Fracture Mechanics Integrity Assessment of Stud Bolts Subjected to Cathodic Protection

2020 ◽  
Author(s):  
Mario A. L. de Castro ◽  
Fabio Alves ◽  
Kumarswamy Karpanan ◽  
Anand Venkatesh
Keyword(s):  
Fracture Mechanics ◽  
Cathodic Protection ◽  
Atomic Hydrogen ◽  
Sea Water ◽  
Circular Crack ◽  
Relative Increase ◽  
Maximum Increase ◽  
Integrity Assessment ◽  
Thread Root ◽  
Metallic Parts

Abstract Exposure of metallic parts to cathodic protection (CP) in sea water leads to production and diffusion of atomic Hydrogen into the metal matrix. Absorption of atomic Hydrogen into the metal could lead to hydrogen embrittlement (HE). In order to study the influence of stresses related to HE, FEA and Fracture Mechanics (FM) assessments were performed on a stud bolt threaded geometry. Effects of manufacturing tolerances, interface between nut and stud bolt and a defect in the form of a semi-circular crack placed in highest stress location of a thread root were also considered. Investigations of stress profiles when tension or bending are applied in test samples for measurement of HE threshold were also done, aiming at showing gaps on ASTM F1624-12 [1]. Tolerance assessment shows a relative maximum increase of 260% of nominal linearized membrane plus bending (NLMB) stresses regarding the nut runout [2] and for the proprietary nut geometry, such relative increase drops to 126% of NLMB stresses. Highest Hydrogen concentrations could be observed in the neighborhood of the first loaded thread root. FEA of cracked geometry shows that Hydrogen concentration could increase by around 283% around the crack tip, when compared to stud bolt in unloaded condition. Integrity assessment according to API 579-1 [3] or BS 7910 [4] and tests conducted according to ASTM F1624-12 [1] show less conservative results.

scholarly journals Cathodic protection of structures in seawater using solar panels

2019 ◽  
Vol 121 ◽  
pp. 02004
Author(s):  
Boris Borisovich Chernov ◽  
Van Mung Vu ◽  
Anac Maskharovich Nugmanov ◽  
Lyudmila Yuryevna Firsova
Keyword(s):  
Cathodic Protection ◽  
Sea Water ◽  
Cathode Current Density ◽  
Complete Agreement ◽  
Low Alloy Steels ◽  
Solar Panels ◽  
Protective Properties ◽  
Theoretical Concepts ◽  
Alloy Steels ◽  
Cathode Current

It is well known that the cathodic protection of structures in seawater is accompanied by the formation of calcareous deposits on them. In current study, we consider the physicochemical modelling of the formation of the deposit composition against cathode current density in seawater. The reliability of the model representations is confirmed by direct experiments. The work also studied the protective properties of the deposits with a different composition for low-alloy steels in natural sea water. It has been shown that the deposits of pure Mg(OH)2 and the deposits of CaCO3 + Mg(OH)2 had better protective ability against corrosion than the deposits of pure CaCO3. However, the deposits of Mg(OH)2 dissolved faster than the deposits of CaCO3 and CaCO3 + Mg(OH)2. Theoretical concepts and experiments on the laws governing the formation of the deposits and their protective properties are in complete agreement with each other. This allows to use the obtained patterns in the cathodic protection of structures in sea water using solar panels, forming standard deviations with predetermined protective properties in the daytime.

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.

Calibration of the AC Potential Dropping System (ACPD) for Determination of Crack Growth in API 5L X65 Steel under Cathodic Protection Effect

2020 ◽  
Vol 1012 ◽  
pp. 412-417
Author(s):  
Misael Souto de Oliveira ◽  
Antonio Almeida Silva ◽  
Marco Antonio dos Santos ◽  
Jorge Antonio Palma Carrasco ◽  
João Vitor de Queiroz Marques
Keyword(s):  
Crack Growth ◽  
Marine Environment ◽  
Cathodic Protection ◽  
Sea Water ◽  
Potential Drop ◽  
Mechanical Tests ◽  
Flow Density ◽  
Type Specimens ◽  
X65 Steel ◽  
Api 5L X65 Steel

