Probabilistic fracture mechanics analysis of reactor pressure vessel with underclad and through-clad cracks under pressurised thermal shock transient

2018 ◽  
Vol 12 (1) ◽  
pp. 45
Author(s):  
Kuen Tsann Chen ◽  
Kuen Ting ◽  
Anh Tuan Nguyen ◽  
Tai Liang Kuo ◽  
Yuan Chih Li ◽  
...  
Keyword(s):  
Fracture Mechanics ◽  
Thermal Shock ◽  
Pressure Vessel ◽  
Reactor Pressure Vessel ◽  
Probabilistic Fracture Mechanics ◽  
Mechanics Analysis ◽  
Reactor Pressure

scholarly journals Recent verification activities on probabilistic fracture mechanics analysis code PASCAL4 for reactor pressure vessel

2020 ◽  
Vol 7 (3) ◽  
pp. 19-00573-19-00573
Author(s):  
Kai LU ◽  
Jinya KATSUYAMA ◽  
Yinsheng LI ◽  
Yuhei MIYAMOTO ◽  
Takatoshi HIROTA ◽  
...  
Keyword(s):  
Fracture Mechanics ◽  
Pressure Vessel ◽  
Reactor Pressure Vessel ◽  
Probabilistic Fracture Mechanics ◽  
Mechanics Analysis ◽  
Reactor Pressure

Fracture-Mechanics Analysis of the Reactor Pressure Vessel Beznau 1 Based on Thermal-Hydraulics Input Data From KWU-MIX and CFD Analyses

2016 ◽  
Author(s):  
Tomas Nicak ◽  
Richard Trewin ◽  
Elisabeth Keim ◽  
Ingo Cremer ◽  
Sebastien Blasset ◽  
...  
Keyword(s):  
Fracture Mechanics ◽  
Thermal Shock ◽  
Pressure Vessel ◽  
Brittle Failure ◽  
Input Data ◽  
Cold Water ◽  
Reactor Pressure Vessel ◽  
Mechanics Analysis ◽  
Cfd Analyses ◽  
Reactor Pressure

The integrity of a reactor pressure vessel (RPV) has to be ensured throughout its entire life in accordance with the applicable regulations. Typically an assessment of the RPV against brittle failure needs to be conducted by taking into account all possible loading cases. One of the most severe loading cases, which can potentially occur during the operating time, is the loss-of-coolant accident, where cold water is injected into the RPV nearly at operating conditions. High pressure in combination with a thermal shock of the ferritic pressure vessel wall caused by the injection of cold water leads to a considerable load at the belt-line area known as Pressurized Thermal Shock (PTS). Usually the assessment against brittle failure is based on a deterministic fracture-mechanics analysis, in which common parameters like J-integral or stress intensity factor are employed to calculate the load path for an assumed (postulated) flaw during the PTS event. The most important input data for the fracture-mechanics analysis is the transient thermal-hydraulics (TH) load of the RPV during the emergency cooling. Such data can be calculated by analytical fluid-mixing codes verified on experiments, such as KWU-MIX, or by numerical Computational Fluid Dynamics (CFD) tools after suitable validation. In KWU-MIX, which is the standard used for TH calculations within PTS analyses, rather conservative analytical models for the quantification of mixing and, depending on the water level, condensation processes in the downcomer (including simplified stripe and plume formations) are utilized. On the contrary, the numerical CFD tools can provide best-estimate results due to the possibility to consider more realistically the stripe and plume formations as well as the geometry of the RPV in detail. In order to quantify the safety margin inherent to the standard approach, two fracture-mechanics analyses of the RPV Beznau 1 based on thermal-hydraulic input data from KWU-MIX and CFD analyses were performed. Subsequently the resulting loading paths were compared between each other and with material properties obtained from the irradiation surveillance program of the RPV to demonstrate the exclusion of brittle-fracture initiation.

Development of Probabilistic Fracture Mechanics Analysis Codes for Reactor Pressure Vessels Considering Recent Embrittlement Model and Calculation Method of SIF: Progress of the Work

2016 ◽  
Author(s):  
Jongmin Kim ◽  
Bongsang Lee ◽  
Taehyun Kim ◽  
Yoonsuk Chang
Keyword(s):  
Stress Intensity ◽  
Fracture Mechanics ◽  
Pressure Vessel ◽  
Calculation Method ◽  
Reactor Pressure Vessel ◽  
Radiation Embrittlement ◽  
Calculation Methods ◽  
Probabilistic Fracture Mechanics ◽  
Mechanics Analysis ◽  
Reactor Pressure

It is widely recognized that the state of knowledge and data for the probabilistic calculations which had been proposed in the early 1980s made a conservative treatment of several key factors and models. Recently, applications of some new radiation embrittlement model, material database, calculation method of stress intensity factors and others which can improve fracture mechanics assessment of reactor pressure vessel (RPV) are introduced. This improvement on the accuracy and reliability of the probabilistic fracture mechanics (PFM) analysis necessitated changes in PFM analysis procedures and calculations. Modification and application of newly developed models and calculation methods are the main target of developing a probabilistic fracture mechanics analysis code based on the structure of existing R-PIE and VISA computer code to reflect the latest technical basis. Failure probabilities of reactor pressure vessel under pressurized thermal shock (PTS) conditions were calculated through finite difference method (FDM) and Monte Carlo simulation techniques with user friendly graphic interface. Moreover, various radiation embrittlement models and calculation methods of stress intensity factor at crack tip based on AFCEN code are applied and verified in the present work.

scholarly journals Development of Probabilistic Fracture Mechanics Analysis Code PASCAL ver.2 for Reactor Pressure Vessel

2007 ◽  
Vol 6 (2) ◽  
pp. 161-171 ◽  
Author(s):  
Kazuya OSAKABE ◽  
Kunio ONIZAWA ◽  
Katsuyuki SHIBATA ◽  
Masahide SUZUKI
Keyword(s):  
Fracture Mechanics ◽  
Pressure Vessel ◽  
Reactor Pressure Vessel ◽  
Probabilistic Fracture Mechanics ◽  
Mechanics Analysis ◽  
Reactor Pressure
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