bluuebunny/crossref_metadata_2025_split huggingface dataset:
Considerations on a Validation Matrix for System Analysis Codes of Pebble Bed Modular HTRs
Authors: Gou Feng , Chen Fubing , Dong Yujie 
Published: October 2017

The modular High Temperature Reactor (HTR) experienced substantial development over the past two decades in China. At present, the 10 MWth experimental reactor HTR-10 is under operation and the 200 MWe demonstration power plant HTR-PM is in its commissioning phase. Both of the HTR-10 and the HTR-PM are pebble bed HTRs designed by the institute of nuclear and new energy technology (INET) of Tsinghua University.
In the design and analysis of modular HTRs, system analysis codes play an important role in predicting the reactor behaviour under both normal operation and accident conditions. According to China’s nuclear safety regulations, codes and models employed in the HTR system analysis must be verified and validated in conformance with a specified procedure. Thus, a flowchart is recommended in this paper for the validation of system analysis codes. As the first step of the flowchart, a phenomena identification and ranking table (PIRT) process has been performed on the subject of the HTR-PM thermal hydraulics and accident analysis. In this paper, the PIRT process is introduced and the PIRT results are presented. With the purpose of developing a validation matrix, test data have been collected mainly from the operation history and various tests of the HTR-10. On such basis, some considerations on developing a validation matrix are discussed to benefit the code validation work.
Analysis and Application of Trace H2 and CH4 in Cycling Process Gas of Fusion Reactors by Gas Chromatography
Authors: Li-ling Yang , Hong-guang Yang , Zhen-xing Liu , Ti-rui Xia 
Published: October 2017

It’s necessary to establish a high-precision analytical method of Gas Chromatography (GC) for achieving the rapid detection of trace hydrogen and methane in cycling process gas of fusion reactors. The analysis of H2, CH4 with concentrations from 0.5ppm to 5000ppm were detected through the application of Discharge Ionization Detector (DID) in high purity helium and the column of molecular sieve, whose parameters was 15m×0.53mm×50μm.
The studies have shown that the retention time of H2, CH4 were 36s and 71s, respectively, their relatively standard deviation (RSD) of area responses were severally below 1.0% and 0.5%, which illustrated good repeatability. Separately, the detection limit (DL) were no more than 10ppb and 2ppb of H2 and CH4 with high column efficiency, while the column temperature was 40°C and the flow rate was 15mL/min. Meanwhile their values of R2 were all larger than 0.999, which included a better linearity relationship between area response and concentration. This method has already applied to the on-line measurement of molecular sieve adsorption, the analysis of electrochemical hydrogen pump and methane decomposition, etc.
Heat Transfer Characteristics of Downward Facing Hot Horizontal Surfaces Using Mist Jet Impingement
Authors: Yadav Ashutosh Kumar , Sharma Parantak , Sharma Avadhesh Kumar , Modak Mayank , Nirgude Vishal , Sahu Santosh K. 
Published: October 2017

Impinging jet cooling technique has been widely used extensively in various industrial processes, namely, cooling and drying of films and papers, processing of metals and glasses, cooling of gas turbine blades and most recently cooling of various components of electronic devices. Due to high heat removal rate the jet impingement cooling of the hot surfaces is being used in nuclear industries. During the loss of coolant accidents (LOCA) in nuclear power plant, an emergency core cooling system (ECCS) cool the cluster of clad tubes using consisting of fuel rods.
Controlled cooling, as an important procedure of thermal-mechanical control processing technology, is helpful to improve the microstructure and mechanical properties of steel. In industries for heat transfer efficiency and homogeneous cooling performance which usually requires a jet impingement with improved heat transfer capacity and controllability. It provides better cooling in comparison to air. Rapid quenching by water jet, sometimes, may lead to formation of cracks and poor ductility to the quenched surface. Spray and mist jet impingement offers an alternative method to uncontrolled rapid cooling, particularly in steel and electronics industries. Mist jet impingement cooling of downward facing hot surface has not been extensively studied in the literature. The present experimental study analyzes the heat transfer characteristics a 0.15mm thick hot horizontal stainless steel (SS-304) foil using Internal mixing full cone (spray angle 20 deg) mist nozzle from the bottom side. Experiments have been performed for the varied range of water pressure (0.7–4.0 bar) and air pressure (0.4–5.8 bar). The effect of water and air inlet pressures, on the surface heat flux has been examined in this study. The maximum surface heat flux is achieved at stagnation point and is not affected by the change in nozzle to plate distance, Air and Water flow rates.
Hierarchical Modeling of GO-FLOW Models for Online Risk Monitoring of Nuclear Power Plants
Authors: Xinyu Dai , Ming Yang , Bowen Zou , Hongxing Lu 
Published: October 2017

