bluuebunny/crossref_metadata_2025_split huggingface dataset:
Detection of adenovirus hepatitis and acute liver failure in allogeneic hematopoietic stem cell transplant patients
Authors: Onda Yoshiyuki , Kanda Junya , Sakamoto Soichiro , Okada Mutsumi , Anzai Naoyuki , Umadome Hiroshi , Tashima Masaro , Haga Hironori , Watanabe Chihiro , Hanaoka Nozomu , Fujimoto Tsuguto , Takaori‐Kondo Akifumi 
Published: October 2020

AbstractHuman adenovirus (HAdV) is an important cause of the common cold and epidemic keratoconjunctivitis in immunocompetent individuals. In immunocompromised patients, HAdV can sometimes cause severe infection such as cystitis, gastroenteritis, pneumonia, encephalitis, hepatitis, or disseminated disease, resulting in significant morbidity and also mortality. In particular, severe cases have been reported in patients after allogeneic hematopoietic stem cell transplantation (allo‐HSCT). Indeed HAdV has been recognized as a pathogen that requires careful monitoring in allo‐HSCT patients. While HAdV hepatitis leading to severe acute liver failure is rare, such liver failure progresses rapidly and is often fatal. Unfortunately, HAdV hepatitis has few characteristic symptoms and physical findings, which makes it difficult to promptly confirm and start treatment. We report here four cases of HAdV hepatitis after allo‐HSCT and their autopsy findings.
Whole Field Measurements to Understand the Role of Varying Depths of Nucleation Site on Vapor Bubble Dynamics and Heat Transfer Rates
Authors: Narayan L Surya , Vijaykumar Pasi , Srivastava Atul 
Published: October 2020

Abstract
This work studies the possible effects of varying depths of cavity on bubbling features and the associated heat transfer rates in nucleate pool boiling regime. A single vapor bubble has been generated on a substrate with a cylindrical cavity at its center that acts as the nucleation site. Experiments have been conducted for three cavity depths (250, 500, and 1000 μm), while keeping its throat diameter constant at 200 μm. With the bulk fluid maintained under saturated conditions, for each cavity depth, surface superheat level has been varied in the range of ΔTsuperheat = 8, 10 and 12 °C. A gradient-based visualization technique, coupled with a high speed camera, has been employed to simultaneously map the changes in thermal gradients during the formation of the vapor bubble as well as bubble dynamic parameters. The image sequence obtained has been qualitatively and quantitatively analyzed to elucidate the dependence of bubbling features and various heat transfer processes on cavity depth. With an increase in the depth of cavity, the net effect of reduction in the available thermal energy due to the increased convection effects and significant depletion of superheated layer are identified as the dominant heat transfer processes that influence the bubbling features. Furthermore, based on the statistics of bubble departure characteristics, the cavity with higher depth (1000 μm) showed a much stable bubble formation with minimal variation in the bubble departure frequency as compared to the bubbling features from a cavity with smaller depth (250 μm). Evaporative heat transfer process has been identified as the primary cause for increased inconsistency of bubbling features at high superheat conditions for experiments performed for low cavity depths.
On the Development of Correlations for Bubble Liftoff Parameters During Subcooled Nucleate Flow Boiling Using Nonintrusive Dynamic Measurements
Authors: Sinha Gulshan Kumar , Srivastava Atul 
Published: October 2020

Abstract
Accurate prediction of bubble dynamic parameters is essential to improve boiling heat transfer models. Considering the complexities and challenges associated with performing a large number of boiling experiments, researchers have realized the importance of experimental correlations for predicting bubble dynamic parameters. In this direction, we report an experimental work concerned with the development of correlations for various bubble liftoff parameters during nucleate flow boiling regime. As a definite advancement, the experimental measurements have been performed in a purely nonintrusive manner, thereby minimizing the errors arising due to the interaction of any external probe with the process under study. The measurement approach makes use of a gradient-based imaging technique to simultaneously map the bubbling features and thermal field around a single vapor bubble generated under subcooled flow boiling conditions. Experiments have been performed in a rectangular channel for a wide range of heat fluxes (q" = 20–50 kW/m2), subcooling level (ΔTsub = 2–9 K), and Reynolds numbers (Re = 600–6000) with water as the working fluid. Results show a strong dependence of bubble liftoff parameters on Reynolds number, subcooling level, and applied heat flux. Based on the experimental measurements, empirical correlations have been developed for various bubble liftoff parameters as a function of Jacob number and Reynolds number. Predictions made through the developed correlations are found to be in good agreement with the measured values as well as with the values reported in the available literature. Of all the bubble parameters, maximum deviation between the predicted and measured values (≈23%) was found to be in bubble release frequency.
Research on Natural Convection of Two-Layer Corium Pools
Authors: Luo Simin , Sun Hongping , Zhang Yapei , Wu Junmei , Qiu Suizheng , Su Guanghui 
Published: October 2020

