bluuebunny/crossref_metadata_2025_split huggingface dataset:
Effect of bilateral carotid body resection on the counterregulatory response to hypoglycaemia in humans
Authors: Wehrwein Erica A. , Limberg Jacqueline K. , Taylor Jennifer L. , Dube Simmi , Basu Ananda , Basu Rita , Rizza Robert A. , Curry Timothy B. , Joyner Michael J. 
Published: November 2014

New Findings

What is the central question of this study?
Hyperoxia blunts hypoglycaemia counterregulation in healthy adults. We hypothesized that this effect is mediated by the carotid bodies and that: (i) hyperoxia would have no effect on hypoglycaemia counterregulation in carotid body‐resected patients; and (ii) carotid body‐resected patients would exhibit an impaired counterregulatory response to hypoglycaemia.

What is the main finding and its importance?
Our data indicate that the effect of hyperoxia on hypoglycaemic counterregulation is mediated by the carotid bodies. However, a relatively normal counterregulatory response to hypoglycaemia in carotid body‐resected patients highlights: (i) the potential for long‐term adaptations after carotid body resection; and (ii) the importance of redundant mechanisms in mediating hypoglycaemia counterregulation.

Hyperoxia reduces hypoglycaemia counterregulation in healthy adults. We hypothesized that this effect is mediated by the carotid bodies and that: (i) hyperoxia would have no effect on hypoglycaemia counterregulation in patients with bilateral carotid body resection; and (ii) carotid body‐resected patients would exhibit an impaired counterregulatory response to hypoglycaemia. Five patients (three male and two female) with bilateral carotid body resection for glomus tumours underwent two 180 min hyperinsulinaemic, hypoglycaemic (∼3.3 mmol l−1) clamps separated by a minimum of 1 week and randomized to either normoxia (21% fractional inspired O2) or hyperoxia (100% fractional inspired O2). Ten healthy adults (seven male and three female) served as control subjects. Hypoglycaemia counterregulation in carotid body‐resected patients was not significantly altered by hyperoxia (area under the curve expressed as a percentage of the normoxic response: glucose infusion rate, 111 ± 10%; cortisol, 94 ± 6%; glucagon, 107 ± 7%; growth hormone, 92 ± 10%; adrenaline, 89 ± 26%; noradrenaline, 79 ± 15%; main effect of condition, P > 0.05). This is in contrast to previously published results from healthy adults. However, the counterregulatory responses to hypoglycaemia during normoxia were not impaired in carotid body‐resected patients when compared with control subjects (main effect of group, P > 0.05). Our data provide further corroborative evidence that the effect of hyperoxia on hypoglycaemic counterregulation is mediated by the carotid bodies. However, relatively normal counterregulatory responses to hypoglycaemia in carotid body‐resected patients highlight the importance of redundant mechanisms in mediating hypoglycaemia counterregulation.
Upward Flame Propagation Experiments on Hydrogen Combustion in a 220 cub. m Vessel
Authors: Kuznetsov Mike , Stern Gerold , Kljenak Ivo , Matkovič Marko , Mavko Borut 
Published: November 2014

During a hypothetical severe accident in a pressurized water reactor (PWR) of nuclear power plant, hydrogen would be generated during the degradation of the reactor core. The ignition and ensuing combustion of hydrogen could cause a sharp pressure increase, which could threaten the integrity of the containment.
The present work describes the execution and results of the Upward Flame Propagation Experiment (UFPE) on hydrogen combustion within the EC LACOMECO project. Two experiments were performed at the HYKA A2 experimental facility (KIT), which is a cylindrical vessel of 220 m3 free volume, 9.1 m height and 6.0 m internal diameter. A homogeneous mixture of hydrogen (11–12 vol.%), steam (20–21 vol.%) and air was established in the vessel. The initial pressure was 1.5 bar, and the average initial temperature was about 90.0 °C. The mixture was ignited at the bottom of the vessel and the ensuing axial and radial flame propagation were observed. Pressure, temperature and flame velocity were measured at different axial and radial locations. The flame propagation was recorded by a high speed video combined with a Background Oriented Schlieren (BOS) method. The experimental results were finally compared to the results of the well-known THAI HD-22 experiment, which was performed at almost similar conditions.
The objective of current work was to investigate the effect of scale on flame propagation regime in order to clarify how can phenomena observed in a scaled-down experimental facility be extrapolated to an actual containment scale. Another goal was to provide detailed experimental data for numerical code validations in a large scale.
Development of Deep Hole Drilling Technology for Tube Sheet of Steam Generator
Authors: Ueda Shinichiro , Yagishita Takuya , Maeno Jun , Okuda Masanari 
Published: November 2014