As a clean and safe energy, nuclear power plays an important role in solving the increasingly prominent environmental problems in the world. Researches on the reliability analysis and risk assessment for the safety operation of Nuclear Power Plants (NPPs) have been paid much more attention due to several accidents in the past decades.
GO-FLOW is a methodology for analyzing the reliability of dynamic systems with common cause failures (CCFs), man-machine failures, and time-dependent mission issues. The traditional GO-FLOW methodology presents a set of operators each of which stands for a relatively basic reliability logic. Complex behavior of a physical component may be simply described by one operator or a combination of operators. Therefore the GO-FLOW chart of a system is much more similar to its physical structure diagram than other reliability models such as Fault Tree.
In a previous study by authors, a Living PSA technology based on GO-FLOW was proposed which utilizes a group of interrelated GO-FLOW operators to describe the potential change of states of a physical component between operation, standby, failure, preventive and corrective maintenance. The proposed Living PSA technology is intended to provide a convenient way for the safety engineers to update the PSA models for risk assessment. However, it will make the PSA models complex, quite different from their physical structure diagram and therefore difficult to check.
This paper presents a hierarchical modeling method of GO-FLOW. The GO-FLOW model of a system will be organized at different layers, including system layer, structure layer and component layer. The proposed Living PSA technology is only applied at component layer and different level of layers are interrelated by introduction of new GO-FLOW operators and equivalent relations. Thus, the GO-FLOW models at the higher layers are more similar to their physical structures of system for understanding and checking. A bottom-up calculation algorithm for the reliability analysis is also presented. An Auxiliary Feedwater System (AFWS) of PWR plant is taken as a study case to show to modeling and analyzing a system for reliability and risk assessment based on this new technology.
Development of the Severe Accident Evaluation Method on Second Coolant Leakages From the PHTS in a Loop-Type Sodium-Cooled Fast Reactor
Authors: Yamada Fumiaki , Imaizumi Yuya , Nishimura Masahiro , Fukano Yoshitaka , Arikawa Mitsuhiro 
Published: October 2017

The loss-of-reactor-level (LORL) is a typical accident condition causing a severe accident (SA) in sodium-cooled fast reactors (SFRs). In the loop-type SFR Monju, important cause of the LORL is the second coolant leakage at the low elevation of the primary heat transport system (PHTS), which occurs in cold standby in a different loop from that of the first coolant leakage in rated power operation because of excessive declining of the sodium level. This study developed an evaluation method for the LORL with considering countermeasures to prevent LORL: i.e., pumping sodium up into reactor vessel (RV) and interruption of sodium flowing out by siphon effect.
To evaluate the effectiveness of the countermeasures, the transient behavior in the RV sodium level was analytically studied in representative accident sequences. The representative sequences with lowest sodium level were selected by considering combinations of possible coolant leakage positions.
The evaluation result clarified that the LORL can be prevented by conducting the above-mentioned countermeasures to maintain the RV sodium level sufficient for the operation of decay heat removal system, even after the second coolant leakage of PHTS.
Passive System Reliability Analysis Based on Improved Support Vector Machine
Authors: Ding Hao , Cai Qi , Zhang Yongfa , Jiang Lizhi 
Published: October 2017