Abstract
As a promising mitigation strategy, In-vessel-retention (IVR) has been an important research focus in realms of nuclear safety analysis. For IVR analysis, it’s of importance to study the natural convection behaviors of the corium pool formed in the lower plenum of the reactor vessel. Because of the different densities between corium compositions, typically the corium pool is believed to evolve to two-layer configurations, with a heavier oxide layer at bottom and a metal layer on top. To know better the convection behaviors of the layered corium pools, research was performed both experimentally and numerically. The test section for the experiment is a two-dimensional semi-ellipsoidal slice, which measures a span of 1.2m, a total pool depth of 0.75m and a thickness of 0.2m. Two immiscible fluid, i.e. fluorinert fluid FC-40 and water, were chosen to simulate the oxide layer and the metal layer respectively. Based on the experimental settings, numerical simulations were performed, and the results were compared with the experimental data. Further, discussions were made on the focusing effect, the relationship between the temperature and the velocity field, and also the coupling modes in the two-layer system. These results offer more details to the knowledge of layered corium pools and can be helpful for IVR analysis.
Study on Eutectic Melting Behavior of Control Rod Materials in Core Disruptive Accidents of Sodium-Cooled Fast Reactors: Project Overview and Progress Until 2018
Authors: Yamano Hidemasa , Takai Toshihide , Furukawa Tomohiro , Kikuchi Shin , Emura Yuki , Kamiyama Kenji , Fukuyama Hiroyuki , Higashi Hideo , Nishi Tsuyoshi , Ohta Hiromichi , Liu Xiaoxing , Morita Koji , Nakamura Kinya 
Published: October 2020

Abstract
One of the key issues in a core disruptive accident (CDA) evaluation in sodium-cooled fast reactors is eutectic reactions between boron carbide (B4C) and stainless steel (SS) as well as its relocation. Such behaviors have never been simulated in CDA numerical analyses in the past, therefore it is necessary to develop a physical model and incorporate the model into the CDA analysis code. This study focuses on B4C-SS eutectic melting experiments, thermophysical property measurement of the eutectic melt, and physical model development for the eutectic melting reaction. The eutectic experiments involve the visualization experiments, eutectic reaction rate experiments and material analyses. The thermophysical properties are measured in a range from solid to liquid state. The physical model is developed for a severe accident computer code based on the measured data of the eutectic reaction rate and the physical properties. This paper describes the project overview and progress of experimental and analytical studies conducted until 2018. Specific results in this paper are boron concentration distributions of solidified B4C-SS eutectic sample in the eutectic melting experiments, which would be used for the validation of the eutectic physical model implemented into the computer code.
A Fault Diagnosis Method Based on Signed Directed Graph and Correlation Analysis for Nuclear Power Plants
Authors: Liu Chongchong , Wu Guohua , Yang Congsong , Li Yunwen , Wu Qian 
Published: October 2020

Abstract
Fault detection and diagnosis (FDD) provides safety alarms and diagnostic functions for a nuclear power plant (NPP), which comprises large and complex systems. NPP has a large number of parameters which make it difficult achieve FDD. Now many diagnosis methods have lack of better explanation for faults and quantitative analysis. Therefore, to overcome the “black box” of FDD based on data-driven methods, this paper adopts signed directed graph (SDG) in knowledge graph for FDD. It can express the cause and effect of accidents through knowledge maps. At same time, this paper uses correlation analysis to conduct a quantitative analysis between parameters and faults.
It this paper, SDG is used to explain the reason of faults. In order to quickly achieve FDD, this paper introduces a quantitative analysis method. It combines expert system and correlation analysis method to analyze the weight of each parameter. On this basis, matrix reasoning is used to achieve the FDD, and the reason is shown in SDG model inference. This paper takes loss of coolant accident as the case study, the case shows that the proposed method is superior to the conventional SDG method and can diagnose the faults timely.
Study on On-Site Calibration Method of Stationary Neutron Dosimeter
Authors: Lv Jinxu , Lv Ning , Guo Huiping , Sun Mingyan , Zhao Kuo , Lv Wenhui , Wei Zhihao , Xiao Qizhan , Li Daji , Liu Yang 
Published: October 2020

