bluuebunny/crossref_metadata_2025_split huggingface dataset:
A New Turbomachine for Clean and Sustainable Hydrocarbon Cracking
Authors: Rubini Dylan , Xu Liping , Rosic Budimir , Johannesdahl Harri 
Published: October 2021

Abstract
Decarbonizing highly energy-intensive industrial processes is imperative if nations are to comply with anthropogenic greenhouse gas emissions targets by 2050. This is a significant challenge for high-temperature industrial processes, such as hydrocarbon cracking, and there have been limited developments thus far. The novel concept presented in this study aims to replace the radiant section of a hydrocarbon cracking plant with a novel turboreactor. This is one of the first major and potentially successful attempts at decarbonizing the petrochemical industry. Rather than using heat from the combustion of natural gas, the novel turboreactor can be driven by an electric motor powered by renewable electricity. Switching the fundamental energy transfer mechanism from surface heat exchange to mechanical energy transfer significantly increases the exergy efficiency of the process. Theoretical analysis and numerical simulations show that the ultrahigh aerodynamic loading rotor is able to impart substantial mechanical energy into the feedstock without excess temperature difference and metal temperature magnitude. A complex shockwave system then transforms the kinetic energy into internal energy over an extremely short distance. The version of the turboreactor developed and presented in this study uses a single rotor row, in which a multistage environment is achieved regeneratively by guiding the flow through a toroidal-shaped vaneless space. This configuration leads to a reduction in reactor volume by more than two orders of magnitude compared with a conventional furnace. A significantly lower wall surface temperature, supersonic gas velocities, and a shorter primary gas path enable a controlled reduction in the residence time for chemical reactions, which optimizes the yield. For the same reasons, the conditions for coke deposition on the turboreactor surfaces are unfavorable, leading to an increase in plant availability. This study demonstrates that the mechanical work input into the feedstock can be dissipated through an intense turbulent mixing process, which maintains an ideal and controlled pressure level for cracking.
Comprehensive Effects of Polymer Flooding on Oil–Water Relative Permeabilities
Authors: Wang Qiong , Jiang Ruizhong 
Published: October 2021

Abstract
Many reports have indicated that, in contrast to waterflooding, oil–water relative permeabilities in polymer flooding are influenced by the polymer properties, such as adsorption and viscoelasticity. Long-term injected polymer solutions flush the reservoir porous media causing wettability alteration and residual oil reduction, which is also an important factor for oil–water relative permeabilities. Therefore, the comprehensive effects of polymer properties and dynamic flushing on two-phase percolation behavior have been the focus of research. In this study, polymer flow experiments, including pre-shearing, relative permeability, and long-term erosion tests, were first conducted. Then, a flushing-interpolation model, including polymer concentration interpolation and surface flux characterization, was developed to dynamically describe the two-phase flow characteristics. Based on the experimental results and the presented dynamic model, polymer-flooding simulations were conducted to analyze the performance of the well and the distribution of the remaining oil. The results indicate that flushing reduces the residual oil saturation, and the viscoelasticity effect further broadens the predominant flushing channels horizontally; both effects lead to higher cumulative oil production and a lower water-cut. Additionally, including the flushing effect in the polymer-flooding model leads to a more reliable numerical prediction of the remaining oil distribution. To account for the comprehensive effects of viscoelasticity and erosion in field applications, water-cut history matching was performed for the studied well group, and the error decreased to 4.39%. Finally, the optimal polymer concentration for high-concentration polymer-flooding programs was determined to be 1450 mg/L.
Water Hammer Simulation in Two-Phase Flow Regimes Using Open Source Code OpenFOAM
Authors: Fuchs Paul , Koch Marco K. 
Published: October 2021

