bluuebunny/crossref_metadata_2025 huggingface dataset:
Development of a Simplified Model for Aerosol Removal by Spray System of Containment
Authors: Huiyu Yu , Haifeng Gu , Junyan Chen , Zhongning Sun
Published: October 2020

The containment spray system is of great importance for the nuclear power plant to reduce the pressure and temperature, as well as concentration of radioactive aerosols released from reactor core under the severe accident. The main aim of present study is to develop a simplified model for aerosol removal by spray system, the model is employed to predict the variation of aerosol concentration in containment and reveal the key mechanism influencing aerosol removal. In the present model, the air flow entrained by spraying droplet is considered which is computed by solving simplified one-dimension momentum equations rather than the standard N-S equation. For the validation of the model, a small size sprinkle experiment facility is built. The facility is equipped with non-intrusive instruments such as particle size spectrometer for aerosol spectrum measurement, HELOS/R for measuring the droplet spectrum. The result of aerosol removal of the present model is compared with the result of the full coverage model and the experimental result. In the experiment, polydisperse aerosol is used and the removal rate of aerosols with different sizes is compared against the result of the present model and the full coverage model. The velocity of entrained gas flow and the distribution of droplets are displayed. The computational result reveals the aerosol removal constant distribution along the height.
Study on Heating Process of Dehumidifying Experiment in HTGR
Authors: Shen Kaiyue , Zheng Wei , Ma Shengchao , Yin Huaqiang , He Xuedong , Ma Tao
Published: October 2020

A large number of carbon materials are used in high temperature gas-cooled reactor (HTGR). As a kind of porous material, the carbon material contains a certain amount of moisture and other impurities. In order to reduce the corrosion of internal material in reactor core of HTGR, the initial core or post-accident core must be strictly heated and dehumidified. The current primary circuit heating mainly relies on the rotation of the primary pump to convert the kinetic energy into thermal energy. Obviously, the current scheme was flawed: (1) Due to the insufficient heat generated by rotation of the primary pump, the temperature rising process of the primary circuit is sluggish; (2) The rotation of the primary pump converts the kinetic energy into thermal energy of the helium, at the meantime, the primary circuit dissipates heat outward. For the above reasons, it is difficult to achieve the desired dehumidification temperature in the heating process. While in this paper, an additional thermal source will be added to the steam generator to heat the primary circuit in a new scheme. A proper flow and heat-transfer model of heating the primary circuit in high-temperature reactor was established based on software COMSOL Multiphysics. The numerical analysis of the primary circuit heating process provides rewarding guidance for the selection of the dehumidification scheme in HTGR.
Experimental Study on Gravity Driven Discharging of Quasi-Two-Dimensional Pebble Bed Based on Mathematical Morphology
Authors: Liu Yujia , Peng Sifan , Gui Nan , Yang Xingtuan , Tu Jiyuan , Jiang Shengyao
Published: October 2020

The pebbles flow is a fundamental issue for both academic investigation and engineering application in reactor core design and safety analysis. In general, experimental methods including spiral X-ray tomography and refractive index matched scanning technique (RIMS) are applied to obtain the identification of particles’ positions within a three-dimensional pebble bed. However, none of the above methods can perform global bed particles’ position identification in a dynamically discharging pebble bed, and the corresponding experimental equipment is difficult to access due to the complication and high expense.

In this research, the experimental study is conducted to observe the gravity driven discharging process in the quasi two-dimensional silos by making use of the high-speed camera and the uniform backlight. A mathematical morphology-based method is applied to the pre-processing of the captured results. After being increased the gray value gradient by the threshold segmentation, the edges of the particles are identified and smoothed by the Sobel algorithm and the morphological opening operation. The particle centroid coordinates are identified according to the Hough circle transformation of the edges. For the whole pebble bed, the self-programmed process has a particle recognition accuracy of more than 99% and a particle centroid position deviation of less than 3%, which can accurately obtain the physical positions of all particles in the entire dynamically discharge process. By analyzing the position evolution of individual particles in consecutive images, velocity field and motion events of particles are observed. The discharging profiles of 5 conditions with different exit are analyzed in this experiment. The results make a contribution to improving the understanding of the mechanism of pebbles flow in nuclear engineering.
Evaluation of Decay Dose Rates for European DEMO Vacuum Vessel Structure
Authors: Breidokaitė Simona , Tidikas Andrius , Stankūnas Gediminas
Published: October 2020

Four types of breeder blanket modules are in consideration for European DEMO pre-conceptual design phase specified by European Fusion Roadmap. In this study two designs (Helium Cooled Pebble Bed (HCPB) and Water-Cooled Lithium Lead (WCLL)) were investigated in terms of their influence on vacuum vessel (VV) nuclear decay dose rates. In both cases, the same vacuum vessel structure is examined. VV model consists of two shells, interspace and thermal shield. The inner shell and outer shell of vacuum vessel are made of Stainless Steel 316L(N)-IG. The interspace layer is made of water for cooling and shielding purposes. Stainless steel constitutes around 60% of all material in vacuum vessel and water makes up the rest. The model of fusion reactor has been developed in the frame of EUROfusion’s Power Plant Physics and Technology (PPPT) programme. This paper presents information about calculation results of the dose rates caused by the neutron irradiation for the vacuum vessel structural elements in DEMO fusion power reactor.

