bluuebunny/crossref_metadata_2025_split huggingface dataset:
Effect of Liquid Viscosity on Two-Phase Flow Development in a Vertical Large Diameter Pipe
Authors: Ibrahim Abubakr , Hewakandamby Buddhika , Yang Zhilin , Azzopardi Barry 
Published: October 2018

Effect of injector geometry on two phase flows is of profound importance to industry. If the injection method is found to vary the flow characteristics dramatically, it can be employed to obtain desirable two phase flow regimes/attributes and avoid rather unsought conditions. This could potentially save a lot of costs in the extraction and transportation of oil and gas as well as in many other applications. Moreover, the issue of flow development and dependency on the injection conditions is essential when modelling two phase flows. A lot of experimental data and empirical models may have been developed based on systems that may not be fully developed. Therefore, inaccurate modelling of the physical interactions of the flow gets adopted, and hence large divergence between models and experimental data produced by different researchers often transpires.
Most of the published studies on entrance effect were conducted on air-water or steam-water systems because of their relevance to heat transfer units in the nuclear industry. This paper presents an extensive experimental investigation into the issue of flow development using two approaches; measuring void fraction at five axial stations along the test section, and using different geometries for bubble injection into the base of the pipe. The study focuses on how the entrance effect is influenced by the liquid viscosity. The experiments were conducted in a 127 mm diameter vertical pipe. The investigation is achieved by contrasting 180 runs produced using three different injector geometries, the runs were repeated using 4 different oil viscosities, making 2160 experimental run. Gas superficial velocity (Ugs) was varied between 0.01–5.40 m/s, while liquid superficial velocity (Uls) between 0.07–0.86 m/s. The viscosities investigated span between 4.0 cP up to 104.6 cP. The void fraction was measured using Electrical Capacitance Tomography (ECT) and the Wire Mesh Sensor (WMS). That in addition to differential pressure measurements.
Calculation of Friction Factors and Nusselt Numbers for Twisted Elliptical Tube Heat Exchangers Using Nek5000
Authors: Shaver Dillon R. , Carasik Lane B. , Merzari Elia , Salpeter Nate , Blandford Edward 
Published: October 2018

The development of fluoride salt-cooled high-temperature reactors (FHRs) for nuclear power generation relies on the development of new technologies. Of the potential options being explored, twisted elliptical tube geometries for heat exchanger design are promising based on usage in other industries. They are expected to offer significant enhancement in heat transfer with only a marginal increase in frictional losses. This allows them to be deployed in relatively compact designs that are well suited for FHRs. The presented work focuses on the computational fluid dynamics (CFD) simulations of heated molten salt flows through various twisted elliptical tube geometries at low modified Froude numbers. The objectives of this work are to evaluate the available correlations at lower Froude numbers and to determine the impact of using non-zero tube to tube spacing to resolve contact points or numerical singularities for future CFD simulations efforts The spectral element CFD code Nek5000 was used for all simulations, which were performed in periodic domains of triangular (hexagonal) and square unit cells surrounding a single tube through a complete twist using an explicit filtering large eddy simulation (LES) method. Simulations were used to para-metrically test the effects of tube-to-tube spacing for laminar and turbulent flow regimes on frictional pressure drop and heat transfer. The tested Reynolds numbers covered both laminar flow and fully developed turbulent flow (90 < Re < 12200). The tested SL/dmax ratios cover the range of 1.02 to 1.08 for both unit cell types.
At moderate Reynolds number and comparitively high modified Froude number, excellent agreement for the Nusselt number was observed between simulations and the applicable correlation. As Froude number was decreased towards the bounds of the correlation, the agreement worsened. Cases were then simulated at low Froude number, testing the effects of tube spacing. It was determined that the laminar case for the square unit cell is the most affected by increasing SL/dmax and the gap size should be minimized to mitigate this. Whereas in the triangular unit cell the laminar flow regime is more significantly impacted by increasing SL/dmax compared to the turbulent flow regime which was only marginally impacted.
Characteristics of Large Vortical Structures in the Shallow Wake Flow With a Gap Near the Bed
Authors: Shinneeb Monsif , Balachandar Ram 
Published: October 2018

