Estimates of Mechanical Properties and Residual Stress of Narrow Gap Weld for Leak-Before-Break Application to Nuclear Piping

2011 ◽  
Vol 133 (2) ◽  
Author(s):  
Jun-Seok Yang ◽  
Chi-Yong Park ◽  
Nam-Su Huh
Keyword(s):  
Residual Stress ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Steam Generator ◽  
Nuclear Power ◽  
Welding Residual Stress ◽  
Fracture Toughness Test ◽  
Narrow Gap ◽  
Main Loop ◽  
Leak Before Break

The present paper addresses the mechanical characteristics of the stainless steel narrow gap weld (NGW) for the leak-before-break (LBB) application to the main loop piping of a nuclear power plant. Recently, in Korea, the connection with the reactor coolant main loop piping and the steam generator has been welded with ER308L NGW after the replacement of a steam generator of a Korean nuclear power plant. The NGW technique has many merits, for instance, the reduction of construction time and the reduction of shrinkage and deformation after welding due to its small groove angle and welding bead width compared with the conventional welds. In this paper, the tensile and fracture toughness test results of the three ER308L test coupons from NGW were presented and compared with those from conventional welds at the operating condition of the nuclear power plant. In addition, the distribution of the welding residual stress as well as the deformation behavior of the ER308L weld due to NGW was predicted through the nonlinear two-dimensional finite element analysis in which the detailed actual welding process of NGW was simulated. The results presented in this paper can be used to evaluate LBB application to nuclear piping with NGW and to provide the important information to perform the flaw evaluation as well as improve the weld procedure of NGW.

Welding Residual Stress in a Large Diameter Nuclear Reactor Pressure Vessel Nozzle

2013 ◽  
Vol 135 (2) ◽  
Author(s):  
Tao Zhang ◽  
Frederick W. Brust ◽  
Gery Wilkowski ◽  
Heqin Xu ◽  
Alfredo A. Betervide ◽  
...  
Keyword(s):  
Residual Stress ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Pressure Vessel ◽  
Nuclear Power ◽  
Large Diameter ◽  
Reactor Pressure Vessel ◽  
Welding Residual Stress ◽  
Stress Fields ◽  
Reactor Pressure

The Atucha II nuclear power plant is a pressurized heavy water reactor (PHWR) being constructed in Argentina. The original plant was designed by Kraftwerk Union (KWU) in the 1970’s using the German methodology of break preclusion. The plant construction was halted for several decades, but a recent need for power was the driver for restarting the construction. Welding residual stresses in nuclear power plant piping can lead to cracking concerns later in the life of the plant, especially for stress-corrosion cracking. Hence, understanding the residual stress distribution from welding is important to evaluate the reliability of pipe and nozzle joints with welds. In this paper, a large-diameter reactor pressure vessel (RPV) hot-leg nozzle was analyzed. This is a nozzle from Atucha II nuclear power plant in Argentina. The main piping material is 20MnMoNi55 with Tenacito 65R weld metal, and inner diameter (ID) welded cladding at the girth weld locations is made of 309L. The special materials and weld geometry will lead interesting welding residual stress fields. In addition, postweld heat treatment (PWHT) of the girth welds and its boundary conditions could also play an important role in determining welding residual stress fields at the plant’s normal operating conditions. Sensitivity analyses were conducted and the technical observations and comments are provided.

Welding Residual Stress in a Large Diameter Nuclear Reactor Pressure Vessel Nozzle

2012 ◽  
Author(s):  
Tao Zhang ◽  
Frederick W. Brust ◽  
Gery Wilkowski ◽  
Heqin Xu ◽  
Alfredo A. Betervide ◽  
...  
Keyword(s):  
Residual Stress ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Pressure Vessel ◽  
Nuclear Power ◽  
Large Diameter ◽  
Reactor Pressure Vessel ◽  
Welding Residual Stress ◽  
Sensitivity Analyses ◽  
Reactor Pressure

The Atucha II nuclear power plant is a pressurized heavy water reactor being constructed in Argentina. The original plant was designed by KWU in the 1970’s using the German methodology of break preclusion. The plant construction was halted for several decades, but a recent need for power was the driver for restarting the construction. Many times welding residual stresses in nuclear power plant can lead to cracking concerns. Hence, understanding the residual stress distribution is important to evaluate the reliability of pipe and nozzle joints with welds. In this paper, a large diameter Reactor Pressure Vessel (RPV) hot-leg nozzle was analyzed. This is a nozzle from Atucha II nuclear power plant in Argentina. The main weld material is 20MnMoNi55 and ID welded cladding is made of 309L. The special materials and weld geometry will lead to interesting welding residual stress field. In addition, post-weld heat treatment (PWHT) region and boundary conditions could also play an important role in determining welding residual stress fields. Sensitivity analyses were conducted and the technical observation and comments will conclude the paper.

