bluuebunny/crossref_metadata_2025_split huggingface dataset:
AtMPK6‐induced phosphorylation of AtERF72 enhances its DNA binding activity and interaction with TGA4/OBF4 in Arabidopsis
Authors: Park H. C. , Park B. O. , Kim H. S. , Kim S. H. , Lee S.W. , Chung W. S. 
Published: October 2020

Abstract

The ethylene‐responsive element binding factor (ERF) family is a large family of transcription factors involved in plant development and environmental stress responses. We previously reported the identification of 29 putative substrates of Mitogen‐activated Protein Kinase3 (AtMPK3), AtMPK4 and AtMPK6, based on a solid‐phase phosphorylation screening using a lambda phage expression library in Arabidopsis thaliana.

In this study, a putative MPK substrate, AtERF72 (At3g16770), was strongly phosphorylated by AtMPK6 on the serine residue at position 151 (Ser151). AtERF72 binds to the GCC box (AGCCGCC) in the promoters of several pathogenesis‐related (PR) genes and activates their transcription. We also show that the DNA‐binding activity of AtERF72 is enhanced upon phosphorylation by AtMPK6 in vitro.

In addition, transient co‐expression experiments in Arabidopsis protoplasts revealed that effector constructs expressing a mutant variant of AtERF72, AtERF72S151D (carrying a Ser to aspartic acid [Asp] substitution at amino acid position 151) showed higher expression of the β‐glucuronidase (GUS) reporter gene driven by the GCC box element than effector constructs expressing the wild‐type AtERF72. Furthermore, yeast two‐hybrid assays revealed that the interaction between AtERF72S151D and TGA4/OBF4 was stronger than that between wild‐type AtERF72 and TGA4/OBF4.

Since AtERF72S151D is equivalent to AtERF72 phosphorylated by AtMPK6 at Ser151, these results suggest that the phosphorylation of AtERF72 by AtMPK6 triggers an event of transcriptional regulation from defence signalling in Arabidopsis.
Galectin‐1 production is elevated in hypertrophic scar
Authors: Kirkpatrick Liam D. , Shupp Jeffrey W. , Smith Robert D. , Alkhalil Abdulnaser , Moffatt Lauren T. , Carney Bonnie C. 
Published: October 2020

AbstractUpon healing, burn wounds often leave hypertrophic scars (HTSs) marked by excess collagen deposition, dermal and epidermal thickening, hypervascularity, and an increased density of fibroblasts. The Galectins, a family of lectins with a conserved carbohydrate recognition domain, function intracellularly and extracellularly to mediate a multitude of biological processes including inflammatory responses, angiogenesis, cell migration and differentiation, and cell‐ECM adhesion. Galectin‐1 (Gal‐1) has been associated with several fibrotic diseases and can induce keratinocyte and fibroblast proliferation, migration, and differentiation into fibroproliferative myofibroblasts. In this study, Gal‐1 expression was assessed in human and porcine HTS. In a microarray, galectins 1, 4, and 12 were upregulated in pig HTS compared to normal skin (fold change = +3.58, +6.11, and +3.03, FDR <0.01). Confirmatory qRT‐PCR demonstrated significant upregulation of Galectin‐1 (LGALS1) transcription in HTS in both human and porcine tissues (fold change = +7.78 and +7.90, P <.05). In pig HTS, this upregulation was maintained throughout scar development and remodeling. Immunofluorescent staining of Gal‐1 in human and porcine HTS showed significantly increased fluorescence (202.5 ± 58.2 vs 35.2 ± 21.0, P <.05 and 276.1 ± 12.7 vs 69.7 ± 25.9, P <.01) compared to normal skin and co‐localization with smooth muscle actin‐expressing myofibroblasts. A strong positive correlation (R = .948) was observed between LGALS1 and Collagen type 1 alpha 1 mRNA expression. Gal‐1 is overexpressed in HTS at the mRNA and protein levels and may have a role in the development of scar phenotypes due to fibroblast over‐proliferation, collagen secretion, and dermal thickening. The role of galectins shows promise for future study and may lead to the development of a pharmacotherapy for treatment of HTS.
Thermofluid Characteristics on Natural and Mixed Convection Heat Transfer From a Vertical Rotating Hollow Cylinder Immersed in Air: A Numerical Exercise
Authors: Rana Basanta Kumar , Singh Bhajneet , Senapati Jnana Ranjan 
Published: October 2020

