An uncertainty analysis method for control room habitability under toxic gas leakage accidents in cogeneration high temperature gas-cooled reactor (HTGR) is proposed to support risk-informed design of the plant. The method is applied to representative toxic gas leakage accidents in a hydrogen production plant by thermochemical Iodine-Sulfur water splitting method coupled to the HTTR gas turbine test plant. Variable parameters are successfully selected for the inputs to uncertainty propagation analysis by sensitivity analysis. Epistemic and aleatory uncertainties for each variable parameter are identified and are propagated using Latin hypercube sampling. The analyses show that the suggested method can successfully characterize and quantify uncertainties in the toxic gas concentration in control room. One important finding is that impact of uncertainty in surface roughness height on toxic gas concentration in control room is significant. The uncertainty is due largely to the simplification of the modeling of obstacles that exists between the reactor building and hydrogen production plant. The results lead us to the conclusion that toxic gas dispersion behavior analysis should combine two evaluation methods: dense gas dispersion model and computational fluid dynamics simulation.

This paper presents an improved empirical method for the prediction of perforation of steel plate reinforced concrete barriers by rigid missiles such as aircraft engines etc. The data of rigid missile experiments newly performed in China in 2018 were collected, and numerical analysis and existing empirical methods of perforation damage analysis were compared. Based on the existing formulas, an improvement is proposed to predict the penetration effect of typical rigid projectiles on steel reinforced concrete barriers in aircraft impact or tornadoes more accurately.

A thermochemical water-splitting iodine-sulfur process (IS process) is one of the candidates for the large-scale production of hydrogen using heat from solar power or nuclear energy. Severe corrosive environment which is thermal decomposition of sulfuric acid exists in the IS process. To achieve the IS process, one of the key factors is the development of structural materials against the severe corrosive environment. A hybrid material with the corrosion-resistance and the ductility was made by a silicon powder plasma spraying and laser treatment. The hybrid materials had excellent corrosion resistance in the condition of 95 mass% boiling sulfuric acid. This was attributed to the formation of SiO₂ on the surface. To confirm the manufacturability of a container using the hybrid technique, the production of container with a welded part, a chamfer and a curved surface was experimentally tried. A configuration of the container is 150mm inside diameter, 120mm in height and 6mm in thickness. In the production process of the container, the treatment of the welded part is very important. The preliminary examination of the welded part was carried out using Hastelloy C276 plates. The distance between fusion line and the plasma spraying and laser treatment area was decided to 15 mm from the results of the preliminary examination. The container firstly divided into the upper and the lower part, the inside of them was conducted the plasma spraying and laser treatment and then they were welded by a tungsten inert gas. Finally, the surface of the welded part was covered by plasma spraying and laser treatment. There was no detachment in the plasma spraying and laser treated layer of the container after the laser treatment. We succeeded in the production of the container using the hybrid technique.

Ratcheting is one common phenomenon under excessive earthquakes. This study focuses on clarifying the ratcheting mechanism of beams subjected to the combination of gravity and seismic loadings. The sinusoidal excitations or their combinations represented the seismic loadings. Gravity acted as load-controlled loading and was applied by the self-weight of the beam as well as the additional mass put at the free end of the beam. A cyclic acceleration was applied at the base of the beam to provide the source of dynamic loading assumed as seismic loading. Equivalent loading conditions were put in experiments as in numerical analysis to validate the nonlinear finite element analyses. The analogy between thermal ratcheting and vibration ratcheting was considered to investigate the characteristics of cyclic accelerations. The frequency effects on the applied alternating accelerations were also investigated in this research. The results show that input waves containing the lower frequency components than the natural frequency are more damageable comparing to that containing higher frequency ones. Besides, the component with lower frequency contributes more to the occurrence of ratcheting. The characteristics of dynamic loading due to base acceleration depend on the loading frequency. The low-frequency acceleration acts like load-controlled loading, while the high-frequency acceleration is close to displacement-controlled loading.

In this study, the SAMPSON code was modified to evaluate severe accidents in a spent fuel pool (SFP). Not only the SFP but also upper spaces of the SFP, walls of the reactor building, and the blowout panel were included. Air oxidation models obtained by the Zircaroy-4 cladding (ANL model) and the Zircaroy-2 cladding (JAEA model) were included in the modified SAMPSON code. Experiments done by Sandia National Laboratory using simulated fuel assemblies equivalent to those of an actual BWR plant were analyzed by the modified SAMPSON code to confirm the function of fuel temperature calculation in the event of loss of fuel cooling in the SFP. The rapid fuel rod temperature rise due to the Zr air oxidation reaction could be reasonably evaluated by the SAMPSON analysis for the radial propagation experiment. The effect of the oxidation reaction model was evaluated by the analysis of the SFP assuming no initial water level. There was almost no difference in the maximum temperature transient of the fuel rod surface between the ANL and JAEA models since the extent of the oxidation reaction was limited by the amount of oxygen supplied in the current analysis conditions. The analysis was conducted with different initial water levels which were no water, water level at bottom of active fuel, and water level at half of active fuel. The present analysis showed that the earliest temperature rise of the fuel rod surface occurred when there was no water in the SFP and natural circulation of air became possible.

In Japan, as in other countries, bentonite-based buffer materials are expected to play roles in reducing stress from rock masses and mitigating nuclide migration in the geological disposal of high-level radioactive waste. These roles are achieved by ensuring buffer density and thickness based on the swelling characteristics of clay. However, in practical construction, we should also consider bentonite buffer piping erosion, a phenomenon in which the buffer surface is destroyed by groundwater flowing between the buffer and rock. Such piping erosion may be a serious issue for maintaining an engineered barrier for radioactive waste in the geological system. In this study, the dynamic behaviors of piping erosion were experimentally investigated. In the experiments, 500 mm φ × 600 mm height compacted bentonite specimens were placed in a cylindrical acrylic cell and distilled water was continuously injected at a flow rate of 0.1 L/min from the bottom and the side of the cell. The amount of bentonite that flowed out of the cell was measured by turbidity of the suspended clay in the drainage water. In the results, while the bentonite specimen swelled between about 10 and 20 days and attached to the inside wall of the cell, a dominant flow channel (piping) was observed between the swelled bentonite specimen and the inside wall of the cell. In addition, the relationship between the accumulated amounts of injected water and eroded bentonite showed a constant slope on double logarithmic plots. Such behaviors indicate that the shear stress due to water flow locally exceeded the swelled bentonite’s shear strength. Furthermore, the slopes were similar to those already reported from a test using a smaller size cell. These results suggest that piping erosion proceeds with a simple regularity and that piping erosion in actual-scale bentonite buffers can be predicted using small-scale data.