The desire for improved safety and simplified designs has led to the increased consideration and rapid development of passive systems. Passive system reliability analysis and evaluation is essential, especially in innovative nuclear reactor designs. Uncertainty related with functional failure should be considered into assessment of passive system reliability based on thermal hydraulics. The low efficiency of thermal hydraulic simulation codes cannot meet the needs of practical application and require a surrogate method to obtain the instant failure probability of passive systems.
In this paper, improved Support Vector Machine (SVM) method with Latin Hypercube Sampling (LHS) is adapted to set an approximation. This method, which can fit the nonlinear elements in thermal hydraulic equation, could determine the failure criterion, a surface classifying operation and failure mode, precisely. Then the calculation of an implicit function is trained and accomplished through the characteristics and features existing in samples. A comparison of polynomial form and neural network method is presented. According to the analysis of parameters effecting the probabilistic assessment, the inputs are adjusted to retrain the implicit function and obtain a better solution.
Status of the JSME Codification Activities on Tornado-Generated Missile Protection for Structures of Nuclear Power Plants
Authors: Shirai Koji , Beppu Masuhiro , Hirai Hideo , Yamada Kazuhiko , Yoshida Shinichi , Shinohara Yohei 
Published: October 2017

In Japan, after the Great East Japan Earthquake, existing nuclear power plants (NPPs) are being re-examined to ensure they can withstand more intense natural phenomena such as earthquakes, tsunamis, volcanic eruptions, forest fires and tornadoes, without any loss of capability to safely maintain their function. One of the representative tornado-generated missiles (tornado missiles) prescribed by the NRA (Nuclear Regular Authority) is 0.2m wide × 0.3m deep × 4.2m long, 135kg in weight with a maximum velocity of 51m/s; a rectangular steel pipe. Several impact design methodologies have been implemented to the safety evaluation during business licensing activities to restart NPPs. However, there is considerable variation among licensees when implementing these impact design methodologies and no clear guidance explaining the requirements necessary to achieve accurate results for the impact design of tornado missiles.
In response, in 2014, the Japan Society of Mechanical Engineers (JSME) formed a Special Task Group (STG) on codification activities for the purpose of developing a guidance document on the impact design of the system, structure and components (SSCs) of the NPPs against tornado missiles. The current guidance undertaken comprises the following five chapters and appendices.
- Section A: This chapter defines the objectives and scope.
- Section B: This chapter outlines the methodology of how to determine the design tornado intensities and design loads against tornado missiles for SSCs.
- Section C: This chapter provides the methodology on how to select representative tornado missile designs from the plant walk-down investigative results and set the design impact velocity based on several vortex models.
- Section D: This chapter notes empirical formulas for the local damage of steel/concrete panels to prevent penetration or scabbing to which tornado missiles are prone. Methods to set design impact loads for hard and soft missiles are also presented, as well as those used to evaluate structural response using an elastic or elastic-plastic single degree of freedom (SDOF) model subjected to impact loads.
- Section E: Modern explicit dynamics finite element computer codes used to design and evaluate SSCs, such as LS-DYNA, AUTODYN, Abaqus/Explicit are sophisticated and robust enough to produce accurate results for tornado missile impact events. This chapter provides guidance on a computational modelling method for explicit dynamics, including the construction of quality finite element models following strain-based acceptance criteria.
- Appendices: as an example of setting tornado missiles and impact velocity, benchmarking of explicit dynamic codes is introduced.
An overview of the contents of this guidance document will be focused on in this paper.
Application of Monte Carlo to PWR Cavity Radiation Streaming Calculation
Authors: Jingru Han , Qiaofeng Liu , Haiying Chen , Chunming Zhang 
Published: October 2017

