bluuebunny/crossref_metadata_2025_split huggingface dataset:
Use of Nuclear Mechanical Design Codes and Fracture Mechanics Procedures for the UK EPR™ at Hinkley Point C
Authors: Gallegillo Martin , Goodfellow Andrew , Philipp Christophe , Miles Lol , Williams Anthony 
Published: January 2014

Structures, systems and components that are important to nuclear safety are designed, manufactured, constructed, installed, commissioned, quality assured, maintained, tested and inspected to the appropriate standards. The codes and standards need to be evaluated to determine their applicability, adequacy and sufficiency.
The consideration of a conservative design is one important element of the Safety Case. This is generally used together with high manufacturing quality, the use of proven materials, inspection, functional testing and defect tolerability. In a UK Safety Case, the assessment of fracture risk of those mechanical components whose likelihood of gross failure is so low that can be discounted requires the determination of the actual critical defect size in order to compare it with the detectable defect size used for inspections.
For the UK EPR™ the 2007 Edition of the RCC-M Code including Addenda 2008, 2009 and 2010 will be the design code of reference and the RSE-M Code version 2010 Appendix 5 will form the basis of the methodology used for standard fracture mechanics assessments of the UK EPR™. Previously, the UK nuclear structural integrity community has used the ASME Code Section III for the UK previous PWR and the R6 procedure for fracture mechanics assessments.
This paper explains the choice of these two codes for the UK EPR™ and the work carried out to date by EDF Energy to demonstrate its compliance with the UK expectations for components critical to safety.
Online Monitoring — From Condition Monitoring to Plant Optimization and Operation Prognosis
Authors: Trieglaff Ralf , Rieck Detlef , Schulz Axel , Wagner Joël , Wernicke Robert 
Published: January 2014

Operation and monitoring of thick-walled components in coal fired power plants demand special requirements: high availability, flexible loads, low maintenance costs and reliability.
In this article it is shown how, based on a strength calculation of thick-walled components potential may be expelled for a flexible, cost effective plant. The aim is to meet the requirements of the volatile energy market successfully.
By different manufacturers developed online monitoring systems are used in the European generation system. The available online monitoring systems differ in terms of the underlying rules and the measuring and evaluation methods.
The common task of all online monitoring systems is the monitoring of the fatigue condition of the power plants. Fatigue calculations represent an essential element of condition monitoring and are necessary to guarantee the plant safety. Furthermore, important business decisions on the availability of facilities are made (periodic inspection and maintenance).
In the article is shown how the existing online monitoring systems can be adapted to the changed requirements.
Potential for smart additional tools based on two planned research projects are demonstrated.
Assessment of Stress Distribution in a Buttress Type Connection for Polyethylene Autoclave Reactor
Authors: Haeri Navid , Lewis John S. , Cornah Brian A. 
Published: January 2014

Background. The authors conducted a study to analyze the stress distribution in an End Cover – Buttress Strap – Cylindrical connection according to Users’ Spec.
Materials and methods. The authors created axi-symmetric finite element models in order to demonstrate the reaction loads and stress distribution on contact surfaces as well as stress distribution in the various regions of the model. They applied different design scenarios according to an operational (users’) spec to the models. The parameters investigated were the reaction loads at contact points and local and linearised stresses. Also each individual buttress groove was studied. Ratcheting and fatigue assessment were also addressed aligned with the stress results.
Results. The study results showed the behaviour of each buttress element under the operating loads. It also showed the location with the greatest local stress levels as well as the locations most vulnerable to ratcheting and fatigue cracking. A safety margin assessment also was carried out picturing the pressure-temperature enhancement possibilities.
Conclusions. The study indicates potential future design options for enhancing buttress type connections. It also contributes to the understanding of existing reactor damage and corrosion assessment, impacts which may be useful in the remaining life assessment and process optimisation. Fracture study also defines the most critical inspection points during the life of the reactor.
Basis for ASME Section XI Code Case for Flaw Tolerance Evaluation of CASS Piping
Authors: Griesbach Timothy J. , Harris David O. , Cofie Nathaniel G. , Chockie Alan , Dedhia Dilip 
Published: January 2014

