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.

A novel method for diagnosis and de-noising of control rod drive mechanism within floating nuclear reactor

2022 ◽  
Vol 244 ◽  
pp. 110398
Author(s):  
Liming Zhang ◽  
Qiao Li ◽  
Jingdong Luo ◽  
Minghui Liu ◽  
Yiming Wang ◽  
...  
Keyword(s):  
Nuclear Reactor ◽  
Control Rod ◽  
Drive Mechanism ◽  
Novel Method

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.

Long-term stability of Sm/sub 2/Co/sub 17/-type magnets for control rod drive mechanism (CRDM) in a nuclear reactor

1995 ◽  
Vol 31 (6) ◽  
pp. 3653-3655 ◽  
Author(s):  
H. Iida ◽  
S. Imayoshi ◽  
K. Morimoto ◽  
M. Watanabe ◽  
N. Komada ◽  
...  
Keyword(s):  
Nuclear Reactor ◽  
Control Rod ◽  
Drive Mechanism ◽  
Term Stability ◽  
Long Term Stability

Dynamic Response Analysis of Control Rod Drive Mechanism Under the Stepping Impact Force for Nuclear Power Plants

2012 ◽  
Author(s):  
Xiaoyao Shen ◽  
Yongcheng Xie
Keyword(s):  
Dynamic Response ◽  
Nuclear Power ◽  
Power Plants ◽  
Structural Integrity ◽  
Impact Force ◽  
Response Analysis ◽  
Control Rod ◽  
Drive Mechanism ◽  
Finite Element Software ◽  
The Impact

The control rod drive mechanism (CRDM) is an important safety-related component in the nuclear power plant (NPP). When CRDM steps upward or downward, the pressure-containing housing of CRDM is shocked axially by an impact force from the engagement of the magnetic pole and the armature. To ensure the structural integrity of the primary coolant loop and the functionality of CRDM, dynamic response of CRDM under the impact force should be studied. In this manuscript, the commercial finite element software ANSYS is chosen to analyze the nonlinear impact problem. A nonlinear model is setup in ANSYS, including main CRDM parts such as the control rod, poles and armatures, as well as nonlinear gaps. The transient analysis method is adopted to calculate CRDM dynamic response when it steps upward. The impact loads and displacements at typical CRDM locations are successfully obtained, which are essential for design and stress analysis of CRDM.

Evaluation and Repair of Primary Water Stress Corrosion Cracking in Alloy 600/182 Control Rod Drive Mechanism Nozzles

2002 ◽  
Author(s):  
Charles R. Frye ◽  
Melvin L. Arey ◽  
Michael R. Robinson ◽  
David E. Whitaker
Keyword(s):  
Water Stress ◽  
Boric Acid ◽  
Stress Corrosion ◽  
Stress Corrosion Cracking ◽  
Base Material ◽  
Corrosion Cracking ◽  
Alloy 600 ◽  
Control Rod ◽  
Drive Mechanism ◽  
Primary Water

In February 2001, a routine visual inspection of the reactor vessel head of Oconee Nuclear Station Unit 3 identified boric acid crystals at nine of sixty-nine locations where control rod drive mechanism housings (CRDM nozzles) penetrate the head. The boric acid deposits resulted from primary coolant leaking from cracks in the nozzle attachment weld and from through-thickness cracks in the nozzle wall. A general overview of the inspection and repair process is presented and results of the metallurgical analysis are discussed in more detail. The analysis confirmed that primary water stress corrosion cracking (PWSCC) is the mechanism of failure of both the Alloy 182 weld filler material and the alloy 600 wrought base material.

