bluuebunny/crossref_metadata_2025_split huggingface dataset:
Numerical Analysis of Single-Phase Thermal Hydraulic Parameters Along Nanostructured Coating Film
Authors: Al-Yahia Omar S. , Addad Yacine , Yoon Ho Joon , Cho Sung Oh 
Published: October 2020

Abstract
In typical pressurized water reactors, zirconium alloys are used as cladding material for the fuel. However, zircalloy is known to face problems with the high temperature steam, due to the chemical process of oxidation, the oxygen molecules will be separated from the water molecules of the coolant leading to hydrogen gas releases. Recently, a research team at KAIST, South Korea suggested a methodology to fabricate nanoporous oxide layer with the aim of preventing the zircalloy outer surface from reacting with the coolant. Although, this new proposal offers a better solution to prevent the potential hydrogen gas generation, it is still not well understood how the nanoporous-layer is going to affect the convective heat transfer rates between the coolant and the fuel. In fact, on one hand the low conductivity of the oxide layer is expected to reduce the conduction heat transfer within the cladding material; but on the other hand, the nanopores on the oxide layer might act as an effective surface roughness, hence affecting both the hydrodynamic and thermal fields within the coolant channels. In this study, a CFD analysis is carried out to investigate the influence of this nanoporous layer on the convective heat transfer rate and pressure drop coefficient. A detailed 2-D steady-state numerical analysis on single-phase model is performed using Star-CCM+ code. The study is conducted using pores with a diameter of 30 to 100 nm. The results obtained from these predictions are then compared with the ones obtained in the case of the smooth surface. Therefore, the main objectives of the present study are to examine the effect of this nanopourous layer on the thermal hydraulic parameters and to produce the corresponding correlations to be used in the system scale thermal-hydraulic codes.
Simulation Analysis on the Pressurizer Degassing Sub-System of Module Small Reactor (ACP100)
Authors: Zhiyun Cai , Yun Ren , Jianyong Lai , Yulong Zhang , Hang Liu , Xianghong Liu , Minghao Liu , Zhu Ye 
Published: October 2020

Abstract
A pressurizer degassing sub-system is designed in Module Small Reactor (ACP100), which can be used to vent the non-condensable gas dissolved in Reactor Coolant during plant shutdown period by opening the remote degassing valve installed on top of the pressurizer. In order to verify the transient characteristic during the whole degassing process, a dedicated transient degassing program is developed based on the improved non-equilibrium multi-region model and the transient degassing model, which is used then to carry out a simulation analysis on this process. Also, the boundary condition models including pressurizer heater power model, pressure spray model, surge line flow model and critical flow model for degassing valve are established. The simulation results show that, under ultimate condition of hot-zero power operation, during the whole degassing process, the Pressurizer’s pressure drops to a maximum 0.038MPa, and the water level is raised 0.0637m above the normal level, which means, both of the pressure and water level are within the normal operation band and shall not initiate any safety signal. Meanwhile, the whole transient process lasts about 16 minutes, and then it enters into the stable degassing period. The gas concentration vs time curve is very smooth. It takes about 5.8 hours to remove the gas dissolved in Reactor Coolant from 35 ml(STP)/kg to 15 ml(STP)/kg. The analysis shows the Pressurizer Degassing Sub-system is safety, effectively and reliable.
Development of a Simplified Model for Aerosol Removal by Spray System of Containment
Authors: Huiyu Yu , Haifeng Gu , Junyan Chen , Zhongning Sun 
Published: October 2020