Advanced PWR nuclear power plant has very-large steam generators of which tube-sheet is diameter 5000mm / thickness 500 mm / weight 100 ton or larger. Drilling of very thick tube sheet needs deep-hole drilling technique with strict dimensional tolerance control. In addition, the tube sheet has 10,000 or more tube holes, which means that stable drilling operation is mandatory for productivity control. Upon such background, steam generator tube sheet drilling operation needs precise dimensional control technique with highly efficient productivity.
In this study, authors did (1) feasibility-study and development of deep hole drilling machine with the Boring and Trepanning Association (BTA) method, (2) cutting parameter evaluation such as feed rate and cutting velocity, (3) tube hole measurement system development, and (4) drill exchange program development with monitoring drilling machine motor loads and setting alarm level for the loads.
Based on techniques developed by this study, IHI has achieved high-accuracy and stable deep-hole drilling technology for very-thick tube sheets applicable to the very-large steam generators.
On the Prospects of Using Deaeratorless Thermal Schemes With Direct Contact Low-Pressure Reheaters at New Nuclear Power Plants
Authors: Gotovsky M. A. , Ermolov V. F. , Mikhailov V. E. , Sukhorukov Yu G. , Trifonov N. N. 
Published: November 2014

Now in Russia a new NPP “Brest – OD - 300” is developing which is to become a head in a series of fast reactors cooled by liquid lead or lead-bismuth alloy. Russian abbreviation OD is translated as Pilot-Demonstrational. The ideas are developed to use in this power plant deaeratorless thermal circuit. Such the schemes are widely and successfully used for conventional power plants in Russia.
Deaeratorless thermal schemes are based on the use of direct contact low-pressure reheaters. These reheaters have deaeration ability. Such schemes improve power plants efficiency by 1.5 %. Some other deaerator functions are distributed among other elements of plant. In particular, the presence of an external, independent supply of water, which is available at nuclear power plants and much higher than the supply of water in the deaerator.
Danger of return flow exists for direct-contact low pressure reheaters. But its design allows to eliminate completely the possibility of dangerous return flow. Compliance with safety was verified by calculation and in the operated power plants.
Study on the Dynamic Characteristics of the Water-Lubricated Tilting-Pad Radial Bearing Considering Temperature-Viscosity Effect
Authors: Li Yu , Wang Dezhong , Gu Weiguo 
Published: November 2014

One water-lubricated tilting-pad radial bearing is studied in this paper to learn the effect of temperature-viscosity effect caused by the water friction loss on the surface of the pads on the carrying capacity and dynamic characteristics of the bearing. The Reynolds equation considering the turbulent stress term and the energy equation are established based on a simplified model of the tilting-pad radial bearing. The influence of the viscosity change to the dynamic characteristics of the bearing is analyzed while considering the temperature rise in case of water friction loss.
Result shows that the temperature and water friction loss will increase in a large degree when shaft neck eccentricity is large. The carrying capacity of the bearing will decrease when the temperature increases. However, the rise of the temperature will not change the stiffness and damping coefficients. The maximal rise of the temperature reaches 293.15K (20 °C) when shaft neck eccentricity angle is 180°and eccentricity ratio is 0.9 leading to the reduction of bearing capacity by 8%. Thus, to maintain the stability and avoid the obvious temperature rise by water friction, the load force upon the shaft should be less than 280kN when the surrounding temperature is 353.15K (80°C) and shaft neck eccentricity angle is 180° and eccentricity ratio is less than 0.8.
Investigation on Coupling Behaviors of Thermal-Hydraulics/Neutronics Under Asymmetrical Inlet Conditions
Authors: Zheng Yong , Peng Min-jun , Xia Geng-lei , Li Ren 
Published: November 2014