In this work the calibration of an Alternative Current Potential Drop (ACPD) system was performed to monitore laboratory mechanical tests on marine environment under cathodic protection. The calibration was done on CT type specimens of API 5L X65 steel dimensioned according to ASTM E1820 standard., The crack propagation during a tensile test with displacement control in an ACPD equipment was monitored through the performs points collection by two channels: one that monitors the crack growth and another that monitors a region free of crack. Using a profile projector and graphical data processing and analysis software, the area of ​​the fracture surface of the specimen was meansured, which allowed to correlate a crack size with a corresponding value of potential drop and the calibration curve. In order to verify verify the efficacy and precision of the technique, step loading tests were performed on API 5L X65 steel test specimens, submerged in synthetic sea water under the overprotection potential of-1300mVAg/AgCl. The results of the calibration showed few dispersed errors, and the main factors of this dispersion may be related to the geometry of the specimen and with variations in current flow density, which is influenced by corners and edges and by the presence of pick-up inductive. The calibration and its effectiveness can be verified through the results of the tests in marine environment, presenting crack lengths close to the actual values, confirming the effectiveness of the ACPD technique.

Deterministic and Probabilistic Assessments of the Reactor Pressure Vessel Structural Integrity: Benchmark Comparisons

2003 ◽  
Author(s):  
Silvia Turato ◽  
Vincent Venturini ◽  
Eric Meister ◽  
B. Richard Bass ◽  
Terry L. Dickson ◽  
...  
Keyword(s):  
Fracture Mechanics ◽  
Pressure Vessel ◽  
Structural Integrity ◽  
Safety Concern ◽  
Nuclear Regulatory Commission ◽  
Reactor Pressure Vessel ◽  
Oak Ridge ◽  
Integrity Assessment ◽  
Probabilistic Fracture Mechanics ◽  
Reactor Pressure

The structural integrity assessment of a nuclear Reactor Pressure Vessel (RPV) during accidental conditions, such as loss-of-coolant accident (LOCA), is a major safety concern. Besides Conventional deterministic calculations to justify the RPV integrity, Electricite´ de France (EDF) carries out probabilistic analyses. Since in the USA the probabilistic fracture mechanics analyses are accepted by the Nuclear Regulatory Commission (NRC), a benchmark has been realized between EDF and Oak Ridge Structural Assessments, Inc. (ORSA) to compare the models and the computational methodologies used in respective deterministic and probabilistic fracture mechanics analyses. Six cases involving two distinct transients imposed on RPVs containing specific flaw configurations (two axial subclad, two circumferential surface-breaking, and two axial surface-braking flaw configurations) were defined for a French vessel. In two separate phases, deterministic and probabilistic, fracture mechanics analyses were performed for these six cases.

NESC-VII: Fracture Mechanics Analyses of WPS Experiments on Large-Scale Cruciform Specimen

2011 ◽  
Author(s):  
Shengjun Yin ◽  
Paul T. Williams ◽  
B. Richard Bass
Keyword(s):  
Fracture Mechanics ◽  
Large Scale ◽  
Structural Integrity ◽  
National Laboratory ◽  
Pressure Vessels ◽  
Cooperative Action ◽  
Oak Ridge ◽  
Integrity Assessment ◽  
Feasibility Test ◽  
Reactor Pressure

This paper describes numerical analyses performed to simulate warm pre-stress (WPS) experiments conducted with large-scale cruciform specimens within the Network for Evaluation of Structural Components (NESC-VII) project. NESC-VII is a European cooperative action in support of WPS application in reactor pressure vessel (RPV) integrity assessment. The project aims in evaluation of the influence of WPS when assessing the structural integrity of RPVs. Advanced fracture mechanics models will be developed and performed to validate experiments concerning the effect of different WPS scenarios on RPV components. The Oak Ridge National Laboratory (ORNL), USA contributes to the Work Package-2 (Analyses of WPS experiments) within the NESC-VII network. A series of WPS type experiments on large-scale cruciform specimens have been conducted at CEA Saclay, France, within the framework of NESC VII project. This paper first describes NESC-VII feasibility test analyses conducted at ORNL. Very good agreement was achieved between AREVA NP SAS and ORNL. Further analyses were conducted to evaluate the NESC-VII WPS tests conducted under Load-Cool-Transient-Fracture (LCTF) and Load-Cool-Fracture (LCF) conditions. This objective of this work is to provide a definitive quantification of WPS effects when assessing the structural integrity of reactor pressure vessels. This information will be utilized to further validate, refine, and improve the WPS models that are being used in probabilistic fracture mechanics computer codes now in use by the NRC staff in their effort to develop risk-informed updates to Title 10 of the U.S. Code of Federal Regulations (CFR), Part 50, Appendix G.