Abstract
The containment spray system is of great importance for the nuclear power plant to reduce the pressure and temperature, as well as concentration of radioactive aerosols released from reactor core under the severe accident. The main aim of present study is to develop a simplified model for aerosol removal by spray system, the model is employed to predict the variation of aerosol concentration in containment and reveal the key mechanism influencing aerosol removal. In the present model, the air flow entrained by spraying droplet is considered which is computed by solving simplified one-dimension momentum equations rather than the standard N-S equation. For the validation of the model, a small size sprinkle experiment facility is built. The facility is equipped with non-intrusive instruments such as particle size spectrometer for aerosol spectrum measurement, HELOS/R for measuring the droplet spectrum. The result of aerosol removal of the present model is compared with the result of the full coverage model and the experimental result. In the experiment, polydisperse aerosol is used and the removal rate of aerosols with different sizes is compared against the result of the present model and the full coverage model. The velocity of entrained gas flow and the distribution of droplets are displayed. The computational result reveals the aerosol removal constant distribution along the height.
Study on Heating Process of Dehumidifying Experiment in HTGR
Authors: Shen Kaiyue , Zheng Wei , Ma Shengchao , Yin Huaqiang , He Xuedong , Ma Tao 
Published: October 2020

Abstract
A large number of carbon materials are used in high temperature gas-cooled reactor (HTGR). As a kind of porous material, the carbon material contains a certain amount of moisture and other impurities. In order to reduce the corrosion of internal material in reactor core of HTGR, the initial core or post-accident core must be strictly heated and dehumidified. The current primary circuit heating mainly relies on the rotation of the primary pump to convert the kinetic energy into thermal energy. Obviously, the current scheme was flawed: (1) Due to the insufficient heat generated by rotation of the primary pump, the temperature rising process of the primary circuit is sluggish; (2) The rotation of the primary pump converts the kinetic energy into thermal energy of the helium, at the meantime, the primary circuit dissipates heat outward. For the above reasons, it is difficult to achieve the desired dehumidification temperature in the heating process. While in this paper, an additional thermal source will be added to the steam generator to heat the primary circuit in a new scheme. A proper flow and heat-transfer model of heating the primary circuit in high-temperature reactor was established based on software COMSOL Multiphysics. The numerical analysis of the primary circuit heating process provides rewarding guidance for the selection of the dehumidification scheme in HTGR.
Experimental Study on Gravity Driven Discharging of Quasi-Two-Dimensional Pebble Bed Based on Mathematical Morphology
Authors: Liu Yujia , Peng Sifan , Gui Nan , Yang Xingtuan , Tu Jiyuan , Jiang Shengyao 
Published: October 2020

Abstract
The pebbles flow is a fundamental issue for both academic investigation and engineering application in reactor core design and safety analysis. In general, experimental methods including spiral X-ray tomography and refractive index matched scanning technique (RIMS) are applied to obtain the identification of particles’ positions within a three-dimensional pebble bed. However, none of the above methods can perform global bed particles’ position identification in a dynamically discharging pebble bed, and the corresponding experimental equipment is difficult to access due to the complication and high expense.
In this research, the experimental study is conducted to observe the gravity driven discharging process in the quasi two-dimensional silos by making use of the high-speed camera and the uniform backlight. A mathematical morphology-based method is applied to the pre-processing of the captured results. After being increased the gray value gradient by the threshold segmentation, the edges of the particles are identified and smoothed by the Sobel algorithm and the morphological opening operation. The particle centroid coordinates are identified according to the Hough circle transformation of the edges. For the whole pebble bed, the self-programmed process has a particle recognition accuracy of more than 99% and a particle centroid position deviation of less than 3%, which can accurately obtain the physical positions of all particles in the entire dynamically discharge process. By analyzing the position evolution of individual particles in consecutive images, velocity field and motion events of particles are observed. The discharging profiles of 5 conditions with different exit are analyzed in this experiment. The results make a contribution to improving the understanding of the mechanism of pebbles flow in nuclear engineering.
Evaluation of Decay Dose Rates for European DEMO Vacuum Vessel Structure
Authors: Breidokaitė Simona , Tidikas Andrius , Stankūnas Gediminas 
Published: October 2020

