RCC-MRx Appendix A16 Methodology for the Analytical J Calculation Under Thermal and Combined Thermal + Mechanical Loadings for Pipes and Elbows and Related Assessment Tool MJSAM

2011 ◽  
Author(s):  
S. Marie ◽  
M. Ne´de´lec ◽  
C. Delaval
Keyword(s):  
Nuclear Power ◽  
Assessment Tool ◽  
Analytical Calculation ◽  
Large Set ◽  
Microsoft Excel ◽  
Defect Assessment ◽  
Initiation And Propagation ◽  
Creep Fatigue ◽  
Assessment Procedures ◽  
2D And 3D

RCC-MRx code provides flaw assessment methodologies and related tools for Nuclear Power Plant cracked components. An important work has been made in particular to develop a large set of compendia for the calculation of the parameter J for various components (plates, pipes, elbows,…) and various defect geometries. Also, CEA in the frame of collaborations with IRSN, developed a methodology for J analytical calculation for cracked pipes and elbows submitted to thermal and combined mechanical and thermal loadings. This paper presents first the development of this methodology and an overview of the validation strategy, based on reference 2D and 3D F.E. calculations. The second part of the paper presents the last version of the MJSAM tools which is based on the 2010 version of the appendix A16 of the RCC-MRx code. All compendia (for KI, J and C* calculation) and all defect assessment procedures have been implemented in the tool: It covers crack initiation and propagation under fatigue, creep, creep-fatigue and ductile tearing situations. Sensitivity and probabilistic analyses can also been performed with this tool, directly linked to Microsoft Excel software for the results exploitation.

Defect Assessment Procedure for Low to High Temperature Range

2003 ◽  
Author(s):  
Bilal Dogan ◽  
Robert Ainsworth
Keyword(s):  
High Temperature ◽  
Growth Conditions ◽  
Life Assessment ◽  
Assessment Procedure ◽  
High Temperature Range ◽  
Temperature Range ◽  
Defect Assessment ◽  
Creep Fatigue ◽  
Recent Developments ◽  
Assessment Procedures

There are many similarities between available procedures used for defect assessment. They have been developed as a result of experience gained from material-specific programs and have often been verified using the same data. One recently updated document covering life assessment procedures under creep and creep/fatigue crack growth conditions is BS 7910. This document takes into account some of the most recent developments in the subject, including some from the British Energy R5 Procedure. Future developments in defect assessment procedures will follow the route of simplified and unified codes covering defect behaviour in the low to high temperature range. In this paper, the relevance of the insignificant creep curves in RCC-MR for defect free structures and the creep exemption criteria in BS7910 are examined. Then, an overview is given of some European developments in defect assessment methods for Fitness-for-Service assessment, based on recent and current projects such as the EC thematic network FITNET.

Fully-Plastic J and CTOD Solutions for Pipes With Circumferential Surface Cracks Subjected to Combined Bending and Tension

2011 ◽  
Author(s):  
Lui´s F. S. Parise ◽  
Claudio Ruggieri
Keyword(s):  
Material Properties ◽  
Tensile Load ◽  
Estimation Procedure ◽  
Large Set ◽  
J Integral ◽  
Surface Cracks ◽  
Defect Assessment ◽  
Wide Range ◽  
Bending Load ◽  
Assessment Procedures

This work provides an estimation procedure to determine the J-integral and CTOD for pipes with circumferential surface cracks subjected to combined bending and tensile load for a wide range of crack geometries and material (hardening) based upon fully-plastic solutions. A summary of the methodology upon which J and CTOD are derived sets the necessary framework to determine nondimensional functions h1 and h2 applicable to a wide range of crack geometries and material properties characteristic of structural, pressure vessel and pipeline steels. The extensive nonlinear, 3-D numerical analyses provide a large set of solutions for J and CTOD which enters directly into fitness-for-service (FFS) analyses and defect assessment procedures of cracked pipes and cylinders subjected to bending load.

