Confinement Effects on Added Mass of Cylindrical Structures in a Potential Flow

2017 ◽  
Author(s):  
R. Capanna ◽  
G. Ricciardi ◽  
C. Eloy ◽  
E. Sarrouy
Keyword(s):  
Mechanical Behaviour ◽  
Potential Flow ◽  
Nuclear Power ◽  
Stokes Equations ◽  
Reactor Core ◽  
Axial Flow ◽  
Added Mass ◽  
Seismic Load ◽  
Slender Body Theory ◽  
Computational Costs

Efficient modelling and accurate knowledge of the mechanical behaviour of the reactor core are needed to estimate the effects of seismic excitation on a nuclear power plant. The fuel assemblies (in the reactor core) are subjected to an axial water flow which modifies their dynamical behaviour. Several fluid-structure models simulating the response of the core to a seismic load has been developed in recent years; most of them require high computational costs. The work which is presented here is a first step in order to simplify the fluid forces modelling, and thus to be able to catch the main features of the mechanical behaviour of reactor core with low computational costs. The main assumption made in this work is to consider the fluid flow as an inviscid potential flow. Thus, the flow can be described only using one scalar function (velocity potential) instead of a vector field and strongly simplifies the fluid mechanics equations, avoiding the necessity to solve Navier-Stokes equations. The pressure distribution around a cylinder is first solved in Fourier space for different values of the parameters (wavenumber, confinement size). The method is applied to a simple geometry (cylinder in an axial flow with a variable confinement) in order to test its effectiveness. The empirical model is then compared to simulations and reference works in literature. The configuration with large confinement has been solved, and results were in agreement with Slender Body Theory. The dependency on the confinement size strongly depends on the wavenumber, but in any case added mass increases as the confinement size decreases. Finally, future perspectives to extend the model to a group of cylinders and to improve the model are discussed (i.e. add viscosity to the model).

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 Steady flow separation patterns in a 45 degree junction

2000 ◽  
Vol 411 ◽  
pp. 1-38 ◽  
Author(s):  
C. ROSS ETHIER ◽  
SUJATA PRAKASH ◽  
DAVID A. STEINMAN ◽  
RICHARD L. LEASK ◽  
GREGORY G. COUCH ◽  
...  
Keyword(s):  
Flow Separation ◽  
Stokes Equations ◽  
Bypass Graft ◽  
Three Dimensional ◽  
Symmetry Plane ◽  
Axial Flow ◽  
Reynolds Numbers ◽  
Secondary Flows ◽  
Measured Velocity ◽  
Curved Tube

Numerical and experimental techniques were used to study the physics of flow separation for steady internal flow in a 45° junction geometry, such as that observed between two pipes or between the downstream end of a bypass graft and an artery. The three-dimensional Navier–Stokes equations were solved using a validated finite element code, and complementary experiments were performed using the photochromic dye tracer technique. Inlet Reynolds numbers in the range 250 to 1650 were considered. An adaptive mesh refinement approach was adopted to ensure grid-independent solutions. Good agreement was observed between the numerical results and the experimentally measured velocity fields; however, the wall shear stress agreement was less satisfactory. Just distal to the ‘toe’ of the junction, axial flow separation was observed for all Reynolds numbers greater than 250. Further downstream (approximately 1.3 diameters from the toe), the axial flow again separated for Re [ges ] 450. The location and structure of axial flow separation in this geometry is controlled by secondary flows, which at sufficiently high Re create free stagnation points on the model symmetry plane. In fact, separation in this flow is best explained by a secondary flow boundary layer collision model, analogous to that proposed for flow in the entry region of a curved tube. Novel features of this flow include axial flow separation at modest Re (as compared to flow in a curved tube, where separation occurs only at much higher Re), and the existence and interaction of two distinct three-dimensional separation zones.

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.

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.