PIV measurements were made to investigate the turbulent wake flow generated by a vertical sharp-edged flat plate suspended in a shallow channel flow with a gap near the bed. The purpose of this study is to investigate the behaviour of large vortical structures in the wake flow. The investigation focused on the horizontal velocity field in the mid-distance between the bottom bed and the top free surface. Two different gap heights between the channel bed and the bottom edge of the bluff body was studied. These two cases were compared to the no-gap flow case which is considered as a reference case. The Reynolds number based on the water depth was 45,000. The large vortical structures were exposed by analyzing the PIV velocity fields using the proper orthogonal decomposition (POD) method. Only few modes were used for the POD reconstruction of the velocity fields to recover ∼50% of the turbulent kinetic energy. A vortex identification algorithm was then employed to quantify the number, size, circulation, and direction of rotation of the exposed vortices. A statistical analysis of the distribution of number, size, and strength of the identified vortices was carried out to explore the characteristics of the vortical structures.
The results revealed that the number of vortical structures increased as a result of the gap flow with a corresponding decrease in the vortex size and strength. This behaviour is attributed to the production of new vortices and the enhancement of the tearing process.
Investigation of Interaction Between Tip Leakage Flow Generated by Unshrouded Impeller and Diffuser Internal Flow
Authors: Ichinose Asuma , Kimura Norio , Yoshimura Mamiko , Hayashi Tomoyuki , Miyagawa Kazuyoshi 
Published: October 2018

Rocket turbo pumps and industrial pumps such as water feed pumps are required to work under high pressure conditions, therefore low specific speed pumps are needed in spite of high rotational speed. In recent years, unshrouded impellers were used because of easy manufacturing and cost reduction. However, when low specific speed unshrouded impellers are used in such conditions, complex tip leakage flow occur and decrease impeller performance. In addition, splitter blades are often used, the internal flow becomes even more complicated. Therefore, such the internal flow of the unshrouded impeller must be clarified. In this research, we have studied such a centrifugal pump, and we have analyzed the internal flow using experiments and CFD (Computational Fluid Dynamics) simulations.
The experimental verification was carried out by measuring the total pressure distribution on the outlet of the impeller and the diffuser. The unsteady static pressure distribution at the shroud side of the impeller was measured to confirm pump performance. We used two types of CFD simulation to evaluate the internal flow in detail. In the first CFD simulation, the unsteady internal flow of an impeller was evaluated by carrying out DES (Detached Eddy Simulation) with a periodic boundary condition that does not contain the diffuser. In the second CFD simulation, interaction between the impeller leakage flow and the diffuser internal flow was evaluated by DES with the whole impeller and diffuser.
From the experimental verification and CFD simulation, it was confirmed that a large-scale vortex structure caused by the tip leakage flow and the secondary flow was observed in the impeller blade-to-blade. And the influence of the impeller leakage flow on the diffuser internal flow and the diffuser performance was evaluated.
From the above studies, it was confirmed that the tip leakage flow has a large influence on the impeller internal flow and the diffuser performance.
The Numerical and Experimental Investigation Into Hydraulic Characteristics of a No-Load Running Check Valve due to Fluid-Structure Interaction
Authors: Zhang Xiang-yuan , Shuai Zhi-jun , Jiang Chen-xing , Li Wan-you , Jian Jie 
Published: October 2018

Valve is a very important unit in pipeline system. The valve flow fluctuation brings about structural vibration and unpopular noise, and even leads to the safety problems and disasters. In this paper, a special no-load running check valve is investigated. The check valve is structural complex with one inlet and two outlets. It can be simplified as a spring-mass system which manipulates the flow rate by combine action of the ambient pressure of medium and the spring deformation.
The three-dimensional model of the valve is established and also the relationship between pressure drops and flow rate of the valve is obtained in various openings and operating conditions. The structure modals were verified by the field tests and thus its fixing boundaries are obtained correctly.
The mechanism causing self-excited vibration of a piping system is determined using a dynamic model which couples the hydraulics of internal flow with the structural motion of a three-ports passive check valve. The coupling is obtained by making the fluid flow coefficient at the check valve to be a function of valve plug displacement. The results are compared with the experimental data, which verifies the correctness of the theoretical results.
It is shown that the special valve has its own hydraulic characteristics, which greatly influence its flow distribution as it has two outlets. It was also testified that the coupling between fluid and structure changes its natural frequencies and has a non-negligible impact on the pressure fluctuation while working.
Operational Impacts and Consequences of Piping Component Failure: A Review of Operating Experience Data As Recorded in CODAP
Authors: Carr Braedon , Lydell Bengt , Riznic Jovica R. 
Published: October 2018