J and CTOD Estimation Procedure for Circumferentially Cracked Pipes Based on Fully-Plastic Solutions

2010 ◽  
Author(s):  
Mario S. G. Chiodo ◽  
Claudio Ruggieri
Keyword(s):  
Pressure Vessel ◽  
Material Properties ◽  
Estimation Procedure ◽  
Large Set ◽  
J Integral ◽  
Surface Cracks ◽  
Defect Assessment ◽  
Wide Range ◽  
Bending Load ◽  
Assessment Procedures

This work provides an estimation procedure to determine the J-integral and CTOD for pipes with circumferential surface cracks subjected to bending load for a wide range of crack geometries and material (hardening) based upon fully-plastic solutions. A summary of the methodology upon which J and CTOD are derived sets the necessary framework to determine nondimensional functions h1 and h2 applicable to a wide range of crack geometries and material properties characteristic of structural, pressure vessel and pipeline steels. The extensive nonlinear, 3-D numerical analyses provide a large set of solutions for J and CTOD which enters directly into fitness-for-service (FFS) analyses and defect assessment procedures of cracked pipes and cylinders subjected to bending load.

Heat Transfer Correlation for Supercritical Carbon Dioxide Flowing in Vertical Bare Tubes

2013 ◽  
Author(s):  
Sahil Gupta ◽  
Eugene Saltanov ◽  
Igor Pioro
Keyword(s):  
Heat Transfer ◽  
Error Analysis ◽  
Nuclear Power ◽  
Nuclear Reactor ◽  
Large Set ◽  
Transfer Coefficients ◽  
Data Points ◽  
Institute Of Technology ◽  
Using Data ◽  
Gen Iv

Canada among many other countries is in pursuit of developing next generation (Generation IV) nuclear-reactor concepts. One of the main objectives of Generation-IV concepts is to achieve high thermal efficiencies (45–50%). It has been proposed to make use of SuperCritical Fluids (SCFs) as the heat-transfer medium in such Gen IV reactor design concepts such as SuperCritical Water-cooled Reactor (SCWR). An important aspect towards development of SCF applications in novel Gen IV Nuclear Power Plant (NPP) designs is to understand the thermodynamic behavior and prediction of Heat Transfer Coefficients (HTCs) at supercritical (SC) conditions. To calculate forced convection HTCs for simple geometries, a number of empirical 1-D correlations have been proposed using dimensional analysis. These 1-D HTC correlations are developed by applying data-fitting techniques to a model equation with dimensionless terms and can be used for rudimentary calculations. Using similar statistical techniques three correlations were proposed by Gupta et al. [1] for Heat Transfer (HT) in SCCO2. These SCCO2 correlations were developed at the University of Ontario Institute of Technology (Canada) by using a large set of experimental SCCO2 data (∼4,000 data-points) obtained at the Chalk River Laboratories (CRL) AECL. These correlations predict HTC values with an accuracy of ±30% and wall temperatures with an accuracy of ±20% for the analyzed dataset. Since these correlations were developed using data from a single source - CRL (AECL), they can be limited in their range of applicability. To investigate the tangible applicability of these SCCO2 correlations it was imperative to perform a thorough error analysis by checking their results against a set of independent SCCO2 tube data. In this paper SCCO2 data are compiled from various sources and within various experimental flow conditions. HTC and wall-temperature values for these data points are calculated using updated correlations presented in [1] and compared to the experimental values. Error analysis is then shown for these datasets to obtain a sense of the applicability of these updated SCCO2 correlations.

