bluuebunny/crossref_metadata_2025_split huggingface dataset:
Flamanville EPR RPV’s Heads Carbon Segregation: Impact of the High Carbon Segregation on the Mechanical Properties of the RPV Steel
Authors: Marie Stéphane , Demarecaux Jérôme , Fichot Philippe , Hajjaj Mejido , Miller Cecile 
Published: October 2018

To characterize the mechanical properties of the residual carbon segregation of the Flamanville EPR RPV heads, a testing program have been organized by Framatome and performed in 3 laboratories.
Three sacrificial pieces have been used for this program, allowing the investigation of different carbon contents, coupled with different quenching conditions (thanks to a specimen sampling at different positions in the heads thickness).
As expected, the increase of the carbon content leads to an increase of the tensile properties, a shift toward the higher temperature of the brittle-to-ductile transition and a lowering of the toughness in the ductile regime. In the worst case, compare to the RTNDT measured in the frame of acceptance tests of the heads, the temperature to be considered to envelop all the experimental results with the toughness curve of the RCC-M code is 20°C higher. The extensive test program allows a quantification of these impacts for the integrity assessments of the RPV.
It is also demonstrated that the segregated material remains homogeneous, with a classical failure mode by cleavage. Master Curve approach is also applicable and used to characterize the carbon content impact on the toughness properties.
Fluid Structure Interaction Homogenization for Tube Bundles: Significant Dissipative Effects
Authors: Artini Gianluca , Broc Daniel 
Published: October 2018

In the nuclear industry, tube arrays immersed in dense fluid are often encountered. These systems have a large amount of tubes necessary to increase the thermal power exchanged and their dynamical analysis for safety assessment and in life operation is one of the major concern of the nuclear industry.
The presence of the fluid creates a strong coupling between tubes which must be taken into account for complete dynamical analysis. However, the description of fluid’s effects on oscillating structures demands great numerical efforts, especially when the tube number increases making any direct numerical simulations impossible to achieve.
In this framework, homogenization methods are a possible solution in order to deal with tube bundle Fluid-Structure Interaction (FSI) problems; in fact, it gives the possibility to analyze the dynamics of the global coupled system in large domain with reasonable degree of detail and faster simulations times.
At the CEA of Saclay a method based on the linearized Euler equations has been developed. It was presented in a previous PVP conference and its main goal is to assess the effect of spatial deformations of the tube bundle displacement field on the dynamic behavior.
In the present paper, after an analysis on the modeling of fluid force where dissipative effects are significant, a homogenized model based on the Navier-Stokes equations is introduced. Simulations in bi-dimensional configurations for different excitations are performed.
Crack Remaining Repair Techniques in Japan
Authors: Ichikawa Hiroya , Kaminaga Takayuki , Miyoshi Yoshiyuki , Kimura Ryuji 
Published: October 2018

Cracking on metal surface including Stress Corrosion Cracking (SCC) is a concern for nuclear power plant reactor system piping and components. Several repair techniques have been developed over the years that help address the cracking issues. In the United States, some crack remaining repair technologies such as Weld Overlay (WOL) technique have already been developed and also some technologies such as Excavate and Weld Repair (EWR) are still being actively discussed. Fitness-for-Service (FFS) codes in Japan also define some repair methods which allow repairing the components without removing the cracks. This paper introduces three repair methods which are described in the Fitness-for Service code in Japan, the WOL, the cap repair technique for Bottom Mounted Instrumentation (BMI) nozzles and seal welding.
WOL is a repair method that maintains structural integrity by depositing Intergranular Stress Corrosion Cracking (IGSCC) resistant weld to the outer surface of cracked circumferential welded joints in primary piping made of austenitic stainless steel. The cap repair technique for BMI nozzles is a method of repairing from the outer surface when leakage due to SCC occurs. The cap develops a new boundary outside the BMI nozzle to prevent the leakage by retaining the leakage inside the cap. The seal welding technique is also a repair method to cover the cracks by welding to prevent propagation of the cracks and to prevent leakage of reactor water.
Provided the cracks are adequately managed and evaluated, these crack remaining repair methods are very beneficial.
FA3 RPV Carbon Segregation Issue: Representativeness and Transferability of Sacrificial Parts
Authors: Lefrançois André , Berger Thierry , Lefever Benoît , Marie Stéphane 
Published: October 2018

