On Safety Assessment of Structures Containing Cracks: Deterministic Analysis

2014 ◽  
Author(s):  
Lingfu Zeng ◽  
Lennart G. Jansson
Keyword(s):  
Stress Corrosion ◽  
Crack Growth ◽  
Nuclear Power ◽  
Safety Assessment ◽  
Radiation Safety ◽  
Tolerance Analysis ◽  
Defect Tolerance ◽  
Deterministic Analysis ◽  
Alternative Procedures ◽  
Effective Assessment

In this paper, safety assessment of structures of nuclear power facilities, containing cracks or crack-like defects, is addressed. A so-called deterministic procedure recommended by Swedish Radiation Safety Agency is reviewed and discussed. The procedure provides a possibility for evaluation of crack growth due to courses of fatigue and stress corrosion, and for prediction of safety reservation margin due to fracture and plastic failure. The review and discussion focus on various assumptions and limitations behind this procedure, and cares that must be taken of when using this procedure for safety assessment. Applications recently conducted for a defect tolerance analysis of aging BWR components are used as examples to address the reliability and accuracy of the assessment and its relevance to rules given by ASME BPV code. It is concluded that for some cases the reliability of such an assessment need to be carefully questioned and there is a great need for alternative procedures for fully achieving a reliable, accurate and effective assessment.

scholarly journals Interaction of Cyclic Loading (Low-Cyclic Fatigue) with Stress Corrosion Cracking (SCC) Growth Rate

2020 ◽  
Vol 2020 ◽  
pp. 1-10
Author(s):  
Rehmat Bashir ◽  
He Xue ◽  
Rui Guo ◽  
Yueqi Bi ◽  
Muhammad Usman
Keyword(s):  
Growth Rate ◽  
Stress Corrosion ◽  
Crack Growth ◽  
Stress Corrosion Cracking ◽  
Nuclear Power ◽  
Power Plants ◽  
Structural Integrity ◽  
Cyclic Fatigue ◽  
Joint Effect ◽  
Corrosion Cracking

The structural integrity analysis of nuclear power plants (NPPs) is an essential procedure since the age of NPPs is increasing constantly while the number of new NPPs is still limited. Low-cyclic fatigue (LCF) and stress corrosion cracking (SSC) are the two main causes of failure in light-water reactors (LWRs). In the last few decades, many types of research studies have been conducted on these two phenomena separately, but the joint effect of these two mechanisms on the same crack has not been discussed yet though these two loads exist simultaneously in the LWRs. SCC is mainly a combination of the loading, the corrosive medium, and the susceptibility of materials while the LCF depends upon the elements such as compression, moisture, contact, and weld. As it is an attempt to combine SCC and LCF, this research focuses on the joint effect of SCC and LCF loading on crack propagation. The simulations are carried out using extended finite element method (XFEM) separately, for the SCC and LCF, on an identical crack. In the case of SCC, da/dt(mm/sec) is converted into da/dNScc (mm/cycle), and results are combined at the end. It has been observed that the separately calculated results for SCC da/dNScc and LCF da/dNm of crack growth rate are different from those of joint/overall effect,  da/dNom. By applying different SCC loads, the overall crack growth is measured as SCC load becomes the main cause of failure in LWRs in some cases particularly in the presence of residual stresses.

Technical Basis for Revision to ASME Section XI Appendix C for Stress Corrosion Crack Growth Rate Equations for Alloy 600 and Associated Welds

2020 ◽  
Author(s):  
Amanda Jenks ◽  
Warren H. Bamford
Keyword(s):  
Growth Rate ◽  
Crack Growth Rate ◽  
Stress Corrosion ◽  
Crack Growth ◽  
Nuclear Power ◽  
Expert Panel ◽  
Degradation Mechanism ◽  
Model Development ◽  
Rate Equations ◽  
Alloy 600