Abstract
Four types of breeder blanket modules are in consideration for European DEMO pre-conceptual design phase specified by European Fusion Roadmap. In this study two designs (Helium Cooled Pebble Bed (HCPB) and Water-Cooled Lithium Lead (WCLL)) were investigated in terms of their influence on vacuum vessel (VV) nuclear decay dose rates. In both cases, the same vacuum vessel structure is examined. VV model consists of two shells, interspace and thermal shield. The inner shell and outer shell of vacuum vessel are made of Stainless Steel 316L(N)-IG. The interspace layer is made of water for cooling and shielding purposes. Stainless steel constitutes around 60% of all material in vacuum vessel and water makes up the rest. The model of fusion reactor has been developed in the frame of EUROfusion’s Power Plant Physics and Technology (PPPT) programme. This paper presents information about calculation results of the dose rates caused by the neutron irradiation for the vacuum vessel structural elements in DEMO fusion power reactor.
In order to perform dose rate calculations, neutron energy spectra are required for different VV segments. Those calculations were performed with MCNP6 code and Joint Evaluated Fission and Fusion (JEFF) 3.2 nuclear data library. Dose rates were calculated with FISPACT-2010 and European Activation File (EAF) 2010 nuclear data library. The dose rates were calculated for European DEMO nuclear fusion reaction after its 20 calendar years lasting operation. Cooling period was considered spanning for up to 1000 years. In addition, radionuclides with contribution of at least 1% to the total value of dose rate were identified.
Highest dose rates are attributed to 56Mn radionuclide immediately after the operation until around one day.60Co is the most dominant isotope during intermediate and long cooling periods. Differences in Interspace layer occur due to 16N isotope.
Experimental Study on Flow Patterns and Pressure Drop of Decaying Swirling Gas-Liquid Flow in a Vertical Pipe
Authors: Zhang Jiarong , Liu Li , Liu Shuai , Gu Hanyang 
Published: October 2020

Abstract
Vertical swirling gas-liquid flow is a kind of complex two-phase flow containing a nonzero tangential velocity component in engineering applications. The accurate flow regime characterization, phase distribution information and pressure drop data about vertical swirling flow are the basis for the optimization of steam generator (SG), which can greatly reduce the cost and improve the safety of nuclear plants. To get these key parameters of swirling vertical flow, we have made a comprehensive visualization experiment in a vertical pipe with 30mm diameter and 5m length by high-speed camera. The experimental pipe is separated into swirling part and non-swirling part. We have set three observation section with different vertical heights in the swirling part. Changing the flow rate of water and gas, different swirling flow pattern photos can be captured by high-speed camera. Based on the photos of different positions and image-processing MATLAB code, we can get three flow regime maps and figure out the decaying law of swirling gas-liquid flow. The pressure drop can be recorded by rotameter at each position. The decaying law of pressrure drop can be concluded from it. These data can be a guide for designing gas-liquid separator in SG to improve the efficiency of nuclear plant.
Seismic Time History Data Precision and Time Interval Requirement
Authors: Li Dali 
Published: October 2020

Abstract
This paper provides the seismic time history data precision and time interval requirement for seismic dynamic analysis. U.S.NRC SRP 3.7.1 “Seismic Design Parameters” Acceptance Criteria for Design Time Histories specifies the power spectral density Nyquist Frequency, time interval, and total duration; however, it does not have the requirement for Response Spectra. The response spectrum bandwidth is inverse-proportional to time interval of the time history. For the time interval of 0.005 seconds, the bandwidth for the response spectrum is between 0.194 Hz and 80.5 Hz; the PSD Nyquist frequency is 100 Hz. For 20.48 seconds time history, 4096 data points are required. The response spectrum between 1.28 Hz and 13.6 Hz has the peak flat magnitude value; the magnitude drops to 0.707 of the peak value from 1.28 Hz to 0.194 Hz and from 13.6 Hz to 80.5 Hz. This paper also provides the time interval requirement for various response spectrum peak flat magnitude value; i.e., the response spectrum highest flat magnitude of 27.2 Hz requires a time interval of 0.0025 seconds time history. For 20.48 seconds time history, 8192 data points are required. For CSDRS, the time interval of 0.005 seconds is adequate for the frequency range of interest between 0.36 Hz and 57.2 Hz. For HRHF, the time interval of 0.0025 seconds is required to analyze the frequency range of interest between 0.36 Hz and 114.4 Hz.
Analysis of Water Retention in Isothermal Vacuum Drying Test
Authors: Salazar Alex , Pulido Ramon J. M. , Lindgren Eric R. , Durbin Samuel G. 
Published: October 2020