A Heuristic for Designing Hybrid Compact Heat Exchangers for Nuclear Applications

2021 ◽  
Author(s):  
Venkata Rajesh Saranam ◽  
Peter Carter ◽  
Kyle Rozman ◽  
Ömer Dogan ◽  
Brian K. Paul
Keyword(s):  
Heat Exchanger ◽  
Diffusion Bonding ◽  
Heat Exchangers ◽  
Nuclear Power ◽  
Nuclear Industry ◽  
Compact Heat Exchangers ◽  
Bonding Parameters ◽  
Heat Exchanger Design ◽  
Creep Fatigue ◽  
Gen Iv

Abstract Hybrid compact heat exchangers (HCHEs) are a potential source of innovation for intermediate heat exchangers in nuclear industry, with HCHEs being designed for Gen-IV nuclear power applications. Compact heat exchangers are commonly fabricated using diffusion bonding, which can provide challenges for HCHEs due to resultant non-uniform stress distributions across hybrid structures during bonding, leading to variations in joint properties that can compromise performance and safety. In this paper, we introduce and evaluate a heuristic for determining whether a feasible set of diffusion bonding conditions exist for producing HCHE designs capable of meeting regulatory requirements under nuclear boiler and pressure vessel codes. A diffusion bonding model for predicting pore elimination and structural analyses are used to inform the heuristic and a heat exchanger design for 316 stainless steel is used to evaluate the efficacy of the heuristic to develop acceptable diffusion bonding parameters. A set of diffusion bonding conditions were identified and validated experimentally by producing various test coupons for evaluating bond strength, ductility, porosity, grain size, creep rupture, creep fatigue and channel deviation. A five-layer hybrid compact heat exchanger structure was fabricated and tensile tested demonstrating that the bonding parameters satisfy all criteria in this paper for diffusion bonding HCHEs with application to the nuclear industry.

scholarly journals Measuring implementation: development of the implementation process assessment tool (IPAT)

2019 ◽  
Vol 19 (1) ◽  
Author(s):  
M. Hartveit ◽  
E. Hovlid ◽  
M. H. A. Nordin ◽  
J. Øvretveit ◽  
G. R. Bond ◽  
...  
Keyword(s):  
Internal Consistency ◽  
Stages Of Change ◽  
Assessment Tool ◽  
Item Analysis ◽  
Behavioural Change ◽  
Implementation Process ◽  
Large Set ◽  
Process Assessment ◽  
Care Providers ◽  
Scale Scores

Abstract Background Implementation science comprises a large set of theories suggesting interacting factors at different organisational levels. Development of literature syntheses and frameworks for implementation have contributed to comprehensive descriptions of implementation. However, corresponding instruments for measuring these comprehensive descriptions are currently lacking. The present study aimed to develop an instrument measuring care providers’ perceptions of an implementation effort, and to explore the instrument’s psychometric properties. Methods Based on existing implementation literature, a questionnaire was designed with items on individual and team factors and on stages of change in an implementation process. The instrument was tested in a Norwegian study on implementation of evidence based practices for psychosis. Item analysis, factor structure, and internal consistency at baseline were examined. Results The 27-item Implementation Process Assessment Tool (IPAT) revealed large variation between mean score of the items. The total scale scores were widely dispersed across respondents. Internal consistency for the total scale was high (Cronbach’s alpha: .962), and all but one item contributed positively to the construct. The results indicated four underlying constructs: individual stages for behavioural change, individual activities and perceived support, collective readiness and support, and individual perceptions of the intervention. Conclusions The IPAT appears to be a feasible instrument for investigating the implementation process from the perspective of those making the change. It can enable examination of the relative importance of factors thought to be essential for implementation outcomes. It may also provide ongoing feedback for leaders tailoring support for teams to improve implementation. However, further research is needed to detect the instrument’s properties later in the implementation process and in different contexts. Trial registration ClinicalTrials.gov code NCT03271242 (retrospective registered September 5, 2017).