The authors are developing the design concept of the pool-type sodium-cooled fast reactor (SFR) that addresses Japan’s specific siting conditions such as earthquakes and meets safety design criteria (SDC) and safety design guidelines (SDG) for Generation IV SFRs. The development of this concept will broaden not only options for reactor types in Japan but also the range and depth of international cooperation. A design concept of 1,500 MWt (650 MWe) class pool-type SFR was thought up by applying design technology obtained from the design of advanced loop-type SFR, named JSFR, equipped with safety measures that reflect results from the feasibility study on commercialized fast reactor cycle systems and fast reactor cycle technology development, improved maintainability and repairability, and lessons learned from the Fukushima Daiichi Nuclear Power Plants accident. The design concepts of a reactor vessel (RV) and its internal structures have been investigated whether they could withstand severe seismic conditions in Japan and thermal loads. The design adopted enhanced RV support structure, enhanced conical-shaped core support structure, a thickened knuckle part of the RV, and a flat plenum separator with ribs. A three-dimensional steady-state thermal hydraulic analysis of the RV revealed that the temperature difference of the upper and lower surfaces of the flat plenum separator could be effectively reduced by installing layers of thermal insulation plates. The authors have also conducted a transient analysis of loss of flow type anticipated transients without scram events to evaluate the feasibility of a self-actuated shutdown system. Moreover, the configuration of the decay heat removal system has been investigated considering sufficient utilization of natural circulation capability of sodium coolant, heat removal capacity of each cooling system, conformance with design requirements, and recommendations of SDC and SDG such as diversity and redundancy of components.

Critical heat flux (CHF) plays an important role as the upper limit of heat dissipation process during the in-vessel reactor external reactor vessel cooling (IVR-ERVC). IVR-ERVC is regarded as a very effective way to release the decay heat after the core melt. This severe accident mitigation countermeasure has already been applied in some advanced PWRs and is considered to be used in some advanced BWRs as well. In this paper, experiments of the macro-fin structure positioned on a slope with 5° and 10° downward inclination angle were carried out. The results of the copper bar experiments were then analyzed and compared with the results obtained from a copper bare block experiments which were conducted previously at the same test facility. It was found that the CHF of the finned structure decreased compared with that of the bare surface. This shows that the current CHF data of the bare surface cannot be used directly to the design of new advanced BWRs. The hot/dry spot theory designated for the copper bar was brought up with an attempt to explain the difference of copper bare block and copper bar. Results show that using this theory can explain the decrease of CHF of the copper bar.

High-temperature control blade degradation tests simulating a beginning phase of a severe accident in BWRs has been comprehensively performed in Japan Atomic Energy Agency (JAEA). In the latest test, a mock-up of BWR bundle material has been investigated under postulated Fukushima Dai-Ichi (1F) unit 2 accident conditions in a complex heating transient scenario including a phase of lack of available steam. The progress in control blade degradation was monitored with help of an in situ video and the detailed analysis of the solidified metallic melt, so-called “metallic debris”, was carried out by conventional SEM EDS method. These results indicated that the composition of the metallic debris at different elevations has been significantly changed due to the redistribution and relocation of steel elements under the influence of B and C, sometimes accompanied by a formation of high-melting-point layers. The results of this paper significantly contribute to the physical understanding of control blade degradation phenomenology during beginning phase of a core degradation for a special case of steam-starved conditions at 1F unit 2.

This paper describes a measurement technique for a two-dimensional (2D) velocity vector profile in the two-phase bubbly flow. The Ultrasonic Velocity Profiler (UVP) method, which is a nonintrusive technique applicable for real-time measurement, is proposed to simultaneously obtain a 2D velocity vector of the bubble and liquid phase in the bubbly flow. To achieve this aim, transducers with special configuration and developed signal processing is applied to the UVP system to reconstruct and decompose a 2D velocity vector of the bubbles and liquid. To confirm the applicability of the improved UVP, the experiment is conducted on a rectangular bubble column flow loop. The 2D velocity vector profile measurement in two-phase bubbly flow is performed experimentally and the applicability of the measurement validated by comparison with other methods.

Core Disruptive Accident (CDA) has been considered as one of the important safety issues in the severe accident evaluation of Sodium-cooled Fast Reactor (SFR), since SFR core is not in the most reactive configuration. Initiating Phase (IP) is the earliest stage of CDA, and SAS4A code was designed to precisely model the IP event progressions. Phenomena Identification and Ranking Table (PIRT) approach was applied to the SAS4A code in order to enhance its reliability in this study. This paper describes the study on the validation method and results of the validation of SAS4A using PIRT approach in a typical SFR. In this study, SAS4A validation was based on the generic PIRT process: (1) definition of issue of SAS4A code, (2) definition of PIRT objectives of using the PIRT approach, (3) selection of potential plant designs, (4) selection of potential scenarios, (5) selection of the figure of merit (FOM), (6) partition of scenario into convenient time phases, (7) definition of considerable physical regions, (8) identification of the phenomena, (9) ranking the important phenomena, (10) development of the code validation test matrix, and (11) test analyses for validation corresponding to the test matrix. Unprotected Loss Of Flow (ULOF) is selected as the specific scenario, since it is one of the most important and typical events in CDA. The average fuel temperature which corresponds to the energy release is selected as the FOM, since the energy release due to power excursion is important to evaluate whether the influence of CDA can be confined in the vessel. The phenomena are identified by the investigation of ULOF event progression and by making the system decomposition and hierarchy. These phenomena are ranked according to importance to the FOM. In the test matrix, the key phenomena are associated with SAS4A models, and the test cases used for validation of each model are shown. The results of the test analysis corresponding to this matrix show that the SAS4A models required for the IP evaluation were sufficiently validated. Furthermore, the validation with this matrix is highly reliable, since this matrix represents the comprehensive validation that also considers the relation among physical phenomena. In this study, the reliability and validity of SAS4A code were significantly enhanced by using PIRT approach to the sufficient level for CDA analyses in SFR.

This paper describes the effect of local plastic component on J-integral and crack opening displacement (COD) evaluation of a circumferential penetrated crack, applicable to the leak before break (LBB) assessment for sodium cooled fast reactor (SFR) components. J-integral and COD evaluation methods are generally formulated as a summation of elastic and plastic components, and so far many evaluation formulae based on these two components have been proposed. However, strictly, the plastic component consists of local plastic and fully plastic components. Many of the conventional evaluation methods consider only the fully plastic component as the plastic component. The reason for this is that the effect of the local plastic component is much smaller than that of the fully plastic component excluding materials with extremely small work hardening. In contrast, for materials with high yield stress and small work hardening, such as modified 9Cr-1Mo steel which is one of the candidate materials for SFR’s piping, the effect of the local plastic component on J-integral and COD cannot be ignored. Therefore, the authors propose formulae taking the effect of local plastic component on J-integral and COD into account, based on finite element analysis (FEA) results, so that it is easy to apply to crack evaluation. The formulae will be employed in the guidelines on LBB assessment for SFR components published from Japan Society of Mechanical Engineers (JSME).

