3D transient coupled thermo-elastic-plastic contact sealing analysis of reactor pressure vessel

2005 ◽  
Vol 18 (04) ◽  
pp. 473
Author(s):  
Xuesong Du
Keyword(s):  
Pressure Vessel ◽  
Reactor Pressure Vessel ◽  
Elastic Plastic ◽  
Reactor Pressure

Reactor Pressure Vessel Integrity Assessment: French Simplified Analysis Method Applied to Underclad Postulated Defect in Nozzle

2016 ◽  
Author(s):  
Adolfo Arrieta-Ruiz ◽  
Eric Meister ◽  
Stéphane Vidard
Keyword(s):  
Fracture Mechanics ◽  
Fracture Risk ◽  
Pressure Vessel ◽  
Reactor Pressure Vessel ◽  
Core Area ◽  
Integrity Assessment ◽  
Elastic Plastic ◽  
The Core ◽  
Simplified Method ◽  
Reactor Pressure

Structural integrity of the Reactor Pressure Vessel (RPV) is one of the main concerns regarding safety and lifetime of Nuclear Power Plants (NPP) since this component is considered as not reasonably replaceable. Fast fracture risk is the main potential damage considered in the integrity assessment of RPV. In France, deterministic integrity assessment for RPV vis-à-vis the brittle fracture risk is based on the crack initiation stage. As regards the core area in particular, the stability of an under-clad postulated flaw is currently evaluated under a Pressurized Thermal Shock (PTS) through a dedicated fracture mechanics simplified method called “beta method”. However, flaw stability analyses are also carried-out in several other areas of the RPV. Thence-forward performing uniform simplified inservice analyses of flaw stability is a major concern for EDF. In this context, 3D finite element elastic-plastic calculations with flaw modelling in the nozzle have been carried out recently and the corresponding results have been compared to those provided by the beta method, codified in the French RSE-M code for under-clad defects in the core area, in the most severe events. The purpose of this work is to validate the employment of the core area fracture mechanics simplified method as a conservative approach for the under-clad postulated flaw stability assessment in the complex geometry of the nozzle. This paper presents both simplified and 3D modelling flaw stability evaluation methods and the corresponding results obtained by running a PTS event. It shows that the employment of the “beta method” provides conservative results in comparison to those produced by elastic-plastic calculations for the cases here studied.

Structural Integrity Research for Reactor Pressure Vessel Under In-Vessel Retention of a Core Melt

2016 ◽  
Author(s):  
Yongjian Gao ◽  
Yinbiao He ◽  
Ming Cao ◽  
Yuebing Li ◽  
Shiyi Bao ◽  
...  
Keyword(s):  
Pressure Vessel ◽  
Material Properties ◽  
Power Plants ◽  
Structural Integrity ◽  
Plastic Material ◽  
Plastic Region ◽  
Reactor Pressure Vessel ◽  
Elastic Plastic ◽  
Elastic Plastic Material ◽  
Reactor Pressure

In-Vessel Retention (IVR) is one of the most important severe accident mitigation strategies of the third generation passive Nuclear Power Plants (NPP). It is intended to demonstrate that in the case of a core melt, the structural integrity of the Reactor Pressure Vessel (RPV) is assured such that there is no leakage of radioactive debris from the RPV. This paper studied the IVR issue using Finite Element Analyses (FEA). Firstly, the tension and creep testing for the SA-508 Gr.3 Cl.1 material in the temperature range of 25°C to 1000°C were performed. Secondly, a FEA model of the RPV lower head was built. Based on the assumption of ideally elastic-plastic material properties derived from the tension testing data, limit analyses were performed under both the thermal and the thermal plus pressure loading conditions where the load bearing capacity was investigated by tracking the propagation of plastic region as a function of pressure increment. Finally, the ideal elastic-plastic material properties incorporating the creep effect are developed from the 100hr isochronous stress-strain curves, limit analyses are carried out as the second step above. The allowable pressures at 0 hr and 100 hr are obtained. This research provides an alternative approach for the structural integrity evaluation for RPV under IVR condition.

