Stress Intensity Factor for Cracks in Control Rod Drive Mechanism Nozzles near the J-weld in Reactor Pressure Vessel Head

2011 ◽  
Vol 47 (20) ◽  
pp. 109
Author(s):  
Zhengjun GU
Keyword(s):  
Stress Intensity Factor ◽  
Stress Intensity ◽  
Intensity Factor ◽  
Pressure Vessel ◽  
Reactor Pressure Vessel ◽  
Control Rod ◽  
Drive Mechanism ◽  
Reactor Pressure

Structural Integrity of Penetration Nozzle of Reactor Pressure Vessel Head of PWR Plant

2008 ◽  
Author(s):  
Kiminobu Hojo ◽  
Naoki Ogawa ◽  
Yoichi Iwamoto ◽  
Kazutoshi Ohoto ◽  
Seiji Asada ◽  
...  
Keyword(s):  
Residual Stress ◽  
Stress Intensity Factor ◽  
Stress Intensity ◽  
Intensity Factor ◽  
Stress Field ◽  
Pressure Vessel ◽  
Structural Integrity ◽  
Complex Structure ◽  
Reactor Pressure Vessel ◽  
Reactor Pressure

A reactor pressure vessel (RPV) head of PWR has penetration holes for the CRDM nozzles, which are connected with the vessel head by J-shaped welds. It is well-known that there is high residual stress field in vicinity of the J-shaped weld and this has potentiality of PWSCC degradation. For assuring stress integrity of welding part of the penetration nozzle of the RPV, it is necessary to evaluate precise residual stress and stress intensity factor based on the stress field. To calculate stress intensity factor K, the most acceptable procedure is numerical analysis, but the penetration nozzle is very complex structure and such a direct procedure takes a lot of time. This paper describes applicability of simplified K calculation method from handbooks by comparing with K values from finite element analysis, especially mentioning crack modeling. According to the verified K values in this paper, fatigue crack extension analysis and brittle fracture evaluation by operation load were performed for initial crack due to PWSCC and finally structural integrity of the penetration nozzle of RPV head was confirmed.

Research on the Fracture Mechanics Analysis Method for Reactor Pressure Vessel Nozzle Corner Flaw

2017 ◽  
Author(s):  
Shen Rui ◽  
Cao Ming ◽  
He Yinbiao ◽  
Tao Hongxin
Keyword(s):  
Stress Intensity Factor ◽  
Stress Intensity ◽  
Fracture Mechanics ◽  
Intensity Factor ◽  
Pressure Vessel ◽  
Solution Method ◽  
Reactor Pressure Vessel ◽  
Design Code ◽  
Outlet Nozzle ◽  
Reactor Pressure

This paper has discussed the stress intensity factor solution method of most popular used nuclear equipment design code and published papers. A series of inlet and outlet nozzle of reactor pressure vessel 3-D FEA fracture mechanics models with different size of corner flaw are created by ABQUAS software. Moreover, the crack front has been specially processed by ZENRCAK software. By compare the stress intensity factor solutions of FEA method and the solutions of influence function method for a 1/4 infinite symmetry plate, the influence functions for PWR reactor pressure vessel inlet and outlet nozzle corner flaw solution are obtained.

Evaluation of Pressure-Temperature Limit Curve Considering Revised Stress Intensity Factor Method

2020 ◽  
Author(s):  
Yunjoo Lee ◽  
Hyosub Yoon ◽  
Kyuwan Kim ◽  
Jongmin Kim ◽  
Hyunmin Kim
Keyword(s):  
Stress Intensity Factor ◽  
Stress Intensity ◽  
Intensity Factor ◽  
Pressure Vessel ◽  
Temperature Limit ◽  
Reactor Pressure Vessel ◽  
Flaw Size ◽  
Limit Curve ◽  
Asme Code ◽  
Reactor Pressure

Abstract Pressure-Temperature limit methodology is based on the rules of Appendix G in Section XI of the ASME Code in accordance with the requirements of 10 CFR 50, Appendix G, and the Appendix G in Section XI method refers to Welding Research Council (WRC) Bulletin 175 (WRC175). Flaw size is an important factor to protect the reactor pressure vessel from brittle failure but is not explicitly documented in WRC175. However, according to the recent change of Appendix G, the ¼ thickness (¼T) flaw size is postulated in the surface of the nozzle inner corner for the evaluation of Pressure-Temperature limit. In this paper, stress intensity factor is computed by using 3D finite element analysis (FEA) considering ¼T corner cracks of inlet nozzle and outlet nozzle in reactor pressure vessel. The result is compared with the stress intensity factor using influence function in the ASME Code. The results of stress intensity factor in accordance with the ASME Code are more conservative than those of the 3-D FEA with a crack. The allowable pressure and operation region in Pressure-Temperature limit curve are affected by the calculation methods of stress intensity factor.

