL-U data obtained from this study, measured by Doi et al. (2012) with D = 30, 45 and 60 mm, and reported by Richter (1981) with D = 19～140 mm. CCFL-U could be expressed by the Kutateladze parameters with the Laplace capillary length for the characteristic length in the range of D ≥ 25 mm. CCFL-U characteristics in small diameters of D ≤ 45 mm, however, strongly depended on the system characteristics such as the shape of the upper tank and the water level in the upper tank. CCFL-U characteristics in this study with D = 20 mm were clearly different with those for D ≥ 25 mm, and were different with those with D = 19 mm reported by Richter (1981). The reason of the differences is not clear.

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This research is the one about the Stream Generator (SG) of Pressurized Water Reactors (PWR). We particularly focus on the economizer and coolant distributions. Economizer is the device which controls the flow of the coolant in secondary loop, and there are two types of economizer in this research. One has both the vertical distribution plate and the baffle plates as the economizer, and the other only has the vertical distribution plate and as the economizer. To analyze the characteristics of SG, we made the SG model and the way to calculate the vapor mass flow rate in SG. Then, we inspected the accuracy of our modeling and the calculation. In addition, we examined the relation between the size of the economizer, the distribution of the coolant in secondary loop and the vapor mass flow rate. Then, we considered the causes of the results in this research based on the calculation.

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Since the downward flow becomes unstable as compared with the upward flow, the condition for the critical heat flux becomes very complicated. In this investigation, the liquid film amount at the critical heat flux condition was evaluated by using the liquid film extraction method. Experiments were carried out with several test sections with different heating lengths. As results the modified correlations of Koizuimi was proposed, and it can estimate the residual amount of the liquid film at the dry out condition in wide range. Although this correlation can explain the qualitative tendency well, the quantitative difference still exists. It might be the fundamental difference of flow structure between upward and downward flows.

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Nozzle guide vanes (NGVs) and rotor blades of gas turbines for thermal power generation are damaged due to erosion through long time operations. The deterioration of the NGVs and rotor blades can reduce of the turbine efficiency and the reliability. To understand the influence of the blade deterioration in detail, aerodynamic characteristics are examined numerically. The numerical simulations of a three-stage gas turbine were conducted for the new blade condition without deterioration and for the old blade condition with deterioration in the first stage. The numerical results were validated by comparing with a heat balance analysis based on the data obtained from a power plant during the operation.

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Experiments and CFD analysis were performed to investigate influences of deterioration gas turbine blades such as an increase of surface roughness of stators and increase of a tip clearance of rotors. It was confirmed that increase of the blade surface roughness affected the blade surface pressure and the total pressure loss in the wake. The relationship between the tip clearance of the rotor blade and fluid forces were compared. Though the numerical and experimental results showed qualitative agreement, further improvement were required to examine the quantitative characteristics.

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Although the high density water jet cooling of high temperature steel sheets is widely used in the rolling process of steel materials, its heat transfer characteristics and heat transfer mechanism in the high velocity water jet region are not sufficiently elucidated. This study is realized the cooling phenomenon of high temperature metal by such high speed water jet in the steady state, and experimentally examined the heat transfer characteristics of the water jet cooling on the high temperature metal. Experiments were carried out by cooling the copper block heated to about 900 ° C with water jet cooling, and the heat transfer characteristics in the vicinity of the stagnation point of water jet were examined from the temperature distribution inside the copper block of heat flux concentrating type. Experiments were conducted with the jet velocity in the range of 4.1 - 9.2 m/s. When the jet velocity was 4.1 m/s, the critical heat flux was found to be about 26 MW m² .

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Water spray is used for cooling the intake air of gas turbine in thermal power plant. Characteristics of spray play important role for efficiently cooling the intake air. The wind tunnel which can adjust temperature and humidity was used. The spray characteristics about two-type nozzles (Hole-nozzle and Pin-jet-nozzle) were measured by a Phase Doppler Anemometry (PDA) with three air-temperature conditions. The temperature drops between upstream and downstream were also measured. Cooling efficiency was calculated from them. The difference in characteristics of spray nozzle can discuss with PDA results of diameter and velocity. Cooling efficiency depends on spray nozzle type, floe rate of spray and air conditions.

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A cooling system utilizing a swirling flow generated downstream of a multi-elbow piping instead of the swirl tube has been proposed to for the ITER divertor cooling, on which an extremely high heat flux more than 10 MW/m² is supposed to impose. In this study, a visualization experiment to research the strength of swirling flow, its continuity and the turbulence statics and a heat transfer experiment under one-sided heating condition are conducted to evaluate the applicability. As a result, heat removal performance until approximately 5 MW/m² was observed although the critical heat flux wasn’t confirmed.