In order to demonstrate that the cover head and bottom head of the FA3 Reactor Pressure Vessel (RPV) exhibits sufficient toughness to withstand operational (from normal up to faulted) conditions, a dedicated experimental program was performed for three sacrificial forged parts (two cover heads and a bottom head) manufactured in Creusot Forge as-well, according to a similar process.
In order to be allowed to consider the mechanical properties (toughness) measured on these sacrificial parts, it was necessary to prove that they were representative of the FA3 ones.
Among all the parameters influencing the toughness properties it was defined that two of them had a major impact: the carbon content and the quenching effects.
For the first one, it was identified that it could be influenced by pouring conditions, forging conditions, including discard and upsetting ratios and machining (thickness).
For the second one, it was identified that it could be influenced by post-heat treatment machining and by quenching conditions.
Regarding these factors, for the five parts: manufacturing parameters and acceptance test results were compared, taking into account uncertainties.
In addition, for the three sacrificial parts, carbon content measured for various depths in the thickness were compared, after a transformation methodology allowing to convert depth inside part to ingot vertical location.
Finally, it was shown that the sacrificial parts could be considered to be representative of the FA3 parts to be used, for their mechanical properties.
Buckling Assessment of Carbon Steel Pressure Vessels During PWHT
Authors: Kataoka Shunji 
Published: October 2018

Post weld heat treatment (PWHT) is often applied to welded pressure vessels and piping to increase the mechanical and metallurgical properties. During the PWHT, a pressure component made of a carbon steel to which PWHT is applied, is heated up to the temperature around 620 deg.C with +/− 25 deg.C tolerances and that elevated temperature is kept for 1 to 2 hours. The local PWHT of the vertical pressure vessel is sometimes planned at a construction site with vertical position, because of some restrictions relating to construction, such as limited access route to the vessel location, and limited space large capacity cranes. In the situation, it is important to assure the structural stability during PWHT. Except wind loads, it is considered that the stress applied in the shell during PWHT is usually not significant, however there have been no published strength criteria corresponds to the PWHT conditions.
In this paper, the new elevated temperature tensile tests and constant stress creep strain tests of A516-Gr.70, a typical carbon steel material for pressure vessels, are conducted. Several examples of the numerical analysis using isochronous stress strain curve considering creep effect are presented.
From the test result, the importance of the consideration of the creep deformation on the structural integrity assessment of PWHT operation is clarified.
Measurements of Stress During Thermal Shock in Clad Reactor Pressure Vessel Material Using Time-Resolved In-Situ Synchrotron X-Ray Diffraction
Authors: Oliver Sam , Simpson Chris , James Andrew , Reinhard Christina , Collins David , Pavier Martyn , Mostafavi Mahmoud 
Published: October 2018

Nuclear reactor pressure vessels must be able to withstand thermal shock due to emergency cooling during a loss of coolant accident. Demonstrating structural integrity during thermal shock is difficult due to the complex interaction between thermal stress, residual stress, and stress caused by internal pressure. Finite element and analytic approaches exist to calculate the combined stress, but validation is limited. This study describes an experiment which aims to measure stress in a slice of clad reactor pressure vessel during thermal shock using time-resolved synchrotron X-ray diffraction.
A test rig was designed to subject specimens to thermal shock, whilst simultaneously enabling synchrotron X-ray diffraction measurements of strain. The specimens were extracted from a block of SA508 Grade 4N reactor pressure vessel steel clad with Alloy 82 nickel-base alloy. Surface cracks were machined in the cladding. Electric heaters heat the specimens to 350°C and then the surface of the cladding is quenched in a bath of cold water, representing thermal shock. Six specimens were subjected to thermal shock on beamline I12 at Diamond Light Source, the UK’s national synchrotron X-ray facility. Time-resolved strain was measured during thermal shock at a single point close to the crack tip at a sample rate of 30 Hz. Hence, stress intensity factor vs time was calculated assuming K-controlled near-tip stress fields. This work describes the experimental method and presents some key results from a preliminary analysis of the data.
Crack Resistance Curve Measurement With Miniaturized CT Specimen
Authors: Chaouadi Rachid , Lambrecht Marlies , Gérard Robert 
Published: October 2018

