PSI - Issue 5

ScienceDirect Available online at www.sciencedirect.com Available o line at www.sciencedire t.com ScienceDirect Structural Integrity Procedia 00 (2016) 000 – 000 Procedia Structu al Integrity 5 (2017) 294–301 Available online at www.sciencedirect.com ScienceDirect StructuralIntegrity Procedia 00 (2017) 000 – 000 Available online at www.sciencedirect.com ScienceDirect StructuralIntegrity Procedia 00 (2017) 000 – 000

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XV Portuguese Conference on Fracture, PCF 2016, 10-12 February 2016, Paço de Arcos, Portugal Thermo-mechanical modeling of a high pressure turbine blade of an airplane gas turbine engine P. Brandão a , V. Infante b , A.M. Deus c * a Department of Mechanical Engineering, Instituto Superior Técnico, Universidade de Lisboa, Av. Rovisco Pais, 1, 1049-001 Lisboa, Portugal b IDMEC, Department of Mechanical Engineering, Instituto Superior Técnico, Universidade de Lisboa, Av. Rovisco Pais, 1, 1049-001 Lisboa, Portugal c CeFEMA, Department of Mechanical Engineering, Instituto Superior Técnico, Universidade de Lisboa, Av. Rovisco Pais, 1, 1049-001 Lisboa, Portugal Abstract During their operation, modern aircraft engine components are subjected to increasingly demanding operating conditions, especially the high pressure turbine (HPT) blades. Such conditions cause these parts to undergo different types of time-dependent degradation, one of which is creep. A model using the finite element method (FEM) was developed, in order to be able to predict the creep behaviour of HPT blades. Flight data records (FDR) for a specific aircraft, provided by a commercial aviation company, were used to obtain thermal and mechanical data for three different flight cycles. In order to create the 3D model needed for the FEM analysis, a HPT blade scrap was scanned, and its chemical composition and material properties were obtained. The data that was gathered was fed into the FEM model and different simulations were run, first with a simplified 3D rectangular block shape, in order to better establish the model, and then with the real 3D mesh obtained from the blade scrap. The overall expected behaviour in terms of displacement was observed, in particular at the trailing edge of the blade. Therefore such a model can be useful in the goal of predicting turbine blade life, given a set of FDR data. 2nd International Conference on Structural Integrity, ICSI 2017, 4-7 September 2017, Funchal, Madeira, Portugal Assessing the Irradiation Defect Induced Changes using Dislocation Based Crystal Plasticity Model for BCC Materials Kulbir Singh a *, C. Robertson b , A.K. Bhaduri a a Indira Gandhi Centre for Atomic Research, Kalpakkam, India 603 102 b DEN-Service de Recherches Metallurgiques Appliquees, CEA, Universite Paris-Saclay, Gif-sur-Yvette, France Radiation effects lead to significant reduction in ductility of nuclear reactor components. Body Centered Cubic (BCC) materials exhibits better thermal properties as compared to Face Centered Cubic (FCC) materials, in addition to excellent resistance to helium embrittlement and void swelling under higher dpa levels, making them the promising materials for future nuclear applications. Fracture toughness in BCC has a well defined temperature dependent transition as compared to FCC materials, which is a major concern restricting their application to high dose conditions. In the present paper such strong temperature dependence of strain rate and flow stress in BCC materials is investigated numerically for both non-irradiated and irradiated conditions. In view of dislocations mobility being a fundamental property to determine the plastic behavior, a dislocation based material model is proposed which is based on physical approach rather than phenomenological. © 2017 The Authors. Published by Elsevier B.V. Peer-review under responsibility of the Scientific Committee of ICSI 2017. Keywords: : Dislocations; Crystal Plasticity; Dislocation mobility; Screw dislocation length; Irradiation defects; Slip system 2nd International Conference on Structural Integrity, ICSI 2017, 4-7 September 2017, Funchal, Madeira, Portugal Assessing the Irradiation Defect Induced Changes using Dislocati n Based Crystal Plasticity Model for BCC Materials Kulbir Singh a *, C. Robertson b , A.K. Bhaduri a a Indira Gandhi Centre for Atomic Research, Kalpakkam, India 603 102 b DEN-Service de Recherches Metallurgiques Appliquees, CEA, Universite Paris-Saclay, Gif-sur-Yvette, France Abstract Radiation effects lead to significant reduction in ductility of nuclear reactor components. Body Centered Cubic (BCC) mat rials xhibits better thermal prop rties as compared to F ce Centered Cubic (FCC) materials, in addition t excellent resistance to helium embrittlement and void swelling under higher dpa levels, maki g them the promising materials for future nuclear applications. Fracture toughness in BCC has a well defined temperature dependent transition as compared to FCC mat rials, which is a major concern restricting their application to high dose conditions. In the present paper such strong temperature dependence of strain rate and flow str ss in BCC materials is i vestigated nu erically for both non-irradiated a d irradiated conditions. In view f dislocations mobility b ing a fundamental property to determin the plastic behavior, a isl cation based material mod l is propose which is based on physical approach rather t an phenomenological. © 2017 The Authors. Published by Elsevier B.V. Peer-review under responsibility of the Scientific Committee of ICSI 2017. Keywords: : Dislocations; Crystal Plasticity; Dislocation mobility; Screw dislocation length; Irradiation defects; Slip system © 2017 The Authors. Published by Elsevier B.V. Peer-review under responsibility of the Scientific Committee of ICSI 2017 © 2016 The Authors. Published by Elsevier B.V. Peer-review under responsibility of the Scientific Committee of PCF 2016. F ritic and f rritic-martensitic steels are onsidered t be promising material for future nuclear reactors (IAEA 2012) (Jayakumar et al. 2013) (Klueh et al. 2007) due to their good thermal properties and excellent resistance to Ferritic and ferritic-martensitic steels are considered to be promising material for future nuclear reactors (IAEA 2012) (Jayakumar et al. 2013) (Klueh et al. 2007) due t their good thermal properties and excellent resistance to Keywords: High Pressure Turbine Blade; Creep; Finite Element Method; 3D Model; Simulation. Abstract 1. Introduction 1. Introduction

2452-3216 © 2016 The Authors. Published by Elsevier B.V. Peer-review under responsibility of the Scientific Committee of PCF 2016. 2452-3216  2017 The Authors. Published by Elsevier B.V. Peer-review under responsibility of the Scientific Committee of ICSI 2017 10.1016/j.prostr.2017.07.136 * Corresponding author. Tel.: +351 218419991. E-mail address: amd@tecnico.ulisboa.pt 2452 3216© 2017 The Authors. Published by Elsevier B.V. Peer-review under responsibility of the Scientific Committee of ICSI 2017. 2452-3216© 2017 The Authors. Published by Elsevier B.V. Peer-review under responsibility of the Scientific Committee of ICSI 2017. * Correspon ing autho . Tel.: +44 27480500 21176; fax: +44 27480104. E-mail address: kulbir@igcar.gov.in * Corresponding author. Tel.: +44 27480500 21176; fax: +44 27480104. E-mail address: kulbir@igcar.gov.in