A Proposal for Passive Wall Treatment Applied in High Performance Axial Flow Compressors

2020 ◽  
Author(s):  
Rubén Bruno Díaz ◽  
Jesuino Takachi Tomita ◽  
Cleverson Bringhenti ◽  
Francisco Carlos Elizio de Paula ◽  
Luiz Henrique Lindquist Whitacker
Keyword(s):  
Stokes Equations ◽  
Axial Compressor ◽  
Axial Flow ◽  
Tip Clearance ◽  
Leakage Flow ◽  
Smooth Wall ◽  
Tip Leakage Flow ◽  
Viscous Effects ◽  
Stall Margin ◽  
Tip Leakage

Abstract Numerical simulations were carried out with the purpose of investigating the effect of applying circumferential grooves at axial compressor casing passive wall treatment to enhance the stall margin and change the tip leakage flow. The tip leakage flow is pointed out as one of the main contributors to stall inception in axial compressors. Hence, it is of major importance to treat appropriately the flow in this region. Circumferential grooves have shown a good performance in enhancing the stall margin in previous researches by changing the flow path in the tip clearance region. In this work, a passive wall treatment with four circumferential grooves was applied in the transonic axial compressor NASA Rotor 37. Its effect on the axial compressor performance and the flow in the tip clearance region was analyzed and set against the results attained for the smooth wall case. A 2.63% increase in the operational range of the axial compressor running at 100%N, was achieved, when compared with the original smooth wall casing configuration. The grooves installed at compressor casing, causes an increase in the flow entropy generation due to the high viscous effects in this gap region, between the rotor tip surface and casing with grooves. These viscous effects cause a drop in the turbomachine efficiency. For the grooves configurations used in this work, an efficiency drop of 0.7% was observed, compared with the original smooth wall. All the simulations were performed based on 3D turbulent flow calculations using Reynolds Averaged Navier-Stokes equations, and the flow eddy viscosity was determined using the two-equation SST turbulence model. The details of the grooves geometrical dimensions and its implementation are described in the paper.

The analysis of emergency situations related to fires at nuclear power plants

2021 ◽  
Vol 30 (5) ◽  
pp. 66-75
Author(s):  
S. A. Titov ◽  
N. M. Barbin ◽  
A. M. Kobelev
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Nuclear Reactor ◽  
Reactor Core ◽  
Emergency Situations ◽  
The Usa ◽  
Large Fires ◽  
Short Circuits ◽  
Nuclear Reactor Core

Introduction. The article provides a system and statistical analysis of emergency situations associated with fires at nuclear power plants (NPPs) in various countries of the world for the period from 1955 to 2019. The countries, where fires occurred at nuclear power plants, were identified (the USA, Great Britain, Switzerland, the USSR, Germany, Spain, Japan, Russia, India and France). Facilities, exposed to fires, are identified; causes of fires are indicated. The types of reactors where accidents and incidents, accompanied by large fires, have been determined.The analysis of major emergency situations at nuclear power plants accompanied by large fires. During the period from 1955 to 2019, 27 large fires were registered at nuclear power plants in 10 countries. The largest number of major fires was registered in 1984 (three fires), all of them occurred in the USSR. Most frequently, emergency situations occurred at transformers and cable channels — 40 %, nuclear reactor core — 15 %, reactor turbine — 11 %, reactor vessel — 7 %, steam pipeline systems, cooling towers — 7 %. The main causes of fires were technical malfunctions — 33 %, fires caused by the personnel — 30 %, fires due to short circuits — 18 %, due to natural disasters (natural conditions) — 15 % and unknown reasons — 4 %. A greater number of fires were registered at RBMK — 6, VVER — 5, BWR — 3, and PWR — 3 reactors.Conclusions. Having analyzed accidents, involving large fires at nuclear power plants during the period from 1955 to 2019, we come to the conclusion that the largest number of large fires was registered in the USSR. Nonetheless, to ensure safety at all stages of the life cycle of a nuclear power plant, it is necessary to apply such measures that would prevent the occurrence of severe fires and ensure the protection of personnel and the general public from the effects of a radiation accident.