In sodium-cooled fast reactors, free liquid surfaces are found in several important components, including reactor vessels; sloshing due to earthquakes is one of the major concerns in design. Especially in cases where seismic isolation systems are installed to prevent or reduce damages to facilities due to earthquakes, periods of vibration are lengthened and become close to natural sloshing periods. As a result, sloshing is more likely to occur. In severe seismic conditions, sloshing waves are considered to even reach a roof slab of a reactor vessel. The structural integrity of roof slabs is required to be evaluated against sloshing impacts. However, there is no widely recognized evaluation method for sloshing impact pressure on flat roofs yet. Therefore, in this paper, a simplified evaluation method is proposed based on Wagner’s theory, which is a well-known classic theory for evaluating impact pressures on rigid wedges dropping on water surfaces. In the proposed method, we assume an equivalent wedge on a flat roof. The impact pressure on the equivalent wedge is evaluated by applying Wagner’s theory. Computational fluid dynamics analysis is conducted to confirm that a key assumption of Wagner’s theory is applicable to the evaluation of sloshing impact on a flat roof. In addition, the predictability of the proposed method is investigated by comparing literature data of sloshing experiments with the estimated values.

A dry method is one of practical methods for decommissioning the TEPCO's Fukushima Daiichi nuclear power station. Japan Atomic Energy Agency (JAEA) has been evaluating the air cooling performance of the fuel debris by using the JUPITER code based on an incompressible fluid model and the CityLBM code based on the lattice Boltzmann method (LBM). However, these codes were based on a uniform Cartesian grid system, and required large computational time and cost to capture complicated debris structures and multi-scale flows at the actual reactor scale. The adaptive mesh refinement (AMR) method is one of the key techniques to accelerate multi-scale simulations. We develop an AMR version of the CityLBM code on GPU based supercomputers and apply it to thermal-hydrodynamics problems. The proposed method is validated against free convective heat transfer experiments at JAEA. Thanks to the AMR method, grid resolution is optimized near the walls where velocity and temperature gradients are large, and the temperature distribution agrees with the experimental data using half the number of grid points. It is also shown that the AMR based CityLBM code on 4 NVIDIA TESLA V100 GPUs gives 6.7x speedup of the time to solution compared with the JUPITER code on 36 Intel Xeon E5-2680v3 CPUs. The results show that the AMR based LBM is promising for accelerating extreme scale thermal convective simulations.

Installation of surveillance cameras in nuclear power plants is critical to protecting the facilities against terrorist attacks or monitoring the reactor operator. This has led to large amounts of video surveillance data, creating a demand for automatic detection of anomalies or suspicious movements. Tracking human motion from video sequences is a notable technique used for detecting anomalies in human behavior and is currently achieved with the use of a depth camera. However, depth cameras require a complicated camera system and their field of view is limited. To overcome this problem, there is a need for recognizing human motion in wide-angle images – a view that often causes distortion. In this study, we devised a method for tracking human motion through wideangle image distortion. The main contribution of this study is a methodology that automatically estimates the transformation parameters needed to improve the accuracy of motion recognition; these parameters are applied to a distorted wide-angle image in every frame. We propose a new multi-layered convolutional neural architecture for estimating the locations of human joints in images and transformation parameters simultaneously. When applied to distorted wide-angle images, the robustness of our method is demonstrated through a quantitative evaluation of human joint location prediction. In addition, we compare our method with a motion tracking system and an infrared-camera-based motion capture system to demonstrate its ability to handle wide-angle and close-range images.

In this study reasonable 2D model was established by using FLUENT for start-up of analysis and evaluation of heat transfer flow characteristics in 1/6 scale model of VCS for HTTR. The pressure vessel temperature was set around 200 ℃, in which the ratio of heat transfer via natural convection has been numerically predicted to be around 20-30% of total heat removal in previous studies. This temperature is useful for the analysis code validation in the prediction of temperature distribution of components such as pressure vessel which is heated up by turbulent flow of natural convection. The numerical results of upper head of pressure vessel by the k-ω-SST intermittency transition model, which can adequately reproduce the separation, re-attachment and transition, reproduced the test results including temperature distribution well in contrast to those by the k-ε model in both cases that helium gas is evacuated or filled in the pressure vessel. It was emerged that any local hot spot did not appear on the top of upper head of pressure vessel where natural convection flow of air is separated in both cases. In addition, the plume of high temperature helium gas generated by the heating of heater was well mixed in the upper head and uniformly heated the inner surface of upper head without generating hot spots.

We extended the first version of fission product (FP) chemistry database named ECUME (Effective Chemistry database of fission products Under Multiphase rEaction). The extended ECUME consists of three kinds of datasets: CRK (dataset for Chemical Reaction Kinetics), EM (Elemental Model set) and TD (ThermoDynamic dataset). The present ECUME is equipped with the CRK for the reaction of Cs-I-B-Mo-O-H system and Ru-N-O-H system in gas phase, the EM for the Cs chemical reaction with stainless steel (SS) (Cs chemisorption onto SS) and the TD for CsBO₂ vapor species and solid Cs₂Si₄O₉ and CsFeSiO₄. A FP chemical reaction calculation in gas phase with the CRK of Cs-I-B-Mo-O-H system has shown the necessity of consideration of chemical reaction kinetics for more accurate estimation of Cs and I release amount into environment. The EM for Cs chemisorption has successfully achieved more accurate estimation of Cs distribution in a reactor by reproducing the effects of CsOH vapor concentration in gas phase and Si content in SS which were not considered by the existing model. The high quality vapor pressure data for CsBO₂ vapor were evaluated based on the result of a high temperature mass spectrometry. Cesium species at high temperature can be estimated by the thermodynamic data with high reliability. Thermodynamic data for solid Cs₂Si₄O₉ and CsFeSiO₄ were successfully evaluated by the experiment and ab-initio based methodology, respectively. These results have shown the validity and importance of the ECUME application for the more accurate evaluation of FP chemistry during transportation in a reactor under a LWR severe accident.

An aerosol particle capturing by a gas-liquid interface has been applied in many industries. One example of this is the radioactive aerosol removal system in the safety system of the nuclear reactor such as filtered venting system. In this study, to clarify the capturing behavior of aerosol particles by the gas-liquid interface, we developed a direct observation technique of the aerosol particle behavior and have been performing the observation experiment of the aerosol particle capturing behavior near the gas-liquid interface. In the experiment, we observed the aerosol particle capturing behavior on the gas-liquid interface of a single droplet. The capturing behavior of the aerosol particles near the droplet interface was observed by using a high-speed and high-resolution video camera and a fiber light. In this observation, we confirmed that aerosol particles were captured at the gas-liquid interface of the droplet after the velocity of the aerosol particles decreases near the droplet. On the other hand, some aerosol particles closing to the droplet did not reach the interface and were not captured. In comparison between the behavior and Stokes number, particles with higher Stokes number were easy to be captured at the gas-liquid interface. Especially, particles with much higher Stokes number penetrated through the gas-liquid interface and were captured inside the droplet. This capturing behavior was not considered in previous studies because the capturing behavior of a “solid” single fiber applied to the capturing behavior of aerosol particles with a droplet. Thus, this study made clear that Stokes number affects capturing behavior. In addition, we confirmed both soluble and insoluble particles in water also were captured in a droplet in high Stokes number. This result means that particles can be captured in a droplet in high Stokes number even if aerosol particles are insoluble in water.