In order to perform dose rate calculations, neutron energy spectra are required for different VV segments. Those calculations were performed with MCNP6 code and Joint Evaluated Fission and Fusion (JEFF) 3.2 nuclear data library. Dose rates were calculated with FISPACT-2010 and European Activation File (EAF) 2010 nuclear data library. The dose rates were calculated for European DEMO nuclear fusion reaction after its 20 calendar years lasting operation. Cooling period was considered spanning for up to 1000 years. In addition, radionuclides with contribution of at least 1% to the total value of dose rate were identified.

Highest dose rates are attributed to 56Mn radionuclide immediately after the operation until around one day.60Co is the most dominant isotope during intermediate and long cooling periods. Differences in Interspace layer occur due to 16N isotope.
Experimental Study on Flow Patterns and Pressure Drop of Decaying Swirling Gas-Liquid Flow in a Vertical Pipe
Authors: Zhang Jiarong , Liu Li , Liu Shuai , Gu Hanyang
Published: October 2020

Vertical swirling gas-liquid flow is a kind of complex two-phase flow containing a nonzero tangential velocity component in engineering applications. The accurate flow regime characterization, phase distribution information and pressure drop data about vertical swirling flow are the basis for the optimization of steam generator (SG), which can greatly reduce the cost and improve the safety of nuclear plants. To get these key parameters of swirling vertical flow, we have made a comprehensive visualization experiment in a vertical pipe with 30mm diameter and 5m length by high-speed camera. The experimental pipe is separated into swirling part and non-swirling part. We have set three observation section with different vertical heights in the swirling part. Changing the flow rate of water and gas, different swirling flow pattern photos can be captured by high-speed camera. Based on the photos of different positions and image-processing MATLAB code, we can get three flow regime maps and figure out the decaying law of swirling gas-liquid flow. The pressure drop can be recorded by rotameter at each position. The decaying law of pressrure drop can be concluded from it. These data can be a guide for designing gas-liquid separator in SG to improve the efficiency of nuclear plant.
Seismic Time History Data Precision and Time Interval Requirement
Authors: Li Dali
Published: October 2020

This paper provides the seismic time history data precision and time interval requirement for seismic dynamic analysis. U.S.NRC SRP 3.7.1 “Seismic Design Parameters” Acceptance Criteria for Design Time Histories specifies the power spectral density Nyquist Frequency, time interval, and total duration; however, it does not have the requirement for Response Spectra. The response spectrum bandwidth is inverse-proportional to time interval of the time history. For the time interval of 0.005 seconds, the bandwidth for the response spectrum is between 0.194 Hz and 80.5 Hz; the PSD Nyquist frequency is 100 Hz. For 20.48 seconds time history, 4096 data points are required. The response spectrum between 1.28 Hz and 13.6 Hz has the peak flat magnitude value; the magnitude drops to 0.707 of the peak value from 1.28 Hz to 0.194 Hz and from 13.6 Hz to 80.5 Hz. This paper also provides the time interval requirement for various response spectrum peak flat magnitude value; i.e., the response spectrum highest flat magnitude of 27.2 Hz requires a time interval of 0.0025 seconds time history. For 20.48 seconds time history, 8192 data points are required. For CSDRS, the time interval of 0.005 seconds is adequate for the frequency range of interest between 0.36 Hz and 57.2 Hz. For HRHF, the time interval of 0.0025 seconds is required to analyze the frequency range of interest between 0.36 Hz and 114.4 Hz.
Analysis of Water Retention in Isothermal Vacuum Drying Test
Authors: Salazar Alex , Pulido Ramon J. M. , Lindgren Eric R. , Durbin Samuel G.
Published: October 2020

Validation of the extent of water removal in a dry storage system using an industrial vacuum drying procedure is needed. Water remaining in casks upon completion of vacuum drying can lead to cladding corrosion, embrittlement, and breaching, as well as fuel degradation. In order to address the lack of time-dependent industrial drying data, this study employs a vacuum drying procedure to evaluate the efficiency of water removal over time in a scaled system. Isothermal conditions are imposed to generate baseline pressure and moisture data for comparison to future tests under heated conditions. A pressure vessel was constructed to allow for the emplacement of controlled quantities of water and connections to a pumping system and instrumentation. Measurements of pressure and moisture content were obtained over time during sequential vacuum hold points, where the vacuum flow rate was throttled to draw pressures from 100 torr down to 0.7 torr. The pressure rebound, dew point, and water content were observed to eventually diminish with increasingly lower hold points, indicating a reduction in retained water.
Experimental Investigations on Thermal Stratification in a Large Pool of Water With Immersed Isolation Condenser
Authors: Kumar Sunil , Vijayan P. K. , Grover R. B. , Kannan U. , Yadav Harekrishna , Agrawal A.
Published: October 2020