Numerical Validation of a Strain-Based Failure Assessment Diagram Approach to Fracture

2009 ◽  
Author(s):  
Peter J. Budden ◽  
Michael C. Smith
Keyword(s):  
Finite Element ◽  
Plastic Strain ◽  
Present Author ◽  
Plastic Relaxation ◽  
Failure Assessment Diagram ◽  
Defect Assessment ◽  
Failure Assessment ◽  
Basic Approaches ◽  
Assessment Procedures ◽  
Diagram Approach

The basic approaches in defect assessment procedures such as R6 consider the stresses on the section containing the flaw. Such approaches can be overly conservative and lead to unacceptably small estimates of limiting defect sizes for cases where the applied loads are due to displacements or strains well in excess of yield, when significant plastic relaxation of stress occurs. The potential for over-conservative assessments has led to a renewed interest in recent years in strain-based assessment methods, in both the power and pipeline industries. Significant levels of plastic strain can be imposed across the flawed section in some cases. Recently, the present author has published a general approach to strain-based fracture that uses a strain-based failure assessment diagram (SB-FAD). This includes a range of Options similar to that of the basic R6 approach. The present paper describes some validation of the SB-FAD approach based on elastic-plastic cracked-body finite element data for plates and cylinders.

Analytical Method for the Calculation of J Parameter in Cracked Components: Presentation of the RSE-M/RCC-MR Procedure and Benchmark Proposal

2010 ◽  
Author(s):  
S. Marie ◽  
B. Drubay ◽  
P. Le Delliou ◽  
S. Chapuliot ◽  
H. Deschanels ◽  
...  
Keyword(s):  
Power Plant ◽  
Nuclear Power ◽  
Strain Curve ◽  
Equivalent Material ◽  
Large Set ◽  
Stress Strain Curve ◽  
Weld Joints ◽  
Joint Geometry ◽  
Material Component ◽  
Definition Of

RSE-M and RCC-MR codes provide flaw assessment methodologies and related tools for Nuclear Power Plant. For these two codes, AREVA, CEA and EDF developed a large set of compendia for the calculation of the parameter J for various components (plates, pipes, elbows, …) and various defect geometries. The last step of these developments deals with the weld joints: since 2004, a methodology have been developed to calculate the J parameter for a defect located in a weld, less conservative than usual methods. This methodology is based on the definition of an equivalent material that leads to the same J value (with same loading conditions and defect geometry) than the bi-material component. The stress-strain curve of this equivalent material is deduced from a combination of the tensile curves of the base metal and of the weld metal. The weigth coefficients applied are specifically defined for the J calculation and generalized to deal with any weld joint geometry.

Methodology to Evaluate Nuclear Power Plant Vulnerability to Events in the Transmission Grid

2014 ◽  
Author(s):  
Alberto Del Rosso ◽  
Jean-François Roy ◽  
Frank Rahn ◽  
Alejandro Capara
Keyword(s):  
Risk Assessment ◽  
Nuclear Power ◽  
Power Plants ◽  
Assessment Tool ◽  
Transmission Grid ◽  
Simulation Tools ◽  
Nuclear Plants ◽  
Western Electricity Coordinating Council ◽  
The Impact ◽  
The U.S

This paper presents a general approach to evaluate the risk of trip or Loss of Off-site Power (LOOP) events in nuclear power plants due to contingencies in the power grid. The proposed methodology is based on the Zone of Vulnerability concept for nuclear plants introduced by EPRI in previous work. The proposed methodology is intended to be part of an integrated probabilistic risk assessment tool that is being developed under ongoing EPRI R&D programs. A detailed analysis of many events occurred in actual nuclear plants has been performed in order to identify, classify and characterize the various vulnerability and type of failures that may affect a nuclear plant. Based the outcome of that analysis, a methodology for evaluating the impact of off-site transmission system events on nuclear plants has been outlined. It includes description of the type of contingencies and conditions that need to be included in the analysis, as well as provisions regarding the simulation tools and models that should be used in each case. The methodology is illustrated in a simplified representation of the Western Electricity Coordinating Council (WECC) system in the U.S.

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