Understanding the effect of uncertainties of Core Disruptive Accident (CDA) scenarios on debris bed coolability on a core catcher is required for decision making on design options to mitigate a CDA consequence. For the understanding, a huge number of calculations are required but are extremely difficult to perform because a huge number of calculations require much calculation time to solve non-steady equations in the coolability calculation model. Thus, we applied Artificial Neural Network (ANN), which is one of models for machine learning, to debris bed coolability calculations. The application of ANN is expected to exponentially improve the calculation speed of debris bed coolability because ANN provides results from experimental rules learned through training without solving non-steady equations. The application is in three steps. Firstly, we created many data for training ANN and validating the trained ANN through coolability calculations parameterizing main dominant inputs (particle diameter of debris bed, porosity of debris bed, etc.) by using Latin hypercube sampling. Secondly, ANN was trained and validated with the created data. The accuracy rate of the results by the ANN to the validation data exceeded 99%. In addition, the calculation time using ANN was micro seconds order. Finally, through demonstration calculations, it was confirmed that we can easily understand the effect of uncertainties of CDA scenarios on debris bed coolability owing to results visualization based on a huge number of parametric calculations using ANN. Thus, the application of ANN to debris bed coolability calculations should contribute to the decision making on design options to mitigate a CDA consequence.

Prediction of the fuel debris location and distribution inside the primary containment vessel (PCV) of the Fukushima Daiichi Nuclear Power Station is important to decide further decommissioning step and strategy. The radiation measurements in the past internal investigations have not yet provided enough information to predict fuel debris location and its distribution inside PCV. To support further measurement efforts, we simulate the detector response inside the PCV and finding possible gamma radiation unique to fuel debris. The calculation result could provide useful information for future gamma detector developments.

Fully natural circulation decay heat removal systems (DHRSs) are adopted for sodium fast reactors, which is a passive safety feature without any electrical pumps. It is needed to grasp the thermal-hydraulic phenomena in the reactor vessel and evaluate the coolability of the core under the natural circulation not only for the normal operating condition but also for severe accident conditions. In this paper, the numerical results of the preliminary analysis for the sodium experimental condition with the PLANDTL-2, in which the core and the upper plenum with a dipped-type direct heat exchanger (DHX) were modeled, are discussed to establish an appropriate numerical models for the reactor core including the gap region among the subassemblies and the DHX. The transient analysis simulating the reactor scram reveals that the 3-dimensional large scale flow structure is developed through the gaps in the whole of the core area during the reactor scram. The steady-state analysis coinciding Richardson number between the PLANDTL-2 and the reactor operation condition reveals that the hot spot and cold spot appear depending on the location of the DHX, which is caused by the complex thermal-hydraulic phenomena driven by the natural circulation. From these preliminary analyses, the characteristics of the thermal-hydraulics behavior in the PLANDTL-2 to be focused are extracted.

If pressurized water or its vapor leaks from a ruptured heat transfer tube in a steam generator of a sodium-cooled fast reactor, a high-velocity, high-temperature, and corrosive jet with sodium-water chemical reaction may cause tube failure propagation. In this study, an analytical method was developed to predict the occurrence of tube failure propagation by overheating rupture. This method consists of the elemental analytical models for a sodium-side temperature distribution formed by a reacting jet, water-side thermal hydraulics, heat transfer between a fluid and a tube, and tube failure by internal pressure. To evaluate the tube failure propagation in a short computation time, these models are based on the experimental data, semi-theoretical correlations, or one-dimensional equations. Applicability of the proposed method was investigated through the numerical analysis of an experiment on water vapor discharging into the liquid sodium. This analysis demonstrated that the method could predict the occurrence of overheating rupture and provide conservative results. While the proposed method is useful for high-speed computations, this method evaluates a high temperature region with a large conservativeness in some cases. To improve this conservativeness, a Lagrangian particle model for the reacting jet was also developed as an alternative method. The numerical analysis by this model showed that the discharged gaseous particles spread with particle-particle and particle-tube interactions.

Seismic buckling of vessels is one of main concerns for the design of fast reactor plants in Japan. Rational design is important because of two conflicting requirements; thicker walls are preferable to prevent seismic buckling of vessels, while excessively thick walls introduce large thermal stress causing unacceptable creep–fatigue interaction damage. In previous studies, we discussed evaluation methods of seismic buckling probability of vessels by taking account of seismic hazards in order to rationalize seismic buckling evaluation, and proposed a rule for seismic buckling of vessels based on the load and resistant factor design method. The proposed rule is expected to widen design window regarding seismic buckling and contribute to more reasonable design of vessels of fast reactors. However, there is still a room for more rational design. The proposed method deals with only seismic load, but in actuality, dead weight and internal pressure also exist. The existence of these loads contributes to reducing the buckling probability because axial compressive load decreases. In this study, the rule was expanded so that dead weight and internal pressure can be taken into account. Furthermore, the influences of dead weight and internal pressure to seismic buckling evaluation were discussed. As result, it was shown that approximately 10 to 20% of further rationalization of allowable seismic load could be achieved by considering dead weight and internal pressure in the evaluation. In addition, it was found that the previously proposed design rule, not considering dead weight and internal pressure, includes approximately 2 to 10 times margins in terms of seismic buckling probability.

Nuclear power plants are equipped with a hardened containment venting system and a filtered containment venting system to avoid a primary containment vessel (PCV) break by over-pressurization. Considering the gas release from these venting systems, it is desirable to delay venting start time or to avoid operation of the systems by ensuring sufficient PCV cooling in view of the time needed to evacuate residents and the half-life of the radioactive materials. We have been developing a new drywell cooler specialized for severe accidents (DWC_SA) as one additional PCV cooling method. The DWC_SA can cool the PCV by continuous steam condensation using downward flow induced by natural force without any electrical devices in the PCV. We have already conducted element tests using a small tube bundle to confirm the occurrence of continuous downward flow with the DWC_SA configuration and to measure heat transfer rate of condensation with non-condensable gas (NCG). The heat transfer model was validated using the element test data. Here, to estimate plant performance of a representative advanced boiling water reactor (ABWR) applying the DWC_SA, we carried out a severe accident analysis using the MAAP code. We originally installed the heat transfer rate map of the DWC_SA as a function of pressure and NCG ratio in the code. The velocities using in the heat transfer model were estimated from CFD analysis. According to the severe accident analysis results, the DWC_SA can extend venting start time sufficiently to allow evacuation of residents in the accident scenario (LOCA with large pipe failure + ECCS failure + SBO) with the external water injection into the PCV and it can avoid operation of the containment venting systems in the case of internal water circulation in the PCV.

