bluuebunny/crossref_metadata_2025 huggingface dataset:
Ecological indicators for assessing ecological success of forest restoration: a world review
Authors: Gatica‐Saavedra Paula , Echeverría Cristian , Nelson Cara R.
Published: October 2017

Restoration is increasingly being used to reverse degradation and destruction of forest ecosystems. With increasing investment in restoration, there is an urgent need to develop effective programs to assess treatment efficacy and effects. We conducted a global review of forest restoration assessments, in order to identify geographic trends in the locations where assessments have been implemented and the specific ecological attributes (ecosystem composition, structure, and function) and indicators being used to measure effects. We found that the number of forest restoration assessments varied by region and was not related to degree of degradation or restoration need. Some regions, like Africa, which have experienced high rates of forest loss and degradation, had few assessments. The majority (43%) of assessments included indicators for only two of three key ecological attributes (composition‐structure or composition‐function) and assessments on average used fewer than three indicators per attribute. The most commonly employed indicators for composition were richness and abundance of plant species and for structure were height and diameter of trees, variables that are generally relatively easy to measure. The use of functional indicators has been increasing over time and they are now more commonly used than structural indicators. The most common functional indicators were soil functions. Most investigators evaluated treatment effects for 6–10 years after implementation. Our findings related to gaps in analysis of ecological indicators can serve as a guide for developing monitoring and assessment protocols for current global forest restoration initiatives by 2020–2030.
The Schwarzian equation for completely replicable functions
Authors: El Basraoui Abdelkrim , McKay John
Published: October 2017

We describe the Schwarzian equations for the 328 completely replicable functions with integral$q$-coefficients [Fordet al., ‘More on replicable functions’,Comm. Algebra22 (1994) no. 13, 5175–5193].
Dynamic Behavior of a Metamaterial Beam With Embedded Membrane-Mass Structures
Authors: Chen Jung-San , Chien I-Ting
Published: October 2017

Flexural propagation behavior of a metamaterial beam with circular membrane-mass structures is presented. Each cell is comprised of a base structure containing circular cavities filled by an elastic membrane with a centrally loaded mass. Numerical results show that there exist two kinds of bandgaps in such a system. One is called Bragg bandgap caused by structural periodicity; the other is called locally resonant (LR) bandgap caused by the resonant behavior of substructures. By altering the properties of the membrane-mass structure, the location of the resonant-type bandgap can be easily tuned. An analytical model is proposed to predict the lowest bandgap location. A good agreement is seen between the theoretical results and finite element (FE) results. Frequencies with negative mass density lie in the resonant-type bandgap.
Based on Human Behavior Process of Human Error Defensive Management Research for NPP
Authors: Yang Shen , Bo Geng , Dan Li
Published: October 2017

According to the research of nuclear power plant human error management, it is found that the traditional human error management are mainly based on the result of human behavior, the event as the point cut of management, there are some drawbacks. In this paper, based on the concept of the human performance management, establish the defensive human error management model, the innovation point is human behavior as the point cut, to reduce the human errors and accomplish a nip in the bud. Based on the model, on the one hand, combined with observation and coach card, to strengthen the human behavior standards expected while acquiring structured behavior data from the nuclear power plant production process; on the other hand, combined with root cause analysis method, obtained structured behavior data from the human factor event, thus forming a human behavior database that show the human performance state picture. According to the data of human behavior, by taking quantitative trending analysis method, the P control chart of observation item and the C control chart of human factor event is set up by Shewhart control chart, to achieve real-time monitoring of the process and result of behavior. At the same time, development Key Performance Indicators timely detection of the worsening trend of human behavior and organizational management. For the human behavior deviation and management issues, carry out the root cause analysis, to take appropriate corrective action or management improvement measures, so as to realize the defense of human error, reduce human factor event probability and improve the performance level of nuclear power plant.
Design and Development of FES’s HMI for HPR1000 NPP
Authors: Zhang Caike , Liu Peibang , Gao Hanjun , Xie Chenglong , Qu Ming
Published: October 2017