Advanced nuclear reactors use large pool of water inventory with an immersed heat exchanger to remove decay heat especially in the case of Station Black Out (SBO). The isolation condenser (IC) immersed in Gravity Driven Water Pool (GDWP) of Advanced Heavy Water Reactor (AHWR) is an example of such systems. Heat rejected by the IC is absorbed in the pool. As a result, water density decreases and moves towards the free surface of pool causing layers of hot water piling up over colder one giving rise to stratified water inventory. consequently, the pool at the free surface starts boiling before the grace period (7 days). In the present paper, thermal stratification has been modeled in a power to volume scaled experimental setup. The study is focused on investigating the effect of heater orientation on suppression of thermal stratification in the pool for both the cases of with and without shrouds around heat exchanger.
Comparison of the Behavior of Different Jet Configurations in a Crossflow
Authors: Lewandowski Michael , Kristo Paul , Weiss Abdullah , Kimber Mark
Published: October 2020

In this study, the well understood single round jet in crossflow is compared to an array of three jets aligned on the central axis of the crossflow, all of which are distinct with independent temperature and flow control. The benefit of this study is to better understand how an array of jets mix with a crossflow in turbulent conditions in an effort to further validate computation models for a variety of multiple-jet applications. The test section involves jets issuing vertically downward into a cross flow provided by a low speed wind tunnel. The jets are spaced two diameters apart, providing for an upstream, middle, and downstream jet. The hydraulic interactions are tracked via a stereoscopic particle image velocimetry (S-PIV) system. Several reference measurements and the uncertainty of the results are discussed to aid analogous computational fluid dynamics (CFD) models in the future. First order statistics between the single and triple jet cases with fixed flow rates are compared. Temporal analysis yields dominant frequencies at distinct regions within the flow. Insights into flow control of multiple jets in a cross flow is discussed. The experimental data sets compare the classical single jet in cross flow to the extension of multiple jets. Several differences in the flow behavior are found. The data sets serve as benchmark cases for future CFD models that will aim to replicate these flow types in different real-world applications such as coolant flows in a gas cooled reactor.
Application of AHP - Fuzzy Evaluation Method in Nuclear Power Plant Site Selection
Authors: Sun Yujie , Chen Changzhi
Published: October 2020

In order to make a comprehensive evaluation of many influencing factors on the site selection of nuclear power plants, and to make a quantitative judgment on the rank of candidate sites, a method of site selection based on analytic hierarchy process (AHP) and fuzzy evaluation is proposed. Firstly, based on the characteristics of design and research institute, the factors influencing the site selection of nuclear power plants are analyzed. The indicator of criterion layer and index layer influencing the site selection of nuclear power plant are determined based on specialty, and the decision-making system of site selection scheme is established. Secondly, based on AHP operation rules, the weights of system indicator are calculated. Thirdly, by determining the evaluation set, the evaluation matrix based on fuzzy theory was established, and the evaluation level of candidate sites was determined based on the principle of maximum membership degree. Finally, a practical nuclear power plant site selection project is used for theoretical verification. The candidate sites of nuclear power plants are ranked by the first and second level fuzzy comprehensive evaluation. The results show that the quantitative ranking of site based on AHP and fuzzy evaluation is consistent with the previous qualitative ranking of the project. The site selection method based on AHP -fuzzy evaluation is efficient.
Hybrid Nodal Integral/Finite Element Method (NI-FEM) for Time-Dependent Convection Diffusion Equation
Authors: Namala Sundar , Uddin Rizwan
Published: October 2020

Nodal integral methods (NIM) are a class of efficient coarse mesh method that use transverse averaging to reduce the governing partial differential equation(s) (PDE) into a set of ordinary differential equations (ODE), and these ODEs or their approximations are analytically solved. Since this method depends on transverse averaging, the standard application of this approach gets restricted to domains that have boundaries that are parallel to one of the coordinate axes (2D) or coordinate planes (3D). The hybrid nodal-integral/finite-element method (NI-FEM) has been developed to extend the application of NIM to arbitrary domains. NI-FEM is based on the idea that the interior region and the regions with boundaries parallel to the coordinate axes (2D) or coordinate planes (3D) can be solved using NIM and the rest of the domain can be solved using FEM. The crux of the hybrid NI-FEM is in developing interfacial conditions at the common interfaces between the regions solved by the NIM and the FEM. Since the discrete variables in the two numerical approaches are different, this requires special treatment of the discrete quantities on the interface between the two different types of discretized elements. We here report the development of hybrid NI-FEM in a parallel framework in Fortran using PETSc for the time-dependent convection-diffusion equation (CDE) in arbitrary domains. Numerical solutions are compared with exact solutions, and the scheme is shown to be second order accurate in both space and time. The order of approximations used for the development of the scheme are also shown to be second order. The hybrid method is efficient compared to standalone conventional numerical schemes like FEM.
Heat Transfer and Fluid Flow Characteristic of One Side Heated Vertical Rectangular Channel That Inserted Thin Metallic Wire
Authors: Suga Gota , Takeda Tetsuaki
Published: October 2020