Abstract
Water hammer phenomena in pipe systems are simulated with the open source code OpenFOAM-v7. Effects like dynamic pressure changes caused by fast closing valves, wave propagation, reflection and interference as well as cavitation and fluid structure interaction are focussed in detailed CFD analysis.
In this paper two experiments conducted at the „Cold Water Hammer Test Facility” (CWHTF) at Helmholtz-Zentrum Dresden-Rossendorf, Germany are simulated. The experiments show large steam bubbles collapse causing a pressure wave to travel within the pipe system. In order to capture cavitation effects the model basis of the Volume of Fluid based, fully compressible, multi fluid solver compressibleMultiphaseInterFoam is modified. The implementation of the mass transfer terms as well as the modification of the Kunz cavitation model is described and first simulation results are presented. Furthermore, an analytical approach is implemented to model effects of fluid structure interaction with the surrounding pipe. This approach computes the liquid sonic speed in relation with structural parameters of the deformable pipe in combination with a modified equation of state.
The simulation results show that the bubble collapsing speed is computed in good accordance to the experiment regarding the timing of the initial pressure peak fits the experimental data. However, the amplitude of the pressure peak is overestimated by 22% neglecting any fluid structure interaction effects. Using the analytical approach to calculate the pipe deformation effects the overestimation of the pressure peak reduces to 8%.
Analyses of Wet and Dry Cavity Strategies for BWR Severe Accident Management With MELCOR-2.2
Authors: Takashima Ayato , Yamaji Akifumi , Li Xin , Fujiwara Daisuke , Shirai Hiroshi , Noujuu Takumi 
Published: October 2021

Abstract
The current severe accident management measure for Boiling Water Reactors (BWRs) in Japan adopts pre-flooding of the drywell (D/W) floor when the reactor pressure vessel (RPV) is expected to fail to prevent failure of the primary containment vessel (PCV) (wet cavity strategy). This prioritizes rapid cooling of the corium and is expected to lead to more steam generation at the time of the RPV failure. The purpose of this study is to evaluate the influences of changes in the accident management measure on PCV integrity / failure mode and the amount of radioactive material released to the environment for BWRs. The accident scenario has been tentatively defined as the TQUV scenario, in which the water injection is assumed to fail after the RPV depressurization. A set of MELCOR analysis models, which had been used for the Fukushima accident analyses has been adopted to represent TQUV event of BWR4/Mark-I. The effects of fuel coolant interaction (FCI) has not been considered in the current modeling. The analysis results show that in the case without pre-flooding D/W, the D/W wall temperature exceeded 200°C (which may be regarded as a rough estimate for the PCV failure by overheating) in about 30 minutes after the RPV failure. On the other hand, if water is continuously poured onto the D/W floor after the RPV failure, overheating failure of the PCV can be avoided, but over-pressure failure may occur within a few hours after the RPV failure.
Simulation Research of Combustion Characteristics of Mixed Sodium Fire in a Columnar Flow
Authors: Ma Yao-long , Zhang Zhi-gang , Wu Qi , Wang Fang 
Published: October 2021

Abstract
Sodium fire accident which was caused by the leakage of liquid sodium is the design basis accidents of sodium-cooled fast reactor. The column flow is a more realistic form of liquid sodium leakage due to the pipe system insulation structure and lower pressure in the sodium-cooled fast reactor. The leaked liquid sodium would combust in the form of columnar fire in the space and pool fire on the ground at the same time to form a mixed sodium fire in a columnar flow. In this paper, a numerical calculation model is established to simulate mixed sodium combustion in a columnar flow. This model is based on the turbulence model, discrete phase model and finite-rate model in Fluent software, and combined with chemical kinetic of sodium combustion. Several variables like spatial temperature, oxygen concentration and pressure during combustion of mixed sodium fire are calculated, and validated with experimental data collected in previous mixed sodium combustion experiment. Besides, the oxygen concentration and the temperature field are calculated for further study of the experiment. The result shows that the spatial temperature calculated by the code coincides well with the experiment, but the chamber pressure cannot be predicted precisely. The oxygen concentration in the center of the sodium flow would decline rapidly in the early stage, then remains at around 5%, but the oxygen concentration in the distance is relatively high. This study could provide reference for evaluation and prevention measures of sodium fire.
Investigation and Design of Energy-Absorbing Structure in Nuclear Fuel Cask
Authors: Hao Yuchen , Li Yue , Wang Jinhua , Wu Bin , Ma Tao , Wang Haitao 
Published: October 2021