The reactor core is a complex system involving the reactor physics, thermal hydraulics and many other aspects. That means the distribution of the core power largely determines the profile of the thermal parameters, meanwhile the local thermal-hydraulics condition will in turn affect the neutronics calculation by moderator temperature effect and Doppler effects. Issues coupling the thermal-hydraulics with neutronics of nuclear plants still challenge the design, safety and the operation of LWR few years ago. Fortunately, the recent availability of powerful computer and computational techniques has enlarged the capabilities of making more realistic simulations of complex phenomena in NPPs.
The current study deals with the development of an integrated thermal-hydraulics/neutronics model for Qinshan phase II NPP project reactor for the analysis of specific plant transients in which the neutronic response of the core is important, application of RELAP5-HD making use of the Helios code to derive the macroscopic cross-sections. Based on the coupled model, the steady state calculation and the transient simulation, involving the abnormal operation mode with asymmetrical coolant flux and temperature on the inlet of reactor, have been performed. The results show that the values obtained from coupled code RELAP5-HD calculation are in good agreement with the available experimental data, and the calculated accident parameters curves can predict all major trends of the transient. Steady state and transient condition calculation results are in accordance with the theoretical analysis from the aspect of coupled thermal-hydraulics/neutronics, this demonstrated a successful best estimate coupled RELAP5-HD model of Qinshan phase II NPP reactor has been developed, and the established model will provide a good foundation for the further analysis of the primary loop. It also can be concluded that the more accurate CFD method coupling three dimensional neutron kinetics code based on neutron diffusion method are necessary for steady-state calculation and analysis of transient/accident conditions when asymmetrical processes take place in the core. It is worth mentioning that RELAP5-HD code has already programmed the human-machine interface and the interface for coupling with other code, hence RELAP5-HD code has a broad application prospect in PWRs safety analysis.
Coupled Simulations of Natural and Forced Circulation Tests in NACIE Facility Using RELAP5 and ANSYS Fluent Codes
Authors: Martelli D. , Forgione N. , Barone G. , Del Nevo A. , Di Piazza I. , Tarantino M. 
Published: November 2014

In this work the activity performed at the DICI (Dipartimento di Ingegneria Civile e Industriale) of the Pisa University in collaboration with the ENEA Brasimone Research Centre is presented. In particular the document deals with the application of an in-house developed coupling methodology between a modified version of RELAP5/Mod3.3 and Fluent commercial CFD code, to the NACIE (Natural Circulation Experiment) LBE experimental loop (built and located at the ENEA Brasimone research centre).
The first part of the document treats the description of the NACIE loop type facility, while in the second part, the developed coupling tool is presented and the obtained numerical results are compared to stand alone RELAP5 results and to data obtained from the NACIE experimental campaign. The experimental tests are performed varying the argon flow rate and the electric power supplied to the heater and both natural and assisted circulation tests are investigated. The numerical model set-up is based on a two-way explicit coupling scheme and 2D and 3D geometrical domain were investigated.
Comparative analyses among numerical and experimental results showed good agreement, giving positive feedback on the feasibility and capability of the developed coupling methodology.
Fuel Dispersion and Flow Blockage Analyses for MYRRHA-FASTEF Reactor by SIMMER Code
Authors: Bandini Giacomino , Eboli Marica , Forgione Nicola 
Published: November 2014