Abstract
Validation of the extent of water removal in a dry storage system using an industrial vacuum drying procedure is needed. Water remaining in casks upon completion of vacuum drying can lead to cladding corrosion, embrittlement, and breaching, as well as fuel degradation. In order to address the lack of time-dependent industrial drying data, this study employs a vacuum drying procedure to evaluate the efficiency of water removal over time in a scaled system. Isothermal conditions are imposed to generate baseline pressure and moisture data for comparison to future tests under heated conditions. A pressure vessel was constructed to allow for the emplacement of controlled quantities of water and connections to a pumping system and instrumentation. Measurements of pressure and moisture content were obtained over time during sequential vacuum hold points, where the vacuum flow rate was throttled to draw pressures from 100 torr down to 0.7 torr. The pressure rebound, dew point, and water content were observed to eventually diminish with increasingly lower hold points, indicating a reduction in retained water.
Experimental Investigations on Thermal Stratification in a Large Pool of Water With Immersed Isolation Condenser
Authors: Kumar Sunil , Vijayan P. K. , Grover R. B. , Kannan U. , Yadav Harekrishna , Agrawal A. 
Published: October 2020

Abstract
Advanced nuclear reactors use large pool of water inventory with an immersed heat exchanger to remove decay heat especially in the case of Station Black Out (SBO). The isolation condenser (IC) immersed in Gravity Driven Water Pool (GDWP) of Advanced Heavy Water Reactor (AHWR) is an example of such systems. Heat rejected by the IC is absorbed in the pool. As a result, water density decreases and moves towards the free surface of pool causing layers of hot water piling up over colder one giving rise to stratified water inventory. consequently, the pool at the free surface starts boiling before the grace period (7 days). In the present paper, thermal stratification has been modeled in a power to volume scaled experimental setup. The study is focused on investigating the effect of heater orientation on suppression of thermal stratification in the pool for both the cases of with and without shrouds around heat exchanger.
Comparison of the Behavior of Different Jet Configurations in a Crossflow
Authors: Lewandowski Michael , Kristo Paul , Weiss Abdullah , Kimber Mark 
Published: October 2020

Abstract
In this study, the well understood single round jet in crossflow is compared to an array of three jets aligned on the central axis of the crossflow, all of which are distinct with independent temperature and flow control. The benefit of this study is to better understand how an array of jets mix with a crossflow in turbulent conditions in an effort to further validate computation models for a variety of multiple-jet applications. The test section involves jets issuing vertically downward into a cross flow provided by a low speed wind tunnel. The jets are spaced two diameters apart, providing for an upstream, middle, and downstream jet. The hydraulic interactions are tracked via a stereoscopic particle image velocimetry (S-PIV) system. Several reference measurements and the uncertainty of the results are discussed to aid analogous computational fluid dynamics (CFD) models in the future. First order statistics between the single and triple jet cases with fixed flow rates are compared. Temporal analysis yields dominant frequencies at distinct regions within the flow. Insights into flow control of multiple jets in a cross flow is discussed. The experimental data sets compare the classical single jet in cross flow to the extension of multiple jets. Several differences in the flow behavior are found. The data sets serve as benchmark cases for future CFD models that will aim to replicate these flow types in different real-world applications such as coolant flows in a gas cooled reactor.
Application of AHP - Fuzzy Evaluation Method in Nuclear Power Plant Site Selection
Authors: Sun Yujie , Chen Changzhi 
Published: October 2020