Introduction of Warm Pre-Stress Concept in French RPV Structural Integrity Assessment: Part II — RSE-M Appendix

2016 ◽  
Author(s):  
Stéphane Marie ◽  
Stéphane Chapuliot ◽  
Dominique Moinereau ◽  
Malik Ait-Bachir ◽  
Clémentine Jacquemoud ◽  
...  
Keyword(s):  
Stress Intensity Factor ◽  
Stress Intensity ◽  
Intensity Factor ◽  
Pressure Vessel ◽  
Structural Integrity ◽  
Loading History ◽  
Main Criterion ◽  
Integrity Assessment ◽  
Service Operation ◽  
Reactor Pressure

A new appendix is introduced in the RSE-M code, devoted to in-service operation on PWRs, dealing with the assessment of a defect in the Reactor Pressure Vessel. This new appendix reflects the current French practice and introduces a second criterion to consider the Warm Pre-Stress (WPS) effect. This appendix is applicable to under clad defects and defects partially in the cladding, and covers nominal, incidental and accidental conditions. The main criterion is the classical comparison between the stress intensity factor (amplified to account for the plasticity of the cladding) and the material toughness (taking into account the irradiation induced ageing). For incidental and accidental situations, if the conventional criterion is not verified, an alternative criterion is proposed to take into account the WPS effect. The criterion corresponds to the ACE criterion developed by AREVA, CEA and EDF taking into account the effective material toughness depending on the loading history. The present paper presents this new RSE-M appendix and provides some basic elements of justification and validation on the ACE criterion.

scholarly journals Analisis Probabilistic Fracture Mechanics Pada Evaluasi Keandalan Bejana Tekan Reaktor Secara 3-D

2018 ◽  
Vol 24 (1) ◽  
Author(s):  
Entin Hartini ◽  
Roziq Himawan ◽  
Mike Susmikanti
Keyword(s):  
Fracture Toughness ◽  
Stress Intensity Factor ◽  
Stress Intensity ◽  
Probability Density Function ◽  
Fracture Mechanics ◽  
Intensity Factor ◽  
Pressure Vessel ◽  
J Integral ◽  
Probabilistic Fracture Mechanics ◽  
Reactor Pressure

Analisis integritas material sangat diperlukan pada Reactor Pressure Vessel (RPV). Komponen tersebut merupakan pressure boundary yang berfungsi untuk mengungkung material radioaktif. Adanya retak pada dinding dapat mempengaruhi integritas RPV tersebut. Penelitian ini bertujuan melakukan analisis fracture mechanics menggunakan model probabilistik untuk evaluasi keandalan RPV. Model probabilistik digunakan untuk pendekatan karakter random dari kuantitas input seperti sifat mekanik material dan lingkungan fisik. Karakter random dari kuantitas input menggunakan teknik sampling berdasarkan probability density function.  Material yang digunakan pada RPV adalah baja feritik (SA 533). Analisis fracture mechanics dilakukan berdasarkan metode elemen hingga (FEM) menggunakan perangkat lunak MSC MARC. Output dari MSC MARC adalah nilai J integral untuk mendapatkan nilai stress intensity factor (SIF) pada evaluasi keandalan bejana tekan reaktor 3D. Hasil perhitungan menunjukan bahwa SIF probabilistik lebih dulu mencapai nilai batas fracture toughness  dibanding  SIF deterministik. Nilai SIF yang dihasilkan dengan metode probabilistik adalah 95,8  MPa m0,5, sedangkan dengan metode deterministik adalah 91,8 MPa m0,5, rasio crack (a/c) semakin kecil akan dihasilkan nilai SIF yang semakin besar.Kata kunci: Probabilistic fracture mechanics, bejana tekan, 3-D.

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