scholarly journals Investigation and Assessment of the CFD for Horizontal Flow in the VHTR Core

2017 ◽  
Vol 2017 ◽  
pp. 1-6
Author(s):  
Yan Shengyuan ◽  
Jean Luc Habiyaremye ◽  
Wei Yingying ◽  
Cong Chi Tran
Keyword(s):  
Nuclear Power ◽  
High Efficiency ◽  
Model Simulation ◽  
Flow Behavior ◽  
Reactor Core ◽  
Field Investigation ◽  
Vital Role ◽  
Working Environment ◽  
Horizontal Flow ◽  
Environment Analysis

A nuclear power station using gas as a cooling medium has attracted so much attention because it offers high efficiency and greater safety. For a nuclear station that operates at a very high temperature, a gas-cooled reactor is fueled by uranium, moderated by graphite, and customarily cooled by helium. Nevertheless, throughout the operation, the bypass flow might be a result of a change in graphite shape that is caused by neutron damage. Core bypass and cross flows are significant elements to consider since the cross gap set hurdles to the flow field that are capable of diverting sufficient amount of coolant from reactor core location and initiating a possible fuel overheating. However, there is a great need to sufficiently validate this method by carrying out a thorough evaluation based on working environment analysis. Comparing the computed results with the existing data from Groehn’s NHDA PMR-200 study was the only way to validate data. A model simulation was performed on a two-prismatic fuel block with a cross gap to examine the gaping size effect. Finally, the prediction methods for horizontal flow phenomena using a CFD technique and the field investigation results from the VHTR core were verified, and the identification of the horizontal flow behavior played a vital role in investigating the coolant velocity and pressure distribution in the horizontal gap.

Experimental and Numerical Simulation on Hydraulic Behavior of Molten Flow in the Lower Part in Reactor Core

2016 ◽  
Author(s):  
Kota Matsuura ◽  
Hideaki Monji ◽  
Susumu Yamashita ◽  
Hiroyuki Yoshida
Keyword(s):  
Numerical Simulation ◽  
Liquid Film ◽  
High Speed ◽  
Nuclear Power ◽  
Power Plants ◽  
Flow Behavior ◽  
Reactor Core ◽  
Video Camera ◽  
Working Fluid ◽  
Simulation Code

In the decommissioning work of nuclear power plants, it is important to grasp the sedimentation place of molten materials. However, the technique to grasp exactly sedimentation place is not established now. Therefore, the detailed and phenomenological numerical simulation code named JUPITER for predicting the molten core behavior is developed. In the study, visualization experiment and numerical simulation were performed to validate the applicability of the JUPITER to the hydraulic relocation behavior in core internals. The test section used in this experiment simulated the structure of the core internals, such as a control rod and a fuel support piece, simply. The working fluid is water under the atmospheric pressure. The experiment uses a high-speed video camera to visualize the flow behavior. The behavior and the speed of the liquid film in a narrow flow channel is obtained. For the numerical analysis carried out prior to the experiment, the behavior of flow down liquid was shown. The typical behavior was also observed that the tip of a liquid film flowing down splits into.

scholarly journals Direct Numerical Simulation of the Flow Through a Randomly Packed Pebble Bed

2020 ◽  
Vol 142 (4) ◽  
Author(s):  
Mustafa Alper Yildiz ◽  
Gerrit Botha ◽  
Haomin Yuan ◽  
Elia Merzari ◽  
Richard C. Kurwitz ◽  
...  
Keyword(s):  
Numerical Simulation ◽  
Direct Numerical Simulation ◽  
Power Systems ◽  
Nuclear Power ◽  
Flow Behavior ◽  
Reactor Core ◽  
Reynolds Numbers ◽  
Fourth Generation ◽  
Flow Through ◽  
Flow Domain

Abstract The proposition for molten salt and high-temperature gas-cooled reactors has increased the focus on the dynamics and physics in randomly packed pebble beds. Research is being conducted on the validity of these designs as a possible contestant for the fourth-generation nuclear power systems. A detailed understanding of the coolant flow behavior is required in order to ensure proper cooling of the reactor core during normal and accident conditions. In order to increase the understanding of the flow through these complex geometries and enhance the accuracy of lower-fidelity modeling, high-fidelity approaches such as direct numerical simulation (DNS) can be utilized. Nek5000, a spectral-element computational fluid dynamics (CFD) code, was used to develop DNS fluid flow data. The flow domain consisted of 147 pebbles enclosed by a bounding wall. In the work presented, the Reynolds numbers ranged from 430 to 1050 based on the pebble diameter and inlet velocity. Characteristics of the flow domain such as volume averaged porosity, axial porosity, and radial porosity were studied and compared with correlations available in the literature. Friction factors from the DNS results for all Reynolds numbers were compared with correlations in the literature. The first- and second-order statistics show good agreement with the available experimental data. Turbulence length scales were analyzed in the flow. Reynolds stress anisotropy was characterized by utilizing invariant analysis. Overall, the results of the analysis in this study provide deeper understanding of the flow behavior and the effect of the wall in packed beds.