Abstract
High Temperature Gas Cooled Reactor is considered as one of the most advanced reactors in the GEN-IV reactor system with inherent safety and high generating efficiency. Now, the project of High Temperature Reactor Pebble bed Module (HTR-PM) designed by Tsinghua university is going to finish construction. The first products of nuclear fuel elements have been transported to the Shidao Bay NPP that is ready to provide fuel power for a subsequent grid-connected generation. HTR-PM employs spherical fuel elements instead of common fuel assemblies. They have significant differences in geometry, structure, and mechanical property. The existing transfer cask for fuel assemblies doesn’t seem appropriate anymore. Therefore, it is necessary to find a new cask to meet this demand for nuclear fuel. One of the most important things is the application of lightweight technology which can promote transport efficiency and enhance the economy. Besides, referring to the safety code required by IAEA, the procedure of radioactive material transportation is strictly controlled, hence we should pay attention to both lightweight and safety. Research has focused on the new energy-absorbing structure and performance.
In this paper, we concentrate on cushioning effect of the thin-walled metal box with foam, rather than the normal structure of metal tube filled with foam. This structure will be assembled in one special type of nuclear fuel container to provide energy-absorbing and cushioning effect. The shape of foam specimen is cuboid, and the metal material is attached to it. First of all, we investigate the drop ball test. This study takes two sets of control group. One is a thin-walled box without foam, the other only has foam. Compared with the control group, the cushioning effect of energy-absorbing box is investigated. Secondly, the energy-absorbing structure is tied with the steel block and drop them from 9m height. Then sensitive analysis has been conducted to study the effect of design parameters and multi-work condition, to reveal some empirical regularities. This paper intends to provide important design experiences and references for energy-absorbing structure of nuclear fuel cask.
Development of Liquid-Particle Image Reconstruction Method in Centrifugal Field by Linear Sensor Wireless Electrical Resistance Tomography (LS-WERT)
Authors: Kimura Kota , Prayitno Yosephus Ardean Kurnianto , Sejati Prima Asmara , Zhao Tong , Iso Yoshiyuki , Takei Masahiro 
Published: October 2021

Abstract
A real-time measurement on the radiochemical processes in nuclear industry is crucial in order to monitoring the process frequently. Recently, a wireless electrical resistance tomography (WERT) with circumferential electrode was developed for the real-time measurement. However, due to the limitation in the circumferential section, we propose a linear sensor-wireless electrical resistance tomography (LS-WERT). This system is detecting the particle deposition thickness in the longitudinal positions. A coupling simulation of smoothed particle hydrodynamics and discrete element model (SPH-DEM) method is used to observe the particle-liquid behavior under centrifugal field. The distribution of particle and liquid phase in this simulation is then measured by an electrical impedance tomography (EIT) simulation. SPH-DEM-EIT coupling includes the converting process from DEM to FEM. An Artificial Neural Network (ANN) is applied with input from the electrical measurement results of the coupling. ANN for LS-WERT gives result in three categories of air, liquid, and particle phase. We evaluate the real-time particle deposition thickness by comparing the LS-WERT result to the high-speed camera images. As a result, LS-WERT has an average accuracy of 2.27% under rotating speed below 220 rpm. In overall, LS-WERT gives a good tendency to the high-speed camera images and effective for the application on real-time measurement of high-speed centrifuge.
Risk Analysis of Gasification Process of Nuclear Fuel Manufacturing Facilities Based on FTA
Authors: Yang Xiaowei , Lyu Dan , Que Ji , Liu Yuntao , Zhao Shangui 
Published: October 2021