Abstract Alloy 600 and its associated weld metals (Alloys 82, 182, and 132) have been used extensively for nuclear power plant construction. These alloys are known to be susceptible to primary water stress corrosion cracking (PWSCC), and reference crack growth rate (CGR) equations for this degradation mechanism were first incorporated into the ASME Boiler and Pressure Vessel Code in the 2009 Addenda of the 2007 Edition via Section XI, Nonmandatory Appendix C. These original equations were developed from a large dataset to which numerous international laboratories contributed. It was reviewed and assessed by an Expert Panel. In 2015, another Expert Panel was organized, including members from the original Expert Panel. Publicly available data produced since the first equations were developed, and data that had not been part of the original model development process, were collected. The data were screened for testing quality and plant applicability by the Expert Panel, and new PWSCC equations were developed. This paper summarizes the development of the revised PWSCC equations for Alloy 600 and Alloy 82/182/132 to support revision of the equations previously incorporated into Appendix C of ASME B&PV Code Section XI.

Crack Growth Safety Assessment of Overlay Weld on the Ω Seal for Control Rod Drive Mechanism

2017 ◽  
Author(s):  
Jiacheng Luo ◽  
Yong Zhang ◽  
Pengzhou Li ◽  
Juan Luo ◽  
Lei Sun
Keyword(s):  
Stress Corrosion ◽  
Crack Growth ◽  
Safety Assessment ◽  
Nuclear Reactor ◽  
Structural Integrity ◽  
Safety Evaluation ◽  
Structure Design ◽  
Control Rod ◽  
Drive Mechanism ◽  
Weld Structure

As a key device, the Control Rod Drive Mechanism (CRDM) is of great important for the safety operation of the nuclear reactor, where the pressure shell of CRDM is exposed to the primary loop pressure, and is installed on the head of the reactor pressure vessel by the Ω seal. During the operation cycle of the reactor, cracks may be induced on the Ω seal weld by primary water stress corrosion cracking which is the major factor leading to leakage of coolant and impacting the safety of the reactor. Therefore, the Ω seal is required to be repaired to maintain structural integrity and safety by using the evaluated overlay weld technology. Based on fracture mechanics, this paper investigates the crack growth safety assessment of the overlay weld structure on the Ω seal of control rod drive mechanism, induced by the operation fatigue and stress corrosion, and the evaluation is completed by the standard code. The calculation method and assessment results can be a reference for the overlay weld structure design and safety evaluation of the Ω seal in CRDM.

scholarly journals Mathematical Formulation and Analytic Solutions for Uncertainty Analysis in Probabilistic Safety Assessment of Nuclear Power Plants

Energies ◽  
2021 ◽  
Vol 14 (4) ◽  
pp. 929
Author(s):  
Gyun Seob Song ◽  
Man Cheol Kim
Keyword(s):  
Monte Carlo ◽  
Uncertainty Analysis ◽  
Probability Density ◽  
Nuclear Power ◽  
Safety Assessment ◽  
Power Plants ◽  
Mathematical Formulation ◽  
Analytic Solutions ◽  
Probabilistic Safety Assessment ◽  
Probabilistic Safety

Monte Carlo simulations are widely used for uncertainty analysis in the probabilistic safety assessment of nuclear power plants. Despite many advantages, such as its general applicability, a Monte Carlo simulation has inherent limitations as a simulation-based approach. This study provides a mathematical formulation and analytic solutions for the uncertainty analysis in a probabilistic safety assessment (PSA). Starting from the definitions of variables, mathematical equations are derived for synthesizing probability density functions for logical AND, logical OR, and logical OR with rare event approximation of two independent events. The equations can be applied consecutively when there exist more than two events. For fail-to-run failures, the probability density function for the unavailability has the same probability distribution as the probability density function (PDF) for the failure rate under specified conditions. The effectiveness of the analytic solutions is demonstrated by applying them to an example system. The resultant probability density functions are in good agreement with the Monte Carlo simulation results, which are in fact approximations for those from the analytic solutions, with errors less than 12.6%. Important theoretical aspects are examined with the analytic solutions such as the validity of the use of a right-unbounded distribution to describe the uncertainty in the unavailability/probability. The analytic solutions for uncertainty analysis can serve as a basis for all other methods, providing deeper insights into uncertainty analyses in probabilistic safety assessment.

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