scholarly journals Eco-Emission Analysis of Multi-Carrier Microgrid Integrated with Compressed Air and Power-to-Gas Energy Storage Technologies

2021 ◽  
Vol 13 (9) ◽  
pp. 4681
Author(s):  
Khashayar Hamedi ◽  
Shahrbanoo Sadeghi ◽  
Saeed Esfandi ◽  
Mahdi Azimian ◽  
Hessam Golmohamadi
Keyword(s):  
Energy Storage ◽  
Power Systems ◽  
High Efficiency ◽  
Objective Assessment ◽  
Energy Systems ◽  
Vital Role ◽  
Compressed Air ◽  
Emission Analysis ◽  
Multi Objective ◽  
Power To Gas

Growing concerns about global greenhouse gas emissions have led power systems to utilize clean and highly efficient resources. In the meantime, renewable energy plays a vital role in energy prospects worldwide. However, the random nature of these resources has increased the demand for energy storage systems. On the other hand, due to the higher efficiency of multi-energy systems compared to single-energy systems, the development of such systems, which are based on different types of energy carriers, will be more attractive for the utilities. Thus, this paper represents a multi-objective assessment for the operation of a multi-carrier microgrid (MCMG) in the presence of high-efficiency technologies comprising compressed air energy storage (CAES) and power-to-gas (P2G) systems. The objective of the model is to minimize the operation cost and environmental pollution. CAES has a simple-cycle mode operation besides the charging and discharging modes to provide more flexibility in the system. Furthermore, the demand response program is employed in the model to mitigate the peaks. The proposed system participates in both electricity and gas markets to supply the energy requirements. The weighted sum approach and fuzzy-based decision-making are employed to compromise the optimum solutions for conflicting objective functions. The multi-objective model is examined on a sample system, and the results for different cases are discussed. The results show that coupling CAES and P2G systems mitigate the wind power curtailment and minimize the cost and pollution up to 14.2% and 9.6%, respectively.

scholarly journals An Approach to Analyze Diagnosis Errors in Advanced Main Control Room Operations Using the Cause-Based Decision Tree Method

Energies ◽  
2021 ◽  
Vol 14 (13) ◽  
pp. 3832
Author(s):  
Awwal Mohammed Arigi ◽  
Gayoung Park ◽  
Jonghyun Kim
Keyword(s):  
Decision Tree ◽  
Nuclear Power ◽  
Power Plants ◽  
Human Error ◽  
Cognitive Activity ◽  
Emergency Operation ◽  
Nuclear Industry ◽  
Working Environment ◽  
Control Rooms ◽  
Diagnosis Errors

Advancements in the nuclear industry have led to the development of fully digitized main control rooms (MCRs)—often termed advanced MCRs—for newly built nuclear power plants (NPPs). Diagnosis is a major part of the cognitive activity in NPP MCRs. Advanced MCRs are expected to improve the working environment and reduce human error, especially during the diagnosis of unexpected scenarios. However, with the introduction of new types of tasks and errors by digital MCRs, a new method to analyze the diagnosis errors in these new types of MCRs is required. Task analysis for operator diagnosis in an advanced MCR based on emergency operation was performed to determine the error modes. The cause-based decision tree (CBDT) method—originally developed for analog control rooms—was then revised to a modified CBDT (MCBDT) based on the error mode categorizations. This work examines the possible adoption of the MCBDT method for the evaluation of diagnosis errors in advanced MCRs. We have also provided examples of the application of the proposed method to some common human failure events in emergency operations. The results show that with some modifications of the CBDT method, the human reliability in advanced MCRs can be reasonably estimated.

scholarly journals Online Simulation of Nuclear Power Plant Primary Systems

2020 ◽  
Vol 2020 ◽  
pp. 1-9
Author(s):  
Hongyun Xie ◽  
Haixia Gu ◽  
Chao Lu ◽  
Jialin Ping
Keyword(s):  
Information System ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Real Time ◽  
Nuclear Power ◽  
Measurement Data ◽  
Reactor Core ◽  
Power Industry ◽  
Nuclear Power Industry ◽  
Online Simulation