This work illustrates the 3D set up model and the results concerning the recent analysis of fuel dispersion in the MYRRHA-FASTEF reactor performed with SIMMER code within the EU-FP7 SEARCH Project. Under severe accidental conditions, the release of fuel in the primary system can occur in case of fuel rod clad failure and degradation. Two cases were therefore taken into account, an imposed fuel release to study key parameters which influence the dispersion phenomenon and a coolant flow blockage in a fuel assembly.
The reactor was simulated by a 3D Cartesian geometry with 65×63×42 cell mesh. Steady-state and transient analyses were performed by SIMMER-IV. Steady-state analysis was performed in order to assess the correct operability of the code and of the model. The results were compared with the design values. The most significant results obtained for temperature trends and profiles, velocity and mass flow rate trends are reported. Transient results were also analysed, i.e. fuel dispersion transients were simulated, comparing the effect of fuel porosity on the fuel dispersion inside the pool. In addition, the effects of the release position and the fuel particle dimension on the dispersion phenomenon were also investigated.
The final section of the paper describes the effects of a flow blockage on the core degradation and dispersion of fuel particles in the primary circuit of the MYRRHA reactor. This simulation, with fuel porosity equal to 5%, started after a preliminary steady state condition. The mass flow rate in one of the inner fuel assemblies was then reduced to about 10% of the initial value.
The results show that the SIMMER-IV code is capable of predicting steady-state results in good agreement with the nominal values, also confirming the correctness of the set up model.
To Analyze the Core Damage Scenario via the RISMC-Based Metatheory and Reliability Theory
Authors: Zheng Gangyang , Gong Yu , Zhang Zhijian , Liu Zibin 
Published: November 2014

With “theory of nuclear safety (TONS)”, this paper intends to explain the Core Damage (CD) scenario of a Nuclear Power Plant (NPP) with the systematic methodology, many notions introduced here can be extended to other types of nuclear installations, as well. This systematic methodology combines the Risk-Informed Safety Margin Characterization (RISMC) Metatheory of TONS, and the basic reliability theory. A “metatheory” of such theories, here, is a theory to analyze the Theory of Nuclear Safety (TONS); in its own theory system, it is designed to summarize the safety of a NPP. Meanwhile, the basic reliability theory, which is decided by the authors, is focus on the mission reliability model (a model can be established by Reliability Block Diagram (RBD)); then the related basic concepts, is simple and clear, and quite mature in NPP field.
The present work outlines the traditional reliability theory and the RISMC-based Metatheory, and these two concepts here are taken as the appropriate TONS to analyze the CD Scenario, after that, a renovate or renew TONS, from these two sides, can be introduced to analyze the fundamental safety of NPP.
Preliminary Severe Accident Analysis of Fusion-Fission Hybrid Reactor Using the MELCOR Code
Authors: Dabin Zhang , Zhou Zhiwei , Xie Heng , Yang Tang 
Published: November 2014

The fusion-fission hybrid conceptual reactor is a proposed means of generating power, which adopts a water cooled fission blanket based on ITER. Due to the water cooled fission blanket, safety performance of the hybrid reactor should be considered, including decay heat remove, core uncovered, core meltdown, core degradation, radioactivity releases, etc., similar with the PWRs. The main goal of this work is to develop the fission blanket model by using MELCOR code, and to evaluate the safety performance under severe accidents preliminarily.
Based on MELCOR 1.8.5, some modification is made for the severe accident analysis of fission blanket. Using modified MELCOR code, an analysis model is developed for the fission blanket and the cooling loop. The strategy of the In-Vessel Retention (IVR) using the ex-vessel cooling method is evaluated during a large break LOCA. The calculation results describes the main phenomena during the severe accident progression, including core dry out, meltdown, relocation, etc.. Simulation result is shown that the decay heat in the fission zone can be removed out by the ex-vessel cooling system merely, and the fuel max temperature will not reach the melting point.
HLM Fuel Pin Bundle Characterization in CIRCE Pool Facility
Authors: Tarantino M. , Martelli D. , Di Piazza I. , Forgione N. , Agostini P. , Coccoluto G. 
Published: November 2014