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A wireless tomography system has been developed to measure the real-time behavior of solid-liquid two phase flow in the centrifuge for controlling a rotational velocity and particle supply rate. The purpose of this study is development of real-time behavior of solid-liquid two phase flow measurement technique and image conductivity distribution in a rotating body because Acentrifugation technology for industry process should be improved to obtain more effective separation, shorten processtime and save energy but a technique for measuring the behavior ofsolid-liquid two phase flow in the centrifuges has not existed yet. The real-time behavior of solid-liquid two phase flow measurement and the rotational velocity control become innovative technologies. As technics to measure typical behavior of solid-liquid two phase, the process tomography is used because the processtomography is suitable for real-time measurement in a rotating body. Processtomography has high temporal resolution. This detector was used with the wireless because electrical cables are not available for centrifuges under high speed rotational condition. This wireless tomography system is used for a lab scale rotating machinery measurement experiment in thisresearch. The wireless tomography system was attached around the lab scale rotating machinery for the behavior of solid-liquid two phase flow measurement experiment. Consequently, we could wirelessly measure the behavior of solid-liquid two phase flow in the lab scale rotating machinery in real time and get data to image the behavior ofsolid-liquid two phase flow in the quantitative. So the wireless tomography system has developed and can measure the behavior of solid-liquid two phase flow in the lab scale rotating machinery in real time.

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In order to improve the efficiency of the generator, it is important to increase the cooling efficiency. The generator is composed of a rotor and a stator and is a grooved rotating double cylinder with grooves cut in the rotor blade axial direction. Therefore, in order to improve the cooling performance of the generator, it is necessary to elucidate the heat transfer flow characteristics due to the rotation of the heated rotor. In previous studies, it has been shown that the heat transfer coefficient of the groove surface tends to be higher than the heat transfer coefficient of the cylindrical surface, but knowledge on the change of the flow due to the heating of the inner cylinder and the temperature distribution of the flow is It is unknown. Therefore, in this research, velocity field and temperature field were visualized by PIV (Particle Image Velocimetry) and LIF (Laser Induced Fluorescence). As a result, it is suggested that the temperature distribution inside the groove measured by LIF may be related to the velocity field measurement result by PIV.

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Toward the improvement of performance of an air-cooled generator, an appropriate design for heat removal is important. An air-cooled generator consists of the outer and slotted inner cylinders, and the effect of the slotted inner cylinder on the heat transfer behavior was unclear. The present study experimentally and numerically clarifies flow behavior and heat transfer characteristics in rotating coaxial cylinders within an air-cooled generator. In the measurement of the temperature distribution, the Nusselt number increased with the increase of the Taylor number, the Nusselt number on the front-side wall of the rotor rotation was higher than that on the back-side wall. In the numerical simulation, the vortex structure and separated flow near the back-side wall was observed in the slit. It was suggested that the heat transfer on the back-side wall was obstructed because of the formation of separated flow.

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In this study, dependence of heat transfer on a heated disc in a cylindrical Hydrogen flow channel on the operating pressure has been investigated experimentally and numerically to understand heat transfer in the chemical vapor deposition reactor. Temperature on the heated disc was almost at steady state at P = 0.1 MPa and slightly higher than that at P = 0.07 MPa. In order to study the heat transfer for wide range of the operating pressure, two dimensional numerical simulation has been conducted. The calculation temperature on the heated disc agreed well with the experimental one except for the leading edge of the disc. At P = 0.1 MPa, the temperature distribution expanded to the gas flow inlet and then strong vortex was formed at the leading edge of the heated disc due to natural convection. On the other hand, the temperature distribution was compressed by the forced convection at P = 0.01 Pa. When mass transfer in the CVD reactor would be controlled by the diffusion of the raw material, the growth rate increased with the decrement of operating pressure and its uniformity was improved at P < 0.03 MPa.

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In the present research, the effect of tilting angle of heat exchanger on drain performance is clarified. A dip testing of a corrugated fin test piece under various tilting angles is conducted. When the test piece was close to a horizontal state, the fins are filled with water by capillary force and reached to a steady condition. On the other hand, when the test piece was tilted to close to a vertical state, filled water was continuously drained out from the test piece with time.