Thermal aging of cast austenitic stainless steel (CASS) piping is a concern for long term operation of nuclear power plants. The effects of aging in susceptible (i.e., high delta ferrite) CASS piping and components must be managed through the license renewal period. In the United States, utilities must follow the guidance for managing age-related degradation in the Generic Aging Lessons Learned (GALL) report, and this may include analyses or inspections to demonstrate that the piping systems remain flaw tolerant. The duplex structure and large grain size of the CASS materials poses challenges with non-destructive examinations (NDE) using ultrasonic inspection (UT) techniques of the CASS components which further complicates the ability to demonstrate that any existing flaws would not compromise piping integrity.
Recent analytical studies of fully aged CF8M piping have resulted in a probabilistic fracture mechanics (PFM) model that has been used to develop allowable flaw sizes based on various levels of probability of failure. This model has been utilized extensively to generate a series of flaw acceptance limits that could be equated to the service levels in the ASME Code. A proposed ASME Section XI Code Case has been developed based on this PFM methodology for flaw tolerance evaluation of CASS piping components considering the effects of aging and uncertainties in material properties. The Code Case is currently under review by the ASME Working Group on Pipe Flaw Evaluation. This paper outlines the proposed Code Case and presents a technical basis for the flaw tolerance evaluation of CASS piping components. The ultimate objective of the flaw tolerance evaluation is to determine allowable flaw sizes in CASS components and determine target flaw sizes for NDE that will ensure safe operation with these components considering possible flaw growth. The flaw tolerance evaluation could be applied for a range of piping systems including pressurizer surge lines and primary system piping made from either statically or centrifugally cast stainless steel.
Behavior of Bolt Force Changes of Flanged Connections Under Thermal Loading
Authors: Hiratsuka Masaaki , Kobayashi Takashi , Nishiura Kengou , Itoi Katsutoyo 
Published: January 2014

When flanged connections are used at a high temperature, one of the issues is the reduction of bolt forces, especially in the cases where PTFE-based gaskets are employed. Bolt forces significantly reduce when flanged connections are heated and re-torque is necessary in many cases. It is important to estimate the residual bolt forces under thermal loading to ensure the integrity of flanged connections.
In this paper, the behavior of bolt force changes under thermal loading was experimentally examined using a flanged connection with expanded PTFE sheet gaskets. The flanged connection was heated and then cooled in an electric oven and the changes of the bolt forces were measured. It is found that the bolt forces reduce at the first thermal loading due to the gasket flow. Once the gasket is settled, the bolt force changes depend on the difference of thermal coefficients of the bolts and the gasket. It is also clarified that thinner gaskets and spiral wound gaskets are effective for elevated temperature applications.
Stress Intensity Factor Coefficients for Circumferential ID Surface Flaws in Cylinders for Appendix A of ASME Section XI
Authors: Cipolla Russell C. , Lee Darrell R. 
Published: January 2014