Real-time Simulation (RTS) has long been used in the nuclear power industry for operator training and engineering purposes. And, online simulation (OLS) is based on RTS and with connection to the plant information system to acquire the measurement data in real time for calibrating the simulation models and following plant operation, for the purpose of analyzing plant events and providing indicative signs of malfunctioning. OLS has been applied in certain industries to improve safety and efficiency. However, it is new to the nuclear power industry. A research project was initiated to implement OLS to assist operators in certain critical nuclear power plant (NPP) operations to avoid faulty conditions. OLS models were developed to simulate the reactor core physics and reactor/steam generator thermal hydraulics in real time, with boundary conditions acquired from plant information system, synchronized in real time. The OLS models then were running in parallel with recorded plant events to validate the models, and the results are presented.

scholarly journals Ultrawide-angle and high-efficiency metalens in hexagonal arrangement

2020 ◽  
Vol 10 (1) ◽  
Author(s):  
Chun-Yuan Fan ◽  
Chia-Ping Lin ◽  
Guo-Dung J. Su
Keyword(s):  
High Efficiency ◽  
Numerical Aperture ◽  
Vital Role ◽  
Optical Systems ◽  
Wide Field ◽  
Wide Angle ◽  
Camera Systems ◽  
Single Lens ◽  
Hexagonal Arrangement ◽  
Virtual And Augmented Reality

Abstract Wide-angle optical systems play a vital role in imaging applications and have been researched for many years. In traditional lenses, attaining a wide field of view (FOV) by using a single optical component is difficult because these lenses have crucial aberrations. In this study, we developed a wide-angle metalens with a numerical aperture of 0.25 that provided a diffraction-limited FOV of over 170° for a wavelength of 532 nm without the need for image stitching or multiple lenses. The designed wide-angle metalens is free of aberration and polarization, and its full width of half maximum is close to the diffraction limit at all angles. Moreover, the metalens which is designed through a hexagonal arrangement exhibits higher focusing efficiency at all angles than most-seen square arrangement. The focusing efficiencies are as high as 82% at a normal incident and 45% at an incident of 85°. Compared with traditional optical components, the proposed metalens exhibits higher FOV and provides a more satisfactory image quality because of aberration correction. Because of the advantages of the proposed metalens, which are difficult to achieve for a traditional single lens, it has the potential to be applied in camera systems and virtual and augmented reality.

Flow and Temperature Fields Measurement Inside Rod Bundle by the Combined Use of PIV and LIF Technique

2018 ◽  
Author(s):  
Xing Li ◽  
Sichao Tan ◽  
Zhengpeng Mi ◽  
Peiyao Qi ◽  
Yunlong Huang
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Great Influence ◽  
Reactor Core ◽  
Transfer Characteristic ◽  
Index Of Refraction ◽  
Temperature Fields ◽  
Spacer Grid ◽  
Rod Bundle ◽  
Combined Use

Thermal hydraulic research of reactor core is important in nuclear energy applications, the flow and heat transfer characteristics of coolant in reactor fuel assembly has a great influence on the performance and safety of nuclear power plants. Particle image velocimetry (PIV) and Laser induced fluorescence (LIF) are the instantaneous, non-intrusive, whole-field fluid mechanics measuring method. In this study, the simultaneous measurement of flow field and temperature field for a rod bundle was conducted using PIV and LIF technique. A facility system, utilizing the matching index of refraction approach, has been designed and constructed for the measurement of velocity and temperature in the rod bundle. In order for further study on complex heat and mass transfer characteristic of rod bundle, the single-phase experiments on the heating conditions are performed. One of unique characteristics of the velocity and temperature distribution downstream the spacer grid was obtained. The experimental results show that the combined use of PIV and LIF technique is applied to the measurement of multi-physical field in a rod bundle is feasible, the measuring characteristics of non-intrusive ensured accuracy of whole field data. The whole field experimental data obtained in rod bundle benefits the design of spacer grid geometry.

scholarly journals Squeezing Force of the Magnetorheological Fluid Isolating Damper for Centrifugal Fan in Nuclear Power Plant

2012 ◽  
Vol 2012 ◽  
pp. 1-6 ◽  
Author(s):  
Jin Huang ◽  
Ping Wang ◽  
Guochao Wang
Keyword(s):  
Nuclear Power ◽  
Theoretical Foundation ◽  
Fluid Dynamic ◽  
Flow Behavior ◽  
Magnetorheological Fluid ◽  
Centrifugal Fan ◽  
Parallel Channel ◽  
Mr Fluid ◽  
Mr Fluids ◽  
Control Devices

Magnetorheological (MR) disk-type isolating dampers are the semi-active control devices that use MR fluids to produce controllable squeezing force. In this paper, the analytical endeavor into the fluid dynamic modeling of an MR isolating damper is reported. The velocity and pressure distribution of an MR fluid operating in an axisymmetric squeeze model are analytically solved using a biviscosity constitutive model. Analytical solutions for the flow behavior of MR fluid flowing through the parallel channel are obtained. The equation for the squeezing force is derived to provide the theoretical foundation for the design of the isolating damper. The result shows that with the increase of the applied magnetic field strength, the squeezing force is increased.