Abstract
Numerical investigations are performed on natural and mixed convection around stationary and rotating vertical heated hollow cylinder with negligible wall thickness suspended in the air. The fluid flow and heat transfer characterization around the hollow cylinder are obtained by varying the following parameters, namely, Rayleigh number (Ra), Reynolds number (ReD), and cylindrical aspect ratio (L/D). The heat transfer quantities are estimated by varying the Rayleigh number (Ra) from 104 to 108 and aspect ratio (L/D) ranging from 1 to 20. Steady mixed convection with active rotation of hollow vertical cylinder is further studied by varying the Reynolds number (ReD) from 0 to 2100. The velocity vectors and temperature contours are shown in order to understand the fluid flow and heat transfer around the vertical hollow cylinder for both rotating and nonrotating cases. The surface average Nusselt number trends are presented for various instances of Ra, ReD, and L/D and found out that the higher rate of heat loss from the cylinder wall occurs at high Ra, low L/D (short cylinder) and high ReD.
Analysis of Chinese People’s Perception of Radiation and Nuclear Power With Insights Into Behavioral Science
Authors: Wu Hsingtzu , Wang Da-Wei 
Published: October 2020

Abstract
Nuclear power has been at the center of public debate for decades, and social acceptance is critical to its development. A questionnaire survey was conducted to gain a better understanding of Chinese citizens’ perception of nuclear power and radiation from the perspectives of behavioral science. In this study the respondents’ cognitive biases regarding nuclear power and radiation were explained with a dual-system model. Effectiveness of outreach programs about nuclear power was recognized. In addition, this study suggests the prejudice against radiation remained strong in the face of statements of fact and monetary incentives. Finally, some suggestions regarding improvement in outreach programs and public relations policy are provided.
Nuclear Data Sensitivity and Uncertainty Analysis for Generalized Response With RMC Code
Authors: Shi Guanlin , Jia Conglong , Huang Shanfang , Li Kaiwen , Zheng Lei 
Published: October 2020

Abstract
With the increase interest in nuclear data sensitivity and uncertainty analysis, Many researches have developed new methods that can be used to perform sensitivity analysis for different response functions. In previous researches, several methods have been developed in the continuous-energy Reactor Monte Carlo code (RMC) to perform sensitivity analysis of the effective multiplication factor (keff), reaction rate ratios and bilinear response functions. Due to the fact that the reaction rates are also common and important generalized response functions, some methods suitable for its sensitivity and uncertainty analysis need to be developed. In this work, the differential operator method was investigated and implemented in RMC to perform sensitivity analysis of reaction rate. The new capability in RMC are tested on Godiva and Flattop benchmark problems. For the Godiva benchmark, the sensitivity coefficients of fission rate and absorption rate calculated by the newly developed differential operator method in RMC are compared with the reference results from McCARD. For the Flattop benchmark, the sensitivity coefficients of fission rate and absorption rate produced by differential operator method are compared with reference results produced by the direct perturbation method. In general, results produced by the differential operator method agree well with reference results, which verifies the correctness of newly developed differential operator method in performing sensitivity analysis of reaction rates. Moreover, based on sensitivity coefficients calculated by the differential operator method (DOM), the first-order uncertainty quantification method are also developed in RMC to perform uncertainty analysis of reaction rates. The uncertainties of fission rate and absorption rate calculated by first-order uncertainty quantification method and stochastic sampling method are compared. Fairly good agreement can be observed from these uncertainty results.
Cyber Resilience Analysis of SCADA Systems in Nuclear Power Plants
Authors: Galiardi Meghan , Gonzales Amanda , Thorpe Jamie , Vugrin Eric , Fasano Raymond , Lamb Christopher 
Published: October 2020

