scholarly journals STUDY ON THERMAL CHARACTERISTICS OF U-SHAPED HEAT PIPE

2021 ◽  
Vol 23 (3) ◽  
pp. 115
Author(s):  
Mukhsinun Hadi Kusuma ◽  
Anhar Riza Antariksawan ◽  
Giarno Giarno ◽  
Dedy Haryanto ◽  
Surip Widodo
Keyword(s):  
Heat Pipe ◽  
Nuclear Power ◽  
Nuclear Reactor ◽  
Thermal Characteristics ◽  
Small Scale ◽  
Air Cooling ◽  
Station Blackout ◽  
Higher Temperature ◽  
Passive Cooling Systems ◽  
Relap5 Code

The latest accident in Japan's nuclear power station became a valuable experience to start engaging passive cooling systems (PCS) more aggressively to improve safety aspects in nuclear power reactors being studied in Indonesia. This investigation is related to the U-shaped heat pipe (UHP) research as PCS of water in the cooling tank (CT). The objective of this research is to study the thermal characteristics of UHP as PCS in the CT. The experiment on small-scale UHP and simulation with RELAP5 code has been conducted to understand the performance of UHP. The experiment results of the small-scale UHP model will be used as a basic understanding of simulating and designing a UHP with big scaling. The study result showed the highest thermal performance of UHP was obtained when it operated on the higher temperature of heat load and higher air cooling velocity. The more UHPs inserted into the cooling pool, the more heat that can be discharged into the environment. This result also shows promising use of UHP for CT PCS. The use of UHP as PCS can enhance the safety aspect of the nuclear reactor, especially in station blackout event.

The Thermal-Hydraulics Behavior Analysis Before the CANDU Fuel Channel Dry-Out

2010 ◽  
Author(s):  
Jingtian Yuan ◽  
Lili Tong ◽  
Xuewu Cao
Keyword(s):  
Nuclear Power ◽  
Fuel Temperature ◽  
Steam Generators ◽  
Fuel Channel ◽  
Whole Process ◽  
Station Blackout ◽  
Main Steam ◽  
Dry Out ◽  
Secondary Side ◽  
Relap5 Code

The CANDU 6 nuclear power plant is modeled with SCDAP/RELAP5 code including the heat transport system (HTS), pressure and inventory control system, calandria vessel (CV) and main steam lines, etc. Through simulation of station blackout (SBO), the results obtained before the fuel channel dry-out are illustrated. Before the steam generators (SGs) dry out, the core decay heat could be removed through the SG secondary side, and fuel temperature, pressure of the HTS could keep low. The loss of the HTS inventory results in the fuel channel dry-out, heat-up of the fuel, the calandria and pressure tubes. The moderator temperature remains low in the whole process.

scholarly journals On the Root Causes of the Fukushima Daiichi Disaster from the Perspective of High Complexity and Tight Coupling in Large-Scale Systems

Symmetry ◽  
2021 ◽  
Vol 13 (3) ◽  
pp. 414
Author(s):  
Atsuo Murata ◽  
Waldemar Karwowski
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Large Scale ◽  
High Complexity ◽  
Tight Coupling ◽  
Large Scale Systems ◽  
Safety Issues ◽  
Loss Of Coolant ◽  
Station Blackout ◽  
Fukushima Daiichi

This study explores the root causes of the Fukushima Daiichi disaster and discusses how the complexity and tight coupling in large-scale systems should be reduced under emergencies such as station blackout (SBO) to prevent future disasters. First, on the basis of a summary of the published literature on the Fukushima Daiichi disaster, we found that the direct causes (i.e., malfunctions and problems) included overlooking the loss of coolant and the nuclear reactor’s failure to cool down. Second, we verified that two characteristics proposed in “normal accident” theory—high complexity and tight coupling—underlay each of the direct causes. These two characteristics were found to have made emergency management more challenging. We discuss how such disasters in large-scale systems with high complexity and tight coupling could be prevented through an organizational and managerial approach that can remove asymmetry of authority and information and foster a climate of openly discussing critical safety issues in nuclear power plants.

scholarly journals Modification of a Solar Thermal Collector to Promote Heat Transfer inside an Evacuated Tube Solar Thermal Absorber

2021 ◽  
Vol 11 (9) ◽  
pp. 4100
Author(s):  
Rasa Supankanok ◽  
Sukanpirom Sriwong ◽  
Phisan Ponpo ◽  
Wei Wu ◽  
Walairat Chandra-ambhorn ◽  
...  
Keyword(s):  
Heat Transfer ◽  
Heat Pipe ◽  
Solar Thermal ◽  
Heat Transfer Fluid ◽  
Small Scale ◽  
Medium Temperature ◽  
Change Behavior ◽  
Set Up ◽  
Evacuated Tube ◽  
Heating Medium