Evaluation on Constraint Effect of Reactor Pressure Vessel Under Pressurized Thermal Shock

2013 ◽  
Author(s):  
Naoki Ogawa ◽  
Kentaro Yoshimoto ◽  
Takatoshi Hirota ◽  
Shohei Sakaguchi ◽  
Toru Oumaya
Keyword(s):  
Fracture Toughness ◽  
Thermal Shock ◽  
Surface Crack ◽  
Pressure Vessel ◽  
Crack Tip ◽  
Reactor Pressure Vessel ◽  
Constraint Effect ◽  
Elastic Plastic ◽  
Pressurized Thermal Shock ◽  
Reactor Pressure

In recent years, the integrity of reactor pressure vessel (RPV) under pressurized thermal shock (PTS) accident has become controversial issue since the larger shift of RTNDT in some higher fluence surveillance data raised a concern on RPV integrity. Under PTS condition, the combination of thermal stress due to a temperature gradient and mechanical stress due to internal pressure causes considerable tensile stress inside the wall of RPV. Currently, RPV integrity is assessed by comparing stress intensity factor on a crack tip under PTS condition and a reference toughness curve based on the fracture toughness data of irradiated compact specimens. Since PTS loading is large enough to cause plastic deformation, a crack tip behavior on the inner surface of RPV can be explained by elastic-plastic fracture mechanics using the J-integral. In this study, 3D elastic plastic finite element analyses were performed to assess the crack tip behavior on surface of a RPV under Loss of coolant Accident, which causes one of the most severe PTS condition. In order to quantify the constraint effect on a surface crack, J-Q approach was applied. The constraint effect of a surface crack was compared with a compact specimen and its influence on the fracture toughness was assessed. As a result, the difference of constraint effect was clearly obtained. And it is recommended to consider constraint effects in the evaluation of structural integrity of RPV under PTS.

Simplified J-Integral Evaluation Method for Evaluation of Reactor Pressure Vessel in the Upper-Shelf Region

2003 ◽  
Author(s):  
Yoshio Urabe ◽  
Seiji Asada ◽  
Takatoshi Hirota ◽  
Morihito Nakano ◽  
Ryuuichi Maeda
Keyword(s):  
Pressure Vessel ◽  
Evaluation Method ◽  
Fe Analysis ◽  
Reactor Pressure Vessel ◽  
J Integral ◽  
Elastic Analysis ◽  
3 Dimensional ◽  
Elastic Plastic ◽  
Shelf Region ◽  
Reactor Pressure

It is well-known that the Upper-Shelf Energy of reactor pressure vessel (RPV) steels is reduced due to neutron irradiation and J-integral (J-applied) is used for the structural integrity evaluation in upper-shelf region. It is very time consuming to calculate the J-integral in detail by using 3-dimensional elastic-plastic FE analysis. U.S. Regulatory Guide 1.161[1] applies a simplified J-applied calculation method based on elastic calculation of the stress intensity factor, KI. However, this method for thermal stress can be applied only for constant cooling rates. From this point of view, a simplified J-applied evaluation method which can be applied to design transients and has high accuracy has been studied. In order to develop the new evaluation method, KI based on 1-dimensional elastic analysis, KI based on 3-dimensional elastic FE analysis and J-applied based on 3-dimensional elastic-plastic FE analysis are calculated and compared with each other. An accurate J estimation method for design transients from 1-dimensional elastic analysis results is proposed and the severest transient in each condition of each RPV group is evaluated in this paper.

Study of the Evolution of Defects in the Structure of Reactor Pressure Vessel by Rate Theory

2014 ◽  
Vol 10 (1) ◽  
pp. 123-127 ◽  
Author(s):  
Gyeong-Geun Lee ◽  
Yong-Bok Lee ◽  
Min-Chul Kim ◽  
Junhyun Kwon
Keyword(s):  
Pressure Vessel ◽  
Reactor Pressure Vessel ◽  
Rate Theory ◽  
Reactor Pressure
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