Effect of Water Height for Modal Characteristics of the Fuel Channel and Stiffness Analysis Using Modal Test Results

2007 ◽  
Vol 353-358 ◽  
pp. 186-189
Author(s):  
Young Shin Lee ◽  
Hyun Soo Kim ◽  
Young Jin Choi ◽  
Jae Hoon Kim ◽  
Jeong Sik Yim
Keyword(s):  
Nuclear Reactor ◽  
Cooling Water ◽  
Test Results ◽  
Analysis Model ◽  
Fuel Channel ◽  
Modal Characteristics ◽  
Modal Test ◽  
Effect Of Water ◽  
Stiffness Characteristics ◽  
Water Height

A fuel channel, which is the major structure of a nuclear reactor, is excited by the flow of cooling water during the operation of the nuclear reactor. This flow of cooling water can cause excessive vibration of the structure by the resonance. So, in the design process of a nuclear structure, the exact evaluation of the effect of water to modal characteristics and stiffness characteristics is very important to generate an exact analysis model. In this study, the effect of water height for modal characterisics of the fuel channel is investigated and the stiffness evaluation of that is conducted using the modal test results.

Optimization of Mixture Formation of Diesel Fuel in a Small CI Engine: The Effect of Water (H2O) Injection Into Intake Port

2017 ◽  
Author(s):  
Se Hun Min ◽  
Jeonghyun Park ◽  
Hyun Kyu Suh
Keyword(s):  
Water Injection ◽  
Injection Timing ◽  
Intake Port ◽  
Cylinder Pressure ◽  
Test Results ◽  
Analysis Model ◽  
Simultaneous Reduction ◽  
Effect Of Water ◽  
Rate Of Heat Release ◽  
Ci Engine

The objective of this study is to investigate the effect of water injection into intake port on the performance of small CI engine. The ECFM-3Z model was applied for the combustion analysis model, and the amount of injected water were varied 10%, 20% and 30% of injected fuel mass. The results of this work were compared in terms of cylinder pressure, rate of heat release (ROHR), and the ISNO and soot emissions. It was found that the cylinder pressure was decreased from 1.2% to 9.2% when the amount of injected water was increased from 10% to 30%. In the results, NO emission significantly decreased from about 24% to about 85% when the amount of injected water increased due to the specific heat and latent heat of water. Considering the test results, the best conditions for the simultaneous reduction of NO and soot is the BTDC 05deg of injection timing and 30% of water injection mass. It can be expected the best IMEP and ISFC characteristics.

An Experimental Investigation on the Behaviour of Pressure Tube Under Symmetrical and Asymmetrical Heating Conditions in an Indian PHWR

2011 ◽  
Author(s):  
Ashwini K. Yadav ◽  
Ravi Kumar ◽  
Akhilesh Gupta ◽  
P. Majumdhar ◽  
B. Chatterjee ◽  
...  
Keyword(s):  
Thermal Stresses ◽  
Nuclear Reactor ◽  
Water System ◽  
Cooling Water ◽  
Pressure Tube ◽  
Temperature Profiles ◽  
Fuel Channel ◽  
Loss Of Coolant Accident ◽  
Heavy Water Reactor ◽  
Nuclear Reactor Safety

Thermal behavior of fuel channel under loss of coolant accident (LOCA) is a major concern for nuclear reactor safety. LOCA along with failure of emergency cooling water system (ECC) may leads to mechanical deformations like sagging, ballooning or even release of containment in open atmosphere due to breaching of pressure tube (PT) under certain depressurization and voiding rates. In order to understand the phenomenon an experiment has been carried out using 19 pin fuel element simulator. Main purpose of the experiment was to trace temperature profiles over the pressure tube, calandria tube and clad tubes of Indian Pressurized Heavy Water Reactor (IPHWR) under symmetrical and asymmetrical heat-up conditions. For simulating the fully voided scenario, symmetrical heating of pressure was carried out by injecting 13.2 KW (2% of nominal power) to all the 19 pins and the temperatures of pressure tube, calandria tube and clad tubes were measured. During symmetrical heating the sagging of fuel channel was initiated at 460 °C and the highest temperature attained by PT was 650 °C. The decay heat from clad tubes was dissipated to moderator mainly by radiation and natural convection. The highest temperature of 680 °C was observed over the outer ring of clad tubes of fuel simulator. Again, to simulate partially voided condition, asymmetrical heating of pressure was carried out by supplying 8.0 kW power to upper 8 pins of fuel simulator and temperature profiles were measured. Along the circumference of pressure tube (PT) the highest temperature difference of 320 °C was observed, which highlights the magnitude of thermal stresses and their role in breaching of pressure tube under partially voided conditions. However, the integrity of pressure tube was intact during both symmetrical and asymmetrical heat-up conditions.