Abstract
In order to make a comprehensive evaluation of many influencing factors on the site selection of nuclear power plants, and to make a quantitative judgment on the rank of candidate sites, a method of site selection based on analytic hierarchy process (AHP) and fuzzy evaluation is proposed. Firstly, based on the characteristics of design and research institute, the factors influencing the site selection of nuclear power plants are analyzed. The indicator of criterion layer and index layer influencing the site selection of nuclear power plant are determined based on specialty, and the decision-making system of site selection scheme is established. Secondly, based on AHP operation rules, the weights of system indicator are calculated. Thirdly, by determining the evaluation set, the evaluation matrix based on fuzzy theory was established, and the evaluation level of candidate sites was determined based on the principle of maximum membership degree. Finally, a practical nuclear power plant site selection project is used for theoretical verification. The candidate sites of nuclear power plants are ranked by the first and second level fuzzy comprehensive evaluation. The results show that the quantitative ranking of site based on AHP and fuzzy evaluation is consistent with the previous qualitative ranking of the project. The site selection method based on AHP -fuzzy evaluation is efficient.
Hybrid Nodal Integral/Finite Element Method (NI-FEM) for Time-Dependent Convection Diffusion Equation
Authors: Namala Sundar , Uddin Rizwan 
Published: October 2020

Abstract
Nodal integral methods (NIM) are a class of efficient coarse mesh method that use transverse averaging to reduce the governing partial differential equation(s) (PDE) into a set of ordinary differential equations (ODE), and these ODEs or their approximations are analytically solved. Since this method depends on transverse averaging, the standard application of this approach gets restricted to domains that have boundaries that are parallel to one of the coordinate axes (2D) or coordinate planes (3D). The hybrid nodal-integral/finite-element method (NI-FEM) has been developed to extend the application of NIM to arbitrary domains. NI-FEM is based on the idea that the interior region and the regions with boundaries parallel to the coordinate axes (2D) or coordinate planes (3D) can be solved using NIM and the rest of the domain can be solved using FEM. The crux of the hybrid NI-FEM is in developing interfacial conditions at the common interfaces between the regions solved by the NIM and the FEM. Since the discrete variables in the two numerical approaches are different, this requires special treatment of the discrete quantities on the interface between the two different types of discretized elements. We here report the development of hybrid NI-FEM in a parallel framework in Fortran using PETSc for the time-dependent convection-diffusion equation (CDE) in arbitrary domains. Numerical solutions are compared with exact solutions, and the scheme is shown to be second order accurate in both space and time. The order of approximations used for the development of the scheme are also shown to be second order. The hybrid method is efficient compared to standalone conventional numerical schemes like FEM.
Heat Transfer and Fluid Flow Characteristic of One Side Heated Vertical Rectangular Channel That Inserted Thin Metallic Wire
Authors: Suga Gota , Takeda Tetsuaki 
Published: October 2020

Abstract
A Very High Temperature Reactor (VHTR) is one of the next generation nuclear systems. From a view point of safety characteristics, a passive cooling system should be designed as the best way of a reactor vessel cooling system (VCS) in the VHTR. Therefore, the gas cooling system with natural circulation is considered as a candidate for the VCS of the VHTR. Japan Atomic Energy Agency (JAEA) is advancing the technology development of the VHTR and is now pursuing design and development of commercial systems such as the 300MWe gas turbine high temperature reactor GTHTR300C (Gas Turbine High Temperature Reactor 300 for Cogeneration). In the VCS of the GTHTR300C, many rectangular flow channels are formed around the reactor pressure vessel (RPV), and a cooling panel utilizing natural convection of air has been proposed. In order to apply the proposed panel to the VCS of the GTHTR300C, it is necessary to clarify the heat transfer and flow characteristics of the proposed channel in the cooling panel. Thus, we carried out an experiment to investigate heat transfer and fluid flow characteristics by natural convection in a vertical rectangular channel heated on one side. Experiments were also carried out to investigate the heat transfer and fluid flow characteristics by natural convection when a porous material with high porosity is inserted into the channel. An experimental apparatus is a vertical rectangular flow channel with a square cross section in which one surface is heated by a rubber heater. Dimensions of the experimental apparatus is 600 mm in height and 50 mm on one side of the square cross section. Air was used as a working fluid and fine copper wire (diameter: 0.5 mm) was used as a porous material. The temperature of the wall surface and gas in the channel were measured by K type thermocouples. We measured the outlet flow rate by hot-wire anemometer which is an omnidirectional spherical probe of diameter 2.5mm. The experiment has been carried out under the condition that a copper wire with a scourer model and a cubic lattice model were inserting into the channel.