A Very High Temperature Reactor (VHTR) is one of the next generation nuclear systems. From a view point of safety characteristics, a passive cooling system should be designed as the best way of a reactor vessel cooling system (VCS) in the VHTR. Therefore, the gas cooling system with natural circulation is considered as a candidate for the VCS of the VHTR. Japan Atomic Energy Agency (JAEA) is advancing the technology development of the VHTR and is now pursuing design and development of commercial systems such as the 300MWe gas turbine high temperature reactor GTHTR300C (Gas Turbine High Temperature Reactor 300 for Cogeneration). In the VCS of the GTHTR300C, many rectangular flow channels are formed around the reactor pressure vessel (RPV), and a cooling panel utilizing natural convection of air has been proposed. In order to apply the proposed panel to the VCS of the GTHTR300C, it is necessary to clarify the heat transfer and flow characteristics of the proposed channel in the cooling panel. Thus, we carried out an experiment to investigate heat transfer and fluid flow characteristics by natural convection in a vertical rectangular channel heated on one side. Experiments were also carried out to investigate the heat transfer and fluid flow characteristics by natural convection when a porous material with high porosity is inserted into the channel. An experimental apparatus is a vertical rectangular flow channel with a square cross section in which one surface is heated by a rubber heater. Dimensions of the experimental apparatus is 600 mm in height and 50 mm on one side of the square cross section. Air was used as a working fluid and fine copper wire (diameter: 0.5 mm) was used as a porous material. The temperature of the wall surface and gas in the channel were measured by K type thermocouples. We measured the outlet flow rate by hot-wire anemometer which is an omnidirectional spherical probe of diameter 2.5mm. The experiment has been carried out under the condition that a copper wire with a scourer model and a cubic lattice model were inserting into the channel.
Development of Three-Dimensional Distribution Visualization Technology for Boron Using Energy Resolved Neutron-Imaging System (RADEN)
Authors: Abe Yuta , Tsuchikawa Yusuke , Kai Tetsuya , Matsumoto Yoshihiro , Parker Joseph D. , Shinohara Takenao , Ohishi Yuji , Kamiyama Takashi , Nagae Yuji , Sato Ikken
Published: October 2020

Boron carbide is used as a neutron-absorbing material in Fukushima-Daiichi Nuclear Power Station (1F), producing borides that are twice as hard as oxides (such as UO2 and ZrO2). The high neutron absorption of boron affects the evaluation of re-criticality during the process of debris retrieval. Therefore, it is important not only to determine the presence of boron but also to investigate the distribution of boron inside the material in a non-destructive manner during decommissioning. To address the uncertainties in the core material relocation behavior of boiling water reactor (BWR) during a severe accident (SA), solidified melt specimens of a simulated fuel assembly were prepared by plasma heating. If core material melting and relocation (CMMR) specimens can be used to estimate the B distribution in 1F Unit-3, that will provide valuable information in the decommissioning of 1F. To address this, the authors focused on the energy-resolved neutron-imaging system, RADEN, which utilizes a wide energy range, from meV to keV. This is an innovative three-dimensional analysis technology for boride distribution that affects the evaluation of hardness and re-criticality. In the calibration standard samples (ZrxB1-x and FexB1-x), there was a good correlation between boron concentration and the energy-dependence of the cross sections of cold and epithermal neutrons. In the CMMR specimens, boron distribution was confirmed from the contrast difference between cold and epithermal neutrons. In the future, the results of calibration standard samples will be applied to the results of CMMR specimens. With this method, three-dimensional boron distribution will be measured, and the understanding of boride distribution 1F Unit-3 will be improved, which may be reflected in an improved SA code.
Analysis of Reactivity Accidents in Modular HTGRs
Authors: Gorton Jacob P. , Brown Nicholas R.
Published: October 2020

A control rod withdrawal (CRW) is a possible reactivity initiated accident (RIA) in modular high-temperature gas-cooled reactors (mHTGRs). The purpose of this study is to perform a sensitivity analysis of a CRW event in an mHTGR model using the systems code RELAP5-3D with point kinetics feedback to demonstrate the impact of uncertainty in heat transfer and reactor kinetic parameters. The adaptive Sobol decomposition method in the uncertainty quantification code RAVEN is used to perform the sensitivity study and to determine which input parameters have the greatest impact on the figures of merit, which in this case are peak reactor power and maximum fuel temperature. This study addresses a need highlighted by the Nuclear Regulatory Commission (NRC) for transient fuel testing by quantifying the impact of uncertainty in heat transfer and reactor kinetic parameters and by generating potential boundary conditions for transient testing of conventional mHTGR fuel.
Proposal of Inspection Rationalization Method and Application for Sodium Cooled Fast Reactor
Authors: Hiroki Yada , Shigeru Takaya , Yasuhiro Enuma
Published: October 2020