Coupled Analysis of Thermal Hydraulics and Neutronics for a Molten Salt Reactor

2017 ◽  
Author(s):  
Jian Ge ◽  
Dalin Zhang ◽  
Wenxi Tian ◽  
Suizheng Qiu ◽  
G. H. Su
Keyword(s):  
Molten Salt ◽  
Nuclear Reactor ◽  
Coupled Analysis ◽  
Molten Salt Reactor ◽  
Thermal Hydraulics ◽  
Delayed Neutron ◽  
Analysis Model ◽  
Liquid Salt ◽  
Energy Equations ◽  
Volume Method

As one of the six selected optional innovative nuclear reactor in the generation IV International Forum (GIF), the Molten Salt Reactor (MSR) adopts liquid salt as nuclear fuel and coolant, which makes the characteristics of thermal hydraulics and neutronics strongly intertwined. Coupling analysis of neutronics and thermal hydraulics has received considerable attention in recent years. In this paper, a new coupling method is introduced based on the Finite Volume Method (FVM), which is widely used in the Computational Fluid Dynamics (CFD) methodology. Neutron diffusion equations and delayed neutron precursors balance equations are discretized and solved by the commercial CFD package FLUENT, along with continuity, momentum and energy equations simultaneously. A Temporal And Spatial Neutronics Analysis Model (TASNAM) is developed using the User Defined Functions (UDF) and User Defined Scalar (UDS) in FLUENT. A neutronics benchmark is adopted to demonstrate the solution capability for neutronics problems using the method above. Furthermore, a steady state coupled analysis of neutronics and thermal hydraulics for the Molten Salt Advanced Reactor Transmuter (MOSART) is performed. Two groups of neutrons and six groups of delayed neutron precursors are adopted. Distributions of the liquid salt velocity, temperature, neutron flux and delayed neutron precursors in the core are obtained and analyzed. This work can provide some valuable information for the design and research of MSRs.

scholarly journals Experimental study of heat transfer through cooling water circuit in a reactor vault by using Al2O3 nano fluid

2018 ◽  
Vol 22 (2) ◽  
pp. 1149-1161 ◽  
Author(s):  
Maria Anish ◽  
Balakrishnan Kanimozh
Keyword(s):  
Heat Transfer ◽  
Temperature Distribution ◽  
Base Fluid ◽  
Nuclear Reactor ◽  
Characteristic Curve ◽  
Reactor Core ◽  
Cooling Water ◽  
Heat Transfer Process ◽  
Nano Fluid ◽  
Al2o3 Nanoparticle

The heat produced in the nuclear reactor due to fission reaction must be kept in control or else it will damage the components in the reactor core. Nuclear plants are using water for the operation dissipation of heat. Instead, some chemical substances which have higher heat transfer coefficient and high thermal conductivity. This experiment aims to find out how efficiently a nanofluid can dissipate heat from the reactor vault. The most commonly used nanofluid is Al2O3 nanoparticle with water or ethylene as base fluid. The Al2O3 has good thermal property and it is easily available. In addition, it can be stabilized in various PH levels. The nanofluid is fed into the reactor?s coolant circuit. The various temperature distribution leads to different characteristic curve that occurs on various valve condition leading to a detailed study on how temperature distribution carries throughout the cooling circuit. As a combination of Al2O3 as a nanoparticle and therminol 55 as base fluid are used for the heat transfer process. The Al2O3 nanoparticle is mixed in therminol 55 at 0.05 vol.% concentration. Numerical analysis on the reactor vault model was carried out by using ABAQUS and the experimental results were compared with numerical results.

scholarly journals Seismic bracing performance of plasterboard timber walls

2019 ◽  
Vol 52 (2) ◽  
pp. 56-66 ◽  
Author(s):  
Angela Liu ◽  
David Carradine
Keyword(s):  
New Zealand ◽  
Seismic Performance ◽  
Residential Buildings ◽  
Three Dimensional ◽  
Test Results ◽  
Analysis Model ◽  
Wall Model ◽  
Seismic Engineering ◽  
Displacement Capacity ◽  
Insight Into