In order to ensure the correctness of the DCS design documents, in the HPR1000 nuclear power plant (referred to as NPP) DCS engineering design phase, the documents need to be verified and validated. Based on this, taking the HPR1000 NPP as a reference unit, the design verification platform which named Full Scope Engineering Simulator (referred to as FES) is designed and developed. The FES project packages are constituted by process system modeling simulation software, DCS control logics simulation software (referred to as Level1), human-machine interface simulation software (referred to as Level2 or HMI), operating panel simulation software, I/O interface simulation software, some computer equipments other hardware, and so on. The FES can be used to validate the HPR1000 NPP DCS control logic diagrams, human-machine interface, and procedures guidelines, and so on. When the design institute submits the DCS design documents (include Level1 and Level2), these documents can be imported into the FES platform through some special import tools, meanwhile the errors and non-compliances are printed out. The errors and non-conformities are feedback to the design institute, according to which the design institute upgrades the design documents. Then the ascending version design documents are feedback and imported into the FES platform, and again printed out errors and non-compliances, so the design documents are verified and validated cycle by cycle, ensuring the correctness of the design documents offered by design institute before the final version as much as possible. This paper mainly explains the design and development of FES human-machine interface, including HMI component library software, special import tools software, the supervisory control and data acquisition (SCADA) software and other auxiliary software design and development. Then the design and development of the communication interface program between the Level2 and Level1 is explained, through which the Level2, Level1, and Level0 (the process model of HRP1000 NPP) are integrated, constituting a close-loop verification system. And explains the verification contents, lists some typical kinds of errors in the validation process, including the component type is not consistent between Level1 and Level2, use the wrong library of the HMI, add the wrong color property of the fluid medium, and the component is not named after rules between the developer and the design institute, etc. This paper also lists some examples of verification, elaborates the necessity of FES in the HPR1000 NPP DCS design process; and throughout the whole design and development process of FES, to sum up some problems in the development, meanwhile making a prospect technological development of the FES.
Research on Simulation Technology for Supervisory Control Software in NPP DCS
Authors: Gao Hanjun , Qu Ming , Zhang CaiKe , Xu Kui , Liu PeiBang , Xie Chenglong
Published: October 2017

The DCS simulator is a crucial component of FSS which is used for operator training and license examination. It is always divided into process control level (short for Level I) and supervisory control level (short for Level II). Among these, the Level II directly faces to operators and engineers. Therefore it requires a high degree of fidelity, reliability and good user experience. Be aimed at the requirement of the Level II, a high-fidelity simulation system based on MVC mode is constructed. And multiple technologies of self-dependent innovation and custom-built software are designed and developed, such as the data middleware, the configuration software, and the SimBase et al. The successful application in the HAINAN’s NPP proves that the DCS Level II’s simulator has a high fidelity, reliability, and accumulates valuable experiences for the development of our own DCS Level II as well.
Nuclear Power Project Performance Management System Research Based on the KPIs
Authors: Zhao Wenzhao , Yu Hang , Tian Limin , Li Xiaobing , Yan Guangyi
Published: October 2017

The foothold of this paper is the systematic mechanism of the performance management. Combining with the company’s current situation and requirements of the nuclear power project management, we build a scientific and effective performance management system of the nuclear power project which using the KPI method as the core. This system includes KPIs library development, performance management plan preparation, process control and supervision, performance evaluation, outcome feedback and application. We aim to improve the project management efficiency, and promote the standardization and systematization of the nuclear power project performance management.

This article has an important reference and guidance for nuclear power project management, and continuous improvement will be made to achieve the best practice in the nuclear industry.
Functional Information of System Components Influenced by Operators’ Actions on Emergency Operating Procedure
Authors: Suryono Tulis Jojok , Gofuku Akio
Published: October 2017

Computer-based emergency operating procedures offer some benefits compared with the paper-based EOP. However, most of them do not fully provide clear and necessary information related to the procedure steps, such as future plant behavior and the affected components after taking an action of each step because of the limited space available on the virtual display unit. This paper investigates a technique to derive and indicate such kind of information using a Multilevel Flow Modeling model of simplified EOP of steam generator tube rupture accident of a pressurized water reactor plant. In this case, the MFM model is built based on operator’s action on each procedure step. Then, by using the influence propagation rules for function primitives of MFM model, the influenced functions of plant components necessary for confirming the procedure step can be gathered. This information will help operators to understand and take the actions in order to achieve the objective of each procedure step.
Technical Architecture About Safety I&C System Software V&V in Nuclear Power Plant
Authors: Wang Shengchao , Gu Pengfei , Ye Wangping , Chen Weihua
Published: October 2017

Digital instrumentation and control (I&C) system has been widely applied in nuclear power plant (NPP), which brings some new problems like software failure. The software failure is easy to come into being the common cause failure (CCF), leading to the failure of the whole protection system for NPP, thus endangering the safety of the NPP. In order to obtain the high quality software for computer important to safety, it is necessary to implement the software verification and validation (V&V) to guarantee that the software can completely and correctly meet the expected requirements of safety functions and performances. Moreover, for the guidance of software V&V and to strengthen the reliability of the software, laws and regulation standards related to the software V&V, were analyzed. On the basis of requirements of the laws and regulations standards and the results of analysis, a technical architecture about nuclear safety digital instrumentation and control system (DCS) software V&V was proposed, which consists of a technical model of software V&V and a general quality assurance (QA) system. The technical model of software V&V mainly illustrated some problems and discussed how to deal with it on the specific implementation of V&V activities and tasks. Meanwhile, in order to support the software V&V activities and tasks carried out efficiently, the general QA system, corresponding to the technical model of software V&V, was established which can supply the visual evidence for quality control process. At this point, it is expected that the results as studied above can provide some technical references in the transformation, formulating or modifying related standards of China and to promote the quality of nuclear safety DCS software.
The Reliability Model for the FPGA-Based Instrument and Control System Using Colored Petri Net
Authors: Ma Zhanguo , Yoshikawa Hidekazu , Yang Ming
Published: October 2017