Since the Lead-cooled Fast Reactor (LFR) has been conceptualized in the frame of GEN IV International Forum (GIF), ENEA is strongly involved on the HLM technology development.
Currently ENEA has implemented large competencies and capabilities in the field of HLM thermal-hydraulic, coolant technology, material for high temperature applications, corrosion and material protection, heat transfer and removal, component development and testing, remote maintenance, procedure definition and coolant handling.
In this frame the Integral Circulation Experiment (ICE) test section has been installed into the CIRCE pool facility, and suitable experiments have been carried out aiming to deeply investigate the pool thermal-hydraulic behavior of a HLM cooled pool reactor.
In particular a fuel pin bundle simulator (FPS) has been installed in the CIRCE pool. It has been conceived with a thermal power of about 1 MW and a linear power up to 25kW/m, relevant values for a LMFR. It consist of 37 fuel pins (electrically simulated) placed on a hexagonal lattice with a pitch to diameter ratio of 1.8. The pins have a diameter of 8.2mm and active lengths of 1 m. Along the FPS, three spacer grid properly designed by ENEA have been installed.
The FPS has been deeply instrumented by several thermocouples. In particular three sections of the FPS have been instrumented to monitor the heat transfer coefficient along the bundle as well as the cladding temperature in different rank of sub-channels.
A full characterization of the FPS has been experimentally achieved both under forced and natural circulation, and the main results gained during the run are reported into the paper.
Moreover the paper reports a preliminary analysis and discussion of such results, also in comparison with CFD calculations performed by CFX code.
Computational Fluid Dynamics Modeling of Two-Phase Boiling Flow and Critical Heat Flux
Authors: Tentner Adrian , Merzari Elia , Vegendla Prasad 
Published: November 2014

This paper presents recent advances in the modeling of two-phase boiling flow and critical heat flux that have been implemented in the Extended Boiling Framework (EBF) [1, 2, 3]. The EBF code was developed as a customized module built on the foundation of the commercial Computational Fluid Dynamics (CFD) code STAR-CD, which provides general two-phase flow modeling capabilities, for the detailed analysis of the two-phase flow and heat transfer phenomena that occur in Boiling Water Reactor (BWR) fuel assemblies. These phenomena include coolant phase changes and multiple flow regimes that directly influence the coolant interaction with the fuel pins and, ultimately, the reactor performance. An effort to expand the EBF two-phase models and to explore their applicability to other CFD codes is currently underway.
The paper presents results of recent CFD analyses of Critical Heat Flux (CHF) experiments that have measured the axial distribution of wall temperature in two-phase upward flow in a vertical channel with a heated wall. The experiments were designed to produce the onset of CHF in the upper half of the heated channel. The simulated axial distribution of wall temperature is compared with experimental data, illustrating the ability of the extended EBF model to capture the onset of CHF for a wide range of thermal-hydraulic conditions relevant for BWRs. The paper concludes with a discussion of results and plans for future work.
Experimental Investigation of Dynamic Characteristics of Swing and Tilted Disc Check Valves
Authors: Marcinkiewicz Jerzy , Adamkowski Adam , Lewandowski Mariusz , Janicki Waldemar 
Published: November 2014

The present work describes an experimental investigation of the dynamic characteristics of check valves, which means experimental examination of closing function and ability to generate pressure transients under different flow decelerations in the pipeline. Two designs of check valves are tested: a swing disc and a tilted disc check valve.
The valves are mounted on discharge pipe of a centrifugal pump. The fluid transient is generated by stopping the pump motor from actual velocity to completely stop in a prescribed time.
Each of the check valves is subjected to tests covering different pressure levels in the upper reservoir, initial flow rates in the pipeline, several decelerations of pump rotation, three settings of torque acting on the valve disc, three values of the moment of friction forces acting on the valve axis and finally free fall of the disc in stagnant water and air.
The test stand, the instrumentation and chosen valves as well as scope and conditions for performing the experiments are described. Selected measured results like angular velocities of the discs, pressures in the pipe at different conditions, and volumetric flow rates are presented and discussed.
The dynamic behaviors of the tested valves were compared with each other.