Stress Analysis and Assess for Nuclear Pipes of Class 1 Based on PIPESTRESS

2013 ◽  
Author(s):  
Rui Liu ◽  
Tie-ping Li ◽  
Ming-yu Wang
Keyword(s):  
High Pressure ◽  
Nuclear Power ◽  
Nuclear Reactor ◽  
Safety Evaluation ◽  
Practical Method ◽  
Mechanical Analysis ◽  
Power Equipment ◽  
Pipe System ◽  
Finite Element Software ◽  
Class 1

In recent years, nuclear pipes play a very important role in the nuclear power equipment. Nuclear pipes especially class 1 usually contain cooling coolant of high temperature and high pressure in the operating nuclear power plant. The integrality of the nuclear pipes is very important to the nuclear reactor. This paper discusses the mechanical analysis and fatigue analysis of Class 1 pipes using finite element software. The soft ware PIPESTRESS is used to analysis the Class 1 pipe system to introduce the procedure for mechanical analysis and safety evaluation. The objective is to satisfy the criteria RCC-M. The results show that the pipe can guarantee the integrity of the pipe under the conditions. The method introduced in this paper can be used as a practical method for mechanical analysis of a nuclear Class 1 pipes.

scholarly journals Neutronic safety analysis of proposed reactor technologies for Ghana’s nuclear power plant using the MCNP code

2019 ◽  
Vol 34 (3) ◽  
pp. 238-242
Author(s):  
Rex Abrefah ◽  
Prince Atsu ◽  
Robert Sogbadji
Keyword(s):  
High Pressure ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Nuclear Reactor ◽  
Atomic Energy Commission ◽  
Clean Energy ◽  
Safety Analysis ◽  
Operating Conditions ◽  
The Government

In pursuance of sufficient, stable and clean energy to solve the ever-looming power crisis in Ghana, the Nuclear Power Institute of the Ghana Atomic Energy Commission has on the agenda to advise the government on the nuclear power to include in the country's energy mix. After consideration of several proposed nuclear reactor technologies, the Nuclear Power Institute considered a high pressure reactor or vodo-vodyanoi energetichesky reactor as the nuclear power technologies for Ghana's first nuclear power plant. As part of technology assessments, neutronic safety parameters of both reactors are investigated. The MCNP neutronic code was employed as a computational tool to analyze the reactivity temperature coefficients, moderator void coefficient, criticality and neutron behavior at various operating conditions. The high pressure reactor which is still under construction and theoretical safety analysis, showed good inherent safety features which are comparable to the already existing European pressurized reactor technology.

Evaluation of the Environmental Fatigue at a Long Radius Elbow of a Piping System under Transient Thermo Mechanical Stresses

2021 ◽  
Vol 412 ◽  
pp. 197-206
Author(s):  
Lenin Ramos-Cantú ◽  
Luis Héctor Hernández-Gómez ◽  
Rafael García-Illescas ◽  
Tanya Nerina Arreola-Valles ◽  
José Javier Moctezuma-Reyes ◽  
...  
Keyword(s):  
Nuclear Power ◽  
High Stress ◽  
Thermal Fluctuations ◽  
Mechanical Analysis ◽  
Piping System ◽  
Transient Effects ◽  
Stress Concentrations ◽  
Pipe System ◽  
Thermal Transients ◽  
Piping Systems

Thermal fatigue widely takes place in light water reactor (LWR) piping systems. It is an important aging mechanism of a nuclear reactor. Thermal transient effects occur at the startup and shutdown of a nuclear power plant. During the thermal transients, local and global cyclic stresses are induced in the piping systems. They are exacerbated by local geometric imperfections and environmental factors, which may lead to crack initiation. The elbows of such piping systems are subject to various combinations of loads (internal pressure, bending, and torsion, as well as thermal fluctuations) during their service life. As can be seen, high-stress concentrations are developed in these piping elements. Therefore, it is important to make a failure evaluation. In this paper, a 12” pipe system segment, which was made with SA 106 Gr C steel, has been considered. It was composed by two straight sections joined by a long radius elbow. Typical start-up and shutdown transient effects of a BWR-5 were considered. A computer-aided thermo-mechanical analysis was carried out using the finite element method. The linearization of the stresses was considered, based on the ASME B & PVC Code Section III, subsection NB. Under these conditions, environmental fatigue was analyzed after 40-and 60-years operation.

Accident Safety Evaluation Method for Spent Fuel Dry Storage Facilities

2017 ◽  
Author(s):  
Zhang Yubin ◽  
Ouyang Yong ◽  
Zhou Yuwei ◽  
Liu Jinlin
Keyword(s):  
Nuclear Power ◽  
Evaluation Method ◽  
Safety Evaluation ◽  
Heat Removal ◽  
Mechanical Analysis ◽  
Spent Fuel ◽  
Dry Storage ◽  
Accident Condition ◽  
Design Characteristics ◽  
Storage Facilities

Dry storage is one of the ways to store spent fuel in the middle of the reactor, which can effectively alleviate the pressure of the storage on the spent fuel pool of nuclear power plant. This paper tries to combine the site of dry storage facilities and the design characteristics to explain and discuss the safety evaluation method under the accident condition, from the mechanical analysis, critical safety, the decay heat removal, the shielding design and so on. Then according to the operating procedures and the accident condition that may be occurred, put forward some possible ways of monitoring and measures of safety protection should be added.

Nuclear Power Equipment (Design)

2014 ◽  
Vol 83 (4) ◽  
pp. 270-274 ◽  
Author(s):  
Masashi KAMEYAMA
Keyword(s):  
Nuclear Power ◽  
Equipment Design ◽  
Power Equipment

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.

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.

Nuclear Power Plant Fires and Explosions: Part IV — Water Hammer Explosion Mechanisms

2017 ◽  
Author(s):  
Robert A. Leishear
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Nuclear Reactor ◽  
Water Hammer ◽  
Compression Process ◽  
Flammable Gas ◽  
Piping Systems ◽  
Fukushima Daiichi ◽  
Accident Conditions

Water hammers, or fluid transients, compress flammable gasses to their autognition temperatures in piping systems to cause fires or explosions. While this statement may be true for many industrial systems, the focus of this research are reactor coolant water systems (RCW) in nuclear power plants, which generate flammable gasses during normal operations and during accident conditions, such as loss of coolant accidents (LOCA’s) or reactor meltdowns. When combustion occurs, the gas will either burn (deflagrate) or explode, depending on the system geometry and the quantity of the flammable gas and oxygen. If there is sufficient oxygen inside the pipe during the compression process, an explosion can ignite immediately. If there is insufficient oxygen to initiate combustion inside the pipe, the flammable gas can only ignite if released to air, an oxygen rich environment. This presentation considers the fundamentals of gas compression and causes of ignition in nuclear reactor systems. In addition to these ignition mechanisms, specific applications are briefly considered. Those applications include a hydrogen fire following the Three Mile Island meltdown, hydrogen explosions following Fukushima Daiichi explosions, and on-going fires and explosions in U.S nuclear power plants. Novel conclusions are presented here as follows. 1. A hydrogen fire was ignited by water hammer at Three Mile Island. 2. Hydrogen explosions were ignited by water hammer at Fukushima Daiichi. 3. Piping damages in U.S. commercial nuclear reactor systems have occurred since reactors were first built. These damages were not caused by water hammer alone, but were caused by water hammer compression of flammable hydrogen and resultant deflagration or detonation inside of the piping.

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