Water chemistry plays an important part in maintaining corrosion resistance in water transport systems throughout nuclear power plants (NPP’s). Small changes in liquid chemistry such as pH, borate concentration, or build-up of crud in reactor cooling water can result in rapid degradation or damage to components and lead to unexpected failures. The Chemical and Volume Control System (CVCS) and Reactor Water Cleanup System (RWCU) are responsible for maintaining these parameters at appropriate levels, and so failure of either of these systems can result in unnecessary stresses on many other reactor systems due to resulting transients. While the major components of these systems all have sufficient redundancy to prevent major accidents, failure of components in these systems can result in failure of other redundant components and affect plant safety [1]. The CVCS and RWCU systems have experienced aging related degradations and failures in the past, and although they have not affected the system’s emergency functions, they have resulted in unnecessary actuation of related systems, and reactor shutdowns [1]. Reactor shutdowns can result in large changes in reactor coolant chemistry such as oxygen and borate concentration transients, and the build-up of corrosion products which can’t be as easily removed during periods of reactor shutdown [2].
In the following analysis of Component Operational Experience Degradation and Ageing Program (CODAP) experience data; causes, impacts, and preventative actions as recorded in CODAP are examined for degradation events which took place in the CVCS and RWCU, of PWRs and BWRs, respectively. The analysis will demonstrate the usefulness of CODAP in examining reactor component failure trends, as well as discuss insights on improvement for the program.
Thermal Hydraulic Analyses for PTS Evaluation: Comparison of Temperature Fields at RPV Predicted by System TH Code and CFD Code
Authors: Kral Pavel , Vyskocil Ladislav 
Published: October 2018

The paper describes methods used for evaluation of pressurized thermal shock (PTS) at UJV Rez and selected results. The UJV Rez participated on the one hand in a number of international programs focused on development of PTS methodology, benchmark testing etc. On the other hand UJV has worked in several projects evaluating PTS for nuclear power plants in the Czech Republic and abroad. The introductory part of the paper also briefly mentions a development of PTS methodology in the world, its current status and trends.
The core of the paper describes thermal hydraulic part of the PTS evaluation process, which usually contains system thermal hydraulic (TH) analysis and mixing calculation. Already at the system TH calculation level, a simple 2D model of reactor downcomer (DC) is used in UJV Rez analyses — as this approach enables the prediction of 2D temperature and velocity fields in DC and improves the prediction of natural circulation flow or its stagnation in individual loops. For the subsequent mixing calculation the Computational Fluid Dynamics (CFD) computer codes are used nowadays. Comparison of 2D temperature fields from 2D system TH analysis and CFD calculation shows surprisingly good agreement for most cases. The paper shows such comparison for several representative cases and takes a more detailed look into one of them.
Synthesis of a CFD Benchmark for the Thermal Mixing in a Sharp Corner T-Junction With a Wall
Authors: Shams Afaque , Edh Nicolas , Angele Kristian 
Published: October 2018

This article reports a CFD-benchmark with the purpose of validating different turbulence modelling approaches for the transient heat transfer due to mixing of hot and cold flow in a T-junction including the wall. This validation exercise has been carried out within the MOTHER project. In the framework of the project, new experiments were performed with a novel measurement sensor allowing the measurements of the fluctuating wall temperature inside the solid pipe wall. The tests were performed for two different Reynolds numbers (Re) 40000 and 60000 and for two different T-junction geometries; a sharp corner and a round corner. The present article reports the synthesis of the CFD validation for a sharp corner T-junction for Re = 40 000. The CFD validation study has been performed using four different CFD softwares, namely STAR-CCM+, Code_Saturne, LESOCC2 and Fluent. In addition, five different turbulence models i.e. wall-function Large Eddy Simulation (LES), Deatched Eddy Simulation (DES), Partially Resolved Numerical Simulation (PRNS), Unsteady Reynolds Averaged Navier-Stokes URANS and RANS were used to perform the CFD computations.
The validation exercise has shown that LES gives the best agreement with the experimental data followed by hybrid (LES/RANS), URANS and RANS models, respectively. The velocity and the thermal fields in the fluid region are correctly predicted by the proper use of the LES modelling, whereas, the accurate prediction of the thermal field in the solid requires very long sampling time in order to achieve a statistically converged solution, which of course requires an enormous computational power. Therefore, the statistical convergence of the thermal field in the solid has been found to be a bottleneck in order to accurately predict the temperature fluctuations in the wall. However, measuring the small amplitude temperature fluctuations is also associated with an uncertainty so the disagreement between CFD and measurements (of the order of 10 %) can also be attributed, in part, to uncertainties in the measurements.
Design of a Closely-Spaced Rod Bundle for a Reference Direct Numerical Simulation
Authors: Shams Afaque , Kwiatkowski Tomasz 
Published: October 2018