The use of miniature compact tension (mini-CT) specimens for fracture mechanics was experimentally demonstrated to allow the characterization of ferritic steels in the transition regime. In particular, the master curve transition temperature T0 can confidently be determined according to the ASTM E1921 standard using mini-CT specimens. This means that specimen size effect is well taken into account if loss of constraint is limited by restricting the test temperature range to remain below the allowed maximum loading level. In the upper shelf ductile regime, where stable crack growth occurs, a number of challenges should be overcome to use such a geometry to derive the crack resistance curve, or JR-curve, transferrable to a structure. Indeed, despite a large scatter, the experimental data on several materials suggest a size effect that underestimates the crack resistance when reducing specimen size.
The crack resistance behavior of several reactor pressure vessel materials was investigated with square-sized ligament compact tension specimens of various size ranging from 1 inch-thickness (B = 25 mm) to the smallest thickness (B = 4.2 mm) of the mini-CT. The main objective of this paper is to estimate the crack resistance behavior of RPV steels that would be obtained with a standard 1T-CT specimen by using mini-CT with the appropriate specimen size correction. After a series of scaling attempts that were not successful, based on a simple dimensional analysis, a simple analytical formulation based on specimen thickness and ligament is suggested to account for specimen size effect for the CT geometry. Reasonable agreement could generally be found on a number of RPV materials between crack resistance measured with mini-CT and standard 1T-CT specimens.
Comparison of Tube-Tube Collision Frequency With and Without the Use of Impingement Plate
Authors: Buzík Jiří , Létal Tomáš , Lošák Pavel , Naď Martin , Pernica Marek 
Published: October 2018

The aim of the present work is to carry out the checking of the tube bundle of heat exchanger for the occurrence of tube-tube collision caused by cross-flow vibration with and without the use of impingement plate. This will be achieved using numerical 2D CFD (computational fluid dynamics) analysis. The 2D analysis is done using ANSYS Fluent software. Tube movement in the shell side is provided by UDF (user-defined function) DEFINE_SDOF_PROPERTIES. By determining the stiffness and weight of the tubes, two-way fluid and tube interaction can be achieved. Due to limitations of 2D CFD analysis, only the occurrence of the tube-tube or tube-shell collisions can be observed. Unfortunately, the first collision causes termination of the simulation due to negative volumes in dynamic mesh. Possible solutions to the issue are also discussed in presented paper.
The analyzed geometry of the shell side is taken from the Heat Exchanger Tube Vibration Data Bank [2]. This publication collects heat exchanger data for which vibration phenomena have been reported. The above-mentioned geometry is a domain with tube bundle at the shell side under the inlet. In the same domain, both the tie rod and the seal strips and the 45° turn of the partitions are considered.
Evaluation of Through Wall Fracture Toughness Distribution of IAEA Reference Material JRQ by Mini-C(T) Specimens and the Master Curve Method
Authors: Yamamoto Masato , Kobayashi Tomohiro 
Published: October 2018