As the FPGA based solution is the software design process but the product is hardware, it is feasible to licensing the FPGA based I&C system by the nuclear regulator. By noticing the advantages of the FPGA technology and gaining the experience in other industrial fields, it is exploited as the design solution in the NPP I&C system, mainly for the safety protection systems. The paper proposes a reliability evaluation model by using the Colored Petri Net for an FPGA based I&C system wherein the fault coverage is considered. And the proposed models are the generic models for I&C system except that the parameters are not same. The models concern on the hardware modules without the logics (software failure) to control and protect the plant. The detection and recovery time influence for the system availability is evaluated in the model. Based on the proposed model, the Monte Carlo simulation is performed as it is the possible way for a complex system model with uncertainty. Consequently, the indicators such as MTTR, MTBF, PFD and so on are calculated. Three main conclusions are obtained from the calculated results i.e., (i) In order to increase the reliability of the FPGA based modules, it is recommended to consider the logic function to separate the FPGA chip to different groups to decrease the failure rate of the module; (ii) It is suggested to shorten the fault tolerance technique required time and to increase the online fault coverage to decrease the MTTR and minimize the failure probability of the system. (iii) The independent failure of the modules and the detection and the repair time delay only cause the two trains failure.
Co-Verification Method of NPP Protection System Based on FPGA Platform and SCADE Model
Authors: Chen Ke , Xu Leilei , Zhou Zhende
Published: October 2017

In recent years, the FPGA-based digital technology is being introduced to replace CPU-based digital platform in the NPPs because of its high reliability, security and easy maintainability. Thus the verification and validation of FPGA-based I&C system has aroused a great attention.

As most safety I&C systems are composed of four divisions, and perform 2-out-of-4 coincidence voting logic to actuate RT and ESFs. The 4 redundant portions are mostly the same. It is uneconomical and time-consuming to set up the full-size verification platform based on 4 redundant channels. In this paper, the co-verification method based on FPGA platform and SCADE model is proposed to efficiently verify the reactor protection logic.

Based on the co-verification platform, it shows the proposed co-verification method is applicable for functional verification. The proposed co-verification method can be used for the preliminary assessment and functional verification of the FPGA-based I&C system and provide a semi-hardware platform for the development of the safety I&C system based on the FPGA platform, which will facilitate the evaluation of FPGA technology in the NPP and accelerate the V&V process.
Research on Analysis and Experiment of Multistage Orifice Throttle in Nuclear Power Plant
Authors: Liu Yan , Yang Xue-long , Wang Xian-yuan , Tian Rui-feng
Published: October 2017

In this work, multi-stage orifice throttling devices in CVS were designed and their resistance coefficient was calculated and analyzed for different working conditions in order to optimize their characteristics. The influence of aperture ratio and hole spacing on the resistance coefficient in a two-stage throttle assembly was analyzed. Using multiple linear regression analysis, relationship between the resistance coefficient and the aperture ratio in a two-stage throttle is obtained and successfully applied to the design of multi-stage throttling devices.
Operation Feasibility Study of Adding Fourth Main Feedwater Pump to CAP1000
Authors: Xiao Ke
Published: October 2017

The standard design of CAP1000 Feedwater system configures 3×33.3% main feedwater pumps. In the operational transient “loss of one main feedwater pump”, the plant will ramp down the turbine load to 70% automatically and remain load balance by the two operational pumps.

The owners of CAP1000 submitted a claim for the fourth main feedwater pump so that the plant can maintain 100% turbine load throughout the transient “loss of one main feedwater pump”. The owners insisted that this design modification avoids partial-load operation and enhance the economic benefit.

In this paper, the trip risk of the addition of the fourth main feedwater pump is evaluated base on the transient analysis of “loss of one main feedwater pump”. And the solution of the fourth pump actuation plan which farthest decrease the reactor trip probability is presented. It is the most balanced plan between economy and feasibility.
Linear Active Disturbance Rejection Control for Fast Nuclear Reactor
Authors: Feng-ming Liu , Cong Shen , Hua-xia Zhang , Shi-liang Zhou , Yu-yan Liu , Yu-rong Li , Xiaoruo Wang , Khalid Mohanned
Published: October 2017

Since dynamics of a fast nuclear power reactor is quite nonlinear, uncertain and time-varying. Moreover, the reactor period depends on the neutron lifetime and it has a crucial importance in reactor power so it is necessary to have a control system for the reactor power. A reliable method for power-level control named linear active disturbance rejection control (LADRC) based on power-level for a fast nuclear power reactor is presented in this paper. The reactor core is simulated based on the point kinetics equations and one delayed neutron group, fast reactor reactivity equation which is very different from thermal reactors and thermal-hydraulics model of the reactor core. This model is completed by calculating initial value of related parameters and system debugging. Finally, it is prove that this model is reasonable and can be used for controlling. The structure of LADRC is very simple and the Bandwidth-Parameterization based controller tuning is easy. It consists of a PD controller and an extended state observer (ESO). Two LADRC systems are designed for power control with model information and without model information respectively. Simulation results show that both the designed controllers have satisfactory performance over the wide range of reactor operating conditions, while LADRC with model information has faster response, less overshoot, better disturbance rejection ability, greater robustness.
The Study on Hydraulic Oil Property Influences on Motion Performances of the Main Steam Isolation Valve and its Driving Device
Authors: Pan Dalei , Kuang Hongbo , Gu Shenjie , Zhong Hua , Guo Guangyue
Published: October 2017