A New Combined Deterministic and Probabilistic Flaw Evaluation Procedure

2003 ◽  
Author(s):  
Peter Dillstro¨m
Keyword(s):  
Residual Stresses ◽  
Computer Program ◽  
Crack Growth ◽  
Evaluation System ◽  
Safety Assessment ◽  
Safety Evaluation ◽  
Limit Load ◽  
Stable Crack Growth ◽  
Evaluation Procedure ◽  
Nuclear Applications

The Swedish handbook and computer program SACC, Safety Assessment of Components with Cracks, has recently been revised. The major differences compared to the last revision are within the following areas: • It is now a combined deterministic and probabilistic flaw evaluation procedure. • A new deterministic safety evaluation system is included. The main purpose is to more realistically handle large secondary stresses within the procedure. • It is now possible to include J-dominated stable crack growth. • New KI and limit load solutions are also included. • The appendices on material data to be used for nuclear applications and on residual stresses are revised. This paper provides an overview on the new handbook, with an emphasis on the probabilistic flaw evaluation procedure.

Experimental Investigations of Control Rod Drive Mechanism

2013 ◽  
Author(s):  
Guangyao Lu ◽  
Zhaohui Lu ◽  
Wenyuan Xiang ◽  
Yonghong Lv ◽  
Wenyou Huang ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Nuclear Reactor ◽  
Reactor Core ◽  
Experimental Investigations ◽  
Control Rod ◽  
Drive Mechanism ◽  
Start Up ◽  
Power Regulation ◽  
Set Up ◽  
Control Rods

The control rod drive mechanism (CRDM) is installed on the CRDM socket in reactor pressure vessel (RPV). Directed by Rod Control and Rod Position Indicating System (RGL), CRDM can impel the control rods move up and down in the nuclear reactor core, which implements the functions of reactor start-up, power regulation, power maintaining, normal reactor shutdown and abnormal (accident) shutdown. CRDM was developed by China Nuclear Power Research Institute (CNPRI). Several design improvements were conducted to solve the problems appeared in the operation of nuclear power station. Test bench was also set up and cold tests were carried out to investigate the characteristics of CRDM. The cold tests included lifting experiment, inserting experiment, rod drop experiment. And studies were carried out to analyze the signals of lifting coil, moving coil, stationary coil and the vibration signals. The test results show that the design of CRDM is reasonable and the operation is reliable.

Flaw Evaluation for PWR and BWR Component Weld Joints Using Advanced FEA Modeling Techniques

2009 ◽  
Author(s):  
T. Hayashi ◽  
S. F. Hankinson ◽  
T. Saito ◽  
C. K. Ng ◽  
W. H. Bamford
Keyword(s):  
Crack Propagation ◽  
Stress Corrosion ◽  
Crack Growth ◽  
Stress Corrosion Cracking ◽  
Structural Integrity ◽  
Corrosion Cracking ◽  
Water Reactor ◽  
Crack Shape ◽  
Weld Joints ◽  
Flaw Shape

Primary Water Stress Corrosion Cracking (PWSCC) of Pressurized Water Reactor (PWR) primary loop piping/nozzle Dissimilar Metal Weld (DMW) joints and Inter Granular Stress Corrosion Cracking (IGSCC) of Boiling Water Reactor (BWR) weld joints is an ongoing issue in the nuclear power industry. Recent field experiences with PWSCC of various DMW joints in US plants led to the development and application of an Advanced Finite Element Analyses (AFEA) methodology that permits crack propagation with a natural flaw shape. Crack growth and fracture evaluations for both PWR and BWR components are generally performed based on a conservative, idealized crack shape model, e.g. semi-ellipse, rectangle, etc., depending on the geometry of the crack and the component. Conventional evaluation methodologies and/or assumptions of this kind, in some cases may provide excessive conservatisms. The use of natural flaw shape development with crack propagation might provide a more realistic assessment of crack growth and structural integrity. The prime purpose of this study is to demonstrate the conservatism/margins in the conventional “idealized crack shape” methodology. A comparison study of crack growth behavior between the applications of the idealized and natural crack shape methodologies has been performed in order to assess the level of conservatism/margins in the conventional crack growth evaluation methodology and the possible impacts on the structural integrity evaluation for both PWR and BWR components. Comparison studies on the impacts of the differences in crack growth law and loading condition used for crack growth evaluations have been performed as well.

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.

Sign in / Sign up

Export Citation Format

Share Document