Abstract
The on-site calibration system of stationary neutron ambient dose equivalent instrument is mainly comprised of a small controllable neutron source and a reference neutron ambient dose equivalent instrument. According to the principle of “relative calibration method”, a small controllable neutron source continuously emits neutrons at the calibration site to construct a neutron radiation field.
The calibration factor (NB) can be obtained by comparing the response numbers from two instrument, the instrument to be calibrated and the reference instrument, which are symmetrically placed in the neutron radiation field. In order to complete the transfer of the ambient dose equivalent calibration coefficient of the reference instrument from National Metrology Center (252CfStandard radiation field) to nuclear facility site, the calibration coefficient needs to be corrected, that is, multiplied by the “energy correction coefficient”. Energy correction coefficient includes: (1) “Instrument Energy Response Correction Coefficient” kε for neutron fluence to neutron counting of the instrument, (2) “Conversion Correction Coefficient” kΦ for neutron fluence to neutron ambient dose equivalent.
Tests have shown that the ambient dose equivalent rate measurement error of the instrument which have been calibrated but without energy correction was 16.5%, by contrast the measurement error was 2.9% with energy correction. It can be seen that the energy correction is necessary and effective for the measuring instrument of neutron ambient dose equivalent during the on-site calibration process.
Study on Eutectic Melting Behavior of Control Rod Materials in Core Disruptive Accidents of Sodium-Cooled Fast Reactors: Validation of a Multi-Phase Model for Eutectic Reaction Between Stainless Steel and Boron Carbide
Authors: Liu Xiaoxing , Morita Koji , Yamano Hidemasa 
Published: October 2020

Abstract
Investigation of the eutectic reaction in a core disruptive accident of sodium cooled reactor is of importance since reactor criticality will be affected by the change in reactivity after eutectic reaction. In our previous study, a two-dimensional fast reactor safety analysis code, SIMMER-III, was extended to include a physical model to simulate the eutectic reaction between stainless steel (SS) and B4C. Based on experimental knowledge on eutectic reaction, the growth of eutectic material was modeled according to a parabolic rate law. Heat and mass transfer behaviors among reactor materials including a eutectic composition in solid and liquid phases were also modeled considering both equilibrium and non-equilibrium processes in phase change. Physical properties of the eutectic composition were also formulated based on experimental measurements for 5 mass% B4C-SS composition. In this study, we extended the eutectic reaction model to SIMMER-IV, a three-dimensional counterpart of SIMMER-III. We performed validation analysis using SIMMER-III and SIMMER-IV with the developed model based on an experiment, where a B4C pellet was immersed into a molten SS pool. Boron concentration in the pool was measured at several time points and the boron concentration after solidification of the molten pool was compared with the experiment post analysis result. Simulation results of boron distribution are comparable to the experimental results.
Design Applicability of the Advanced Spectrum Method Assisted by Time History Analysis for Multiply Supported Piping System
Authors: Yoshida Ayaka , Takayama Yoshihiro , Shudo Hiromichi 
Published: October 2020

Abstract
The independent support motion response spectrum method (ISM) is currently used for seismic analysis to calculate the response of multiply supported piping system with independent inputs of support excitations. This approach may derive considerable overestimation in the combination of group responses under the absolute sum rule of NUREG-1061. To reduce the excessive overestimation, an advanced method named “Spectrum Method Assisted by Time History Analysis (SATH)” has been proposed. In the SATH method, modal responses by multiple excitations are combined with correlation coefficients which are calculated from the covariances and the standard deviations of time history responses of the oscillators having each of modal frequencies of piping system by each of the floor acceleration time history excitations.
In this paper, the applicability of the SATH method to actual design which uses broadened Floor Response Spectra (FRS) was examined. Conservativeness of the SATH method is also discussed with presenting that responses derived from the SATH method tend to exceed responses by the time history analysis even when many different time intervals of acceleration time histories are used within the range of the FRS broadening.
Numerical Simulation of Liquid Jet Behavior in Shallow Pool by Interface Tracking Method
Authors: Suzuki Takayuki , Yoshida Hiroyuki , Horiguchi Naoki , Yamamura Sota , Abe Yutaka 
Published: October 2020

Abstract
In the severe accident (SA) of nuclear reactors, fuel and components melt, and melted materials fall to a lower part of a reactor vessel. In the lower part of a reactor vessel, in some sections of the SAs, it is considered that there is a water pool. Then, the melted core materials fall into a water pool in the lower plenum as a jet. The molten material jet is broken up, and heat transfer between molten material and coolant may occur. This process is called a fuel-coolant interaction (FCI). FCI is one of the important phenomena to consider the coolability and distribution of core materials.
In this study, the numerical simulation of jet breakup phenomena with a shallow pool was performed by using the developed method (TPFIT). We try to understand the hydrodynamic interaction under various, such as penetration, reach to the bottom, spread, accumulation of the molten material jet. Also, we evaluated a detailed jet spread behavior and examined the influence of lattice resolution and the contact angle. Furthermore, the diameters of atomized droplets were evaluated by using numerical simulation data.
Experimental Investigation of Behavior of Impinging Liquid Jet in a Shallow Pool by 3D-LIF
Authors: Yamamura Sota , Kimura Fumihito , Yoshida Hiroyuki , Kaneko Akiko , Abe Yutaka 
Published: October 2020