Hydraulic oil is an important part of the pneumatic-hydraulic type driving device of the main steam isolation valve, whose physical property effects on the electromechanical equipment performances were never studied. This paper reports the research on the physical property effects. The effects of the environment on hydraulic oil physical properties are discussed and the plant environment are introduced. Based on the mechanic-electric-hydraulic co-simulation model of the equipment, the motion performance indexes of the driving devices are illustrated and the cycle simulation methodology is proposed. The influencing laws of hydraulic oil properties on the motion performance indexes are revealed and illustrated. The results show that hydraulic oil properties have important effects on the motion performances of the equipment. The results provide important references for designing the equipment and experimental verification, which make the equipment more reliable.
Design Verification for Control Logic in Nuclear Power Plant
Authors: Li Hui , Zhang Guangxin
Published: October 2017

The control function for process, HVAC and electrical systems in nuclear power plant (NPP) are represented by control logic diagram. To develop Distributed Control System (DCS), the designer and supplier should complete the activities of control logic configuration, testing and verification which are based on control logic diagram. Design Verification is an effective method to ensure the correctness of control logic design. This paper represents an overview of the procedure for control logic design verification in NPP I&C system, as well as the common function of platform/system to implement design verification. It focuses on the design verification for component control logic, to confirm the design outputs meet the inputs — the control requirements of plant’s systems. Finally, a conclusion about the design verification system as well as future scenarios is given.
Prestressed Time-Limited Aging Analyses of Concrete Containment Structure
Authors: Cai Dahua , Wang Yonghuan , Zhang Jiangtao , Yang Lin , Rong Hua , Li Jiwa , Wu Zhiming
Published: October 2017

For prestressed concrete containment structure, prestress loss is a key factor that affects the performance of containment structure. Therefore, prestressed time-limited aging analysis (TLAA) is essential for containment structures. The main objective of prestressed TLAA is to assess the safety of containment structures after prestress loss occurred over time. This paper takes the in-service containment structure as an example to investigate the method of TLAA for grounted prestressed containment structure. Firstly, it introduces methods for prestressed TLAA. Secondly, a finite element model of containment structure is established to calculate the minimum required value (MRV) of prestress. The numerical model is verified by the pressure test results. Thirdly, prestress loss of tendons is calculated. Finally, the residual prestress of tendons are compared with the MRV of prestress to confirm whether the containment can service in a certain period. This study can provide guidance for goouted prestressed TLAA of containment structures.
Design of a Human Error Analysis Experiment for Soft Control in MCR of Nuclear Power Plant Using Eye Tracking Technology
Authors: Jun-bo Wang , Kun-ze Yang , Yu-xin Zhang , Yoshikawa Hidekazu
Published: October 2017

With the rapid development of digitalized technology, adoption of advanced Main Control Room (RCM) in Nuclear Power Plants (NPPs) has become an inevitable trend. However, the Human Machine Interface (HMI) and Man-machine Interaction (MMI) based on digitalized technologies will also introduce new types of human error and new risks into main control room. Traditional human error analysis is based on post-test interview and questionnaire survey which suffers from difficulty in backtracking the experiment for tracing to source of human errors of operators. This paper presents an experimental research on utilizing an eye tracking device for studying the characteristics of operators during they interact with NPP through HMIs by soft control. The presented approach can also be applied for the statistics of human errors and analysis of causes which will useful for the Probabilistic Safety Assessment (PSA), improvement of HMI design and operator training.
Analysis of Impact on Core Level Monitoring System due to Judgement of Main Coolant Pump Status During LOCA
Authors: Xialan Jiang , Yi Liu , Taili Liu , Xuan Zhang
Published: October 2017

CCMS (Core Cooling Monitoring System) is widely used in current CPR1000 (Improved Chinese PWR) nuclear power plant to calculate core level and subcooling margin in the reactor vessel. In order to validate the effectiveness of core level calculation under condition of LOCA (Loss of Coolant Accident), and quantitatively analyze water level calculation difference of two judgment methods of main coolant pump status in control system, the measuring principles of CCMS are analyzed. The level and error of CCMS calculation under these two conditions is calculated using RELAP5 (Reactor Excursion and Leak Analysis Program) to simulated thermal hydraulic part of CPR1000 unit and virtual DCS (Distributed Control System) to emulate control system. The result shows that CCMS system will introduce large error of water level between calculation value and real level in reactor during LOCA, and operator may be misled by the incorrect information to an improper operation procedure.
Design of a Platform for Safety Justification of Safety-Critical Software
Authors: Jia Guo , Ming Yang
Published: October 2017