The stress intensity factor (KI) equations for a surface crack in ASME Section XI, Appendix A are based on non-dimensional coefficients (Gi) that allow for the calculation of stress intensity factors for a cubic varying stress field. Currently, the coefficients are in tabular format for the case of a surface crack in a flat plate geometry. The tabular form makes the computation of KI tedious when determination of KI for various crack sizes is required and a flat plate geometry is conservative when applied to a cylindrical geometry.
In this paper, closed-form equations are developed based on tabular data from API 579 (2007 Edition) [1] for circumferential cracks on the ID surface of cylinders. The equations presented, represent a complete set of Ri/t, a/t, and a/l ratios and include those presented in the 2012 PVP paper [8]. The closed-form equations provide G0 and G1 coefficients while G2 through G4 are obtained using a weight function representation for the KI solutions for a surface crack. These equations permit the calculation of the Gi coefficients without the need to perform tabular interpolation. The equations are complete up to a fourth order polynomial representation of applied stress, so that the procedures in Appendix A have been expanded. The fourth-order representation for stress will allow for more accurate fitting of highly non-linear stress distributions, such as those depicting high thermal gradients and weld residual stress fields.
The equations developed in this paper will be added to the Appendix A procedures in the next major revision to ASME Section XI. With the inclusion of equations to represent Gi, the procedures of Appendix A for the determination of KI can be performed more efficiently without the conservatism of using flat plate solutions. This is especially useful when performing flaw growth evaluations where repetitive calculations are required in the computations of crack size versus time.
The equations are relatively simple in format so that the KI computations can be performed by either spreadsheet analysis or by simple computer programming. The format of the equations is generic in that KI solutions for other geometries can be added to Appendix A relatively easily.
Characterisation of A508 Steel Cleavage Initiation Sites and Local Plasticity Using Surface Matching Technique
Authors: McKenzie F. , Smith R. J. , Scenini F. , Sherry A. H. 
Published: January 2014

Cleavage fracture in ferritic pressure vessel steels is preceded by local plastic deformation that initiates micro-cracks at carbides or second phase particles within the highly stressed region ahead of a crack tip. The objective of this study is to locate initiation sites of failed Compact Tension fracture mechanics specimens, tested at temperatures within the lower ductile to brittle transition region. A surface matching approach was developed to identify regions of local plastic deformation on the fracture surfaces of tested specimens, using confocal microscopy to acquire accurately mapped topographic images of both fractured surfaces, and then subtracting each pair of images in a virtual environment. This methodology is conceptually similar to the fracture surface topography analysis (FRASTA) technique but uses a home developed MATLAB software based on image correlation. The residual mismatch (interference) between the two datasets was used to identify the regions of local plasticity on both fracture surfaces of each test specimen studied that is believed to be associated with the cleavage initiation sites.
The size and location of the localised plasticity was found to be consistent with scanning electron microscopy observations of cleavage initiation sites located directly ahead of the fatigue pre-crack tips in tested specimens. Finite element modelling was used to identify the magnitude of stress at the identified regions of cleavage initiation for the specimens studied, providing new insights into the mechanism of cleavage initiation in the lower transition region. This study suggests that this is a promising methodology for the identification of the initiation sites of cleavage fracture.
Fundamental Study on Health Monitoring for Pipe by Vibration Analysis
Authors: Kawazu Shunsuke , Fujita Satoshi , Minagawa Keisuke 
Published: January 2014

In the operation of industrial plant, it is important to ensure the piping integrity. One of the serious problems of pipe is wall-thinning because of erosion (mechanical process) and corrosion (chemical process). Wall-thinning causes several kind of damage to pipes, such as the wholesale destruction of pipes due to corrosion, which may result in great accident and seriously harm human and environment. In general, wall thinning of pipes is measured by ultrasonic testing. However, it takes too much time for the test because there are many observation spots due to the recent growth of plant size. For this reason, this method of inspection of wall thinning has been reconsidered.
In this study, evaluation technique of pipe integrity by measuring the flow-induced by vibration of the pipe with acceleration sensors and analyzing the vibration characteristics is proposed. If the method of this study is successfully carried out, it is expected that the condition of pipes can be figured out in a short time and the examination can be made efficiently. This paper shows some/the research findings of fundamental study.
Effect of Material Microstructure on Creep Damage Formation Behavior for Ni-Base Directionally Solidified Superalloy
Authors: Ozeki Go , Sugiura Ryuji , Yokobori A. Toshimitsu , Nagumo Yoshiko , Takeuchi Hiroaki , Matsuzaki Takashi 
Published: January 2014