Instrumentation and control (I&C) systems act as the central nervous system of the Nuclear Power Plant as the erroneous functioning of them can become a reason for an inevitable casualty. The information from these instruments is crucial for plant operators to assess and continuously monitor the plant status. This fact highlights the significance of instrument availability during all possible environmental conditions in a Nuclear Power Plant (NPP). Along with several instruments used in an NPP, transmitters have gained momentous significance in the instrumentation loop as they provide critical information about the plant status. Therefore, the CN0289 pressure transmitter, which has been used extensively in industrial applications, is taken as an object of study for this paper. The instrumentation availability analysis depends on its evaluation under the faulted environment that may be caused by severe accidents. So, the aberrations in the output of the transmitter circuit under harsh environmental conditions have been observed in a well privileged simulating environment MULTISIM. The effect of accidental conditions like elevated temperature, humidity and high pressure have been analyzed on electronic components of the circuit and components most critical for the instrument availability have been identified. Later, a discriminant classifier has been implemented to analyze the instrument condition under different environmental scenarios for its availability. For this purpose, faulty signals are generated in MATLAB by using the statistical formulas of the faults observed in the output of the transmitter circuit. A set of timedomain distinguishing features have been selected to evaluate the instrument usability by the discriminant classifier. Based on these distinguishing features, a quadratic curve separates the faulty signals from the healthy ones indicating the availability of the instrument in a particular scenario. So, this analysis aids in assessing the instrument performance and later a guideline to ameliorate its design for operation under harsh environmental conditions.

Gas–liquid two-phase flow in a stagnant pool is an important phenomenon in designing and operating industrial facilities. When gas is mixed or boiling occurs in stagnant water, the actual water level appears higher than the original water level. The actual water level is called a two-phase mixture level and largely depends on the flow channel geometries, dimensions, and flow conditions. This study focuses on the influence of channel geometries, circular pipes and rod bundles, on the two-phase mixture level and its fluctuation behavior. An air–water experiment using circular pipes with inner diameters of 50 and 224 mm and 5 × 5 and 10 × 10 rod bundles was conducted, and the two-phase mixture level swell was visually observed. As the inlet gas flow rate increased, the two-phase mixture level basically increased regardless of the channel geometry. The fluctuation amplitude was remarkably increased by formulating the slug bubbles covering the entire diameter in the small pipe with a diameter of up to 50 mm. In the rod bundles and large pipe with a diameter of 224 mm, no slug bubble was sustained, and the two-phase water level and its fluctuation amplitude were relatively small compared with those of the small pipe.

Nuclear power plants and chemical plants contain many piping, which degrades with age. As a result of a reduction in structural strength, guillotine breaking and rupturing can occur, leading to severe damage to plants. Piping is subjected to various types of tension and bending, but the influence of the load history on the integrity of piping containing a flaw is poorly understood. The goal of the present study was to develop an improved integrity assessment method for a stainless steel pipe with a part-through notch that takes into account the load history. This involved first determining the feasibility of the leak-before-break (LBB) concept for piping subjected to a combined load and then evaluating the stress at the crack penetration point and estimating the plastic collapse strength. An austenitic stainless steel pipe (SUS 304) with a length of 110 mm, a diameter of 32 mm, and a wall thickness of 3 mm was used as a specimen. A part-through notch with a notch angle of 90° was cut in the center of the specimen by wire electric discharge machining. Loading tests were carried out using statically indeterminate fracture mechanics testing equipment developed by ourselves. This allowed the loading history to be varied because the equipment is capable of applying arbitrary sequences of tension and bending. The LBB concept is considered to be applicable if the crack penetrates before reaching the maximum stress. The plastic collapse strength was then determined using the double elastic slope method. The LBB concept was found to be applicable. In addition, the stress at the crack penetration point was approximately equal to the maximum stress. The results of the present study indicated that the plastic collapse strength of a part-through notched specimen can be safely estimated using the theoretical plastic collapse strength of a through-wall notched pipe.

Since the Fukushima Daiichi nuclear power plant accident, we have been developing a failure evaluation method that considers creep damage mechanisms using detailed three-dimensional finite element analysis model of lower head including penetration, stub tubes, and weld parts, etc., for the early completion of the decommissioning of the nuclear power plants in Fukushima Daiichi. For the finite element analysis, we have been obtaining material properties for which no data are provided in existing databases or in the literature. In particular, creep data corresponding to the high temperature region near the melting point of materials is important in evaluating creep deformation under severe accident conditions. In this study, we obtained the uniaxial tensile and creep properties for low-alloy steel, stainless steel, and Ni-based alloy. In particular, creep test data with long rupture times at high temperatures are expanded using a tensile test machine that can measure the elongation of test specimens in a noncontact measurement system. The parameters related to the failure evaluation were improved on the basis of the expanded creep database.

In this work, we have experimentally studied the effect of particle size of the oxidized zircaloy-4 on the observed yield of Hydrogen generation by Co-60 gamma radiolysis of water. The oxidation products were obtained by high-temperature oxidation of zircaloy-4 under dry air atmosphere for 2 and 4 hours. We crushed them into various sizes from 3 to 484 μm after the oxidation. The crushed particles of the product of 4 hours oxidation consisted of almost monoclinic zirconium oxide and the 2 hours oxidation product contained from 26 to 56% of tetragonal zirconium oxide. These particles had specific surface areas from 0.73 to 2.58 m²/g and band gaps 5.1 eV all. The oxidation products of the weight fraction of 10% and distilled water were mixed into a 4 ml glass vial to examine the Hydrogen generation by Co-60 gamma radiolysis. The experiments have presented that the observed hydrogen yield, G(H₂) was increased by decreasing the median particle diameter of both the oxidation products. Especially, the G(H₂) of water added the particles with 3 or 5 μm became larger than that of only distilled water. This enhancement was not caused by the bandgap of the oxidation product. Calculated G(H₂) per surface area of the oxidation product, G(H₂)/S was constant on all particle sizes. Furthermore, G(H₂)/S of the water added the oxidation products with the median particle diameter 3 or 5 μm were larger than the commercial zirconium oxides (m-ZrO₂, t-ZrO₂) whose median particle diameters are 3 μm. In conclusion, Hydrogen generation of water by Co-60 gamma radiolysis was influenced by the particle size added into water. We have considered that the surface area of the oxidation product was the predominant factor of this effect because the G(H₂) per specific surface area was constant regardless of particle size of the crushed oxidation products in this experiment.

A calcium silicate insulator material was investigated for its interaction with non-radioactive cesium at room temperature by batch-type adsorption experiments. An industrial-grade calcium silicate was weighted and ground into powder with an average size of 15.4 ± 0.1 μm, which was then placed into the polypropylene (PP) tubes containing cesium chloride aqueous solution. Cesium concentrations were range between 1.3×10^-6–3.5×10^-3 M. Adsorption kinetics of cesium was evaluated based on pseudo-first order and pseudo-second order models, while the adsorption isotherm was modeled using the Langmuir model and modified BET model. Physicochemical properties of calcium silicate before and after the adsorption was investigated using XRD, SEM/EDS, and FT-IR. Chemisorption was found as the underlying sorption process between calcium silicate insulator material and cesium.