Since safety-critical software is crucial to nuclear safety in the occurrence of accident, it is required to have rather higher requirements in both reliability and safety than the non-safety one. However, since the complexity of a software product, how to ensure the reliability and safety of a software product is still a challenging work. The paper presents a design of a platform for safety justification of safety-critical software of nuclear power plants. A syllogism referred as to Claim, Argument and Evidence (CAE) is applied to clarify the key factors that will affect software reliability and the dependencies between them. The proposed safety justification platform offers a user-friendly graphical interface to help construct a CAE model by a drag and drop way. The proposed safety justification platform could be used for the rigorous argument of various factors that may affect the reliability of a safety-critical software product during different phases of its life cycle and establishing their causalities. In this way, it could greatly improve its creditability and applicability and lowering the uncertainties in software development and application, and therefore has a significant engineering values in ensuring and improving the quality and reliability of nuclear software products.
Corrosion Cause Analysis of Electric Feedwater Pump Motor Shaft in Nuclear Power Plants and Relevant Corrective Actions
Authors: Yinhui Che , Jianjun Guan , Shuai Zu , Qiang Chen
Published: October 2017

Electric feedwater pump is an important feedwater equipment of nuclear power plants, and its reliability is directly related to the safe and steady operation of nuclear power plants and also economic benefits. In fact, corrosion of electric feedwater pump motor shaft occurs repeatedly, and even bearing shell in the motor can be burned out happen sometimes. This text sets out to analyze the cause of corrosion of electric feedwater pump motor shaft, identify the root cause, and further work out pertinent corrective actions based on the structure of the feedwater pump.
Root Cause Analysis of Pipeline Rupture in the Condensate Extraction System of Nuclear Power Plant
Authors: Qiang Chen , Ning Han , Weihan Chen , Yinhui Che
Published: October 2017

Pipe rupture in the Condensate Extraction system (CEX) of nuclear power plants will lead to high oxygen content in the secondary circuit and therefore exacerbates equipment and pipeline corrosion. At the same time, there is a risk of loss of vacuum in the system, which has a direct impact on the safe and stable operation of nuclear power plants and will affect the economic benefits. In this paper, Equipment Failure Root Cause Analysis (ERCA) methodology is employed combined with metallography analysis (SEM analysis, XRD) and finite element simulation analysis, to investigate the root cause for drainage pipeline rupture in CEX System of Pressurized Water Reactor (PWR). Detailed analysis process of ERCA was introduced including RCA project establishment, data collection, failure modes analysis and so on. The most probable failure mode is pointed out through the investigation and evidence analyzed. It suggests that the improper design and the installation of limiting orifice plate should be the root cause. And corresponding corrective actions are put forward in details to prevent the recurrence.
Design of the Communication Independence for ACPR1000 Nuclear Power Plant Digital Safety I&C System
Authors: Na Sun , Gui-lian Shi , Yi-qin Xie , Gang Li , Guo-jin Jiang
Published: October 2017

Communication independence is one of the key criteria of digital safety I&C system design. This paper mainly analyzes the requirements for communication independence in safety regulations and standards, and then introduces the architecture and design features, including communication failure processing measures, of communication networks of ACPR1000 nuclear power plant safety digital protection system based on FirmSys platform developed by CTEC. The communication design meets the regulations requirements and effectively improves the safety and reliability of the system, and it is successfully applied in reactor protection system (RPS) of Yang Jiang nuclear power plant unit 5&6. In addition this design can provide reference for communication designs of other NPPs and industries.
Evaluation and Monitoring of Geotechnical Hazards Using UAV Based Digital Photogrammetry
Authors: García García Hugo Alberto , Prieto Rafael , Abello Samuel
Published: October 2017

Unmanned aerial vehicles (UAV), also known as drones, have become a very reliable and convenient tool to many engineering applications. Pipeline corridors are often exposed to geotechnical risk that may interrupt proper service. The risk sources vary, and range from steep hills to many site conditions such as ancient landslides, changes on phreatic water level, creeping soils, or geomorphological changes on the environment. Knowing and understanding the level of risk is highly dependent on identifying any geotechnical hazard in due time, before it materializes. Pipeline corridors are measured in kilometers, and the area that may adversely affect the corridor can vary from a few meters to several hundred on each side of the corridor. This paper presents an application of UAV based digital photogrammetry to evaluate and monitor the hazard and evolution of the site PK 469 OCENSA km 149 ODC. The site is maintained by Oleoducto Central S.A. (OCENSA), and the paper describes the different geotechnical studies and works carried out to maintain uninterrupted service of the pipeline. The paper includes technical information supporting the processing and characteristics of the photogrammetry based evaluation and monitoring method, to highlight the efficiency and accuracy of the proposed method.
Thermo-Mechanical Performance Assessment of Selected Plates From MP-1 High Power Experiments
Authors: Ozaltun Hakan , Rabin Barry H.
Published: October 2017