In order to rationalize maintenance for nuclear power plants, it is necessary to develop optimize maintenance plan by considering characteristics of each plant. In sodium-cooled fast reactor, there are constraints on inspections due to the specialty of handling sodium equipment, that is one of the important points when considering rationalization of maintenance. To solve this problem, we proposed a basic concept of maintenance optimization scheme that is a design support tool in order to develop maintenance strategy, based on “system based code (SBC)”. SBC is a concept to optimize the reliability of a nuclear power plant by consideration of all related technical requirements. “ASME Code Case N-875” and “ASME Boiler and Pressure Vessel Code, Section XI, Division 2 (RIM)” based on system based code were already developed as standards for inspection. One of the proposed scheme goals is to make a concrete way of necessary assessment method. Another is to provide several combinations of design and maintenance, and information for owner in order to choose the acceptable combination. In the beginning, we are working to develop the scheme that can be applied to sodium fast reactor as the main concept of next generation reactor. In this context, primary heat transfer system (PHTS) piping of fast reactor was evaluated by the scheme. This piping was chosen because it is major significant component and the inspection have constraint conditions that need preparation work. As a result, design candidate (e.g. single and double wall piping) and inspection candidate (e.g. ultrasonic testing and continues leakage monitoring) combinations along with benefit of each cases were provided.
Determining Airborne Release Fraction From DOT 7A Drums Exposed to a Thermal Insult
Authors: Mendoza Hector , Figueroa Victor G. , Gill Walter , Sanborn Scott
Published: October 2020

Fire suppression systems for transuranic (TRU) waste facilities are designed to minimize radioactive material release to the public and to facility employees in the event of a fire. Currently, facilities with Department of Transportation (DOT) 7A drums filled with TRU waste follow guidelines that assume a fraction of the drums experience lid ejection in case of a fire. This lid loss is assumed to result in significant TRU waste material from the drum experiencing an unconfined burn during the fire, and fire suppression systems are thus designed to respond and mitigate potential radioactive material release. However, recent preliminary tests where the standard lid filters of 7A drums were replaced with a UT-9424S filter suggest that the drums could retain their lid if equipped with this filter. The retention of the drum lid could thus result in a very different airborne release fraction (ARF) of a 7A drum’s contents when exposed to a pool fire than what is assumed in current safety basis documents. This potentially different ARF is currently unknown because, while studies have been performed in the past to quantify ARF for 7A drums in a fire, no comprehensive measurements have been performed for drums equipped with a UT-9424S filter. If the ARF is lower than what is currently assumed, it could change the way TRU waste facilities operate. Sandia National Laboratories has thus developed a set of tests and techniques to help determine an ARF value for 7A drums filled with TRU waste and equipped with a UT-9424S filter when exposed to the hypothetical accident conditions (HAC) of a 30-minute hydrocarbon pool fire. In this multi-phase test series, SNL has accomplished the following: (1) performed a thermogravimetric analysis (TGA) on various combustible materials typically found in 7A drums in order to identify a conservative load for 7A drums in a pool fire; (2) performed a 30-minute pool fire test to (a) determine if lid ejection is possible under extreme conditions despite the UT-9424S filter, and (b) to measure key parameters in order to replicate the fire environment using a radiant heat setup; and (3) designed a radiant heat setup to demonstrate capability of reproducing the fire environment with a system that would facilitate measurements of ARF. This manuscript thus discusses the techniques, approach, and unique capabilities SNL has developed to help determine an ARF value for DOT 7A drums exposed to a 30-minute fully engulfing pool fire while equipped with a UT-9424S filter on the drum lid.
Risk-Informed Analysis for Accident Tolerant Fuel in Pressurized Water Reactors
Authors: Ma Zhegang , Parisi Carlo , Davis Cliff , Zhang Sai , Zhang Hongbin
Published: October 2020

This paper presents the research activities performed by Idaho National Laboratory (INL) for the Department of Energy (DOE) Light Water Reactor Sustainability (LWRS) Program, Risk-Informed System Analysis (RISA) Pathway, Enhanced Resilient Plant (ERP) Systems research, using the probabilistic risk assessment (PRA) tool SAPHIRE and the deterministic best estimate tool RELAP5-3D for risk-informed analysis. The ERP research supports DOE and industry initiatives by developing Accident Tolerant Fuel (ATF), the Diverse and Flexible Coping Strategy (FLEX), and passive cooling system designs to enhance existing reactors’ safety features (both active and passive) and to substantially reduce operating costs of nuclear power plants (NPPs) through risk-informed approaches to analyze the plant enhancements and their characterization.

The risk-informed analysis used SAPHIRE and RELAP5-3D to evaluate the risk impacts from near-term ATF (FeCrAl and Chromium-coated clads) on a generic Westinghouse three-loop pressurized water reactor (PWR) under the following accident scenarios: station blackout (SBO), loss of feedwater (LOFW), steam generator tube rupture (SGTR), loss-of-coolant accidents (LOCAs), locked rotor transient, turbine trip transient, anticipated transient without scram (ATWS), and main steam line break (MSLB). The RELAP5-3D simulations included the time to core damage, time to 0.5 kilograms hydrogen generation, and total hydrogen generation. The simulation results show there are modest gains of coping time (delay of time to core damage) due to efficacy of the near-term ATF designs in various accident scenarios. The risk benefits on behalf of the core damage frequency (CDF) brought by the ATF designs would be small for most of the scenarios. However, results revealing much less hydrogen being produced at the time of core damage show a clear benefit in adopting ATFs.
Experimental and Numerical Investigation of Thermal Performance of Synthetic Jet Impingement
Authors: Singh Pushpanjay K. , Kothari Rohit , Sahu Santosh K. , Upadhyay Prabhat K. , Singh Shashwat
Published: October 2020