Ni-base directionally solidified superalloy strengthened by γ′ precipitates have been developed as a gas turbine blade. However, it is difficult to detect creep damage such as creep voids by conventional observation techniques. It is important to clarify the creep damage behavior for Ni-base superalloy.
In this study, creep crack growth tests for Ni-base directionally solidified superalloy CM247LC were conducted using the in-situ observational system. Additionally, the metallographical investigation was conducted on crept specimen using EBSD analysis and relationship between creep crack growth path and material microstructure were clarified. And, in order to clarify the difference of creep crack growth behavior, the designed two-dimensional elastic-plastic creep finite element analyses were conducted for the model with grain distribution obtained by EBSD analysis.
Study on Piping Response Under Multiple Excitation: Part 1 — Triple Shaking Table Test of Piping Having Three-Supporting Points
Authors: Watakabe Tomoyoshi , Kaneko Naoaki , Aida Shigekazu , Otani Akihito , Moriizumi Makoto , Tsukimori Kazuyuki , Kitamura Seiji 
Published: January 2014

The piping in a nuclear power plant is laid across multiple floors of a single building or two buildings, which are supported at many points. As the piping is excited by multiple inputs from the supporting points during an earthquake, seismic response analysis by multiple excitations is needed to obtain the exact seismic response of the piping. However, few experiments involving such multiple excitations have been performed to verify the validity of multiple excitation analysis. Therefore, analysis of the seismic design of piping in Japan is performed by the enveloped Floor Response Spectrum (FRS), which covers all floor response spectra at all supporting points. The piping response estimated by enveloped FRS is conservative in most cases compared with the actual seismic response by multiple excitations. To perform rational seismic design and evaluation, it is important to investigate the seismic response by multiple excitations and verify the validity of the analysis method by multiple-excitation test.
This paper reports on the result of the shaking test using triple uni-axial shaking tables and a 3-dimensional piping model (89.1mm in diameter and 5.5mm thickness). The piping model was fixed to three shaking tables, meaning three. Different inputs were possible. By the shaking test, dynamic behavior under multiple excitations was confirmed, and data to verify multiple-excitation analysis was obtained.
Relevance of a Size Effect Correction on a Minimum Toughness Value in the Ductile-to-Brittle Transition of Ferritic Steel to Study the French PWR RPV Resistance
Authors: Delvallée-Nunio Isabelle , Balestreri François 
Published: January 2014

The integrity of Reactor Pressure Vessel (RPV) which ensures radioactivity confinement is reevaluated regularly to take into account updated knowledge and to verify the high level of safety required for this component. The French 900-MWe RPV integrity demonstration was reviewed by the French Safety Authority (ASN) and the Institute for Radiological Protection and Nuclear Safety (IRSN).
This paper focuses on the validity of the fracture criterion used in this demonstration realized by the operator.
In the RPV operating temperature range, ferritic steel can have a ductile-to-brittle behaviour due to irradiation embrittlement. In this range, toughness results from normalized test specimens exhibit a size effect that could be related to the crack length. According to this, the operator has considered that the toughness of the RPV containing a given crack is equal to the minimum toughness defined on the RCC-M curve —similar to the ASME curve— reduced by a factor function of the crack length. For IRSN, this criterion is not relevant. This conclusion is based on the statistical analysis of toughness results available in the literature which shows that although taking into account a crack length correction is relevant on the average toughness, it is not the case for the minimum toughness which is the value used in the mechanical analysis: There is no size effect on the minimal toughness. Following the IRSN’s recommendation, the RPV service ability of the 900-MWe reactors has been demonstrated by the operator without applying a crack length correction on the toughness.
Experimental Fatigue Evaluation of Aluminum Brazed Plate Fin Heat Exchangers
Authors: Hölzl Reinhold , Flüggen Rainer 
Published: January 2014