Monolithic plate-type fuel is a fuel form that is being developed for the conversion of high performance research and test reactors to low-enrichment uranium fuels. These fuel-plates are comprised of a high density, low enrichment, U-Mo alloy based fuel foil encapsulated in an aluminum cladding. To benchmark this new design, number of plates has been irradiated with satisfactory performance. As a part of continuing evaluation efforts, a set of plates covering range of operational parameters is scheduled to be tested during MP-1 irradiation experiments. It is necessary to evaluate the thermo-mechanical performance of plates during irradiation. For this, selected plates with distinct operational histories; covering low power, high power and high fission density were simulated. Fully coupled three-dimensional models of plates with a capability to evolve mechanical and thermal properties of constituent materials with irradiation time and burn-up were developed. The models input used projected parameters, including plate geometry, irradiation history and coolant conditions as input. The model output included temperature, displacement and stresses in the fuel, cladding and diffusion barrier. The fuel behavioral model considered inelastic behavior including volumetric swelling due to solid and gaseous products, irradiation induced creep, thermal expansion, conductivity degradation and plasticity. A visco-plastic behavioral model was used for the cladding that included thermal creep, irradiation hardening, growth due to fast neutrons and Mises plasticity. The plates were then simulated by using projected irradiation parameters. The resulting temperature, displacement and stress-strains were comparatively evaluated on the selected paths. The results were then compared with those of plates from previous RERTR experiments.
Solution of Governing Equations for Six-Field System Code
Authors: Roth Glenn A. , Mesina George L. , Aydogan Fatih
Published: October 2017

Modeling of two phase flows in nuclear power plants is very important for design, licensing, and operator training and therefore must be performed accurately. As requirements have increased, the form and accuracy of the models and computer codes have improved along with them. Early formulations for the field equations include: single phase liquid with algebraic drift flux for the gas phase, modeled with mass, momentum and energy governing equations; and two separate fields, liquid phase and gas phase, typically modeled with six governing equations. These lump bubbles and droplets into the gas and liquid phases respectively and use flow regime maps based upon available experimental data. However, the experiments do not cover the entire spectrum of reactor conditions, so that transitions and extrapolations, which are inherently inaccurate, must be employed. Further, some reactor scenarios, such as boiling and condensation, can be more accurately resolved by modeling bubbles or droplets separately from the continuous fields. Introduction of an additional field, droplet or bubble, apart from the continuous liquid and gas fields, generally uses nine governing equations.

Despite the successful development of the above-mentioned methods for modeling reactor coolant flow in modern software, such as RELAP, TRAC, TRACE, CATHARE and many others, there remain reactor scenarios that require greater resolution to model. This is particularly true of conditions during reflood, where emergency spray flows dominate the cooling profile within the core.

Existing system codes use a lumped approach for two phase flows that groups the fields by their phase, thereby losing track of the physical interactions between the discrete fluid fields. The accuracy of these accident analysis system codes can be improved by characterizing the interactions between additional coolant fields. To capture the effect of the various field interactions, governing equations involving six-fields have been developed. The six fields are 1) continuous liquid, 2) continuous vapor, 3) large droplets, 4) small droplets, 5) large bubbles and 6) small bubbles. The additional fields and the related governing equations introduce additional variables and source terms that require new closure relationships and primary variables. This article presents the equations and variables and develops the discrete set of 18 equations that must be solved to model the system.
Flow Testing and Analysis of the FSP-1 Experiment
Authors: Hawkes Grant L. , Jones Warren F. , Marcum Wade , Weiss Aaron , Howard Trevor
Published: October 2017

The U.S. High Performance Research Reactor conversions fuel development team is focused on developing and qualifying the uranium-molybdenum (U-Mo) alloy monolithic fuel to support conversion of domestic research reactors to low enriched uranium. Several previous irradiations have demonstrated the favorable behavior of the monolithic fuel.

The Full Size Plate 1 (FSP-1) fuel plate experiment will be irradiated in the northeast (NE) flux trap of the Advanced Test Reactor (ATR). This fueled experiment contains six aluminum-clad fuel plates consisting of monolithic U-Mo fuel meat. Flow testing experimentation and hydraulic analysis have been performed on the FSP-1 experiment to be irradiated in the ATR at the Idaho National Laboratory (INL). A flow test experiment mockup of the FSP-1 experiment was completed at Oregon State University. Results of several flow test experiments are compared with analyses. This paper reports and shows hydraulic analyses are nearly identical to the flow test results.

A water channel velocity of 14.0 meters per second is targeted between the fuel plates. Comparisons between FSP-1 measurements and this target will be discussed. This flow rate dominates the flow characteristics of the experiment and model. Separate branch flows have minimal effect on the overall experiment. A square flow orifice was placed to control the flowrate through the experiment. Four different orifices were tested. A pressure differential versus flow rate curve for each orifice is reported herein. Fuel plates with depleted uranium in the fuel meat zone were used in one of the flow tests. This test was performed to evaluate flow test vibration with actual fuel meat densities and reported.
Experimental Study on Heat Transfer to Supercritical CO2 Flowing in Vertical Upward Tube at Medium Mass Flux
Authors: Zhang Qian , Li Huixiong , Kong Xiangfei , Zhang Jun , Lei Xianliang , Zhang Weiqiang
Published: October 2017