Abstract
Aging plants, efficiency goals, and safety needs are driving increased digitalization in nuclear power plants (NPP). Security has always been a key design consideration for NPP architectures, but increased digitalization and the emergence of malware such as Stuxnet, CRASHOVERRIDE, and TRITON that specifically target industrial control systems have heightened concerns about the susceptibility of NPPs to cyber attacks. The cyber security community has come to realize the impossibility of guaranteeing the security of these plants with 100% certainty, so demand for including resilience in NPP architectures is increasing. Whereas cyber security design features often focus on preventing access by cyber threats and ensuring confidentiality, integrity, and availability (CIA) of control systems, cyber resilience design features complement security features by limiting damage, enabling continued operations, and facilitating a rapid recovery from the attack in the event control systems are compromised.
This paper introduces the REsilience VeRification UNit (RevRun) toolset, a software platform that was prototyped to support cyber resilience analysis of NPP architectures. Researchers at Sandia National Laboratories have recently developed models of NPP control and SCADA systems using the SCEPTRE platform. SCEPTRE integrates simulation, virtual hardware, software, and actual hardware to model the operation of cyber-physical systems. RevRun can be used to extract data from SCEPTRE experiments and to process that data to produce quantitative resilience metrics of the NPP architecture modeled in SCEPTRE. This paper details how RevRun calculates these metrics in a customizable, repeatable, and automated fashion that limits the burden placed upon the analyst. This paper describes RevRun’s application and use in the context of a hypothetical attack on an NPP control system. The use case specifies the control system and a series of attacks and explores the resilience of the system to the attacks. The use case further shows how to configure RevRun to run experiments, how resilience metrics are calculated, and how the resilience metrics and RevRun tool can be used to conduct the related resilience analysis.
Module Design Layout and Equipment Analysis for Off-Site Prefabrication Manufacture and Assembly in a Small Modular Reactor
Authors: Wrigley P. A. , Wood P. , O’Neill S. , Hall R. , Robertson D. 
Published: October 2020

Abstract
Current large scale, generation 3 nuclear power plants (in the western world) are mostly expensive, delayed and over budget [1] [2]. Modularisation in industrial plants has shown cost savings of between 7–20% and schedule savings between 20–50% [3]. Small Modular Reactors (SMR) defined as “shop fabricated and then transported as modules to the sites for installation” [4] might enable further cost reductions and there are over 50+ SMR designs currently in development internationally.
This paper establishes the module envelope constraints for road transportable factory built SMR Modules and analyses equipment data from a Pressurised Water Reactor (PWR) reactor for these module envelope constraints. The paper found that two tanks and a pump might need to be redesigned for road transport or they may be utilised as one off logistical operations, depending on costs. The 900MWe PWR reactor equipment items were then scaled to reflect a 450MWe plant [5], [6]. If the equipment items are scaled, all equipment would fit into the 5m module space envelope at least in one direction. It may be advantageous to redesign the two tanks. Finally, a simple Mixed Integer Linear Programming (MILP) modularisation optimisation was utilised to layout equipment into a (6 × 50 × 4.5m) module. The objective function was reduced from 14661 for the original Lapp (1989) plant to 2958 for the single module.
Aerosol Source Terms Characterization During Cutting of Fuel Debris Simulants With Different Tools in the Context of Fukushima Daiichi Decommissioning
Authors: Porcheron Emmanuel , Dazon Claire , Leblois Yohan , Gelain Thomas , Chagnot Christophe , Doyen Ioana , Journeau Christophe , Suteau Christophe , Roulet Damien 
Published: October 2020

Abstract
Assessing the production and dispersion of aerosols carrying contamination during Fukushima fuel debris retrieval is IRSN’s contribution to a research project undertaken by a French consortium jointly with ONET Technologies and CEA on behalf of METI and managed by the Mitsubishi Research Institute (Georges et al. 2017). The objective is to obtain quantified data for evaluating the risk of disseminating contamination when implementing cutting tools such as laser or others such as mechanical one, over the next few years, in the process of decommissioning the damaged reactors at the Fukushima-Daiichi plant.
Modelling of the H2020 INSPYRE Fuel Creep Experiment
Authors: Fedorov Alexander , Zwijsen Kevin , van Til Sander 
Published: October 2020

