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

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.

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Estimation of Mechanical Characteristics of Rubber Bearings with Multi Lead Plugs for Nuclear Power Plants Based on Finite Element Method

TRANSACTIONS OF THE JAPAN SOCIETY OF MECHANICAL ENGINEERS Series C ◽

10.1299/kikaic.72.2739 ◽

2006 ◽

Vol 72 (721) ◽

pp. 2739-2747

Author(s):

Akihiro MATSUDA ◽

Masato OMOTO ◽

Kazuta HIRATA

Keyword(s):

Finite Element Method ◽

Finite Element ◽

Nuclear Power ◽

Power Plants ◽

Nuclear Power Plants ◽

Mechanical Characteristics ◽

Rubber Bearings ◽

Element Method

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

International Journal of Modern Physics B ◽

10.1142/s0217979210065659 ◽

2010 ◽

Vol 24 (15n16) ◽

pp. 2797-2802 ◽

Cited By ~ 1

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.

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Experimental Investigation on Buckling of Steel Nuclear Containment With Elliptical Head

Volume 5: High-Pressure Technology; Rudy Scavuzzo Student Paper Symposium and 24th Annual Student Paper Competition; ASME Nondestructive Evaluation, Diagnosis and Prognosis Division (NDPD); Electric Power Research Institute (EPRI) Creep Fatigue Workshop ◽

10.1115/pvp2016-63498 ◽

2016 ◽

Cited By ~ 1

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.

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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

NTU KhPI Bulletin Power and heat engineering processes and equipment ◽

10.20998/2078-774x.2021.03.05 ◽

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.

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