A large amount of cesium (Cs) chemisorbed onto stainless steel is predicted to be present especially in the upper region of reactor pressure vessel (RPV) during light water reactor severe accident. A chemisorption model was developed for estimation of such amounts of Cs for stainless steel type 304 (SS304) exposed to cesium hydroxide (CsOH) vapor. However, this existing chemisorption model cannot accurately reproduce experimental results and is considered not to be suitable for the estimation of the Cs-chemisorbed amounts under various conditions experienced in Fukushima Dai-ichi nuclear power station. Our recent laboratory study indicated that the surface reaction rate constant used in the exiting chemisorption model depended on both of silicon content in SS304 and concentration of gaseous CsOH as well as on temperature. Therefore, in this study, a modified Cs chemisorption model which accounts for these effects was constructed by combining penetration theory for gas-liquid mass transfer with chemical reaction and mass action law for CsOH decomposition at interface between gaseous and solid phases. As a result, it was found that the modified model was able to adequately describe effects on concentration of CsOH in gaseous phase and Si content in SS304 and more accurately reproduce the experimental data than the existing model.

Most impact research has been presented on the basis of impact tests and numerical analysis performed by rigid projectile impact perpendicular to the target structure. On the other hand, there are only few reports on impacts at an oblique angle. To evaluate more realistic conditions regarding issues related to oblique impacts to reinforced concrete (RC) structures, we have proposed an analytical method to estimate the local damage to RC structures by oblique impact and have validated the evaluation approach by comparison with experimental results. At present, we have finalized simulation analyses of oblique impact assessments on RC panels using rigid/soft projectiles with flat nose shape utilizing the validated approach. In this study, we focus on impacts caused by rigid/soft projectiles with hemispherical nose shape. The same analytical method is applied to simulate the structural damage of the RC panel due to rigid/soft projectile with hemispherical nose shape. Results on the penetration depth of the RC structure and the energy-contribution ratio are presented. By comparing the results of local damage to RC structure caused by projectiles with flat and hemispherical nose shapes, the influence of the nose shape of projectile on local damage of RC panel has been investigated.

Non-destructive examinations (NDEs) have an important role in assurance of the structural integrity of nuclear components including pipe lines. NDEs are performed for welds in piping in accordance with the rules, which prescribe the requirements of NDE such as interval and extent of examination. In general, there are two kinds of sampling method for selecting welds to be examined in each interval considering the specified extent of examination. The first method is the fixed location sampling method, in which welds for NDEs are same in each interval, while the second method is the random location sampling method, in which welds for NDEs are selected from those not examined in the last interval. The selection of the sampling method is important in assuring the structural integrity of piping. Probabilistic fracture mechanics (PFM) analysis, which is one of rational structural integrity assessment methods, can quantitatively calculate failure probability of welds in piping by considering aging degradation mechanisms, such as stress corrosion cracking and fatigue, as well as crack detections and repair of cracked welds through NDE. In some countries, especially the United States, PFM approach has been applied to NDE, such as risk-informed in-service inspection. From such backgrounds, further application of PFM approach is increasingly expected. In this study, to investigate the applicability of PFM approach on NDE, we focused on the difference in sampling method in NDE for piping and calculated the failure probability of a typical nuclear piping considering NDEs based on the two sampling methods through PFM analysis. From the results, the quantitative influence of the difference in sampling method on failure probability was clarified.

Dose estimation for workers and the public was conducted to use the recycled soil removed from the Fukushima prefecture for land reclamation. The land was availed to the public as a greenery-covered area. In the case of revegetation with trees, the absorption of Cs by the trees was considered. The exposure from trees, trimmed or thinned trees, and the organic deposits from litter fall were evaluated. From the results, the concentration levels of radiocesium, ¹³⁴Cs and ¹³⁷Cs, for which the annual effective exposure dose does not exceed 1 mSv/y was calculated. Moreover, the thickness of the cover soil required to maintain the exposure dose lower than 10 μSv/y for the public during servicing was ascertained. Furthermore, disasters were considered; for instance, by considering that tsunamis, fires, and intense heavy rain can increase the exposure doses based on the changes in the conditions of the reclamation land. We confirmed that the additional exposure dose during the disaster and rehabilitation of the area was lower than 1 mSv/y. From these evaluation results, we determined the amenable concentration levels for recycling.

JAEA has been progressing to design HTGR fuels for not only small-type practical HTGRs but also VHTR proposed in GIF which can be utilized for various purposes with high-temperature heat at 750 to 950 °C. To increase economy of these HTGRs, JAEA has been upgrading the design method for the HTGR fuel, which can maintain their integrities at the burnup of three to four times higher than that of the conventional HTTR fuel. Design principles and specifications of various concepts of the high burnup HTGR fuels designed by JAEA are reported. As the latest results on post-irradiation examinations of the high burnup HTGR fuel progressing in a framework of international collaboration with Kazakhstan, irradiation shrinkage rate of the fuel compact as a function of fast neutron fluence was obtained at around 100 GWd/thm. Furthermore, the future R&Ds needed for the high burnup HTGR fuel are described based on these experimental results.

Probabilistic fracture mechanics (PFM) is considered a promising methodology in assessing the integrity of structural components in nuclear power plants because it can rationally represent the influence parameters in their probabilistic distributions without over-conservativeness. In Japan, Japan Atomic Energy Agency has developed a PFM analysis code PASCAL4 (PFM Analysis of Structural Components in Aging LWRs Version 4) which enables the probabilistic integrity assessment of reactor pressure vessels (RPVs) considering neutron irradiation embrittlement and pressurized thermal shock events. Several efforts have been made to verify PASCAL4 to ensure that this code can provide reliable analysis results. In particular, a Japanese working group, which consists of different participants from the industry and from universities and institutes, has been established to conduct the verification studies. This paper summarizes verification activities of the working group in the past two years. Based on those verification activities, the reliability and applicability of PASCAL4 for structural integrity assessments of Japanese RPVs have been confirmed with great confidence.