Abstract
To better understand irradiation creep of nuclear fuel, NRG has prepared, as part of the H2020 European project INSPYRE, a separate effect irradiation experiment in the High Flux Reactor (HFR) in Petten (the Netherlands) aiming to measure fuel creep in-pile as a function of temperature, flux, burn-up and axial pressure load. This continuous type of measurement will supply a large data set, leading to more detailed knowledge on fuel behaviour during irradiation. To support the experiment and make optimal use of the generated data, a model is created of the experiment to better predict the behaviour of the fuel samples during irradiation. The current paper describes the numerical model, which couples the 1.5D fuel performance code TRANSURANUS (TU) with a Finite Element Analysis (FEA). The thermal analysis of the experiment is carried out using the FEA. Such approach enables to model a rather complex geometry of the experiment, and to include axial heat transport, which is not implemented in TU. TU is modified in order to use the fuel pellet temperatures obtained using the FEA and to include the axial load present in the experiment. The model is validated against several test cases and used to predict the fuel behaviour during a selection of foreseen irradiation scenario’s. Results of the model will be used in the future for optimization of the irradiation parameters used in the experiment and for analysis of the data obtained during the irradiation.
Coordinated Control Scheme Design of a Nuclear Heating Reactor Cogeneration Plant for Balancing Renewables
Authors: Dong Zhe 
Published: October 2020

Abstract
To balance the intermittent renewable energy (IRE), it is necessary for nuclear power plants (NPPs) to operate flexibly. The flexibility can be given by cogeneration, where the main steam flow are used to drive several thermal loads such as the turbine and seawater desalination process. The electric power of a nuclear cogeneration plant (NCP) can be regulated through adjusting the excess steam flow to the cogeneration processes. Due to the fluctuation of IRE generation, the distribution of main steam to different thermal loads in a NCP varies with net demand frequently, which in turn results in a floating feedwater temperature, and further forms a disturbance to the operation of nuclear reactor. To mitigate the fluctuation of process variables caused by balancing IRE generation, it is necessary to study the coordinated control of flexible nuclear cogeneration plants. In this paper, the scheme of a NCP form by a 200MWth nuclear heating reactor NHR-200II based nuclear steam supply system (NSSS) driving a turbine and a seawater desalination process is first introduced. To balance the IRE generation, a plant coordinated control strategy is newly proposed. The feasibility of this control is verified through numerical simulation, which shows that this NCP can be adopted for balancing IRE.
Cogeneration of Multi-Modular High Temperature Gas-Cooled Reactor Based on Cu-Cl Cycle and High Temperature Electrolysis
Authors: Miao Liu , Zhe Dong , Bowen Li , Di Jiang , Xiaojin Huang 
Published: October 2020

Abstract
With the transformation of global energy, the development of hybrid energy system of nuclear energy and other clean energy is a new research direction. As a representative of the generation IV nuclear energy system, modular high temperature gas-cooled reactor (MHTGR) can provide high quality steam for multiple usage. In this paper, the copper-chlorine (Cu-Cl) cycle and high-temperature electrolysis (HTE) hydrogen production are proposed as the backup processes of the multiple MHTGR unit, and the dynamic model of the cogeneration plant is established according to the conservation of mass, energy and momentum. Furthermore, the coordinated control system of cogeneration plant is proposed to realize load following control. It is programmed on MATLAB/Simulink platform and verified by numerical simulation, and the results show that the model and control scheme of cogeneration plant proposed in this paper is feasible to realize the load following function of stable grid frequency.
Power Rising and Descending Transient for the OTSG of a Small PWR
Authors: Jiang Baihui , Zhou Zhiwei , Xia Zhaoyang , Sun Qian 
Published: October 2020

Abstract
As key heat transfer system in small and medium size pressurized water reactors, once-through steam generators are important parts of energy exchange between primary and secondary circuits, and are very important for the design and operation of reactors. However, two-phase flow and heat transfer in once-through steam generators are very complicated. When a reactor experience power rising and descending transient, the heat removal of once-through steam generator, the flow rate, the inlet fluid temperature and outlet steam temperature will all change accordingly. Especially when a reactor is running at a low power, the flow rate of the secondary side of OTSG is extremely small and the single-phase region of the secondary side of OTSGs is also too small. The two-phase flow instability may occur, which has a serious impact on reactor operation and safety. So, a reasonable power-up and power-down transient scheme is required to ensure operational stability when starting up and shutting down a reactor.
RELAP5/MOD4.0 is a commercial software developed by Innovative System Software, LCC for transient analysis of light water reactors (LWR). After years of development and improvement, RELAP5 has been a basic tool for analysis and calculation of various simulators of nuclear power plants. Scholars all over the world have carried out a large number of analysis of two-phase flow stability using RELAP5, and the results are reliable.
This paper takes once through steam generators with given structural parameters as the research object, and uses RELAP5 as the calculation tool. The influencing factors of flow instability are discussed in this paper, and the operating parameters of the fluid on the primary and secondary sides are designed to satisfy the flow stability under different powers. And a set of power-up and power-down schemes for stable operation is proposed.
Consideration of Relationship Between Decommissioning With Digital-Twin and Knowledge Management
Authors: Taruta Yasuyoshi , Yanagihara Satoshi , Hashimoto Takashi , Kobayashi Shigeto , Iguchi Yukihiro , Kitamura Koichi , Kouda Yuya , Tomoda Koichi 
Published: October 2020