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A kettle reboiler in which horizontal tube bundle is set in a liquid pool of a working fluid is usually used as an evaporator in geothermal binary power system. Boiling heat transfer performance depends on not only vapor generation on the heat transfer surface of each tube but also agitation of liquid induced by rising bubbles. In the design of the kettle reboiler, it is important to understand the interaction between boiling flow behaviors and heat transfer coefficient in the tube bundle. In this study, forced boiling heat transfer experiments in a 3×8 horizontal tube bundle were conducted. Copper heating tubes with the outer diameter of 15 mm were horizontally installed in a vertical duct of the cross-section of 90 × 90 mm² in a lattice arrangement with the pitch of 22.5 mm. The cylinders were heated by inserted cartridge heater with constant heat flux. As the results, it was clearly observed that the heat transfer around the tubes in the downstream part was enhanced by rising bubble agitation. The effect was observed especially under the low heat flux condition. On the other hand, in the comparison among the tubes along the center, the heat transfer coefficient of the second tube from the inlet was enhanced under the high heat flux condition, and the value became higher than that of the 4th tube. The enhancement might be caused by increasing bubble nucleation site density, because superheated liquid was supplied from the 1st tube with partial boiling.

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A concentric-tube two-phase thermosyphon to use sodium as a refrigerant for heat transportation use of the small nuclear reactor for Mars is promising. For the design, the inside of the thermosyphon using the low boiling point refrigerant manufactured for the transparent material was observed. Then, even if the heat transport rate was close to the maximum, it turned out that the flow in the adiabatic section of outside of the concentric tube is two-phase flow. Moreover, it checked that about 70% of performance of the calculated value of ideal conditions was realizable by measuring maximum heat transport rate and comparing with the model calculated value.

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Conventional boiling systems, such as boilers, are composed of vertical parallel channels. Among various types of parallelchannel systems, In a parallel-channel system with multi-branches structure is anticipated peculiar behaviors in flow distribution and dynamic behavior, i.e. oscillation mode. Experimental investigation was conducted on the oscillatory flow instability in such system with complicated structure,. The experimental apparatus composed of two parallel units having two parallel channels in each, consequently four parallel channels. Experimental results revealed the effect of the location of inlet flow restriction on the flow stability and the flow oscillation modes. The threshold of flow instability was well estimated in some extent using modified G. Guido’s method.

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The final goal of this study is to clarify the enhancement of wall surface wettability of metal-material due to the Radiation Induced Surface Activation (RISA) in the basic design guidelines of the supercritical pressure water-cooled reactor (SCWR). Two specimens of the PNC1520 which is considered as a potential material of fuel-cladding tube of SCWR and SUS304 under high-temperature and high-pressure condition were used to measure the effect of surface wettability by γ-ray irradiation. Consequently, the RISA effect decreased when the specimen temperature over 250oC. Due to RISA enhancement, the surface wettability on the oxidized specimen was significantly improved rather than on the non-oxidized specimen.

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General heat transfer enhancement method between fluid and channels is to set extended heat transfer surface such a fin in a channel. However, it takes much cost and labor to produce a complicated channel. Therefore, the new heat transfer enhancement method to insert porous materials in a channel is considered. In this case, we can set porous materials easily and additionally. We used copper wires as porous materials and air as working fluids. Copper wires were inserted in a circular stainless tube intermittently and air was blown inside a tube. We changed s size of copper wires such the heat transfer surface and porosity, a diameter and measured an amount of heat transfer and pressure drop in the each conditions. Furthermore, we examined a degree of variation of an amount of heat transfer and pressure drop by finding the coefficients of variation. As a result, we found it is possible to reduce pressure drop largely while maintaining an amount of heat transfer by lowering the heat transfer surface of copper wires and porosity of an entire channel to some extent. And we can adjust an amount of heat transfer and pressure drop by compressing or expanding a porous material even if an amount of porous material is constant.

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There is a heat transfer enhancement method of inserting a porous material in flow channel, different from methods such as attaching the conventional laminated fin, etc. Therefore, evaluating the heat transfer and fluid flow performance of air-water heat exchanger inserting a porous material as a heat transfer accelerator in the flow channels. In this study, inserting the thin metallic wires as a porous material instead of the conventional laminated fin. The test section is a horizontal rectangular tube with 63 mm in width, 60 mm in height and, 240 mm in length. The test section is divided into 8 flow channels. The dimension per the flow channel is 4.5 mm in width and 60 mm in height. The length of the insertion region of a porous material is 200 mm. Using air as the working fluid. It is sent to a heater as forced convection by a ring blower, heated by a heater, and then sent to the test section. A cooling water flow channel is provided in the test section, and water at a constant temperature is flowed from the constant temperature bath by a pump. It was found the pressure drop can be further reduced, although the amount of heat removal decreases when inserting the thin metal wires, comparing to the amount of heat removal when using the laminated fin.

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