The load and temperature history during pressurized thermal shock (PTS) event is highly depending on the crack edge location in wall thickness direction of a reactor pressure vessel (RPV) beltline region. Therefore, the consideration of plant specific through-wall fracture toughness distribution, which is not considered in the current codes and regulations [1,2], may improve the structural integrity assessment for PTS event.
The Master Curve (MC) method [3,4] is one of the methods, which can directory evaluate the fracture toughness of ferritic materials with relatively low number of any size of specimens. CRIEPI has proposed the use of very small C(T) (Mini-C(T)) specimens for the MC method. The appropriateness of Mini-C(T) technology has been demonstrated through a series of researches and round robin activities [5, 6, 7, 8, 9].
The present study evaluated the through-wall fracture toughness distribution of irradiated IAEA reference material (JRQ) by means of combination of MC method and Mini-C(T) specimens. Four thickness locations between inner surface to 1/4-T was selected. Those four layers were separately subjected to the Mini-C(T) MC evaluation in two different laboratories. Both laboratories could separately obtain valid and consistent reference temperature, To, from all the tested layers. Inner most layer exhibits 80 °C lower To compared to the 1/4-T location even though the layer has the highest fluence of 5.38 × 1019 n/cm2, while that in 1/4-T location is 2.54 × 1019 n/cm2. The results demonstrate that initial toughness distribution is dominant in the general trend of fracture toughness distribution even after the material was highly irradiated.
A Methodology for Constructing Complex 3D Cracked Body Finite Element Models Efficiently
Authors: Vorster Willem J. J. , Mann Alex 
Published: October 2018

Design code design by analysis routes often require knowledge of collapse parameters while structural integrity fitness for purpose assessments require detailed knowledge of both collapse and fracture parameters. These parameters can be calculated using detailed Finite Element Analysis (FEA) simulation.
However, the process of constructing Finite Element (FE) models of complex components such as branches and branches located on pipework bends is often time consuming when considering aspects of weld design around branches attached to bends or the requirement for specialist knowledge when inserting crack-like features into FE models close to welds or areas with notable curvature such as that found in bends. This paper presents a technique which can be used to construct finite element models quickly so as to shift the effort of the structural integrity expert from modelling to assessment.
This paper provides a basic procedure for creating and manipulating FE models to enable inserting cracks in areas of structural discontinuities or in areas of significant curvature. The usefulness of this method is illustrated by considering the results of 3D defect free and cracked FE analyses of pipe bends. It illustrates the use of the method to create defect free branches and can be extended to include crack in these components. The analyses reported here use the commercial FEA software Abaqus. However, this technique can be adapted to any modern FEA software. This methodology would be of particular interest to industries relying on pipework and pressure vessel components and where the failure of such component can have serious safety and commercial implications.
Evaluation of Bending Limit of 9Cr-1Mo-V Steel by Master Curve and Failure Assessment Diagram Method: Evaluation of Welded Metal
Authors: Nakatani Mitsuyoshi , Oda Kazuo , Tanaka Tomohiro , Abe Masamitsu , Takashima Yasuhito , Minami Fumiyoshi 
Published: October 2018

In this study, the bending and rolling limit of 9Cr-1Mo-V steel plate welded using submerged arc welding (SAW) was investigated. Hereafter in this report, the bending limit refers to bending by press or by roller. Since heat treatment after welding and before the bending process may be effective in preventing brittle fracture during bending, the effect of postweld heat treatment (PWHT) for improving toughness was evaluated by Larson-Miller parameter (LMP). Charpy impact tests of SAW welded parts made of 9Cr-1Mo-V steel plates were performed, and the absorbed energy of the weld metal was lower than that of the heat affected zone (HAZ) and the base metal, however the absorbed energy was recovered by PWHT. The absorbed energy increased as LMP increased.
The reference temperature T0 for the master curve method was estimated by the results of Charpy impact tests. The bending limit was estimated by the method proposed in the previous study. The fracture probability during the bending process seems to decrease as LMP increases.
Prediction for Plastic Collapse Stresses for Pipes With Inner and Outer Circumferential Flaws
Authors: Hasegawa Kunio , Li Yinsheng , Mares Vratislav , Lacroix Valery 
Published: October 2018