The goal of this study is to develop a racking model of plasterboard-sheathed timber walls as part of the efforts towards performance-based seismic engineering of low-rise light timber-framed (LTF) residential buildings in New Zealand. Residential buildings in New Zealand are primarily stand-alone low-rise LTF buildings, and their bracing elements are commonly plasterboard-sheathed LTF walls. It is an essential part of performance-based seismic designs of LTF buildings to be able to simulate the racking performance of plasterboard walls. In this study, racking test results of 12 plasterboard walls were collected and studied to gain insight into the seismic performance of plasterboard-sheathed LTF walls. The racking performance of these walls was examined in terms of stiffness/strength degradation, displacement capacity, superposition applicability and failure mechanisms. Subsequently, a mathematical analysis model for simulating racking performance of LTF plasterboard walls is developed and presented. The developed racking model is a closed-form wall model and could be easily used for conducting three-dimensional non-linear push-over studies of seismic performance of LTF buildings.

Effect of Neutron Irradiation on the Mechanical Properties of an A508 CL2 and 15Kh2NMFA Irradiated in the NOMAD_3 Rig in the BR2 Cooling Water

2021 ◽  
Author(s):  
Inge Uytdenhouwen ◽  
Rachid Chaouadi
Keyword(s):  
Neutron Irradiation ◽  
Nuclear Reactor ◽  
Cooling Water ◽  
Charpy Impact ◽  
Tensile Tests ◽  
Significant Loss ◽  
Irradiation Temperature ◽  
Irradiation Damage ◽  
Transition Curves ◽  
Reactor Pressure

Abstract The typical operating temperatures of a nuclear reactor pressure vessel in a PWR are between 290°C and 300°C. However, many BWRs and some PWRs operate at slightly lower temperatures down to 260°C. Most of the literature and neutron irradiation damage is therefore focused on those irradiation temperatures. It is well-known that the lower the irradiation temperature, the more neutron irradiation damage occurs, because no appreciable annealing happens below approximately 230°C. The NOMAD_3 irradiation consisted in total of 24 Charpy sized samples from an A508 Cl.2 forging and a 15Kh2NMFA material. They were irradiated to three various fluences between 1.55 and 7.90 × 1019 n/cm2 (E > 1MeV) at approximately 100°C. The hardening of the A508 Cl.2 was between 260 and 400 MPa which was much higher than the NOMAD_0 properties which were irradiated at approximately 280°C. The tensile tests of irradiated materials are all characterized by a significant loss of work hardening capacity leading to plastic flow localization promptly after the yield strength is reached. This affects also the shape of the Charpy impact transition curves. The radiation embrittlement derived from Charpy impact tests, ΔT41J, is up to 156°C for the highest fluence. For this irradiation, the embrittlement to hardening ratio was also around 0.43 +/−0.2°C/MPa as it was found in the previous campaign NOMAD_0. This paper discusses the tensile, hardness and impact properties of the NOMAD_3 irradiation campaign. It is compared to the NOMAD_0 with respect to effect of irradiation temperature and annealing recovery.

An effective method for damage assessment based on limited measured locations in skeletal structures

2020 ◽  
Vol 24 (1) ◽  
pp. 183-195 ◽  
Author(s):  
Parsa Ghannadi ◽  
Seyed Sina Kourehli
Keyword(s):  
Neural Network ◽  
Artificial Neural Network ◽  
Damage Detection ◽  
Degrees Of Freedom ◽  
Truss Structures ◽  
Mode Shapes ◽  
Analysis Model ◽  
Test Analysis ◽  
Modal Test ◽  
Artificial Neural

This article proposes a new damage detection method using Modal Test Analysis Model and artificial neural networks. A challenge in damage detection problems is lack of measured degrees of freedom, as well as limitations of attached sensors. Modal Test Analysis Model has been used in order to estimate unmeasured degrees of freedom. An experimental cantilever beam was used to show Modal Test Analysis Model’s efficiency in estimation of unmeasured mode shapes. To solve the inverse problem of damage detection, mode shapes estimated by Modal Test Analysis Model were used as inputs, and characteristics of the damage served as outputs of the artificial neural network. The sensitivity analysis carried out for each example showing the performance of artificial neural network after mode shape expansion was efficiently improved. Three numerical examples for plane and space truss structures are considered, in order to verify effectiveness of the proposed method. Results demonstrate a high accuracy of Modal Test Analysis Model and artificial neural network for structural damage detection.

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