Abstract
In some scenarios of severe accidents, the core materials melt and fall into a water pool in the lower plenum as a jet. The molten material jet is broken up, and heat transfer between molten material and coolant occurs. The aim of this study is to clarify the behavior of liquid jet falling into a shallow pool. In a previous study, it is clarified that, in a shallow pool, the jet spread radially after bottoming, and the atomization occurs with high flow velocity in a shallow pool. the detail of atomization and the spreading of the jet cannot be measured by the limitation of a 2D visualization method.
In this study, a 3D-LIF method is used to obtain the detail 3D shape data of the jet. The 3D visualization of the jet is conducted. Using 3D shape data, the liquid film and the atomized droplet are measured. The initial jet velocity is selected as a parameter. As a result, following knowledge is obtained. The thickness of liquid film increases suddenly, and the radius of thin liquid flow increases with the increase of the initial jet velocity. The number of atomized droplets increases with the increase of the initial jet velocity. However, the size of the droplets are not influenced by the initial jet velocity.
Classification Analysis of Communication System of Nuclear Power Plant
Authors: Yun Yu , Shen Zheng , Jing Liu 
Published: October 2020

Abstract
The communication system of nuclear power plants in China is not a safety class system, but it plays an important role in the safe operation of nuclear power plants. Under emergency state, the communication system is a prerequisite for accident management. In order to ensure communication on-site and off-site, diverse communication sub-systems are designed throughout the nuclear power plant, including various communication means for voice, data and images. For an advanced generation II pressurized water reactor (PWR) nuclear power plant (NPP) in China, there are various subsystems, including normal telephone system, safety telephone system, grid telephone system and so on.
Although NPPs have designed diverse communication sub-systems, there is not any clear classification of the sub-systems, which is not enough for the reliability of communication sub-systems under accident conditions. Therefore, it can hardly ensure effective communications between different emergency response organizations and this will influence the mitigation of the accident. In order to identify the importance of different communication sub-systems, to optimize the design of communication system, and to improve the reliability and efficiency of nuclear power plant communication system, it’s necessary to analyze the function and operation of each sub-system, as well as to develop the classification method of nuclear power plant communication system.
Considering the availability and reliability of onsite and offsite communication under emergency conditions, slightly considering economic issue, this paper determines 7 assessment factors and develops a set of scoring methods for communication system classification. On this basis, this paper completes the classification of the communication system for an advanced generation II PWR NPP, which provides a reference for communication system classification and provides the technical basis for design modification of the communication system.
Walk-Away Safe Modular Small and Micro Reactors for Electricity Generation and Process Heat
Authors: El-Genk Mohamed S. , Schriener Timothy 
Published: October 2020

Abstract
The walk-away safe, SLIMM and VSLLIM small and micro reactor designs have recently been developed at the University of New Mexico’s Institute for Space and Nuclear Power studies. These passive operation reactors concepts could generate 10–100 MWth and 1.0–10 MWth without refueling for 6.3–68 full power years (FPYs) and 5.8–92 FPY, respectively. These factory constructed, assembled and sealed reactors would be transported to designated sites by truck, rail or barge, and installed below ground, to protect against impact by an airplane or a missile, and mounted on seismic insolation bearings, to resist earthquakes.
This paper describes the design, operation and passive safety features of the SLIMM and VSLLIM reactors and presents estimates of operation lives and post-operational biological dose rates. Results of passive removal of decay heat by natural circulation of ambient air along the outer surface of the guard vessel wall will also presented and discussed.
A Real-Time Hard X-Ray Tomographic System Designed for HL-2A Fast Electron Bremsstrahlung Radiation
Authors: Yuan Chen , Wu Jun 
Published: October 2020

Abstract
A real-time hard X-ray (HXR) tomographic system is designed for HL-2A tokamak, which is dedicated to the real-time tomography of fast electron bremsstrahlung radiation during the lower hybrid (LH) driven mode within the energy range of 20keV to 200keV. This system has realized the investigation of HXR energy from 12 different chords on the equatorial plane of the reaction region. The spatial and temporal resolutions of the system are 2cm and 10ms, separately. HXR detection is accomplished by a self-designed detector array, with a structure of 12 arc arranged cadmium telluride (CdTe) semiconductors and their corresponding collimators. The real-time HXR acquisition and processing is achieved by the main electronic system, which is comprised of a high speed analog-to-digital module and a high performance signal processing unit. Due to high HXR flux and the real-time demand in measurement, the HXR tomography is accomplished by several customized digital processing algorithms based on FPGA logic resources, such as the digital real-time spectrum measurement, the trapezoidal shaper, the pile up filter, and the baseline restorer, etc. This system has been proved to be qualified as a dependable platform of fast electron bremsstrahlung radiation research during LH mode on HL-2A, which provides indispensable parameters for plasma state during fusion reaction.