Synthetic jet potentially useful in electronics cooling is investigated both numerically and experimentally. In the present study, a confined three dimensional synthetic jet with sinusoidal moving wall is considered. Computations are carried out using the FLUENT software with the coupled user defined function describing the diaphragm movement. In this study the effect of various geometrical parameters influencing the flow field and heat transfer are investigated. The effects of change in orifice geometry (circular, square and rectangular), orifice aspect ratio, and jet-to-plate distance are studied for a given hydraulic diameter. The heat transfer results obtained from the synthetic jet is compared with the continuous jet. An electromagnetic actuator is used as an oscillating diaphragm for the generation of synthetic jet. A stainless steel foil with 0.05 mm thickness is used as the test specimen. The surface temperature of the test specimen is measured by using a thermal imaging technique during synthetic jet impingement and a constant temperature anemometer has been employed for velocity measurement. Tests are carried out for Reynolds number of 5448, varied range of jet-to-plate distance (1–18). The maximum value of the heat transfer coefficient is found to be 16 times more than the heat transfer coefficient for natural convection.
Research on Hydraulic Model Test of Pumping Station Forebay
Authors: Jian Jiale , Wang Fang , Zhang Rongyong , Tang Benjing
Published: October 2020

A physical model with a scale of 1:20 was used to study the hydraulic characteristics of the flow in the forebay area of the pump station in Tian Wan Nuclear Power Plant Project Unit 5&6. The original layout scheme of the pump station forebay is limited by space, the turning radius of the intake gallery is small and the straight line section is short, the uniformity of water distribution is poor after seawater enters the forebay, the flow in the forebay cannot fully diffuse, the mainstream is concentrated, the flow distribution in the inlet section of each channel of the pump station is uneven, and the flow pattern is poor. Firstly, the distribution of water flow and uniformity of water distribution in the pump station forebay under different turning radius of tunnel are compared, and the turning radius of intake corridor is determined by calculating the head loss in different parts of tunnel. A variety of rectification facilities are arranged in the pumping station forebay, including modifying the grid type, setting diversion wall, diversion plate, energy dissipation beam, rectification bottom sill, etc. The recommended scheme is determined by hydraulic model test. The scheme can satisfy the requirement of uniformity of water distribution in the forebay, and the flow pattern along each section of the channel is also good.
Inter-Unit Common Cause Failure Analysis Based on Data From Intra-Unit Cases
Authors: Higo Eishiro , Soga Shota , Miura Hiromichi
Published: October 2020

This study provides an inter-unit common cause failure (CCF) analysis method utilizing the international CCF database. The inter-unit CCF is one of the possible major risks at a site with multiple units. The conventional CCF analysis is based on the CCF database, in which inter-unit CCF events rarely occur in real nuclear power plants. The conventional approach cannot be directly applied to inter-unit CCF cases because of the lack of data. The method ignores the asymmetricity among units, so it may lead to overly conservative outcomes if it is applied to inter-unit CCF. We have proposed a new concept, “commonality factor,” which represents the degree of similarity among components in different units and showed a concept that the inter-unit CCF probability can be expressed in terms of the intra-unit CCF probability with the commonality factor. The commonality factor is approximated based on CCF coupling factors, which tie two or more failures together as a CCF. This study explains how to estimate the commonality factor by analyzing the conventional CCF database and judging commonality among components from different units. The proposed method is demonstrated through two preliminary examples.
Sensitivity Analysis of External Exposure Dose for Future Burial Measures of Decontamination Soil Generated Outside Fukushima Prefecture
Authors: Shimada Asako , Sawaguchi Takuma , Takeda Seiji
Published: October 2020

A large area of east Japan was contaminated by radiocesium following a nuclear accident at the Fukushima Daiichi Nuclear Power Station. Following decontamination of the soil, external effective dose conversion factors were calculated by changing the volume of decontamination soil, depth of cover soil, and distance of the evaluation point from the decontamination soil. The decrement of the factors with an increase of the distance was larger for the smaller volume of decontamination soil. The factors decrease exponentially with an increase of the depth of cover soil in all cases. When there was no cover soil, annual external exposure doses for residents at 1 m from the repository site and public entry were over 10 μSv/y, even for the smallest size (2m × 2m × 1m) and 50 percentile value of radiation concentration (700 Bq/kg). When the surface was covered by 30 cm of non-contaminated soil, the annual external exposure doses were less than 10 μSv/y for the largest size (200m × 200m × 10m) and 95 percentile concentration (2500 Bq/kg).
Numerical Simulation of the Air-Water Two-Phase Flow Across a 90-Degree Vertical-Upward Elbow
Authors: Qiao Shouxu , Zhong Wenyi , Hao Sijia , Qi Peiyao , Tan Sichao
Published: October 2020

