Solution of the Lambda modes problem of a nuclear power reactor using an h–p finite element method

2014 ◽  
Vol 72 ◽  
pp. 338-349 ◽  
Author(s):  
A. Vidal-Ferrandiz ◽  
R. Fayez ◽  
D. Ginestar ◽  
G. Verdú
Keyword(s):  
Finite Element Method ◽  
Finite Element ◽  
Nuclear Power ◽  
Power Reactor ◽  
Nuclear Power Reactor ◽  
Element Method

scholarly journals CONCEPTION OF REFERENCE BOOKS OF DEFECTS IS IN EQUIPMENT AND PIPELINES OF POWER AND TRANSPORT SYSTEMS

2020 ◽  
Vol 1 (46) ◽  
pp. 387-404
Author(s):  
Kharytonova L ◽  
◽  
Kutsenko O ◽  
Kadenko I ◽  
◽  
...  
Keyword(s):  
Finite Element Method ◽  
Finite Element ◽  
Power Systems ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Transport Systems ◽  
Object Of Study ◽  
Service Inspection ◽  
Element Method

The paper focuses on the one of the persperctive approaches to the increasing of thje safety of Nuclear Power Plants - Flaw Handbook Concept. Object of study - equipment and piping of Nuclear Power Plants. Purpose of study - the description of the Flaw Handbook Concept and the application of the concept for the specific example. Method of the study - numerical procedures of the finite-element method and fracture mechanics. In the modern economics the optimization of the performance and operation of industry and power systems is of the main importance. The Flaw Handbook Concept is considered in the paper. This concept is applied on the nuclear power plants in the leading states with the aim of the optimization of the procedures of in-service inspection and repair. The main steps of the concept are considered and applied for the specific example. An example of Flaw Handbook using is analysed. The results of the paper can be incorporated into the procedures of in-service inspection for the safety-significant equipment and piping. KEYWORDS: FLAW HANDBOOK, BRITTLE FRACTURE, FATIGUE, FINITE-ELEMENT METHOD, SURGE PIPE.

ANALYSIS OF RESIDUAL STRESS FOR NARROW GAP WELDING USING FINITE ELEMENT METHOD

2010 ◽  
Vol 24 (15n16) ◽  
pp. 2797-2802 ◽  
Author(s):  
CHOON YEOL LEE ◽  
JAE KEUN HWANG ◽  
JOON WOO BAE
Keyword(s):  
Finite Element Method ◽  
Residual Stress ◽  
Finite Element ◽  
Nuclear Power ◽  
Power Plants ◽  
Large Diameter ◽  
Narrow Gap ◽  
Narrow Gap Welding ◽  
Welding Processes ◽  
Element Method

Reactor coolant loop (RCL) pipes circulating the heat generated in a nuclear power plant consist of so large diameter pipes that the installation of these pipes is one of the major construction processes. Conventionally, a shield metal arc welding (SMAW) process has been mainly used in RCL piping installations, which sometimes caused severe deformations, dislocation of main equipments and various other complications due to excessive heat input in welding processes. Hence, automation of the work of welding is required and narrow-gap welding (NGW) process is being reviewed for new nuclear power plants as an alternative method of welding. In this study, transient heat transfer and thermo-elastic-plastic analyses have been performed for the residual stress distribution on the narrow gap weldment of RCL by finite element method under various conditions including surface heat flux and temperature dependent thermo-physical properties.

Estimation of Surface Fault Displacement by High Performance Computing

2018 ◽  
Vol 12 (04) ◽  
pp. 1841003 ◽  
Author(s):  
Masataka Sawada ◽  
Kazumoto Haba ◽  
Muneo Hori
Keyword(s):  
Finite Element Method ◽  
Finite Element ◽  
Nuclear Power ◽  
High Performance ◽  
Time Integration ◽  
Three Dimensional ◽  
Shear Stiffness ◽  
Dynamic Simulations ◽  
Three Dimensional Models ◽  
Element Method

Reliable estimation of surface fault displacements is crucial to the safety of nuclear power plant facilities. It is necessary to develop a numerical method for the estimation. In the study, we develop a finite element method in which the following two functions are implemented: (1) a symplectic time integration of an explicit scheme to properly conserve the energy of the system; and (2) rigorously formulated joint elements of high order. The finite element method is enhanced with parallel computing capability. We apply the developed method to solve simple three-dimensional models of faults embedded in a rock mass. It includes a comparison of results from quasi-static and dynamic simulations and investigation of the sensitivity of results to the shear stiffness on faults. In the study, we propose capacity computing with a quasi-static simulation for uncertainty quantification.