Estimation of Through-Wall Cracking Frequency of RPV Under PTS Events Using PFM Analysis Method for Identifying Conservatism Included in Current Japanese Code

2014 ◽  
Author(s):  
Kazuya Osakabe ◽  
Koichi Masaki ◽  
Jinya Katsuyama ◽  
Genshichiro Katsumata ◽  
Kunio Onizawa
Keyword(s):  
Fracture Toughness ◽  
Fracture Mechanics ◽  
Crack Initiation ◽  
Structural Integrity ◽  
Probabilistic Approach ◽  
Pressure Vessels ◽  
Assessment Procedure ◽  
Analysis Method ◽  
Integrity Assessment ◽  
The U.S

To assess the structural integrity of reactor pressure vessels (RPVs) during pressurized thermal shock (PTS) events, the deterministic fracture mechanics approach prescribed in Japanese code JEAC 4206-2007 [1] has been used in Japan. The structural integrity is judged to be maintained if the stress intensity factor (SIF) at the crack tip during PTS events is smaller than fracture toughness KIc. On the other hand, the application of a probabilistic fracture mechanics (PFM) analysis method for the structural reliability assessment of pressure components has become attractive recently because uncertainties related to influence parameters can be incorporated rationally. A probabilistic approach has already been adopted as the regulation on fracture toughness requirements against PTS events in the U.S. According to the PFM analysis method in the U.S., through-wall cracking frequencies (TWCFs) are estimated taking frequencies of event occurrence and crack arrest after crack initiation into consideration. In this study, in order to identify the conservatism in the current RPV integrity assessment procedure in the code, probabilistic analyses on TWCF have been performed for certain model of RPVs. The result shows that the current assumption in JEAC 4206-2007, that a semi-elliptic axial crack is postulated on the inside surface of RPV wall, is conservative as compared with realistic conditions. Effects of variation of PTS transients on crack initiation frequency and TWCF have been also discussed.

The Experimental Work in Support to ATLAS+ Project

2018 ◽  
Author(s):  
Dominique Moinereau ◽  
Patrick Le Delliou ◽  
Anna Dahl ◽  
Yann Kayser ◽  
Szabolcs Szavai ◽  
...  
Keyword(s):  
Fracture Mechanics ◽  
Experimental Work ◽  
Large Scale ◽  
Nuclear Reactor ◽  
Structural Integrity ◽  
Assessment Tools ◽  
Ductile Material ◽  
Small Scale ◽  
Term Operation ◽  
Integrity Assessment

The 4-years European project ATLAS+ project was launched in June 2017. Its main objective is to develop advanced structural assessment tools to address the remaining technology gaps for the safe and long term operation of nuclear reactor pressure coolant boundary systems. The transferability of ductile material properties from small scale fracture mechanics specimens to large scale components is one of the topics of the project. A large programme of experimental work is to be conducted in support of the development and validation of advanced tools for structural integrity assessment within the framework of the work-package 1 (WP 1): Design and execution of simulation oriented experiments to validate models at different scales. The experimental work is based on a full set of fracture mechanics experiments conducted on standard specimens and large scale components (several pipes and one mock-up), including a full materials characterization. Three materials are considered: • a ferritic steel 15NiCuMoNb5 (WB 36) • an aged austenitic stainless steel weld • a VVER (eastern PWR) dissimilar metal weld (DMW) The paper presents the WP 1, the experimental programme and summarizes the first results.

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