Abstract
The process of nuclear fuel manufacturing is complicated. The process conditions are strict, and some materials required for the production process are flammable, explosive, chemically toxic and corrosive. As the scale of production increases, the automation of production equipment increases, it is necessary to assess risk of nuclear fuel manufacturing facilities, identify the risk factors of equipment or production processes, analyse hazards, and improve measures, with the objective of protecting staff, the public and environment.
Risk assessment methods mainly include safety checklist analysis, what-if analysis, Hazard and Operability Analysis (HAZOP), Failure Mode and Effects Analysis (FMEA), Fault Tree Analysis (FTA), Event Tree Analysis (ETA), etc. Various analytical methods are applicable to different stages of the engineering development process.[1]
At present, nuclear fuel manufacturing facilities in China do not use risk assessment methods to identify hazards in the design, construction and operation stages. It is only Experience-based engineering judgment and lack of systematicity. In this paper, the gasification chest UF6 leakage is the top event, and the fault tree model is established for the gasification process. Qualitative and preliminary quantitative analysis is carried out to identify the key links that need to strengthen the design. It provides suggestions for improving the reliability of the system and equipment.
Study on Deposition Motion of Naturally Circulating Particulate Matter in Supercritical Water Based on Factor and Correspondence Analysis
Authors: Qi Tian , Zhou Tao , Chen Ning , Chen Juan 
Published: October 2021

Abstract
It is very important to study the deposition of particles in natural circulation of supercritical water to ensure the safe and stable operation of supercritical water reactor. The data of natural circulation loop calculated by ANSYS-CFX simulation software were analyzed by factorial analysis method, and the effects of axial distance, initial particle volume fraction, heating power and particle size on particle deposition were obtained. The results show that the contribution rate of particle size to the deposition rate is the largest, about 36.3%, and the contribution rate of initial particle concentration to the deposition rate is about 15.1%; the interaction between axial distance and heating power is the most obvious, and the interaction effect is the pipe temperature distribution. Through correspondence analysis, the main influencing factors of particle deposition rate at each level were analyzed. The results show that: when the deposition rate is small, the small change of axial distance will also have a greater impact on the deposition of particles; when the deposition rate is further increased, the change of initial particle volume fraction will significantly affect the deposition of particles; when the deposition rate is large, the particle size plays a leading role in the deposition of particles. Both of the two analysis methods show that: in the influence on the deposition of particles in supercritical water natural circulation, the influence degree is particle size &gt; concentration &gt; axial distance &gt; heating power. Based on the two analysis methods, an analysis regression model is established and the volume proportion of particles in natural circulation is predicted.
Numerical Investigation of Safety System Parameters in Molten Salt Reactor: Wall Effect on Freeze Valve Opening Time
Authors: Ilham Muhammad , Aji Indarta Kuncoro , Okawa Tomio 
Published: October 2021

Abstract
Molten Salt Reactor (MSRs) is one of the fourth generation Nuclear Power Plants with better capabilities and potentialities compared to previous generation, the enthusiasm for molten fuel reactor has been increasing around the world. MSRs has passive safety where if the core is overheating cause by accident event, the liquid salt fuel was required to be moved to the safety drain tank underneath the core vessel by gravity force. During this occasion, the freeze valve (FV) that formed in the pipe located between the core and drain tank must be melt out promptly to prevent the vessel to reach it is melting point. In this paper, we conduct on thermal analysis of the freeze valve at the solidification and melting process based on finite elements methods. The enthalpy-porosity method adopted by ANSYS Fluent was used to simulated the designed system at specified condition. The Oak Ridge National Laboratory of Molten Salt Reactor Experiment Freeze valve system was used as a references for parameters investigation. Using pipe wall thickness of 5 mm, 10 mm, and 15 mm to examined the wall effect to thermal properties of the designed freeze valve. The wall pipe for FV systems material was also investigate in order to examine its effect to the opening time. Further, the temperature distributions of the valve system were obtained and analyzed. It was found that the wall effect has significant impact to the solidification and melting process.
Prediction of Iodine Peak and Iodine Purification Time in PWR Nuclear Power Plant With Defective Fuel Rods
Authors: Wang Liang , Liu Zhiyuan , Liu Fei , Zhang Chunming , Liu Fudong , Ye Yuanlv 
Published: October 2021