Leak behavior of SCC degraded steam generator tubings of nuclear power plant

2005 ◽  
Vol 235 (23) ◽  
pp. 2477-2484 ◽  
Author(s):  
Seong Sik Hwang ◽  
Hong Pyo Kim ◽  
Joung Soo Kim ◽  
Kenneth E. Kasza ◽  
Jangyul Park ◽  
...  
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Steam Generator ◽  
Nuclear Power

scholarly journals Aplicación de la Teoría de Perturbación – Método Diferencial- al Análisis de Sensibilidad en Generadores de Vapor de Centrales Nucleares PWR-Caso Angra I

2019 ◽  
pp. 119-126
Keyword(s):  
Perturbation Theory ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Steam Generator ◽  
Nuclear Power ◽  
Homogeneous Model ◽  
Differential Method ◽  
Differential Approach ◽  
Good Potential ◽  
Direct Calculations

Aplicación de la Teoría de Perturbación – Método Diferencial- al Análisis de Sensibilidad en Generadores de Vapor de Centrales Nucleares PWR-Caso Angra I Aplication of the Perturbation Theory- Differential Methodto Sensibility Análisis in PWR Nuclear Power Plant Steam Generator- Angra I Giol Sanders R, Andrade de Lima F, Marques A, Gallardo A, Bruna M, Zúñiga A Institución Peruano de Energía Nuclear Universidad Federal de Rio De Janeiro-Brasil DOI: https://doi.org/10.33017/RevECIPeru2011.0033/ RESUMEN En este trabajo basado en la tesis del Magíster Roberto Giol S. [1] presenta una aplicación del formalismo diferencial de la teoría de perturbación a un modelo termohidráulico homogéneo de simulación del comportamiento estacionario de uno de los generadores de vapor de la Central Nuclear tipo PWR Angra I del Brasil. Se desarrolla un programa de cálculo PERGEVAP tomando como base el código GEVAP de Souza[2]. El programa PERGEVAP permite realizar cálculos de sensibilidad de funcionales lineales (temperatura media del primario)y no lineales (flujo de calor medio a través de las paredes de los tubos del generador) con relación a las variaciones de ciertos parámetros termo-hidráulicos(flujo másico del primario, calor específico, etc), Los resultados obtenidos con este formalismo son luego comparados con los obtenidos del cálculo directo con el propio código GEVAP, pudiéndose verificar una excelente concordancia. Este método se muestra promisorio para efectuar cálculos repetitivos asociados al diseño y análisis de Seguridad de los componentes de las Centrales Nucleares. Descriptores: teoría de perturbación, método diferencial, sensibilidad, generador de vapor, central nuclear PWR. ABSTRACT This report presents an application of the differential approach of the perturbation theory to an homogeneous model of a PWR steam generator in the Angra 1 Nuclear Power Plan in Brazil under steady-state conditions. Program PERGEVAP was built fom the code GEVAP developed by Souza and allows sensitivity calculations of linear (average primary loop temperature) and non-linear (average heat flux) functionals due to variations in some thermo-hydraulics parameters (flow rate, specific heat, , etc). Results obtained with this approach are then compared with direct calculations performed using the GEVAP code, with excellent agreements. The method has good potential to treat repeated calculations needed in the design and safety analysis of the Nuclear Plant components. Keywords: perturbation theory, differential method, steam generator, PWR nuclear Power Plant.

Water Level Control of Steam Generator in Nuclear Power Plant Based on Intelligent MFAC-PID

2021 ◽  
Author(s):  
Yonglu Du ◽  
Haotian Li ◽  
Minrui Fei ◽  
Ling Wang ◽  
Pinggai Zhang ◽  
...  
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Water Level ◽  
Steam Generator ◽  
Nuclear Power ◽  
Level Control
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