Experimental Investigation on Buckling of Steel Nuclear Containment With Elliptical Head

2016 ◽  
Author(s):  
Keming Li ◽  
Jinyang Zheng ◽  
Zekun Zhang
Keyword(s):  
Finite Element Method ◽  
Finite Element ◽  
Internal Pressure ◽  
Nuclear Power ◽  
Power Plants ◽  
Nonlinear Finite Element ◽  
Nonlinear Finite Element Method ◽  
Initial Shape ◽  
Nuclear Containment ◽  
Element Method

Thanks to relative ease of fabrication and erection, steel containments with a torispherical or elliptical head are common in large nuclear power plants with relatively high internal pressure. Buckling failure mode is critical in the design of a steel containment with elliptical head under internal pressure. An experiment on buckling of elliptical head under internal pressure is presented in this paper. A test vessel was designed to provide a substantial margin of safety which permitted testing the head to rupture. The head has a diameter of 4797 mm, radius-to-height ratio of 1.728 and nominal thickness of 5.5 mm. Initial shape and shell displacement measurements of the head were carried out by using 3D laser scanners. The details of the buckling behavior are given. Initial buckling pressure was predicted by nonlinear finite element method considering the measured initial shape of the head. The agreement between the initial experimental buckling pressure and that predicted by the nonlinear finite element method is good. It is very useful in the buckling evaluation and the development of design rules for steel nuclear containment with elliptical head.

scholarly journals Assessment of the Technical Condition of the Emergency Heat Exchanger for the Reactor Plant of the B-320 Type in Order to Extend Its Service Life

2021 ◽  
pp. 36-43
Author(s):  
Valeriy Konshin ◽  
Mikola Zaiats
Keyword(s):  
Finite Element Method ◽  
Finite Element ◽  
Heat Exchanger ◽  
Nuclear Power ◽  
Power Plants ◽  
Technical Condition ◽  
Operating Conditions ◽  
Preliminary Evaluation ◽  
The Finite Element Method ◽  
Element Method

Extending the life of nuclear power plants in Ukraine during in the super-project period, as in most countries operating nuclear power units, is an accepted strategy and is being implemented practically. In this regard, there is a need for verification calculation of the main elements The calculated analysis of the stress-strain state of the heat exchanger is carried out using the finite element method of power equipment that determine the resource characteristics. The technical condition of the emergency cooling heat exchanger  for the power unit no. 3 of the SUNPP has been evaluated. The analysis of design, technical and operational documentation in the amount of preliminary evaluation of technical condition was performed. Potential mechanisms of wear of heat exchanger elements were determined. The technique of carrying out verification calculations for static, cyclic and seismic stability was described. The emergency cooling heat exchanger calculation model is made in the APM Structure 3D calculation code. The tense-deformed state of the heat exchanger is calculated using the finite-element method of resampling the design region.  The results of the verification calculation of the emergency cooling heat exchanger in the calculated states corresponding to normal operating conditions, hydrotests and under seismic impacts in conditions of the maximum design earthquake were presented.  The correspondence of the actual stresses in the calculation zones of the heat exchanger to the permissible values, specified in the current regulatory documentation was established. The amount of damageability to the heat exchanger elements was determined for the permissible number of load cycles. The cyclical strength of the elements of the emergency cooling heat exchanger, taking into account the period of application equal to 60 years inclusive, is ensured in accordance with the requirements.

Thermal Hydraulics of an ADSS With Sharp 180° Turn and Straight Flow Guide Using Finite Element Method

2005 ◽  
Author(s):  
K. Arul Prakash ◽  
B. V. Rathish Kumar ◽  
G. Biswas
Keyword(s):  
Heat Transfer ◽  
Finite Element Method ◽  
Finite Element ◽  
Nuclear Power ◽  
Critical Conditions ◽  
Critical System ◽  
Flow And Heat Transfer ◽  
Straight Flow ◽  
Flow Guide ◽  
Element Method

ADSS (Accelerator Driven Sub-critical System) nuclear reactors are expected to play a significant role in nuclear power generation in future due to their ability to utilize Thorium as the nuclear fuel, operating in sub-critical conditions and to transmute radioactive nuclear wastes. Numerical investigation of fluid flow and heat transfer characteristics of an ADSS has been accomplished using a finite element method based on Streamline Upwind Petrov-Galerkin (SUPG) technique. The time-dependent governing equations for conservation of mass, momentum and energy are solved. The simulations have been carried out to predict the heat transfer in the spallation regime. At the first place, laminar regime of the flow is considered for the ADSS geometry. The Reynolds number of interest were varied over a specified range.

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