An experimental study was performed on heat transfer characteristics of supercritical pressure CO2 (SC-CO2) flowing at medium mass flux conditions in a vertically-upward tube of 16 mm inner diameter at the Heat Transfer and Flow test loop of Supercritical CO2 (HTF-SCO2) in Xi’an Jiaotong University. Experimental parameters included the pressure ranging from 7.5 to 10.5 MPa, the mass flux of 400–600 kg/m2s, and the heat flux of 20–100 kW/m2. Based on the experimental data, effects of mass flux, heat flux and operation pressure on heat transfer characteristics of SC-CO2 were thoroughly discussed. With the decrease of mass flux and increase of heat flux, heat transfer characteristics of SC-CO2 becomes worse and worse. The wall temperature rises to high levels with the occurrence of a wall temperature peak and the wall temperature peak also rises remarkably with the decrease in mass flux and increase in heat flux. Especially, effect of pressures on the heat transfer of SC-CO2 was found to be quite different from that previously reported in literature. When the heat flux is low (such as 30 kW/m2), the HTD was diminished with the increase in pressures, but when the heat flux is up to 50 kW/m2, the HTD is surprisingly intensified by the increase of pressure. The buoyancy effect was considered to explain this distinct influence of pressure on the heat transfer of SC-CO2 by employed a non-dimensional parameter Bu. With the increase of pressure, buoyancy effect was diminished owing to the decrease of density difference between fluids near the wall and the center. When heat flux was lower, the Bu was located between 5×10−6 and 10−4, where buoyancy effect impaired heat transfer, so the heat transfer coefficient increased by rising pressure. But when heat flux was larger, the Bu was above 10−4, where buoyancy effect began to enhance heat transfer, as a result, the heat transfer coefficient was reduced by weakened buoyancy effect because of the increase of pressure. (CSPE)
Generation and elimination of silicon pitting for 300 mm CMOS process technologies
Authors: Pani Santosh Kumar , Hogan Royston Hugh , Pandurangan Madhavan , Zhang Jian , Koesun Pak
Published: October 2017

The complementary metal oxide semiconductor (CMOS) process technology in a 300 mm wafer fab experienced wafer center yield loss. In-line wafer defect inspection revealed gross silicon pitting at the wafer center as the root cause of the yield loss. Affected dies showed pitting at the interfacial corner between the silicon substrate and the isolation oxide. The mechanism of silicon pitting involves creation of silicon microdefects during several high-temperature furnace anneal process steps (liner oxide and posthigh energy phosphorus implant anneals) because of local die and global wafer level thermally induced mechanical stress at the silicon substrate to isolation oxide interfacial corner. In addition, implant-induced silicon microdefects from a high energy (>2 MeV) phosphorus implant, extending to the silicon surface provide a further significant contribution to the silicon microdefect population. The overall microdefect population is further aggravated by Ostwald ripening during a subsequent thick silicon-oxide furnace growth process, resulting in sufficient corner silicon microstructure damage to enhance wet-etching during a subsequent wet-clean leading to gross silicon pitting. Silicon pitting is eliminated by lowering either the liner oxide or postimplant anneal temperatures or skipping the high energy phosphorus implant. Incorporating a reduced liner oxide anneal temperature into the CMOS process flow eliminated the wafer-level yield loss at the wafer center associated with gross silicon pitting defects.
Engineering the Enantioselectivity and Thermostability of a (+)-γ-Lactamase from Microbacterium hydrocarbonoxydans for Kinetic Resolution of Vince Lactam (2-Azabicyclo[2.2.1]hept-5-en-3-one)
Authors: Gao Shuaihua , Zhu Shaozhou , Huang Rong , Li Hongxia , Wang Hao , Zheng Guojun
Published: October 2017

To produce promising biocatalysts, natural enzymes often need to be engineered to increase their catalytic performance. In this study, the enantioselectivity and thermostability of a (+)-γ-lactamase from
Microbacterium hydrocarbonoxydans
as the catalyst in the kinetic resolution of Vince lactam (2-azabicyclo[2.2.1]hept-5-en-3-one) were improved. Enantiomerically pure (−)-Vince lactam is the key synthon in the synthesis of antiviral drugs, such as carbovir and abacavir, which are used to fight against HIV and hepatitis B virus. The work was initialized by using the combinatorial active-site saturation test strategy to engineer the enantioselectivity of the enzyme. The approach resulted in two mutants, Val54Ser and Val54Leu, which catalyzed the hydrolysis of Vince lactam to give (−)-Vince lactam, with 99.2% (enantiomeric ratio [E] > 200) enantiomeric excess (ee) and 99.5% ee (E > 200), respectively. To improve the thermostability of the enzyme, 11 residues with high temperature factors (B-factors) calculated by B-FITTER or high root mean square fluctuation (RMSF) values from the molecular dynamics simulation were selected. Six mutants with increased thermostability were obtained. Finally, the mutants generated with improved enantioselectivity and mutants evolved for enhanced thermostability were combined. Several variants showing (+)-selectivity (E value > 200) and improved thermostability were observed. These engineered enzymes are good candidates to serve as enantioselective catalysts for the preparation of enantiomerically pure Vince lactam.