Abstract
For light water reactor (LWR), if fuel rods failure occurs, the fission products (FP) will be released into the primary coolant, which will affect radiation safety, cause adverse environmental impact and economic loss. When fuel defects occur in nuclear power units, the prediction of iodine peak usually refers to the results of NRC research of the US Nuclear Regulatory Commission: the peak effect during the shutdown can make the activity value of primary loop FP (iodine and RIG) be 10–1000 times of that during normal operation This method is evaluated from a statistical point of view and has little significance for the actual operation of the nuclear power plants. The main reason is that the release rate from the cladding to the primary circuit is not considered in the calculation of the iodine peak purification time. Combined with the actual power operation data of the WWER units, the difference between RIG escape and iodine release is analyzed from the mechanism. Based on the actual power operation data, the calculation model of fission nuclide activity concentration in the primary circuit is established, and a method to predict the iodine peak size and purification time is finally obtained. In the case of fuel rods failure, the peak value of iodine-131 during the shutdown was estimated, and the purification time and final concentration from purification to reactor opening were estimated, which ensured the overhaul progress of the nuclear power plant.
Results of a LES Application to LBE Turbulent Flow in a Wire-Wrapped Single Rod Channel
Authors: Pucciarelli Andrea 
Published: October 2021

Abstract
Innovative Liquid Metal Fast Reactors are among the most promising concepts for the upcoming Nuclear Gen IV; nevertheless, several challenges still have to be overcome before achieving a sufficiently mature understanding of all the involved phenomena. Concerning Thermal-Hydraulics aspects, the correct estimation of the turbulent heat flux contributions and their impact on the predicted heat transfer is still being investigated. Particularly, this becomes relevant when addressing complex geometries, such as the wire-wrapped rod bundles considered for the reactor core, in which simple assumptions may no more be sufficient for a suitable representation of the actual turbulent quantities’ distributions.
The present paper reports on the results of a LES application to a wire-wrapped single rod channel performed adopting the STAR-CCM+ code. The considered fluid is Lead-Bismuth Eutectic Alloy (LBE), often adopted as the working fluid in several experimental facilities built to support the development of Gen. IV LFRs while the considered Reynolds number is in the range of 8600 and the heated length is 0.1 m. The obtained results provide relevant information to be considered for the development and tuning of RANS turbulence models specialized in addressing operating conditions involving liquid-metals: the distributions of the turbulent heat fluxes (THF) is here particularly taken into account.
Codes and Methods Improvements for VVER Comprehensive Safety Assessment: The CAMIVVER H2020 Project
Authors: Verrier Denis , Vezzoni Barbara , Calgaro Barbara , Bernard Olivier , Previti Alberto , Lafaurie Clément , Hashymov Artur , Groudev Pavlin , Stefanova Antoaneta , Zaharieva Neli , Damian Frédéric , Mosca Pietro , Tomatis Daniele , Bieder Ulrich , Willien Adrien , Dos Santos Nicolas , Mercatali Luigi , Sanchez-Espinoza Victor Hugo , Forgione Nicola , Paci Sandro 
Published: October 2021

Abstract
CAMIVVER (Codes And Methods Improvements for VVER comprehensive safety assessment) is a three years project launched under the European Union research program HORIZON 2020. The main aim of the Project is to develop and improve computer codes and methods for VVER comprehensive safety assessment.
VVER reactors constitute a significant and dynamic part of the European energy market. Their safe Long Term Operation (maintenance, refuelling, safety-upgrade, revamping) relies on the industrial use of neutronics and thermal-hydraulics codes and methods that allow studying the behaviour of the plant in normal and accidental conditions.
Computer codes development is strongly required for the following reasons:
1) Current codes and methods used for VVER safety assessment are subjected to growing international export controls from outside EU, threatening the EU sovereignty and security in terms of energy supply.
2) A new generation of innovative codes and methods are under development within Europe. They are improving 3D-multiphysics modelling and uncertainty quantification capabilities and are worth being transferred from lab to industry as they will substantially improve the physics comprehension of PWR and VVER.
3) European codes and methods development for VVER safety assessment will open the VVER market to the European nuclear industry.
CAMIVVER will perform developments required for the new generation codes and will generalize 3D-multiphysics coupling, performing benchmark against current industrial codes used for PWR and VVER safety assessment. CAMIVVER will demonstrate CFD assets and compatibility with uncertainty propagation in the frame of a safety assessment.