The cavity streaming is the neutron beam from the reactor core through the tunnel, which is between the external surface of the pressure vessel and the shield inner surface.
Reactor cavity streaming calculation is a typical deep penetration problem with complex geometry. The accurate calculation of neutron radiation streaming is a key problem to the reactor shielding calculation, for which the Monte Carlo method and the discrete ordinate method are two popular methods. The speed of discrete ordinate method calculation is fast, but it is hard to describe the complex pile of cavity; the Monte Carlo method can accurately describe the complex geometry, it has a high calculation precision, but with a low direct simulation efficiency. Based on a pressurized water reactor nuclear power plant, this paper presents a detailed model realized by Monte Carlo code, with continuous energy points cross section libraries. The neutron flux density distribution of PWR reactor cavity streaming can directly be calculated by a three-dimensional simulation. For such an actual deep penetration problem, a variety of variance reduction techniques are studied, an effective variance reduction technique is used to obtain results with small statistic errors for a Monte Carlo simulation, which effectively solves the problem of large-scale deep penetrating convergence difficulty, the cavity radiation streaming calculation and analysis are completed. The result shows that the Monte Carlo method can be used as a powerful tool to solve the problem of cavity streaming leakage.
Development of Experimental and Computational Procedures for Nuclear Power Plant Component Testing Under Flooding Conditions
Authors: Tahhan Antonio , Muchmore Cody , Nichols Larinda , Wells Alison , Roberts Gregory , Ryan Emerald , Suresh Sneha , Bhandari Bishwo , Pope Chad 
Published: October 2017

Idaho State University (ISU), with support from Idaho National Laboratory, is actively engaged in enhancing nuclear power plant risk modeling. The ISU team is significantly increasing the understanding of non-containment, nuclear power plant component performance under flooding conditions. The work involves experimentation activities and development of mathematical models, using data from component flooding experiments. The research consists in developing experimentation procedures that comprised small scale component testing, followed by simple and then complex full scale component testing. The research is taking place in the Component Flooding Evaluation Laboratory (CFEL). Tests in CFEL will include water rise, spray, and wave impact experiments on passive and active components.
Initial development work focused on small scale components, radios and simulated doors, that served as a low-risk and low-cost proof-of-concept options. Following these tests, full-scale component tests were performed in the Portal Evaluation Tank (PET). The PET is a semi-cylindrical 7500-1 capacity steel tank, with an opening to the environment of 2.4 m. × 2.4 m. The opening allows installation of doors, feedthroughs, pipes, or other components. The first set of experiments with the PET were conducted in 2016 using hollow doors subjected to a water rise scenario. Data collected during the door tests is being analyzed using Bayesian regression methods to determine the parameters of influence and inform future experiments.
A practical method of simulating full scale wave impacts on components and structures is also being researched to further enhance CFEL capabilities. Early on, the team determined full scale wave impacts could not be simulated using traditional wave flumes or pools; therefore, closed conduit flow is under consideration. Computational fluid dynamics software is being used to simulate fluid velocities associated with tsunami waves of heights up to 6-m, and to design a wave impact simulation device capable of accurately recreating a near vertical wave section with variable height and fluid velocity.
The component flooding simulation activities associated with this project involve use of smoothed particle dynamics codes. These particle-based simulation methods do not require a mesh to be applied to the fluid, which allows for more natural flows to be simulated.
Finally, CFEL can be described as a pioneering element, comprised of several ongoing research and experimental projects, that are vital to the development of risk analysis methods for the nuclear industry.
Risk Assessment of Hydrocarbon Pipelines Facing Natural Hazards
Authors: Amórtegui Gil José Vicente 
Published: October 2017