The present study investigates the air-water two-phase flow across a 90-degree vertical-upward elbow with the computational fluid dynamics (CFD) simulation. The Eulerian-Eulerian two-fluid model and the Multi Size Group (MUSIG) model are used to predict the development of the detailed interfacial structures between the two phases. The axial development of the void fraction and the interfacial area concentration are investigated and benchmarked with the experimental data measured using the four-sensor conductivity probe. It is concluded that CFD simulation can predict the characteristics distributions of void fraction and interfacial area concentration and their development downstream of the elbow. The double-peaked void fraction distribution is found to be caused by the secondary flow induced by the elbow. The liquid phase on the outer curvature moves to the inner curvature and forms a double counter rotating vortex, entraining the bubbles to form a double-peaked distribution. The elbow effects become dissipated between 33 and 63 hydraulic diameters. The simulation results of liquid-phase and gas-phase parameters can be used to develop the theoretical two-phase flow models for the elbow region.
Simulation for Predicting Condition of Plant Equipment
Authors: Kashiwase Shoichi , Osaki Kenji , Hatakeyama Makoto , Kaneko Tomokazu
Published: October 2020

Condition-based maintenance (CBM) is a maintenance program that recommends maintenance decisions based on maintenance information by condition monitoring. To apply CBM to industrial systems, it is important to diagnose condition of equipment appropriately based on the knowledge of abnormal condition of the equipment. In nuclear power plant, however, it is quite difficult to obtain the knowledge of abnormal condition. For this reason, simulation is expected to be an effective approach to estimate conditions of equipment instead of data acquisition. In this study, we focused on and built models for a motor operated valve and a pump, since the inspection cost is generally expensive due to the number of them in nuclear power plant. We considered a pump system and analyzed the measurement values in normal and abnormal conditions. Our analytical result reproduces the behavior of the measurement values and shows the characteristic deviation from normal condition in abnormal conditions.
A Series of Investigations Inside of Primary Containment Vessel of Fukushima Daiichi Nuclear Power Station Unit2
Authors: Kobayashi Taiji , Yasuda Toshihiro , Sugiura Tessai , Terai Fujio , Hirose Yukinori , Uranishi Atsuyoshi , Sakamoto Naoya
Published: October 2020

In January 2012, Toshiba Energy Systems & Solutions Corporation inserted a camera into the damaged Primary Containment Vessel (PCV) of Fukushima Daiichi Nuclear Power Station unit 2 for the first time since the Great East Japan Earthquake. Since then, we, as a member of International Research Institute for Nuclear Decommissioning (IRID), has developed several devices and has investigated inside of PCV.

The investigation devices for unit 2 are roughly categorized into three types. For the investigations in early stages, just a camera and sensors with simple support tools were inserted through a small hole. Then small crawler type robots were developed to expand the investigation area to inside of Pedestal which exists the central part of PCV. Since the devices found that there were not enough areas to run around by the robots in the Pedestal, a long telescopic guiding pipe device was chosen to deliver cameras, sensors, and tools into the Pedestal. This type of device took movie of wide area of inside of the Pedestal. An image processing software was also developed to get a high quality and extensive picture from the movie taken by a low resolution camera and insufficient lighting. In February 2019, the other telescopic guiding pipe device finally delivered a camera and fingers like tong to grab deposits fallen on the bottom of pedestal for the first time after the earthquake, and found that some of deposits can be lift up.

Some of projects were implemented as subsidized projects of “Decommissioning and Contaminated Water Management” founded by the Ministry of Economy, Trade and Industry.
DeepLofargram: A deep learning based fluctuating dim frequency line detection and recovery
Authors: Han Yina , Li Yuyan , Liu Qingyu , Ma Yuanliang
Published: October 2020

This paper investigates the problem of dim frequency line detection and recovery in the so-called lofargram. Theoretically, long enough time integration can always enhance the detection characteristic. But this does not hold for irregularly fluctuating lines. Deep learning has been shown to perform very well for sophisticated visual inference tasks. With the composition of multiple processing layers, very complex high level representations that amplify the important aspects of input while suppressing irrelevant variations can be learned. Hence, DeepLofargram is proposed, composed of a deep convolutional neural network and its visualization counterpart. Plugging into specifically designed multi-task loss, an end-to-end training jointly learns to detect and recover the spatial location of potential lines. Leveraging on this deep architecture, performance limits of low SNR can be achieved as low as −24 dB on average and −26 dB for some. This is far beyond the perception of human vision and significantly improves the state-of-the-art.
Gut Microbiota: the Emerging Link to Lung Homeostasis and Disease
Authors: Zhou An , Lei Yuanyuan , Tang Li , Hu Shiping , Yang Min , Wu Lingyi , Yang Shiming , Tang Bo
Published: October 2020

The gut microbiota plays a crucial role in the development of the immune system and confers benefits or disease susceptibility to the host. Emerging studies have indicated the gut microbiota could aﬀect pulmonary health and disease through cross talk between the gut microbiota and the lungs. Gut microbiota dysbiosis could lead to acute or chronic lung disease, such as asthma, tuberculosis, and lung cancer. In addition, the composition of the gut microbiota may be associated with different lung diseases, the prevalence of which also varies by age.
Metagenomic Sequencing To Detect Respiratory Viruses in Persons under Investigation for COVID-19
Authors: Babiker Ahmed , Bradley Heath L. , Stittleburg Victoria D. , Ingersoll Jessica M. , Key Autum , Kraft Colleen S. , Waggoner Jesse J. , Piantadosi Anne
Published: October 2020