A High Effective Parallel Method for the Coupling Between Neutronics and Thermal-Hydraulic

2016 ◽  
Author(s):  
Xiaomeng Dong ◽  
Zhijian Zhang ◽  
Zhaofei Tian ◽  
Lei Li ◽  
Guangliang Chen
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Large Scale ◽  
Nuclear Reactor ◽  
Reactor Core ◽  
Decisive Role ◽  
Reactor Power ◽  
Coupling Method ◽  
Outlet Temperature

Multi-physics coupling analysis is one of the most important fields among the analysis of nuclear power plant. The basis of multi-physics coupling is the coupling between neutronics and thermal-hydraulic because it plays a decisive role in the computation of reactor power, outlet temperature of the reactor core and pressure of vessel, which determines the economy and security of the nuclear power plant. This paper develops a coupling method which uses OPENFOAM and the REMARK code. OPENFOAM is a 3-dimension CFD open-source code for thermal-hydraulic, and the REMARK code (produced by GSE Systems) is a real-time simulation multi-group core model for neutronics while it solves diffusion equations. Additionally, a coupled computation using these two codes is new and has not been done. The method is tested and verified using data of the QINSHAN Phase II typical nuclear reactor which will have 16 × 121 elements. The coupled code has been modified to adapt unlimited CPUs after parallelization. With the further development and additional testing, this coupling method has the potential to extend to a more large-scale and accurate computation.

Stability of a String in Axial Flow

1985 ◽  
Vol 107 (4) ◽  
pp. 421-425 ◽  
Author(s):  
G. S. Triantafyllou ◽  
C. Chryssostomidis
Keyword(s):  
Stability Analysis ◽  
Frictional Coefficient ◽  
Axial Flow ◽  
Added Mass ◽  
Equation Of Motion ◽  
Bending Stiffness ◽  
Diameter Ratio ◽  
Constant Speed ◽  
Hamilton’S Principle ◽  
The Stability

The equation of motion of a long slender beam submerged in an infinite fluid moving with constant speed is derived using Hamilton’s principle. The upstream end of the beam is pinned and the downstream end is free to move. The resulting equation of motion is then used to perform the stability analysis of a string, i.e., a beam with negligible bending stiffness. It is found that the string is stable if (a) the external tension at the free end exceeds the value of a U2, where a is the “added mass” of the string and U the fluid speed; or (b) the length-over-diameter ratio exceeds the value 2Cf/π, where Cf is the frictional coefficient of the string.

Annular Axial Flow-Induced Vibration of Cylindrical Shells by Flu¨gge’s Shell Theory

2006 ◽  
Author(s):  
Katsuhisa Fujita ◽  
Makoto Katou
Keyword(s):  
Shell Theory ◽  
Stokes Equations ◽  
Cylindrical Shells ◽  
Axial Flow ◽  
Navier Stokes ◽  
Complex Eigenvalue ◽  
Flow Induced Vibration ◽  
Navier Stokes Equations ◽  
Flowing Fluid ◽  
Narrow Passage

The unstable phenomena of thin cylindrical shells subjected to annular axial flow are investigated. In this paper, the analytical model is composed of an elastic axisymmetric shell and a rigid one which are arranged co-axially. Considering the fluid structure interaction between shells and fluid flowing through an annular narrow passage, the coupled equation of motion is derived using Flu¨gge’s shell theory and Navier-Stokes equations. The unstable phenomena of thin cylindrical shells are clarified by using the root locus based on the complex eigenvalue analysis. The numerical parameter studies on the shells with a freely supported end and a rigid one, and with both simply supported ends, are performed taking the dimensins of shells, the characteristics of flowing fluid so on as parameters. The influence of these parameters on the threshold of instability of the coupled vibration between thin cylindrical shells and annular axial flowing fluid are investigated and discussed.

Sign in / Sign up

Export Citation Format

Share Document