Hydrocarbon pipelines are exposed to hazards from natural processes, which may affect their integrity and trigger processes that have consequences on the environment.
Among the natural hazards are the effects of the earthquakes, the neotectonic activity, the volcanism, the weathering of soils and rocks, the landslides, the flows or avalanches of mud or debris, the processes related to sediment transport such as the erosion, the scour by streams, the floods and the sloughing due to rains. Those processes are sometimes related to each other, e.g. the earthquakes can produce slides, or movement of geological faults, or soil liquefaction; the rain can trigger landslides and can cause avalanches and mudslides or debris flow; the volcanic eruptions can originate landslides and avalanches, or pyroclastic flows. Human activities can also induce or accelerate “natural” processes that affect the integrity of the pipelines.
The strength and stiffness of the pipelines allow them to tolerate the effects of natural hazards for some period of time. The amount of time depends on the strength and deformability, the stress state, the age, the conditions of installation and operation of the pipelines and their geometric arrangement with regard to the hazardous processes.
In the programs for pipeline integrity management, the risk is defined as a function that relates the probability of the pipeline rupture and the consequences of the failure. However, some people define risk as the summation of the indicators of probability and consequences, such as a RAM matrix. Others define the risk as the product of the probability of failure times the cost of the consequences, while the overall function used to evaluate the rupture probability of a pipeline facing hazards considered in the ASME b31.8 S standard includes all the elements involved in the failure process. In that standard, for the specific analysis of natural hazards, it is proposed that the function is separated in the two following principal elements: the probability of occurrence of the threatening process (hazard) and the pipeline’s capacity to tolerate it. In this paper a general function is proposed, which is the product of the probability of occurrence of the threatening process, the vulnerability of the pipeline (expressed as the fraction of the potential damage the pipe can undergo), and the consequences of the pipeline failure (represented in the summation of the costs of the spilled product, its collection, the pipeline repair and the damages made by the rupture).
An Initial Approach for Pluviometric Monitoring Along ROWs
Authors: Langone Marcelo Júlio , Piedade Célio Roberto Campos , Oliveira Hudson Régis , Cordeiro Guilherme Licodiedoff , de Paula Henrique Xavier 
Published: October 2017

Employing published information and sharing monitoring apparatus and studies, were the initial solution adopted to reach a rain monitoring process in Transpetro (Petrobras Transporte S.A.) southern ROWs. This type of approach can be easily establish and allows the operator to reach satisfactory results, especially for those that has no controlled process at all, where a small amount of information leads for some result even if it is to clarify the need and the resources for such kind of approach.
Even being quite conservative and bit superficially, the process recently established in Transpetro, has already brought a slight economy and safety. Aiming to show the process flow, this paper does not concentrates in reflecting the trigger point, instead it reflects the experience and general methods applied for checking forecasts and rainfall index.
Prediction of Local Transmission of Mycobacterium tuberculosis Isolates of a Predominantly Beijing Lineage by Use of a Variable-Number Tandem-Repeat Typing Method Incorporating a Consensus Set of Hypervariable Loci
Authors: Murase Yoshiro , Izumi Kiyohiko , Ohkado Akihiro , Aono Akio , Chikamatsu Kinuyo , Yamada Hiroyuki , Igarashi Yuriko , Takaki Akiko , Mitarai Satoshi 
Published: October 2017

ABSTRACT

Strain genotyping based on the variable-number tandem repeat (VNTR) is widely applied for identifying the transmission of
Mycobacterium tuberculosis
. A consensus set of four hypervariable loci (1982, 3232, 3820, and 4120) has been proposed to improve the discrimination of Beijing lineage strains. Herein, we evaluated the utility of these four hypervariable loci for tracing local tuberculosis transmission in 981 cases over a 14-month period in Japan (2010 to 2011). We used six different VNTR systems, with or without the four hypervariable loci. Patient ages and weighted standard distances (a measure of the dispersion of genotype-clustered cases) were used as proxies for estimating local tuberculosis transmission. The highest levels of isolate discrimination were achieved with VNTR systems that incorporated the four hypervariable loci (i.e., the Japan Anti-Tuberculosis Association [JATA]18-VNTR, mycobacterial interspersed repetitive unit [MIRU]28-VNTR, and 24
Beijing
-VNTR). The clustering rates by JATA12-VNTR, MIRU15-VNTR, JATA15-VNTR, JATA18-VNTR, MIRU28-VNTR, and 24
Beijing
-VNTR systems were 52.2%, 51.0%, 39.0%, 24.1%, 23.1%, and 22.0%, respectively. As the discriminative power increased, the median weighted standard distances of the clusters tended to decrease (from 311 to 80 km,
P
< 0.001, Jonckheere-Terpstra trend test). Concurrently, the median ages of patients in the clusters tended to decrease (from 68 to 60 years,
P
< 0.001, Jonckheere-Terpstra trend test). These findings suggest that strain typing using the four hypervariable loci improves the prediction of active local tuberculosis transmission. The four-locus set can therefore contribute to the targeted control of tuberculosis in settings with high prevalence of Beijing lineage strains.