Broad testing for respiratory viruses among persons under investigation (PUIs) for SARS-CoV-2 has been performed inconsistently, limiting our understanding of alternative viral infections and coinfections in these patients. RNA metagenomic next-generation sequencing (mNGS) offers an agnostic tool for the detection of both SARS-CoV-2 and other RNA respiratory viruses in PUIs. Here, we used RNA mNGS to assess the frequencies of alternative viral infections in SARS-CoV-2 RT-PCR-negative PUIs (
n
 = 30) and viral coinfections in SARS-CoV-2 RT-PCR-positive PUIs (
n
 = 45).
IP7-SPX Domain Interaction Controls Fungal Virulence by Stabilizing Phosphate Signaling Machinery
Authors: Desmarini Desmarini , Lev Sophie , Furkert David , Crossett Ben , Saiardi Adolfo , Kaufman-Francis Keren , Li Cecilia , Sorrell Tania C. , Wilkinson-White Lorna , Matthews Jacqueline , Fiedler Dorothea , Djordjevic Julianne Teresa
Published: October 2020

Invasive fungal diseases pose a serious threat to human health globally with >1.5 million deaths occurring annually, 180,000 of which are attributable to the AIDS-related pathogen,Cryptococcus neoformans. Here, we demonstrate that interaction of the inositol pyrophosphate, IP7, with the CDK inhibitor protein, Pho81, is instrumental in promoting fungal virulence. IP7-Pho81 interaction stabilizes Pho81 association with other CDK complex components to promote PHO pathway activation and phosphate acquisition. Our data demonstrating that blocking IP7-Pho81 interaction or preventing Pho81 production leads to a dramatic loss in fungal virulence, coupled with Pho81 having no homologue in humans, highlights Pho81 function as a potential target for the development of urgently needed antifungal drugs.
Long-Term Temporal Stability of the Resistome in Sewage from Copenhagen
Authors: Brinch Christian , Leekitcharoenphon Pimlapas , Duarte Ana S. R. , Svendsen Christina A. , Jensen Jacob D. , Aarestrup Frank M.
Published: October 2020

The Copenhagen sewage resistome is surprisingly stable in time. The implication is that, at least for cities that are comparable to Copenhagen in terms of sewer infrastructure, few or even single samples provide a robust picture of the resistome within a city.
A Simplified and Efficient Method for Himar-1 Transposon Sequencing in Bacteria, Demonstrated by Creation and Analysis of a Saturated Transposon-Mutant Library in Mycobacterium abscessus
Authors: Foreman Mark , Gershoni Moran , Barkan Daniel
Published: October 2020

Transposon insertion sequencing is a powerful tool, but many researchers are discouraged by the apparent technical complexity of preparing the genomic library for deep sequencing and by the complicated computational analysis needed for insertion site identification. Our proposed method makes the preparation of the library easy and straightforward, relying on well-known molecular biology techniques. In addition, the results obtained from the deep sequencing are easily analyzed in terms of transposon insertion site identification, placing library preparation and analysis within the reach of more researchers in the microbiology community, including those with less computational and bioinformatic resources and experience. This is demonstrated by analysis of the most saturated Tn-mutant library created to date in the emerging pathogen
Mycobacterium abscessus
.
Raising the Resurrection plate from an unfolded-slab plate tectonic reconstruction of northwestern North America since early Cenozoic time
Authors: Fuston Spencer , Wu Jonny
Published: October 2020

The configuration of mid-ocean ridges subducted below North America prior to Oligocene time is unconstrained by seafloor isochrons and has been primarily inferred from upper-plate geology, including near-trench magmatism. However, many tectonic models are permitted from these constraints. We present a fully kinematic, plate tectonic reconstruction of the NW Cordillera since 60 Ma built by structurally unfolding subducted slabs, imaged by mantle tomography, back to Earth’s surface. We map in three-dimensions the attached Alaska and Cascadia slabs, and a detached slab below western Yukon (Canada) at 400–600 km depth that we call the “Yukon Slab.” Our restoration of these lower plates within a global plate model indicates the Alaska slab accounts for Pacific-Kula subduction since ca. 60 Ma below the Aleutian Islands whereas the Cascadia slab accounts for Farallon subduction since at least ca. 75 Ma below southern California, USA. However, intermediate areas show two reconstruction gaps that persist until 40 Ma. We show that these reconstruction gaps correlate spatiotemporally to published NW Cordillera near-trench magmatism, even considering possible terrane translation. We attribute these gaps to thermal erosion related to ridge subduction and model mid-ocean ridges within these reconstruction gap mid-points. Our reconstructions show two coeval ridge-trench intersections that bound an additional “Resurrection”-like plate along the NW Cordillera prior to 40 Ma. In this model, the Yukon slab represents a thermally eroded remnant of the Resurrection plate. Our reconstructions support a “northern option” Farallon ridge geometry and allow up to ∼1200 km Chugach terrane translation since Paleocene time, providing a new “tomographic piercing point” for the Baja-British Columbia debate.