Detailed knowledge of a coolant flow in a fuel assembly of a reactor core has always been a major factor in the design of new nuclear systems. In this regard, traditionally adopted subchannel analysis codes cannot take into account local phenomena, which are quite essential. On the other hand, Computational Fluid Dynamic (CFD) is being recognized as a valuable research tool for thermal-hydraulics phenomenon in the fuel assembly geometries. Because of the high Reynolds number and geometric complexities, the practical CFD calculations are mostly limited to pragmatic Reynolds Averaged Navier-Stokes (RANS) type modelling approaches. A good prediction of the flow and heat transport inside the fuel rod bundle is a challenge for such RANS turbulence models and these models need to be validated. Although the measurement techniques are constantly getting improved, however, the CFD-grade experiments of flow mixing and heat transfer in the subchannel scale are often impossible or quite costly to be performed. In addition, lack of experimental databases makes it impossible to validate and/or calibrate the available RANS turbulence models for certain flow situations. In that context, Direct Numerical Simulation (DNS) can serve as a reference for model development and validation.
The aim of this work is to design a numerical experiment in order to generate a high quality DNS database for a tight lattice bare rod bundle, which will serve as a reference for the validation purpose. The considered geometric design is based on the well-known Hooper experiment, which contains a bare rod bundle with pitch-to-diameter ratio of P/D = 1.107. Performing a DNS computation corresponding to the Hooper experiment requires a huge computational power. Hence, a wide range of unsteady RANS (URANS) study has been performed to scale-down the Reynolds number such that it is feasible for a DNS computation and at the same time it still preserves the main flow characteristics. In addition to the flow field, a parametric study for three different passive scalars is performed to take into account the heat transfer analysis. These passive scalars correspond to the Prandtl numbers of air, water and liquid metal fluids. The heat transfer of these three fluids has been studied in combination with two different boundary conditions at the walls, i.e. a constant temperature and a constant heat flux. Finally, the obtained URANS results are used to compute the Kolmogorov and Batchelor length scales in order to estimate the overall meshing requirements for the targeted DNS.
Nuclear Facility Safety Enhancement Using Sandia National Laboratories’ Computer Codes
Authors: Louie David L. Y. 
Published: October 2018

This paper describes the ongoing study of nuclear facility safety enhancement using Sandia National Laboratories’ (SNL) computer codes, supported by U.S. Department of Energy (DOE) Nuclear Safety Research and Development (NSR&D) Program. Continued DOE NSR&D support, since 2014 has allowed the use of the SNL engineering code suite (SIERRA Mechanics) to further substantiate data in the DOE Handbook published in 1994: DOE-HDBK-3010-94, “Airborne Release Fractions/Rates and Respirable Fractions for Nonreactor Nuclear Facilities.” The use of SIERRA codes allows for a better understanding of the mechanics, dynamics, chemistry and overall physics of airborne release scenarios. SIERRA codes provide insights into the contributing phenomena of source term releases from events such as liquid fires. The 1994 Handbook documents small-scaled, bench-top and limited experiments involving liquid fires, powder spills, pressurized releases, and mechanical insult-induced fragmentation scenarios. Data recorded from these scenarios has been substantiated using SIERRA solid mechanics and fluid mechanics codes.
Data passing among multi-physics SIERRA codes predicted the contaminant release from a drum rupture due to fire even though there is no experimental data available. In the anticipated revision effort of the Handbook by DOE, these computational capabilities could enhance the data in a broader usage and could provide confidence in the safety analysis SIERRA codes can provide the initial source term to be used in the leak path factor (LPF) analyses, which predicts the ST release out of the facility. Typical LPF analysis is done using the MELCOR code, developed at SNL for the U.S. Nuclear Regulatory Commission. Widely used in nuclear reactor applications, MELCOR is a toolbox safety code in the DOE’s central registry for LPF applications. A recent LPF guidance study done by SNL indicated that MELCOR 2.1, along with updated guidance, should replace the obsolete MELCOR 1.8.5 guidance. This new guidance is significantly improved over the previous guidance, utilizing extensive MELCOR validation, including applicable reactor experiments and experiments described in the DOE-HDBK-3010-94 for LPF applications. The latest version of MELCOR should be included in DOE’s central registry, and should be used by safety analysts for LPF analyses.
Study on Description of Plant Status at Fukushima Accident by Emergency Action Level
Authors: Nagashima Kazufumi , Yasuda Nakahiro 
Published: October 2018

This paper aims at verifying the current Japanese Emergency Response Guideline, especially the notification scheme of Emergency Action Level (EAL), through the analysis of the progress of Fukushima nuclear accident.
We compared timing and emergency classification between two data sets of the plant statuses which expressed by the old prediction-based notification and the latest EAL-based notification, in order to assess the current EAL scheme along the effectiveness of protective action for the local residents.
We observed that the plant statuses expressed by the current EAL-based notification gave more engineering insights in the earliest accident phase. We also identified a potential improvement area of the guideline in the following severe accident management (SAM) phase after the trigger of first precautionary action, where we are required to reduce uncertainties in both processes of the operator’s notification and the government’s decision-making, in order to compensate for the abstention of utilizing predictive information.