Evacuated-tube solar collector (ETSC) is developed to achieve high heating medium temperature. Heat transfer fluid contained inside a copper heat pipe directly affects the heating medium temperature. A 10 mol% of ethylene-glycol in water is the heat transfer fluid in this system. The purpose of this study is to modify inner structure of the evacuated tube for promoting heat transfer through aluminum fin to the copper heat pipe by inserting stainless-steel scrubbers in the evacuated tube to increase heat conduction surface area. The experiment is set up to measure the temperature of heat transfer fluid at a heat pipe tip which is a heat exchange area between heat transfer fluid and heating medium. The vapor/ liquid equilibrium (VLE) theory is applied to investigate phase change behavior of the heat transfer fluid. Mathematical model validated with 6 experimental results is set up to investigate the performance of ETSC system and evaluate the feasibility of applying the modified ETSC in small-scale industries. The results indicate that the average temperature of heat transfer fluid in a modified tube increased to 160.32 °C which is higher than a standard tube by approximately 22 °C leading to the increase in its efficiency by 34.96%.

scholarly journals Computational Fluid Dynamics Modeling of Steam Condensation on Nuclear Containment Wall Surfaces Based on Semiempirical Generalized Correlations

2012 ◽  
Vol 2012 ◽  
pp. 1-7 ◽  
Author(s):  
Pavan K. Sharma ◽  
B. Gera ◽  
R. K. Singh ◽  
K. K. Vaze
Keyword(s):  
Nuclear Power ◽  
Nuclear Reactor ◽  
Pressure Rise ◽  
Dynamics Modeling ◽  
Steam Condensation ◽  
Noncondensable Gases ◽  
Nuclear Containment ◽  
Design Basis ◽  
Lumped Parameter ◽  
Transport Calculation

In water-cooled nuclear power reactors, significant quantities of steam and hydrogen could be produced within the primary containment following the postulated design basis accidents (DBA) or beyond design basis accidents (BDBA). For accurate calculation of the temperature/pressure rise and hydrogen transport calculation in nuclear reactor containment due to such scenarios, wall condensation heat transfer coefficient (HTC) is used. In the present work, the adaptation of a commercial CFD code with the implementation of models for steam condensation on wall surfaces in presence of noncondensable gases is explained. Steam condensation has been modeled using the empirical average HTC, which was originally developed to be used for “lumped-parameter” (volume-averaged) modeling of steam condensation in the presence of noncondensable gases. The present paper suggests a generalized HTC based on curve fitting of most of the reported semiempirical condensation models, which are valid for specific wall conditions. The present methodology has been validated against limited reported experimental data from the COPAIN experimental facility. This is the first step towards the CFD-based generalized analysis procedure for condensation modeling applicable for containment wall surfaces that is being evolved further for specific wall surfaces within the multicompartment containment atmosphere.

Compressed Air Energy Storage System Exergy Analysis and its Combined Operation with Nuclear Power Plants

2013 ◽  
Vol 448-453 ◽  
pp. 2786-2789 ◽  
Author(s):  
Jin Li ◽  
Chu Fu Li ◽  
Yan Xia Zhang ◽  
Hui Guo Yue
Keyword(s):  
Energy Storage ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Exergy Analysis ◽  
Compressed Air ◽  
Air Cooling ◽  
Compressed Air Energy Storage ◽  
Tracking Mode ◽  
Combined Operation

Nuclear plants are facing more and more peaking pressure, and combined operation with compressed air energy storage (CAES) systems is an effective approach to improve its peaking capacity. This work first simulates and conducts the exergy analysis for the CAES system. The results show that exergy efficiency of the CAES system is about 51.7%, as well as the exergy loss are primary in the fuel combustion and compressed air cooling processes, accounted for 25.4% and 11.3% of total exergy, respectively. Subsequently, three combined operation modes between CAES system and nuclear power plants for power grid peaking are investigated, which shows that three section tracking mode and incomplete tracking mode can achieve the balance between peaking effects and peaking cost.