The demand for high separation efficiency needs an advanced device that can be installed in the separation machinery. Parameters in the separation system, such as feeding flow rate and rotational speed, influence the sedimentation formation. A wireless electrical resistance detector (WERD) was developed with an ability to detect the change of the electrical properties of the suspension inside the centrifugal separation domain. The main focus is to observe the particle sedimentation thickness in the specific positions inside an industrial-scale centrifuge. This wireless instrument has an excellent performance in ultra-high rotation operating speed due to its lightweight and flexibility. WERD transmitted the measured resistance to the processing PC, which then processed the data using the particle thickness formula. The code refers to the resistance strength and normalization method under the boundary of the particle volume fraction in the sedimentation state. Simulation and experimental studies solved the constant of the particle thickness formula. In these studies, the sediment layer was represented by the suspended microsphere acrylic particles with a diameter of 10 μm. The suspension was a mixture of the aqueous Sodium Chloride and the acrylic particles. The real-scale experiment was conducted on industrial centrifuge with a constant relative centrifugal force of RCF = 2,130 G. As a result, the distribution of particle sedimentation thickness during the centrifugation was successfully observed by WERD. The results showed as the feeding rate increases; the particle sedimentation thickness increased up to near the feeding point. Under constant relative centrifugal force, the particle distribution rate under a low feeding flow rate gave a smoother thickening distribution. In the high feeding flow rate, the distribution was thicker and faster especially at the position nearer to the feeding point. These findings of the WERD application is useful for the non-invasive sedimentation monitoring in the separation system.

Two-phase thermosyphons, which use sodium at 960 K as a refrigerant for heat transportation in small nuclear reactors, are promising for manned exploration to Mars. This is because the concentric-tube type thermosyphon may not have a flooding limit, and so the heat transfer performance per unit volume is comparatively large. Moreover, since its external form is a single tube by inserting a pipe through which the liquid flows into a pipe through which the gas flows, the reactor core can be made smaller. Experimental investigations of heat transport characteristics using water, R113, ethanol, and nitrogen as refrigerants and development of prediction formulas are progressing for establishing the design of the concentric-tube two-phase thermosyphon. However, it is necessary to use sodium at 960 K for application to heat transportation of a reactor, and further elucidation of the flow phenomenon is necessary to establish its performance prediction model. We have proposed a model based on bubble pump theory in order to take into consideration the rise of liquid level of the heating section. We conducted an experimental study on the flow inside a thermosyphon made of transparent material, and evaluated the maximum heat transfer rate using a low boiling point refrigerant, HFE-7100. As a result, even when the heat transfer rate was close to the maximum, it was shown that the flow regime in the outer tube of the adiabatic section of the concentric tube is two-phase flow. However, the experimental value of the maximum heat transfer rate was found to be about 17 % smaller than the calculated value. We therefore investigated the cause of this difference by making pressure measurements and flow observations, and found that bubble entrainment in the inner tube of the concentric tube of the adiabatic section greatly influences the pressure distribution. It is thought that taking mixing into consideration would help improve the model.

Elbow pipes present a complex stress field during deformation. The stress field becomes more complicated due to the wall thinning. Thus, predicting their failure behavior is difficult. In previous studies, low-cycle fatigue tests were performed on carbon steel elbow pipes, and a revised universal slope method was proposed to predict their fatigue lives. However, the accuracy of predictions has not been verified on materials other than carbon steel. Therefore, in this study, low-cycle fatigue tests and a finite-element analysis were performed on carbon steel (STPT410) and stainless steel (SUS304) elbows. The low-cycle fatigue life of SUS304 elbows was observed to be three to four times longer than that of STPT410 elbows. Analysis of results revealed that the fatigue life predicted by the revised universal slope method shows a tendency similar to the experimental results, and the fatigue life of SUS304 can be predicted with high accuracy as well as STPT410.

An evaluation methodology of a thermal-hydraulics based on a mechanism in light water reactors (LWRs) is needed from a viewpoint of the safety analysis during normal operation and unanticipated transient such as under a severe accident. Currently, the evaluation of safety for the nuclear reactor has been implemented by a best estimate (BE) code and subchannel analysis code. These analysis codes contain models and empirical correlations. Therefore, the full-scale mock-up tests are needed to evaluate the reliability and validation of code. And the model and empirical correlation are allowed to be applied only in the range where the experiments were implemented. The large mock-up tests are once again needed in order to consider the new geometry and boundary conditions when the design of components is changed. Hence, the 3D detailed numerical simulation by the mechanistically based method is expected to be applied for the preliminary analysis to improve the design of fuel assemblies and evaluate the safety. This 3D detailed numerical simulation can reduce the large mock-up tests. The detailed numerical simulation method can provide much information relating to the two-phase flow such as the bubble size, its velocity, and detailed void distribution which, for example, are needed to predict the critical heat flux based on the mechanism. Moreover, JAEA is implementing the development of the nuclear-thermal-coupling code by using a detailed two-phase flow analysis code based on the VOF method like a JUPITER code. In this study, the numerical simulation of two-phase flow in the 4x4 bundle was examined by numerical simulation code JUPITER in order to examine the possibility of the JUPITER code for the large scale two-phase flow analysis. The simulation results are verified by the previous experimental data of two-phase flow.

Regarding a severe accident of supercritical water-cooled reactor (SCWR), phase change between subcritical and supercritical conditions is crucial since heat transfer rate changes massively causing a dryout accident. Fundamental knowledge on surface wettability and boiling heat transfer on metals at subcritical conditions under radiation are, thus, important in thermal-hydraulic design and safety analysis of reactor core in light water reactors including a supercritical water-cooled reactor. The radiation induced surface activation (RISA) which enhances wettability and anticorrosive effect on the metal surface was first revealed by authors in 1999. In the earlier studies, significant improvements of surface wettability and boiling heat transfer on oxide film coated-materials by the RISA were observed in a room temperature condition. The purpose of this study is to evaluate the effect of oxidized metal and γ-ray irradiation on metal surface wettability in high pressure and high temperature conditions. In this experiment, the test section was pressurized at 12 MPa with nitrogen gas using pressure vessel and was heated up to temperatures of 20, 150, 200, 250 and 290 centigrade. Two types of material; a stainless-304 and austenitic stainless steel named PNC1520, which is considered as a potential material of fuel-cladding tube of the SCWR, were used as specimens. The oxide film on the specimen was formed in supercritical water at 380 centigrade and 22 MPa. About 600 kGy Co-60 γ-ray source was used for irradiation. The results showed that the difference of oxidization on wettability was insignificant at room temperature before γ-ray irradiation while contact angles on the oxidized specimen decreased at high temperatures. The water growth rate on oxidized material slightly lower compare to non-oxidized material. This result suggests oxide film formation on metal surface plays an important role in surface wettability enhancement by the RISA.

Within the ALISA (Access to Large Infrastructures for Severe Accidents) cooperation between the European Union and China on SA (Severe Accident) facilities, CNPRI (China Nuclear Power technology Research Institute) has proposed to CEA (Commissariat à l’Energie Atomique et aux Energies Alternatives) a series of experiments to determine the density and surface tension of different oxidic mixtures representative from both in-vessel and ex-vessel retention scenarios. This article describes those tests performed in the VITI facility in in CEA severe accident experimental platform, PLINIUS, by means of the Maximum Bubble Pressure (MBP) method. The measurement procedure using the Maximum Bubble Pressure method is described. A first estimate of in-vessel corium surface tension is given, while both density and surface tension measurements are successfully performed for both ex-vessel corium compositions, which constitutes to our best knowledge the first measurements ever performed for ex-vessel (MCCI) liquid corium compositions.