Bending stresses at incipient plastic collapse for pipes with circumferential surface flaws are predicted by net-section stress approach. Appendix C-5320 of ASME B&PV Code Section XI provides a formula of bending stress at the plastic collapse, where the formula is applicable for both inner and outer surface flaws. That is, the collapse stresses for pipes with inner and outer surface flaws are the same, because of the pipe mean radius at the flawed section being entirely the same.
Authors considered the separated pipe mean radii at the flawed ligament and at the un-flawed ligament. Based on the balances of axial force and bending moment, formulas of plastic collapse stresses for each inner and outer flawed pipe were obtained. It is found that, when the flaw angle and depth are the same, the collapse stress for inner flawed pipe is slightly higher than that calculated by Appendix C-5320 formula, and the collapse stress for outer flawed pipe is slightly lower than that by Appendix C-5320 formula, as can be expected. The collapse stresses derived from the three formulas are almost the same in most instances. For less common case where the flaw angle and depth are very large for thick wall pipes, the differences amongst the three collapse stresses become large.
3D FEM Modeling of Crack Initiation in a Reactor Pressure Vessel During a PTS Event (LBLOCA)
Authors: Mora Méndez Diego Fernando , Niffenegger Markus , Qian Guian , Jaros Michal , Niceno Bojan 
Published: October 2018

To perform the integrity assessment of a reactor pressure vessel (RPV) related to Pressurized Thermal Shock (PTS), we model the RPV using the 3D finite element method (FEM). Accurate prediction of temperature and stress fields is determined by using 2-Phase computational fluid dynamics (CFD) simulation in combination with an appropriate finite element discretization of the RPV wall. The cladding and the ferritic low alloy steel are considered as two separated layers, which can be intersected by superficial cracks. The calculation of the stress intensity factor (SIF) in mode I is based on the linear fracture mechanics theory (LEFM) and hypothetical cracks are located in different locations to consider the most critical cases.
In the present study, the sub-modeling technique is implemented to refine the mesh required by the fracture analysis in the region of interest. Three types of cracks are considered: axial, circumferential and inclined. The performed integrity assessment uses the master curve approach. The stress intensity factor in the deepest point of a surface crack was compared with the material’s fracture toughness. In previous studies the integrity of the RPV subjected to medium and small break Loss-of-Coolant Accident (MBLOCA and SBLOCA, respectively) has been evaluated, therefore the concern in this contribution is the large break of Loss-Of-Coolant Accident (LBLOCA). The combination of 3D FEM with CFD simulations allows us to study the influence of the dynamic cooling plume on the stress intensity in more detail than with the conventional one dimensional method.
Experimental Investigation of Interfacial and Permeation Leak Rates in Sheet Gaskets and Valve Stem Packing
Authors: Salah Omar Aweimer Ali , Bouzid Abdel-Hakim 
Published: October 2018

The quantities of leak rate through sealing systems are being regulated because of the global concern on the hazardous pollutants being released into the atmosphere and their consequences on the environment and health. The maximum tolerated leak is becoming a design criterion, and the leak rate for an application under specific conditions is required to be estimated with reasonable accuracy. In this respect, experimental and theoretical studies are being conducted to characterize the gas flow through gaskets and packing rings. The amount of the total leak that is present in a gasketed joint or a valve stem packing is the sum of the permeation leak through the sealing material and the interfacial leak at the mating surfaces between the sealing element and mechanical clamp assembly. The existing models used to predict leakage do not separate these two types of leaks.
This paper deals with a study based on experimental testing that quantifies the amount of these two types of leaks in bolted gasketed joints and packed stuffing boxes. It shows the contribution of interfacial leak for low and high contact surface stresses and the influence of the surface finish as a result of a 32 and 250 micro-inch RAAH phonographic finish in the case of a bolted flange joint. The results indicate that most of the leak is interfacial reaching 99% at the low stress while the interfacial leak is in the same order of magnitude of the permeation leak at high stress reaching 10−6 and 10−8 mg/s in both packing and gaskets, respectively.