Preliminary feasibility analysis of Heat Pipe Cooled Bimodal Space Nuclear Reactor

2021 ◽  
pp. 103817
Author(s):  
Xiaoyan Tian ◽  
Huaqi Li ◽  
Duoyu Jiang ◽  
Lei Zhu ◽  
Sen Chen ◽  
...  
Keyword(s):  
Heat Pipe ◽  
Nuclear Reactor ◽  
Feasibility Analysis

Micro Fluidic Jets for Thermal Management of Electronics

Volume 4 ◽  
2004 ◽  
Author(s):  
Jivtesh Garg ◽  
Mehmet Arik ◽  
Stanton Weaver ◽  
Seyed Saddoughi
Keyword(s):  
Heat Transfer ◽  
Natural Convection ◽  
Heated Surface ◽  
Turbulent Jets ◽  
Harmonic Signal ◽  
Synthetic Jet ◽  
Synthetic Jets ◽  
Small Scale ◽  
Air Cooling ◽  
Infrared Thermal Imaging

Micro fluidics devices are conventionally used for boundary layer control in many aerospace applications. Synthetic Jets are intense small scale turbulent jets formed from entrainment and expulsion of the fluid in which they are embedded. The idea of using synthetic jets in confined electronic cooling applications started in late 1990s. These micro fluidic devices offer very efficient, high magnitude direct air-cooling on the heated surface. A proprietary synthetic jet designed in General Electric Company was able to provide a maximum air velocity of 90 m/s from a 1.2 mm hydraulic diameter rectangular orifice. An experimental study for determining the thermal performance of a meso scale synthetic jet was carried out. The synthetic jets are driven by a time harmonic signal. During the experiments, the operating frequency for jets was set between 3 and 4.5 kHz. The resonance frequency for a particular jet was determined through the effect on the exit velocity magnitude. An infrared thermal imaging technique was used to acquire fine scale temperature measurements. A square heater with a surface area of 156 mm2 was used to mimic the hot component and extensive temperature maps were obtained. The parameters varied during the experiments were jet location, driving jet voltage, driving jet frequency and heater power. The output parameters were point wise temperatures (pixel size = 30 μm), and heat transfer enhancement over natural convection. A maximum of approximately 8 times enhancement over natural convection heat transfer was measured. The maximum coefficient of cooling performance obtained was approximately 6.6 due to the low power consumption of the synthetic jets.

Heat Transfer Correlation for Supercritical Carbon Dioxide Flowing in Vertical Bare Tubes

2013 ◽  
Author(s):  
Sahil Gupta ◽  
Eugene Saltanov ◽  
Igor Pioro
Keyword(s):  
Heat Transfer ◽  
Error Analysis ◽  
Nuclear Power ◽  
Nuclear Reactor ◽  
Large Set ◽  
Transfer Coefficients ◽  
Data Points ◽  
Institute Of Technology ◽  
Using Data ◽  
Gen Iv

Canada among many other countries is in pursuit of developing next generation (Generation IV) nuclear-reactor concepts. One of the main objectives of Generation-IV concepts is to achieve high thermal efficiencies (45–50%). It has been proposed to make use of SuperCritical Fluids (SCFs) as the heat-transfer medium in such Gen IV reactor design concepts such as SuperCritical Water-cooled Reactor (SCWR). An important aspect towards development of SCF applications in novel Gen IV Nuclear Power Plant (NPP) designs is to understand the thermodynamic behavior and prediction of Heat Transfer Coefficients (HTCs) at supercritical (SC) conditions. To calculate forced convection HTCs for simple geometries, a number of empirical 1-D correlations have been proposed using dimensional analysis. These 1-D HTC correlations are developed by applying data-fitting techniques to a model equation with dimensionless terms and can be used for rudimentary calculations. Using similar statistical techniques three correlations were proposed by Gupta et al. [1] for Heat Transfer (HT) in SCCO2. These SCCO2 correlations were developed at the University of Ontario Institute of Technology (Canada) by using a large set of experimental SCCO2 data (∼4,000 data-points) obtained at the Chalk River Laboratories (CRL) AECL. These correlations predict HTC values with an accuracy of ±30% and wall temperatures with an accuracy of ±20% for the analyzed dataset. Since these correlations were developed using data from a single source - CRL (AECL), they can be limited in their range of applicability. To investigate the tangible applicability of these SCCO2 correlations it was imperative to perform a thorough error analysis by checking their results against a set of independent SCCO2 tube data. In this paper SCCO2 data are compiled from various sources and within various experimental flow conditions. HTC and wall-temperature values for these data points are calculated using updated correlations presented in [1] and compared to the experimental values. Error analysis is then shown for these datasets to obtain a sense of the applicability of these updated SCCO2 correlations.

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