Sintered materials are superior in productivity because of their simple process, but they are not applied to high-load gears because of their insufficient strength. To improve the fatigue strength of sintered materials, the authors have developed liquid-phase sintering which can achieve high-density without using secondary processing. In this study, the effect of boron addition (0-0.4 mass%) on the rolling contact fatigue strength of Fe-Ni-Mo-B-C sintered and carburized material was evaluated. In addition, in order to evaluate only the boron addition effect excluding the influence of density, the sintered density of each specimen was controlled to be the same. In the test range of this study, the rolling contact fatigue limit (p_max)_lim of material with an additional quantity of boron of 0.1 mass% showed the highest value exceeding 1700 MPa. This value was not only significantly higher than the (p_max)_lim of the boron-free material (1100 MPa), but also an extremely high value comparable to the (p_max)_lim of wrought steel (1900 MPa). The reason why the (p_max)_lim of a 0.1B roller was remarkably high was investigated from the viewpoints of both pore structure and material structure. As for the pore structure, the pore shape of the boron-free roller was irregular, whereas the pore shape of the 0.1B roller was spherical. As a result of CAE analysis of the orthogonal shear stress inside the roller during the rolling contact fatigue test, it was found that the maximum value of orthogonal shear stress around the pores of the 0.1B roller was about 35 % lower than that of the boron-free roller. This result suggests that cracks are less likely to occur in the 0.1B roller than in the boron-free roller. In other words, it is thought that the pore shape of 0.1B material affects the improvement of rolling contact fatigue strength.

Three bump sizes, six pressures and five placement angles of bump were used in this paper to conduct enveloping stiffness experiments. The experimental tire displacement-load curves showed that the enveloping stiffness of the tire under the same pressure did not change with changes of the bump size. And the enveloping stiffness of the tire were changed with changes of pressures and placement angles; Models consider pressure and placement angle of bump were established respectively, and established a model consider pressure and placement angle simultaneously by least squares, those models can be used to estimate enveloping stiffness force without experiments. Abaqus finite element software simulations showed that the stress distribution in the tread is H-shaped and the deformation distribution in the tread is oval; By analyzing the force distribution of the ply found that the distribution of force changes with the change of placement angles. The lateral displacement(U1) of the belts increases linearly with increasing angle. The end of cord is prone to failure at 0° and 45°, and the cord near middle of ply is also prone to failure at all angles.

A high-pressure field is generated in a circular tube by introducing an unsteady jet from its open end. The head of this jet acts as a piston, driving compression waves ahead of it. The peak value of the induced overpressure is evaluated as a solution of a Riemann problem, wherein the jet head is equivalent to a piston head. The jet head of the driver gas, with a filling pressure of 400 kPa, is equivalent to a piston head moving at 160 m/s. This high-pressure generation scheme through the “piston effect” is useful for industrial applications, including filter cleaning in dust collectors, and as an interesting example of unsteady, compressible fluid dynamics.

In the present study, the authors investigate the simplest model of cross-flow impellers, in order to provide the fundamental flow features of cross-flow impellers. The model is a two-dimensional cross-flow impeller with a very-high aspect ratio, and rotates in stationary fluid. More strictly, we occasionally place a splitter plate with minimal influence instead of casings for more precise observations and measurements. Instantaneous measurements of flow-velocity vectors on the mid-span plane are conducted using a CCD camera and a double-pulse YAG laser with a particle-image-velocimetry technique, in addition to conventional flow visualisations and hot-wire-velocimetry measurements. As a result, the authors show the revolutional motion of such an eccentric vortex as a large-scale re-circulating flow structure inside the cross-flow impeller, and reveal the detail of flow features at Reynolds number Re = 300-2500. Furthermore, the authors show the relations between Re and out-flow rate from the impeller.

In recent years, master–slave systems with pneumatic actuators, which are necessary to realize a safe and secure society, have attracted increasing attention and are currently being used in the medical and welfare fields. In addition, they are used in disaster-recovery support through application to remote-controlled robots. The master–slave system proposed in this study aims at improving safety and reliability by controlling a pneumatic actuator with mechanical feedback. A characteristic analysis is essential to improve the performance and expand the possibilities for incorporation into practical systems. The purpose of this study is to clarify the performance of a master–slave system comprising only mechanical components and consider possible applications. We have developed a pneumatic master–slave system that uses a one-axis stage and a precision pressure regulator as the master and a pneumatic positioning device as the slave and performed characteristic confirmation experiments. First, we modeled the pneumatic positioning device and compared it with the actual step response. Next, we made a prototype of the master–slave system and conducted characteristic confirmation experiments such as step response and response to repeated inputs. Our experimental results suggest that this system is suitable for applications that require relatively low speeds and positioning accuracies of several millimeters, such as nursing-care assisting systems, transportation-support systems at production sites, and robots with expanded human abilities.

This paper addresses the compensation of the vibration caused by flow disturbance, which is the pressure fluctuation of compressed air supplied to a pneumatic anti-vibration apparatus (AVA). The pressure of the compressed air changes periodically during compression process. For this reason, proportional-integral-sinusoidal (PIS) control is utilized so that the flow disturbance can be attenuated. In this control method, a Sinusoidal compensator (S compensator) provides a periodic control input signal whose frequency is the same as the frequency of the disturbance. However, when the PIS control is applied to the vibration attenuation for the pneumatic AVA, there arise two practical problems that have to be solved. First, the vibration of an isolated table occurs in transient state at the start-up/shut-down of the S compensator. This is due to the switching of the S compensator. To deal with the first practical problem, soft switching approach is presented. Second, high-frequency vibration is observed in steady state after the S compensator is started up. Since the source of the high-frequency vibration is the phase-lag of the S compensator, a phase-lead type PIS compensator is employed. By using this compensator, the phase margin of a control system can be increased and the high-frequency vibration can be attenuated. Moreover, in this paper, vibration transmissibility is analyzed so as to investigate effects of the PIS control on the isolation from floor vibration. It is shown that 1) when the PIS control is used, anti-resonance and resonance are excited and 2) there is a trade-off between flow disturbance attenuation and vibration transmissibility reduction.

The present paper proposes a model of muscle fatigue and recovery considering both the behavior of slow- and fast-twitch muscle fibers and the role of antagonistic muscles. The proposed model can be used to predict the degree of muscle fatigue of each muscle in the upper arm when the muscle activation pattern of the muscles changes because of the variation of the output force direction at the distal extremity. Furthermore, it can predict the variation in the muscle fatigue not only under maximum voluntary construction but also under any constant applied force or alternating periods of constant output force and rest. To validate the proposed model, case studies were conducted. The parameters necessary to predict the degree of fatigue of each muscle can be determined from a few preliminary experiments in which a participant outputs force in only four directions. The prediction results from the case studies showed good agreement with the measurement results. Therefore, the proposed model can be applied in cases where the output force direction changes with any force magnitude or alternates with periods rest, which is not the case for existing models.