Expectations for Inservice Testing Programs at New Nuclear Power Plants

2017 ◽  
Author(s):  
Thomas G. Scarbrough
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Reactor Core ◽  
Nuclear Regulatory Commission ◽  
Lessons Learned ◽  
Mechanical Equipment ◽  
American Society ◽  
Regulatory Commission ◽  
Safety Systems

In a series of Commission papers, the U.S. Nuclear Regulatory Commission (NRC) described its policy for inservice testing (IST) programs to be developed and implemented at nuclear power plants licensed under 10 CFR Part 52. This paper discusses the expectations for IST programs based on those Commission policy papers as applied in the NRC staff review of combined license (COL) applications for new reactors. For example, the design and qualification of pumps, valves, and dynamic restraints through implementation of American Society of Mechanical Engineers (ASME) Standard QME-1-2007, “Qualification of Active Mechanical Equipment Used in Nuclear Power Plants,” as accepted in NRC Regulatory Guide (RG) 1.100 (Revision 3), “Seismic Qualification of Electrical and Active Mechanical Equipment and Functional Qualification of Active Mechanical Equipment for Nuclear Power Plants,” will enable IST activities to assess the operational readiness of those components to perform their intended functions. ASME has updated the Operation and Maintenance of Nuclear Power Plants (OM Code) to improve the IST provisions for pumps, valves, and dynamic restraints that are incorporated by reference in the NRC regulations with applicable conditions. In addition, lessons learned from performance experience and testing of motor-operated valves (MOVs) will be implemented as part of the IST programs together with application of those lessons learned to other power-operated valves (POVs). Licensee programs for the Regulatory Treatment of Non-Safety Systems (RTNSS) will be implemented for components in active nonsafety-related systems that are the first line of defense in new reactors that rely on passive systems to provide reactor core and containment cooling in the event of a plant transient. This paper also discusses the overlapping testing provisions specified in ASME Standard QME-1-2007; plant-specific inspections, tests, analyses, and acceptance criteria; the applicable ASME OM Code as incorporated by reference in the NRC regulations; specific license conditions; and Initial Test Programs as described in the final safety analysis report and applicable RGs. Paper published with permission.

scholarly journals Reactor Core Conceptual Design for a Scalable Heating Experimental Reactor, LUTHER

2021 ◽  
Vol 2 (2) ◽  
pp. 207-214
Author(s):  
Thinh Truong ◽  
Heikki Suikkanen ◽  
Juhani Hyvärinen
Keyword(s):  
Low Temperature ◽  
Conceptual Design ◽  
Nuclear Power ◽  
Water Pressure ◽  
Reactor Core ◽  
Electricity Production ◽  
Design Parameters ◽  
Experimental Reactor ◽  
Fuel Assemblies ◽  
Enriched Uranium

In this paper, the conceptual design and a preliminary study of the LUT Heating Experimental Reactor (LUTHER) for 2 MWth power are presented. Additionally, commercially sized designs for 24 MWth and 120 MWth powers are briefly discussed. LUTHER is a scalable light-water pressure-channel reactor designed to operate at low temperature, low pressure, and low core power density. The LUTHER core utilizes low enriched uranium (LEU) to produce low-temperature output, targeting the district heating demand in Finland. Nuclear power needs to contribute to the decarbonizing of the heating and cooling sector, which is a much more significant greenhouse gas emitter than electricity production in the Nordic countries. The main principle in the development of LUTHER is to simplify the core design and safety systems, which, along with using commercially available reactor components, would lead to lower fabrication costs and enhanced safety. LUTHER also features a unique design with movable individual fuel assembly for reactivity control and burnup compensation. Two-dimensional (2D) and three-dimensional (3D) fuel assemblies and reactor cores are modeled with the Serpent Monte Carlo reactor physics code. Different reactor design parameters and safety configurations are explored and assessed. The preliminary results show an optimal basic core design, a good neutronic performance, and the feasibility of controlling reactivity by moving fuel assemblies.

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