Abstract
Decommissioning of Nuclear Power Plant is a long-term project during which generations are expected to change. Therefore, it is necessary to appropriately transfer knowledge, technology and skills to the next generation. In recent years, in the world of decommissioning, attempts have been made to apply advanced technologies such as utilization of knowledge management and digital technology.
This study describes adaptation in decommissioning from viewpoint of utilizing IT technology called digital twin and aspect of knowledge management.
Single Point Vulnerabilities Management Strategy of Qinshan Nuclear Power Plant
Authors: Hanlin Wang 
Published: October 2020

Abstract
A SPV (Single Point Vulnerability) analysis consists of a list of trains, components or tag IDs, any which could result in a plant shutdown or significant load reduction were the affected equipment be unable to perform its function in supporting generation. Usually, a SPV related components or tag IDs that are identified as a part of an AP-913 based component criticality analysis. But how can we build a solid SPV component management strategy is a key issue for a heathy equipment reliability process. This paper developed a good practices to integrated SPV which related to the generation risk management into the equipment reliability process.
CFD Analysis of S-Gamma Model Coupled With Two-Group Interfacial Area Transport Equations and AMUSIG Model for a Large Diameter Pipe
Authors: Hong Sungje , Schlegel Joshua P. , Sharma Subash L. 
Published: October 2020

Abstract
This paper describes the modeling of flow regimes beyond bubbly flows in a large diameter channel considering polydispersity and bubble induced turbulence using the Eulerian two-fluid approach. A two-bubble-group approach with two-group interfacial area transport equations (IATEs) is used to demonstrate flow phenomena in a large diameter pipe. Source and sink terms for mass and momentum exchanges between the two groups of bubbles and for bubble coalescence and breakup mechanisms are implemented. For predicting particle size and its distribution, S-Gamma (Sγ) model is used. The Sγ model with two-group IATEs are evaluated by comparing local distributions of void fractions and Sauter mean diameters with results of adaptive-multiple-size-group (AMUSIG) models and experimental dataset developed by Schlegel et al., (2012) for model validations. It shows that two-group IATEs with Sγ model predict reasonably accurate flow characteristics of beyond bubbly flow regimes, but also show shortcomings in their accuracies predicting local distributions, which imply that further studies for modeling of interfacial force are needed.
Stratification and Heat Transfer of Molten Corium Pool for In-Vessel Retention
Authors: Xiang Qingan , Deng Jian , Zhu Dahuan , Wu Xiaoli , Bi Jinsheng , Chen Baowen , Cai Rong , Qian Libo , Ma Yugao 
Published: October 2020

Abstract
In-vessel retention (IVR) consists in cooling the molten corium contained in the lower head of reactor vessel by natural convection and reactor cavity flooding. The general approach which is used to study IVR problems is a “bounding” approach which consists in assuming a specified corium pool stratification in the vessel and then demonstrating that the vessel can cope with the resulting thermal and mechanical loads. Thermal loading on the vessel is controlled by the convective heat transfer inside the molten corium pool. Traditionally molten corium pool in the lower head was expected to stratify into two-layer with the dense oxide pool at the bottom and the light metal layer on the top. Based on the MASCA experiments, the increased density of the metal layer attributed to a transfer of uranium metal leads to inverse stratification with a heavy metal layer relocating below the oxide pool. This behavior can be explained by physicochemical interaction between the oxidic and metallic phases of the corium pool. Therefore, a methodology which couples physicochemical effects and thermal hydraulics has been developed to address the IVR issue. The main purpose of this paper is to present this methodology and its application for calculate stratification probability of two-layer and three-layer configuration, analyze the safety margin of IVR in two-layer and three-layer configuration, and evaluate the lower head heat thermal failure probability.