For the verification of a finite element based calculation procedure for the strength and fatigue design of aluminum brazed Plate Fin Heat Exchangers (PFHE) an extensive experimental investigation program was planned and executed.
In the first step high temperature loadings as single events were applied to the PFHEs to recognize, whether damage (a leak) can be produced by a severe single loading. After various tests on different geometries the investigations were enlarged to thermal cycling tests on small PFHEs until leaks were detected.
The tests were executed with three blocks in parallel to cover a statistical influence and the tests were repeated with three new blocks under identical conditions. All six blocks gave nearly the same number of load cycles until first detection of leaks. In addition three blocks of the same geometry were loaded with lower thermal stress, giving higher load cycle numbers until a leak occurred.
The experimentally determined cycle numbers were compared to calculated cycle numbers out of the finite element modeling and showed very good agreement. Those tests and comparisons are a strong basis for fatigue and lifetime considerations on large scale PFHEs in the process industry.
Improving the Traditional RCA Methodology to Design a Most Efficient and Reliable Valve for LDPE Hypercompressors Leveraging Experimental Data From Test Bench Simulation: A Real Case Study
Authors: Maggi Carmelo , Tognarelli Leonardo , Bagagli Riccardo , Wojnar Jan 
Published: January 2014

The behavior of the valves of Hypercompressors on LDPE plants is challenging to predict because it depends on many factors and often the expected and macroscopic gas parameters, such as pressure, temperature and gas composition are not sufficient to properly evaluate the valve behavior in the field.
In fact valve operation is highly dependent on local phenomena such as localized pressure losses and presence of vortexes which are in turn influenced by the geometry of the valve and by its behavior.
To better understand all these phenomena it is needed to characterize these valves through experimental tests aimed at defining, with a good accuracy, the valve dimensionless parameters Cd (drag coefficient) and Ks (flow coefficient) as a function of the geometry of the valve itself.
If the coefficients Cd and Ks are not accurate, the expected behavior of the valve may be completely different from the evidence of the field and could not properly explain certain types of failure modes.
With a more accurate evaluation of Cd and Ks, some types of damage which in first hypothesis would seem caused by factors external to the valve, in reality are proven to be intrinsically related to valve design and often dependent on valve malfunctioning.
As a final step, through to a deep understanding of the valve behavior in the field an improvement of valve reliability and efficiency can be achieved through optimization of the design for various operating conditions.
Crack Arrest Test Results of Unirradiated and Irradiated German RPV Steels
Authors: Obermeier Florian , Barthelmes Julia , Keim Elisabeth , Hein Hieronymus , Schnabel Hilmar , Kaiser Marco 
Published: January 2014

In the CARISMA[1] and CARINA[2] projects comprehensive tensile, Charpy-impact and fracture toughness tests were performed for unirradiated and irradiated original reactor pressure vessel (RPV) steel specimens from German pressurized water reactors (PWR) up to neutron fluences in the range of 60 operational years and beyond.
In addition, crack arrest fracture toughness tests were performed to demonstrate the crack arrest behavior of the materials. To determine the crack arrest properties of ferritic steels, the designated test method according to ASTM E1221 [3] was used. However, in particular for irradiated reactor pressure vessel materials with higher irradiation embrittlement, the prescribed standard test specimen does not always provide adequate test results. During starter notch preparation annealing effects occurred in the heat affected zone (HAZ) of the brittle weld of the starter notch causing crack arrest in the HAZ after unstable crack initiation. Therefore a modified test method to perform crack arrest tests with so called duplex specimens was investigated.
In this paper this modified method and the test results of five base and four weld metals with a fluence up to 4,69E+19 cm−2 (E >1 MeV) are discussed.
The available test results show that the duplex specimen is an appropriate alternative to the standard compact crack arrest (CCA) specimen. The measured KIa fracture toughness data are enveloped by the “lower bound” of the ASME KIa-curve indexed with RTNDTj or TKIa but not all data are enveloped by indexing the “lower bound” curve with RTT0 like described in the ASME Code Case N-629 [4]. Furthermore correlations of the crack arrest test results with Charpy-impact and fracture toughness test results will be shown.