IMPORTANCE
Enzymatic kinetic resolution of the racemic Vince lactam using (+)-γ-lactamase is the most often utilized means of resolving the enantiomers for the preparation of carbocyclic nucleoside compounds. The efficiency of the native enzymes could be improved by using protein engineering methods, such as directed evolution and rational design. In our study, two properties (enantioselectivity and thermostability) of a γ-lactamase identified from
Microbacterium hydrocarbonoxydans
were tackled using a semirational design. The protein engineering was initialized by combinatorial active-site saturation test to improve the enantioselectivity. At the same time, two strategies were applied to identify mutation candidates to enhance the thermostability based on calculations from both a static (B-FITTER based on the crystal structure) and a dynamic (root mean square fluctuation [RMSF] values based on molecular dynamics simulations) way. After combining the mutants, we successfully obtained the final mutants showing better properties in both properties. The engineered (+)-lactamase could be a candidate for the preparation of (−)-Vince lactam.
Complete Genome Sequences of Four Foot-and-Mouth Disease Viruses of Serotype South African Territories 1 (SAT 1), Topotype X, Isolated from Cattle in Nigeria in 2015
Authors: Vandenbussche Frank , Mathijs Elisabeth , Ularamu Hussaini G. , Ehizibolo David O. , Haegeman Andy , Lefebvre David , Lazarus David D. , Wungak Yiltawe S. , De Vleeschauwer Annebel R. , Van Borm Steven , De Clercq Kris
Published: October 2017

The complete genome sequences of four foot-and-mouth disease viruses of South African territories 1 (SAT 1) serotype are reported. These viruses originate from an outbreak in Nigeria in 2015 and belong to the novel SAT 1 topotype X from the west and central African virus pool.
Draft Genome Sequences of One Methicillin-Sensitive and Seven Methicillin-Resistant Staphylococcus aureus Sequence Type 5 Isolates Obtained in California
Authors: Hau Samantha J. , Bayles Darrell O. , Alt David P. , Nicholson Tracy L.
Published: October 2017

Staphylococcus aureus
is a commensal bacterium of humans that can cause a spectrum of diseases. An isolate’s capacity to cause disease is partially attributed to the acquisition of novel mobile genetic elements. This report provides the draft genome sequence of one methicillin-susceptible and seven methicillin-resistant clinical human
S. aureus
isolates.
Complete Genome Sequence of a Colistin-Resistant Escherichia coli Strain Harboring
mcr-1
on an IncHI2 Plasmid in the United States
Authors: Gilrane Victoria L. , Lobo Stephen , Huang Weihua , Zhuge Jian , Yin Changhong , Chen Donald , Alvarez Kimberly J. , Budhai Alexandra , Nadelman Ilana , Dimitrova Nevenka , Fallon John T. , Wang Guiqing
Published: October 2017

We report here the incidental detection and complete genome sequence of a urinary
Escherichia coli
strain harboring
mcr-1
and resistant to colistin in a New York patient returning from Portugal in 2016. This strain, with sequence type 1485 (ST1485), was a non-extended-spectrum beta-lactamase (ESBL) and non-carbapenemase producer and carried the
mcr-1
gene on an IncHI2 plasmid.
Metagenome of the Microbial Community of Anammox Granules in a Nitritation/Anammox Wastewater Treatment System
Authors: Mardanov Andrey V. , Beletsky Alexey V. , Nikolaev Yury , Kotlyarov Roman Y. , Kallistova Anna , Pimenov Nikolai V. , Ravin Nikolai V.
Published: October 2017

We sequenced the metagenome of a granular sludge in a nitritation/anammox bioreactor used for the treatment of ammonium-rich wastewater.
Proteobacteria
,
Planctomycetes
,
Bacteroidetes
,
Chloroflexi
,
Ignavibacteriae
, and
Acidobacteria
were the predominant phyla in the studied bioreactor. Binning of contigs yielded a near-complete genome of the dominant anammox bacterium assigned to the candidate genus
Brocadia
.
Complete Mitochondrial Genome of the Human Liver Cancer HA22T Cell Line
Authors: Huang Deng-Gao , Cheng Gang , Wu Jin-Cai
Published: October 2017

Liver cancer remains one of the most lethal malignancies worldwide. We sequenced the complete mitochondrial genome of a liver cancer HA22T cell line. The sequenced mitogenome of 16,603 bp will aid in the use of the HA22T cell line for liver cancer study.
Complete Genome Sequence of
Porcine respirovirus 1
Strain USA/MN25890NS/2016, Isolated in the United States
Authors: Park Jie Yeun , Welch Michael , Harmon Karen M. , Zhang Jianqiang , Piñeyro Pablo E. , Li Ganwu , Gauger Phillip C.
Published: October 2017

A porcine parainfluenza virus type 1 (species
Porcine respirovirus 1
) cell culture isolate, USA/MN25890NS/2016, was obtained from porcine nasal swabs, and its complete genome sequence (GenBank accession number MF681710) was determined to help further characterize this virus.